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Generic Environmental Impact Statement for License Renewal of Nuclear Plants (NUREG-1437 Vol. 2)


Table of Contents


Publication Information


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Final Report

Manuscript Completed: April 1996
Date Published: May 1996

Division of Regulatory Applications
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001


Figures


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B.1 PWR major refurbishment outage schedule

B.2 BWR major refurbishment schedule

B.3 Typical program impacts relative to corresponding conservative case impacts

B.4 Incremental labor hours

B.5 Outage average incremental on-site staff

B.6 Additional personnel required to perform conservative case pressurized-water reactor license renewal major refurbishment outage activities

B.7 Additional personnel required to perform conservative case boiling-water reactor license renewal current-term outage activities

B.8 Additional personnel required to perform typical case pressurized-water reactor license renewal current-term outage activities

B.9 Additional personnel required to perform typical case boiling-water reactor license renewal major refurbishment activities

B.10 Incremental low-level waste generated

B.11 Incremental occupational radiation exposure

B.12 Incremental waste disposal costs

B.13 Incremental capital and labor costs

B.14 Total license renewal costs

B-1.1 Outage duration and personnel exposure in seven steam generator replacement projects

C.1 Population categories, by sparseness and proximity

C.2 The seven case study nuclear plants

C.3 Conservative scenario refurbishment work force estimates (PWR)

C.4 Socioeconomic impact area associated with Arkansas Nuclear One refurbishment: Pope County

C.5 Region surrounding the Arkansas Nuclear One nuclear plant

C.6 Socioeconomic impact area associated with D. C. Cook refurbishment, including Berrien County, Lake Township, and Bridgman

C.7 Region surrounding the D. C. Cook nuclear plant

C.8 Socioeconomic impact area associated with Diablo Canyon refurbishment: San Luis Obispo County

C.9 Region surrounding the Diablo Canyon nuclear plant

C.10 Socioeconomic impact area associated with Indian Point refurbishment: Westchester and Dutchess counties

C.11 Region surrounding the Indian Point nuclear plant

C.12 Socioeconomic impact area associated with Oconee Nuclear Station refurbishment: Oconee County

C.13 Region surrounding the Oconee Nuclear Station nuclear plant

C.14 Socioeconomic impact area associated with Three Mile Island refurbishment: Middletown, Royalton, and Londonderry Township

C.15 Region surrounding the Three Mile Island nuclear plant

C.16 Socioeconomic impact area associated with Wolf Creek Generating Station refurbishment: Coffey County

C.17 Region surrounding the Wolf Creek Generating Station nuclear plant

E.C.1 Person-rem per year for Arkansas One

E.C.2 Person-rem per year for Beaver Valley

E.C.3 Person-rem per year for Big Rock Point

E.C.4 Person-rem per year for Calvert Cliffs

F.1 Example information request letter sent to state fish and wildlife resource agencies, state water pollution control agencies, and regions of the U.S. Fish and Wildlife Service, National Marine Fisheries Service, and Environmental Protection Agency.

G.1 Residuals from regression of the log of early fatality (average deaths per reactor year) on the log of 16-km (10-mile) exposure index of persons at risk.

G.2 Log plot of early fatalities per reactor year within 16 km (10 miles) of 21 nuclear power plants [3300 MW(t) or greater], resulting from postulated accidents, regressed on log of exposure index (EI) for 16 km (10 miles)

G.3 Residuals from regression of log of normalized latent fatality (average deaths per 1000-MW reactor-year) on the log of 240-km (150-mile) exposure index of persons at risk

G.4 Residuals from regression of the log of normalized total dose (rem per 1000-MW reactor-year) on the log of 240-km (150-mile) exposure index of persons at risk

G.5 Log plot of normalized latent fatalities per 1000 MW(t) per reactor year of 28 nuclear power plants resulting from postulated accidents, regressed on log of exposure index (EI) at 240 km (150 miles)

G.6 Log plot of normalized total dose in person-rem per 1000 MW(t) per reactor year within 240 km (150 miles) of 28 nuclear power plants [3300 MW(t) or greater] resulting from postulated accidents, regressed on log of exposure index (EI)

G.7 Residuals from regression of the log of normalized expected cost (dollars per 1000-MW reactor-year) on the log of 240-km (150-mile) exposure index of persons at risk

G.8 Log plot of normalized expected cost per 1000 MW(t) per reactor year of 27 nuclear power plants [3300 MW(t) or greater] resulting from postulated accidents, regressed on the log of exposure index (EI).

G.9 Log plot of early fatalities (average deaths per reactor-year) for final environmental statement pressurized-water reactor plants, fitted regression line, and 95 percent normal-theory upper prediction confidence bounds.

G.10 Log plot of early fatalities (average deaths per reactor-year) for final environmental statement boiling-water reactor plants, fitted regression line, and 95 percent normal-theory upper prediction confidence bounds

G.11 Log plot of normalized latent fatalities (average deaths per 1000-MW reactor-year) for final environmental statement pressurized-water reactor plants, fitted regression line, and 95 percent distribution-free upper prediction confidence bounds

G.12 Log plot of normalized latent fatalities (average deaths per 1000-MW reactor-year) for final environmental statement boiling-water reactor plants, fitted regression line, and 95 percent distribution-free upper prediction confidence bounds

G.13 Log plot of normalized total dose (person-rem per 1000-MW reactor-year) for final environmental statement pressurized-water reactor plants, fitted regression line, and 95 percent distribution-free upper prediction confidence bounds

G.14 Log plot of normalized total dose (person-rem per 1000-MW reactor-year) for final environmental statement boiling-water reactor plants, fitted regression line, and 95 percent distribution-free upper prediction confidence bounds

G.15 Log plot of normalized expected cost (dollars per 1000-MW reactor-year) for final environmental statement pressurized-water reactor plants, fitted regression line, and 95 percent distribution-free upper prediction confidence bounds

G.16 Log plot of normalized expected cost (dollars per 1000-MW reactor-year) for final environmental statement boiling-water reactor plants, fitted regression line, and 95 percent distribution-free upper prediction confidence bounds

G.17 Cumulative proportions of the midyear license date for 16-km (10-mile) exposure index of persons at risk for final environmental statement plants and all other plants

G.18 Cumulative proportions of the midyear license for 240-km (150-mile) exposure index of persons at risk for final environmental statement plants and all other plants


Acronyms and Abbreviations


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ADS automatic depressurization system
AEA Atomic Energy Act of 1954
AEC U.S. Atomic Energy Commission
AEO Atomic Energy Outlook 1990
AFUDC allowance for funds used during construction
AGA American Gas Association
AGR advanced gas-cooled reactor
AIRFA American Indian Religious Freedom Act
ALARA as low as reasonably achievable
ALI annual limits on intake
A/m amps per meter
AML acute myelogenous leukemia
ANO Arkansas Nuclear One
ANOVA analysis of variance
ANSI American National Standards Institute
AP&L Arkansas Power and Light
ASME American Society of Mechanical Engineers
ATWS anticipated transit without scram
 
BAU business-as-usual
BEIR Biological Effects of Ionizing Radiation
BIG/GT biomass-gasifier/gas turbine
BRC below regulatory concern
BSD Burlington School District
B&W Babcock and Wilcox
BWR boiling-water reactor
 
° C degrees centigrade (Celsius)
CAA Clean Air Act
CAAA Clean Air Act Amendments of 1990
CCC California Coastal Commission
CDE committed dose equivalent
CDF core damage frequencies
CE Combustion Engineering
CEDE committed effective dose equivalent
CEQ Council on Environmental Quality
CERCLA Comprehensive Environmental Response, Compensation, and Liability Act
CFC chlorofluorocarbon
CFR Code of Federal Regulations
Ci curie
CML chronic myelogenous leukemia
CMSA consolidated metropolitan statistical area
CNS central nervous system
CO carbon monoxide
ConEd Consolidated Edison
CPI containment performance improvement
CPW continuous polymer wire
CRAC Consequence (of) Reactor Accident Code
CRD control rod drive
CWA Clean Water Act of 1977
CZMA Coastal Zone Management Act
 
DAC derived air concentrations
DAW dry active waste
DE dose equivalent
DECON a nuclear plant decommissioning method
DER Florida Department of Environmental Regulation
DFA direct fluorescent antibody
DMBA dimethylbenzanthracene
DNR Florida Department of Natural Resources
DO dissolved oxygen
DOE U.S. Department of Energy
DOI Department of Interior
DRBC Delaware River Basin Commission
DREF dose rate effectiveness factor
DRI Data Resources Incorporated
DSC dry shielded canister
DSM demand-side management
 
E electric field
EA environmental assessment
EAB exclusion area boundary
EDE effective dose equivalent
EEC European Economic Community
EEDB Energy Economic Data Base
EEG electroencephalogram
EEI Edison Electric Institute
E-field electric-field
EI exposure index
EIA Energy Information Administration
EIS environmental impact statement
EKG electrocardiogram
ELF extremely low frequency
EM electromagnetic
EMF electromagnetic field
ENTOMB a nuclear plant decommissioning method
EO Executive Order
EPA U.S. Environmental Protection Agency
EPACT Energy Policy Act of 1992
EPCRA Emergency Planning and and Community Right-to-Know Act
EPRI Electric Power Research Institute
EPZ emergency planning zone
ESA Endangered Species Act
ESEERCO Empire State Electric Energy Research Corporation
 
FDA U.S. Food and Drug Administration
FEMA U.S. Federal Emergency Management Agency
FERC Federal Energy Regulatory Commission
FES final environmental statement
FFCA Federal Facilities Compliance Agreement
FIFRA Federal Insecticide, Fungicide, and Rodenticide Act
FIS federal interim storage
FONSI finding of low significant impact
FPC Florida Power Commission
FP&L Florida Power & Light
FR Federal Register
FSAR final safety analysis report
FWCA Fish and Wildlife Coordination Act
FWS U.S. Fish and Wildlife Service
 
GBD gas bubble disease
GCHWR gas-cooled heavy-water-moderated reactor
GCR gas-cooled reactor
GE General Electric Company
GEIS generic environmental impact statement
g/m2/s gallons per square meter per second
GNP gross national product
GNSI General Nuclear Systems, Inc.
GPU General Public Utilities Corporation
GRI Gas Research Institute
GTCC greater-than-class-C
GW gigawatt
GWd gigawatt-days
 
HC hydrocarbons
HL&P Houston Lighting and Power Company
HLW high-level radioactive waste
HP health physics
HPOF high-pressure oil-filled
HRS hazard ranking system
HSM horizontal storage module
HSWA Hazardous and Solid Waste Amendments of 1984
HWR heavy-water reactor
 
ICRP International Commission on Radiological Protection
IGSCC intergranular stress-cracking corrosion
IMP intramembranous protein particle
INIRC International Non-Ionizing Radiation Protection Association
INPO Institute of Nuclear Power Operations
IOR ion exchange resin
IPA integrated plant assessment
IPE individual plant examination
IRPA International Radiation Protection Association
ISFSI independent spent-fuel storage installation
ISI in-service inspection
ISTM inspection, surveillance, testing, and maintenance
 
kV kilovolt
kV/m kilovolts per meter
kW kilowatt
kWh kilowatt-hour
 
LD Legionnaires' disease
LDR land disposal restrictions
LDSD Lower Dauphin School District
LET linear energy transfer
LLRWPAA Low-Level Radioactive Waste Policy Amendments Act of 1985
LLW low-level radioactive waste
LMFBR liquid-metal first breeder reactor
LOCA loss-of-coolant accident
LOS level of service
LPGS Liquid Pathway Generic Study
LPZ low population zone
LWR light-water reactor
 
m meter
mA milliamperes
MACCS MELCOR Accident Consequence Code System
MANOVA multivariate analyses of covariance
MAP Methodologies Applications Program
MASD Middletown Area School District
mCi milliCurie
MCLG maximum contaminant goal levels
MDNR Maryland Department of Natural Resources
MFD magnetic flux density
mG milligauss
mM millimole
MMPA Marine Mammals Protection Act
MPC maximum permissible concentration
MPRSA Marine Protection, Research, and Sanctuaries Act
MPOB maximum permissible organ burden
MRC Marine Review Committee
mrem millirem
MRS monitored retrievable storage
m3/s cubic meters per second
MSA metropolitan statistical area
MSW municipal solid waste
mT millitesla
MTIHM metric tons of initial heavy metal
MTU metric tons of uranium
mV/m millivolts per meter
MW megawatt
MWd megawatt-days
MW(e) megawatt (electrical)
MW(t) megawatt (thermal)
MYL middle year of license
MYR middle year of relicense
m g/g micrograms per gram
m m micron
 
NAA nonattainment area
NAAQS National Ambient Air Quality Standards
NAS National Academy of Sciences
NBS National Bureau of Standards (now NIST)
NCA National Coal Association
NCRP National Council on Radiation Protection and Measurements
NEC normalized expected cost
NEPA National Environmental Policy Act of 1969
NERC North American Electric Reliability Council
NESC National Electric Safety Code
NESHAP National Emission Standards for Hazardous Air Pollutants
NGS nuclear generating station
NHPA National Historic Preservation Act of 1966
NIEHS National Institute of Environmental Health Sciences
NIOSH National Institute for Occupational Safety and Health
NIST National Institute of Standards and Technology
NLF normalized latent facility
NMFS National Marine Fisheries Service
NMR nuclear magnetic resonance
NOx nitrogen oxide(s)
NPA National Planning Association
NPDES National Pollutant Discharge Elimination System
NPP nuclear power plant
NRC U.S. Nuclear Regulatory Commission
NSPS new source performance standards
NSSS nuclear steam supply system
NTD normalized total dose
NUHOMS Nutech Horizontal Modular System
NUMARC Nuclear Utilities Management and Resources Council
NUREG an NRC reports category
NUS NUS Corporation
NWPA Nuclear Waste Policy Act of 1982
NYSDEC New York State Department of Environmental Conservation
 
ODC ornithine decarboxylase
OHMS hydroxy melatonin sulfate
OL operating license
O&M operation and maintenance
ONS Oconee Nuclear Station
OPEC Organization of Petroleum Exporting Countries
OR odds ratio
ORNL Oak Ridge National Laboratory
OSHA Occupational Safety and Health Administration
OTA Office of Technology Assessment
OTEC ocean thermal energy conversion
 
PAME primary amoebic meningoencephalitis
PASNY Power Authority for the State of New York
PCB polychlorinated biphenyl
PG&E Pacific Gas and Electric
pH hydrogen-ion concentration
PHWR pressurized heavy-water reactor
PLEX plant life extension
PM particulate matter
PMR proportionate mortality ratios
ppm parts per million
PSD prevention of significant deterioration
PRA probabilistic risk assessment
PTH parathyroid hormone
PURPA Public Utility Regulatory Policies Act of 1978
PURTA Public Utilities Realty Tax Assessment of 1970
PV solar photovoltaic
PWR pressurized-water reactor
 
QA quality assurance
 
RBE relative biological effectiveness
RCB reactor containment building
RCRA Resource Conservation and Recovery Act of 1976
RD&D 1. research, design, and development
2. research, development, and demonstration
RERF Radiation Effects Research Council
RET renewable energy technology
RF radio frequency
RHR residual heat removal
RIMS Regional Industrial Multiplier System
rms root mean square
ROW right(s) of way
RPV reactor pressure vessel
RRY reference reactor year
RSD Russellville (Ark.) School District
RSS Reactor Safety Study
RV recreational vehicle
RY reactor-year
 
SAFSTOR a nuclear plant decommissioning method
SAMDA severe accident mitigation design alternative
SAND Data Resource Incorporated's detailed electricity sector model
SAND NUPLEX SAND generating capacity projections
SAR safety analysis report
SARA Superfund Amendments and Reauthorization Act
SCE Southern California Edison
SCM Surface Compartment Model
SDG&E San Diego Gas & Electric Company
SDWA Safe Drinking Water Act
SEA Science and Engineering Associates, Inc.
SER safety evaluation report
SERI Solar Energy Research Institute
SEV state equalized value
SF spent fuel
SHPO state historic preservation office
SI International System
SIR standardized incidence ratio
SLB shallow land burial
SMR standardized mortality ratio
SMITTR surveillance, on-line monitoring, inspections, testing, trending, and recordkeeping
SMSA standard metropolitan statistical area
SO2 sulfur dioxide
SOK San Onofre kelp bed
SONGS San Onofre Nuclear Generating Station
SRBC Susquehanna River Basin Commission
SSC systems, structures, and components
 
t metric tons
TDE total dose equivalent
TDS total dissolved solids
TEDE total effective dose equivalent
TMI Three Mile Island (nuclear plant)
TRU transuranic
TSCA Toxic Substances Control Act
TVA Tennessee Valley Authority
 
UCB upper confidence bound
UFC uranium fuel cycle
UHV ultra-high voltage
UNSCEAR United Nations Scientific Committee on the Effects of Atomic Radiation
USD Unified School District
USGS U.S. Geological Survey
USI unresolved safety issue
 
VDT video display terminal
VR volume reduction
VRF volume reduction factor
 
W watt
WCGS Wolf Creek Generating Station
WHO World Health Organization
WNP-2 Washington Nuclear Project
WTE® Whole Tree Energy®


Appendix A : General Characteristics and Environmental Settings of Domestic Nuclear Power Plants


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This section contains brief descriptions of each nuclear power plant site in the United States. The information was compiled from: (1) the plant safety analysis reports (SARs), (2) the Nuclear Regulatory Commission "Gray Book" (NUREG-0020), (3) site environmental reports, (4) environmental impact statements, (5) environmental assessments used to check data for cooling water system and site information, and (6) WASH-1319 used for selected data. Specific data that could not be found in these six sources were obtained from ORNL-NSIC-55.

Specific data sources are listed on the following page.

Source for General Information

Plant Name: SAR

Location: County and distance and direction from nearest town or city: NUREG-0020

Latitude and longitude: List provided by R. Rush, ORNL

Licensee: Utility as listed in NUREG-0020

Source for Information on Unit

Docket Number: NUREG-0020

Construction Permit: Nuclear Safety Journal, Power Reactor Licensing Activity

Operating License: Table A.1 of SECY-90-160 (NUREG-0020)

Commercial Operation: NUREG-0020

License Expiration: Table A.1 of SECY-90-160 (NUREG-0020)

Licensed Thermal Power [MW(t)]: NUREG-0020

Design Electrical Rating [net MW(e)]: NUREG-0020

Type of Reactor: NUREG-0020

Nuclear Steam Supply System Vendor: NUREG-0020

Source for Information on Cooling Water System

Type: SAR, NUREG-0020

Source: NUREG-0020

Source Temperature Range: SAR, ORNL-NSIC-55

Condenser Flow Rate: SAR, ORNL-NSIC-55

Design Condenser Temperature Rise: SAR, ORNL-NSIC-55

Intake Structure : SAR

Discharge Structure: SAR

Source for Information on Site

Total Area: SAR, WASH-1319

Exclusion Distance: SAR

Low Population Zone: SAR

Nearest City: SAR; 1980 population:*

Site Topography: SAR

Surrounding Area Topography: SAR

Land Use within 8 km (5 miles): SAR

Nearby Features: SAR

Area of Transmission Line Corridor: WASH-1319

Population within an 80-km (50-mile) radius:*

*Population data are taken from population projections developed for NRC by MITRE Corporation and made available to GEIS project.


Arkansas Nuclear One


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Location:

Pope County, Arkansas

10 km (6 miles) WNW of Russellville

latitude 35.3100°N; longitude 93.2308°W

Licensee: Arkansas Power and Light Co.

Unit Information Unit 1 Unit 2
Docket Number 50-313 50- 368
Construction Permit 1968 1972
Operating License 1974 1978
Commercial Operation 1974 1980
License Expiration 2014 2018
Licensed Thermal Power [MW(t)] 2568 2815
Design Electrical Rating [net MW(e)] 850 912
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor B&W CE

Cooling Water System

Type: Unit 1, once through

Source: Dardanelle Reservoir Unit 2, natural draft cooling tower

Source Temperature Range: 4-28°C (40-83°F)

Condenser Flow Rate:

48.3 m3/s (765,000 gal/min) for Unit 1

26.6 m3/s (422,000 gal/min) for Unit 2

Design Condenser Temperature Rise:

8.3°C (15°F) for Unit 1

17.1°C (30.7°F) for Unit 2

Intake Structure : 981-m (3220-ft) canal

Discharge Structure: 160-m (520-ft) canal

Site Information

Total Area: 469 ha (1160 acres)

Exclusion Distance: 1.05-km (0.65-mile) radius

Low Population Zone: 6.44-km (4.00-mile) radius

Nearest City: Little Rock; 1980 population: 159,159

Site Topography: flat

Surrounding Area Topography: hilly to mountainous

Land Use within 8 km (5 miles): wooded

Nearby Features:

The nearest town is Mill Creek 3 km (2 miles) NE. The size of the Dardanelle Reservoir is 15,000 ha (37,000 acres). The reservoir is part of the Arkansas River. The Missouri Pacific Railroad and U.S. Highway I-40 are just N of the site.

Area of Transmission Line Corridor: 1500 ha (3700 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
200,000 210,000 220,000 250,000 270,000


Beaver Valley Power Station


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Location:

Beaver County, Pennsylvania

40 km (25 miles) NW of Pittsburgh

latitude 40.6219°N; longitude 80.4339°W

Licensee: Duquesne Light Co.

Unit Information Unit 1 Unit 2
Docket Number 50-334 50-412
Construction Permit 1970 1974
Operating License 1976 1987
Commercial Operation 1976 1987
License Expiration 2016 2027
Licensed Thermal Power [MW(t)] 2652 2652
Design Electrical Rating [net MW(e)] 835 836
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: natural draft cooling towers

Source: Ohio River

Source Temperature Range: 1.1-28°C (34-83°F)

Condenser Flow Rate: 30.31 m3/s (480,400 gal/min) each unit

Design Condenser Temperature Rise: 14°C (26°F)

Intake Structure : concrete structure at river edge

Discharge Structure: at river edge

Site Information

Total Area: 203 ha (501 acres)

Exclusion Distance: 0.45 km (0.28 mile)

Low Population Zone: 5.79 km (3.60 miles)

Nearest City: Pittsburgh; 1980 population: 423,959

Site Topography: flat

Surrounding Area Topography: hilly

Land Use within 8 km (5 miles): industrial and residential

Nearby Features:

The nearest town is Midland 1.6 km (1 mile) NW. A large industrial area is about 1.6 km (1 mile) WNW. The Penn Central Railroad is adjacent to the site. Beaver Creek and

Raccoon Creek State Parks are within 16 km (10 miles).

Area of Transmission Line Corridor: uses existing corridor

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
3,740,000 3,840,000 3,910,000 4,040,000 4,170,000


Bellefonte Nuclear Plant


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Location:

Jackson County, Alabama

11 km (7 miles) ENE of Scottsboro

latitude 34.7089°N; longitude 85.9275°W

Licensee: Tennessee Valley Authority

Unit Information Unit 1 Unit 2
Docket Number 50-438 50-439
Construction Permit 1974 1974
Operating License -- --
Commercial Operation -- --
License Expiration -- --
Design Thermal Power [MW(t)] 3760 3760
Design Electrical Rating [net MW(e)] 1213 1213
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor B&W B&W

Cooling Water System

Type: natural draft cooling towers

Source: Guntersville Lake

Source Temperature Range: 5-27°C (41-81°F)

Condenser Flow Rate: 26 m3/s (410,000 gal/min) each unit

Design Condenser Temperature Rise: 20°C (36°F)

Intake Structure : intake channel

Discharge Structure: submerged multi-port diffuser

Site Information

Total Area: 610 ha (1500 acres)

Exclusion Distance: 0.92-km (0.57-mile) minimum

Low Population Zone: 3.22 km (2.00 miles)

Nearest City: Huntsville; 1980 population: 142,513

Site Topography: flat valley

Surrounding Area Topography: hilly out of valley

Land Use within 8 km (5 miles): agricultural and wooded

Nearby Features:

The nearest town is Hollywood 5 km (3 miles) WNW. The Widows Creek coal-fired plant is 24 km (15 miles) NE. Guntersville Lake is on the Tennessee River.

Area of Transmission Line Corridor: 1200 ha (2900 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,070,000 1,150,000 1,230,000 1,340,000 1,470,000


Big Rock Point Nuclear Plant


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Location:

Charlevoix County, Michigan

6 km (4 miles) NE of Charlevoix

latitude 45.3592°N; longitude 85.1947°W

Licensee: Consumers Power Co.

Unit Information Unit 1
Docket Number 50-155
Construction Permit 1960
Operating License 1962
Commercial Operation 1963
License Expiration 2002
Licensed Thermal Power [MW(t] 240
Design Electrical Rating [net MW(e)] 72
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: once through

Source: Lake Michigan

Source Temperature Range: 3-20°C (38-68°F)

Condenser Flow Rate: 3.1 m3/s (49,000 gal/min)

Design Condenser Temperature Rise: 11°C (20°F)

Intake Structure : underwater crib

Discharge Structure: open discharge canal

Site Information

Total Area: 240 ha (600 acres)

Exclusion Distance: 0.82 km (0.51 mile)

Low Population Zone: 4.02 km (2.50 miles)

Nearest City: Sault Ste. Marie, Canada; 1980 population: 81,048

Site Topography: gently sloping

Surrounding Area Topography: gently sloping

Land Use within 8 km (5 miles): commercial and industrial

Nearby Features:

The nearest town is Charlevoix 6 km (4 miles) SW. The C&O Railroad is about 1.6 km (1 mile) SE. Lake Charlevoix is 5 km (3 miles) S.

Area of Transmission Line Corridor:

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
200,000 210,000 210,000 230,000 240,000


Braidwood Station


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Location:

Will County, Illinois

39 km (24 miles) SSW of Joliet

latitude 41.2436°N; longitude 88.2297°W

Licensee: Commonwealth Edison Co.

Unit Information Unit 1 Unit 2
Docket Number 50-456 50- 457
Construction Permit 1975 1975
Operating License 1987 1988
Commercial Operation 1988 1988
License Expiration 2027 2028
Licensed Thermal Power [MW(t)] 3411 3411
Design Electrical Rating [net MW(e)] 1120 1120
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: closed cycle cooling pond

Source: Kankakee River

Source Temperature Range: 0-31°C (32-87°F)

Condenser Flow Rate: 46.05 m3/s (729,800 gal/min) each unit

Design Condenser Temperature Rise: 12°C (21°F)

Intake Structure : concrete structure at lake shore

Discharge Structure: surface discharge flume to lake

Site Information

Total Area: 1804 ha (4457 acres)

Exclusion Distance: 0.48-km (0.30-mile) minimum

Low Population Zone: 1.810 km (1.125 mile) radius

Nearest City: Joliet; 1980 population: 77,956

Site Topography: flat to rolling

Surrounding Area Topography: flat to rolling

Land Use within 8 km (5 miles): agricultural

Nearby Features:

The nearest town is Godley 0.8 km (0.5 mile) SW. There are 4 state parks within 16 km (10 miles). Joliet Arsenal is about 13 km (8 miles) NE. Dresden Nuclear Power Station is about 16 km (10 miles) N and La Salle County Station (nuclear) is about 32 km (20 miles) WSW. The Illinois Central Gulf Railroad is just NW. U.S. Highway I-55 is about 3 km (2 miles) NW.

Area of Transmission Line Corridor: 961.5 ha (2376 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
4,510,000 4,650,000 4,750,000 4,920,000 5,090,000


Browns Ferry Nuclear Power Station


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Location:

Limestone County, Alabama

16 km (10 miles) NW of Decatur

latitude 34.7042°N; longitude 87.1186°W

Licensee: Tennessee Valley Authority

Unit Information Unit 1 Unit 2 Unit 3
Docket Number 50-259 50- 260 50-296
Construction Permit 1967 1967 1968
Operating License 1973 1974 1976
Commercial Operation 1974 1975 1977
License Expiration 2013 2014 2016
Licensed Thermal Power [MW(t)] 3293 3293 3293
Design Electrical Rating [net MW(e)] 1065 1065 1065
Type of Reactor BWR BWR BWR
Nuclear Steam Supply System Vendor GE GE GE

Cooling Water System

Type: once through and helper towers

Source: Tennessee River

Source Temperature Range: 4-32°C (40-90°F)

Condenser Flow Rate: 40 m3/s (630,000 gal/min) each unit

Design Condenser Temperature Rise: 14°C (25°F)

Intake Structure : concrete structure in small inlet

Discharge Structure: diffuser pipes

Site Information

Total Area: 340 ha (840 acres)

Exclusion Distance: 1.22-km (0.76-mile) radius

Low Population Zone: 11.3 km (7.00 miles)

Nearest City: Huntsville; 1980 population: 142,513

Site Topography: flat

Surrounding Area Topography: flat to rolling

Land Use within 8 km (5 miles): agricultural

Nearby Features:

The nearest town is Lawngate 1.6 km (1 mile) NE. The Redstone Arsenal is 40 km (25 miles) E. The Southern Railroad is 10 km (6 miles) S and the Louisville and Nashville Railroad is 10 km (6 miles) E.

Area of Transmission Line Corridor: 546 ha (1350 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
760,000 810,000 850,000 930,000 1,010,000


Brunswick Steam Electric Plant


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Location:

Brunswick County, North Carolina

26 km (16 miles) S of Wilmington

latitude 33.9583°N; longitude 78.0106°W

Licensee: Carolina Power and Light Co.

Unit Information Unit 1 Unit 2
Docket Number 50-325 50- 324
Construction Permit 1967 1968
Operating License 1976 1974
Commercial Operation 1977 1975
License Expiration 2016 2014
Licensed Thermal Power [MW(t)] 2436 2436
Design Electrical Rating [net MW(e)] 821 821
Type of Reactor BWR BWR
Nuclear Steam Supply System Vendor GE GE

Cooling Water System

Type: once through

Source: Cape Fear River

Source Temperature Range: 4-30°C (40-86°F)

Condenser Flow Rate: 42.6 m3/s (675,000 gal/min) each unit

Design Condenser Temperature Rise: 9°C (17°F)

Intake Structure : 5-km (3-mile) canal from Cape Fear River

Discharge Structure: 10-km (6-mile) canal to Atlantic Ocean

Site Information

Total Area: 490 ha (1200 acres)

Exclusion Distance: 0.92 km (0.57 mile)

Low Population Zone: 3.22 km (2.00 miles)

Nearest City: Wilmington; 1980 population: 44,000

Site Topography: flat

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): less than one-half agricultural, remainder swamps or wooded

Nearby Features:

Nearest town is Southport 5 km (3 miles) S. Sunny Point Military Ocean Terminal is about 8 km (5 miles) N.

Area of Transmission Line Corridor: 1400 ha (3500 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
230,000 250,000 270,000 300,000 340,000


Byron Station


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Location:

Ogle County, Illinois

27 km (17 miles) SW of Rockford

latitude 42.0750°N; longitude 89.2811°W

Licensee: Commonwealth Edison Co.

Unit Information Unit 1 Unit 2
Docket Number 50-454 50- 455
Construction Permit 1975 1975
Operating License 1985 1987
Commercial Operation 1985 1987
License Expiration 2025 2027
Licensed Thermal Power [MW(t)] 3411 3411
Design Electrical Rating [net MW(e)] 1120 1120
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: natural draft cooling towers

Source: Rock River

Source Temperature Range:

Condenser Flow Rate: 39.9 m3/s (632,000 gal/min) each unit

Design Condenser Temperature Rise: 13°C (24°F)

Intake Structure : concrete structure on river bank

Discharge Structure: discharged to river

Site Information

Total Area: 565.8 ha (1398 acres)

Exclusion Distance: 0.42-km (0.26-mile) minimum

Low Population Zone: 4.83 km (3.00 miles)

Nearest City: Rockford; 1980 population: 139,712

Site Topography: rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): agricultural

Nearby Features:

The nearest town is Byron about 5 km (3 miles) NNE. The Chicago Milwaukee and the St. Paul and Pacific Railroads are about 6 km (4 miles) NNE. White Pines State Park is about 18 km (11 miles) WSW.

Area of Transmission Line Corridor: 800 ha (2000 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,000,000 1,030,000 1,060,000 1,100,000 1,100,000


Callaway Plant


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Location:

Callaway County, Missouri

16 km (10 miles) SE of Fulton

latitude 38.7622°N; longitude 91.7817°W

Licensee: Union Electric Co.

Unit Information Unit 1
Docket Number 50-483
Construction Permit 1976
Operating License 1984
Commercial Operation 1984
License Expiration 2024
Licensed Thermal Power [MW(t)] 3565
Design Electrical Rating [net MW(e)] 1171
Type of Reactor PWR
Nuclear Steam Supply System Vendor WEST

Cooling Water System

Type: natural draft cooling tower

Source: Missouri River

Source Temperature Range:

Condenser Flow Rate: 33 m3/s (530,000 gal/min)

Design Condenser Temperature Rise: 17°C (30°F)

Intake Structure : intake from river

Discharge Structure: discharged to river

Site Information

Total Area: 1290 ha (3188 acres)

Exclusion Distance: 1.21-km (0.75-mile) radius

Low Population Zone: 4.02 ha (2.50 miles)

Nearest City: Columbia; 1980 population: 62,061

Site Topography: flat, on a small plateau

Surrounding Area Topography: rolling to hilly

Land Use within 8 km (5 miles): wooded, agricultural, and pasture

Nearby Features:

The nearest town is Portland 8 km (5 miles) SE. The Missouri River is about 8 km (5 miles) S. The Missouri, Kansas, and Texas Railroad is about 5 km (3 miles) S and the Missouri Pacific Railroad is about 10 km (6 miles) S. U.S. Highway I-70 is about 16 km (10 miles) N.

Area of Transmission Line Corridor: 461 ha (1140 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
400,000 420,000 430,000 460,000 500,000


Calvert Cliffs Nuclear Power Plant


[ Prev | Next | Top of file ]

Location:

Calvert County, Maryland

56 km (35 miles) S of Annapolis

latitude 38.4347°N; longitude 76.4419°W

Licensee: Baltimore Gas and Electric Co.

Unit Information Unit 1 Unit 2
Docket Number 50-317 50- 318
Construction Permit 1969 1969
Operating License 1974 1976
Commercial Operation 1975 1977
License Expiration 2014 2016
Licensed Thermal Power [MW(t)] 2700 2700
Design Electrical Rating [net MW(e)] 845 845
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor CE CE

Cooling Water System

Type: once through

Source: Chesapeake Bay

Source Temperature Range: 1-31°C (34-87°F)

Condenser Flow Rate: 76 m3/s (1,200,000 gal/min) each unit

Design Condenser Temperature Rise: 6°C (10°F)

Intake Structure : about 170 m (560 ft) from shore

Discharge Structure: about 260 m (850 ft) from shore

Site Information

Total Area: 459 ha (1135 acres)

Exclusion Distance: 1.08-km (0.67-mile) radius

Low Population Zone:

Nearest City: Washington, D.C.; 1980 population: 638,432

Site Topography: rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): agricultural and wooded

Nearby Features:

The nearest town is Long Beach 1.6 km (1 mile) NNW. Calvert Cliffs State Park is about 6 km (4 miles) SSE. A naval ordnance facility is 11 km (7 miles) SSW. Washington, D.C., is 72 km (45 miles) NW.

Area of Transmission Line Corridor: 805 ha (1990 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
3,030,000 3,140,000 3,260,000 3,480,000 3,720,000


Catawba Nuclear Station


[ Prev | Next | Top of file ]

Location:

York County, South Carolina

10 km (6 miles) NNW of Rock Hill

latitude 35.0514°N; longitude 81.0708°W

Licensee: Duke Power Co.

Unit Information Unit 1 Unit 2
Docket Number 50-413 50- 414
Construction Permit 1975 1975
Operating License 1985 1986
Commercial Operation 1985 1986
License Expiration 2025 2026
Licensed Thermal Power [MW(t)] 3411 3411
Design Electrical Rating [net MW(e)] 1145 1145
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: mechanical draft cooling towers

Source: Lake Wylie

Source Temperature Range: 6-28°C (43-83°F)

Condenser Flow Rate: 42 m3/s (660,000 gal/min) each unit

Design Condenser Temperature Rise: 13°C (24°F)

Intake Structure : skimmer wall on cove of the lake

Discharge Structure: on another cove of the lake

Site Information

Total Area: 158 ha (391 acres)

Exclusion Distance: 0.76-km (0.47-mile) radius

Low Population Zone: 6.12-km (3.80-mile) radius

Nearest City: Charlotte, North Carolina; 1980 population: 315,474

Site Topography: rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): wooded with recreational and permanent homes along the lake

Nearby Features:

The nearest town is Rock Hill 10 km (6 miles) SSE. U.S. Highway I-77 is about 10 km (6 miles) E and I-85 is about 27 km (17 miles) N. The Southern Railway is 8 km (5 miles) S.

Area of Transmission Line Corridor: 236 ha (584 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,590,000 1,730,000 1,860,000 2,090,000 2,340,000


Clinton Power Station


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Location:

De Witt County, Illinois

10 km (6 miles) E of Clinton

latitude 40.1731°N; longitude 88.8342°W

Licensee: Illinois Power Co.

Unit Information Unit 1
Docket Number 50-461
Construction Permit 1976
Operating License 1987
Commercial Operation 1987
License Expiration 2027
Licensed Thermal Power [MW(t)] 2894
Design Electrical Rating [net MW(e)] 933
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: once through

Source: Salt Creek

Source Temperature Range: 0-28°C (32-83°F)

Condenser Flow Rate: 35.8850 m3/s (568,701 gal/min)

Design Condenser Temperature Rise: 13°C (23°F)

Intake Structure : concrete structure at shoreline of North Fork Salt Creek

Discharge Structure: 5-km (3-mile) flume discharging to Salt Creek

Site Information

Total Area: 5702 ha (14,090 acres)

Exclusion Distance: 0.97-km (0.60-mile) radius

Low Population Zone: 4.02-km (2.50-mile) radius

Nearest City: Decatur; 1980 population: 93,939

Site Topography: flat

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): agricultural

Nearby Features:

The nearest town is De Witt 3 km (2 miles) ENE. Weldon Springs State Park is 10 km (6 miles) SW. The Illinois Central Gulf Railroad crosses the site. U.S. Highway I-74 is 18 km (11 miles) NE. A dam on Salt Creek near the site creates the reservoir for the cooling water system.

