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Generic Environmental Impact Statement for License Renewal of Nuclear Plants (NUREG-1437 Vol. 1, Part 2)

2. Description of Nuclear Power Plants and Sites, Plant Interaction with the Environment, and Environmental Impact Initiators Associated with License Renewal

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2.1 Introduction

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Currently, 118¹ commercial nuclear power plants are located at 74 sites in 33 of the contiguous United States. Of these, 57 sites are located east of the Mississippi River, with most of this nuclear capacity located in the Northeast (New England states, New York, and Pennsylvania); the Midwest (Illinois, Michigan, and Wisconsin); and the Southeast (the Carolinas, Georgia, Florida, and Alabama). No commercial nuclear power plants are located in Alaska or Hawaii. Approximately half of these 74 sites contain two or three nuclear units per site. Three of the 118 plants have been shut down and will be decommissioned. The plant characteristics and environmental settings for these nuclear power plant sites are provided in Appendix A. Table 2.1 provides a summary overview of the plants considered in preparing this Generic Environmental Impact Statement (GEIS).

The total capacity of generating U.S. commercial nuclear power plants is approximately 99 GW(e), with plant generating capacities ranging from 67 MW(e) to 1270 MW(e). In 1992, the U.S. electric utility industry generated about 2.8 x 10¹² kWh, 21.6 percent of which was supplied by nuclear power. The range of annual electricity production for these plants is approximately 390 x 10⁶ kWh/year to 6900 x 10⁶ kWh/year using an assumed annual capacity factor of 62 percent. It is anticipated that the electric utility industry will seek to operate many of these nuclear power plants beyond the current operating license term of 40 years. This GEIS examines how these plants and their interactions with the environment would change if such plants were allowed to operate (under the proposed license renewal regulation 10 CFR Part 54) for a maximum of 20 years past the term of the original plant license of 40 years.

The purpose of this section is to provide an orientation from the perspective of environmental considerations and assessments. Section 2.2 describes commercial nuclear power plants and their major features and plant systems. Section 2.3 describes the ways nuclear power plants interact with and affect the environment. The license renewal rule, particularly its requirements that may result in changes to nuclear plant environmental impacts, is discussed in Section 2.4. Section 2.5 reviews the generation of particular environment impacts, or precursors to such impacts, that are typical of current nuclear plant operation. It discusses the "baseline" values to be used in comparing incremental effects resulting from license renewal. Section 2.6 describes major refurbishment activities and changes that could occur at nuclear power plants during license renewal refurbishment and the extended years of operation. This section provides the background for more thorough evaluations and environmental impact assessments discussed in Sections 3 through 10.

2.2 Plant and Site Description and Plant Operation

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2.2.1 External Appearance and Setting

Nuclear power plants generally contain four main buildings or structures:

Containment or reactor building. A massive containment structure that houses the reactor vessel, the suppression pool [boiling-water reactors (BWRs) only], steam generators, pressurizer [pressurized-water reactors (PWRs) only], pumps, and associated piping. The building is generally designed to withstand such disasters as hurricanes, earthquakes, and aircraft collisions. The containment's ability to withstand such disasters, as well as the effects of accidents initiated by system failures, is the principal deterrent to release of radioactive materials to the environment.
Turbine building. Plant structures that house the steam turbine and generator, condenser, waste heat rejection system, pumps, and equipment that supports those systems.
Auxiliary buildings. Buildings that house such support systems as the ventilation system, the emergency core cooling system, the water treatment system, and the waste treatment system, along with fuel storage facilities and the plant control room.
Cooling towers. Structures designed to remove excess heat from the condenser without dumping such heat directly into water bodies.

A plant site also contains a large switchyard, where the electric voltage is stepped up and fed into the regional power distribution system, and may also include various administrative and security buildings. During the operating life of a plant, its basic appearance remains unchanged.

Typically, nuclear power plant sites and the surrounding area are flat-to-rolling countryside in wooded or agricultural areas. More than 50 percent of the sites have 80-km (50-mile) population densities of less than 200 persons per square mile, and over 80 percent have 80-km (50-mile) densities of less than 500 persons per square mile. The most notable exception is the Indian Point Station, located within 80 km (50 miles) of New York City, which has a projected 1990 population density within 80 km (50 miles) of almost 2000 persons per square mile.

Site areas range from 34 ha (84 acres) for the San Onofre Nuclear Generating Station in California to 12,000 ha (30,000 acres) for the McGuire Nuclear Station in North Carolina. As shown in Table 2.1, 28 site areas range from 200 to 400 ha (500 to 1000 acres), and an additional 12 sites are in the 400- to 800-ha (1000- to 2000-acre) range. Thus, almost 60 percent of the plant sites encompass 200 to 800 ha (500 to 2000 acres). Larger land-use areas are associated with plant cooling systems that include reservoirs, artificial lakes, and buffer areas.

2.2.2 Reactor Systems

U.S. reactors employed for domestic electric power generation are conventional (thermal) light-water reactors (LWRs), using water as moderator and coolant. The two types of LWRs are PWRs (Figure 2.1) and BWRs (Figure 2.2). Of the 118 power reactors in the United States, 80 are PWRs and 38 are BWRs.