Area of Transmission Line Corridor: 367 ha (906 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
730,000 770,000 790,000 830,000 870,000


Comanche Peak Steam Electric Station


[ Prev | Next | Top of file ]

Location:

Somervell County, Texas

64 km (40 miles) SW of Fort Worth

latitude 32.2983°N; longitude 97.7856°W

Licensee: Texas Utilities Electric Co.

Unit Information Unit 1 Unit 2
Docket Number 50-445 50- 446
Construction Permit 1974 1974
Operating License 1990 1993
Commercial Operation 1990 1993
License Expiration 2030 2033
Design Thermal Power [MW(t)] 3411 3411
Design Electrical Rating [net MW(e)] 1150 1150
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: once through

Source: Squaw Creek Reservoir

Source Temperature Range:

Condenser Flow Rate: 65 m3/s (1,030,000 gal/min) each unit

Design Condenser Temperature Rise: 8°C (15°F)

Intake Structure : on shore of reservoir

Discharge Structure: canal to reservoir

Site Information

Total Area: 3104 ha (7669 acres)

Exclusion Distance: 1.54-km (0.96-mile) minimum

Low Population Zone: 6.44-km (4.00-mile) radius

Nearest City: Fort Worth; 1980 population: 385,164

Site Topography: flat with hills rising from the reservoir

Surrounding Area Topography: rolling to hilly

Land Use within 8 km (5 miles): agricultural, farm/ranch land, and range land

Nearby Features:

The nearest town is Glen Rose 8 km (5 miles) SSE. Dinosaur Valley State Park is 8 km (5 miles) SW. A 66-cm (26-inch) oil pipeline is very near the site and a 91-cm (36-inch) natural gas line is about 3 km (2 miles) from the site.

Area of Transmission Line Corridor: 185 ha (458 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,130,000 1,310,000 1,460,000 1,650,000 1,880,000


Donald C. Cook Nuclear Power Plant


[ Prev | Next | Top of file ]

Location:

Berrien County, Michigan

16 km (10 miles) S of St. Joseph

latitude 41.9761°N; longitude 86.5664°W

Licensee: Indiana and Michigan Electric Co.

Unit Information Unit 1 Unit 2
Docket Number 50-315 50- 316
Construction Permit 1969 1969
Operating License 1974 1977
Commercial Operation 1975 1978
License Expiration 2014 2017
Licensed Thermal Power [MW(t)] 3250 3411
Design Electrical Rating [net MW(e)] 1030 1100
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: once through

Source: Lake Michigan

Source Temperature Range: 1-23°C (34-74°F)

Condenser Flow Rate: 50 m3/s (800,000 gal/min) each unit

Design Condenser Temperature Rise: 12°C (21°F)

Intake Structure : intake cribs 686 m (2250 ft) from shore

Discharge Structure: 381 m (1250 ft) from shore

Site Information

Total Area: 260 ha (650 acres)

Exclusion Distance: 0.61 km (0.38 mile)

Low Population Zone: 3.22 km (2.00 miles)

Nearest City: South Bend, Indiana; 1980 population: 109,727

Site Topography: rolling

Surrounding Area Topography: flat to rolling

Land Use within 8 km (5 miles): agricultural and wooded

Nearby Features:

The nearest town is Livingston 1.6 km (1 mile) SW. The Chesapeake and Ohio Railroad and U.S. Highway I-94 are just E of the site. Warren Dunes State Park is about 8 km (5 miles) SSW.

Area of Transmission Line Corridor: 1340 ha (3300 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,250,000 1,310,000 1,350,000 1,440,000 1,530,000


Cooper Nuclear Station


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Location:

Nemaha County, Nebraska

37 km (23 miles) S of Nebraska City

latitude 40.3619°N; longitude 95.6411°W

Licensee: Nebraska Public Power District

Unit Information Unit 1
Docket Number 50-298
Construction Permit 1968
Operating License 1974
Commercial Operation 1974
License Expiration 2014
Licensed Thermal Power [MW(t)] 2381
Design Electrical Rating [net MW(e)] 778
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: once through

Source: Missouri River

Source Temperature Range: 1-23°C (34-73°F)

Condenser Flow Rate: 39.8 m3/s (631,000 gal/min)

Design Condenser Temperature Rise: 10°C (18°F)

Intake Structure : at shoreline

Discharge Structure: at shoreline

Site Information

Total Area: 441 ha (1090 acres)

Exclusion Distance: 1.09 (0.68 mile)

Low Population Zone: 1.61-km (1.00-mile) radius

Nearest City: Lincoln; 1980 population: 171,932

Site Topography: flat

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): agricultural

Nearby Features:

The nearest town is Nemaha about 1.6 km (1 mile) S. A railroad runs just W of the site. Indian Cave State Park is about 13 km (8 miles) SSE.

Area of Transmission Line Corridor: 2777 ha (6862 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
180,000 190,000 200,000 220,000 230,000


Crystal River Nuclear Plant


[ Prev | Next | Top of file ]

Location:

Citrus County, Florida

11 km (7 miles) NW of Crystal River

latitude 28.9572°N; longitude 82.6989°W

Licensee: Florida Power Corp.

Unit Information Unit 3
Docket Number 50-302
Construction Permit 1968
Operating License 1977
Commercial Operation 1977
License Expiration 2017
Licensed Thermal Power [MW(t)] 2544
Design Electrical Rating [net MW(e)] 825
Type of Reactor PWR
Nuclear Steam Supply System Vendor B&W

Cooling Water System

Type: once through

Source: Gulf of Mexico

Source Temperature Range: 31°C (87°F) maximum

Condenser Flow Rate: 43 m3/s (680,000 gal/min)

Design Condenser Temperature Rise: 9.5°C (17.1°F)

Intake Structure : 4900 m (16,000 ft) from shoreline

Discharge Structure: 4000-m (13,000-ft) canal

Site Information

Total Area: 1917 ha (4738 acres)

Exclusion Distance: 1.34-km (0.83-mile) radius

Low Population Zone: 8.05 km (5.00 miles)

Nearest City: Gainesville; 1980 population: 81,371

Site Topography: swamps and marshland

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): wooded and pasture land

Nearby Features:

The nearest town is Crystal River about 11 km (7 miles) SE. Units 1 and 2 are coal-fired plants and share a common intake and discharge with the nuclear unit.

Area of Transmission Line Corridor: 866 ha (2140 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
440,000 490,000 550,000 660,000 790,000


Davis-Besse Nuclear Power Station


[ Prev | Next | Top of file ]

Location:

Ottawa County, Ohio

34 km (21 miles) E of Toledo

latitude 41.5972°N; longitude 83.0864°W

Licensee: Toledo Edison Co.

Unit Information Unit 1
Docket Number 50-346
Construction Permit 1971
Operating License 1977
Commercial Operation 1978
License Expiration 2017
Licensed Thermal Power [MW(t)] 2772
Design Electrical Rating [net MW(e)] 906
Type of Reactor PWR
Nuclear Steam Supply System Vendor B&W

Cooling Water System

Type: natural draft cooling tower

Source: Lake Erie

Source Temperature Range: 1-23°C (34-74°F)

Condenser Flow Rate: 30 m3/s (480,000 gal/min)

Design Condenser Temperature Rise: 14°C (26°F)

Intake Structure : submerged intake about 900 m (3000 ft) offshore

Discharge Structure: submerged discharge about 280 m (930 ft) offshore

Site Information

Total Area: 386 ha (954 acres)

Exclusion Distance: 0.72-km (0.45-mile) radius

Low Population Zone: 3.22 km (2.00 miles)

Nearest City: Toledo; 1980 population: 354,635

Site Topography: flat

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): agricultural with marshland around site

Nearby Features:

The nearest town is Oak Harbor about 10 km (6 miles) SW. Several wildlife refuge areas are within 8 km (5 miles) of the site.

Area of Transmission Line Corridor: 730 ha (1800 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,920,000 1,990,000 2,050,000 2,170,000 2,290,000


Diablo Canyon Nuclear Power Plant


[ Prev | Next | Top of file ]

Location:

San Luis Obispo County, California

19 km (12 miles) W of San Luis Obispo

latitude 35.2117°N; longitude 120.8544°W

Licensee: Pacific Gas and Electric Co.

Unit Information Unit 1 Unit 2
Docket Number 50-275 50-323
Construction Permit 1968 1970
Operating License 1984 1985
Commercial Operation 1985 1986
License Expiration 2024 2025
Licensed Thermal Power [MW(t)] 3338 3411
Design Electrical Rating [net MW(e)] 1086 1119
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: once through

Source: Pacific Ocean

Source Temperature Range: 10-17°C (50-63°F)

Condenser Flow Rate: 54.5 m3/s (863,000 gal/min) each unit

Design Condenser Temperature Rise: 10°C (18°F)

Intake Structure : reinforced-concrete structure located at shore line in a cove with artificial breakwater wall

Discharge Structure: reinforced-concrete structure drops water in stair step type weir overflow from elevation 21 m (70 ft) to the ocean and discharges on the surface at the shore line

Site Information

Total Area: 300 ha (750 acres)

Exclusion Distance: 0.80 km (0.50 mile)

Low Population Zone: 9.66 km (6.00 miles)

Nearest City: Santa Barbara; 1980 population: 74,542

Site Topography: hilly

Surrounding Area Topography: hilly to mountainous

Land Use within 8 km (5 miles): undeveloped and wooded

Nearby Features:

Site is remote, the nearest town being San Luis Obispo 19 km (12 miles) E. Beaches 11-24 km (7-15 miles) ESE have an influx of summer visitors. Pismo Beach State Park and Morro Bay State Park are within 24 km (15 miles). Vandenberg Air Base is 56 km (35 miles) ESE.

Area of Transmission Line Corridor: 2400 ha (6000 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
300,000 330,000 350,000 380,000 420,000


Dresden Nuclear Power Station


[ Prev | Next | Top of file ]

Location:

Grundy County, Illinois

14 km (9 miles) E of Morris

latitude 41.3897°N; longitude 88.2711°W

Licensee: Commonwealth Edison Co.

Unit Information Unit 2 Unit 3
Docket Number 50-237 50-249
Construction Permit 1966 1966
Operating License 1969 1971
Commercial Operation 1970 1971
License Expiration 2010 2011
Licensed Thermal Power [MW(t)] 2527 2527
Design Electrical Rating [net MW(e)] 794 794
Type of Reactor BWR BWR
Nuclear Steam Supply System Vendor GE GE

Cooling Water System

Type: cooling lake & spray canal

Source: Kankakee River

Source Temperature Range: 4-29°C (40-85°F)

Condenser Flow Rate: 29.7 m3/s (471,000 gal/min) each unit

Design Condenser Temperature Rise:

Intake Structure: canal from Kankakee River to a crib house

Discharge Structure: A canal carries water to a cooling lake of about 520 ha (1275 acres) with a hold-up time of about 3 days. The water then divides, some going to the Illinois River and some returns to the plant. Spray modules are floated in the canals.

Site Information

Total Area: 386 ha (953 acres) plus 516-ha (1275-acre) cooling lake

Exclusion Distance: 0.80-km (0.50-mile) radius

Low Population Zone: 8.00 km (4.97 miles)

Nearest City: Joliet; 1980 population: 77,956

Site Topography: flat

Surrounding Area Topography: rolling prairie

Land Use within 8 km (5 miles): agriculture

Nearby Features:

The nearest town is Channahon 5 km (3 miles) NNE. The General Electric Nuclear Power Plant Training Center is S of the site. A large abandoned strip mine is located in the area. Braidwood Station nuclear plant is about 16 km (10 miles) S and La Salle County Station nuclear plant is about 35 km (22 miles) SW. An army ammunition plant is about 11 km (7 miles) E.

Area of Transmission Line Corridor: 911 ha (2250 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
6,820,000 7,050,000 7,200,000 7,450,000 7,710,000


Duane Arnold Energy Center


[ Prev | Next | Top of file ]

Location:

Linn County, Iowa

13 km (8 miles) NW of Cedar Rapids

latitude 42.1006°N; longitude 91.7772°W

Licensee: Iowa Electric Light and Power Co.

Unit Information Unit 1
Docket Number 50-331
Construction Permit 1970
Operating License 1974
Commercial Operation 1975
License Expiration 2014
Licensed Thermal Power [MW(t)] 1658
Design Electrical Rating [net MW(e)] 538
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: mechanical draft cooling towers

Source: Cedar River

Source Temperature Range: 0-32°C (32-89°F)

Condenser Flow Rate: 18 m3/s (290,000 gal/min)

Design Condenser Temperature Rise: 14°C (25°F)

Intake Structure : structure on river shoreline

Discharge Structure: canal to shoreline

Site Information

Total Area: 200 ha (500 acres)

Exclusion Distance: 0.43 km (0.27 mile)

Low Population Zone: 9.66 km (6.00 miles)

Nearest City: Cedar Rapids; 1980 population: 110,243

Site Topography: flat

Surrounding Area Topography: rolling and hilly

Land Use within 8 km (5 miles): agricultural

Nearby Features:

The nearest town is Palo about 3 km (2 miles) SW. Several wildlife refuge areas are within 16 km (10 miles) of the site.

Area of Transmission Line Corridor: 469 ha (1160 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
620,000 660,000 690,000 750,000 820,000


Joseph M. Farley Nuclear Plant


[ Prev | Next | Top of file ]

Location:

Houston County, Alabama

26 km (16 miles) E of Dothan

latitude 31.2228°N; longitude 85.1125°W

Licensee: Alabama Power Co.

Unit Information Unit 1 Unit 2
Docket Number 50-348 50-364
Construction Permit 1972 1972
Operating License 1977 1981
Commercial Operation 1977 1981
License Expiration 2017 2021
Licensed Thermal Power [MW(t)] 2652 2652
Design Electrical Rating [net MW(e)] 829 829
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: mechanical draft cooling towers

Source: Chattahoochee River

Source Temperature Range: 30°C (86°F) maximum

Condenser Flow Rate: 40.1 m3/s (635,000 gal/min) each unit

Design Condenser Temperature Rise: 11°C (20°F)

Intake Structure : intake from river bank via a storage pond

Discharge Structure: at river bank

Site Information

Total Area: 749 ha (1850 acres)

Exclusion Distance: 1.26 km (0.78 mile)

Low Population Zone: 3.22 km (2.00 miles)

Nearest City: Columbus, Georgia; 1980 population: 169,441

Site Topography: flat to rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): agricultural and wooded

Nearby Features:

The nearest town is Columbia about 6 km (4 miles) N. Chattahoochee State Park is about 19 km (12 miles) S.

Area of Transmission Line Corridor: 2140 ha (5300 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
390,000 410,000 440,000 490,000 540,000


Enrico Fermi Atomic Power Plant


[ Prev | Next | Top of file ]

Location:

Monroe County, Michigan

48 km (30 miles) SW of Detroit

latitude 41.9631°N; longitude 83.2578°W

Licensee: Detroit Edison Co.

Unit Information Unit 2
Docket Number 50-341
Construction Permit 1972
Operating License 1985
Commercial Operation 1988
License Expiration 2025
Licensed Thermal Power [MW(t)] 3292
Design Electrical Rating [net MW(e)] 1093
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: natural draft cooling towers

Source: Lake Erie

Source Temperature Range: 1-24°C (34-76°F)

Condenser Flow Rate: 52.80 m3/s (836,700 gal/min)

Design Condenser Temperature Rise: 10°C (18°F)

Intake Structure : at edge of lake

Discharge Structure: to the lake via a 20-ha (50-acre) pond

Site Information

Total Area: 453 ha (1120 acres)

Exclusion Distance: 0.92 km (0.57 mile)

Low Population Zone: 4.83 km (3.00 miles)

Nearest City: Detroit; 1980 population: 1,203,368

Site Topography: flat

Surrounding Area Topography: flat to rolling

Land Use within 8 km (5 miles): mostly agricultural

Nearby Features:

The town of Stony Point is adjacent to the site to the S. Sterling State Park and General Custer Historical Site are about 8 km (5 miles) SW.

Area of Transmission Line Corridor: 73 ha (180 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
5,370,000 5,630,000 5,840,000 6,230,000 6,650,000


James A. Fitzpatrick Nuclear Power Plant


[ Prev | Next | Top of file ]

Location:

Oswego County, New York

10 km (6 miles) NE of Oswego

latitude 43.5239°N; longitude 76.3983°W

Licensee: Power Authority of the State Of New York

Unit Information
Docket Number 50-333
Construction Permit 1970
Operating License 1974
Commercial Operation 1975
License Expiration 2014
Licensed Thermal Power [MW(t)] 2436
Design Electrical Rating [net MW(e)] 816
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: once through

Source: Lake Ontario

Source Temperature Range: 3-19°C (37-67°F)

Condenser Flow Rate: 22.25 m3/s (352,600 gal/min)

Design Condenser Temperature Rise: 18°C (32°F)

Intake Structure : intake from the lake

Discharge Structure: discharge to the lake

Site Information

Total Area: 284 ha (702 acres)

Exclusion Distance: 0.92 km (0.57 mile)

Low Population Zone: 5.47 km (3.40 miles)

Nearest City: Syracuse; 1980 population: 170,105

Site Topography: flat to rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): agricultural, industrial, residential, and recreational

Nearby Features:

The nearest town is Lakeview about 1.6 km (1 mile) WSW. Fort Ontario is about 8 km (5 miles) SW. Nine Mile Point Nuclear Station is about 0.8 km (0.5 mile) W.

Area of Transmission Line Corridor: 400 ha (1000 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
820,000 810,000 800,000 800,000 810,000


Fort Calhoun Station


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Location:

Washington County, Nebraska

31 km (19 miles) N of Omaha

latitude 41.5208°N; longitude 96.0767°W

Licensee: Omaha Public Power District

Unit Information Unit 1
Docket Number 50-285
Construction Permit 1968
Operating License 1973
Commercial Operation 1974
License Expiration 2013
Licensed Thermal Power [MW(t)] 1500
Design Electrical Rating [net MW(e)] 478
Type of Reactor PWR
Nuclear Steam Supply System Vendor CE

Cooling Water System

Type: once through

Source: Missouri River

Source Temperature Range: 0-27°C (32-80°F)

Condenser Flow Rate: 23 m3/s (360,000 gal/min)

Design Condenser Temperature Rise: 9.94°C (17.9°F)

Intake Structure : concrete structure at river shore

Discharge Structure: at river shore

Site Information

Total Area: 270 ha (660 acres)

Exclusion Distance: 0.92-km (0.57-mile) minimum

Low Population Zone: 8.05 km (5.00 miles)

Nearest City: Omaha; 1980 population: 313,939

Site Topography: flat to rolling

Surrounding Area Topography: flat and rolling

Land Use within 8 km (5 miles): agricultural

Nearby Features:

The nearest town is De Soto 3 km (2 miles) SSE. De Soto National Wildlife Refuge is about 1.6 km (1 mile) E. Wilson Island State Park is about 6 km (4 miles) SE.

Area of Transmission Line Corridor: 75.3 ha (186 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
770,000 800,000 830,000 890,000 950,000


Robert Emmett Ginna Nuclear Power Plant


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Location:

Wayne County, New York

32 km (20 miles) NE of Rochester

latitude 43.2778°N; longitude 77.3089°W

Licensee: Rochester Gas and Electric Corp.

Unit Information Unit 1
Docket Number 50-244
Construction Permit 1966
Operating License 1969
Commercial Operation 1970
License Expiration 2009
Licensed Thermal Power [MW(t)] 1520
Design Electrical Rating [net MW(e)] 470
Type of Reactor PWR
Nuclear Steam Supply System Vendor WEST

Cooling Water System

Type: once through

Source: Lake Ontario

Source Temperature Range: 0-27°C (32-80°F)

Condenser Flow Rate: 22.5 m3/s (356,000 gal/min)

Design Condenser Temperature Rise: 10.9°C (19.6°F)

Intake Structure : Structure is located on lake bottom 940 m (3100 ft) from shore. Water flows to screenhouse via a 3-m (10-ft) diameter tunnel in bedrock.

Discharge Structure: open canal to Lake Ontario

Site Information

Total Area: 137 ha (338 acres)

Exclusion Distance: 0.47-1.38 km (0.29-0.85 mile)

Low Population Zone: 4.83 km (3.00 miles)

Nearest City: Rochester; 1980 population: 241,741

Site Topography: gently rolling to flat

Surrounding Area Topography: sloping

Land Use within 8 km (5 miles): agricultural, orchards

Nearby Features:

The nearest town is Lakeside 3 km (2 miles) SW. The N.Y. Central Railroad is about 5 km (3 miles) S.

Area of Transmission Line Corridor: 110 ha (280 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,140,000 1,120,000 1,100,000 1,110,000 1,120,000


Grand Gulf Nuclear Station


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Location:

Claiborne County, Mississippi

40 km (25 miles) S of Vicksburg

latitude 32.0075°N; longitude 91.0475°W

Licensee: System Energy Resources, Inc.

Unit Information Unit 1
Docket Number 50-416
Construction Permit 1974
Operating License 1984
Commercial Operation 1985
License Expiration 2024
Licensed Thermal Power [MW(t)] 3833
Design Electrical Rating [net MW(e)] 1250
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: natural draft cooling towers

Source: Mississippi River

Source Temperature Range: 1-28°C (33-82°F)

Condenser Flow Rate: 36.1 m3/s (572,000 gal/min)

Design Condenser Temperature Rise: 17°C (30°F)

Intake Structure : a series of radial-collector wells along the shoreline

Discharge Structure: discharge to river via a barge slip

Site Information

Total Area: 850 ha (2100 acres)

Exclusion Distance: 0.69-km (0.43-mile) radius

Low Population Zone: 3.22 km (2.00 miles)

Nearest City: Jackson; 1980 population: 202,895

Site Topography: flat to rolling

Surrounding Area Topography: flat and rolling

Land Use within 8 km (5 miles): wooded and recreational

Nearby Features:

The nearest town is Grand Gulf 3 km (2 miles) N. The Natchez Trace Parkway is about 10 km (6 miles) SE. The Grand Gulf Military Park is just N of the site. There are numerous hunting lodges near the site.

Area of Transmission Line Corridor: 930 ha (2300 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
350,000> 380,000 410,000 450,000 500,000


Haddam Neck Plant (Connecticut Yankee)


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Location:

Middlesex County, Connecticut

21 km (13 miles) E of Meriden

latitude 41.4819°N; longitude 72.4992°W

Licensee: Connecticut Yankee Atomic Power Co.

Unit Information
Docket Number 50-213
Construction Permit 1964
Operating License 1967
Commercial Operation 1968
License Expiration 2007
Licensed Thermal Power [MW(t)] 1825
Design Electrical Rating [net MW(e)] 582
Type of Reactor PWR
Nuclear Steam Supply System Vendor WEST

Cooling Water System

Type: once through

Source: Connecticut River

Source Temperature Range: 1-29°C (34-85°F)

Condenser Flow Rate: 23.5 m3/s (372,000 gal/min)

Design Condenser Temperature Rise: 12.4°C (22.4°F)

Intake Structure : at shoreline

Discharge Structure: discharge canal to Connecticut River about 1.6 km (1 mile) downriver

Site Information

Total Area: 212 ha (525 acres)

Exclusion Distance: 0.53 km (0.33 mile)

Low Population Zone: 4.35 km (2.70 miles)

Nearest City: Meriden; 1980 population: 57,118

Site Topography: level with steep slopes up from river

Surrounding Area Topography: mostly hilly

Land Use within 8 km (5 miles): wooded

Nearby Features:

The nearest town is Haddam 1.6 km (1 mile) WSW. Haddam Meadows State Park is within 1.6 km (1 mile). The New York, New Haven, and Hartford Railroad runs along the pposite river bank. The Millstone Nuclear Power Station is 32 km (20 miles) SE.

Area of Transmission Line Corridor: 399 ha (985 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
3,630,000 3,770,000 3,910,000 4,140,000 4,380,000


Shearon Harris Nuclear Power Plant


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Location:

Wake County, North Carolina

32 km (20 miles) SW Raleigh

latitude 35.6336°N; longitude 78.9564°W

Licensee: Carolina Power and Light Co.

Unit Information Unit 1
Docket Number 50-400
Construction Permit 1978
Operating License 1987
Commercial Operation 1987
License Expiration 2027
Licensed Thermal Power [MW(t)] 2775
Design Electrical Rating [net MW(e)] 900
Type of Reactor PWR
Nuclear Steam Supply System Vendor WEST

Cooling Water System

Type: natural draft cooling tower

Source: Buckhorn Creek

Source Temperature Range: 5-27°C (41-81°F)

Condenser Flow Rate: 30.5 m3/s (483,000 gal/min)

Design Condenser Temperature Rise: 14.3°C (25.7°F)

Intake Structure : at shoreline of reservoir on Buckhorn Creek

Discharge Structure: discharged to reservoir

Site Information

Total Area: 4348 ha (10,744 acres)

Exclusion Distance: 2.03-km (1.26-miles) minimum

Low Population Zone: 4.83 km (3.00 miles)

Nearest City: Raleigh; 1980 population: 149,771

Site Topography: rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): mostly wooded with some agricultural

Nearby Features:

The nearest town is Bonsal 3 km (2 miles) NW. The Seaboard Coast Line Railroad is 3 km (2 miles) NW. Buckhorn Creek feeds into the Cape Fear River.

Area of Transmission Line Corridor: 1400 ha (3500 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,430,000 1,570,000 1,690,000 1,890,000 2,120,000


Edwin I. Hatch Nuclear Plant


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Location:

Appling County Georgia

18 km (11 miles) N of Baxley

latitude 31.9342°N; longitude 82.3444°W

Licensee: Georgia Power Co.

Unit Information Unit 1 Unit 2
Docket Number 50-321 50-366
Construction Permit 1969 1972
Operating License 1974 1978
Commercial Operation 1975 1979
License Expiration 2014 2018
Licensed Thermal Power [MW(t)] 2436 2436
Design Electrical Rating [net MW(e)] 776 784
Type of Reactor BWR BWR
Nuclear Steam Supply System Vendor GE GE

Cooling Water System

Type: mechanical draft cooling towers

Source: Altamaha River

Source Temperature Range: 6-32°C (43-90°F)

Condenser Flow Rate: 35.1 m3/s (556,000 gal/min) each unit

Design Condenser Temperature Rise: 11°C (20°F)

Intake Structure : at edge of river

Discharge Structure: 37 m (120 ft) from shore

Site Information

Total Area: 908 ha (2244 acres)

Exclusion Distance: 1.26 km (0.78 mile)

Low Population Zone: 1.26 km (0.78 mile)

Nearest City: Savannah; 1980 population: 141,654

Site Topography: flat to rolling

Surrounding Area Topography: flat to rolling

Land Use within 8 km (5 miles): mostly wooded

Nearby Features:

The nearest town is Cedar Crossing about 11 km (7 miles) NNW. U.S. Highway 1 is just west of the site.

Area of Transmission Line Corridor: 1898 ha (4691 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
330,000 360,000 380,000 420,000 460,000


Hope Creek Generating Station


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Location:

Salem County, New Jersey

13 km (8 miles) SW of Salem

latitude 39.4678°N; longitude 75.5381°W

Licensee: Public Service Electric and Gas Co.

Unit Information Unit 1
Docket Number 50-354
Construction Permit 1974
Operating License 1986
Commercial Operation 1986
License Expiration 2026
Licensed Thermal Power [MW(t)] 3293
Design Electrical Rating [net MW(e)] 1067
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: natural draft cooling tower

Source: Delaware River

Source Temperature Range: 1-27°C (34-81°F)

Condenser Flow Rate: 34.8 m3/s (552,000 gal/min)

Design Condenser Temperature Rise: 16°C (28°F)

Intake Structure : at edge of river

Discharge Structure: pipe 3 m (10 ft) offshore

Site Information

Total Area: 300 ha (740 acres)

Exclusion Distance: 0.90-km (0.56-mile) radius

Low Population Zone: 8.05-km (5.00-mile) radius

Nearest City: Wilmington, Delaware; 1980 population: 70,195

Site Topography: flat

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): tidal marshes and grasslands

Nearby Features:

The nearest town is Port Penn about 6 km (4 miles) NW in Delaware. The nearest railroad is 13 km (8 miles) NE. The plant is on the same site as the Salem Nuclear Generating Sation.

Area of Transmission Line Corridor: 369 ha (912 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
4,850,000 4,960,000 5,050,000 5,230,000 5,420,000


Indian Point Station


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Location:

Westchester County, New York

39 km (24 miles) N of New York City

latitude 41.2714°N; longitude 73.9525°W

Licensee: Consolidated Edison Co. of New York, Inc. (Unit 2) Power Authority of the State of New York (Unit 3)

Unit Information Unit 2 Unit 3
Docket Number 50-247 50-286
Construction Permit 1966 1969
Operating License 1973 1976
Commercial Operation 1974 1976
License Expiration 2013 2016
Licensed Thermal Power [MW(t)] 2758 3025
Design Electrical Rating [net MW(e)] 873 965
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: once through

Source: Hudson River

Source Temperature Range: 0-26°C (32-78°F)

Condenser Flow Rate: 53 m3/s (840,000 gal/min) each unit

Design Condenser Temperature Rise: 9.2°C (16.6°F)

Intake Structure : concrete structure at river bank

Discharge Structure: discharge channel to river exiting through 12 ports

Site Information

Total Area: 96.7 ha (239 acres)

Exclusion Distance: 0.32-km (0.20-mile) radius

Low Population Zone: 1.05-km (0.65-mile) radius

Nearest City: White Plains; 1980 population: 46,999

Site Topography: hilly

Surrounding Area Topography: hilly to mountainous

Land Use within 8 km (5 miles): residential, parks, military reservations

Nearby Features:

The nearest town is Buchannan 3 km (2 miles) ESE. Camp Smith (military) is 1.6 km (1 mile) N and West Point is 13 km (8 miles) N.

Area of Transmission Line Corridor: 4 ha (10 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
15,190,000 15,000,000 14,890,000 15,200,000 15,520,000


Kewaunee Nuclear Power Plant


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Location:

Kewaunee County, Wisconsin

43 km (27 miles) E of Green Bay

latitude 44.3431°N; longitude 87.5361°W

Licensee: Wisconsin Public Service Corp.

Unit Information
Docket Number 50-305
Construction Permit 1968
Operating License 1973
Commercial Operation 1974
License Expiration 2013
Licensed Thermal Power [MW(t)] 1650
Design Electrical Rating [net MW(e)] 535
Type of Reactor PWR
Nuclear Steam Supply System Vendor WEST

Cooling Water System

Type: once through

Source: Lake Michigan

Source Temperature Range: 1-19°C (34-67°F)

Condenser Flow Rate: 27 m3/s (420,000 gal/min)

Design Condenser Temperature Rise: 11°C (19°F)

Intake Structure : intake crib 4.6 km (15 ft) deep 533 m (1750 ft) from shore

Discharge Structure: at shoreline

Site Information

Total Area: 367 ha (908 acres)

Exclusion Distance: 1.21 km (0.75 mile)

Low Population Zone: 4.83-km (3.00-mile) radius

Nearest City: Green Bay; 1980 population: 87,899

Site Topography: flat to rolling

Surrounding Area Topography: flat to rolling

Land Use within 8 km (5 miles): agricultural and dairy farming

Nearby Features:

The nearest town is Two Creeks about 5 km (3 miles) S. Point Beach Nuclear Plant is about 8 km (5 miles) S.

Area of Transmission Line Corridor: 431.4 ha (1066 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
640,000 670,000 690,000 730,000 780,000


La Salle County Station


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Location:

La Salle County, Illinois

18 km (11 miles) SE of Ottawa

latitude 41.2439°N; longitude 88.6708°W

Licensee: Commonwealth Edison Co.

Unit Information Unit 1 Unit 2
Docket Number 50-373 50-374
Construction Permit 1973 1973
Operating License 1982 1984
Commercial Operation 1984 1984
License Expiration 2022 2024
Licensed Thermal Power [MW(t)] 3323 3323
Design Electrical Rating [net MW(e)] 1078 1078
Type of Reactor BWR BWR
Nuclear Steam Supply System Vendor GE GE

Cooling Water System

Type: cooling pond

Source: Illinois River

Source Temperature Range: 8-29°C (47-85°F)

Condenser Flow Rate: 40.7 m3/s (645,000 gal/min) each unit

Design Condenser Temperature Rise: 13°C (24°F)

Intake Structure : intake from 832.8-ha (2058-acre) cooling pond, makeup from river

Discharge Structure: discharge to cooling pond

Site Information

Total Area: 1240 ha (3060 acres)

Exclusion Distance: 0.51 km (0.32 mile)

Low Population Zone: 6.41 km (3.98 miles)

Nearest City: Joliet; 1980 population: 77,956

Site Topography: flat

Surrounding Area Topography: flat with hills along river

Land Use within 8 km (5 miles): agricultural

Nearby Features:

The nearest town is Seneca about 8 km (5 miles) NNE. Braidwood Station (nuclear plant) is about 32 km (20 miles) ENE and Dresden Nuclear Power Station is about 35 km (22 miles) NE.

Area of Transmission Line Corridor: 921.9 ha (2278 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,160,000 1,220,000 1,260,000 1,310,000 1,370,000


Limerick Generating Station


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Location:

Montgomery County, Pennsylvania

34 km (21 miles) NW of Philadelphia

latitude 40.2200°N; longitude 75.5900°W

Licensee: Philadelphia Electric Co.

Unit Information Unit 1 Unit 2
Docket Number 50-352 50-353
Construction Permit 1974 1974
Operating License 1985 1990
Commercial Operation 1986 1990
License Expiration 2025 2020
Licensed Thermal Power [MW(t)] 3293 3293
Design Electrical Rating [net MW(e)] 1055 1055
Type of Reactor BWR BWR
Nuclear Steam Supply System Vendor GE GE

Cooling Water System

Type: natural draft cooling towers

Source: Schuylkill River

Source Temperature Range: 6-28°C (42-82°F)

Condenser Flow Rate: 28 m3/s (450,000 gal/min) each unit

Design Condenser Temperature Rise: 17°C (30°F)

Intake Structure : intake from river

Discharge Structure: discharge to river

Site Information

Total Area: 241 ha (595 acres)

Exclusion Distance: 0.76 km (0.47 mile)

Low Population Zone: 2.09-km (1.30-mile) radius

Nearest City: Reading; 1980 population: 78,686

Site Topography: rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): agricultural and undeveloped

Nearby Features:

The nearest town is Linfield about 1.6 km (1 mile) SE. Valley Forge State Park is 16 km (10 miles) SSE. U.S. Highway I-76 is about 16 km (10 miles) S.

Area of Transmission Line Corridor: 3 ha (7 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
6,970,000 7,070,000 7,170,000 7,390,000 7,620,000


Maine Yankee Atomic Power Plant


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Location:

Lincoln County, Maine

16 km (10 miles) NE of Bath

latitude 43.9506°N; longitude 69.6961°W

Licensee: Maine Yankee Atomic Power Co.

Unit Information
Docket Number 50-309
Construction Permit 1968
Operating License 1973
Commercial Operation 1972
License Expiration 2013
Licensed Thermal Power [MW(t) 2700
Design Electrical Rating [net MW(e)] 825
Type of Reactor PWR
Nuclear Steam Supply System Vendor CE

Cooling Water System

Type: once through

Source: Back River

Source Temperature Range: 3-14°C (37-57°F)

Condenser Flow Rate: 26.9 m3/s (426,000 gal/min)

Design Condenser Temperature Rise: 14.2°C (25.6°F)

Intake Structure : at river bank

Discharge Structure: to Montsweag Bay on Back River

Site Information

Total Area: 300 ha (740 acres)

Exclusion Distance: 0.61-km (0.38-mile) radius

Low Population Zone: 9.66-km (6.00-mile) radius

Nearest City: Portland; 1980 population: 61,572

Site Topography: flat to rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): wooded and some idle farm land

Nearby Features:

The nearest town is Edgecomb about 5 km (3 miles) E. U.S. Highway 1 and the Maine Central Railroad are about 1.6 km (1 mile) NE.

Area of Transmission Line Corridor: 89 ha (220 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
640,000 700,000 750,000 830,000 920,000


William B. McGuire Nuclear Station


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Location:

Mecklenburg County, North Carolina

27 km (17 miles) NNW of Charlotte

latitude 35.4322°N; longitude 80.9483°W

Licensee: Duke Power Co.

Unit Information Unit 1 Unit 2
Docket Number 50-369 50-370
Construction Permit 1973 1973
Operating License 1981 1983
Commercial Operation 1981 1984
License Expiration 2021 2023
Licensed Thermal Power [MW(t)] 3411 3411
Design Electrical Rating [net MW(e)] 1180 1180
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: once through

Source: Lake Norman

Source Temperature Range: 3-32°C (38-89°F)

Condenser Flow Rate: 42.6 m3/s (675,000 gal/min) each unit

Design Condenser Temperature Rise: 12.3°C (22.1°F)

Intake Structure : submerged and surface intakes at shoreline

Discharge Structure: 610-m (2000-ft) discharge canal

Site Information

Total Area: 12,100 ha (30,000 acres)

Exclusion Distance: 0.76-km (0.47-mile) radius

Low Population Zone: 8.85 km (5.50 miles)

Nearest City: Charlotte; 1980 population: 315,474

Site Topography: rolling

Surrounding Area Topography: hilly

Land Use within 8 km (5 miles): agricultural and wooded

Nearby Features:

The nearest town is Lowesville about 5 km (3 miles) W. The dam forming Lake Norman and a hydro powerplant are adjacent to the site.

Area of Transmission Line Corridor: 25 ha (62 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,750,000 1,900,000 2,040,000 2,280,000 2,540,000


Millstone Nuclear Power Station


[ Prev | Next | Top of file ]

Location:

New London County, Connecticut 5 km (3 miles) WSW of New London latitude 41.3086°N; longitude 72.1681°W

Licensee: Northeast Utilities

Unit Information Unit 1 Unit 2 Unit 3
Docket Number 50-245 50-336 50-423
Construction Permit 1966 1970 1974
Operating License 1970 1975 1986
Commercial Operation 1971 1975 1986
License Expiration 2010 2015 2026
Licensed Thermal Power [MW(t)] 2011 2700 3411
Design Electrical Rating [net MW(e)] 660 870 1154
Type of Reactor BWR PWR PWR
Nuclear Steam Supply System Vendor GE CE WEST

Cooling Water System

Type: once through

Source: Long Island Sound

Source Temperature Range: 2-22°C (36-72°F)

Condenser Flow Rate: 27 m3/s (420,000 gal/min) for Unit 1 32.97 m3/s (522,500 gal/min) for Unit 2 57.2108 m3/s (906,668 gal/min) for Unit 3

Design Condenser Temperature Rise: 12°C (21°F) for Unit 1 13°C (24°F) for Unit 2 9.7°C (17.5°F) for Unit 3

Intake Structure: on shore of Niantic Bay off Long Island Sound

Discharge Structure: discharge to Niantic Bay via holding pond

Site Information

Total Area: 200 ha (500 acres)

Exclusion Distance: 0.55-km (0.34-mile) minimum

Low Population Zone: 3.86-km (2.40-mile) radius

Nearest City: New Haven; 1980 population: 126,089

Site Topography: flat

Surrounding Area Topography: flat to rolling

Land Use within 8 km (5 miles): mostly undeveloped with some recreational, agricultural, and residential

Nearby Features:

The nearest town is Niantic 3 km (2 miles) NW. U.S. Highway I-95 is about 6 km (4 miles) NNE. Stone Ranch Military Reservation is about 10 km (6 miles) NW. Harkness Memorial State Park, Bluff Point State Park, and Rocky Neck State Park are within 8 km (5 miles) of the site. The U.S. Dept. of Agriculture Plum Island facility is 16 km (10 miles) S in Long Island Sound. The Haddam Neck Plant (nuclear) is 32 km (20 miles) NW.

Area of Transmission Line Corridor: 375 ha (927 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
2,760,000 2,860,000 2,960,000 3,140,000 3,330,000


Monticello Nuclear Generating Plant


[ Prev | Next | Top of file ]

Location:

Wright County, Minnesota 56 km (35 miles) NW of Minneapolis latitude 45.3333°N; longitude 93.8483°W

Licensee: Northern States Power Co.

Unit Information
Docket Number 50-263
Construction Permit 1967
Operating License 1970
Commercial Operation 1971
License Expiration 2010
Licensed Thermal Power [MW(t)] 1670
Design Electrical Rating [net MW(e)] 545
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: once through and helper towers

Source: Mississippi River

Source Temperature Range: 0-29°C (32-85°F)

Condenser Flow Rate: 18 m3/s (280,000 gal/min)

Design Condenser Temperature Rise: 14.9°C (26.8°F)

Intake Structure: canal

Discharge Structure: canal

Site Information

Total Area: 860 ha (2150 acres)

Exclusion Distance: 0.48 km (0.30 mile)

Low Population Zone: 1.61 km (1.00 mile)

Nearest City: Minneapolis; 1980 population: 370,951

Site Topography: flat terraces

Surrounding Area Topography: flat to gently sloping

Land Use within 8 km (5 miles): agricultural and dairy farming

Nearby Features:

The business district of Monticello is about 3.2 km (2 miles) SE. Sherburne National Wildlife Refuge is about 14 km (9 miles) N. Lake Maria State Park is about 10 km (6 miles) WSW and Sand Dunes State Forest and campground are 14 km (9 miles) NE.

Area of Transmission Line Corridor: 588.4 ha (1454 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
2,170,000 2,360,000 2,520,000 2,820,000 3,150,000


North Anna Power Station


[ Prev | Next | Top of file ]

Location:

Louisa County, Virginia 64 km (40 miles) NW of Richmond latitude 38.0608°N; longitude 77.7906°W

Licensee: Virginia Electric and Power Co.

Unit Information Unit 1 Unit 2
Docket Number 50-338 50-339
Construction Permit 1971 1971
Operating License 1978 1980
Commercial Operation 1978 1980
License Expiration 2018 2020
Licensed Thermal Power [MW(t)] 2893 2893
Design Electrical Rating [net MW(e)] 907 907
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: once through

Source: Lake Anna

Source Temperature Range: 9-28°C (48-83°F)

Condenser Flow Rate: 59.33 m3/s (940,300 gal/min) each unit

Design Condenser Temperature Rise: 8°C (14°F)

Intake Structure: intake on lake shore

Discharge Structure: discharged to lake via a 1400-ha (3400-acre) cooling pond.

Site Information

Total Area: 7545 ha (18,643 acres)

Exclusion Distance: 1.35 km (0.84 mile)

Low Population Zone: 9.66 km (6.00 miles)

Nearest City: Richmond; 1980 population: 219,214

Site Topography: rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): agricultural and wooded

Nearby Features:

The nearest town is Centreville 1.6 km (1 mile) SW. Fredericksburg and Spotsylvania National Military Park is about 24 km (15 miles) NE.

Area of Transmission Line Corridor: 1428 ha (3528 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,150,000 1,250,000 1,340,000 1,480,000 1,630,000


Nine Mile Point Nuclear Station


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Location:

Oswego County, New York 10 km (6 miles) NE of Oswego latitude 43.5222°N; longitude 76.4100°W

Licensee: Niagra Mohawk Power Corp.

Unit Information Unit 1 Unit 2
Docket Number 50-220 50-410
Construction Permit 1965 1974
Operating License 1968 1987
Commercial Operation 1969 1988
License Expiration 2008 2027
Licensed Thermal Power [MW(t)] 1850 3323
Design Electrical Rating [net MW(e)] 620 1080
Type of Reactor BWR BWR
Nuclear Steam Supply System Vendor GE GE

Cooling Water System

Type: Unit 1 - once through Unit 2 - natural draft cooling tower

Source: Lake Ontario

Source Temperature Range: 1-25°C (33-77°F)

Condenser Flow Rate: 16 m3/s (250,000 gal/min) for Unit 1 37 m3/s (580,000 gal/min) for Unit 2

Design Condenser Temperature Rise: 18°C (32°F) for Unit 1 15°C (27°F) for Unit 2

Intake Structure: separate submerged pipelines about 300 m (1000 ft) offshore

Discharge Structure: diffuser pipe 169 m (555 ft) long serving both units

Site Information

Total Area: 360 ha (900 acres)

Exclusion Distance: 1.19 km (0.74 mile) minimum

Low Population Zone: 6.44 km (4.00 mile) radius

Nearest City: Syracuse; 1980 population: 170,105

Site Topography: flat to rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): agricultural, industrial, residential, and recreational

Nearby Features:

The nearest town is Lakeview about 1.6 km (1 mile) WSW. Fort Ontario is about 10 km (6 miles) SW. James A. FitzPatrick Nuclear Power Plant is 0.8 km (0.5 mile) E.

Area of Transmission Line Corridor: 664 ha (1640 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
820,000 810,000 790,000 800,000 810,000


Oconee Nuclear Station


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Location:

Oconee County, South Carolina 42 km (26 miles) W of Greenville latitude 34.7917°N; longitude 82.8986°W

Licensee: Duke Power Co.

Unit Information Unit 1 Unit 2 Unit 3
Docket Number 50-269 50-270 50-287
Construction Permit 1967 1967 1967
Operating License 1973 1973 1974
Commercial Operation 1973 1974 1974
License Expiration 2013 2013 2014
Licensed Thermal Power [MW(t)] 2568 2568 2568
Design Electrical Rating [net MW(e)] 887 887 887
Type of Reactor PWR PWR PWR
Nuclear Steam Supply System Vendor B&W B&W B&W

Cooling Water System

Type: once through

Source: Lake Keowee

Source Temperature Range: 7-25°C (44-77°F)

Condenser Flow Rate: 43 m3/s (680,000 gal/min) for each unit

Design Condenser Temperature Rise: 9.6°C (17.2°F)

Intake Structure: A skimmer wall draws water from depths of 216-223 m (710-733 ft) at a velocity of 0.2 m/s (0.6 ft/s).

Discharge Structure: All three units discharge through one structure near Keowee dam. Discharge is underwater at a depth of 233 m (765 ft).

Site Information

Total Area: 210 ha (510 acres)

Exclusion Distance: 1.61-km (1.00-mile) radius

Low Population Zone: 9.66 km (6.00 miles)

Nearest City: Greenville; 1980 population: 58,242

Site Topography: flat to rolling

Surrounding Area Topography: hilly

Land Use within 8 km (5 miles): wooded

Nearby Features:

The nearest town is Six Mile 6 km (4 miles) ENE. Keowee dam is close to the plant. Chattahoochee National Forest is about 24 km (15 miles) W.

Area of Transmission Line Corridor: 3160 ha (7800 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
990,000 1,080,000 1,170,000 1,310,000 1,470,000


Oyster Creek Nuclear Generating Station


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Location:

Ocean County, New Jersey 14 km (9 miles) S of Toms River latitude 39.8142°N; longitude 74.2064°W

Licensee: GPU Nuclear Corp.

Unit Information Unit 1
Docket Number 50-219
Construction Permit 1964
Operating License 1969
Commercial Operation 1969
License Expiration 2009
Licensed Thermal Power [MW(t)] 1930
Design Electrical Rating [net MW(e)] 650
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: once through

Source: Barnegat Bay

Source Temperature Range: 2-28°C (35-83°F)

Condenser Flow Rate: 29 m3/s (460,000 gal/min)

Design Condenser Temperature Rise: 8°C (14°F)

Intake Structure: Forked River serves as a canal for intake and discharge to Barnegat Bay.

Discharge Structure: Forked River serves as a canal for intake and discharge to Barnegat Bay.

Site Information

Total Area: 573 ha (1416 acres)

Exclusion Distance: 0.40 km (0.25 mile)

Low Population Zone: 3.22 km (2.00 miles)

Nearest City: Atlantic City; 1980 population: 40,199

Site Topography: flat

Surrounding Area Topography: rolling plains to flat lowlands

Land Use within 8 km (5 miles): mostly undeveloped

Nearby Features:

The nearest town is Forked River about 3 km (2 miles) N. The Garden State Parkway is 1.6 km (1 mile) W. There is a large influx of people seeking recreation in the summer.

Area of Transmission Line Corridor: 130 ha (322 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
4,030,000 4,190,000 4,300,000 4,560,000 4,840,000


Palisades Nuclear Plant


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Location:

Van Buren County, Michigan 56 km (35 miles) W of Kalamazoo latitude 42.3222°N; longitude 86.3153°W

Licensee: Consumers Power Co.

Unit Information Unit 1
Docket Number 50-255
Construction Permit 1967
Operating License 1972
Commercial Operation 1973
License Expiration 2012
Licensed Thermal Power [MW(t)] 2530
Design Electrical Rating [net MW(e)] 805
Type of Reactor PWR
Nuclear Steam Supply System Vendor CE

Cooling Water System

Type: mechanical draft cooling towers

Source: Lake Michigan

Source Temperature Range: 2-24°C (35-75°F)

Condenser Flow Rate: 25.6 m3/s (405,000 gal/min)

Design Condenser Temperature Rise: 14°C (25°F)

Intake Structure: intake crib 1000 m (3300 ft) from shore

Discharge Structure: canal 33 m (108 ft) long

Site Information

Total Area: 197 ha (487 acres)

Exclusion Distance: 0.71-km (0.44-mile) radius

Low Population Zone:

Nearest City: Kalamazoo; 1980 population: 79,722

Site Topography: flat to rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): agricultural, wooded, berry farms, and orchards

Nearby Features:

The nearest town is South Haven about 6 km (4 miles) N. Van Buren State Park joins the plant on the north. Many tourists come to the beaches in the summer. The C&O Railway is about 3 km (2 miles) E. Highway I-196 is about 1.6 km (1 mile) E.

Area of Transmission Line Corridor: 910 ha (2250 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,170,000 1,220,000 1,260,000 1,340,000 1,420,000


Palo Verde Nuclear Generating Station


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Location:

Maricopa County, Arizona 55 km (34 miles) W of Phoenix latitude 33.3881°N; longitude 112.8644°W

Licensee: Arizona Public Service Co.

Unit Information Unit 1 Unit 2 Unit 3
Docket Number 50-528 50-529 50-530
Construction Permit 1976 1976 1976
Operating License 1985 1986 1987
Commercial Operation 1986 1986 1988
License Expiration 2025 2026 2027
Licensed Thermal Power [MW(t)] 3800 3800 3800
Design Electrical Rating [net MW(e)] 1270 1270 1270
Type of Reactor PWR PWR PWR
Nuclear Steam Supply System Vendor CE CE CE

Cooling Water System

Type: mechanical draft cooling towers treatment plant

Source: Phoenix city sewage Source Temperature Range:

Condenser Flow Rate: 35 m3/s (560,000 gal/min) each unit

Design Condenser Temperature Rise: 17.8°C (32.1°F)

Intake Structure: 56-km (35-mile) underground pipeline from Phoenix 91st Avenue Sewage Treatment Plant

Discharge Structure: blowdown from the circulating water system is directed to on-site evaporation ponds without requiring any off-site discharge

Site Information

Total Area: 1640 ha (4050 acres)

Exclusion Distance: 0.87-km (0.54-mile) minimum

Low Population Zone: 6.44-km (4.00-mile) radius

Nearest City: Phoenix; 1980 population: 789,704

Site Topography: flat with hills

Surrounding Area Topography: flat with hills

Land Use within 8 km (5 miles): open desert with some agriculture

Nearby Features:

The nearest town is Wintersburg about 5 km (3 miles) N. U.S. Highway I-10 is about 11 km (7 miles) N. The Southern Pacific Railroad is about 8 km (5 miles) SE.

Area of Transmission Line Corridor: 6720 ha (16,600 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,180,000 1,330,000 1,450,000 1,690,000 1,970,000


Peach Bottom Atomic Power Station


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Location:

York County, Pennsylvania 29 km (18 miles) S of Lancaster latitude 39.7589°N; longitude 76.2692°W

Licensee: Philadelphia Electric Co.

Unit Information Unit 2 Unit 3
Docket Number 50-277 50-278
Construction Permit 1968 1968
Operating License 1973 1974
Commercial Operation 1974 1974
License Expiration 2013 2014
Licensed Thermal Power [MW(t)] 3293 3293
Design Electrical Rating [net MW(e)] 1065 1065
Type of Reactor BWR BWR
Nuclear Steam Supply System Vendor GE GE

Cooling Water System

Type: once through with helper towers

Source: Conowingo Pond

Source Temperature Range: 1-27°C (34-80°F)

Condenser Flow Rate: 47 m3/s (750,000 gal/min) each unit

Design Condenser Temperature Rise: 11.6°C (20.8°F)

Intake Structure: intake from Conowingo Pond through a small intake pond

Discharge Structure: 1520-m (5000-ft) canal to Conowingo Pond

Site Information

Total Area: 250 ha (620 acres)

Exclusion Distance: 0.82-km (0.51-mile) minimum

Low Population Zone: 2.22 km (1.38 miles)

Nearest City: Lancaster; 1980 population: 54,725

Site Topography: rolling to hilly

Surrounding Area Topography: rolling to hilly

Land Use within 8 km (5 miles): agricultural and wooded

Nearby Features:

The nearest town is Slate Hill 3 km (2 miles) SW. Susquehanna State Park is about 5 km (3 miles) N. U.S. Highway I- 95 is about 24 km (15 miles) SE. Conowingo Dam, about 13 km (8 miles) SE on the Susquehanna River, forms Conowingo Pond. Unit 1 is a 40 Mwe nuclear plant on the same site and was retired from service in 1974. Three Mile Island Nuclear Station is 56 km (35 miles) upstream on the Susquehanna River.

Area of Transmission Line Corridor: 417 ha (1030 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
4,660,000 4,850,000 5,010,000 5,280,000 5,570,000


Perry Nuclear Power Plant


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Location:

Lake County, Ohio 11 km (7 miles) NE of Painesville latitude 41.8008°N; longitude 81.1442°W

Licensee: Cleveland Electric Illuminating Co.

Unit Information Unit 1
Docket Number 50-440
Construction Permit 1977
Operating License 1986
Commercial Operation 1987
License Expiration 2026
Licensed Thermal Power [MW(t)] 3579
Design Electrical Rating [net MW(e)] 1205
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: natural draft cooling tower

Source: Lake Erie

Source Temperature Range: 0-26°C (32-79°F)

Condenser Flow Rate: 34.41 m3/s (545,400 gal/min)

Design Condenser Temperature Rise: 18°C (32°F)

Intake Structure: submerged multiport structure 777 m (2550 ft) offshore

Discharge Structure: submerged diffuser 503 m (1650 ft) offshore

Site Information

Total Area: 450 ha (1100 acres)

Exclusion Distance: 0.89-km (0.55-mile) radius

Low Population Zone: 4.02 km (2.50 miles)

Nearest City: Euclid; 1980 population: 59,999

Site Topography: flat

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): forest land, agricultural (horticulture), residential, industrial, and some recreational

Nearby Features:

The nearest town is North Perry 1.6 km (1 mile) SW. The Penn Central Railroad is about 5 km (3 miles) S. U.S. Highway I- 90 is about 8 km (5 miles) S.

Area of Transmission Line Corridor: 610 ha (1500 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
2,480,000 2,530,000 2,570,000 2,670,000 2,770,000


Pilgrim Nuclear Power Station


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Location:

Plymouth County, Massachusetts 6 km (4 miles) SE of Plymouth latitude 41.9444°N; longitude 70.5794°W

Licensee: Boston Edison Co.

Unit Information Unit 1
Docket Number 50-293
Construction Permit 1968
Operating License 1972
Commercial Operation 1972
License Expiration 2012
Licensed Thermal Power [MW(t)] 1998
Design Electrical Rating [net MW(e)] 655
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: once through

Source: Cape Cod Bay

Source Temperature Range: 0-28°C (32-83°F)

Condenser Flow Rate: 19.6 m3/s (311,000 gal/min)

Design Condenser Temperature Rise: 16°C (29°F)

Intake Structure: concrete structure at edge of bay protected by a breakwater

Discharge Structure: canal about 260 m (850 ft) long

Site Information

Total Area: 209 ha (517 acres)

Exclusion Distance: 0.53 km (0.33 mile)

Low Population Zone: 6.76 km (4.20 miles)

Nearest City: Brockton; 1980 population: 95,172

Site Topography: flat to rolling

Surrounding Area Topography: rolling to hilly

Land Use within 8 km (5 miles): mostly undeveloped

Nearby Features:

The nearest town is Plymouth about 6 km (4 miles) NW. Miles Standish State Forest is about 10 km (6 miles) SW. Plymouth Rock and Plimoth Plantation historical sites are about 8 km (5 miles) W.

Area of Transmission Line Corridor: 70 ha (174 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
4,440,000 4,590,000 4,690,000 4,880,000 5,080,000


Point Beach Nuclear Plant


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Location:

Manitowoc County, Wisconsin 21 km (13 miles) NNW of Manitowoc latitude 44.2808°N; longitude 87.5361°W

Licensee: Wisconsin Electric Power Co.

Unit Information Unit 1 Unit 2
Docket Number 50-266 50-301
Construction Permit 1967 1968
Operating License 1970 1972
Commercial Operation 1970 1972
License Expiration 2010 2012
Licensed Thermal Power [MW(t)] 1519 1519
Design Electrical Rating [net MW(e)] 497 497
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: once through

Source: Lake Michigan

Source Temperature Range:

Condenser Flow Rate: 22 m3/s (350,000 gal/min) each unit

Design Condenser Temperature Rise: 10.7°C (19.3°F)

Intake Structure: Structure is 533 m (1750 ft) from shore in 7-m (22-ft) deep water. Top elevation is 2.4 m (8 ft) above normal lake level. Intake to plant is through 38 pipes located 1.5 m (5 ft) above lake bed.

Discharge Structure: 2 flumes projecting about 46 m (150 ft) from shore

Site Information

Total Area: 836 ha (2065 acres)

Exclusion Distance: 1.19-km (0.74-mile) radius

Low Population Zone: 9.01 km (5.60 miles)

Nearest City: Green Bay; 1980 population: 87,899

Site Topography: flat to rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): agricultural, dairy farming, vegetable canning

Nearby Features:

The nearest town is Two Creeks 1.6 km (1 mile) NNW. Point Beach State Forest is just S of site. The Kewaunee Nuclear Power Plant is about 8 km (5 miles) N.

Area of Transmission Line Corridor: 1344 ha (3321 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
610,000 640,000 660,000 700,000 740,000


Prairie Island Nuclear Generating Plant


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Location:

Goodhue County, Minnesota 45 km (28 miles) SE of Minneapolis latitude 44.6219°N; longitude 92.6331°W

Licensee: Northern States Power Co.

Unit Information Unit 1 Unit 2
Docket Number 50-282 50-306
Construction Permit 1968 1968
Operating License 1973 1974
Commercial Operation 1973 1974
License Expiration 2013 2014
Licensed Thermal Power [MW(t)] 1650 1650
Design Electrical Rating [net MW(e)] 530 530
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: mechanical draft and/or once cooling towers

Source: Mississippi River

Source Temperature Range: 0-28°C (32-82°F)

Condenser Flow Rate: 18.6 m3/s (294,000 gal/min) each unit

Design Condenser Temperature Rise: 15°C (27°F)

Intake Structure: short canal

Discharge Structure: Discharges to a basin then to towers and/or river.

Site Information

Total Area: 230 ha (560 acres)

Exclusion Distance: 0.69-km (0.43-mile) radius

Low Population Zone: 2.41 km (1.50 miles)

Nearest City: Minneapolis; 1980 population: 370,951

Site Topography: flat to rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): dairy farming and agricultural

Nearby Features:

The business district of the town of Red Wing is 9.6 km (6 miles) SE. A railroad line is just SW of the site.

Area of Transmission Line Corridor: 394 ha (973 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
2,290,000 2,490,000 2,650,000 2,960,000 3,310,000


Quad-Cities Station


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Location:

Rock Island County, Illinois 32 km (20 miles) NE of Moline latitude 41.7261°N; longitude 90.3100°W

Licensee: Commonwealth Edison Co.

Unit Information Unit 1 Unit 2
Docket Number 50-254 50-265
Construction Permit 1967 1967
Operating License 1972 1972
Commercial Operation 1973 1973
License Expiration 2012 2012
Licensed Thermal Power [MW(t)] 2511 2511
Design Electrical Rating [net MW(e)] 789 789
Type of Reactor BWR BWR
Nuclear Steam Supply System Vendor GE GE

Cooling Water System

Type: once through

Source: Mississippi River

Source Temperature Range: 0-29°C (32-85°F)

Condenser Flow Rate: 29.7 m3/s (471,000 gal/min) each unit

Design Condenser Temperature Rise: 13°C (24°F)

Intake Structure: crib house at edge of river

Discharge Structure: 4300-m (14,000-ft) spray canal

Site Information

Total Area: 317 ha (784 acres)

Exclusion Distance: 0.80 km (0.50 mile)

Low Population Zone: 4.83 km (3.00 miles)

Nearest City: Davenport, Iowa; 1980 population: 103,264

Site Topography: flat

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): agricultural and small industrial park

Nearby Features:

The nearest town is Folletts 5 km (3 miles) NW. The Rock Island Railroad is 3 km (2 miles) W and the Chicago, Milwaukee, and St. Paul Railroad is 1.6 km (1 mile) E. The Rock Island Arsenal is about 24 km (15 miles) SW.

Area of Transmission Line Corridor: 570 ha (1400 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
740,000 760,000 780,000 810,000 850,000


Rancho Seco Nuclear Generating Station


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Location:

Sacramento County, California 40 km (25 miles) SE of Sacramento latitude 38.3444°N; longitude 121.1200°W

Licensee: Sacramento Municipal Utility District

Unit Information Unit 1
Docket Number 50-312
Construction Permit 1968
Operating License 1974
Commercial Operation 1975
License Expiration 2014
Licensed Thermal Power [MW(t)] 2772
Design Electrical Rating [net MW(e)] 918
Type of Reactor PWR
Nuclear Steam Supply System Vendor B&W

Cooling Water System

Type: natural draft cooling

Source: Folsom Canal towers

Source Temperature Range: 10-21°C (50-70°F)

Condenser Flow Rate: 28.1 m3/s (446,000 gal/min)

Design Condenser Temperature Rise: 16°C (28°F)

Intake Structure: 5.6-km (3.5-mile) pipeline from Folsom Canal

Discharge Structure: 2.4-km (1.5-mile) pipeline to reservoir

Site Information

Total Area: 1000 ha (2480 acres)

Exclusion Distance: 0.64-km (0.40-mile) radius

Low Population Zone: 8.05 km (5.00 miles)

Nearest City: Sacramento; 1980 population: 275,741

Site Topography: flat to rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): agricultural and grazing land

Nearby Features:

The nearest town is Clay 3 km (2 miles) WSW. The Southern Pacific Railroad is about 1.6 km (1 mile) N.

Area of Transmission Line Corridor: 350 ha (870 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
2,010,000 2,200,000 2,360,000 2,590,000 2,850,000

Note: This plant was shut down as the result of a public referendum in June 1989.


River Bend Station


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Location:

West Feliciana County, Louisiana 39 km (24 miles) NNW of Baton Rouge latitude 30.7569°N; longitude 91.3314°W

Licensee: Gulf States Utility Co.

Unit Information Unit 1
Docket Number 50-458
Construction Permit 1977
Operating License 1985
Commercial Operation 1986
License Expiration 2025
Licensed Thermal Power [MW(t)] 2894
Design Electrical Rating [net MW(e)] 936
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: mechanical draft cooling towers

Source: Mississippi River

Source Temperature Range: Condenser Flow Rate: 32.084 m3/s (508,470 gal/min)

Design Condenser Temperature Rise: 15°C (27°F)

Intake Structure: at river bank

Discharge Structure: pipe extending into the river

Site Information

Total Area: 1352 ha (3342 acres)

Exclusion Distance: 0.92-km (0.57-mile) radius

Low Population Zone: 4.02-km (2.50-mile) radius

Nearest City: Baton Rouge; 1980 population: 220,394

Site Topography: flat

Surrounding Area Topography: flat to rolling

Land Use within 8 km (5 miles): agricultural and forest

Nearby Features:

The nearest town is St. Francisville 5 km (3 miles) NW. Audubon Memorial State Park is about 5 km (3 miles) NNE. The Illinois Central Railroad crosses the site.

Area of Transmission Line Corridor: 410 ha (1014 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
800,000 860,000 920,000 1,010,000 1,110,000


H. B. Robinson Plant


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Location:

Darlington County, South Carolina 42 km (26 miles) NE of Florence latitude 34.4025°N; longitude 80.1586°W

Licensee: Carolina Power and Light Co.

Unit Information Unit 2
Docket Number 50-261
Construction Permit 1967
Operating License 1970
Commercial Operation 1971
License Expiration 2010
Licensed Thermal Power [MW(t)] 2300
Design Electrical Rating [net MW(e)] 700
Type of Reactor PWR
Nuclear Steam Supply System Vendor WEST

Cooling Water System

Type: once through

Source: Lake Robinson

Source Temperature Range: 8-29°C (46-85°F)

Condenser Flow Rate: 30.42 m3/s (482,100 gal/min)

Design Condenser Temperature Rise: 10°C (18°F)

Intake Structure: concrete structure on edge of lake

Discharge Structure: 6.8-km (4.2-mile) canal discharging about 6 km (4 miles) upstream from intake

Site Information

Total Area: 2000 ha (5000 acres)

Exclusion Distance: 0.43-km (0.27-mile) radius

Low Population Zone: 7.24 km (4.50 miles)

Nearest City: Columbia; 1980 population: 101,229

Site Topography: rolling

Surrounding Area Topography: rolling

Land Use within 8 km (5 miles): agricultural and wooded, some recreational

Nearby Features:

The nearest town is Hartsville 8 km (5 miles) SE. Unit 1 is an adjacent 185 MW(e) capacity coal-fired plant. Sand Hills State Forest is about 6 km (4 miles) N. The Carolina

Sandhills National Wildlife Refuge is about 8 km (5 miles) NNW.

Area of Transmission Line Corridor: 414 ha (1024 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
740,000 810,000 880,000 990,000 1,120,000


Salem Nuclear Generating Station


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Location:

Salem County, New Jersey 13 km (8 miles) SW of Salem latitude 39.4628°N; longitude 75.5358°W

Licensee: Public Service Electric and Gas Co.

Unit Information Unit 1 Unit 2
Docket Number 50-272 50-311
Construction Permit 1968 1968
Operating License 1976 1981
Commercial Operation 1977 1981
License Expiration 2016 2021
Licensed Thermal Power [MW(t)] 3411 3411
Design Electrical Rating [net MW(e)] 1115 1115
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: once through

Source: Delaware River

Source Temperature Range: 1-26°C (33-79°F)

Condenser Flow Rate: 69 m3/s (1,100,000 gal/min) each unit

Design Condenser Temperature Rise: 7.6°C (13.6°F)

Intake Structure: 12 bay structure on edge of river

Discharge Structure: submerged pipes extending 150 m (500 ft) into the river

Site Information

Total Area: 280 ha (700 acres)

Exclusion Distance: 1.29 km (0.80 mile)

Low Population Zone: 8.05 km (5.00 miles)

Nearest City: Wilmington, Delaware; 1980 population: 70,195

Site Topography: flat

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): tidal marshes and grasslands

Nearby Features:

The nearest town is Port Penn about 6 km (4 miles) NW in Delaware. The nearest railroad is 13 km (8 miles) NE. The plant is on the same site as the Hope Creek Generating Station (nuclear).

Area of Transmission Line Corridor: 1600 ha (3900 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
4,810,000 4,910,000 5,000,000 5,180,000 5,370,000


San Onofre Nuclear Generating Station


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Location:

San Diego County, California 8 km (5 miles) SE of San Clemente latitude 33.3703°N; longitude 117.5569°W

Licensee: Southern California Edison Co.

Unit Information Unit 1 Unit 2 Unit 3
Docket Number 50-206 50-361 50-362
Construction Permit 1964 1973 1973
Operating License 1967 1982 1983
Commercial Operation 1968 1983 1984
License Expiration 2007 2022 2023
Licensed Thermal Power [MW(t)] 1347 3390 3390
Design Electrical Rating [net MW(e)] 436 1070 1080
Type of Reactor PWR PWR PWR
Nuclear Steam Supply System Vendor WEST CE CE

Cooling Water System

Type: once through

Source: Pacific Ocean

Source Temperature Range: 12-23°C (54-73°F)

Condenser Flow Rate: 21.51 m3/s (340,900 gal/min) for Unit 1 50.3 m3/s (797,000 gal/min) each for Units 2 & 3

Design Condenser Temperature Rise: 11°C (19°F) for Unit 1 11°C (20°F) For Units 2 & 3

Intake Structure: Unit 1--intake 980 m (3200 ft) from shore; Units 2 & 3--velocity-cap structure about 1040 m (3400 ft) from shore in water 9 m (30 ft) deep

Discharge Structure: Unit 1-discharged 790 m (2600 ft) from shore in water 7.3 m (24 ft) deep; Units 2 & 3- diffuser port systems extending 1160 m to 2590 m (3800 to 8500 ft) from shore

Site Information

Total Area: 34 ha (84 acres)

Exclusion Distance: 0.60 (0.37 mile)

Low Population Zone: 3.14 km (1.95 miles)

Nearest City: Oceanside; 1980 population: 76,698

Site Topography: narrow sloping coastal plain and sea cliffs

Surrounding Area Topography: hilly

Land Use within 8 km (5 miles): military reservation

Nearby Features:

The nearest town is San Clemente 8 km (5 miles) NW. The site is surrounded by Camp Pendleton Marine Base. Camps on the base are 2.4 km (1.5 miles) or more from the site. U.S. Highway I-5 and the Atchison, Topeka, and Santa Fe Railroad are adjacent to the site to the east.

Area of Transmission Line Corridor: 450 ha (1100 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
5,430,000 5,950,000 6,400,000 7,050,000 7,760,000


Seabrook Station


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Location:

Rockingham County, New Hampshire 21 km (13 miles) SSW of Portsmouth latitude 42.8983°N; longitude 70.8497°W

Licensee: Public Service Company of New Hampshire

Unit Information Unit 1
Docket Number 50-443
Construction Permit 1976
Operating License 1990
Commercial Operation --
License Expiration 2032
Design Thermal Power [MW(t)] 3411
Design Electrical Rating [net MW(e)] 1198
Type of Reactor PWR
Nuclear Steam Supply System Vendor WEST

Cooling Water System

Type: once through

Source: Atlantic Ocean

Source Temperature Range: 3-13°C (37-55°F)

Condenser Flow Rate: 25.2 m3/s (399,000 gal/min)

Design Condenser Temperature Rise: 21°C (38°F)

Intake Structure: 3 structures 15 m (50 ft) below sea level with pipeline submerged about 50 m (175 ft) below mean sea level and extending about 2100 m (7000 ft) offshore

Discharge Structure: submerged pipeline ending in a diffuser located about 1675 m (5500 ft) offshore and about 1525 m (5000 ft) S of intake

Site Information

Total Area: 363 ha (896 acres)

Exclusion Distance: 0.92-km (0.57-mile) minimum

Low Population Zone: 2.01 km (1.25 miles)

Nearest City: Lawrence, Massachusetts; 1980 population: 63,175

Site Topography: flat

Surrounding Area Topography: flat to rolling

Land Use within 8 km (5 miles): undeveloped salt-water marshes with some industrial, residential, and recreational

Nearby Features:

The nearest town is Seabrook 1.6 km (1 mile) W. U.S. Highway I-95 is about 1.6 km (1 mile) W. The Boston and Maine Railroad is adjacent to the site. Hampton Beach State Park is 3 km (2 miles) E.

Area of Transmission Line Corridor: 625 ha (1545 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
3,760,000 3,900,000 4,010,000 4,220,000 4,450,000


Sequoyah Nuclear Plant


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Location:

Hamilton County, Tennessee 16 km (10 miles) NE of Chattanooga latitude 35.2233°N; longitude 85.0878°W

Licensee: Tennessee Valley Authority

Unit Information Unit 1 Unit 2
Docket Number 50-327 50-328
Construction Permit 1970 1970
Operating License 1980 1981
Commercial Operation 1981 1982
License Expiration 2020 2021
Licensed Thermal Power [MW(t)] 3411 3411
Design Electrical Rating [net MW(e)] 1148 1148
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: once through and/or natural draft cooling towers

Source: Chickamauga Lake

Source Temperature Range: 6-28°C (42-83°F)

Condenser Flow Rate: 32.9 m3/s (522,000 gal/min) each unit

Design Condenser Temperature Rise: 17°C (30°F)

Intake Structure: intake from lake

Discharge Structure: discharge to lake

Site Information

Total Area: 212 ha (525 acres)

Exclusion Distance: 0.56 km (0.35 mile)

Low Population Zone: 4.83 km (3.00 miles)

Nearest City: Chattanooga; 1980 population: 169,514

Site Topography: rolling

Surrounding Area Topography: hilly

Land Use within 8 km (5 miles): some residential and recreational

Nearby Features:

The nearest town is Shady Grove about 3 km (2 miles) NW. Harrison Bay State Park is 5 km (3 miles) S. The Volunteer Ordnance Works is about 15 km (9 miles) S. Chickamauga Lake is part of the Tennessee River.

Area of Transmission Line Corridor: 510 ha (1260 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
930,000 1,020,000 1,090,000 1,210,000 1,330,000


Shoreham Nuclear Power Station


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Location:

Suffolk County, New York 19 km (12 miles) NW of Riverhead latitude 40.9583°N; longitude 72.8667°W

Licensee: Long Island Lighting Co.

Unit Information
Docket Number 50-322
Construction Permit 1973
Operating License 1989
Commercial Operation --
License Expiration 2013
Design Thermal Power [MW(t)] 2436
Design Electrical Rating [net MW(e)] 819
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: once through

Source: Long Island Sound

Source Temperature Range: 2-23°C (36-74°F)

Condenser Flow Rate: 36.19 m3/s (573,600 gal/min)

Design Condenser Temperature Rise: 11°C (20°F)

Intake Structure: intake canal

Discharge Structure: diffuser system

Site Information

Total Area: 202 ha (499 acres)

Exclusion Distance: 0.31 km (0.19 mile)

Low Population Zone: 3.22 km (2.00 miles)

Nearest City: New Haven, Connecticut; 1980 population: 126,089

Site Topography: flat to rolling

Surrounding Area Topography: rolling to hilly

Land Use within 8 km (5 miles): some residential and recreational

Nearby Features:

The nearest town is Shoreham 3 km (2 miles) W. Brookhaven State Park is about 3 km (2 miles) S. Brookhaven National Laboratory is about 11 km (7 miles) S. Grumman Peconic River Airport is about 10 km (6 miles) SE. Wildwood State Park is about 6 km (4 miles) E.

Area of Transmission Line Corridor: 16 ha (39 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
5,390,000 5,400,000 5,420,000 5,550,000 5,690,000

Note: This plant has not been allowed to operate due to litigation concerning emergency response.


South Texas Project


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Location:

Matagorda County, Texas 19 km (12 miles) SSW of Bay City latitude 28.7950°N; longitude 96.0481°W

Licensee: Houston Lighting and Power Co.