Table 2.1 Nuclear power plant baseline information

Plant	Unit	Oper- ating license	Lic- ense exp- iration	Elec- trical rating [MW(e)]	Reactor type (a)	Steam supply system vendor (b)	Cool- ing sys- tem (c)	Cooling water source	Con- den- ser flow rate (10³ gal/ min)	Intake structure	Dis- charge struc- ture	Total site area (ac.)	Nearest city	Trans- mission corridor (ac.)	1990 population (50 miles)
Arkansas Nuclear One	1 2	1974 1978	2014 2018	850 912	PWR PWR	B&W CE	OT NDCT	Dardanelle Reservoir	765 422	3220-ft canal	520-ft canal	1,160	Little Rock, Ark.	3,700	200,000
Beaver Valley	1 2	1976 1987	2016 2027	835 836	PWR PWR	WEST WEST	NDCT NDCT	Ohio River	480 480	At river edge	At river edge	501	Pittsburgh	Uses existing corridor	3,740,000
Bellefonte Nuclear Plant	1 2	-- --	-- --	1,213 1,213	PWR PWR	B&W B&W	NDCT NDCT	Guntersville Lake	410 410	Intake channel	Submerged diffuser	1,500	Huntsville, Ala.	2,900	1,070,000
Big Rock Point Nuclear Plant	1	1962	2002	72	BWR	GE	OT	Lake Michigan	49	Underwater crib	Open discharge canal	600	Sault Ste. Marie, Canada	--	200,000
Braidwood Station	1 2	1987 1988	2027 2028	1,120 1,120	PWR PWR	WEST WEST	CCCP CCCP	Kankakee River	730 730	At lake shore	Surface flume	4,457	Joliet, Ill.	2,376	4,510,000
Browns Ferry Nuclear Power Station	1 2 3	1973 1974 1976	2013 2014 2016	1,065 1,065 1,065	BWR BWR BWR	GE GE GE	OT with towers	Tennessee River	630 630 630	In small river inlet	Diffuser pipes	840	Huntsville, Ala.	1,350	760,000
Brunswick Steam Electric Plant	1 2	1976 1974	2016 2014	821 821	BWR BWR	GE GE	OT OT	Cape Fear River	675 675	3-mile canal from river	6-mile canal to Atlantic Ocean	1,200	Wilmington, N.C.	3,500	230,000
Byron Station	1 2	1985 1987	2025 2027	1,120 1,120	PWR PWR	WEST WEST	NDCT NDCT	Rock River	632 632	On river bank	Discharge to river	1,398	Rockford, Ill.	2,000	1,000,000
Callaway Plant	1	1984	2024	1,171	PWR	WEST	NDCT	Missouri River	530	From river	To river	3,188	Columbia, Mo.	1,140	400,000
Calvert Cliffs Nuclear Power Plant	1 2	1974 1976	2014 2016	845 845	PWR PWR	CE CE	OT OT	Chesapeake Bay	1,200 1,200	560 ft from shore	850 ft from shore	1,135	Washington, D.C.	1,990	3,030,000
Catawba Nuclear Station	1 2	1985 1986	2025 2026	1,145 1,145	PWR PWR	WEST WEST	MDCT MDCT	Lake Wylie	660 660	Skimmer wall	Cove of lake	391	Charlotte, N.C.	584	1,590,000
Clinton Power Station	1	1987	2027	933	BWR	GE	OT	Salt Creek	569	Shoreline of creek	3-mile flume	14,090	Decatur, Ill.	906	730,000
Comanche Peak Steam Electric Station	1 2	1989 --	2029 --	1,150 1,150	PWR PWR	WEST WEST	OT OT	Squaw Creek Reservoir	1,030 1,030	Shore of reservoir	Canal to reservoir	7,669	Ft. Worth, Tex.	458	1,130,000
Donald C. Cook Nuclear Power Plant	1 2	1974 1977	2014 2017	1,030 1,100	PWR PWR	WEST WEST	OT OT	Lake Michigan	800 800	2,250 ft from shore	1,250 ft from shore	650	South Bend, Ind.	3,300	1,250,000
Cooper Nuclear Station	--	1974	2014	778	BWR	GE	OT	Missouri River	631	At shoreline	At shoreline	1,090	Lincoln, Neb.	6,862	180,000
Crystal River Nuclear Plant	3	1977	2017	825	PWR	B&W	OT	Gulf of Mexico	680	16,000 ft from shore	13,000 ft canal	4,738	Gainesville, Fla.	2,140	440,000
Davis-Besse Nuclear Power Station	1	1977	2017	906	PWR	B&W	NDCT	Lake Erie	480	Submerged 3,000 ft off shore	Submerged 900 ft off shore	954	Toledo, Ohio	1,800	1,920,000
Diablo Canyon Nuclear Power Plant	1 2	1984 1985	2024 2025	1,086 1,119	PWR PWR	WEST WEST	OT OT	Pacific Ocean	863 863	At shore with break wall	Surface to ocean	750	Santa Barbara, Calif.	6,000	300,000
Dresden Nuclear Power Station	2 3	1969 1971	2010 2011	794 794	BWR BWR	GE GE	Cooling lake and spray canal	Kankakee River	471 471	Canal from Kankakee River	Cooling lake to Illinois River	953 + 1,274 cooling pond	Joliet, Ill.	2,250	6,820,000
Duane Arnold Energy Center	1	1974	2014	538	BWR	GE	MDCT	Cedar River	290	Shoreline	Canal to shoreline	500	Cedar Rapids, Iowa	1,160	620,000
Joseph M. Farley Nuclear Plant	1 2	1977 1981	2017 2021	829 829	PWR PWR	WEST WEST	MDCT MDCT	Chattahoochee River	635 635	River to storage pond	At river bank	1,850	Columbus, Ga.	5,300	390,000
Enrico Fermi Atomic Power Plant	2	1985	2025	1,093	BWR	GE	NDCT	Lake Erie	837	At edge of lake	Pond to lake	1,120	Detroit	180	5,370,000
James A. FitzPatrick Nuclear Power Plant	--	1974	2014	816	BWR	GE	OT	Lake Ontario	353	From lake	To lake	702	Syracuse, N.Y.	1,000	820,000
Fort Calhoun Station	1	1973	2013	478	PWR	CE	OT	Missouri River	360	At shore	At shore	660	Omaha, Neb.	186	770,000
Robert Emmett Ginna Nuclear Power Plant	1	1969	2009	470	PWR	WEST	OT	Lake Ontario	356	Lake bottom	Open canal	338	Rochester, N.Y.	280	1,140,000
Grand Gulf Nuclear Station	1	1984	2024	1,250	BWR	GE	NDCT	Mississippi River	572	Collector wells	Discharge via barge slip	2,100	Jackson, Miss.	2,300	350,000
Haddam Neck (Connecticut Yankee)	--	1967	2007	582	PWR	WEST	OT	Connecticut River	372	Shoreline	Canal to river	525	Meridian, Conn.	985	3,630,000
Shearon Harris Nuclear Power Plant	1	1987	2027	900	PWR	WEST	NDCT	Buckhorn Creek	483	Reservoir on creek	To reservoir	10,744	Raleigh, N.C.	3,500	1,430,000
Edwin I. Hatch Nuclear Plant	1 2	1974 1978	2014 2018	776 784	BWR BWR	GE GE	MDCT	Altamaha River	556	Edge of river	120 ft from shore	2,244	Savannah, Ga.	4,691	330,000
Hope Creek Generating Station	1	1986	2026	1,067	BWR	GE	NDCT	Delaware River	552	Edge of river	10 ft from shore	740	Wilmington, Del.	912	4,850,000
Indian Point Station	2 3	1973 1976	2013 2016	873 965	PWR PWR	WEST WEST	OT	Hudson River	840	At river bank	Channel to river	239	White Plains, N.Y.	10	15,190,000
Kewaunee Nuclear Power Plant	--	1973	2013	535	PWR	WEST	OT	Lake Michigan	420	1,750 ft from shore	At shoreline	908	Green Bay, Wisc.	1,066	640,000
La Salle County Station	1 2	1982 1984	2022 2024	1,078 1,078	BWR BWR	GE GE	Cooling pond	Illinois River	645	From cooling pond	To cooling pond	3,060	Joliet, Ill.	2,278	1,160,000
Limerick Generating Station	1 2	1985 1990	2025 2030	1,055 1,055	BWR BWR	GE GE	NDCT NDCT	Schuylkill River	450	From river	To river	595	Reading, Pa.	7	6,970,000
Maine Yankee Atomic Plant	--	1973	2013	825	PWR	CE	OT	Back River	426	River bank	Bay on Back River	740	Portland, Maine	220	640,000
William B. McGuire Nuclear Station	1 2	1981 1983	2021 2023	1,180 1,180	PWR PWR	WEST WEST	OT	Lake Norman	675	Submerged and surface at shoreline	2,000-ft canal discharge	30,000	Charlotte, N.C.	62	1,750,000