Unit Information Unit 1 Unit 2
Docket Number 50-498 50-499
Construction Permit 1975 1975
Operating License 1988 1989
Commercial Operation 1988 1989
License Expiration 2028 2029
Licensed Thermal Power [MW(t)] 3800 3800
Design Electrical Rating [net MW(e)] 1250 1250
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: closed cycle cooling reservoir

Source: Colorado River

Source Temperature Range: 14-29°C (58-84°F)

Condenser Flow Rate: 57.26 m3/s (907,400 gal/min) each unit

Design Condenser Temperature Rise: 11°C (19°F)

Intake Structure: on bank of Colorado River

Discharge Structure: on bank of Colorado River

Site Information

Total Area: 4998 ha (12,350 acres)

Exclusion Distance: 1.43-km (0.89-mile) minimum

Low Population Zone: 4.83 km (3.00 miles)

Nearest City: Galveston; 1980 population: 61,902

Site Topography: flat

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): agricultural

Nearby Features:

The nearest town is Matagorda 13 km (8 miles) SE. The Missouri Pacific Railroad is about 8 km (5 miles) NNE. A 40-cm (16-inch) natural gas pipeline is about 3 km (2 miles) NW.

Area of Transmission Line Corridor: 1932 ha (4773 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
270,000 300,000 320,000 350,000 380,000


St. Lucie Plant


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Location:

St. Lucie County, Florida 11 km (7 miles) SE of Fort Pierce latitude 27.3486°N; longitude 80.2464°W

Licensee: Florida Power and Light Co.

Unit Information Unit 1 Unit 2
Docket Number 50-335 50-389
Construction Permit 1970 1977
Operating License 1976 1983
Commercial Operation 1976 1983
License Expiration 2016 2023
Licensed Thermal Power [MW(t)] 2700 2700
Design Electrical Rating [net MW(e)] 830 830
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor CE CE

Cooling Water System

Type: once through

Source: Atlantic Ocean

Source Temperature Range: 31°C (87°F) maximum

Condenser Flow Rate: 30.96 m3/s (490,600 gal/min) each unit

Design Condenser Temperature Rise: 14°C (25°F)

Intake Structure: 370 m (1200 ft) offshore

Discharge Structure: Unit 1 is 370 m (1200 ft) offshore; Unit 2 is a multiport discharge 900 m (3000 ft) offshore; both structures are 730 (2400 ft) from the intake structures.

Site Information

Total Area: 458 ha (1132 acres)

Exclusion Distance: 1.56-km (0.97-mile) radius

Low Population Zone: 1.61 km (1.00 mile)

Nearest City: West Palm Beach; 1980 population: 62,530

Site Topography: flat land and water

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): expanding residential and some recreational

Nearby Features:

The nearest town is Ankona 3 km (2 miles) W. The Florida East Coast Railroad is about 3 km (2 miles) W. The plant is on Hutchinson Island which is separated from the mainland by the Indian River which is part of the intercoastal waterway. A causeway to the mainland is about 10 km (6 miles) SSE.

Area of Transmission Line Corridor: 310 ha (760 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
690,000 780,000 860,000 1,040,000 1,250,000


Virgil C. Summer Nuclear Station


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Location:

Fairfield County, South Carolina 42 km (26 miles) NW of Columbia latitude 34.2958°N; longitude 81.3203°W

Licensee: South Carolina Electric and Gas Co.

Unit Information Unit 1
Docket Number 50-395
Construction Permit 1973
Operating License 1982
Commercial Operation 1984
License Expiration 2022
Licensed Thermal Power [MW(t)] 2775
Design Electrical Rating [net MW(e)] 900
Type of Reactor PWR
Nuclear Steam Supply System Vendor WEST

Cooling Water System

Type: once through

Source: Lake Monticello

Source Temperature Range: 11-33°C (52-91°F)

Condenser Flow Rate: 30.6 m3/s (485,000 gal/min)

Design Condenser Temperature Rise: 14°C (25°F)

Intake Structure: intake at shoreline

Discharge Structure: discharge to lake via a discharge pond

Site Information

Total Area: 890 ha (2200 acres)

Exclusion Distance: 1.63-km (1.01-mile) radius

Low Population Zone: 4.83 (3.00 miles)

Nearest City: Columbia; 1980 population: 101,229

Site Topography: rolling

Surrounding Area Topography: rolling to hilly

Land Use within 8 km (5 miles): mostly wooded with some agricultural

Nearby Features:

The nearest town is Jenkinsville 5 km (3 miles) SE. U.S. Highway I-26 is 11 km (7 miles) SSW. The Southern Railroad is 1.6 km (1 mile) W. The Fairfield pumped storage hydrostation is about 1.6 km (1 mile) NW and uses Lake Monticello as well as the Parr Reservoir.

Area of Transmission Line Corridor: 638 ha (1576 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
910,000 990,000 1,080,000 1,220,000 1,390,000


Surry Power Station


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Location:

Surry County, Virginia 27 km (17 miles) NW of Newport News latitude 37.1656°N; longitude 76.6983°W

Licensee: Virginia Electric and Power Co.

Unit Information Unit 1 Unit 2
Docket Number 50-280 50-281
Construction Permit 1968 1968
Operating License 1972 1973
Commercial Operation 1972 1973
License Expiration 2012 2013
Licensed Thermal Power [MW(t)] 2441 2441
Design Electrical Rating [net MW(e)] 788 788
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: once through

Source: James River

Source Temperature Range: 2-29°C (35-84°F)

Condenser Flow Rate: 53 m3/s (840,000 gal/min) each unit

Design Condenser Temperature Rise: 8°C (14°F)

Intake Structure: 2.7-km (1.7-mile) concrete canal

Discharge Structure: 880-m (2900-ft) canal

Site Information

Total Area: 340 ha (840 acres)

Exclusion Distance: 0.50 km (0.31 mile)

Low Population Zone: 4.83 km (3.00 miles)

Nearest City: Newport News; 1980 population: 144,903

Site Topography: flat

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): agriculture, military reservations, recreation

Nearby Features:

The nearest town is Scotland 8 km (5 miles) W. Jamestown Island, a Federal park, is 6 km (4 miles) NW. Chippokes Plantation, a state park, is 5 km (3 miles) WSW. Jamestown National Historical Park is 8 km (5 miles) WNW. Colonial Williamsburg is 11 km (7 miles) NNW. These numerous attractions bring many visitors to the area. Adjacent to the site on the north is Hog Island, a waterfowl refuge. U.S. Highway I-64 is 19 km (12 miles) NW.

Area of Transmission Line Corridor: 1790 ha (4420 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,900,000 2,080,000 2,240,000 2,510,000 2,800,000


Susquehanna Steam Electric Station


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Location:

Luzerne County, Pennsylvania 11 km (7 miles) NE of Berwick latitude 41.0922°N; longitude 76.1467°W

Licensee: Pennsylvania Power and Light Co.

Unit Information Unit 1 Unit 2
Docket Number 50-387 50-388
Construction Permit 1973 1973
Operating License 1982 1984
Commercial Operation 1983 1985
License Expiration 2022 2024
Licensed Thermal Power [MW(t)] 3293 3293
Design Electrical Rating [net MW(e)] 1050 1050
Type of Reactor BWR BWR
Nuclear Steam Supply System Vendor GE GE

Cooling Water System

Type: natural draft cooling towers

Source: Susquehanna River

Source Temperature Range:

Condenser Flow Rate: 28.3 m3/s (448,000 gal/min) each unit

Design Condenser Temperature Rise: 8°C (14°F)

Intake Structure: at river bank

Discharge Structure: diffuser pipe 73 m (240 ft) from river bank

Site Information

Total Area: 435 ha (1075 acres)

Exclusion Distance: 0.55-km (0.34-mile) radius

Low Population Zone: 4.83 km (3.00 miles)

Nearest City: Wilkes-Barre; 1980 population 51,551

Site Topography: rolling

Surrounding Area Topography: hilly with flat river valley

Land Use within 8 km (5 miles): wooded and agricultural

Nearby Features:

The nearest town is Beach Haven about 1.6 km (1 mile) SW. U.S. Highway I-80 is 8 km (5 miles) S. The ConRail Railroad is 0.8 km (0.5) mile E and the Delaware and Hudson Railroad is 1.6 km (1 mile) E.

Area of Transmission Line Corridor: 730 ha (1800 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,500,000 1,510,000 1,530,000 1,550,000 1,580,000


Three Mile Island Nuclear Station


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Location:

Dauphin County, Pennsylvania 16 km (10 miles) SE of Harrisburg latitude 40.1531°N; longitude 76.7250°W

Licensee: Metropolitan Edison Co.

Unit Information Unit 1
Docket Number 50-289
Construction Permit 1968
Operating License 1974
Commercial Operation 1974
License Expiration 2014
Licensed Thermal Power [MW(t)] 2568
Design Electrical Rating [net MW(e)] 819
Type of Reactor PWR
Nuclear Steam Supply System Vendor B&W

Cooling Water System

Type: natural draft cooling towers

Source: Susquehanna River

Source Temperature Range: 1-29°C (33-85°F)

Condenser Flow Rate: 27 m3/s (430,000 gal/min)

Design Condenser Temperature Rise:

Intake Structure: concrete structure on river bank

Discharge Structure: discharged at the shoreline

Site Information

Total Area: 191 ha (472 acres)

Exclusion Distance: 0.61-km (0.38-mile) radius

Low Population Zone: 3.22 km (2.00 miles)

Nearest City: Harrisburg; 1980 population: 53,264

Site Topography: flat

Surrounding Area Topography: rolling to hilly

Land Use within 8 km (5 miles): agricultural

Nearby Features:

The nearest town is Middletown 6 km (4 miles) N. Harrisburg-York airport is 13 km (8 miles) WNW. Unit 2 ceased operation after an accident in 1979. Peach Bottom Atomic Power Station is 56 km (35 miles) downstream.

Area of Transmission Line Corridor: 725 ha (1790 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
2,170,000 2,210,000 2,240,000 2,290,000 2,350,000


Trojan Nuclear Plant


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Location:

Columbia County, Oregon 51 km (32 miles) N of Portland latitude 46.0408°N; longitude 122.8844°W

Licensee: Portland General Electric Co.

Unit Information Unit 1
Docket Number 50-344
Construction Permit 1971
Operating License 1975
Commercial Operation 1976
License Expiration 2015
Licensed Thermal Power [MW(t)] 3411
Design Electrical Rating [net MW(e)] 1130
Type of Reactor PWR
Nuclear Steam Supply System Vendor WEST

Cooling Water System

Type: natural draft cooling tower

Source: Columbia River

Source Temperature Range:

Condenser Flow Rate: 27.04 m3/s (428,600 gal/min)

Design Condenser Temperature Rise: 25°C (45°F)

Intake Structure: at river bank

Discharge Structure: submerged pipe extending 110 m (350 ft) from river bank

Site Information

Total Area: 257 ha (635 acres)

Exclusion Distance: 0.66-km (0.41-mile) minimum

Low Population Zone: 4.02-km (2.50-mile) radius

Nearest City: Portland; 1980 population: 368,148

Site Topography: flat to rolling

Surrounding Area Topography: hilly to mountainous

Land Use within 8 km (5 miles): wooded

Nearby Features:

The nearest town is Prescott 0.8 km (0.5 mile) N. The Burlington Northern Railroad is just W of the site. Gifford Pinchot National Forest and Mount St. Helens National Monument are about 48 km (30 miles) ENE.

Area of Transmission Line Corridor: 510 ha (1260 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,850,000 2,160,000 2,430,000 2,820,000 3,780,000


Turkey Point Plant


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Location:

Dade County, Florida 40 km (25 miles) S of Miami latitude 25.4350°N; longitude 80.3314°W

Licensee: Florida Power and Light Co.

Unit Information Unit 3 Unit 4
Docket Number 50-250 50-251
Construction Permit 1967 1967
Operating License 1972 1973
Commercial Operation 1972 1973
License Expiration 2012 2013
Licensed Thermal Power [MW(t)] 2200 2200
Design Electrical Rating [net MW(e)] 693 693
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: closed cycle canal

Source: Biscayne Bay

Source Temperature Range: 12-32°C (54-90°F)

Condenser Flow Rate: 39.4 m3/s (624,000 gal/min) each unit

Design Condenser Temperature Rise: 9°C (16°F)

Intake Structure: intake canal and barge canal

Discharge Structure: canal system covering about 1600 ha (4000 acres)

Site Information

Total Area: 9700 ha (24,000 acres)

Exclusion Distance: 1.27 km (0.79 mile)

Low Population Zone: 8.05 km (5.00 miles)

Nearest City: Miami; 1980 population: 346,681

Site Topography: flat

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): mostly undeveloped

Nearby Features:

The nearest town is Florida City about 14 km (9 miles) W. Hawk Missile Base is 1.6 km (1 mile) NW. Homestead recreation park is about 3 km (2 miles) NNW. The Florida East Coast Railroad is about 14 km (9 miles) NW. Units 1 and 2 are coal fired and adjacent to the site.

Area of Transmission Line Corridor: 331 ha (817 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
2,700,000 3,070,000 3,420,000 4,160,000 5,050,000


Vermont Yankee Nuclear Power Station


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Location:

Windham County, Vermont 8 km (5 miles) S of Brattleboro latitude 42.7803°N; longitude 72.5158°W

Licensee: Vermont Yankee Nuclear Power Corp.

Unit Information Unit 1
Docket Number 50-271
Construction Permit 1967
Operating License 1973
Commercial Operation 1972
License Expiration 2013
Licensed Thermal Power [MW(t)] 1593
Design Electrical Rating [net MW(e)] 540
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: once through & helper towers

Source: Connecticut River

Source Temperature Range: 0-23°C (32-74°F)

Condenser Flow Rate: 23.1 m3/s (366,000 gal/min)

Design Condenser Temperature Rise: 11°C (20°F)

Intake Structure: concrete structure at edge of river

Discharge Structure: aerating structure discharging at edge of river

Site Information

Total Area: 50.6 ha (125 acres)

Exclusion Distance: 0.27 km (0.17 mile)

Low Population Zone: 8.05 km (95.00 miles)

Nearest City: Holyoke, Massachusetts: 1980 population: 44,678

Site Topography: flat

Surrounding Area Topography: rolling to hilly

Land Use within 8 km (5 miles): mostly wooded, some agricultural and industrial

Nearby Features:

The nearest town is Vernon about 1.6 km (1 mile) W. Vernon Dam is 1 km (0.7 mile) downstream from the site. The Yankee Nuclear Power Station is about 32 km (20 miles) WSW.

Area of Transmission Line Corridor: 627 ha (1550 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,510,000 1,580,000 1,620,000 1,710,000 1,800,000


Vogtle Electric Generating Plant


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Location:

Burke County, Georgia 42 km (26 miles) SE of Augusta latitude 33.1414°N; longitude 81.7625°W

Licensee: Georgia Power Co.

Unit Information Unit 1 Unit 2
Docket Number 50-424 50-425
Construction Permit 1974 1974
Operating License 1987 1989
Commercial Operation 1987 1989
License Expiration 2027 2029
Licensed Thermal Power [MW(t)] 3411 3411
Design Electrical Rating [net MW(e)] 1101 1160
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: natural draft cooling towers

Source: Savannah River

Source Temperature Range: 4-30°C (39-86°F)

Condenser Flow Rate: 32.16 m3/s (509,600 gal/min) each unit

Design Condenser Temperature Rise: 18°C (33°F)

Intake Structure: at river bank

Discharge Structure: single-point discharge pipe near the shoreline

Site Information

Total Area: 1282 ha (3169 acres)

Exclusion Distance: 1.09-km (0.68-mile) minimum

Low Population Zone: 3.22-km (2.00-mile) radius

Nearest City: Augusta; 1980 population: 47,532

Site Topography: rolling

Surrounding Area Topography: rolling, river flood plain

Land Use within 8 km (5 miles): Department of Energy Savannah River Plant, some farming and wooded

Nearby Features:

The nearest town is Shell Bluff about 11 km (7 miles) W. The Seaboard Coast Line Railroad is about 6 km (4 miles) NE. The Department of Energy Savannah River Plant is about 16 km (10 miles) NNE.

Area of Transmission Line Corridor:

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
630,000 690,000 750,000 840,000 930,000


Washington Nuclear Project 2 (WNP-2)


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Location:

Benton County, Washington 19 km (12 miles) NW of Richland latitude 46.4714°N; longitude 119.3331°W

Licensee: Washington Public Power Supply System

Unit Information Unit 2
Docket Number 50-397
Construction Permit 1973
Operating License 1984
Commercial Operation 1984
License Expiration 2024
Licensed Thermal Power [MW(t)] 3323
Design Electrical Rating [net MW(e)] 1100
Type of Reactor BWR
Nuclear Steam Supply System Vendor GE

Cooling Water System

Type: mechanical draft cooling towers

Source: Columbia River

Source Temperature Range: 3-18°C (38-64°F)

Condenser Flow Rate: 35 m3/s (550,000 gal/min)

Design Condenser Temperature Rise: 15.9°C (28.7°F)

Intake Structure: 2 perforated pipe inlets supported offshore above the river bed 270 m (900 ft) from pump structure on river bank

Discharge Structure: buried 5-km (3-mile) pipeline terminating at the river bed 53 m (175 ft) from the shoreline

Site Information

Total Area: on Department of Energy Hanford Reservation

Exclusion Distance: 1.95-km (1.21-mile) radius

Low Population Zone: 4.83 km (3.00 miles)

Nearest City: Spokane; 1980 population: 171,300

Site Topography: flat

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): Hanford Reservation and agricultural

Nearby Features:

The nearest town is Richland 14 km (9 miles) S. The site is in the SE part of the Hanford Reservation.

Area of Transmission Line Corridor: on Hanford Reservation

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
280,000 310,000 330,000 370,000 410,000


Waterford Steam Electric Station


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Location:

St. Charles County, Louisiana 32 km (20 miles) W of New Orleans latitude 29.9947°N; longitude 90.4711°W

Licensee: Louisiana Power and Light Co.

Unit Information Unit 3
Docket Number 50-382
Construction Permit 1974
Operating License 1985
Commercial Operation 1985
License Expiration 2025
Licensed Thermal Power [MW(t)] 3390
Design Electrical Rating [net MW(e)] 1104
Type of Reactor PWR
Nuclear Steam Supply System Vendor CE

Cooling Water System

Type: once through

Source: Mississippi River

Source Temperature Range: 8-28°C (46-82°F)

Condenser Flow Rate: 61.53 m3/s (975,100 gal/min)

Design Condenser Temperature Rise: 9°C (16°F)

Intake Structure: at river bank

Discharge Structure: at river bank

Site Information

Total Area: 1441 ha (3561 acres)

Exclusion Distance: 0.92-km (90.57-mile) radius

Low Population Zone: 3.22 km (2.00 miles)

Nearest City: New Orleans; 1980 population: 557,927

Site Topography: flat

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): industrial, agricultural, recreational, and residential

Nearby Features:

The nearest town is Killona 1.6 km (1 mile) WNW. U.S. Highway I-10 is about 11 km (7 miles) NE and I-90 about 11 km (7 miles) SE. Several active and abandoned gas and oil fields are with in 16 km (10 miles). Lake Pontchartrain is about 11 km (7 miles) NE. The Missouri Pacific Railroad is just S of the site and the Southern Pacific Railroad is about 13 km (8 miles) SE.

Area of Transmission Line Corridor: 110 ha (280 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,970,000 2,130,000 2,290,000 2,520,000 2,780,000


Watts Bar Nuclear Plant


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Location:

Rhea County, Tennessee 11 km (7 miles) SSE of Spring City latitude 35.6022°N; longitude 84.7894°W

Licensee: Tennessee Valley Authority

Unit Information Unit 1 Unit 2
Docket Number 50-390 50-391
Construction Permit 1973 1973
Operating License -- --
Commercial Operation -- --
License Expiration -- --
Design Thermal Power [MW(t)] 3411 3411
Design Electrical Rating [net MW(e)] 1170 1170
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: natural draft cooling towers

Source: Chickamauga Lake

Source Temperature Range: 6-28°C (43-82°F)

Condenser Flow Rate: 26 m3/s (410,000 gal/min) each unit

Design Condenser Temperature Rise: 21°C (38°F)

Intake Structure: at lake bank

Discharge Structure: to lake via a holding pond

Site Information

Total Area: 716 ha (1770 acres)

Exclusion Distance: 1.21 km (0.75 mile) radius

Low Population Zone: 4.83 km (3.00 miles)

Nearest City: Chattanooga; 1980 population: 169,514

Site Topography: flat to rolling

Surrounding Area Topography: rolling to hilly

Land Use within 8 km (5 miles): wooded with some agricultural

Nearby Features:

The nearest town is Peakland 3 km (2 miles) NE. Watts Bar Dam is 1.6 km (1 mile) N. A fossil-fired steam plant is just N of the site. U. S. Highway I-75 is about 18 km (11 miles) SE. The New Orleans and Texas Pacific Railroad is 11 km (7 miles) NW. Chickamauga Lake is on the Tennessee River.

Area of Transmission Line Corridor: 1281 ha (3165 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
950,000 1,040,000 1,120,000 1,240,000 1,370,000


Wolf Creek Generating Station


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Location:

Coffey County, Kansas 6 km (4 miles) NE of Burlington latitude 38.2386°N; longitude 95.6894°W

Licensee: Wolf Creek Nuclear Operating Corp.

Unit Information Unit 1
Docket Number 50-482
Construction Permit 1977
Operating License 1985
Commercial Operation 1985
License Expiration 2025
Licensed Thermal Power [MW(t)] 3411
Design Electrical Rating [net MW(e)] 1170
Type of Reactor PWR
Nuclear Steam Supply System Vendor WEST

Cooling Water System

Type: closed cycle cooling lake

Source: Wolf Creek

Source Temperature Range: 0-31°C (32-87°F)

Condenser Flow Rate: 30 m3/s (500,000 gal/min)

Design Condenser Temperature Rise: 17.5°C (31.5°F)

Intake Structure: structure on shore of cooling lake

Discharge Structure: discharged to 2060-ha (5090-acre) cooling lake into an embayment separated from the intake

Site Information

Total Area: 3973 ha (9818 acres)

Exclusion Distance: 1.21-km (0.75-mile) radius

Low Population Zone: 4.02-km (2.50-mile) radius

Nearest City: Topeka; 1980 population: 118,690

Site Topography: flat to rolling

Surrounding Area Topography: flat to rolling

Land Use within 8 km (5 miles): agricultural and range land

Nearby Features:

The nearest town is Sharpe about 3 km (2 miles) N. The Flint Hills National Wildlife Refuge is about 11 km (7 miles) W. The John Redmond Reservoir is about 6 km (4 miles) W. U.S. Highway I-35 is 23 km (14 miles) N. The cooling lake is formed by a dam on Wolf Creek.

Area of Transmission Line Corridor: 1200 ha (2900 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
200,000 210,000 220,000 250,000 270,000


Yankee Nuclear Power Station


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Location:

Franklin County, Massachusetts 34 km (21 miles) NE of Pittsfield latitude 42.7281°N; longitude 72.9289°W

Licensee: Yankee Atomic Electric Co.

Unit Information Unit 1
Docket Number 50-029
Construction Permit 1957
Operating License 1960
Commercial Operation 1961
License Expiration 2000
Licensed Thermal Power [MW(t)] 600
Design Electrical Rating [net MW(e)] 175
Type of Reactor PWR
Nuclear Steam Supply System Vendor WEST

Cooling Water System

Type: once through

Source: Deerfield River

Source Temperature Range: 2-20°C (35-68°F)

Condenser Flow Rate: 8.8 m3/s (140,000 gal/min)

Design Condenser Temperature Rise: 13°C (24°F)

Intake Structure: intake from Sherman Pond about 27 m (90 ft) below normal pond level.

Discharge Structure: discharge to Sherman Pond

Site Information

Total Area: 800 ha (2000 acres)

Exclusion Distance: 0.95 km (0.59 mile)

Low Population Zone: 8.05 km (95.00 miles)

Nearest City: Pittsfield; 1980 population: 51,974

Site Topography: hilly

Surrounding Area Topography: very hilly

Land Use within 8 km (5 miles): some maple syrup production

Nearby Features:

The nearest town is Monroe Bridge 1.6 km (1 mile) WSW. Sherman Pond is adjacent to the site and discharges to the Deerfield River. A hydro station is just below the dam. Vermont Yankee Nuclear Power Station is about 32 km (20 miles) ENE. There are many ski resorts in the area.

Area of Transmission Line Corridor:

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
1,720,000 1,760,000 1,800,000 1,870,000 1,950,000


Zion Nuclear Plant


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Location:

Lake County, Illinois 10 km (6 miles) N of Waukegan latitude 42.4456°N; longitude 87.8022°W

Licensee: Commonwealth Edison Co.

Unit Information Unit 1 Unit 2
Docket Number 50-295 50-304
Construction Permit 1968 1968
Operating License 1973 1973
Commercial Operation 1973 1974
License Expiration 2013 2013
Licensed Thermal Power [MW(t)] 3250 3250
Design Electrical Rating [net MW(e)] 1040 1040
Type of Reactor PWR PWR
Nuclear Steam Supply System Vendor WEST WEST

Cooling Water System

Type: once through

Source: Lake Michigan

Source Temperature Range: 0-19°C (32-66°F)

Condenser Flow Rate: 46.4 m3/s (735,000 gal/min) each unit

Design Condenser Temperature Rise: 11°C (20°F)

Intake Structure: Intake is located 790 m (2600 ft) offshore in water 6.7 m (22 ft) deep. Intake cap is 3 m (10 ft) below normal lake surface.

Discharge Structure: Each unit has a separate discharge structure 230 m (760 ft) from shoreline.

Site Information

Total Area: 100 ha (250 acres)

Exclusion Distance: 0.40-km (0.25-mile) radius

Low Population Zone: 1.61-km (1.00-mile) radius

Nearest City: Waukegan; 1980 population: 67,653

Site Topography: flat

Surrounding Area Topography: flat

Land Use within 8 km (5 miles): residential, industrial, agricultural, and recreational

Nearby Features:

Site is bounded by the Illinois Beach State Park on the south, a city park on the north, the town of Zion on the west, and Lake Michigan on the east. A railroad runs along the western site boundary. U.S. Highway I-94 is 10 km (6 miles) W.

Area of Transmission Line Corridor: 58.7 ha (145 acres)

Population within an 80-km (50-mile) radius:

1990 2000 2010 2030 2050
7,480,000 7,720,000 7,900,000 8,200,000 8,520,000


References


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Nuclear Safety Journal, various issues 1957--1979.

NUREG-0020, Vol. 9, Licensed Operating Reactors, Summary Status Report: Data as of 8-31-85, U.S. Nuclear Regulatory Commission, Office of Resource Management, Division of Budget and Analysis, October 1985.

ORNL-NSIC-55, Vols. 1 and 2, F. A. Heddleson, Design Data and Safety Features of Commercial Nuclear Power Plants, Oak Ridge National Laboratory, Oak Ridge, Tennessee, November 1973.

WASH-1319, W. Ramsay and P. R. Reed, Land Use and Nuclear Power Plants: Case Studies of Siting Problems, U.S. Atomic Energy Commission, October


Appendix B : Definition of Impact Initiators for Nuclear Plant License Renewal Generic Environmental Impact Study


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B.1 Introduction


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Chapter 2 described the nuclear plant programs characterized for the purpose of assessing possible environmental impacts associated with license renewal. Both typical and conservative programs for both boiling- water reactors (BWRs) and pressurized- water reactors (PWRs) were described, together with the underlying assumptions and bases used in the development of these programs. Chapter 2 also presented estimates of the incremental environmental impact initiators associated with nuclear power plant license renewal.

This appendix provides additional discussion of impact initiator estimates. Additional factors and details are discussed, and comparisons are provided with license renewal-related impact initiator estimates derived from other sources. This appendix also compares the differences in impact initiators between the typical and conservative programs.

As noted in Chapter 2, license renewal for a particular plant will be based on ensuring compliance by the licensee with the current licensing basis for that plant (i.e., the original plant licensing basis as amended during the initial license term). In addition, the licensees will be required to demonstrate for certain important systems, structures, and components (SSC) that the effects of aging will be managed in the renewal period in a manner such that the important functions of these SSCs will be maintained. The SSCs of concern in the renewal period are those which traditionally do not have as readily monitorable performance or condition characteristics and include most passive, long-lived plant SSCs. Therefore, the Nuclear Regulatory Commission's (NRC's) license renewal rule requires a systematic review of, as a minimum, passive, long-lived SSCs that support safety or other critical functions of a nuclear power facility. To make these determinations regarding these SSCs, it is expected that licensees will implement aging management activities for SSCs for which current programs are not adequate to ensure continued functionality in the renewal term. These aging management activities are expected to include surveillance, on-line monitoring, inspections, testing, trending, and recordkeeping (SMITTR) as appropriate. This enhanced activity, together with updated aging assessments, is intended to ensure that aging-related degradation of important SSCs is detected and mitigated in a timely manner. The satisfactory fulfillment of NRC requirements for license renewal may necessitate repairs or modifications to the facility or its operations which are incremental to corresponding actions being performed during the term of the current license. Note that the license renewal rule does not require any specific modifications to a facility.

In addition to those actions required by 10 CFR Part 54 or other licensing requirements, licensees may undertake various refurbishment and upgrade activities at nuclear plants to better ensure economic and reliable power generation from these facilities. These activities performed for safety and/or economic reasons can result in environmental initiators which are different from those incurred in the original licensing term.

B.1.1 Purpose

The primary objective of the effort discussed here was the development of quantitative estimates of selected license renewal-related environmental impact initiators. The term "impact initiators" was defined in Chapter 2. The resulting impact initiator estimates were used in developing the Generic Environmental Impact Statement (GEIS) to support nuclear plant license renewal rulemaking. All initiators characterized in this appendix are incremental relative to those already experienced with current nuclear plant operation. The incremental environmental impact initiators expected to result from license renewal-related activities are as follows:

As noted in Chapter 2, the impact initiators cited above are those which result from nuclear plant incremental aging management activities. These are the incremental activities performed to support license renewal and extended plant operation. Also, the focus is on changes in impact initiators originating from plant activities as opposed to changes in the plant environs or receptors (e.g., changes in the population affected by the plant). The impact initiators assessed herein form a sufficient set from which to assess most license renewal-related environmental impacts.

Two types of license renewal program estimates are developed herein. The first applies to "typical" license renewal programs and is intended to be representative of the type of programs that most plants seeking license renewal might implement. The second is more encompassing and is intended to be an upper bound as to the impacts likely to be generated at any particular plant.

Both types of estimates are useful. The typical scenarios are useful for estimating impacts from an "average" license renewal program and for estimating total nuclear plant population impacts on the nation as a whole. The typical programs are intended to be representative of plants that have been reasonably well maintained and that have already undertaken most major refurbishment activities that might have been necessary. The conservative scenario estimates, on the other hand, are useful for estimating the maximum impacts likely to result from any individual plant's license renewal program.

B.1.2 Scope and Organization

This appendix presents estimates of potential environmental impact initiators that may result from nuclear plant license renewal. These quantitative estimates apply to an assumed approach to aging management for two specific reactor plant types, BWRs and PWRs. Postulated sets of license renewal activities, with separate implementation schedules, have been defined for each reactor type and for both the typical and conservative scenarios. This appendix also presents the bases and assumptions used in developing the information.

More specifically, the results include the following:

This appendix presents and describes all of these results. In addition, estimates are provided of the impact initiators attributable to satisfying the proposed revision to the license renewal rule [FR 59, no. 174, 46574 (September 9, 1994)]. Possible off-site labor costs are also quantified, as are replacement energy costs for the incremental downtime needed to perform aging management activities.

To encompass the full range of individual plant license renewal actions, additional candidate programs could have been defined and characterized. These could have been developed based on other approaches to plant aging management. For example, the programs used in this analysis are characterized by extensive refurbishment and replacement of SSCs as a means of managing aging. An alternative program might be one with reliance on more extensive SMITTR activities and less reliance on refurbishment. The approach followed in this evaluation is more conservative because it results in higher estimates of impact quantities. Alternative approaches to license renewal will likely be proposed by some nuclear utilities. However, the staff believes the programs characterized here are reasonably comprehensive and provide reasonable estimates of both typical program impacts characteristic of the reactor population as a whole, and upper bound impacts associated with what might be required by a few outlier plants seeking license renewal.

Section B.2 discusses the technical approach and bases used in the development of environmental impact initiator estimates. The specific SMITTR and major refurbishment activities included in the typical and conservative license renewal programs are reviewed in Section B.3, as are additional details of the data and information development. The results of the analysis are presented in Section B.4. That section also compares the results and estimates of license renewal-related costs developed here with similar information developed by industry.


B.2 Technical Approach and Bases


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The overall plan for support of the GEIS was to develop, by plant category, expert estimates for the various environmental impact initiators associated with nuclear plant license renewal. Plant categories were defined based on the characteristics deemed important in determining environmental impacts. The environmental impact initiators for the two basic plant categories of interest were estimated by first defining a representative set of activities to be pursued to achieve license renewal and extended plant operation. Impact initiators (labor, radiation exposure, radioactive wastes, etc.) were then identified and quantified for each activity. These activity impacts were summed to provide an estimate of overall environmental impacts associated with each plant type and each program type.

B.2.1 Technical Approach

The work undertaken to define and characterize impact initiators in support of the GEIS development was divided into three primary technical areas. These are briefly discussed below.

B.2.1.1 Definition of Information Requirements

This effort addressed two key aspects to ensure complete support for the GEIS: (1) development of candidate lists of activities with potential environmental consequences and (2) identification of environmental attributes (impact initiators) associated with those activities.

A comprehensive list of possible license renewal-related activities with potential environmental impacts was developed. Emphasis was placed on defining those activities clearly associated with license renewal (i.e., those activities which would not be included in a continuation or extrapolation of the activities that occurred during the original licensing term). The types of activities considered range from enhanced inspection programs to component replacement, and they include the list of activities originally developed for the License Renewal Rule Regulatory Analysis (NUREG-1362). The list of activities developed for that regulatory analysis was modified to reflect the proposed changes to the license renewal rule (10 CFR Part 54). In turn, the potential environmental impact initiators of each identified activity were examined and analyzed. Typical attributes included labor force requirements, low-level waste generation, capital costs, and worker radiation exposure.

B.2.1.2 Design of Database Extension and Application

Work performed in support of the 10 CFR Part 54 License Renewal Regulatory Analysis had initiated the development of a database of aging management and aging mitigation activities. To maintain control over the quality of the data and the effort required, the data were managed with a state-of-the-art relational database program on a microcomputer. This database application incorporated models of SMITTR effectiveness, permitting assessment of proposed aging management programs. The relational database facilitated the organization, archiving, and retrieval of the generic SMITTR data. The microcomputer database design was expanded to cover the more comprehensive information requirements related to assessing license renewal environmental impacts.

B.2.1.3 Review and Development of Data

Estimates of the potential incremental environmental attributes or challenges (i.e., impact initiators) created by license renewal- related activities were prepared for a generic BWR and a generic PWR. The plant features utilized were based on representative 1000-MW(e) plant designs. The plant designs were briefly discussed in Chapter 2. All attributes were quantified using actual data, industry estimates, or NRC's generic estimating methods. In addition, schedules were developed for implementing each activity of each program. Many activities carried out in support of license renewal and extended plant life are repeated at given intervals. For these types of activities, the repetition frequency and implementation schedule were also established.

The Part 54 Regulatory Analysis was reviewed for applicability and updated with more recent or more accurate information if available. New data requirements were evaluated and information sources identified. All information was reviewed in detail to ascertain its accuracy and entered into a database system. The database was then sorted into the requisite plant categories and the information provided for performance of the environmental impact assessment.

B.2.1.4 Accounting for the Effects of Other NRC Regulations

All activities were reviewed for possible overlap with actions that may be undertaken to satisfy other licensee requirements, such as those imposed by the Maintenance Rule. For the typical license renewal programs, any activity potentially required by regulations other than the license renewal rule was deleted from the programs. In certain cases, activities which met this criterion were retained to encompass what licensees might do to better ensure reliable and economical plant performance, and thus to account for enhanced or additional actions performed on non-safety-related SSCs. Whenever such activities were retained, the numbers of SSCs to which these activities applied were reduced to reflect that fraction of the time that the actions would be performed in response to Maintenance Rule or other rule requirements.

Note that this type of review was performed for the typical scenarios only. For the conservative scenarios, this type of refinement to the programs would have had a negligible effect on the overall estimates of impact initiator quantities.

B.2.2 Assumptions and Bases

B.2.2.1 Bases for Reference License Renewal Programs

Most of the assumptions and bases used in developing the license renewal program environmental impact initiator estimates were discussed in Chapter 2. Additional aspects are presented here.

The typical and conservative license renewal programs characterize actions a licensee may take to ensure both safe and economic operation of its plant beyond the current 40-year license period. In reality, each plant's program and the specific refurbishment or repairs made for extended life will depend on many factors, including the original plant design, repairs already undertaken in the original license period, operating conditions and unusual occurrences, and plant management philosophy. The set of actions actually undertaken for license renewal, therefore, are expected to vary by plant because of specific plant designs, vintages, and classes. The staff believes the range of estimates developed here reasonably bound the impacts likely to actually accrue at any individual plant site.

The typical programs are intended to be representative of the typical or "average" plant's activities in support of license renewal. However, as noted in Chapter 2, the typical programs are still somewhat conservative.

The conservative license renewal scenarios are intended to capture what might occur for those outlier plants whose impacts will be considerably greater than what is typical of the reactor population as a whole. Because these conservative programs are quite comprehensive, they encompass impacts from more typical programs. The primary bases and assumptions used were discussed in Chapter 2.

The typical programs for both BWRs and PWRs are similar, except for the differences caused by reactor design and technology. This is also the case for the conservative programs.

B.2.2.2 Aging Management Programs: Descriptions, Assumptions, and Bases

Key aspects of the license renewal environmental impact assessments were discussed in Chapter 2. Additional factors and considerations are presented in the following discussions.