Millstone Nuclear Power Plant	1 2 3	1970 1975 1986	2010 2015 2026	660 870 1,154	BWR PWR PWR	GE CE WEST	OT OT OT	Long Island Sound	420 523 907	Niantic Bay	Via holding ponds	500	New Haven, Conn.	927	2,760,000
Monticello Nuclear Generating Plant	--	1970	2010	545	BWR	GE	OT with towers	Mississippi River	280	Canal	Canal	1,325	Minneapolis, Minn.	1,454	2,170,000
North Anna Power Station	1 2	1978 1980	2018 2020	907 907	PWR PWR	WEST WEST	OT	Lake Anna	940	Lake shore	Via cooling pond	18,643	Richmond, Va.	3,528	1,150,000
Nine Mile Point Nuclear Station	1 2	1968 1987	2008 2027	620 1,080	BWR BWR	GE GE	OT NDCT	Lake Ontario	250 580	Pipelines 1,000 ft off shore	Diffuser pipe	900	Syracuse, N.Y.	1,640	820,000
Oconee Nuclear Station	1 2 3	1973 1973 1974	2013 2013 2014	887 887 887	PWR PWR PWR	B&W B&W B&W	OT	Lake Keowee	680	710-ft deepskimmer wall	765 ft deep	510	Greenville, S.C.	7,800	990,000
Oyster Creek Generating Station	1	1969	2009	650	BWR	GE	OT	Barnegat Bay	460	Forked River from bay	Forked River to bay	1,416	Atlantic City, N.J.	322	4,030,000
Palisades Nuclear Plant	1	1972	2012	805	PWR	CE	MDCT	Lake Michigan	405	Crib 3,300 ft from shore	108-ft canal	487	Kalamazoo, Mich.	2,250	1,170,000
Palo Verde Generating Station	1 2 3	1985 1986 1987	2025 2026 2027	1,270 1,270 1,270	PWR PWR PWR	CE CE CE	MDCT	Phoenix City Sewage Treatment Plant	560	35-mile pipe	Evaporation ponds	4,050	Phoenix, Ariz.	16,600	1,180,000
Peach Bottom Atomic Power Station	2 3	1973 1974	2013 2014	1,065 1,065	BWR BWR	GE GE	OT with towers	Conowingo Pond	750	Small intake pond	5,000-ft canal	620	Lancaster, Pa.	1,030	4,660,000
Perry Nuclear Power Station	1	1986	2026	1,205	BWR	GE	NDCT	Lake Erie	545	Multiport 2,250 ft off shore	Diffuser 1,650 ft off shore	1,100	Euclid, Ohio	1,500	2,480,000
Pilgrim Nuclear Power Station	1	1972	2012	655	BWR	GE	OT	Cape Cod Bay	311	Edge of bay	850-ft canal	517	Brockton, Mass.	174	4,440,000
Point Beach Nuclear Plant	1 2	1970 1972	2010 2012	497 497	PWR PWR	WEST WEST	OT	Lake Michigan	350	1,750 ft from shore	Flumes 150 ft from shore	2,065	Green Bay, Wisc.	3,321	610,000
Prairie Island Nuclear Generating Plant	1 2	1973 1974	2013 2014	530 530	PWR PWR	WEST WEST	MDCT or OT	Mississippi River	294	Short canal	Basin to towers and/or river	560	Minneapolis, Minn.	973	2,290,000
Quad-Cities Station	1 2	1972 1972	2012 2012	789 789	BWR BWR	GE GE	OT	Mississippi River	471	Edge of river	14,000-ft spray canal	784	Davenport, Iowa	1,400	740,000
Rancho Seco Nuclear Station	1	1974	2014	918	PWR	B&W	NDCT	Folsom Canal	446	3.5-mile pipe	1.5-mile pipe to reservoir	2,480	Sacramento, Calif.	870	2,010,000
River Bend Station	1	1985	2025	936	BWR	GE	MDCT	Mississippi River	508	At river bank	Into river	3,342	Baton Rouge, La.	1,014	800,000
H. B. Robinson Plant	2	1970	2010	700	PWR	WEST	OT	Lake Robinson	482	Edge of lake	4.2-mile canal	5,000	Columbia, S.C.	1,024	740,000
Salem Nuclear Generating Station	1 2	1976 1981	2016 2021	1,115 1,115	PWR PWR	WEST WEST	OT	Delaware River	1,100	Edge of river	500 ft into river	700	Wilmington, Del.	3,900	4,810,000
San Onofre Nuclear Generating Station	1 2 3	1967 1982 1983	2007 2022 2023	436 1,070 1,080	PWR PWR PWR	WEST CE CE	OT	Pacific Ocean	341 797 797	3,200 to 3,400 ft off shore	2,600 to 8,500 ft from shore	84	Oceanside, Calif.	1,100	5,430,000
Seabrook Station	1	1990	2032	1,198	PWR	WEST	OT	Atlantic Ocean	399	7,000 ft off shore	5,500 ft off shore	896	Lawrence, Mass.	1,545	3,760,000
Sequoyah Nuclear Plant	1 2	1980 1981	2020 2021	1,148 1,148	PWR PWR	WEST WEST	OT and/or NDCT	Chickamauga Lake	522	From lake	To lake	525	Chattanooga, Tenn.	1,260	930,000
Shoreham Nuclear Power Station	--	--	----	819	BWR	GE	OT	Long Island Sound	574	Intake canal	Diffuser system	499	New Haven, Conn.	39	5,390,000
South Texas Project	1 2	1988 1989	2028 2029	1,250 1,250	PWR PWR	WEST WEST	CCCP	Colorado River	907	Bank of river	Bank of river	12,350	Galveston, Texas	4,773	270,000
St. Lucie Plant	1 2	1976 1983	2016 2023	830 830	PWR PWR	CE CE	OT	Atlantic Ocean	491	1,200 ft off shore	>1,200 ft off shore	1,132	West Palm Beach, Fla.	760	690,000
Virgil C. Summer Nuclear Station	1	1982	2022	900	PWR	WEST	OT	Lake Monticello	485	Intake at shoreline	Discharge pond to lake	2,200	Columbia, S.C.	1,576	910,000
Surry Power Station	1 2	1972 1973	2012 2013	788 788	PWR PWR	WEST WEST	OT	James River	840	1.7-mile canal	2900-ft canal	840	Newport News, Va.	4,420	1,900,000
Susquehanna Steam Electric Station	1 2	1982 1984	2022 2024	1,050 1,050	BWR BWR	GE GE	NDCT	Susquehanna River	448	River bank	240 ft from bank	1,075	Wilkes-Barre, Pa.	1,800	1,500,000
Three Mile Island Nuclear Station	1	1974	2014	819	PWR	B&W	NDCT	Susquehanna River	430	At river bank	At shoreline	472	Harrisburg, Pa.	1,790	2,170,000
Trojan Nuclear Plant	1	1975	2015	1,130	PWR	WEST	NDCT	Columbia River	429	At river bank	350 ft from bank	635	Portland, Ore.	1,260	1,850,000
Turkey Point Plant	3 4	1972 1973	2012 2013	693 693	PWR PWR	WEST WEST	Closed- cycle- canal	Biscane Bay	624	Intake canal and barge canal	Canal system	24,000	Miami	817	2,700,000
Vermont Yankee Nuclear Power Station	1	1973	2013	540	BWR	GE	OT and towers	Connecticut River	366	Edge of river	Edge of river	125	Holyoke, Mass.	1,550	1,510,000
Vogtle Electric Generating Plant	1 2	1987 1989	2027 2029	1,101 1,160	PWR PWR	WEST WEST	NDCT	Savannah River	510	At river bank	Near shoreline	3,169	Augusta, Ga.	--	630,000
Waterford Steam Electric Station	3	1985	2025	1,104	PWR	CE	OT	Mississippi River	975	At river bank	At river bank	3,561	New Orleans	280	1,970,000
Watts Bar Nuclear Plant	1 2	-- --	-- --	1,170 1,170	PWR PWR	WEST WEST	NDCT NDCT	Chickamauga Lake	410	At lake bank	Holding pond to lake	1,170	Chattanooga, Tenn.	3,165	950,000
Washington Nuclear Project (WNP)	2	1984	2024	1,100	BWR	GE	MDCT	Columbia River	550	Offshore	175 ft from shoreline	Depart-ment of Energy, Hanford Reservation	Richland, Wash.	Hanford Reservation	280,000
Wolf Creek Generation Station	1	1985	2025	1,170	PWR	WEST	CCCP	Wolf Creek	500	Cooling lake	Cooling lake to embayment	9,818	Topeka, Kansas	2,900	200,000
Yankee Nuclear Power Station	1	1960	2000	175	PWR	WEST	OT	Deerfield River	140	Sherman Pond, 90 ft below surface	Sherman Pond	2,000	Pittsfield, Mass.	--	1,720,000
Zion Nuclear Plant	1 2	1973 1973	2013 2013	1,040 1,040	PWR PWR	WEST WEST	OT OT	Lake Michigan	735	2600 ft off shore	760 ft off shore	250	Waukegan, Ill.	145	7,480,000

^aPWR = pressurized-water reactor; BWR = boiling-water reactor.

^bB-W = Babcock and Wilcox; GE = General Electric; WEST = Westinghouse; C-E = Combustion-Engineering.

^cOT = once through; NDCT = natural draft cooling tower; MDCT = mechanical draft cooling tower; CCCP = closed cycle cooling pond, lake, or reservoir.

Figure 2.1 Pressurized-water-reactor power generation system.

Figure 2.2 Boiling-water-reactor generating system.

In the PWR, reactor heat is transferred from the primary coolant to a secondary coolant loop that is at a lower pressure, allowing steam to be generated in the steam generator. The steam then flows to a turbine for power production. In contrast, the BWR generates steam directly within the reactor core, which passes through moisture separators and steam dryers and then flows to the turbine.

All domestic power reactors employ a containment structure as a major safety feature to prevent the release of radionuclides in the event of an accident. PWRs employ three types of containments: (1) large, dry containments; (2) subatmospheric containments; and (3) ice condenser containments. Of the 80 U.S. PWRs, 65 have large, dry containments; 7 have subatmospheric containments; and 8 have ice condenser containments. BWR containments typically are composed of a suppression pool and dry well. Three types of BWR containments (Mark I, Mark II, and Mark III) have evolved. There are 24 Mark I, 10 Mark II, and 4 Mark III containment designs in the United States.

NUREG/CR-5640 provides a comprehensive overview and description of U.S. commercial nuclear power plant systems.

2.2.3 Cooling and Auxiliary Water Systems

The predominant water use at a nuclear power plant is for removing excess heat generated in the reactor by condenser cooling. The quantity of water used for condenser cooling is a function of several factors, including the capacity rating of the plant and the increase in cooling water temperature from the intake to the discharge. The larger the plant, the greater the quantity of waste heat to be dissipated, and the greater the quantity of cooling water required.

In addition to removing heat from the reactor, cooling water is also provided to the service water system and to the auxiliary cooling water system. The volume of water required for these systems for once-through cooling is usually less than 15 percent of the volume required for condenser cooling. In closed-cycle cooling, the additional water needed is usually less than 5 percent of that needed for condenser cooling.

Of the 118 nuclear reactors, 48 use closed-cycle cooling systems (see Table 2.2, which groups the 74 plant sites into three broad categories according to environment). Most closed-cycle systems use cooling towers. Some closed-cycle system units use a cooling lake or canals for transferring heat to the atmosphere. Once-through cooling systems are used at 70 units. A few of these systems are augmented with helper cooling towers to reduce the temperature of the effluent released to the adjacent body of water.

In closed-cycle systems, the cooling water is recirculated through the condenser after the waste heat is removed by dissipation to the atmosphere, usually by circulating the water through large cooling towers constructed for that purpose. Several types of closed-cycle cooling systems are currently used by the nuclear power industry. Recirculating cooling systems consist of either natural draft or mechanical draft cooling towers, cooling ponds, cooling lakes, or cooling canals. Because the predominant cooling mechanism associated with closed-cycle systems is evaporation, most of the water used for cooling is consumed and is not returned to a water source.

Table 2.2 Types of cooling systems used at nuclear power sites
Plant site	State	Cooling system^a
Coastal or estuarine environment
Diablo Canyon Nuclear Power Plant	California	Once through
San Onofre Nuclear Generating Station	California	Once through
Millstone Nuclear Power Plant	Connecticut	Once through
Crystal River Nuclear Plant	Florida	Once through
St. Lucie Plant	Florida	Once through
Turkey Point Plant	Florida	Cooling canal
Maine Yankee Atomic River Plant	Maine	Once through
Calvert Cliffs Nuclear Power Plant	Maryland	Once through
Pilgrim Nuclear Power Plant	Massachusetts	Once through
Seabrook Station	New Hampshire	Once through
Hope Creek Generating Station	New Jersey	Towers (natural draft)
Oyster Creek Generating Station	New Jersey	Once through
Salem Nuclear Generating Station	New Jersey	Once through
Indian Point Station	New York	Once through
Shoreham Nuclear Power Station	New York	Once through
Brunswick Steam Electric Plant	North Carolina	Once through
South Texas Project	Texas	Cooling pond
Surry Power Station	Virginia	Once through
Great Lakes shoreline environment
Zion Nuclear Plant	Illinois	Once through
Big Rock Point Nuclear Plant	Michigan	Once through
Donald C. Cook Nuclear Power Plant	Michigan	Once through
Enrico Fermi Atomic Power Plant	Michigan	Towers (natural draft) and pond
Palisades Nuclear Plant	Michigan	Towers (mechanical draft)
James A. FitzPatrick Nuclear Power Plant	New York	Once through
Robert Emmett Ginna Nuclear Power Plant	New York	Once through
Nine Mile Point Nuclear Station	New York	Once through and towers
Davis-Besse Nuclear Power Station	Ohio	Towers (natural draft)
Perry Nuclear Power Station	Ohio	Towers (natural draft)
Kewaunee Nuclear Power Plant	Wisconsin	Once through
Point Beach Nuclear Plant	Wisconsin	Once through
Freshwater riverine or impoundment environment
Bellefonte Nuclear Plant	Alabama	Towers (natural draft)
Browns Ferry Nuclear Power Plant	Alabama	Once through and helper towers
Joseph M. Farley Nuclear Plant	Alabama	Towers (mechanical draft)
Palo Verde Generating Station	Arizona	Towers (mechanical draft)
Arkansas Nuclear One	Arkansas	Once through and towers
Rancho Seco Nuclear Station	California	Towers (natural draft)
Haddam Neck Plant (Connecticut Yankee)	Connecticut	Once through
Edwin I. Hatch Nuclear Plant	Georgia	Towers (mechanical draft)
Vogtle Electric Generating Plant	Georgia	Towers (natural draft)
Braidwood Station	Illinois	Cooling pond
Byron Station	Illinois	Towers (natural draft)
Clinton Power Station	Illinois	Cooling pond
Dresden Nuclear Power Station	Illinois	Spray canal and cooling pond
La Salle Country Station	Illinois	Cooling pond
Quad Cities Station	Illinois	Once through
Duane Arnold Energy Center	Iowa	Towers (mechanical draft)
Wolf Creek Generation Station	Kansas	Cooling pond
River Bend Station	Louisiana	Towers (mechanical draft)
Waterford Steam Electric Station	Louisiana	Once through
Yankee Nuclear Power Station	Massachusetts	Once through
Monticello Nuclear Generating Plant	Minnesota	Variable (mechanical draft)
Prairie Island Nuclear Generating Plant	Minnesota	Variable (mechanical draft)
Grand Gulf Nuclear Station	Mississippi	Towers (natural draft)
Callaway Plant	Missouri	Towers (natural draft)
Cooper Nuclear Station	Nebraska	Once through
Fort Calhoun Station	Nebraska	Once through
Shearon Harris Nuclear Power Plant	North Carolina	Towers (natural draft)
William B. McGuire Nuclear Station	North Carolina	Once through
Trojan Nuclear Plant	Oregon	Towers (natural draft)
Beaver Valley	Pennsylvania	Variable (natural draft)
Limerick Generating Station	Pennsylvania	Towers (natural draft)
Peach Bottom Atomic Power Station	Pennsylvania	Once through and towers (mechanical draft)
Susquehanna Steam Plant Station	Pennsylvania	Towers (natural draft)
Three Mile Island Nuclear Station	Pennsylvania	Towers (natural draft)
Catawba Nuclear Station	South Carolina	Towers (mechanical draft)
Oconee Nuclear Station	South Carolina	Once through
H. B. Robinson Plant	South Carolina	Cooling pond
Virgil C. Summer Nuclear Station	South Carolina	Cooling pond
Sequoyah Nuclear Plant	Tennessee	Variable (natural draft)
Watts Bar Nuclear Plant	Tennessee	Towers (natural draft)
Comanche Peak	Texas	Once through
Vermont Yankee Nuclear Power Station	Vermont	Once through and helper towers
North Anna Power Station	Virginia	Once through
Washington Nuclear Project-2	Washington	Towers (mechanical draft)