B.2.2.2.1 Sources of Information

Activities assumed to occur under each plant operational or outage mode were based on information available in industry lead and pilot plant life extension studies (EPRI NP-5181SP and NP-5181M; EPRI NP-5289P; EPRI NP-5002), NRC's Nuclear Plant Aging Research program results (NUREG/CR-5284; NUREG/CR-4731), previous and ongoing NRC license renewal regulatory analysis efforts (Sciacca 1989; MITRE 1988; Sciacca January 25, 1990; Sciacca February 20, 1990), discussions of major repair activities undertaken at operating nuclear power plants as reported in technical literature (Forest 1988; Katz 1988; Miselis 1988), and discussions with industry and nuclear equipment suppliers. Discussions were also held with lead plant personnel to further ascertain the results of their life extension and license renewal evaluations (Sciacca January 3, 1993; Attachment 1). Estimates of labor and routine occupational exposure incurred in the performance of these activities were largely based on information provided in those sources. Where such estimates were not available, they were derived using the generic estimating methods developed by the NRC (NUREG/CR-4627; NUREG/CR-5236; NUREG/CR-5035; NUREG/CR-4555). The assessment of available information included an extensive literature search of actual industry data of relevant SMITTR and refurbishment/replacement activities. The information found, and the sources investigated, are discussed in Attachment 1 to this appendix. The Maintenance Rule (10 CFR 50.65) was also reviewed to assess the effects of this requirement relative to detecting and mitigating aging degradation of important SSCs.

B.2.2.2.2 Major Refurbishment Schedules

Impact initiators were initially developed for two different schedules for major refurbishment or replacement activities (Sciacca 1990). The reference schedule assumes that major refurbishment activities associated with license renewal are started shortly after the new license is granted and that these are accomplished over several successive outages. They are completed by the time the plant completes its 40th year of operation, which is about 10 years into the new license term. A second schedule was explored which was based on the assumption that all major activities of this type occur at the end of the current 40-year license period, either by preference or because the license renewal is not expected in time to schedule activities earlier during the current period. This major refurbishment outage would necessitate a longer duration than that called for by the reference schedule. Because of the complexity of accomplishing all of the major refurbishment activities called for in the example aging management programs at a single outage, this latter scenario was dropped from consideration.

The schedule for performing any major refurbishment activities will undoubtedly be highly plant-specific, and such activities could well be spread throughout the term of the renewed license. Earlier timing of these activities provides the utilities with more time to recover the cost of the investment through the sale of energy produced. Thus, the schedules utilized for the present evaluations are reasonable, but alternative schedules are also possible.

The schedules utilized were similar for both the BWR and PWR programs. However, typical programs have little need for an extended outage because the extent of major refurbishment activities is relatively modest. The "major refurbishment outage" duration for typical programs was reduced compared with that deemed necessary for the conservative case scenarios.

B.2.2.2.3 Outage Types and Durations

Chapter 2 noted that activities carried out in support of license renewal and extended plant life were assumed to be performed primarily during selected outages. Five types of outages were used; they are referred to as normal refueling, 5-year in-service inspection (ISI) refueling, 10-year ISI refueling, current term refurbishment outages, and major refurbishment outages.

Outage types and durations were established to allow estimation of the rates at which environmental impacts might be generated as a result of license renewal activities. Of greatest concern from this standpoint are the projections of the number of temporary workers needed to accomplish license renewal activities. The number of workers required at a site for a given outage depends on the amount of work to be performed (labor hours), the time available to accomplish the work, and the number of labor hours expended per person-week or person-day. The number of workers so identified, in turn, allows estimation of potential socioeconomic and other impacts to affected communities.

Certain aging management activities were assumed to be performed during full power operation. These activities will add to the plant full-time staff requirements.

In the reference BWR and PWR programs, the initial period of the renewed license was characterized by the major refurbishment outage as well as by several shorter outages referred to as current-term outages. The duration of the major refurbishment outage for the conservative case scenarios was set at 9 months for both reactor types. This duration was established based on the most limiting activity taking place during that period. For the PWRs, the most limiting activity was steam generator replacement. The limiting activity for the BWRs was the replacement of reactor recirculation piping. Recent experience indicates that both of these major activities can be accomplished in 9 months or less.

For the conservative scenarios, the 10- year ISI was given a duration of 4 months, with other 5-year ISIs lasting 3 months. Most other refuelings were assumed to be 2-month outages. Current-term outages were assumed to have a duration of 4 months each.

For the typical scenarios, the duration of the major refurbishment outage was set at 4 months. This duration was adequate to accomplish the limited number of major refurbishment activities included in these programs. For these scenarios the 10-year and 5-year ISIs, as well as current-term outages, were given a duration of 3 months each.

Assignments of outage duration were based on experience prevalent in the nuclear industry.

In reality, all outage durations will be established based on both economic considerations (e.g., cost of replacement power) and what can practically be accomplished during each outage. The short outages in which many major activities (including refueling, ISIs, major component replacement, etc.) are assumed to be undertaken simultaneously may require very large (possibly unreasonably so) labor forces. No attempt was made in this limited effort to optimize outage schedules or durations. However, preliminary work schedules were developed for the conservative scenario major refurbishment outages for both BWRs and PWRs to assess whether the major activities slated for this period could reasonably be accomplished in the allotted time. This assessment indicated it is feasible to accomplish the example refurbishment during the 9-month duration assumed.

B.2.2.2.4 Labor Categories

Labor necessary to accomplish the inspection, surveillance, testing, maintenance (ISTM), and major refurbishment/ replacement activities associated with license renewal and plant life extension (PLEX) were estimated separately for each activity. Labor was subdivided into the categories of engineering, administrative, skilled crafts, and laborers. Each labor category's hours in different radiation fields were estimated by activity. In addition, health physics-related support service labor was separately estimated for all activities performed in a radioactive environment.

B.2.2.2.5 Activity Repetitions

The number of times a given activity was performed was determined based on the intervals between the times when a given activity (such as a particular inspection or refurbishment) would be performed on a given component and the number of such components in the plant subject to those actions. Quantities of similar components were determined from reviews of representative plant piping and instrumentation diagrams, key system schematics, plant descriptions, and detailed material take-offs available for various plants. Frequencies for activities such as major refurbishment or replacements might occur only once in the plant's lifetime (e.g., BWR recirculation pipe replacement), or they might occur several times (e.g., valve refurbishment or replacement). Only incremental aging management activities, those which are in addition to those currently performed, were included here. Lead plant program information was also used in establishing activity repetitions and frequencies.

B.2.2.2.6 Radioactive Waste Generation

Volumes and types of waste generated were estimated on an activity-by-activity basis. For refurbishment, overhaul, or replacement activities, estimates of noncompactible wastes were based on the size of the items involved (i.e., the physical dimensions of the target items). Associated compactible wastes were estimated based on typical ratios of compactible-to- noncompactible volumes. In addition, compactible and noncompactible waste volumes were derived from information found in published reports of major repairs undertaken at nuclear power plants. Fluid volumes generated as a result of decontamination activities were estimated based on typical volumes generated for similar activities. All fluids used in these processes were assumed to be processed through filters or resin beds to remove contamination so that no radioactive liquids needed to be disposed of. The resulting resins or filters are disposed of as dry wastes.

All inspection, surveillance, and test activities conducted on radioactive systems or in radiation areas were also assumed to generate radioactive wastes. For such activities, compactible dry active wastes (DAW) were assumed to be generated at the rate of 0.012 m3 (0.4 ft3) per craft labor hour (as-generated volume). These result from the laundering and disposal of anti- contamination clothing and other protective equipment.

B.2.2.2.7 Waste Disposal Costs

Costs associated with the disposal of low-level radioactive wastes generated from license renewal-related activities were estimated separately for BWRs and PWRs. These estimates took into account the projected volumes of noncompactible and compactible DAW generated by each reactor type in the conduct of license renewal- related activities. The disposal costs were calculated using the NRC's generic estimating methodology (NUREG/CR- 4555), but the bases were updated to reflect the rapid escalation in burial costs arising from the formation of regional compacts and the likely closure or limited availability of the existing low-level waste disposal sites. The basis for estimating waste disposal costs is discussed more fully in Section B.3.2.4.

B.2.2.3 Approach to Estimating Impact Initiators

The estimation of impact initiators first required that the generic license renewal programs be defined in terms of the specific activities and activity repetitions making up each program. Next, the median value of each impact for each individual SMITTR or refurbishment activity was estimated for that activity taken over the full range of plants and potential circumstances. For major refurbishment activities, however, surveys were performed of pertinent, recent industry experience. Experience has shown that strong learning curve effects exist (i.e., subsequent work benefits from the experience of prior similar activities), even when the activities of interest are performed by different nuclear plant licensees. These learning curve effects suggest that, especially for major repair/refurbishment activities, it is appropriate to use information reflecting recent experience rather than median or average experience. Impact estimates for activities of this type were based on recent experience. Once values were established for each activity included in a license renewal program, the values were summed for all activities making up the program.

The particular aging management approach assumed for assessing environmental impact initiators relies more heavily on refurbishment, replacement, and monitoring than on extensive inspection, surveillance, and testing. The approach taken for these example programs tends to concentrate the impact initiators during initial refurbishment periods. For the conservative case outages, they represent an envelope that captures the activities that might be performed at essentially any U.S. nuclear power plant in support of license renewal and extended plant life. They are intended to present fairly robust scenarios in terms of environmental impacts incurred during the refurbishment outages.


B.3 Data Development


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The primary objective of this effort was to provide quantitative estimates for license renewal-related initiators which could produce incremental environmental hazards or impacts because of extended operation of nuclear power plants beyond the original 40-year term. That objective was accomplished using the following basic approach. First, candidate lists of plant SSCs susceptible to aging degradation were identified. Next, prototypic license renewal and aging management programs were defined in terms of the activities which could be carried out to manage the aging of these SSCs. These were the incremental activities carried out to support license renewal and extended plant life but not required or impacted by other NRC requirements. Each activity performed on each SSC was evaluated to estimate the potential impact initiators resulting from the conduct of that activity. Finally, total program impacts were estimated by summing the impacts from the individual activities making up a license renewal program. As noted previously, these programs of activities were defined and evaluated separately for BWRs and PWRs, each with both a typical and a conservative scenario. This section discusses the methods and bases used to establish the quantitative estimates of impact initiators.

As indicated in Section B.2.2.2.1, many different sources of information were drawn upon to establish the characteristics and content of the prototypic license renewal aging management programs and to estimate the impacts associated with each. These sources helped, in particular, to characterize the types of aging management programs that might be needed to support extended plant life during the license renewal term. The activities carried out under these programs will be needed to maintain the current licensing basis of the plants and to provide for their economical operation, as well as to satisfy the aging management requirements stipulated in the license renewal rule.

In the discussions which follow, Section B.3.1 describes the key aging management programs used to assess potential environmental impacts, and Section B.3.2 presents the specific impact initiators and describes the quantification of each initiator.

B.3.1 Aging Management Activities

The SSCs of interest for the example license renewal programs were presented in Chapter 2. The following discussions elaborate on representative aging management activities likely to be carried out on these SSCs.

The incremental aging management activities carried out to allow operation of a nuclear power plant beyond the original 40- year license term will be from one of two broad categories. These two categories of activities are (1) SMITTR actions, most of which are repeated at regular intervals, and (2) major refurbishment or replacement actions that usually occur fairly infrequently, or possibly only once, in the life of the plant for any given item.

B.3.1.1 SMITTR Aging Management Activities

Most of the SMITTR activities included in the present assessment were taken from the Safety-Centered Aging Management program defined previously and utilized for the 10 CFR Part 54 License Renewal Regulatory Analysis (NUREG-1362). However, the current effort includes additional items and activities, because the previous analysis focused only on SSCs important to safety, whereas licensees will also perform actions aimed at ensuring reliable and efficient electrical power production. Thus, many balance-of-plant SSCs are included here which were not included in the 10 CFR Part 54 evaluations.

In certain cases an SMITTR activity could involve replacement or refurbishment of the SSC being addressed. Any such SMITTR replacement/refurbishment activities for a particular item typically occurred more than once in the extended life of the plant.

Table B.1 lists the incremental SMITTR actions used as the basis for estimating license renewal environmental impacts. It indicates the specific aging detection and mitigation actions performed on each SSC of concern. The table also indicates the actions included in the typical scenarios, as well as those in the conservative case scenarios.

Table B.1 indicates the specific SMITTR activities included in each type of program, but it does not indicate the number of SSCs subject to a particular activity. The programs defined for the conservative case scenarios in all instances match or exceed the number of SSCs included in the corresponding typical license renewal programs.

B.3.1.2 Major Refurbishment Aging Management Activities

The list of major replacement and refurbishment activities included here was derived largely from areas of concern identified in the industry pilot and lead plant life extension studies, for both the conservative and typical scenarios. Those studies did not necessarily indicate that all of the items addressed should be replaced or undergo major overhauls. However, for all items addressed there was sufficient concern over their long-term integrity that investigators thought that, as a minimum, additional analysis was warranted.

Although replacement may not have been indicated for the pilot and lead plants, at least a few plants may well face extensive actions of this type to ensure safe and economical operation throughout the renewal term. Therefore, regardless of the specific determinations for the pilot and lead plants, the SSCs of concern identified in those studies form a representative list of candidate items for inclusion in major replacement and refurbishment actions for outlier plants, and thus for the conservative scenarios. Other items included in this list were drawn from actions that have already occurred at one or several operating power plants. BWR recirculation piping replacement and PWR steam generator replacement fall into this category. Although many plants will undertake the replacement of such items during the current license term, there may well be other plants which would undertake such tasks only to allow for extended plant operation. Inclusion of these activities in the conservative case scenario evaluations provides for a conservative estimate of what at least a few plants may require.

Table B.1 Incremental SMITTRa enhancement activities

SMITTR action Conservative/typical program
BWRb SMITTR Enhancements  
Bellows
Inspect one refueling and dry well bellows assembly

Both
Control Rod Drive Mechanism
Discharge and vent value tests of one mechanism

Both
Recirculation Pump and Motor
Conduct detailed inspection (disassembly/reassembly) of one pump and motor

Both
Metal Containment Including Suppression Chamber
Inspect suppression pool and vent system exterior
Renew protective coating on containment structure

Both
C
RPVc Internals  
Conduct underwater inspection of core plate for IGSCC,d jet pump brace and safe ends, shroud-to-shroud flange and access hold cover, bolt inspection method, and ultrasonic testing of top guide. Both
Conduct ultrasonic testing of top guide in central core region for IGSCC, shroud-shroud support cylinder welds, core spray inlet tee attachment, jet pump riser elbow to thermal sleeve weld region, and jet pump diffuser-to-adapter weld joint Both
PWRe SMITTR Enhancements  
Critical Concrete Structure--Containment
Renew all concrete protective coating on containment structure
Both
Reactor Coolant pump
Conduct detailed inspection (disassembly, reassembly) of PWR coolant pump, shaft, and motor
Both
RPV Internals
Inspect core support plate, core shroud, top guide using visual and ultrasonic testing or similar methods, and welds and critical areas
Both
Enhancements to Components of Both PWRs and BWRs  
AC or DC Bus
Inspect one medium-voltage breaker per manufacturer's recommendations
Both
Actuation and Instrumentation Channel
Inspect connectors and penetrations for one channel
Both
Building Crane
Perform load lift program on one crane, comprehensive SMITTR of crane or hoist
Both
Check Valve
Re-grind one valve seat; replace moving parts mechanisms
Both
Compressed Air System
Perform frequent inspection of compressed air system elements, including filter _P and leakage checks
Both
Containment
Examine fabrication welds (ultrasonic testing and visual) and base and concrete core sample (remove and replace a 6-in. square of concrete)
Both
Emergency Diesel Generator  
Inspect main bearings for wear and connecting rods for fatigue damage; also check for gear fatigue and wear Both
Conduct turbocharger drive gearing surveillance for one emergency diesel generator Both
Fan Cooler
Inspect one fan motor for break down during run (megger); perform visual check of fan running, vibration
Both
Fuel Pool
Conduct visual inspection of liner
Both
Heat Exchanger
Conduct comprehensive efficiency test on one heat exchanger
Both
Heating, Ventilation, and Air Conditioning (HVAC)  
Inspect ducting, fans and motors, flex-joints, and dampers for degradation Both
Conduct SMITTR of HVAC of one building Both
Hydraulic or Air-Operated Valve
Refurbish operator on one valve; regrind valve seat
Both
Main Condenser
Inspect wall thickness of condenser
Both
Main Generator
Inspect rotor of one main generator
Both
Main Turbine
Conduct ultrasonic test of casing for one turbine
C
Motor Operated Valve
Refurbish one valve, replacing internals
Both
Motor-Driven Pump and Motor
Conduct detailed disassembly-inspection-reassembly for one pump and motor internals
Both
Nuclear Steam Supply System Supports
Torque statistical sample of component support anchor bolts
Both
RPV  
Visually assess condition of entire vessel exterior; inspect/evaluate one specimen for fracture toughness and tensile strength Both
Inspect condition of dry lubricants in sliding foot area Both
Turbine-Driven Pump and Turbine
Conduct detailed disassembly-inspection- reassembly of one pump and turbine internals
Both

a SMITTR = Surveillance, On-Line Monitoring, Inspections, Testing, Trending, and Recordkeeping

b BWR = boiling-water reactor

c RPV = reactor pressure vessel

d IGSCC = intergranular stress-cracking corrosion

e PWR = pressurized-water reactor

Table B.2 Major refurbishment/replacement activities

Refurbishment/replacement action Conservative/typical program
Activities Common to Both BWRsa and PWRsb  
• General refurbishment and repair of turbine building, primary auxiliary building, waste processing building, fuel storage building, and feedwater pipe enclosures C
• Major overhaul and upgrade for buildings C
• Major repair/refurbishment of main generator Both
• Overhaul one crane C
• Refurbish 25 percent of liquid rad waste system C
• Refurbish coating of one condensate storage tanks C
• Refurbish main station switchgear C
• Refurbish main steam valves C
• Renew protective coating on containment structure Both
• Repair/refurbish 5 percent of reactor containment building interior concrete (or equivalent repairs) C
• Repair/refurbish turbine pedestal C
• Repair/replace major concrete imbedments in reactor containment building C
• Repair/replace portions of nuclear steam supply system major piping and component supports C
• Repair ultimate heat sink structure C
• Repair/replace 20 percent of main steam, feedwater, condensate, and circulating water system piping C
• Replace approximately half of the feedwater heaters C
• Replace closure stud bolts Both
• Replace containment electrical penetrations Both
• Replace containment sensors and instrumentation C
• Replace diesel generators C
• Replace turbine rotor Both
• Replace portions of electrical cabling both inside and outside of containment C
• Replace/repair electrical raceways and supports C
Activities Unique to BWRs  
• Replace all shroud head bolts in reactor vessel C
• Replace recirculation pump shaft and impeller, refurbish casing--of each pump C
• Replace entire BWR recirculation piping system and safe-ends C
• Replace one-half of the jet pump assemblies in the reactor vessel C
• Replace upper and lower core structure Both
Activities Unique to PWRs  
• Anneal the reactor vessel C
• Replace approximately half of reactor pressure vessel lower internal structures Both
• Replace steam generators C
• Replace pressurizer C
• Replace reactor coolant pump internals and refurbish pump C

a BWR = boiling-water reactor

b PWR = pressurized-water reactor

Table B.2 lists the major refurbishment or replacement activities used to estimate environmental impacts. Both typical and conservative case activities are indicated. The table indicates the fractions or portions of the SSCs involved which are subject to the stated actions. Unless otherwise noted, 100 percent of an SSC was assumed to be replaced or refurbished. As with the list of actions cited in Table B.1, the quantities assumed were based in part on the information provided in the industry pilot and lead plant studies (EPRI NP-5181SP; EPRI NP-5181M) and from reported existing industry experience on major refurbishment (Forest 1988; Katz 1988; Miselis 1988; North Anna-1 1993; Rippon 1990). In other cases, engineering judgment provided the basis for the portions of the systems or structures being replaced or refurbished. The actual industry experience to date with similar activities indicates that theactions listed and quantities represented in Table B.2 for the conservative case scenarios are quite conservative in that no individual plant has had to undertake the comprehensive set of actions shown. An even more conservative approach could have been taken whereby the list of activities could have been expanded and/or the portions of the SSCs involved could have been increased (e.g., replace 100 percent of feedwater heaters rather than 50 percent). However, such an approach was judged to be highly unrealistic and would have resulted in unrealistically high estimates of license renewal environmental impacts.

Table B.2 indicates that relatively few major refurbishment activities have been included in the typical license renewal programs. The activities of this type that were retained were based in part on a review of the lead plant license renewal program plans. The typical programs are based on the assumption that most plants will be maintained and operated in a manner that reduces the need for all but a few major refurbishment activities that must be undertaken sometime during the term of the renewed license. In reality, many plants will have undertaken various major refurbishment activities during the term of the current license.

B.3.1.3 Outage and Operational Modes

The bulk of the incremental activities making up the example license renewal programs must be performed when the plants are shut down. As indicated in Section B.2.2.2.3, five different types of outages were used for defining the schedule for conducting these activities. These modes are referred to as current-term outages, refurbishment outages, 5-year ISI outages, 10-year ISI outages, and normal refueling outages. In addition, certain incremental inspection and surveillance activities were assumed to be conducted during power operation. This is referred to as the full power mode. The five outage modes are characterized in the following sections.

Figure 2.3 in Chapter 2 indicated the points in a representative license renewal schedule at which these various outage modes were assumed to occur.

B.3.1.3.1 Current Term Outages

Many of the major refurbishment and replacement activities undertaken to support license renewal can be performed in stages and need not be accomplished in a single outage. This would apply, for example, to activities such as electrical cable replacement and structural upgrades. For the example programs used herein, such activities are assumed to commence shortly after the renewed license is granted by the NRC. The current analysis assumes that four current- term outages occurring within the first 10 years under the new license will be used to accomplish the bulk of the major upgrades that can be spread out in time. These outages had an assumed duration of 4 months each for the conservative case scenarios and 3 months each for the typical scenarios.

B.3.1.3.2 5-Year In-service Inspections

Two 5-year ISIs will be performed during the renewal term, corresponding to years 5 and 15 of the extended period. Certain incremental activities are assumed to be performed in addition to the 5-year ISI actions currently required of nuclear plant licensees. The incremental SMITTR activities performed during the normal refueling outages of the extended term are also carried out for the 5-year ISI outages. These outages have durations of 3 months each for all programs.

B.3.1.3.3 10-Year In-service Inspection

A single 10-year ISI is assumed to be performed midway through the extended license period. The activities assumed to occur at this outage are incremental to current 10-year ISI requirements, and also include all actions undertaken during normal outages of the extended term of plant operation. A 4-month outage duration is assumed for the conservative case scenarios, and 3 months each for the typical scenarios.

B.3.1.3.4 Refueling Outages

In addition to the 5- and 10-year ISI outages, the 20-year renewal term is assumed to be characterized by eight normal refueling outages with a duration of 2 months each for each license renewal program. Incremental SMITTR activities are performed at each of these outages.

B.3.1.3.5 Refurbishment Outage

Certain major plant upgrades, replacements, and refurbishment must realistically be accomplished during a single outage period. Replacement of steam generators in a PWR and recirculation piping in a BWR fall into this category. To accommodate major activities such as these, a single extended outage is assumed to occur at the end of the 40-year current term of operation for the conservative scenarios. For both BWRs and PWRs this conservative case outage was assumed to have a duration of 9 months. The refurbishment activities in the typical license renewal scenarios are modest compared to those in the conservative case scenarios and can be accomplished in less time. A 4-month duration was judged to be adequate for this outage for the typical scenario. Other major activities that were initiated during the current-term outages are assumed to be completed at this refurbishment outage.

A preliminary check was performed as to the reasonableness of the 9-month duration for the major refurbishment outages of both the BWR and the PWR conservative case license renewal programs. This check entailed identifying the critical path activities slated for accomplishment at this outage, assessing the time required to perform each activity, and developing an overall schedule. Figures B.1 and B.2 display the results of this evaluation. Figure B.1 shows a possible schedule for the PWR for completing the critical path activities. The comparable information for the BWR is displayed in Figure B.2. Recent industry experience, where available, was used to estimate the duration requirements for each critical path activity. These assessments also focused primarily on in-containment activities, and the assumption was made that outside- containment work was less limiting and allowed greater scheduling flexibility than the in-containment work. Both schedules allow for complete defueling of the reactor before initiating major refurbishment activities in the containment buildings. These assessments, although preliminary, suggest that the assumed 9-month duration for the conservative case major refurbishment outages is feasible.

Note that recent industry experience with major refurbishment activities such as steam generator replacement indicates that these large efforts can be accomplished in periods ranging from about 3 to 5 months, rather than the 9 months assumed for the current conservative program evaluations (North Anna-1 1993; Rippon 1990). The 9-month major refurbishment outage duration was retained to more realistically accommodate the large number of refurbishment activities assumed to proceed simultaneously during this outage.

For the typical license renewal scenarios, the most limiting activities undertaken during the major refurbishment outage were replacement of certain reactor vessel internal components and repairs to the main turbine-generator. A 4-month outage duration was judged to be sufficient to accomplish these activities.

Figure B.1 PWR major refurbishment outage schedule.

Figure B.2 BWR major refurbishment schedule.

Table B.3 Outage duration summary

Outage type Outage duration (months)
Conservative Typical
Refueling 2 2
5-year in-service inspection 3 3
10-year in-service inspection 4 3
Current-term outage (refurbishment) 4 3
Major refurbishment outage 9 4

Table B.3 summarizes the different outage types and durations for both reactor types and for both the typical and conservative license renewal scenarios.

In addition to the aging assessment and management activities performed during plant shutdown, certain incremental SMITTR activities can also be performed during full power operation. The current assessment identified only a limited number of activities of this type. This activity mode is referred to as the full power mode.

B.3.2 Quantification of Impact Initiators

Three primary types of impact initiators related to license renewal activities were quantified in this assessment: on-site labor, occupational radiation exposure, and radioactive waste generation. Other possible contributors to socioeconomic and/or environmental impacts were also assessed: capital costs, radioactive waste disposal costs, additional off-site labor requirements, and plant down time and replacement energy costs. The following sections discuss the basis for the impact quantification associated with conducting the license renewal-related activities.

B.3.2.1 Labor

This assessment developed three aspects of labor required to carry out aging management activities in support of license renewal. These aspects include labor hours, labor costs, and the number of individuals needed during a given period to perform the activities. In addition, the five labor categories of administrative personnel, engineering, craft workers, unskilled laborers, and health physics support staffing were treated.

This labor quantification effort first defined the number of craft and/or unskilled labor hours needed to perform each specific activity encompassed by any of the aging management programs. The labor estimates associated with the conduct of the SMITTR activities were taken largely from the license renewal regulatory analysis developed previously (Sciacca January 25, 1990; Sciacca February 20, 1990). The labor estimates for each activity were reviewed. Changes to the original estimates were made if new information indicated the need for revision. In many cases, the number of times a given activity was carried out was increased relative to the estimates used in the regulatory analysis. This approach was taken because the current effort encompasses actions undertaken to address both plant safety and economics, whereas the regulatory analysis dealt strictly with safety-related activities. This is particularly true for the conservative scenarios. Also, the broader scope of the current effort required the inclusion of many balance-of-plant SSCs that need to be addressed to provide economical and reliable electrical power generation over the extended term of operation.

The labor estimates required to accomplish major refurbishment or replacements were derived in two ways. If the activity of concern had already been performed in U.S. nuclear plants, and if the actual labor expenditures were reported in available documentation, this actual experience was used as a basis. Adjustments were made in certain cases to reflect learning curve effects where future repetitions of that activity could benefit from the earlier examples. However, many major refurbishment/replacement activities postulated in the current assessment have not been performed previously, or if they have, the labor required to accomplish these actions is not available. For activities in this category, NRC's generic cost estimation methods were used (NUREG/CR-4627; NUREG/CR-5236). This method uses "greenfield" or new construction labor estimates as a starting point. Labor to remove and replace given SSCs is then estimated by factoring in operating plant constraints which affect labor productivity. These constraints include factors such as access restrictions, congestion, interference with non-target systems and structures, radiation impacts, and manageability aspects. For those activities for which this method was used, the present estimates of labor requirements to accomplish major refurbishment and replacement activities took into account the specific environment under which the work would be performed. This included area-specific radiation dose rates if the work had to be performed in a radioactive environment (NUREG/CR-5035).

Distinctions between craft and unskilled workers were also developed on an activity- by-activity basis. Most of the SMITTR activities are assumed to be performed by trained technicians, who were treated as being the equivalent of skilled craftsmen. Unskilled labor was assumed to be used for some of the major replacement/refurbishment work. The ratio of craft to unskilled labor was estimated based on engineering experience. This ratio was determined separately for each activity for which some mix of both craft and unskilled labor could realistically be assumed.

Estimates of on-site administrative and engineering labor requirements for each activity were derived from the estimates of craft and unskilled labor hours. For most activities, engineering labor was assumed to be 15 percent of the craft and unskilled labor hours, and administrative efforts were taken to be 5 percent. However, engineering labor for certain activities was based on estimates presented in the industry pilot and lead plant studies on plant life extension.

Health physics (HP) support labor efforts were estimated based on the occupational radiation exposure incurred in the conduct of activities performed in a radiation environment. Previous studies (NUREG/CR-5236) indicated that typical nuclear plant expenditures for radiation protection services are in excess of $10,500 per person-rem1 of radiation exposure incurred. Of this amount, about 85 percent is labor expenditures, and the balance is attributable to materials, equipment, etc. The hourly cost of providing HP support was assumed to be $63.00 (NUREG/CR-4627). Using these figures, an estimate of 127 person-hours per person-rem of exposure was established and used in estimating HP labor hours.

All labor estimates reflect incremental on-site personnel requirements only. Additional engineering and administrative support would very likely be required for some activities, especially for major refurbishment and replacement efforts. These efforts are assumed to take place at locations remote from the reactor plant sites and would not contribute to local environmental or socioeconomic impacts. However, these off-site costs are separately accounted for to make the estimates of license renewal costs more comprehensive.

Labor costs were derived once the labor hour estimates were established. The hourly rates used for each labor category were as follows:

Administrative = $40.80
Craft = $41.30
Engineering = $45.80
Health Physics = $63.00

These hourly rates are fully burdened to reflect fringe benefits and indirect or overhead costs. They are based on electric utility wage surveys conducted by the Bureau of Labor Statistics and reflect 1994 dollars. The rates used are U.S. averages; higher or lower rates may prevail in specific geographic regions.

B.3.2.2 Occupational Radiation Exposure

Occupational radiation exposures were estimated for all activities involving radioactive systems or work in radioactive areas. Three equivalent average dose rates were assumed for the activities considered in this assessment. These rates were 0.015 rem/h for high radiation zone activities, 0.0075 rem/h for average or medium conditions, and 0.0025 rem/h for low radiation zones. These dose rate ranges were derived from a review of actual experience for both major replacement/refurbishment activities and routine surveillance and inspection activities. The rates as derived are based on the total labor hours taken to accomplish a task and the total exposure recorded for that task. As such, they take into account both the time spent in radiation zones and that spent in nonradioactive areas associated with the conduct of a particular activity. They also reflect actions taken to reduce exposures, such as application of shielding and decontamination. Exposures were determined by multiplying the total labor hours for craft and unskilled workers for a given activity by the high, medium, or low dose rates.

Estimates of occupational radiation exposure for most of the SMITTR activities included in the present assessment used the foregoing average exposure rates. Particular dose rates were assigned to a given activity based on the location and relative radiation levels of the SSC addressed by the SMITTR action.

Exposure estimates for major refurbishment/replacement actions were typically handled on an activity-by-activity basis. Estimates for activities for which data from actual experience were available used that actual experience. Steam generator replacement and recirculation piping replacement provide examples of activities for which actual radiation exposure data are available. For such cases, the total labor hours were spread among the three standard dose rates in a manner which resulted in total job exposures matching those from actual experience.

Major activities for which no actual exposure data were available employed a slightly different approach for exposure estimates. For most of these activities total labor estimates were derived using NRC's generic cost estimation methods (NUREG/CR-4627; NUREG/CR-5236). Job-specific radiation dose rates prevalent for each activity were assessed based on surveys of typical conditions for both BWRs and PWRs (NUREG/CR-5035). These estimates took into account the likelihood of decontamination or other dose-reduction measures being applied, and the raw dose rate data were adjusted accordingly. Similarly, these cases took into account the time actually spent in the radiation field.

B.3.2.3 Radioactive Waste Generation

This effort initially sought to define radioactive waste generation according to classical designations of dry wastes, Class A, B, or C; dry mixed wastes (radioactive wastes mixed with hazardous chemicals); wet wastes, Class A, B, or C; or wet mixed wastes. A review of current practices for the nuclear industry indicated that essentially none of the radioactive wastes presently shipped from nuclear plants for burial are wet wastes. Radioactive liquids are decontaminated or solidified on-site or at contractor facilities, eliminating the need for burial of the liquids. A review of the types of dry wastes likely to be generated by the activities carried out in support of license renewal and plant life extension indicated that most of these could be considered as dry Class A waste. No Class B or C wastes were identified. However, certain activities are expected to produce some greater-than- Class C (GTCC) dry wastes. This waste will result from the removal of neutron-activated materials from the reactor vessel or from the removal of materials located sufficiently close to the reactor core that activation is a problem.

The assessment of the volumes of radioactive waste to be disposed of, and the estimation of labor requirements associated with the in-plant handling of the waste, requires that DAW be defined or classified as compactible or noncompactible. Compactible DAW is amenable to significant volume reduction by compaction, incineration, or other processes. The as- shipped or as-processed volume of this waste is typically factors of 5 to 100 less than the as-generated volume. Noncompactible DAW, on the other hand, typically has an as-packaged volume which is greater than the as-generated volume because of the difficulty of achieving high packing factors with the noncompactible materials involved. The extent of volume reduction achieved is typically referred to as the volume reduction factor (VRF). This factor is defined as:

The current assessment used a VRF of 10 for compactible DAW to estimate as- shipped volumes from the as-generated values. This VRF is reasonably representative of current industry experience, and it assumes a modest amount of improvement in waste processing in the future for the industry as a whole. A VRF of 0.8 (i.e., a volume increase) was used for estimating the as-shipped volume of noncompactible DAW requiring disposal. This factor also assumes the use of state-of- the-art technology in the packaging of the noncompactible wastes.

Volumes and types of waste generated were estimated on an activity-by-activity basis. For refurbishment, overhaul, or replacement activities, estimates of noncompactible wastes were based on the size of the items involved (i.e., the physical dimensions of the target items). Associated compactible wastes were estimated based on typical ratios of compactible-to- noncompactible volumes. In addition, compactible and noncompactible waste volumes were derived from information found in published reports for major repairs undertaken at nuclear power plants.

Fluid volumes generated as a result of decontamination activities were estimated based on typical volumes generated for similar activities. All fluids used in these processes were assumed to be processed through filters or resin beds to remove contamination, with the result that no radioactive liquids needed to be disposed of. The resulting resins or filters are solidified and disposed of as dry wastes.

All inspection, surveillance, and test activities conducted on radioactive systems or in radiation areas were also assumed to generate radioactive wastes. For such activities, compactible DAW was assumed to be generated at the rate of 0.0113 m3 (0.4 ft3) per craft or unskilled worker labor hour (as-generated volume). This generation rate represents a rough average of waste production based on experience with both major replacement activities and with more routine SMITTR activities. These wastes result from the laundering and disposal of anti-contamination clothing and other protective equipment, from plastic sheeting used to restrict the spread of airborne contamination, and from the use of other such materials.

Some site labor must be expended in handling and processing the wastes generated by the activities performed in support of license renewal. In addition, some incremental radiation exposure will be incurred by those workers handling these wastes. The current assessment estimated the labor using the following rates:

Noncompactible DAW = 10.6 h/m3 (0.3 h/ft3)
Compactible DAW= 17.7 h/m3 (0.5 h/ft3)

These rates apply to the as-generated volumes of wastes. Similarly, radiation exposure incurred in the in-plant handling of radioactive wastes was estimated using a rate of 0.0012 person-rem per cubic foot of waste in the as-shipped form, and this rate applies to both compactible and noncompactible types of waste. (These rates were obtained from NUREG/CR-4627).

At least some of the waste processing activities can occur during reactor operating periods rather than being completed during the outage times when the wastes are generated. Such processing will reduce (or at least not add to) the labor burden on-site during the outages when large work forces are needed. However, the current estimates assume that the waste handling efforts occur during the same periods that the wastes are generated. This approach adds somewhat to the conservatism of the impact production rates presented here.

B.3.2.4 Waste Disposal Cost

Costs associated with the disposal of low-level radioactive wastes generated as a result of license renewal-related activities were estimated separately for BWRs and PWRs, taking into account the projected volumes of noncompactible and compactible DAW generated by each reactor type for each license renewal program. The estimates utilized the base information developed in NUREG/CR-4555, Rev. 1. However, the costs were modified to reflect the rapid escalation in burial costs resulting from the formation of regional compacts and the likely closure or limited availability of the existing low-level waste disposal sites. The analysis performed indicated that burial costs at the regional compact sites are projected to be in the range of $7,000 to $16,000/m3 ($200 to $450/ft3). The current estimates used $12,000/m3 ($340/ft3) for burial. The costs associated with handling, on-site temporary storage, and transportation of the DAW were added to the burial costs. These generic estimates were based on an assumed plant-to-burial-site distance of 1,600 km (1000 miles) and the wastes were assumed to have relatively high activity levels for the purpose of estimating costs.

Steam generators replaced as part of the PWR conservative case license renewal program are contaminated and could be disposed of as low-level radioactive waste. Their volume is quite large (approximately 1,130 m3 or 40,000 ft3), however, and the spent units are typically stored on-site rather than buried at an approved waste disposal site. Special storage buildings have been constructed at the affected reactor sites to house the spent steam generators. The cost of the storage buildings is estimated to be about $1 million and is included in the overall waste disposal cost estimates.