^a Of the 48 plants with closed-cycle cooling systems, 15 use mechanical draft cooling towers, 25 use natural draft cooling towers, 4 use a canal system, and 4 use a cooling lake. Of the 70 plants with once-through cooling systems, 24 discharge to a river, 11 discharge to the Great Lakes, 19 discharge to the ocean or an estuary, and 16 discharge to a reservoir or lake. Five of the once-through plants can also switch to cooling towers.

In a once-through cooling system, circulating water for condenser cooling is drawn from an adjacent body of water, such as a lake or river, passed through the condenser tubes, and returned at a higher temperature to the adjacent body of water. The waste heat is dissipated to the atmosphere mainly by evaporation from the water body and, to a much smaller extent, by conduction, convection, and thermal radiation loss.

All sites with two or three reactors use the same cooling system for all reactors, except for two sites: Arkansas Nuclear One in Arkansas and Nine Mile Point in New York. These two sites use once-through cooling for one unit and closed-cycle for the other.

For both once-through and closed-cycle cooling systems, the water intake and discharge structures are of various configurations to accommodate the source water body and to minimize impact to the aquatic ecosystem. The intake structures are generally located along the shoreline of the body of water and are equipped with fish protection devices (ORNL/TM-6472). The discharge structures are generally of the jet or diffuser outfall type and are designed to promote rapid mixing of the effluent stream with the receiving body of water. Biocides and other chemicals used for corrosion control and for other water treatment purposes are mixed with the condenser cooling water and discharged from the system.

In addition to surface water sources, some nuclear power plants use groundwater as a source for service water, makeup water, or potable water. Other plants operate dewatering systems to intentionally lower the groundwater table, either by pumping or by using a system of drains, in the vicinity of building foundations.

2.2.4 Radioactive Waste Treatment Systems

During the fission process, a large inventory of radioactive fission products builds up within the fuel. Virtually all of the fission products are contained within the fuel pellets. The fuel pellets are enclosed in hollow metal rods (cladding), which are hermetically sealed to further prevent the release of fission products. However, a small fraction of the fission products escapes the fuel rods and contaminates the reactor coolant. The primary system coolant also has radioactive contaminants as a result of neutron activation. The radioactivity in the reactor coolant is the source of gaseous, liquid, and solid radioactive wastes at LWRs.

The following sections describe the basic design and operation of PWR and BWR radioactive-waste-treatment systems.

2.2.4.1 Gaseous Radioactive Waste

For BWRs, the sources of routine radioactive gaseous emissions to the atmosphere are the air ejector, which removes noncondensable gases from the coolant to improve power conversion efficiency, and gaseous and vapor leakages, which, after monitoring and filtering, are discharged to the atmosphere via the building ventilation systems.

The off-gas treatment system collects noncondensable gases and vapors that are exhausted at the condenser via the air ejectors. These off-gases are processed through a series of delay systems and filters to remove airborne radioactive particulates and halogens, thereby minimizing the quantities of the radionuclides that might be released. Building ventilation system exhausts are another source of gaseous radioactive wastes for BWRs.

PWRs have three primary sources of gaseous radioactive emissions:

discharges from the gaseous waste management system;
discharges associated with the exhaust of noncondensable gases at the main condenser if a primary-to-secondary system leak exists; and
radioactive gaseous discharges from the building ventilation exhaust, including the reactor building, reactor auxiliary building, and fuel-handling building.

The gaseous waste management system collects fission products, mainly noble gases, that accumulate in the primary coolant. A small portion of the primary coolant flow is continually diverted to the primary coolant purification, volume, and chemical control system to remove contaminants and adjust the coolant chemistry and volume. During this process, noncondensable gases are stripped and routed to the gaseous waste management system, which consists of a series of gas storage tanks. The storage tanks allow the short-half-life radioactive gases to decay, leaving only relatively small quantities of long-half-life radionuclides to be released to the atmosphere. Some PWRs are using charcoal delay systems rather than gas storage tanks (e.g., Seabrook).

2.2.4.2 Liquid Radioactive Waste

Radionuclide contaminants in the primary coolant are the source of liquid radioactive waste in LWRs. The specific sources of these wastes, the modes of collection and treatment, and the types and quantities of liquid radioactive wastes released to the environment are in many respects similar in BWRs and PWRs. Accordingly, the following discussion applies to both BWRs and PWRs, with distinctions made only where important differences exist.

Liquid wastes resulting from LWR operation may be placed into the following categories: clean wastes, dirty wastes, detergent wastes, turbine building floor-drain water, and steam generator blowdown (PWRs only). Clean wastes include all liquid wastes with a normally low conductivity and variable radioactivity content. They consist of reactor grade water, which is amenable to processing for reuse as reactor coolant makeup water. Clean wastes are collected from equipment leaks and drains, certain valve and pump seal leaks not collected in the reactor coolant drain tank, and other aerated leakage sources. These wastes also include primary coolant. Dirty wastes include all liquid wastes with a moderate conductivity and variable radioactivity content that, after processing, may be used as reactor coolant makeup water. Dirty wastes consist of liquid wastes collected in the containment building sump, auxiliary building sumps and drains, laboratory drains, sample station drains, and other miscellaneous floor drains. Detergent wastes consist principally of laundry wastes and personnel and equipment decontamination wastes and normally have a low radioactivity content. Turbine building floor-drain wastes usually have high conductivity and low radionuclide content. In PWRs, steam generator blowdown can have relatively high concentrations of radionuclides depending on the amount of primary-to-secondary leakage. Following processing, the water may be reused or discharged.

Each of these sources of liquid wastes receives varying degrees and types of treatment before storage for reuse or discharge to the environment under the site National Pollutant Discharge Elimination System (NPDES) permit. The extent and types of treatment depend on the chemical and radionuclide content of the waste; to increase the efficiency of waste processing, wastes of similar characteristics are batched before treatment.

The degree of processing, storing, and recycling of liquid radioactive waste has steadily increased among operating plants. For example, extensive recycling of steam generator blowdown in PWRs is now the typical mode of operation, and secondary side wastewater is routinely treated. In addition, the plant systems used to process wastes are often augmented with the use of commercial mobile processing systems. As a result, radionuclide releases in liquid effluent from LWRs have generally declined or remained the same.

2.2.4.3 Solid Radioactive Waste

Solid low-level radioactive waste (LLW) from nuclear power plants is generated by removal of radionuclides from liquid waste streams, filtration of airborne gaseous emissions, and removal of contaminated material from various reactor areas. Liquid contaminated with radionuclides comes from primary and secondary coolant systems, spent-fuel pools, decontaminated wastewater, and laboratory operations. Concentrated liquids, filter sludges, waste oils, and other liquid sources are segregated by type, flushed to storage tanks, stabilized for packaging in a solid form by dewatering, slurried into 55-gal steel drums, and stored on-site in shielded Butler-style buildings or other facilities until suitable for off-site disposal (NUREG/CR-2907). These buildings usually contain volume reduction facilities to reduce the volume of LLW requiring off-site disposal (EPRI NP-5526-V1).

High-efficiency particulate filters are used to remove radioactive material from gaseous plant effluents. These filters are compacted in volume reduction facilities that have volume reduction equipment and are disposed of as solid wastes.

Solid LLW consists of contaminated protective clothing, paper, rags, glassware, compactible and noncompactible trash, and non-fuel-irradiated reactor components and equipment. Most of this waste comes from plant modifications and routine maintenance activities. Additional sources include tools and other material exposed to the reactor environment (EPRI-NP-5526-V1; EPRI NP-5526-V2). Before disposal, compactible trash is usually taken to on- or off-site VR facilities. Compacted dry active waste is the largest single form of LLW disposed from nuclear plants, comprising one-half and one-third of total average annual volumes from PWRs and BWRs, respectively (EPRI NP-5526V1).

Volume reduction efforts have been undertaken in response to increased disposal costs and the passage of the 1980 Low Level Radioactive Waste Policy Act and the 1985 Low Level Radioactive Waste Policy Amendments Act (LLRWPAA) (Pub. L. 96-573; Pub. L. 99-240), which require LLW disposal allocation systems for nuclear plants (see Section 6.3). Volume reduction is performed both on- and off-site. The most common on-site volume reduction techniques are high-pressure compacting of waste drums, dewatering and evaporating wet wastes, monitoring waste streams to segregate wastes, minimizing the exposure of routine equipment to contamination, and decontaminating and sorting radioactive or nonradioactive batches before off-site shipment. Off-site waste management vendors compact compactible wastes at ultra-high pressure (supercompaction); incinerate dry active waste; separate and incinerate oily, organic wastes; solidify the ash; and occasionally undertake waste crystallization and asphalt solidification of resins and sludges (EPRI NP-6163; EPRI NP-5526-V1; EPRI NP-5526-V2; DOE/RW-0220).

Spent fuel contains fission products and actinides produced when nuclear fuel is irradiated in reactors, as well as any unburned, unfissioned nuclear fuel remaining after the fuel rods have been removed from the reactor core. After spent fuel is removed from reactors, it is stored in racks placed in storage pools to isolate it from the environment. Delays in siting an interim monitored retrievable storage (MRS) facility or permanent repository, coupled with rapidly filling spent-fuel pools, have led utilities to seek other storage solutions, including expansion of existing pools, aboveground dry storage, longer fuel burnup, and shipment of spent fuel to other plants (Gerstberger 1987; DOE RW-0220).

Pool storage has been increased through (1) enlarging the capacity of spent-fuel racks, (2) adding racks to existing pool arrays ("dense-racking"), (3) reconfiguring spent fuel with neutron-absorbing racks, and (4) employing double-tiered storage (installing a second tier of racks above those on the pool floor).

Efforts are under way to develop dry storage technologies; these ude casks, silos, dry wells, and vaults (DOE December 1989). Dry storage facilities are simpler and more readily maintained than fuel pools. They are growing in favor because they offer a more stable means of storage and require relatively little land area (less than 0.2 ha--half an acre in most cases) (Johnson 1989). Dry storage is currently in use at about 5 percent of the sites.

2.2.4.4 Transportation of Radioactive Materials

There are four types of radioactive material shipments to and from nuclear plants: (1) routine and refurbishment-generated LLW transported from plants to disposal facilities, (2) routine LLW shipped to off-site facilities for volume reduction, (3) nuclear fuel shipments from fuel fabrication facilities to plants for loading into reactors (generally occurring on a 12- to 18-month cycle), and (4) spent-fuel shipments to other nuclear power plants with available storage space (an infrequent occurrence usually limited to plants owned by the same utility).

Workers and others are protected from exposure during radioactive material transport by the waste packaging. Operational restrictions on transport vehicles, ambient radiation monitoring, imposition of licensing standards (which ensure proper waste certification by testing and analysis of packages), waste solidification, and training of emergency personnel to respond to mishaps are also used (NUREG-0170; O'Sullivan 1988). Additional regulations may be imposed by states and communities along transportation corridors (Pub. L. 93-633; OTA-SET-304).

A typical PWR makes approximately 44 shipments of LLW per year; an average BWR makes 104 shipments per year (EPRI NP-5983). Most of this LLW is Class A waste packaged in 55-gal drums or other "Type A" containers and shipped to disposal facilities on flatbed trucks (DOE August 1989). (A "Type A" container permits no release of radioactive material under normal transportation conditions and must maintain sufficient shielding to limit radiation exposure to handling personnel). LLW shipments require manifests that describe the contents of the packages to permit inspection by state, local, and facility personnel and to ensure that the waste is suitable for a particular disposal facility (NUREG-0945).

Currently, the only spent-fuel shipments from nuclear plants are to other plants. A few spent-fuel shipments have, in the past, been made to fuel reprocessing plants. These shipments are packaged in "Type B" casks designed to retain the highly radioactive contents under normal and accident conditions. These containers range in size from 23-36 metric tons (25-40 tons) for truck shipment (each cask is capable of holding seven fuel assemblies) to 109 metric tons (120 tons) for rail transport (with a capacity for 36 assemblies) (DOE/RW-0065). The casks are resistant to both small-arms fire and high-explosive detonation (NUREG-0170).

2.2.5 Nonradioactive Waste Systems

Nonradioactive wastes from nuclear power plants include boiler blowdown (continual or periodic purging of impurities from plant boilers), water treatment wastes (sludges and high saline streams whose residues are disposed of as solid waste and biocides), boiler metal cleaning wastes, floor and yard drains, and stormwater runoff. Principal chemical and biocide waste sources include the following:

Boric acid used to control reactor power and lithium hydroxide used to control pH in the coolant. (These chemicals could be inadvertently released because of pipe or steam generator leakage.)
Sulfuric acid, which is added to the circulating water system to control scale.
Hydrazine, which is used for corrosion control. (It is released in steam generator blowdown.)
Sodium hydroxide and sulfuric acid, which are used to regenerate resins. (These are discharged after neutralization.)
Phosphate in cleaning solutions.
Biocides used for condenser defouling.