B.3.2.5 Capital Costs

Capital costs were estimated for those activities involving the application or installation of new equipment, materials, and hardware. Wherever available, the estimates were established based on recent industry experience for the addition or replacement of the items of concern. Where such cost information was not available, two other approaches were used. The first relied on NRC's generic cost estimation methods and databases (NUREG/CR-4627). This methodology draws on the Energy Economic Data Base (EEDB) developed by the U.S. Department of Energy (DOE/NE- 0051/1; DOE/CR-5764). The EEDB provides estimates of both labor and material/equipment quantities and cost. This information has been developed for modern, large PWR and BWR plant designs. The EEDB presents reasonably detailed information which covers most areas of the plant, including both the nuclear steam supply system (NSSS) and the balance of plant. However, this cost base does not include any detail of the NSSS equipment or hardware capital costs. The second alternative approach to estimating capital cost where no recent industry experience was available was based on the use of detailed, actual construction cost breakdowns from a U.S. nuclear plant constructed several years ago. This cost base provided sufficient breakdown of the entire plant, including detail on the NSSS component and subcomponent cost. Where this base was used, the costs reported were escalated to 1994 dollars, and, where appropriate, the costs were adjusted to reflect size differences between this base plant and the 1000-MW(e) reference size adopted for the current estimates.

B.3.2.6 Other Costs

Two other cost elements were considered to define license renewal-related costs in a more comprehensive manner: home office costs and replacement energy costs. The home office costs account for off- site engineering and quality assurance (QA) expenditures. This category of costs accounts for the design, analysis, safety review, and documentation efforts typically associated with modifications at nuclear power plants. Home office costs also allow for QA functions and activities carried on to support these modifications. Home office engineering and QA efforts were estimated using NRC's generic cost estimation methodology (NUREG/CR-4921). Based on surveys of a wide range of actual physical modifications made to operating nuclear power plants, this methodology has established that, on average, the engineering and QA functions typically amount to about 25 percent of the direct modification costs. This basis accounts for both on-site and off- site engineering and QA functions. The direct costs include direct (unburdened) labor as well as the cost of materials, equipment, and hardware associated with a particular modification. Because the on-site efforts were separately accounted for as described in Section B.3.2.1, estimates of the off-site work were developed using the 25 percent of direct costs approach and subtracting from this the estimate of on-site engineering costs.

Replacement energy costs can be a major contributor to overall project cost if plants remain out of service for extended periods. An assessment was made of replacement energy costs as they relate to the example license renewal programs. This evaluation reviewed the replacement energy costs per day of plant downtime (NUREG/CR-4012 1992) separately for BWRs and PWRs. Weighted averages were taken for several plants whose electrical generating capacity was nearest to 1000 MW(e). For PWRs, the daily replacement energy cost estimated on this basis was $342,000 (1994 dollars). For BWRs, this figure was estimated to be $287,000 (1994 dollars) per day. Replacement energy cost depends on several factors, including plant size, location and load pool, season, and cost fluctuations in non-nuclear alternative energy sources. However, the estimates cited here are representative of U.S. plants in the 1000-MW(e) size range.


B.4 Results


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This section summarizes the quantitative results developed in this evaluation. All key impact initiators are discussed, including labor, occupational radiation exposures, capital costs, radioactive waste generation, and waste disposal costs associated with the conduct of activities carried out in support of license renewal and extended plant life. This section also discusses elements of license renewal which do not necessarily contribute to environmental or socioeconomic impacts but which play important roles in assessing the overall economic viability of license renewal. These are the elements of off-site costs and replacement energy costs. Finally, this section provides a comparison of industry- developed license renewal cost estimates with those developed in this assessment. Both typical and conservative case scenario results are discussed.

B.4.1 BWR and PWR License Renewal Program Impact Initiators

As noted previously, the typical license renewal program scenarios presented herein are intended to be representative of those the majority of nuclear plants seeking license renewal might experience regarding major refurbishment and enhanced SMITTR activities needed to satisfy NRC requirements and better ensure reliable and economical plant life extension. The conservative case scenarios, on the other hand, are intended to reflect what might occur at a few outlier plants requiring much more extensive refurbishment/replacement activities than are typical of the reactor population as a whole. As such, the typical programs are estimated to have rather modest environmental impacts compared with those expected for the conservative case scenarios.

Tables B.4 and B.5 present summaries of the license renewal program impact initiator quantities for the typical and conservative case license renewal scenarios, respectively. Each table shows the quantities separately for BWRs and PWRs. Similarly, each table shows the impact quantities generated during the different plant modes. Note that the impact quantities are presented on a per-occurrence basis for refueling outages, current-term outages, and the 5- year ISI outage, each of which occurs more than once. The totals, however, reflect the summation over all occurrences of all activities performed in support of license renewal. Tables B.4 and B.5 also show the labor and costs associated with incremental activities performed during full-power plant operation. The labor hours and costs for this category represent the totals accrued over the entire period of the renewed licenses. All values shown are intended to capture only incremental effects associated with license renewal, and they exclude baseline activities which represent a continuation or evolution of current practices related to the operation and maintenance of nuclear power plants.

The types and extent of activities included in the conservative case programs, especially the extensive major replacement/refurbishment activities included and their resulting impact initiator estimates as reflected in Table B.5, are thought to reasonably bound what might be needed for any individual nuclear plant site in pursuit of license renewal and plant life extension.

The values in Tables B.4 and B.5 indicate that most of the environmental impact initiators accrue during the major refurbishment outages. For the conservative scenarios, the current-term outages also result in considerably higher levels of impact quantities being generated compared with the more routine outages. The current-term and major refurbishment outages are the periods when major replacement and refurbishment activities performed in support of license renewal and extended plant life are assumed to occur. For the conservative case scenarios, the impacts produced are primarily from activities performed to ensure that current safety and licensing bases are maintained, as well as to help ensure that plant economic and availability/reliability goals are met. Relatively few of the conservative case impacts are attributable to the enhanced aging management of SSCs important to license renewal called for by the License Renewal Rule. The rule requirements have a relatively greater impact on the typical programs, because these programs have fewer major refurbishments compared with the conservative case scenarios. The specific effects of the 10 CFR 54 rulemaking on the impact quantities are discussed in Section B.4.3.

A comparison of the figures in Tables B.4 and B.5 shows that the typical license renewal program impact initiators are on the order of 15 to 25 percent of the quantities estimated for the conservative case scenarios. Figure B.3 graphically illustrates the overall fraction of the total impacts for the typical programs relative to the conservative case scenario totals. The values shown represent a linear, composite average of the various impact category totals listed in Table B.4 relative to the totals presented in Table B.5. Thus, the conservative case scenarios are estimated to have five to six times the impact quantities of typical license renewal programs. This result is to be expected because of the extensive major refurbishment activities assumed to be undertaken by a few outlier plants represented by the conservative case scenarios.

Figure B.3 Typical program impacts relative to corresponding conservative case impacts.

Note that the additional on-site personnel figures cited in Tables B.4 and B.5 represent the average incremental work force sizes needed to accomplish license renewal-related activities assuming this work is uniformly spread over the entire duration of each separate outage. Peak work force sizes for each outage will be higher, as discussed in Section B.4.1.1.

B.4.1.1 Labor Hours and On-Site Staffing

The estimates of incremental labor shown in Tables B.4 and B.5 indicate that roughly 0.5 to 0.8 million labor hours could be expended for typical license renewal activities for both BWRs and PWRs, whereas the corresponding labor hour estimates for the conservative case scenarios are on the order of 5 to 7 million. These estimates include administrative, engineering, health physics, craft, and nonskilled labor. For the conservative case scenarios of Table B.5, about 95 percent of the labor hours for both BWRs and PWRs are attributable to the major activities that occur during the current-term outages and the major refurbishment outage. Thus, for the conservative case scenarios, these major activities tend to dominate the impact quantities compared with the more routine activities occurring at normal refueling and at the 5- and 10-year ISI outages. The labor values shown are greater for the conservative case PWR than for the corresponding BWR primarily because of the large amount of labor associated with the replacement of all four steam generators assumed in the reference PWR plant design. Table B.4 indicates that the typical case BWR labor hours are about 30 percent greater than the corresponding PWR estimates. The differences here are primarily the result of a few additional SMITTR activities being performed for the BWR over the remaining life of the plant, as well as a greater number of components that are subject to these activities.

The labor hour estimates for the different license renewal scenarios are illustrated in Figure B.4.

The additional on-site personnel estimates reflect both the labor estimates and the assumed outage durations. The conservative case license renewal programs assumed that the major refurbishment outage would be 9 months long. As discussed in Section B.3.1.3, this duration appears to be reasonable. The conservative case major refurbishment outage would require about 870 additional on-site staff for the BWR and about 1700 incremental on-site personnel for the PWR to accomplish the example program aging management activities in the allotted time. As previously noted, the larger work force required for the PWR primarily results from the large effort associated with steam generator replacement. These estimates address personnel needed over and above those who would be on-site to perform normal refueling and maintenance tasks. Most of the other outages require roughly the same number of incremental on- site personnel for both reactor types. Note that for each type of outage, the staffing indicated is in all cases incremental to the staff needed to carry out current practices. Also, these estimates reflect average incremental staff assuming the work is spread uniformly over the entire outage duration.

Figure B.4 Incremental labor hours.

For the typical scenarios, the incremental on-site staffing requirements are relatively modest. The largest staff increment is required for the major refurbishment outage, as this is the time when the few major refurbishment or replacement activities in these programs are assumed to be carried out.

The number of on-site personnel estimated in Tables B.4 and B.5 is not strictly proportional to the outage duration and the total labor hours expended during a given outage. This is because a 50-hour work week was assumed for craft, health physics, and nonskilled workers, whereas the engineering and administrative personnel were assumed to have a 40-hour work week. Also, the ratio of engineering and administrative hours to craft, health physics, and nonskilled worker hours varied from activity to activity.

Figure B.5 graphically indicates the highest average number of temporary workers needed to carry out license renewal- related activities for each of the four license renewal scenarios. This figure shows the largest requirement for each scenario as identified in Tables B.4 and B.5. Note that all estimates of incremental on-site personnel displayed in Tables B.4 and B.5, and in Figure B.5, were arrived at assuming level staffing for the entire duration of a given outage.

The extent of certain socioeconomic impacts such as housing will depend on peak numbers of personnel on-site rather than on the average numbers employed over a given outage. Therefore, additional analyses were performed to define probable staffing profiles throughout the major outages. Because the outages of interest would also include defueling/refueling and work typically conducted during present-day outages (e.g., ISIs, routine maintenance), the temporary workers needed to accomplish these routine activities must also be considered in estimating peak work force sizes. Table 2.4 of Chapter 2 noted that, based on a recent industry survey, there are typically 750 to 800 additional workers on- site at a nuclear plant during routine planned outages. The assumption was made that these workers are needed for a period of 2 months per outage. Therefore, these more routine efforts performed by temporary workers add up to a total of about 1600 person-months of effort. This effort needed to accomplish more routine outage activities was added to the license renewal-related labor efforts identified in Tables B.4 and B.5 to arrive at estimates of peak work force sizes.

Figure B.5 Outage average incremental on-site staff.

Figure B.6 Additional personnel required to perform conservative case pressurized-water reactor license renewal major refurbishment outage activities.

Figure B.7 Additional personnel required to perform conservative case boiling-water reactor license renewal current-term outage activities.

Figure B.8 Additional personnel required to perform typical case pressurized- water reactor license renewal current-term outage activities.

Figure B.9 Additional personnel required to perform typical case boiling- water reactor license renewal major refurbishment activities.

Figures B.6 through B.9 present monthly projections of temporary worker staffing needed to carry out both license renewal activities and routine refueling, and ISTM activities. The most limiting cases are shown for each license renewal scenario. Figure B.6 shows the projected number of temporary personnel needed during the major refurbishment outage for the conservative PWR license renewal scenario. This outage was assumed to require 9 months. The monthly staffing needs were arrived at by developing a schedule for carrying out each of the different activities to be accomplished during this outage. These schedules were similar to those presented in Figures B.1 and B.2, but they were more complete in that all activities slated for the outage of interest were included. An effort was made to average out the work force over the entire outage duration to the extent possible. However, considerable peaking does occur because not all activities can proceed simultaneously. Figure B.6 separately identifies temporary personnel needed to accomplish license renewal activities versus those needed for more routine outage activities. The upper figures on the bars in Figure B.6 represent the total number of temporary workers needed during each particular month of the outage; the lower figure, where present, is the number needed to accomplish the incremental license renewal-related activities only. Based on this projection, the temporary work force needed during the major refurbishment outage for the conservative PWR license renewal scenario is estimated to be about 2300. This contrasts with the 1700 additional workers averaged over the entire outage as identified in Table B.5, which excluded consideration of the work force needed to carry out refueling and other routine activities. The month-to- month temporary staffing needs presented in Figure B.6 are by no means optimized, but they do indicate that the peak numbers of workers considerably exceeds estimates based on averages over the entire outage duration.

Figure B.7 presents estimates of peak temporary worker staffing needs for the BWR conservative license renewal scenario. In this case, the highest staffing needs are projected to occur during the current-term outages rather than during the major refurbishment outage. This is the most limiting BWR outage, because although the number of temporary workers on-site needed to accomplish incremental license renewal-related activities was about equal for both the current-term outage and the major refurbishment outage (see Table B.5), the 1600 person-months of effort needed for refueling and routine outage tasks must be accomplished in a 4-month period rather than a 9-month period, giving a greater overall total for the current-term outages. The projections in Figure B.7 indicate that the peak temporary work force needed for this BWR license renewal scenario is 1440 personnel.

Temporary worker needs for the typical license renewal scenarios are shown in Figures B.8 and B.9 for PWRs and BWRs, respectively. The peak staffing needs for the PWR occur during the 3-month current-term outage. In this scenario, the number of temporary workers needed during this outage to accomplish incremental license renewal-related activities is very modest, and the majority of the temporary staff would be needed to carry out refueling and more routine outage activities. The peak staffing needs are only about one-third of those needed for the conservative PWR license renewal scenario. Figure B.9 for the typical BWR license renewal scenario indicates that slightly more than 1000 temporary workers would be needed during the peak period of the major refurbishment outage. This is the most limiting outage staffing need for the typical BWR case.

B.4.1.2 Radioactive Waste Volumes

The waste volumes shown in Tables B.4 and B.5 include all types of low-level radioactive waste generated as a result of incremental license renewal and plant life extension activities. The volumes are those which remain after the wastes have been processed for storage or burial, and they include the volume of the burial or storage containers. The compactible waste items are assumed to undergo volume reduction. An average VRF of about 10 was used. This VRF is consistent with the use of currently available supercompactor technology. Even higher VRFs are achievable with incineration techniques, but these were not assumed here to preserve the conservative nature of the overall estimates. The noncompactible items of waste, on the other hand, are assumed to require a burial or storage volume which is about 20 percent greater than the initial volume of the solid article resulting from less-than-perfect packing factors associated with this type of waste.

Table B.4 indicates that the typical case BWR scenario is estimated to produce about 226 m3 (8000 ft3) of low-level radioactive waste as a result of license renewal-related activities. The corresponding volume for the PWR is about 170 m3 (6000 ft3). The greater volume for the BWR is because of the slightly greater number of SMITTR activities and the greater number of SSCs subject to these activities compared with the PWR. In addition, activities on turbine plant equipment for the BWR generate radioactive waste, whereas similar activities for the PWR do not.

The considerably larger volume of waste noted in Table B.5 for the PWR conservative case compared to the BWR conservative case is almost solely due to the effects of steam generator replacement in the PWR. These very large items contribute about 1,130 m3 (40,000 ft3) to the total PWR waste volume, and there are no comparable items in the BWR. The steam generators that have been removed to date from operating reactors have typically been stored on-site in special facilities constructed for that purpose rather than being disposed of at licensed burial facilities.

Total waste generation quantities are illustrated in Figure B.10 for both the typical and conservative case scenarios for each plant type. The example license renewal programs generated small amounts of GTCC wastes. These wastes are neutron-activated materials removed from the reactor vessels and/or reactor internals. The estimated amounts for the typical scenarios are 28 m3 (1000 ft3) for BWRs and 14 m3 (500 ft3) for PWRs, and for the conservative case scenarios about 44 m3 (1540 ft3) for BWRs and 14 m3 (500 ft3) for PWRs. These GTCC wastes were not included in the volumes cited in Tables B.4 or B.5. They are assumed to be retained on-site rather than shipped off-site for burial.

B.4.1.3 Occupational Radiation Exposure

Figure B.11 displays the estimates of incremental occupational radiation exposure incurred in carrying out license renewal activities. As shown in Figure B.11 and as indicated in Tables B.4 and B.5, incremental radiation exposure is projected to be on the order of 250 to 450 person-rem for the typical case scenarios and about 2500 person-rem for both reactor types for the conservative case scenarios. Because current average annual exposures for U. S. nuclear power plants are about 500 person-rem, the license renewal-related incremental occupational exposure for the typical scenarios represents the equivalent of about one additional year of operation. Given a 20-year incremental operating period, the license renewal-related activities appear to add about 5 percent to the cumulative exposure that would otherwise be expected over that period of extended plant life. For the conservative case scenarios of Table B.5, the estimated incremental exposure of roughly 2500 person-rem represents about five times the currently experienced annual exposure. However, the estimates from the conservative case license renewal programs are highly conservative because they encompass a greater variety and extent of activities than is expected from most plants pursuing license renewal. The largest fraction of the radiation exposure is expected to accrue from the major refurbishment activities for both reactor types. The bulk of the exposure is estimated to occur during the major refurbishment outage, and to a lesser extent during the current-term outages. The BWRs are expected to incur somewhat greater occupational radiation exposures than are the PWRs.

Figure B.10 Incremental low-level waste generated.

Figure B.11 Incremental occupational radiation exposure.

B.4.1.4 Waste Disposal Costs

The costs for disposing of low-level radioactive wastes generated as a result of license renewal-related activities are estimated to be about $3 million for the typical case scenarios and about $26 million to $37 million for the conservative case scenarios. Relative to the total costs associated with license renewal activities, these costs represent about 3 to 4 percent of the totals. As noted in Section B.3.2.4, waste disposal costs include charges for waste handling and packaging, short-term on-site storage, transportation, and burial. For the PWR conservative case scenario, the spent steam generators are assumed to be stored on-site rather than sent to an approved burial site for permanent disposal.

A cost of roughly $1 million has been assumed for the on-site steam generator storage facility, and this cost has been added to the overall waste disposal cost for the conservative case PWR. Waste disposal costs are graphically displayed in Figure B.12.

B.4.1.5 Capital Costs and On-Site Labor Costs

In addition to waste disposal costs, Tables B.4 and B.5 display labor costs and capital costs (costs associated with the purchase of materials, equipment, and hardware). The labor costs include those attributable to all categories of on-site labor, including administrative, engineering, craft, unskilled, and health physics. The costs are based on wage rates appropriate to each labor category and the labor mix as discussed in previous sections.

The values in Table B.4 indicate that, for the typical cases, capital costs are roughly twice the labor costs. Total on-site (labor plus capital) costs are estimated to be about $90 million for the typical BWR and about $80 million for the typical PWR. The higher costs for the BWR are consistent with the greater number of incremental SMITTR activities and greater number of SSCs included in the BWR program.

For the conservative case results displayed in Table B.5, the trends of capital versus labor costs are essentially reversed. That is, labor costs are higher than the capital costs for both reactor types. This relatively higher labor cost is attributable to the fact that many of the major refurbishment/replacement activities of the conservative cases involve radioactive SSCs and work in radiation zones. Work in radiation zones is less productive than work in nonradiation zones, and relatively more labor hours must be expended to accomplish a given activity. Capital costs, on the other hand, are essentially independent of whether the equipment or materials involved are in radiation zones. The combined labor and capital costs for the conservative case are estimated to be about $400 million for the BWR and about $460 million for the PWR. The higher costs for the PWR are primarily due to the large labor cost associated with steam generator replacement.

Labor and capital costs for both the typical and conservative case license renewal scenarios are illustrated in Figure B.13.

B.4.1.6 Off-Site Labor Costs

Off-site engineering and QA work are estimated to cost about $13 million-$15 million for the typical cases and about $38 million-$42 million for the conservative case scenarios. These are the off-site costs in this category carried out in support of the SMITTR and refurbishment activities. Off- site labor costs are depicted in Figure B.13.

Figure B.12 Incremental waste disposal costs.

Figure B.13 Incremental capital and labor costs.

B.4.1.7 Total Costs

All costs shown in Tables B.4 and B.5 are in 1994 dollars. They also are presented as "overnight" (undiscounted) costs. That is, they represent costs as if all activities of each program were performed in a very short period of time rather than being spread over approximately 28 years as is envisioned for the actual scenarios. Time-value-of-money effects are not included in Tables B.4 or B.5, and no allowance has been included for costs of financing during the construction/refurbishment stages. Also, replacement energy costs are excluded from the figures presented in these tables. Those costs are discussed below.

Table B.4 indicates that the total program costs for the typical BWR are estimated to be about $110 million, and the corresponding PWR costs are about $90 million. Based on a 1000-MW(e) reference plant size, these estimates indicate license renewal-related costs of roughly $100/KW(e) for the typical renewal cases. Table B.5 indicates that the conservative case program costs are estimated to be in the range of $440 million to $500 million, with corresponding unit costs between $440/KW(e) and $500/KW(e).

B.4.1.8 Replacement Energy Costs

Replacement energy costs were estimated based on a rate of $290,000 per day of plant downtime for BWRs and $340,000 per day for PWRs (NUREG/CR- 4012 1992). The typical BWR and PWR license renewal programs have a cumulative incremental downtime of 5 months, whereas for the conservative case scenarios the incremental downtime is estimated to be about 15 months. This is the time required to accomplish the SMITTR and major refurbishment activities making up the programs. Cumulative downtime costs, therefore, are estimated to be about $44 million for BWRs and $52 million for PWRs for the typical scenarios and about $130 million to $155 million per plant for BWRs and PWRs, respectively, for the conservative case scenarios (overnight costs).

Figure B.14 illustrates the total estimated license renewal-related costs previously discussed, including replacement energy costs. This figure indicates the relative magnitude of the major cost components.

B.4.1.9 Local Purchases

Of the capital costs reported in Tables B.4 and B.5, a small fraction may possibly be spent locally. Items such as concrete, rebar, formwork, certain electrical wire and cables, and similar materials could conceivably be purchased from local suppliers in the vicinity of nuclear plants. The cost of these items used here for the typical programs was estimated to be less than $1 million for each plant type, and possibly about $5 million total for each conservative case scenario. These purchases occur for activities performed during the current-term outages and the major refurbishment outage.

B.4.2 Comparisons to Industry Costs for Plant Life Extension

The nuclear power industry, the U. S. Department of Energy, and other entities have evaluated the benefits and costs of nuclear plant life extension. They have estimated both the likely costs associated with plant life extension and the break-even costs. The break-even costs indicate the point at which nuclear plant life extension is as costly as would be the construction of alternative power sources such as a new coal plant or a new nuclear plant.

Figure B.14 Total license renewal costs.

Table B.6 compares the costs of license renewal and extended plant life developed for the GEIS with estimates prepared by industry. The table includes both typical and conservative case estimates. The GEIS estimates as presented in the table are all given on an overnight basis (i.e., both financing costs and present-worth effects have been excluded from the figures). The GEIS estimates include all cost elements presented in Tables B.4 and B.5, and they include replacement energy costs as well. These are shown separately for BWRs and PWRs. Table B.6 indicates that the GEIS estimates for plant lift extension costs range from about $150 million to $570 million for the BWR typical and conservative case license renewal programs, respectively, and about $140 million to $650 million for the corresponding PWR programs. On a dollar per kilowatt basis, and based on the reference 1000-MW(e) plant size, these costs are in the range of $140 to $150/kw for the typical scenarios and from about $570 to $650/kw for the conservative case scenarios. The available industry estimates for plant life extension are shown in Table B.6 in dollars per kilowatt (SAND88- 7095 1988; McCutchan 1988). They range from about $230/kw to almost $700/kw. The GEIS estimates fall roughly within the range developed by the industry sources. The typical case scenarios fall somewhat below the industry estimates, whereas the conservative case estimates are at the higher end of the industry estimates.

B.4.3 Other Impacts and Considerations

This section briefly discusses two aspects related to license renewal program costs. These are the time-value-of-money effects (present worth) and the portions of the programs directly attributable to meeting the new aging management requirements imposed by 10 CFR Part 54, Rules for Nuclear Plant License Renewal.

B.4.3.1 Present Worth Considerations

The estimates presented in Tables B.4, B.5, and B.6 were given on an "overnight" cost basis, and did not account for the fact that the expenditures are spread out in time over a considerable period. Table B.7 shows the effects of considering the time-value-of- money on the total program costs. Present value program costs are shown for three discount rates: 0 percent, 5 percent, and 10 percent. All costs are given in 1994 dollars. The datum time used to develop the values in Table B.7 is a representative point in a program at which a licensee would submit the application for license renewal to the NRC. As was shown in Figure 2.3 of Chapter 2, this point is assumed to occur 12 years before the expiration of the initial 40-year license period, and is 32 years before the end of plant life, assuming a total plant life of 60 years. The figures shown for the 0 percent discount rate are the same as those presented in Tables B.4, B.5, and B.6, and they include off-site labor costs and replacement energy costs.

The example license renewal programs incurred the major portion of the costs in the first 12 years following the submittal of a license renewal application. This is the period when major refurbishment/replacement activities are assumed to take place. In spite of the fact that these expenditures are not assumed to occur further out in time relative to the datum time point used, discounting does significantly reduce the effective cost of the license renewal programs.

Table B.6 Comparisons of industry plant life extension cost estimates (all costs in millions of 1994 dollars)
 

PWRa

BWRb

 

Conservative

Typical

Conservative

Typical
GEIS estimates (million dollars)        
On-site labor cost 269 21 202 28
Capital costs 155 54 171 63
Total on-site costs 461 78 399 92
Off-site labor 38 13 42 15
Incremental replacement energy costsc 155 52 132 44
Total estimated costs 654 143 573 151
$/kw 654 143 573 151
         
Industry estimates ($/kw)d        
Monticello       634
General Electric       230
Surry   1331    
Westinghouse   698    
a PWR = pressurized-water reactor
b BWR = boiling-water reactor
c @$ 287,000per day (BWR) or $342,000 per day (PWR)
d SAND88-7095, escalated to 1994 dollars

Table B.7 Time-value-of-money effects on nuclear plant license renewal program costs
 

Program costs (million dollars)

 

PWRa

BWRb

Discount rate
(%)
Conservative Typical Conservative Typical
0 694 183 605 186
5 436 107 381 107
10 291 68 256 67

a PWR = pressurized-water reactor
b BWR = boiling-water reactor

B.4.3.2 10 CFR Part 54 Impacts

Certain of the aging management activities making up the example programs used here are incremental requirements called for by 10 CFR Part 54. The example list of activities assumed attributable to the Part 54 rule are identified in Table B.8, and they represent a subset of those presented in Tables B.1 and B.2. The list of activities in Table B.8 was derived from the 10 CFR Part 54 Regulatory Analysis (NUREG/CR-1362) and from evaluations of the actions contemplated for the lead plant programs (Sciacca January 3, 1993; January 13, 1993). Other interpretations of these sources could yield different results. However, the example lists of activities presented in Table B.8 are thought to be reasonably representative of what might be needed to satisfy the requirements of the license renewal rule. This list is adequate for estimating impacts attributable to 10 CFR Part 54.

Many of the activity descriptions in Table B.8 are presented on a per-SSC basis, but the total program includes many repetitions of each activity to cover multiple similar SSCs as well as repeat actions on the same SSC. The activities listed are all SMITTR actions, as opposed to major refurbishment activities. Also, these actions address only those SSCs which are important to license renewal, and the Part 54 impacts exclude actions likely to be taken by licensees solely for economic and plant availability purposes.

Table B.9 presents estimates of the impact initiators attributable to the enhanced aging management activities called for by 10 CFR Part 54. The figures indicate that the impacts attributable to the Part 54 rule are only a small fraction of the conservative values shown in Tables B.4 and B.5. The costs shown in Table B.9 represent overnight costs, and do not reflect any discounting for expenditures incurred over the 30 or more years assumed for the conduct of the subject activities.

The capital costs reflect the addition of new or enhanced monitoring and surveillance systems and equipment, as well as the costs of replacement hardware for SSCs on a routine basis. An example of the latter is valve internal components.

Table B. 8 and B. 9

B.4.4 Consideration of Other License Renewal Programs

The current effort focused on license renewal environmental impact initiators for two generic light-water reactor types: BWRs and PWRs. The resulting estimates are believed to encompass a high percentage of the potential environmental impact initiators which may accrue as plants undertake license renewal and plant life extension activities. The estimates of environmental impacts associated with nuclear plant license renewal can be refined in a number of ways. Because no commercial light-water reactors have yet applied for license renewal, the nature and characteristics of actual programs have yet to be fully defined. As noted previously, each plant's program is expected to be somewhat unique.

Alternative programs for license renewal and plant life extension are certainly likely, and the impacts of these alternative programs could be evaluated. Differences from the reference programs used to develop the current estimates are likely for both major refurbishment activities and SMITTR activities. The SMITTR programs used here are based on the safety-centered approach developed for the License Renewal Regulatory Analysis (Sciacca January 25, 1990; Sciacca February 20, 1990). Other ISTM programs are certainly possible, some of which may have greater impacts than those defined herein. Similarly, different major refurbishment programs can also produce significantly different environmental impacts from the representative programs used here. For example, a program could be based on a much more extensive use of ISTM, with less reliance on major refurbishment, than the programs used in the current assessments.

Although alternative programs could have been evaluated, the license renewal programs used to develop the current estimates of environmental impact initiators are believed to bound what might actually occur at most plants.


B.5 Endnotes


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A discussion of the SI units used in measuring radioactivity and radiation dose is given in Appendix E, Section E.A.3.


B.6 References


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The following references include those for Attachment 1.

Abbott, S. L., et al., "Westinghouse Reactor Vessel Life Extension Variance Study," p. 210 in Proceedings of the Topical Meeting on Nuclear Power Plant Life Extension, Vol. 1, American Nuclear Society, Snowbird, Utah, July 31-August 3, 1988.

DOE/CR-5764, Phase VIII Update (1986) BWR Supplement for the Energy Economic Data Base Program, United Engineers and Constructors, August 1986.

DOE/NE-0051/1, Phase VIII Update (1986) Report for the Energy Economic Data Base Program, United Engineers and Constructors, August 1986.

Eckert, G. (Kraftwerk Union Aktiengesellschaft, Offenbach, Federal Republic of Germany), "Replacement of Steam Generators and Recirculation Pipes in LWRs as Long-Lasting Measures to Extend Plant Lifetime," pp. 237-50 in Proceedings of an NEA Symposium Held in Cooperation with the IAEA Nuclear Plant Life Extension, February 1987.

EPRI NP-2418, Extended Life Operation of Light Water Reactors: Economic and Technological Review, Vols. 1-2, Electric Power Research Institute, Palo Alto, California, June 1982.

EPRI NP-4208, The Longevity of Nuclear Power Systems, Electric Power Research Institute, Palo Alto, California, August 1985.

EPRI NP-5002, Virginia Power Company et al., LWR Plant Life Extension, Electric Power Research Institute, Palo Alto, California, January 1987.

EPRI NP-5181SP and EPRI NP-5181M, Northern States Power Company, BWR Pilot Plant Life Extension Study at the Monticello Plant: Phase 1, Electric Power Research Institute, Palo Alto, California, May 1987.

EPRI NP-5289P, Virginia Power Company et al., PWR Pilot Plant Life Extension Study at Surry Unit 1: Phase 1, Electric Power Research Institute, Palo Alto, California, July 1987.

Katz, L. R., "The PLEX Aspects of the Storage or Disposal of Replaced Nuclear Components," p. 276 in Proceedings of the Topical Meeting on Nuclear Power Plant Life Extension, Vol. 1, American Nuclear Society, Snowbird, Utah, July 31-August 3, 1988.

Lott, R. G., and T. R. Mager, "Reactor Vessel Life Extension by Thermal Annealing," pp. 127-32 in Proceedings of the ASME Pressure Vessel and Piping Conference, Honolulu, July 1989.

Massie, H. W., et al., "A Systematic and Cost-Effective Approach to Nuclear Plant Life Extension," pp. 31-34 in Proceedings of the International Conference on Nuclear Power Plant Aging, Availability Factor, and Reliability Analysis, San Diego, 1985.

McBrien, H. W., "Pipes Replaced at Vermont Yankee in Record Time," Nuclear Engineering International (UK), 32, 28 (April 1987).

McCutchan, D. A., et al., "Analysis of Integrated Plant Upgrading/Life Extension Programs," p. 232 in Proceedings of the Topical Meeting on Nuclear Power Plant Life Extension, Vol. 1, July 31-August 3, 1988, Snowbird Utah.

Miselis, V., "The Westinghouse Reactor Vessel Life Extension Variance Study," Proceedings of the Topical Meeting on Nuclear Power Plant Life Extension, American Nuclear Society, Snowbird, Utah, July 31- August 3, 1988.

MITRE Corp., A Topical Study on the Use of Nuclear Plant Aging Information in the Regulatory Analysis of License Renewal Alternatives, WP-8800477, December 1988.

Moore, T., "Nuclear Plants: Life after 40," EPRI Journal, pp. 20-29, October/November 1990.

Morency, K. R., and M. S. McGough, "The Evolution of Steam Generator Replacement Projects in the United States," Nuclear Plant Journal, pp. 72, May/June 1989.

North Anna-1 Steam Generator Changeout Beats Goals," Nuclear News, p. 38, June 1993.

NUREG-1362, Regulatory Analysis for Proposed Rule on Nuclear Power Plant License Renewal, U.S. Nuclear Regulatory Commission, Washington, D.C., July 1990.

NUREG/CR-4012, J. C. Van Kuiken et al., Replacement Energy Costs for Nuclear Electricity- Generating Units in the United States: 1987-1991, Vol. 2, prepared by Argonne National Laboratory for the U.S. Nuclear Regulatory Commission, Washington, D.C., 1986.

NUREG/CR-4555, R. Clark et al., Generic Cost Estimates for the Disposal of Radioactive Wastes, Rev. 1, prepared by Science and Engineering Associates, Inc., for the U.S. Nuclear Regulatory Commission, Washington, D.C., September 1988.

NUREG/CR-4627, E. Claiborne et al., Generic Cost Estimates: Abstracts from Generic Studies for Use in Preparing Regulatory Impact Analyses, Rev. 1, prepared by Science and Engineering Associates, Inc., for the U.S. Nuclear Regulatory Commission, Washington, D.C., December 1988.

NUREG/CR-4731, V. N. Shah and P. E. MacDonald, Residual Life Assessment of Major Light Water Reactor Components: Overview, prepared by Idaho National Engineering Laboratory for the U.S. Nuclear Regulatory Commission, Washington, D.C., June 1987.

NUREG/CR-4921, H. M. Smith and E. J. Ziegler, Engineering and Quality Assurance Cost Factors Associated with Nuclear Plant Modification, U.S. Nuclear Regulatory Commission, Washington, D.C., April 1987.

NUREG/CR-5035, S. K. Beal et al., Database of System-Average Dose Rates at Nuclear Power Plants, prepared by Science and Engineering Associates, Inc., for the U.S. Nuclear Regulatory Commission, Washington, D.C., October 1987.

NUREG/CR-5236, F. W. Sciacca et al., Radiation-Related Impacts for Physical Modifications at Nuclear Power Plants, prepared by Science and Engineering Associates, Inc., for the U.S. Nuclear Regulatory Commission, Washington, D.C., October 1989.

NUREG/CR-5248, I. Levy et al., Prioritization of TIRGALEX- Recommended Components for Further Aging Research, prepared by Battelle Pacific Northwest Laboratories for the U.S. Nuclear Regulatory Commission, October 1988.

NYPA (New York Power Authority), 1989 Outage ALARA Report, NYPA (Indian Point 3), 1989

Rippon, S., "The Dampierre-1 Steam Generator Replacement," Nuclear News, p. 84, September 1990.

SAND88-7095, Forest, L. R. Jr. et al., Cost Savings from Extended Life Nuclear Plants, September 1988.

Sciacca, F. W., Science and Engineering Associates, Inc., letter to D. Cleary, U.S. Nuclear Regulatory Commission, "Transmittal of Discussion of Direct Consequences for Inclusion in Draft License Renewal Regulatory Analysis," January 25, 1990.

Sciacca, F. W., Science and Engineering Associates, Inc., letter to D. Cleary, U.S. Nuclear Regulatory Commission, "Transmittal of Appendix D of Draft License Renewal Regulatory Analysis," February 20, 1990.

Sciacca, F. W., Science and Engineering Associates, Inc., letter to D. Cleary, U.S. Nuclear Regulatory Commission, "Letter Report Presenting Base Case and Typical License Renewal Program Impact Driver Summaries," January 3, 1993.

Sciacca, F. W., Science and Engineering Associates, Inc., letter to D. Cleary, U.S. Nuclear Regulatory Commission, "Bases and Assumptions Used in Developing Updated Base Case and Typical License Renewal Program Scenarios," January 13, 1993.

Sciacca, F. W., et al., Science and Engineering Associates, Inc., Cost Analysis of Alternative LWR License Renewal Regulations, Report SEA 87-288-09-A:1, U.S. Nuclear Regulatory Commission, March 1989.

Sciacca, F. W., et al., Impact Driver Definition for Nuclear Plant License Renewal Generic Environmental Impact Study, Report SEA 89-461-09-A:1, U.S. Nuclear Regulatory Commission, August 1990.

"Steam Generator Replacement at Dampierre 1, France,"

Nuclear Plant Journal, pp. 93-95, May/June 1990.

Zachary, J., et al., "ALARA at the Peach Bottom-3 Pipe Replacement Project," Nuclear News, pp. 34- 37, September 1989.