Other small volumes of wastewater are released from other plant systems depending on the design of each plant. These are discharged from such sources as the service water and auxiliary cooling systems, water treatment plant, laboratory and sampling wastes, boiler blowdown, floor drains, stormwater runoff, and metal treatment wastes. These waste streams are discharged as separate point sources or are combined with the cooling water discharges.

2.2.6 Nuclear Power Plant Operation and Maintenance

Nuclear power reactors are capable of generating electricity continuously for long periods of time. However, they operate neither at maximum capacity nor continuously for the entire term of their license. Plants can typically operate continuously for periods of time ranging from 1 year to 18 months on a single fuel load. Scheduled and unscheduled maintenance outages and less than peak power generation resulting from diminished consumer demand, or operational decisions, have reduced the power output for the U.S. nuclear power industry as a whole to an average annual capacity of between 58 and 73 percent of the maximum capability for the years 1975 through 1993, inclusive (NUREG-1350, vol. 6).

Maintenance activities are routinely performed on systems and components to help ensure the safe and reliable operation of the plant. In addition, inspection, testing, and surveillance activities are conducted throughout the operational life of a nuclear power plant to maintain the current licensing basis of the plant and ensure compliance with federal, state, and local requirements regarding the environment and public safety.

Nuclear power plants must periodically discontinue the production of electricity for refueling, periodic in-service inspection (ISI), and scheduled maintenance. Refueling cycles occur approximately every 12 to 18 months. The duration of a refueling outage is typically on the order of 2 months. Enhanced or expanded inspection and surveillance activities are typically performed at 5- and 10-year intervals. These enhanced inspections are performed to comply with Nuclear Regulatory Commission (NRC) and/or industry standards or requirements such as the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. Five-year ISIs are scheduled for the 5th, 15th, 25th, and 35th years of operation, and 10-year ISIs are performed in the 10th, 20th, and 30th years. Each of these outages typically requires 2 to 4 months of down time for the plant. For economic reasons, many of these activities are conducted simultaneously (e.g., refueling activities typically coincide with the ISI and maintenance activities).

Many plants also undertake various major refurbishment activities during their operational lives. These activities are performed to ensure both that the plant can be operated safely and that the capacity and reliability of the plant remain at acceptable levels. Typical major refurbishments that have occurred in the past include replacing PWR steam generators, replacing BWR recirculation piping, and rebuilding main steam turbine stages. The need to perform major refurbishments is highly plant-specific and depends on factors such as design features, operational history, and construction and fabrication details. The plants may remain out of service for extended periods of time, ranging from a few months to more than a year, while these major refurbishments are accomplished. Outage durations vary considerably, depending on factors such as the scope of the repairs or modifications undertaken, the effectiveness of the outage planning, and the availability of replacement parts and components.

Each nuclear power plant is part of a utility system that may own several nuclear power plants, fossil-fired plants, or other means of generating electricity. An on-site staff is responsible for the actual operation of each plant, and an off-site staff may be headquartered at the plant site or some other location. Typically, from 800 to 2300 people are employed at nuclear power plant sites during periods of normal operation, depending on the number of operating reactors located at a particular site. The permanent on-site work force is usually in the range of 600 to 800 people per reactor unit. However, during outage periods, the on-site work force typically increases by 200 to 900 additional workers. The additional workers include engineering support staff, technicians, specialty craftspersons, and laborers called in both to perform specialized repairs, maintenance, tests, and inspections and to assist the permanent staff with the more routine activities carried out during plant outages.

2.2.7 Power-Transmission Systems

Power-transmission systems associated with nuclear power plants consist of switching stations (or substations) located on the plant site and transmission lines located primarily off-site. These systems are required to transfer power from the generating station to the utility's network of power lines in its service area.

Switching stations transfer power from generating sources to power lines and regulate the operation of the power system. Transformers in switching stations convert the generated voltage to voltage levels appropriate for the power lines. Equipment for regulating system operation includes switches, power circuit breakers, meters, relays, microwave communication equipment, capacitors, and a variety of other electrical equipment. This equipment meters and controls power flow; improves performance characteristics of the generated power; and protects generating equipment from short circuits, lightning strikes, and switching surges that may occur along the power lines. Switching stations occupy on-site areas generally two to four times as large as areas occupied by reactor and generator buildings, but are not as visible as the plant buildings.

The length of power transmission lines constructed for nuclear plants varies from a few miles for some plants to hundreds of miles for others. Power line systems include towers (structures), insulator strings, conductors, and ground wires strung between towers. Power lines associated with nuclear plants usually have voltages of 230 kV, 345 kV, 500 kV, or 765 kV (see Section 4.5.1). They operate at a low frequency of 60 Hz (60 cycles per second) compared with frequencies of 55-890 MHz for television transmitters and 1000 MHz and greater for microwaves.

Most power line towers are double wooden poles ("H-frame" structure) or metal lattice structures that support one or two sets of conductors (three conductors per set; see Section 4.5.1). Tower height, usually between 21 and 51 m (70 and 170 ft), increases with line voltage. Strings of insulators connect the conductors to the towers. The tops of the towers support two ground wires that transmit the energy of lightning strikes to the ground. Thus, the ground wires prevent lightning strikes to the conductors, minimize the occurrence of power system outages, and protect vital power system components that could be damaged by lightning-caused power surges on the conductors.

2.3 Plant Interaction with the Environment

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This section describes how nuclear plants interact with the environment. Nuclear power plants are sited, designed, and operated to minimize impacts to the environment, including plant workers. Land that could be used for other purposes is dedicated to electric power production for the life of the plant. The aesthetics of the landscape are altered because of the new plant structures; the surface and groundwater hydrology and terrestrial and aquatic ecology may be affected; the air quality may be affected; and, finally, the community infrastructure and services are altered to accommodate the influx of workers into the area. The environmental impact from plant operation is determined largely by waste effluent streams (gaseous, liquid, and solid); the plant cooling systems; the exposure of plant workers to radiation; and plant expenditures, taxes, and jobs.

Operational activities associated with nuclear power plants, including maintenance actions, often produce liquid discharges that are released to the surrounding environment. The major liquid effluent occurs in once-through cooling systems which discharge heat and chemicals into a receiving body of water, but all nuclear power plants have liquid effluents to some extent. To operate, power plants must obtain an NPDES permit that specifies discharge standards and monitoring requirements, and they are required to be strictly in compliance with the limits set by the permit. NPDES permits are issued by the Environmental Protection Agency (EPA) or a designated state water quality agency. They must be renewed every 5 years.

Any gaseous effluents generated are similarly controlled by the EPA and state permitting agencies, which require compliance with the Clean Air Act and any amendments added by the states. On-site incineration of waste products is controlled in this manner.

2.3.1 Land Use

Nuclear power plants are large physical entities. Land requirements generally amount to several hundred hectares for the plant site, of which 20 to 40 ha (50 to 100 acres) may actually be disturbed during plant construction. Other land commitments can amount to many thousands of hectares for transmission line rights-of-way (ROWs) and cooling lakes, when such a cooling option is used.

Nuclear power plants that began initial operation after the promulgation of the National Environmental Policy Act of 1969 (Pub. L. 91-190) or the Endangered Species Act of 1973 (Pub. L. 93-205) are sited and operate in compliance with these laws. Any modifications to the plants after the effective dates of these acts must be in compliance with the requirements of these laws. The Endangered Species Act applies to both terrestrial and aquatic biota. The individual states may also have requirements regarding threatened and endangered species; the state-listed species may vary from those on the federal lists.