Appendix B -- Attachment 1 : Recent Industry Experience for Major Refurbishments and Estimates for Plant Life Extension Costs


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A literature search for plant life extension and license renewal related cost information was conducted in support of the draft Generic Environmental Impact Statement published for comment in August 1991. That literature search focused on obtaining industry-derived data on inspection, surveillance, test, and maintenance (ISTM) actions and on major repair, replacement, or refurbishment actions undertaken in the past or planned in support of license renewal and plant life extension (PLEX) activities. The information collected is discussed in this attachment. Since the search was performed in 1991, all information cited dates from 1991 or earlier. The cost information has not been updated to 1994 dollars.

For each activity of interest to this evaluation (e.g., reactor pressure vessel replacement, steam generator replacement), information such as capital cost, labor, radiation exposure incurred, radioactive waste type and volume generated, and outage duration was obtained, if available. This search resulted in the identification of numerous references; however, most did not provide the needed detail on the aforementioned items. Many of the references presented overall PLEX cost in dollars per kilowatt but did not provide a breakdown of individual activities. It is important to note that most references recognized that the PLEX costs will vary significantly depending upon the reactor type and the vintage.

The following sections present information organized by specific component or topic. Following these sections is a list of references reviewed.

STEAM GENERATORS

Although several utilities have replaced their steam generators, the majority are searching for ways to extend the life of their operating steam generators, such as heat treating and sleeving to avoid the cost of replacement. In "Steam Generator Replacement at Dampierre 1, France," Nuclear Plant Journal May/June 1990, it is reported that a steam generator replacement at Dampierre 1, France, had a cost estimate of $106 million, including $3.5 million for three steam generators. Additionally, Rippon 1990 reported the steam generator replacement took 200,000 work hours and resulted in an exposure of 220 person-rem.

In Eckert (1987), the replacement of steam generators in a two-loop plant was estimated by Kraftwerk Union to require 2.5 months using 140,000 work hours and result in a total dose of 700 person-rem. Additionally, the planning of the steam generator replacement took 45,000 work hours.

Item Palisades Turkey Point Surry
Direct cost (million dollars) 75 102 81
Replacement power cost (million dollars) 200 124 --
Total (million dollars) 275 226 --
Outage time 2 yrs 207 days 8.5 mos

In SAND88-7095, the replacement of steam generators is assigned a probability. This probability will be zero if the component has been replaced, or low if the component is of current design. However, older plants will have a high probability of replacement. The cost of steam generator replacement has been estimated at $20 million (1986 dollars) multiplied by the number of loops multiplied by the probability of replacement. EPRI NP-2418 provides the following information on steam generator replacement or partial steam generator replacement at three plants.

For Turkey Point and Surry, the operation involved a partial replacement of the steam generators.

In EPRI NP-4208, the following table outlines steam generator replacement cost, outage time, and the collective dose.

In NYPA 1989, information was presented on worker exposure in person-rem for the steam generator replacement at Indian Point 3. For an outage which included refueling and maintenance and steam generator replacement, the total dose was 852 person-rem. Of the 852 person- rem, 541 person-rem were attributed to the steam generator replacement.

Plant

Rating [MW(e)] Replacement year Outage (months) Cost (million dollars) Collective dose (person-rem)
Surry 2

775

1979-80

10

80 2140
Surry 1

775

1980

--

-- 1760
Turkey Point 3

666

1981-82

10

  2150
Turkey Point 4

666

1982-83

7

190a 1305
Point Beach 1

497

1983-84

6

50 590
Robinson 2

665

1984-85

8

85 1207
a For both Turkey Point 3 and 4.

Figure B-1.1 presents a comparison of seven steam generator replacement projects with respect to outage duration and personnel exposure. It was presented in Morency and McGough 1989.

Figure B-1.1. Outage duration and personnel exposure in seven steam generator replacement projects.

Katz (1988) indicates that for steam generator replacement in a two-loop plant, the exposure in person-rem and the labor in work hours are estimated to be 1,387 and 624,000, respectively. The estimated cost for on-site storage of the steam generator is $735,000.

The cost to immediately cut up and ship the steam generator is estimated to be $20,980,000.

The following table provides information on steam generator size, weight, and storage volumes.

Plant

Length (ft)

Diameter min/max (ft) Nozzle size (in) Dry weight (tons) Total storage volume (ft3) Portion of storage allotmenta (%)
Plant A (4 loop)

46

10/12

29

209 37,340 56
Plant B (2 loop)

63

10.6/13/6

31

305 31,653 48
Plant C (3 loop)

63

10.6/13/6

31

305 47,470 67
Plant D (3 loop)

67.6

11.3/14/6

31

331 57,864 82
a The portion of storage allotment refers to allocations for the disposal of low-level radioactive wastes at existing U.S. civilian disposal sites.

Note: To convert ft to m, multiply by 0.305.
To convert in. to cm, multiply by 2.54.
To convert ft3 to m3, multiply by 0.03.

BWR RECIRCULATION PIPE REPLACEMENT

Eckert (1987) has estimated that the exchange of two recirculation loops and six risers will require 2.5 months, 50,000 work hours for the preplanning, and 220,000- 380,000 work hours to execute the activities (including training) and will result in 500- 800 person-rem of total dose.

In Zachary et al. (September 1989), the radiation exposure for BWR major pipe replacement was reported for Peach Bottom 2 and 3 and compared with other plants.

In SAND88-7095, the replacement of piping is assigned a probability. This probability will be zero if the component has been replaced, or low if the component is of current design.

Radiation exposure incurred for boiling-water major pipe replacement (person- rem)
Peach Bottom 2 2200
Peach Bottom 3 1074
PL-A 1900
PL-B 1785
PL-C 1785
PL-D 1638
PL-E 1580

Note: To convert person-rem to person-sievert, multiply by .01.

However, older plants will have a high probability of replacement. The cost of boiling-water reactor (BWR) piping replacement costs were estimated to be $75 million multiplied by the probability of replacement. BWRs with Mark I designs probably will be more problematic to refurbish than those with Mark II or Mark III designs. A rough estimate is that refurbishment of the Mark I design will cost $25 million more than refurbishment of the Mark II or Mark III.

In EPRI NP-4208, the following table is presented which outlines BWR piping estimated replacement cost and outage time.

The costs and downtime that have been reported for BWR recirculation pipe replacements range from $19 million to $65 million and from 6 to 10 months, respectively.

According to McBrien (April 1987), during a 9-month refueling outage in 1986 at Vermont Yankee, the entire recirculation piping and part of the plant's residual heat removal system were replaced at a cost of approximately $60 million dollars. Workers at the plant acquired a total exposure of 1786 person-rem.

Plant Rating [MW(e)] Estimated costs (million dollars) Estimated outage (months)
Browns Ferry 1067 42 (budget) 6
Dresden 3 794 40 (budget) 9-10
Hatch 2 806 -- 6
Monticello 536 19 (budget) 6
Nine Mile Point 1 610 65 10
Pilgrim 1 670 40 (budget) 9-10

PRESSURE VESSEL COSTS AND THERMAL ANNEALING

In EPRI-4208, it is reported that a dry anneal at 850?F for 168 hours will restore most fracture toughness properties lost during irradiation embrittlement. However, this indicates that the vessel internals must be removed and a heating system installed. However, there are problems related to post-anneal behavior that need to be resolved. These include the following.

In EPRI NP-2418, the following reactor pressure vessel replacement information is presented. The total cost is in 1979 dollars and excludes the cost of money and replacement power. The total time of replacement is approximately 5 months.

Reactor pressure vessel replacement costs (million dollars)

Direct costs  
Material 34
Labor ($25-35/h) 17
Indirect Costs  
Occupational exposure 13
Project supervisors 17
Consultants, management ($40 h)  
Subtotal 81
Contingency (50%) 41
Total 122

In EPRI NP-4208, reactor pressure vessel replacement has been estimated to cost $100 million to $150 million (1979 dollars) and to require 2-3 years.

Abbot et al. (1988) report a cost of $20 million to $50 million for three potential tasks as outlined:

In Lott and Mager (1984), the estimated costs for a severely embrittled reactor vessel and a moderately embrittled vessel are presented. In some circumstances it may be advantageous to anneal the vessel earlier in plant life to accrue the benefits of annealing immediately. A severely embrittled vessel is one in which the embrittlement surpasses reasonably acceptable limits. For the severely embrittled reactor vessel, if the vessel can be annealed for less than $200 million then the cost of annealing is less than the savings associated with deferring the construction of a replacement plant by 1 year. This assessment is based upon a modest replacement cost of $2 billion for the power station and an annual cost of capital of $200 million, based upon financing rates of 10 percent. The moderately embrittled vessel is one for which embrittlement is projected to exceed acceptable limits before the end of the useful life of the reactor. There is a clear savings from deferring the large costs associated with the annealing. However, if the annealing can be used to increase the plant availability, then there is a clear benefit to annealing. An analysis conducted indicates that an increase in total plant availability of 1 day is approximately equivalent to $500,000. To make performing an early anneal financially advantageous, the annual cost benefit from annealing should exceed the financing costs associated with the anneal. Assuming an interest rate of 10 percent, the plant availability would have to increase by 2 days per year to justify a $10 million expense on annealing.

NUPLEX CAPITAL COSTS AND REPLACEMENT POWER COSTS

In SAND88-7095, a range of $250- $500/kW (1986 dollars) for nuclear plant- life extension (NUPLEX) refurbishment on an overnight basis ($300-$600/kW including financing) is reported. The overnight costs for Surry Unit 1 pressurized-water reactor have been estimated at $250/kW and for the Monticello BWR the costs have been estimated at $500/kW. More recent information presented in Moore (1990), indicates that the Monticello overall cost for extending operation is estimated to be $200/kW. That same source cites $150/kW as the capital cost for the Yankee Rowe plant for running a 20-year renewal term. Westinghouse has estimated the cost of NUPLEX refurbishment as ranging from $240/kW to $900-$1000/kW based upon the amount of refurbishment needed. The higher estimate is a result of replacement of most major components and annealing of the pressure vessel.

McCutchen et al. (1988) reports a life extension program cost of $270/kW or $318 million.

Massie et al. (1985) assumed in their calculations a typical Westinghouse three-loop plant rated at 775 MW(e), with a mid-life in year 1992, and replacement power costs of $350,000/day in 1985 dollars.

BIBLIOGRAPHY FOR APPENDIX B ATTACHMENT 1

Braun, C., "Economic Factors Related to Life Extension of Nuclear and Coal-Fired Plants," Transactions of the American Nuclear Society, 46, 558-59 (June 1984).

Byron, J., and M. Lapides, "Guidelines for the Life Extension of Nuclear Power Plants," Draft EPRI Document, July 1988.

"Interim Draft for Generic Environmental Impact Statement for License Renewal of Nuclear Power Plants," Prepared for U.S. Nuclear Regulatory Commission, October 15, 1990.

Katz, L. R., "The Codes and Standards Aspects of Nuclear Plant Life Extension," pp. 131-38 in Proceedings of the ASME Pressure Vessel and Piping Conference, Pittsburgh, June 19-23, 1988.

Lapides, M. E., "Making Decisions on Plant Life Extension," Nuclear Engineering International (UK), 34, 26-27 (August 1989). Moylan, M. F., et al., "Reactor Vessel Life Extension," in Proceedings of the ASME Pressure Vessel and Piping Conference, San Diego, June 28- July 2, 1987.

Stancavage, P. P., "BWR Life Extension Economics," pp. 309-315 in Proceedings of the 1986 Joint ASME/ANS Nuclear Power Conference: Safe and Reliable Nuclear Power Plants, Philadelphia. Tipping, P., et al., "Study of the Mechanical Property Changes of Irradiation Embrittled Pressure Vessel Steels and Their Response to Annealing Treatments," in Transactions of the 9th International Conference on Structural Mechanics in Reactor Technology, Vol. G, Fracture Mechanics and NDE, pp. 115-27, Lausanne, Switzerland, August 1987.


Appendix C : Socioeconomics


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C.1 Research Methods


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The social impact assessment methods employed in this project were designed to identify the significance level of potential socioeconomic impacts during refurbishment and the license renewal term and to identify relationships between key social factors (impact predictors) and the intensity of impacts. The research methods used consisted of a literature review, a search of newspaper citations, a survey of all nuclear utilities, and seven detailed case studies.

The impact categories examined were limited primarily to those socioeconomic effects associated with project-induced employment (direct and indirect), population growth, expenditures, and tax payments. This approach is traditionally followed in preparing environmental impact statements (EISs) involving the construction and operation of nuclear power plants. The key socioeconomic topics suggested for in-depth examination by the literature search and citation review were (1) population, (2) housing, (3) taxes, (4) public services, (5) off-site land use, (6) economic structure, and (7) historic and aesthetic resources.

The following sections provide additional detail on the literature review, the review of newspaper citations, the utility survey, the seven case studies, and the techniques used to analyze the past and projected impacts associated with nuclear power plants. C.1.1 Literature Review

The purpose of the literature review was to identify important socioeconomic issues, to obtain an industry-wide summary of the impacts that had occurred in those subject areas as a result of past nuclear plant construction and operations, and to identify possible causal factors related to those impacts. The literature review focused largely on EISs prepared for nuclear power plants at the time of their application for an operating license [Atomic Energy Commission (AEC) and U.S. Nuclear Regulatory Commission (NUREG) final environmental statements]. In addition to projecting the future impacts of plant operation, many of these documents summarize the impacts that occurred during plant construction. Along with these EISs, several detailed retrospective studies of impacts that had occurred at specific nuclear power plants were examined. Section C.2.1 provides a detailed discussion of the literature review.

C.1.2 Review of Newspaper Citations

Citations from five major metropolitan newspapers and a national wire service were examined to check the completeness of the socioeconomic impact categories suggested through the literature review. The newspapers were the Atlanta Constitution, the Houston Post, the Los Angeles Times, the New York Times, and the Washington (D.C.) Post. The wire service was United Press International. The search spanned 1989 and the first 5 months of 1990. Potentially relevant articles were identified through a computer database search, using the key words "nuclear power" and "nuclear power plant" in conjunction with a number of other words and phrases including "public reaction," "public concern," and "public opinion." Over 400 articles were identified through this search, although upon review, many were found to be irrelevant for this study. Overall, the traditional socioeconomic impact areas described have received little recent attention in the print media.

C.1.3 Survey of Utilities

Two written surveys of the nation's nuclear utilities were conducted. The survey instruments were designed by Oak Ridge National Laboratory (ORNL), with substantial input from the Nuclear Regulatory Commission (NRC) and the Nuclear Management and Resources Council (NUMARC). Both were administered by NUMARC.

The first survey instrument, sent to all U.S. nuclear utilities, was designed to elicit important descriptive information on plant operations. The respondents provided an industry-wide picture of current and past numbers of plant workers and of nuclear plant financial contributions to host communities. Usable data were received for some portion of these questions for 66 of the 74 nuclear plant sites nationwide.

The second survey instrument was sent to the utilities that operate the seven socioeconomic case study plants, and responses were received from all seven utilities. The purpose of these items was to gather detailed information on worker residential location, plant expenditures, and tax payments to local communities so that the causal factors related to past impacts could be identified and future impacts could be predicted.

C.1.4 Case Studies

The seven nuclear plants were chosen for detailed study as representative of all U.S. nuclear plants in terms of the socioeconomic characteristics of their host communities. The site-selection methodology and the plants chosen are described below. The case study examination was designed to provide detailed information on past impacts at a sample of nuclear power plants that represent the range of plants nationwide and to allow the projection of future impacts in key issue areas.

Detailed information was obtained on the seven case study sites through a review of EISs and site-specific NUREG reports, as well as through telephone interviews conducted with state and local officials and other expert sources. The sources were chosen for their expertise in the socioeconomic issue areas addressed (e.g., housing, land use) and included employees of local planning agencies, chambers of commerce, and economic development agencies; local tax assessors and treasurers; officials at state employment offices; and local media personnel. Nearly 300 telephone interviews were conducted at the seven case study sites. A detailed telephone protocol was used to collect data at the five sites previously studied by Mountain West Research, InC. (NUREG/CR-2749, vols. 1, 4, 5, 7, 12), in a postlicensing study conducted for the NRC. A more exhaustive protocol was used for the two case study sites that had not been studied previously by Mountain West and for which more information was needed. Section C.7 contains the questions asked in these interview protocols.

The seven case study sites chosen represent roughly 10 percent of the U.S. nuclear power plant sites. The primary factor considered when selecting sites for socioeconomic study was the population of the area surrounding the plant. Population was chosen as the primary factor because the literature reviewed and other previous experience suggested a strong relationship between an area's remoteness and the magnitude of impacts related to population growth, employment, expenditures, and taxes. Plant age and location were also considered so that the sample includes sites representing various licensing dates, population characteristics, and geographic sections of the United States.

In considering plants for this study, preference was given to those sites for which detailed historical data about socioeconomic impacts were available. Thus, 12 plants studied by Mountain West Research, InC. (NUREG/CR-2749), and 2 plants studied by the Electric Power Research Institute (EPRI) were considered first: Arkansas Nuclear One (ANO), Bellefonte, Calvert Cliffs, D. C. Cook, Crystal River, Diablo Canyon, Nine Mile Point, Oconee, Peach Bottom, Rancho Seco, St. Lucie, San Onofre, Surry, and Three Mile Island (TMI).

Each of these plants was classified according to the remoteness of its location, based on a classification scheme developed by Battelle Human Affairs Research Centers for Sandia National Laboratories (NUREG/CR-2239). Site remoteness involves population density in the area near the plant and the plant's distance from large cities. Battelle combined both these factors to measure "sparseness" and "proximity." Sparseness measures population density and city size within 32 km (20 miles) of the site, whereas proximity focuses on density and city size within 80 km (50 miles). Each measure involves four categories. Although Battelle expressed these categories in terms of numbers of people within 32- and 80-km (20- and 50-mile) radii, the absolute numbers were converted to the number of persons per square kilometer so that 1986 census data could be used for comparison and site selection. The sparseness and proximity measures used to classify potential case study sites are shown in Table C.1.

Three population classifications take into account the combination of the four-point sparseness and proximity scales. The three population classes are low, medium, and high population. Low corresponds to the most sparse population category and sites not in close proximity to large cities, whereas high corresponds to the least sparse population category and sites that are in close proximity to large cities. The bounds of each population classification are shown in Figure C.1.

Because only 1 of the 14 previously studied plants listed earlier fell into the low population category, 4 additional low-population sites were considered for inclusion in this study: Big Rock Point, Cooper, Wolf Creek, and Hatch. The Indian Point site also was added because it is located in an area with high population density and in close proximity to New York City. The applicable population classification for each of the 19 potential case study sites is shown in Table C.2.

The seven case study sites selected from those listed above were chosen to represent a broad range of population remoteness, geographic location, and plant age. The major characteristics of the plants chosen are shown in Table C.3. Their geographic distribution is illustrated in Figure C.2. All the sample plants have pressurized water reactors (PWRs), which in the bounding case scenario will require 84 percent more workers for refurbishment than will boiling water reactors (BWRs). The main analysis of potential socioeconomic impacts is based on this bounding case scenario. In the typical case scenario, the work force required to refurbish BWRs is projected to be 72 percent larger than that required for PWRs. (See Section 3.3.1.1 for details about work force projections.)

Figure C.1 Population categories, by sparseness and proximity.

Figure C.2 The seven case study nuclear plants.

C.1.5 Analysis of Impacts

C.1.5.1 Defining Significance Levels for Each Impact Category

For each socioeconomic topic, the characteristics of small, moderate, and large levels of impact were defined. These definitions were developed on the basis of Council on Environmental Quality regulations (40 CFR Part 1500), information from site-specific nuclear EISs and NUREG studies, interviews with local information sources, studies concerning nity response to nuclear and non-nuclear technologies, and best professional judgment. The definitions of significance for each socioeconomic topic are presented in Sections 3.7.2 through 3.7.7 and Sections 4.7.2 through 4.7.7.

C.1.5.2 Characterizing Past Impact Levels and Identifying Impact Predictors

Descriptions of past impacts in all socioeconomic issue areas were gathered through the data-collection methods described in Sections C.1.1 through C.1.4. These impacts then were characterized as small, moderate, or large on a site-specific basis, using the significance level definitions discussed in Chapters 3 and 4. A description of past impacts for each of the case study sites is presented in Sections C.4.1 through C.4.7. From these site-specific characterizations of the representative case study plants, generalizations were made concerning the range of past impacts for all nuclear plants nationwide.

In examining the impacts identified in available reports and EISs and through the detailed case studies, it is apparent that the extent to which socioeconomic impacts are experienced at a given project site would depend on several factors. These factors, which will be referred to as impact "predictors," consist of characteristics of the project (called "drivers" in Appendix B) as well as characteristics of the area in which a plant is located. The specific predictors identified include local population characteristics; the employment, expenditures, and tax revenues generated by the project; and the existing infrastructure of the project's host community or communities. By looking at impact predictors and the resulting impacts that occurred at many different sites during plant construction and operations, the relationships between predictor magnitude and impact significance were identified. These relationships are discussed in Sections 3.7.2.3 through 3.7.7.3 and in Sections 4.7.2.3 through 4.7.7.3.

C.1.5.3 Projecting Future Impacts

The first step in projecting impacts was to obtain projections of the number of direct workers required for refurbishment-related activities and for operations during the license renewal term. Projections of the refurbishment work force were prepared by Science and Engineering Associates, InC. (SEA 1994); they are presented in Chapter 2 and in greater detail in Appendix B, and are discussed in Section C.3.1.1. The number of refueling and maintenance workers employed at a typical plant during past outages was obtained from the survey of all U.S. nuclear utilities (Section C.1.3) and verified by SEA. SEA used information from the survey and the literature regarding the number of person-months required for normal refueling outages to develop an estimate of the number of refueling workers likely to be on-site during current-term and final refurbishment outages. The number of operations workers currently employed at each case study plant was obtained from the survey of all U.S. nuclear utilities, whereas the number of additional permanent workers required for plant operation during the license renewal term came from descriptions of proposed inspection, surveillance, testing, maintenance (ISTM) tasks (Section C.3.1.2). Additional detail about work forces required during refurbishment outages and the license renewal term are provided in Sections C.3.1.1 and C.3.1.2, respectively.

To project the maximum impacts likely to result from a plant's license renewal refurbishment activities, the staff based its socioeconomic impact analysis on the bounding case work forces projected by SEA (1994). The conservative nature of the bounding case scenario is described in Appendix B. Because the bounding case work force estimate for PWRs (2273 workers at peak) is larger than the estimate for BWRs (1482 workers), the staff conducted its primary analysis of potential socioeconomic impacts using the projected PWR work force. This analysis has identified some issues for which moderate or large adverse impacts are possible. For these issues, the potential for less severe impacts associated with the bounding case work force at BWR sites has been considered. For these issues, an analysis of the potential impacts at BWR sites is provided and is based on the 1500-person work force associated with the bounding case BWR refurbishment scenario. For those issues where moderate or large adverse impacts were determined to be possible with 1500 workers, an analysis of the typical case refurbishment work force (1017 workers at BWR sites) has been conducted. Those issues for which moderate or large adverse impacts were found to be possible with a work force smaller than 1017 (i.e., operations-period refueling work forces) have not been subjected to these additional analysis.

Using the work force projections, the staff determined the number of indirect jobs that would be created as a result of refurbishment and license renewal. Indirect employment was projected using the Regional Industrial Multiplier System (RIMS) direct/indirect job ratios calculated for each plant in the NUREG/CR-2749 study. Using the employment projections for direct and indirect workers, projected changes in local population were calculated based on patterns of worker residential location, in-migration, and family size identified in the site-specific NUREG reports. Patterns for refurbishment and refueling/maintenance workers were assumed to follow those established by plant construction workers, and patterns for additional permanent license renewal term workers were assumed to follow those established by current term plant staff (Section C.4.1.1). Population changes caused by temporary refueling and maintenance workers involved in periodic plant outages during the license renewal term were not studied in detail because these workers would be employed for very short periods of time, but evidence about past effects during such outages was collected and considered in the analysis.

The projections of direct and indirect employment associated with refurbishment and routine (nonoutage) operations were used to assess the economic impacts of refurbishment and license renewal. Economic impacts were projected by comparing estimated plant-related employment (direct and indirect) with projections by the National Planning Association (NPA) of total employment for the study areas during the refurbishment period and the license renewal term (Section C.4.1.6). For the refurbishment period, employment estimates for all refurbishment-related workers (including refueling/maintenance workers) were used. For the operations period, estimates of all permanent (nonoutage) workers were used; this includes additional jobs and those continuing from past operations.

Because many socioeconomic impacts are driven by population growth, the next step in projecting the impacts of refurbishment and license renewal was to use the population growth projected for each case study area in assessing impacts to housing, public services, and off-site land use. For each of these topic areas, impact predictions were made by comparing levels of impact significance associated with past plant-related population growth to projected population growth and by examining projected conditions of key infrastructure components. The analysis assumes that no other major construction project will occur concurrently with plant refurbishment and subsequent refueling and maintenance activities. If other large construction projects are ongoing when these activities occur, impacts could be greater than those predicted. Housing impacts were projected for refurbishment and continued operations, respectively, by comparing the housing demand expected to result from the in-migration of refurbishment-related workers and of additional permanent operations workers with projections concerning local housing markets (number of units and vacancies) generated from U.S. Census data (Section C.4.1.2). In addition, evidence concerning past impacts associated with the influx of refueling and maintenance workers during plant outages was gathered and used as an indicator of possible outage-related impacts during the license renewal term at the one site where significant growth-induced housing impacts were predicted for the refurbishment period. Public service impacts were projected by comparing the number of refurbishment-related workers and additional permanent operations workers expected to in-migrate with the local communities' projected capacity to provide public services, as indicated by local information sources (Section C.4.1.4). As with housing, evidence was gathered at one site concerning outage-induced transportation impacts. For off-site land use, impacts were projected by examining the size of anticipated population growth resulting from refurbishment-related workers and additional permanent operations workers relative to state data center projections of a study area's total population. Potential changes in land-use patterns caused by plant tax payments were also considered in projecting the impacts of license renewal (Section C.4.1.5).

Unlike the subjects already discussed, impacts to taxes and to historic resources and aesthetic resources are not driven primarily by changes in employment and population. For these three topics, impacts examined for the refurbishment period are those that result solely from changes induced by refurbishment-related activities (i.e., increased tax assessments and modified plant structure). In contrast, the assessment of license renewal term impacts includes continuing impacts from past operations and the new impacts already discussed. A detailed discussion of the methods used to predict impacts in each of these subject areas is presented in Sections C.4.1.3 and C.4.1.7, respectively.

Socioeconomic impacts identified and analyzed here are site-specific; they have no statewide or nationwide consequence. Therefore, simultaneous relicensing of several nuclear power plants will not result in cumulative regional or national impacts.

Judgments on whether or not potential environmental impacts in each subject area would need to be further addressed in each individual license renewal application were made based on the nature of the projected impact and its level of significance. These conclusions are not discussed in Appendix C but are presented in the body of the Generic Environmental Impact Statement (GEIS). Because of uncertainty surrounding the number of workers that would actually be required for plant refurbishment, a sensitivity analysis was performed wherein socioeconomic impacts were predicted in response to a work force roughly 50 percent larger than the estimated peak work force for the major refurbishment outage provided in Section C.3.1.1 (even though the estimate given in that section was designed to be an upper bound for a typical plant). The discussion of conclusions for each socioeconomic topic in the body of the GEIS states whether or not the conclusion category (1 or 2) expected for the preferred estimate would change in response to the larger work force.

Sections C.4.1 through C.4.7 present a detailed discussion of projected socioeconomic impacts in each of the above subject areas for each case study site, and Sections 3.7.2 through 3.7.7 and Sections 4.7.2 through 4.7.7 summarize these impacts and project impacts for all nuclear plant sites based on the case study predictions. These nationwide predictions are considered valid because the impacts predicted at the case study sites represent the range of potential impacts at existing nuclear plants. Population, which is considered an impact predictor rather than an actual impact, is discussed in Sections C.4.1.1 through C.4.7.1, 3.7.1, and 4.7.1.


C.2 Baseline Description


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C.2.1 Overview of Past Population- and Tax-Driven Nuclear Plant Impacts

C.2.1.1 Objectives

This literature review summarizes the results of previous case studies that examined the socioeconomic impacts of nuclear power plants. The objective of the review was to identify

Socioeconomic impacts occurring during either construction or operation are of interest. Construction impacts provide an upper bound to what might happen during a major plant refurbishment, whereas operations impacts typify what would occur after license renewal, allowing for adjustments for refurbishment-induced changes in work forces, taxes, or other impact drivers.

C.2.1.2 Literature Reviewed

Several categories of literature were reviewed. One major category is EISs for nuclear plant operating licenses (OLs) (Section C.5.3 lists the EISs reviewed). This category is potentially useful because the EISs not only consider impacts from plant operations but often summarize impacts that occurred during construction. They are official documents in support of the NRC's regulatory process and, presumably, carry a measure of credibility in respect to what the regulatory process requires in terms of data, findings, and content. The second category of literature includes case studies commissioned by organizations closely involved with the nuclear industry. The NRC has conducted several such plant-specific studies (NUREG/CR-2750, ORNL/NUREG/TM-22, and NUREG/CR-0916). EPRI also conducted a series of case studies of power plants, two of which were nuclear (EPRI/EA-2228). The third category of literature encompasses studies of single nuclear plants that are usually sponsored by utilities as part of the regulatory process or by some other organization interested in documenting socioeconomic impacts.

C.2.1.3 Types of Impacts

This literature review reveals no population- or tax-driven socioeconomic impacts other than those typically assessed in environmental impact documents. Those documents written in support of the National Environmental Policy Act of 1969 (NEPA) process almost always focus on readily quantifiable impacts to public services, housing, the economy, and land use. Exceptions in which the assessment is qualitative include aesthetic and cultural resources impacts, which are almost always considered in EISs, and recreational impacts, which are discussed in 44 of the 78 (56 percent) OL EISs examined. Impacts to these resources, however, are seldom found to be population- or tax-driven. Consequently, they are defined by general statements about appearances of the plant, transmission lines (sometimes the lines are rerouted to avoid negative impact to residents in the vicinity), and compatibility with nearby cultural resources. Recreation impacts are generally defined as positive contributions such as visitor centers; boating, fishing, and hiking activities; and dedication of land on the plant site to natural resources conservation and education. Typically, non-NEPA documents do not examine these resources.

Among public service impacts identified, the assessments focus on schools, transportation, and public safety; less emphasis is placed on utilities, water and sewer facilities, and health and welfare services. These same impacts are covered in the other case studies examined in the literature, and there is a strong consensus that all should be treated as valid kinds of socioeconomic impacts under NEPA. Housing is another impact that understandably receives considerable emphasis in NEPA and non-NEPA assessments, with residential distribution being foremost in importance, followed by housing type and costs. Generally, housing impacts are treated before public service impacts because most services generally support people by place of residence.

Economic impacts are almost always assessed in environmental documents and related case studies. They are readily quantifiable and tangible impacts that are easy to understand. Minimally, the project work force total and annual payrolls are included (although early NRC EISs did not note these basic impacts). Emphasis normally is on totals of direct employment and payroll generated by the nuclear plant, indirect jobs in the local economy, amounts of money contributed to the regional economy, and tax contributions to the local tax base--particularly property and sales taxes. Additional types of taxes and amounts of revenue that flow to the state governments generally are not considered in NEPA documentation, although case study reports produced by NRC more adequately assess these latter impacts.

Land-use impacts created directly by the plant itself and its transmission lines are assessed in 78 percent of OL EISs. Generally, such impacts are considered in terms of acres disturbed, people relocated, and land flooded for cooling lakes. Only rarely is attention directed at what effect plant siting would have on broader community land use and associated growth. These generalizations hold true for the other literature categories reviewed, with the exception of the NRC's series of 12 case studies, which gives considerable attention to land use and associated community growth and finds that land-use changes related to plant siting and worker in-migration "strengthened overall patterns of change and development" (NUREG/CR-2750).

C.2.1.4 Causal Factors of Impacts

Most of the socioeconomic impacts created by nuclear plants are driven by the plant-related population or taxes. Nuclear power plants require large numbers of workers to construct and, to a lesser extent, operate, and they generally pay significant amounts of taxes. The amounts of jobs and taxes tend to correlate fairly directly with the size of the plant. In many communities where nuclear plants are located, the plant is very likely to be one of the largest, or even the largest, employer and contributor to the tax base. Therefore, its workers create impacts on schools, public services, housing, utilities, transportation, and health and welfare services. Of particular importance are in-migrating workers and their families, who must be accommodated by expansion of such services. If the communities are small and the plant site is located beyond commuting distance of a reasonably large population center, then worker in-migration would be higher and resulting demands of services would be increased--perhaps to the point that major expansion is required.

Taxes from the plant and its workers provide a major benefit to local communities in helping to pay for public services. Once built, a nuclear plant typically contributes millions of dollars annually to the local tax base. The range may be as low as $1 or 2 million and as high as $42 million (1990 dollars), depending upon the assessed plant value and tax structure, according to the 69 percent of OL EISs that discuss the issue. An even broader range of tax payments was reported by utility respondents to a recent questionnaire (Section C.2.2). Although the effects of plant tax payments are primarily positive, potentially negative tax-related issues can involve (1) the timing of tax revenues that may be too late to pay for construction-period impacts caused by in-migrating workers; (2) discontinuities between jurisdictions that can tax the plant and jurisdictions in which many of the employees reside; and (3) the disproportionately large amount of the tax base represented by the power plant, which can pose a major problem in the future for local real estate tax revenues. A fourth issue is how to pay for public services in the case of nuclear plants owned by the government that pay no local property taxes and only small payments in lieu of taxes. As in the case of population impacts, tax-related impacts have the potential for being much larger in rural areas with small tax bases.

C.2.1.5 Impact Thresholds

In developing a license renewal rule, the emphasis is on identifying socioeconomic impacts that could be particularly problematic for local communities and the conditions under which these could occur. For example, is the labor requirement so large and the community so remote from a population center that in-migration to the plant's area would be large enough to significantly stress the public infrastructure? Or, in regard to taxes, is the assessed value of the plant so large relative to the existing tax base that local governments would be highly dependent on plant-related revenues? An impact threshold can be thought of as the set of conditions (e.g., a particular number of plant-related workers in conjunction with a host community's population and distance from major urban areas) under which significant impacts can be expected.

Before addressing the issue of thresholds, it is useful to generalize very briefly about the severity of these kinds of impacts that could result from nuclear plants. The literature that specifically deals with the issue (NUREG/CR-2750; NUREG/CR-0916) notes that nuclear plants seldom, if ever, create massive (boomtown-level) impacts to community infrastructure on the scale of mine-mouth coal plants or hydroelectric dams because nuclear plants are not sited in remote western regions of the country where such severe impacts can occur. This finding is supported by the OL EISs, which fail to identify any potential or actual case of a boomtown-level impact from a nuclear plant. Indeed, the overall findings from the literature reviewed argue strongly for the proposition that population- and tax-driven impacts of nuclear plants during the construction period overwhelmingly are small to noticeable for affected communities and well below boomtown proportions. Significant negative impacts have occurred, however, at a few plant sites. During plant operation, employment and tax revenues can be a substantial benefit for local communities, and their loss would be equally significant. Of particular concern would be cases in which nuclear plants make up a large percentage of the tax base.

In regard to the identification of impact thresholds during plant refurbishment and continued operations, the literature alone does not give any clear answers. However, in conjunction with the case studies detailed in Section C.4, a number of impact predictors were identified that can be used to indicate whether significant impacts are likely to occur at a given site in several socioeconomic subject areas. These impact predictors are discussed in Sections 3.7 and 4.7.

C.2.2 Overview of Current and Past Socioeconomic Characteristics for All Plants

This section summarizes information on selected socioeconomic characteristics for U.S. nuclear plants. Specific topics include the plants' operating period employment; characteristics of typical planned outages, in-service inspections (ISIs), and largest single outages; assessed value; and tax payments. The section is intended to provide an overview of the entire U.S. commercial nuclear industry. Information used to prepare this report was obtained through questionnaires mailed to all utilities that operate nuclear plants.

Considerable differences exist among nuclear plants in terms of the size of their operating work forces. Table C.4 provides a summary of data concerning current operating-period employment at nuclear plants grouped by the number of units at each plant. Although the employment figures are intended to represent the number of permanent personnel on-site, they might overstate that figure because it is likely that temporary workers were included in some utilities' responses. The incremental increase in the mean number of workers per unit is not linear, because the mean for one-unit plants constitutes over 66 percent of the mean for two-unit plants, whereas the mean for two-unit plants represents only 52 percent of the mean at three-unit plants. The number of units at each station represents those licensed by 1990; therefore, for some two- and three-unit plants, employment may have been considerably lower in past years depending upon the number of units that were actually on-line and their levels of operation. Table C.5 depicts changes in mean operating-period employment at the plants from 1975 to 1990.

U.S. nuclear plants also differ in the number of workers, the costs, and the length of time required per unit for various types of outages. Table C.6 depicts the minimum, maximum, and mean number of workers, costs, and time required per unit for a typical planned outage during which refueling and other routine tasks are performed. Table C.7 presents the same information for an ISI outage. The two tables should be read in columns, not rows, because the plant that had the minimum number of workers, for example, is not necessarily the same as the plant with the lowest cost or the shortest outage. The numbers presented in Tables C.6 and C.7 might be high because some utilities probably included permanent operations workers in their count of workers used during outages, even though they were asked to list only additional on-site workers. Also, the numbers count each worker who came on-site at some time during the outage, regardless of the duration of the stay. These numbers, therefore, are almost certainly higher than the peak number of workers on-site during a single day or week. The maximum number of 2600 is particularly suspect; the next highest number of workers given for a typical planned outage was 1500.

Table C.8 summarizes information on the largest single outage experienced for one unit at those nuclear plants providing data for this topiC. The tasks performed during these large outages vary but can include steam generator replacement, core support barrel repair, recirculation system piping replacement, and refueling. The table, which contains data on the number of additional workers, the total costs, and the time involved in the responding plants' largest outages, is designed to be read in columns. As with preceding tables, Table C.8 presents numbers that are higher than actual work force peaks because some utilities included operations workers in their count of additional workers and because the numbers count each worker who came on-site, regardless of the duration of the stay. In general, the amount by which these numbers overstate actual peaks can be expected to increase with the length of the outage.

Sizeable differences also exist among nuclear plants in terms of their assessed values. Table C.9 provides a summary of data concerning current assessed values at nuclear plants grouped by the number of units at each plant. The incremental increase in the mean assessed value per unit is not linear, because the mean for one-unit plants constitutes almost 66 percent of the mean for two-unit plants, whereas the mean for two-unit plants represents only 26 percent of the mean at three-unit plants. The number of units at each plant represents those licensed by 1990; therefore, for some two- and three-unit plants, assessed values may have been considerably lower in past years depending upon the number of units actually on-line. Table C.10 depicts changes in the minimum, maximum, and mean assessed values from 1980 to 1985.

The amount of local and state taxes paid on nuclear plants by the utilities that own them also varies considerably. Table C.11 depicts information about the local and state taxes paid on nuclear plants grouped by the number of units at each plant. Once again, the incremental increase in the mean total amount of taxes paid is not linear, because the mean for one-unit plants constitutes over 65 percent of the mean for two-unit plants, whereas the mean for two-unit plants represents only 28 percent of the mean at three-unit plants. For some two- and three-unit plants, tax payments were probably lower in past years because one or more of the plants' units may not have been on-line. Table C.12 depicts changes in the minimum, maximum, and mean amounts of local and state taxes paid from 1980 to 1985. A relatively small number of utilities explicitly mentioned paying state taxes; however, where such taxes are paid, the mean payment is substantially greater than the mean for local taxes. Because a large number of utilities reported only total tax payments without specifying the jurisdictions to which the taxes are paid, it is possible that the utilities pay more state taxes than are indicated here.


C.3 Description of License Renewal


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C.3.1 Work Force and Expenditures Required for Plant Refurbishment and the License Renewal Term

License renewal for a commercial nuclear power plant will involve two time periods: the refurbishment period and the license renewal term. The length of the refurbishment period, the number of additional workers who would be on-site to perform refurbishment tasks, and the costs of refurbishment would vary among nuclear plants. The license renewal term will be 20 years, but the number of additional personnel that will be required to operate a plant will vary. This section describes the estimates of refurbishment period length, costs, and work force sizes used in the GEIS to assess the socioeconomic impacts of license renewal. Section C.3.1.1 describes the estimates of refurbishment outage length, work force, and costs provided by SEA (Appendix B; SEA 1994). Section C.3.1.2 describes the estimates of the license renewal term work force.

C.3.1.1 The Refurbishment Period

For a nuclear power plant, the refurbishment period is expected to begin several years before a unit's original operating license expires. Plant refurbishment would probably involve bringing additional workers to the site to perform certain tasks during four regularly scheduled current term outages and one major refurbishment outage. GEIS predictions are based on a bounding case (conservative) refurbishment work force scenario prepared by SEA (1994). The scenario provides refurbishment work force estimates that are expected to represent the upper bound of work force requirements for a typical plant. The impact assessment performed for the GEIS is based on the projected work force required to refurbish a PWR because that represents the worst-case situation and because all the socioeconomic case study plants are equipped with PWRs. It is assumed that refurbishment activities for multiple units at a nuclear power plant would be performed sequentially, even where two units' licenses expire in the same year (Table 2.1), because the utilities are not expected to shut down more than one unit at a time. For a PWR, the peak number of 2273 refurbishment workers is expected to be reached during the major refurbishment outage (SEA 1994). For a BWR, the peak work force (1500 persons, including refurbishment and refueling workers) would be on-site during the current term outages (Chapter 2, Appendix B).

A second work force scenario--the typical case--has been developed by SEA (1994). In this scenario, the peak work force at PWRs (874 persons) will occur during current term outages, while the BWR peak work force (1017 persons) will occur during the final refurbishment outage.

Four current term outages per unit are expected, starting 8 to 10 years before the original operating license expires (Figure C.3). Each outage would last approximately 4 months in the bounding case scenario, and it is assumed that outages would be separated from each other by 18 months of normal operation. All the current term outages would consist of refurbishment activities conducted while the reactor is shut down for refueling and routine maintenance.

During current term outages, only the refurbishment workers required for license renewal can be expected to cause new impacts, because the refueling and maintenance workers' presence is related to continued operations under the original license. However, normal plant staff and refueling and maintenance workers who are on-site during refurbishment can be expected to contribute to the overall magnitude of impacts.

One major outage is assumed in the year before the expiration of a unit's operating license, to allow performance of any remaining refurbishment tasks not completed during the previous four current term outages. This assumed refurbishment outage is expected to last 9 months and require, at its peak, approximately 2273 direct refurbishment workers at a PWR.1

C.3.1.2 The License Renewal Term

The license renewal term for a nuclear power plant would begin 10 years before the end of the initial 40-year license period (see Figure 2.3). The renewed operating license would allow 20 additional years of plant operation subsequent to the 40-year operating license. The license renewal term work force is expected to include those personnel on-site during the original operating period. In addition, it is expected that continued operations during the license renewal term could require some additional workers because of the requirement for more frequent surveillance and inspection (NUREG-1398). Past descriptions of proposed ISTM tasks indicate that these are likely to require between 20 and 60 additional workers per unit. To provide a realistic upper bound to potential population-driven impacts associated with continued plant operations, the high end of the projected range of ISTM workers (60 per unit) is used to approximate the number of additional permanent workers required for ongoing ISTM tasks during the license renewal term.2

Figure C.3 Conservative scenario refurbishment work force estimates (PWR).

As with the original operating period at all nuclear power plants, periodic refueling and maintenance would be performed during the license renewal term. In addition, each unit would undergo two 5-year ISIs and one 10-year ISI during the license renewal term. The conditions under which renewed operating licenses would be granted are expected to require more maintenance and ISI workers to perform these tasks than during the term of the original licenses. It is estimated that each of the 8 normal refueling outages that would occur during the license renewal term for both PWRs and BWRs would require approximately 30 more workers for refueling and maintenance than are currently required. In addition, it is estimated that each license renewal term refueling would cost approximately $1 million more than refueling during the original term. Further, it is projected that each of the two license renewal term 5-year ISI outages (1) would require approximately 30 more workers at a PWR and 70 more workers at a BWR and (2) would cost approximately $1.5 million more at a PWR and $3 million more at a BWR than 5-year ISIs during the original term. Finally, it is estimated that the one license renewal term 10-year ISI outage (1) would require about 50 more workers at a PWR and 110 more workers at a BWR and (2) would cost approximately $2.5 and $5 million more, respectively, than a 10-year ISI outage during current operations. The GEIS does not systematically assess the impacts associated with these periodic outage workers because such workers would not be permanent plant staff and their presence in the community is likely to be very short-lived. However, as noted earlier, evidence about past effects during such outages was collected and considered in the analysis.

C.3.2 Changes in Taxable Value of the Plant and in Tax Distributions Following Refurbishment

The taxable value of nuclear power plants is expected to increase early in the license renewal term because of improvements made during refurbishment. Subsequent depreciation is possible, although this depends on the basis of the assessed value and is likely to be gradual during the 20-year license renewal term. Furthermore, inflation would offset the effects of depreciation so that the assessed value may decrease some in real terms but would increase or remain stable in nominal terms. Overall, tax payments to local jurisdictions are expected to remain roughly similar (with some increase) to those made during current plant operations. Also, future increases in value would accompany any additional plant improvements.

Two case study sites in this evaluation illustrate past increases in the taxable assessed value of nuclear plants during normal operation. The ANO facility had a taxable assessed value of $139 million in 1980, which rose to $184 million in 1989. This increase is 3.2 percent compounded yearly and is close to the inflation rate of this time period [the implicit price deflator of the gross national product (GNP) for this 9-year period is 4.4 percent]. The D. C. Cook nuclear facility increased in taxable assessed value from $365 million to $520 million during this same time period, for a compounded growth rate of 4.0 percent. These annual increases in assessed value are the result of the continued maintenance and replacement of equipment and the general inflation level driving replacement value and income-earning ability of the plant higher.

Taxing policies of the relevant state and local governments also affect the taxable assessed value of a nuclear plant. For example, the Oconee plant is exempt from payment of property taxes on pollution control equipment installed at the plant during its operation, resulting in somewhat smaller tax payments than would otherwise be required. Although county governments often assess the taxable value of nuclear plants, their assessments are frequently based on state guidelines. Additionally, some nuclear plant sites are assessed only by the state, and the local taxing authorities apply their own millage rates to these assessments.

The ANO, Diablo Canyon, and Wolf Creek nuclear plants are typical of plants that are assessed by rules mandated by state tax departments. The local taxing authorities of Arkansas, California, and Kansas employ the unitary approach method to develop the annual taxable assessed value for nuclear plant sites. This method bases the plant valuation on a reasonable value that an investor or business would pay for the plant. The assessed value is based on the following weighted factors: the cost that the parent utility would need to acquire the plant assets, the income-earning ability of the plant, and the stock market valuation of the parent utility (with the market value of the plant apportioned from the value of the utility). The taxable assessed value determined by the state is then multiplied by the individual millage rates of the local taxing authorities to calculate the nuclear plant tax payment.

The increase in taxable assessed value resulting from refurbishment is likely to be greater than past increases in the taxable value of nuclear plants. Although capital expenditures for replacement of plant equipment and maintenance expenditures have occurred during normal operation of the plant, expenditures are likely to be made at a higher level during refurbishment. This would cause the assessed value of the plant to increase at a higher-than-normal growth rate immediately following refurbishment.

The trend in distribution of property taxes paid by the case study nuclear plants to local taxing authorities varies considerably depending on the particular circumstances affecting each plant. If the growth of the local economy is sufficiently large, as in the case of the Oconee plant, the proportion of total local property taxes contributed by the plant would probably decrease. In some cases, the millage levy for various taxing authorities changes over time. For example, the property taxes assessed on the Wolf Creek nuclear facility have been increased at a 17.7 percent annual rate by Coffey County over the past decade, whereas the Burlington School District has had its property tax assessments increase at a smaller, 8.0 percent annual rate since 1980. At the ANO site, tax rates on the nuclear plant for the county and the local school district have been lowered. This was the result of changes in state tax laws in 1986 that caused a rollback on millages, resulting in lower property taxes. This has caused the county and the local school district to receive lower property tax payments in the past 4 years and to consider general tax increases to avoid deficits. In most cases, however, periodic capital expenditures made by the case study nuclear plants have allowed their property tax payments to remain at least stable in real terms over the past decade and to increase in nominal terms.


C.4 Description of Case Study Sites


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The following sections detail impacts that occurred during construction and operation of each case study plant and project impacts for the refurbishment period and operations during the 20 years following the expiration of the initial license. The ANO case study includes a discussion of the methodology used to assess each impact category at each case study site.

C.4.1 Arkansas Nuclear One

The impact area (those places in which the most pronounced socioeconomic impacts might result from refurbishment and license renewal) for the ANO plant consists of Pope County, Arkansas, and the largest community within Pope County, Russellville. The selection of this area is based on worker residence patterns, employment, expenditures, and tax payments. Figure C.4 depicts the impact area, and Figure C.5 shows the region in which it is located.

C.4.1.1 Population

This section discusses the local population growth associated with the construction, operation, and license renewal of the ANO plant. Plant-related population growth is driven by the number of workers who migrate into nearby communities to work at a nuclear plant. These individuals and their families, and other persons and their families who move into the area to work in indirect jobs generated by the plant's presence, add to the communities' population totals. Such increases in population constitute the main driver of public service, housing, and land-use impacts, as well as many local economic impacts. Thus, to predict the socioeconomic impacts of a nuclear plant's license renewal, it is necessary to calculate projections of plant-related population growth.

Figure C.4 Socioeconomic impact area associated with Arkansas Nuclear One refurbishment: Pope County.

Figure C.5 Region surrounding the Arkansas Nuclear One nuclear plant.

The projections of population growth calculated in the GEIS are based on a number of assumptions. First, it is assumed that certain key characteristics of the refurbishment and license renewal term work forces would be analogous to those of the original construction and operating work forces, respectively. These key characteristics are (1) the percentage of the work force residing in the study area, (2) the percentage of the work force who in-migrated to the study area, (3) the percentage of in-migrants accompanied by their families, and (4) the ratio of direct to indirect jobs created by work force in-migration. Second, future population growth is represented as occurring during the peak refurbishment year and in the first year of the license renewal term. The population growth because of license renewal would result from the influx of workers over the entire license renewal period, which could last up to 30 years (10 years for refurbishment and 20 years for the license renewal term). But population growth is projected for a single year to provide a worst-case estimate for predicting population-driven impacts. Finally, population growth is projected using U.S. Census 1990 estimates of average family size for the case study states.

Given these assumptions, data concerning construction and operating work force characteristics, and estimates of refurbishment and license renewal term work force sizes, the staff has projected population growth associated with license renewal. Tables throughout this section illustrate the calculations involved in making the projections. Data used to prepare this section were obtained from Socioeconomic Impacts of Nuclear Generating Stations: Arkansas Nuclear One Station Case Study (NUREG/CR-2749, vol. 1); Environmental Assessment Proposed Rule for Nuclear Plant License Renewal (NUREG-1398); SEA refurbishment work force estimates (Appendix B); population projections by the Arkansas State Data Center; and the Arkansas Power and Light Company (AP&L;) (AP&L; 1990).

The discussion of population growth is organized into two time periods. Section C.4.1.1.1 identifies the population growth that Pope County experienced as a result of the construction and operation of ANO from 1969 to 1989. Section C.4.1.1.2 projects the population growth expected to result from ANO's refurbishment period and license renewal term operations beginning in 2014 (Unit 1) based on the growth associated with the plant's initial construction. Also, Section C.4.1.1.2 projects the population growth expected to result from ANO's license renewal term based on the growth associated with operations in the past.

C.4.1.1.1 Growth Resulting from Plant Construction and Operations

ANO's construction resulted in large population increases in Pope County (Table C.13). During the peak construction year, 1974, ANO personnel and their families who migrated to the area to work at the plant, and others who moved into the area to work in jobs generated by the plant's presence, totalled approximately 2756 persons. This influx of new residents represented 8.3 percent of Pope County's total population in 1974 (NUREG/CR-2749, vol. 1, p. 86).

Operations at the ANO plant also have resulted in large population increases in Pope County. In 1989, 2205 permanent plant staff were on-site at ANO; additional contract workers were on-site during outages, but they have not been included because their presence at the plant was temporary. Of the permanent work force, 90 percent (1985) lived in Pope County (AP&L; 1990). Based on the residential settlement pattern of ANO's 1977 work force, the staff estimated that 43.8 percent (869) of those residing in Pope County in 1989 were prior residents who obtained jobs and that 56.2 percent (1116) were workers who migrated into the area for jobs (Table C.14). Also following the pattern set during plant operations, it is estimated that 60 percent of the in-migrants (670) were accompanied by their families. Assuming the 1990 Arkansas average family size of 3.06 persons, this represents a total in-migration of 2496 residents for the county. Based on the ratio of nonplant jobs created in Pope County in 1977, it is estimated that ANO's 1989 operations created an additional 860 indirect jobs in service industries supported by the spending of ANO workers (NUREG/CR-2749, vol. 1, pp. 56-86). As a result of these indirect jobs, an estimated 454 additional workers and their families (a total of 922 persons) moved into Pope County (Table C.14). In all, approximately 3418 new residents are estimated to have moved into Pope County as a result of ANO's 1989 operations. These new residents made up about 7.7 percent of Pope County's 1989 population of 44,534 (NUREG/CR-2749, vol. 1, pp. 56-86; McFarland 1990).

C.4.1.1.2 Predicted Growth Resulting from License Renewal

As discussed in Section C.3.1, ANO's license renewal would require the completion of a number of refurbishment tasks for Units 1 and 2. Many of the refurbishment tasks are expected to be completed during scheduled refueling outages at each unit during the 10 years that precede the expiration of the initial license. However, the final refurbishment work is expected to be completed during one large refurbishment outage scheduled for each unit in the year before the unit's initial operating license expires.

Assuming the refurbishment schedule as described in Section C.3.1, the peak refurbishment year for ANO Unit 1 is expected to be 2013, and the peak refurbishment year for ANO Unit 2 is expected to be 2017. For each unit, the on-site refurbishment work force would be about the same size, and the work force would be on-site for approximately the same period of time. However, because uncertainties exist concerning the length of the outage and the size of the work force required to complete the refurbishment of a given unit, this section examines a bounding case work force scenario as described in Section C.3.1.

Given the work force scenario detailed in Section C.3.1, it is estimated that 2273 workers would be on-site to complete refurbishment of ANO 1 in 2013 and ANO 2 in 2017 (SEA 1994). Further assuming that the residential distribution of refurbishment workers would be similar to that of the 1974 ANO construction work force, it is estimated that 65 percent (1477) would reside in Pope County. Based on plant construction and operating experience, it is projected that 43.8 percent (516) of those residing in Pope County would be prior residents who obtain refurbishment jobs and that 56.2 percent (830) would be workers who migrate into the area for refurbishment jobs (Table C.15). Also following the pattern set during plant construction and operations, it is assumed that 60 percent of the in-migrants (498) would be accompanied by families. Using the Arkansas average family size of 3.06 persons, total refurbishment worker in-migration would result in 1856 new residents for the county. Based on the ratio of nonplant jobs created in Pope County in 1974, ANO's refurbishment is projected to create an additional 473 indirect jobs in service industries supported by the spending of ANO refurbishment workers. As a result of these indirect jobs, an estimated 246 additional workers and their families (a total of 499 persons) would be projected to move into Pope County (Table C.15). In all, approximately 2355 new residents would be expected to move into Pope County as a result of ANO's refurbishment under the work force scenario. That would represent a 3.7 percent increase in Pope County's projected population of 63,395 in 2014 (NUREG/ CR-2749, vol. 1, pp. 58-71, 82-83).

Once plant refurbishment is completed for ANO Units 1 and 2, the work force would consist mostly of permanent plant staff. There would be additional refurbishment/refueling workers temporarily on-site approximately every 2 years, but they would not be permanent, on-site plant staff; and many of them are expected to commute from outside the study area. It is expected that a maximum of 60 additional permanent workers per unit would be required during the license renewal term, adding 120 workers to ANO's existing work force. Assuming that the new workers' residential distribution would be the same as that of the current workers, approximately 90 percent (108) would reside in Pope County. Based on worker in-migration in 1977, it is expected that 43.8 percent (47) of those residing in Pope County would be prior residents who obtain jobs and 56.2 percent (61) would be workers who migrate into the area for jobs (Table C.16). Also following the pattern set during plant operations, 60 percent of the in-migrants (37) would be accompanied by their families. Using the Arkansas average family size of 3.06 people, total in-migration would result in 137 new residents for the county. Based on the ratio of nonplant jobs created in Pope County in 1977, ANO's license renewal term is projected to create an additional 47 indirect jobs in service industries supported by the spending of ANO workers. As a result of these indirect jobs, an estimated 25 additional workers and their families (a total of 52 persons) would be projected to move into Pope County (Table C.16). In all, approximately 189 new residents would be expected to move into Pope County as a result of ANO's license renewal term. That would represent 0.3 percent of Pope County's projected population in 2014 (NUREG/ CR-2749, vol. 1, pp. 58-71, 80-82).

C.4.1.2 Housing

The following sections examine the housing impacts that occurred in Pope County during construction and operation of the ANO plant and predict housing impacts that would result from refurbishment activities and continued operation. Possible impacts to housing include changes in the number of housing units, particularly the rate of growth of the housing stock; changes in occupancy rates; changes in the characteristics of the housing stock; and changes in rental rates or property values. The general methodology used to assess past impacts and predict refurbishment- related housing impacts is discussed in Section C.1.5. U.S. Census information; local agencies' housing data; and interviews with local government officials, planners, and realtors provided information about impacts that resulted from the construction and operation of the seven nuclear power plants used as case study sites. These sources provided information about past impacts of a known magnitude that resulted from a known number of in-migrating workers. This provided a basis of comparison when predicting future impacts of refurbishment. Refurbishment-related housing impacts are predicted by comparing refurbishment-related housing demand to the projected housing market (number of units and vacancies). Project-related housing demand is based on the assumption that some unaccompanied workers would share accommodations and is determined by the following equation:

project-related housing demand = workers with families + 0.85 x unaccompanied workers.

Projections of the number of housing units present in the study area at peak refurbishment time are based on historical growth rates of the local housing market. This assumes that average growth rates would remain constant. Non-project-related housing demand at the time of refurbishment is determined by dividing projected population by average household size. The 1990 household size is used in this calculation. Household size is expected to continue its gradual decline, thus suggesting a greater demand for housing. It is believed, however, that the housing market would adequately respond to such a gradual change; therefore, housing vacancies, even though household size decreases, would be very much the same as those predicted using the known 1990 household size.

C.4.1.2.1 Impacts from Plant Construction and Operation

The following discussion begins with a description of the housing market at the time of ANO construction and details project-related housing demand in the study area. A discussion of changes that occurred in the housing market and impacts on housing induced by plant construction follows. Finally, impacts from the operation of ANO on local housing are assessed.

Between 1970 and 1978, 4361 new housing units were added to the existing housing stock of Pope County (based on the number of electrical connections), bringing the total number of units to 14,243 (NUREG/CR-2749, vol. 1; U.S. Bureau of the Census 1972). This 44.3 percent increase represents a rate of growth consistent with census reports of a 50.8 percent increase in housing in Pope County during the 1970-80 intercensal period. Nine hundred of these new units, or 21 percent, were located in Russellville (Figure C.4).

Project-related demand for housing in Pope County has been estimated according to the number of construction workers who moved to the area (NUREG/CR-2749, vol. 1). The ANO work force peaked in 1977 at an average annual employment of 1445 persons. Project-related demand for housing in Pope County peaked in 1977 at 858 units (6.25 percent of the 1977 housing). At this time, ANO Unit 1 had begun commercial operation and ANO Unit 2 was under construction. New housing units added to the Pope County market totaled 3486 between 1969, when the project began, and 1977, when it peaked. In 1970, 391 housing units were either for rent or for sale in Pope County. In Russellville, the homeowner vacancy rate was 2.6 percent and the rental vacancy rate was 10.4 percent. Housing shortages may have occurred infrequently and lasted for only a short duration (NUREG/CR-2749, vol. 1), but the existing vacancies and the rapidly expanding housing stock for the most part kept pace with project- and non- project-related demand.

The construction of ANO was an important factor in the rapid growth of the Pope County housing stock. Other factors included non-project-related population growth resulting from economic opportunities and the expansion of Arkansas Tech University in Russellville (NUREG/CR-2749, vol. 1). Several housing projects were undertaken during and possibly in response to ANO construction. A 35-ha (87-acre), multi-unit project was begun in 1967 after the announcement of ANO. Widely held local belief is that this development was linked to ANO; however, developers and local realtors indicated that it occurred in response to general population growth that had begun to occur before ANO (NUREG/ CR-2749, vol. 1). Five new mobile home parks were established during plant construction and, along with existing mobile home parks, accommodated as many as one-third of the construction workers and their families. Another development related to construction workers' demand for rental units was the conversion of old single-family homes into apartments (NUREG/CR-2749, vol. 1).

Between 1970 and 1977, considerable construction of multifamily units occurred. In Russellville, where approximately 75 percent of the construction workers located, 50 percent of the new units were multifamily units. Although single-family housing increased 16 percent between 1970 and 1977, multifamily units increased by 42.7 percent.

During the 1970s, when the project-related demand for housing might have affected housing values, the increase in the median value and median rent of housing in Pope County was comparable to that experienced in the state. Median value rose 181 percent in Pope County and 190 percent in the state of Arkansas, whereas median rent rose 73.3 percent in Pope County and 76.4 percent in Arkansas. However, local residents and officials have indicated that during peak ANO construction years, housing values escalated to levels above national trends and rents increased in response to construction workers' demands for housing (NUREG/CR-2749, vol. 1). The addition of multifamily structures in the middle and late 1970s brought housing values and rental rates once again in line with normal inflationary increases occurring statewide.

The end of construction at ANO did not have a destabilizing effect on the housing market. The project-related demand declined gradually and was abated by the gradual in-migration of the operations work force. By 1980, when both units were in commercial operation, housing vacancy rates in Pope County were comparable to those in the state of Arkansas. The home-owner vacancy rate was 2.1 percent in Pope County and 1.6 percent in Arkansas, whereas rental vacancy rate was 8.0 percent in Pope County and 8.8 percent in Arkansas.

Operation of ANO has had little effect, if any, on housing in the area. The roads and water and gas lines associated with the plant have facilitated residential development in areas neighboring the plant but have not been as big an attraction as the aesthetic quality of Lake Dardanelle. Indirectly, the plant may have had some effects on property values because the good wages employees receive have enabled them to buy or build homes that are considered expensive relative to other homes in the area.

In summary, substantial changes occurred in the housing market, housing characteristics, and property values during the construction period of ANO. The conversion of large homes into apartments, the increase in multifamily housing, and the temporary increase in housing values and rental rates are examples of this change. ANO may have been the impetus for, or a contributing factor to, these changes; however, other industrial development and the growth of the local college also spawned some of this change. For example, the tremendous growth in the housing market had begun before the construction of ANO. Also, housing occupied by construction workers was absorbed into the market and occupied by non-project-related population. Considering all these factors, the impact on housing during ANO construction was moderate.

C.4.1.2.2 Predicted Impacts of License Renewal

Project-related population increase and the commensurate housing demand would be the cause of new housing-related impacts during refurbishment activities. A summary of recent and anticipated growth in housing is provided. This is followed by predictions of possible impacts during refurbishment and the license renewal term.

In the period 1980-90, housing in Pope County increased 23.8 percent above the 1980 level. Assuming this rate of growth will continue, there would be approximately 30,900 housing units in Pope County in 2014. Based on a projected population of 63,395 and a 1990 average household size of 2.61 persons, 24,259 housing units would be required to accommodate Pope County's 2014 population. This suggests that there will be available housing, possibly as many as 6500 units in 2014. Even if Pope County's growth were to slow considerably, e.g., to 1.6 percent annually (a rate equal to the average annual rate that occurred between 1980 and 1986), there will be about 25,650 housing units in 2014 and over 1300 vacancies.

According to the estimate of the number of refurbishment workers required for the project and based on plant construction experience, 830 workers of the total work force of 2273 are expected to migrate to Pope County for refurbishment jobs. Of these in-migrants, 498 are expected to be accompanied by families. Some doubling-up is expected to occur among the 332 unaccompanied workers, so that each unaccompanied mover would require 0.85 housing unit. In 2013, the refurbishment-related housing demand would be 780 housing units (where refurbishment-related housing demand = workers with families + 0.85 ? unaccompanied workers). In addition, numerous indirect jobs are expected to result from project workers' spending. An additional 196 indirect workers are projected to move to Pope County, bringing the total project-related housing demand in the peak year of refurbishment to 976 units.

The projected refurbishment-related housing demand is larger than the original construction-related housing demand of 858 units, but the number of housing units in the study area will have increased 86 to 117 percent under the conservative and current growth rates, respectively, between peak construction and refurbishment periods. Refurbishment-related housing demand would account for 3.8 or 3.2 percent (under the conservative and current growth rates, respectively) of the projected housing units in the study area in 2014, compared to construction- related demand in 1977 accounting for 6.25 percent of the housing units. Changes in the characteristics of the Pope County housing market that have occurred during or since plant construction should improve the accommodation of refurbishment-related workers. These changes include a greater proportion of multifamily units and the addition of mobile home parks. Some of the demand may be met by construction workers' recreational vehicles or mobile homes; this may require, however, the temporary addition or expansion of mobile home parks. However, no substantial construction of new housing units is expected to occur during refurbishment activities unless other economic and industrial growth warrants it, as was the case before and during ANO construction. Because housing demand would be small relative to the existing housing market, would not exceed projected vacancies, and would be even less than that experienced during construction, refurbishment-related housing demand is expected to have a small new impact on the study area housing market.

Housing impacts related to refueling activities and housing value and marketability that would occur as a result of continued plant operation during the license renewal term are the same as those currently being experienced (Section C.4.1.2.1). The 120 additional operations workers (60 per unit) and the commensurate housing demand would cause only small new housing impacts.

C.4.1.3 Taxes

The local impact of plant-related property taxes is presented here and in the other six case study presentations. Where information is available, the assessed valuation of the nuclear plant and the study area is presented to show the importance to the tax base from the start of construction to the current period. The impact of taxes on specific taxing authorities, such as local school districts, is presented. For these jurisdictions, the magnitude of plant-related property taxes relative to total local jurisdiction revenues is shown, again from the beginning of construction to the latest tax period in which information was available.

Each case study lists (1) the taxing authorities receiving revenues from the nuclear plants and (2) the property tax payments and tax or millage rates from the nuclear plants. At case study sites where the state assesses the value of the nuclear plant, the state tax valuation method is described. Tax reform legislation affecting the tax revenues from nuclear plants has been enacted in a few of the states where case study nuclear plant sites are located. The impact on total tax revenues and the taxing authority in general is described in sites where such legislation has been passed.

Tax and total revenue information was obtained directly from the governments that tax the case study nuclear plants. This information was obtained, where available, for the years 1980, 1985, and 1989. This longitudinal tax and revenue data allowed the evaluation of trends in nuclear plant tax revenue impacts over the past decade.

C.4.1.3.1 Impacts from Plant Construction and Operation

The jurisdictions that receive the bulk of the taxes paid by AP&L; for ANO station are Pope County and the Russellville School District. Property taxes are the principal source of revenue for Arkansas counties and municipalities. Table C.17 shows AP&L;'s annual tax payments to Pope County for ANO during the 1968-89 period (NUREG/CR-2749, vol. 1).

From 1968 through 1989, Pope County's assessed valuation increased at an annual rate of 10.1 percent in real terms. During this same time period, ANO had its assessed valuation increase at a 21.9 percent real annual rate. ANO's portion of Pope County's total assessed valuation increased sharply from 1968 to 1980, from 5.4 to 73.6 percent. Thereafter, ANO's portion has dropped to 46.2 percent in 1989. Taxes paid to Pope County increased considerably as construction of the plant progressed in the early 1970s and more than tripled between 1972 and 1976 once construction was completed.

In 1980, the state legislative passed Amendment 59, which prevented reduction in taxes on utility properties for the first 5 years after the amendment's passage. It required a gradual reduction (to occur over the succeeding 5 years) in the millage rates assessed against utility property. Because of Amendment 59, ANO's tax revenues to the county have steadily decreased from approximately $1.6 million in 1985 to $1.2 million in 1989.

The recipient of the largest tax payments within Pope County was the Russellville School District. In 1978, property within the jurisdiction of the Russellville School District was assessed at a tax rate of 50 mills, whereas the tax rate for Pope County was 9 mills (Arkansas State Department of Education 1990). However, this millage rate for the school district has been falling throughout the 1980s. In 1985 the combined millage rate for real estate and personal property components in the Russellville School District was 48.1, but by 1989 it had fallen to 22.5. During the 1980s, the assessed value of property within the district rose steadily from $176.5 million in 1980 to $275.6 million in 1985 and to $341.1 million in 1989. In real terms, the assessed valuation in the Russellville School District grew at an annual rate of 3.1 percent. Table C.18 shows the revenue impact of ANO.

To compare the amount of taxes paid to the Russellville School District in real terms during the 1980s, the assessed valuation of the school district is deflated in real terms by the GNP deflator and then multiplied by the millage rate for the school district for the year in question. The resulting estimated taxes for Russellville School District increased at a 3 percent annual real rate from 1980 through 1985 but then declined at a 15.4 percent annual real rate from 1985 to 1989. Decreased millage rates resulting from Amendment 59 are largely the cause for the decreased revenue. As tax revenues decline, the school district will likely seek a tax increase in the future.

The Russellville School District ranked 66th out of the 329 school districts in the state of Arkansas for expenses per student in 1989. This is up from a ranking of 132 in 1988. The district is currently ranked 7th out of 329 in teachers' salaries in 1989 (the comparable ranking in 1988 was 25th).

Currently, Pope County is in a period of transition from a farm-oriented community to an area of light industrial development. Industrial development has increased substantially over the last 20 years. Undoubtedly, some of the development in the county is associated with the substantial tax revenues from ANO; however, the introduction of Interstate 40 through the county has had a major impact on development in the area. Officials at Pope County and Russellville School District have indicated that improvements in the county and school district and substantially reduced tax rates were possible because of ANO.

C.4.1.3.2 Predicted Impacts of License Renewal

A new tax-related impact is expected to occur during refurbishment of ANO. This new impact does not involve capital improvements that take place during the final refurbishment outage. Rather, it results from capital improvements that are undertaken during the current term outages, which would increase the assessed value of the plant during this time and, thus, increase ANO's tax payments to local jurisdictions. The magnitude of the impact depends on AP&L;'s decision about which improvements would occur early on and which would be done during the final outage. For example, if the steam generator is replaced during a current term outage, the assessed value may increase considerably before the license renewal term begins. If steam generator replacement and other major capital improvements are not undertaken early on, the increase in assessed valuation may be only minor. The increase, in either case, is expected to cause only an small to moderate new tax impact.

During the license renewal term, the tax-related impact would be primarily the continuation of tax payments ANO is currently making to local jurisdictions. A new impact would result from the increase in tax payments from improvement made at ANO during the final refurbishment period. Thus, tax revenues would increase in absolute terms but may remain constant as a percentage of total revenues of Pope County and the Russellville School District. ANO's contribution to the county's total revenues has fallen during the past decade, from 49 percent of the total revenues in 1980 to 26 percent in 1989. The additional tax payments during the license renewal term may halt this trend. Based on current conditions, ANO tax revenues--the continuing and additional payments combined--are expected to continue to make up a large share of the total revenues. Decreased millages that have resulted from ANO's substantial tax contributions may remain, although Amendment 59 does allow for millage increases. The large tax-related impact currently being experienced would continue during the license renewal term.

C.4.1.4 Public Services

The general methodological approach used to predict future impacts is discussed in Section C.1. For most public services, impacts were calculated based on the projected number of in-migrating workers and on the projected state of the local infrastructure. The expected number of in-migrants was calculated separately for each case study site, based on the in-migration patterns observed in past studies at these same sites. Where historic data were not available, in-migration rates were estimated on the basis of comparisons with sites that were similar in terms of population density and proximity to metropolitan areas. Only in the area of transportation was in-migration considered unimportant, since all project workers (and plant-related equipment) will use local roads to access the project site.

To project impacts to local educational systems, two important factors were the number of in-migrating workers accompanied by their families and the associated family size. Assumptions about these key variables were based on past patterns observed at the case study sites. Specifically, the number of in-migrating workers expected to bring their families with them at any given site was calculated based on the percentage of past workers accompanied by their families at the same site. Refurbishment workers were assumed to follow the same pattern as past construction workers, and future operations workers were assumed to be the same as past operation workers. Average household size for each site was determined from current state- specific data. For each family at a given site, the number of children was considered to be this average family size minus the two parents. The total number of additional children of plant workers was calculated by multiplying the number of in-migrating families by the expected number of children per family. Assuming that dependent children were equally distributed between the ages of 0 and 18, 68.4 percent of the children were projected to be of school age (6-18 years). This was the number of additional children expected to be enrolled in local schools.

C.4.1.4.1 Impacts from Plant Construction and Operations

ANO has affected public services in several surrounding counties, municipalities, and school districts, but three jurisdictions have been affected more than others: the Russellville School District, Pope County, and the city of Russellville. Each entity provides different services and has been affected in varying ways by ANO, as discussed below. A few construction impacts were noticeable in education and public utilities, but projected impacts from relicensing should be less significant. Information regarding expenditures is discussed in detail in Section C.4.1.3.

Education

The Russellville School District has seen much change as a result of the ANO plant. During the 1960s, the school district was facing severe financial difficulty and overcrowding. Even though Russellville's economy was growing and the population was growing with it, the Russellville School District had problems coping with the rise in enrollment. Student/teacher ratios reached as high as 35 to 1 during the 1960s, and a tax hike was approved to fund the building of a new high school.

Local residents saw the ANO plant as a solution to the problem. Taxes from ANO in its first 3 years of construction helped to pay for the new high school, but it was not until several years later (about 1973) that the Russellville School District's situation stabilized. It was difficult to accommodate new students brought in by ANO's construction and other growth, but once the high school was complete, assimilation was easy (NUREG/CR-2749, vol. 1, p. 116).

The student/teacher ratio began falling steadily after 1968; by 1980, it had fallen to 20 to 1 and the Russellville School District teachers were being paid more than others in Arkansas. Through ANO's tax payments, the biggest impact on the system, the district was able to recruit highly qualified teachers, which played a part in encouraging further economic growth in Russellville (NUREG/CR-2749, vol. 1, p. 116).

Like the plant's construction, operations have generated economic growth in Russellville, which in turn has affected the Russellville School District. Several informants noted that firms have transferred employees to Pope County, and some new businesses have appeared because of ANO's location. For instance, one recent company move into Russellville is expected to increase school enrollment in the area by 50 to 100 students. Refueling activities also have an effect on the Russellville School District, but the concurrent rises in enrollment are minor and short-lived.

Although the positive financial impacts of the ANO plant were tremendous, the district is once again experiencing financial difficulty. The state's constitution was amended in 1980 to modify its taxation policies. The new taxation policies caused a reduction in utility tax payments beginning in 1985. This resulted in lower revenues for the school district. As ANO taxes are a major source of monies for the Russellville School District, the drop in funds has left the school system in a financial dilemma. The district will likely seek a tax increase in the future, because of program expansion during the affluent years.