2.3.2 Water Use

Nuclear power plants withdraw large amounts of mainly surface water to meet a variety of plant needs (Section 2.2.3). Water withdrawal rates are large from adjacent bodies of water for plants with once-through cooling systems. Flow through the condenser for a 1,000-MW(e) plant may be 45 to 65 m³/s (700,000 to 1,000,000 gal/min). Water lost by evaporation from the heated discharge is about 60 percent of that which is lost through cooling towers. Additional water needs for service water, auxiliary systems, and radioactive waste systems account for 1 to 15 percent of that needed for condenser cooling.

Water withdrawal from adjacent bodies of water for plants with closed-cycle cooling systems is 5 to 10 percent of that for plants with once-through cooling systems, with much of this water being used for makeup of water by evaporation. With once-through cooling systems, evaporative losses are about 40 percent less but occur externally in the adjacent body of water instead of in the closed-cycle system. The average makeup water withdrawals for several recently constructed plants having closed-cycle cooling, normalized to 1,000 MW(e), are about 0.9 to 1.1 m³/s (14,000 to 18,000 gal/min). Variation results from cooling tower design, concentration factor of recirculated water, climate at the site, plant operating conditions, and other plant-specific factors.

Consumptive loss normalized to 1,000 MW(e) is about 0.7 m³/s (11,200 gal/min), which is about 80 percent of the water volume taken in. Consumptive water losses remove surface water from other uses downstream. In those areas experiencing water availability problems, nuclear power plant consumption may conflict with other existing or potential closed-cycle uses (e.g., municipal and agricultural water withdrawals) and in-stream uses (e.g., adequate in-stream flows to protect aquatic biota, recreation, and riparian communities). The environmental impacts of consumptive water use are considered in Sections 4.2.1 and 4.2.2.

As discussed in Section 2.2.3, some nuclear power plants use groundwater as an additional source of water. The rate of usage varies greatly among users. Many plants use groundwater only for the potable water system and require less than 0.006 m³/s (100 gal/min); however, withdrawals at other sites can range from 0.02 to 0.2 m³/s (400 to 3000 gal/min). Impacts associated with groundwater use are discussed in Sections 4.2.2, 4.3.2, and 4.4.3.

Nuclear plant water usage must comply with state and local regulations. Most states require permits for surface water usage. Groundwater usage regulations vary considerably from state to state, and permits are typically required.

2.3.3 Water Quality

Water quality is impacted by the numerous nonradioactive liquid effluents discharged from nuclear power plants (Section 2.1.6). Discharges from the heat dissipation system account for the largest volumes of water and usually the greatest potential impacts to water quality and aquatic systems, although other systems may contribute heat and toxic chemical contaminants to the effluent. The relatively small volumes of water required for the service water and auxiliary cooling water systems do not generally raise concerns about thermal or chemical impacts to the receiving body of water. However, because effluents from these systems contain contaminants that could be toxic to aquatic biota, their concentrations are regulated under the power plant's NPDES discharge permit. The quality of groundwater may also be diminished by water from cooling ponds seeping into the underlying groundwater table.

Sewage wastes and cleaning solvents, including phosphate cleaning solutions, are treated as sanitary wastes. They are treated before release to the environment so that, after release, their environmental impacts are minimized. In cases where nonradioactive sanitary or other wastes cannot be processed by on-site water treatment systems, the wastes are collected by independent contractors and trucked to off-site treatment facilities. Water quality issues relate to the following: NPDES permit system for regulating low-volume wastewater, adequate wastewater treatment capacity to handle increased flow and loading associated with operational changes to the plant and discharges of wastes through emission of phosphates from utility laundries, suspended solids and coliforms from sewage treatment discharges, and other effluents that cause excessive biological oxygen demand.

Many power plants are periodically treated with biocidal chemicals (most commonly some form of chlorine) to control fouling and bacterial slimes. Discharge of these chemicals to the receiving body of water can have toxic effects on aquatic organisms. The biological and water quality impacts of discharges from the discharge systems are considered in Sections 4.2, 4.3, and 4.4.

Chlorine is used widely as a biocide at nuclear power plants and represents the largest potential source of chemically toxic release to the aquatic environment. Chlorine application as a cooling system biocide is typically by injection in one of several different forms, including chlorine gas or sodium hypochlorite. It may be injected at the intake or targeted at various points (such as the condensers) on an intermittent or continuous basis. Such treatments control certain pest organisms such as the Asiatic clam or the growth of bacterial or fungal slime (TVA 1978). The control of biological pests or growths is critical to maintaining optimum system performance and minimizing operating costs (EPRI CS-3748).

Because of the evolution of the guidelines pertaining to chlorine and changes in biocide technologies over the past 15 years, the potential for any adverse impacts of chlorine has been decreasing. Improvements in dechlorination technologies are likely to significantly reduce the level of chlorine in the aquatic environment. Given the critical need for controlling biofouling in the cooling system, both alternative and chlorine treatment technologies are expected to keep pace with regulatory requirements.

All effluent discharges are regulated under the provisions of the Clean Water Act and the implementing effluent guidelines, limitations, and standards established by EPA and the states. Conditions of discharge for each plant are specified in its NPDES permit issued by the state or EPA.

2.3.4 Air Quality

Transmission lines have been associated with the production of minute amounts of ozone and oxides of nitrogen. These issues are associated with corona, the breakdown of air very near the high-voltage conductors. Corona is most noticeable for the higher-voltage lines and during foul weather. Through the years, line designs have been developed that greatly reduce corona effects.

The effluents created and released from the incineration of any waste products must comply with EPA and state requirements regarding air quality. Permits for release of controlled amounts of these effluents to the atmosphere are controlled by state permitting agencies. Because nuclear power plants generally do not produce gaseous effluents, the impact on air quality is minimal.

2.3.5 Aquatic Resources

Operation of the once-through (condenser cooling) system requires large amounts of water that are withdrawn directly from surface waters. These surface waters contain aquatic organisms that may be injured or killed through their interactions with the power plant. Aquatic organisms that are too large to pass through the intake debris screens, which commonly have a 1-cm (0.4-in.) mesh, and that cannot move away from the intake, may be impinged against the screens. If the organisms are held against the screen for long periods, they will suffocate; if they receive severe abrasions, they may die. Impingement can harm large numbers of fish and large invertebrates (e.g., crabs, shrimp, and jellyfish).

Aquatic organisms that are small enough to pass through the debris screens will travel through the entire condenser cooling system and be exposed to heat, mechanical, and pressure stresses, and possibly biocidal chemicals, before being discharged back to the body of water. This process, called entrainment, may affect a wide variety of small plants (phytoplankton), invertebrates (zooplankton), fish eggs, and larvae (ichthyoplankton). Entrainment mortality is variable. Conditions at some plants with once-through cooling may result in relatively low levels of mortality, although at such plants the volumes of water (and numbers of entrained organisms) are often high. On the other hand, generally no aquatic organisms survive at plants with closed-cycle cooling that recirculate water through cooling towers, although the volumes of water withdrawn are relatively low. Biological effects of entrainment and impingement are considered in Section 4.2.3.

Discharges from the plant heat rejection system may affect the receiving body of water through heat loading and chemical contaminants, most notably chlorine or other biocides. Heated effluents can kill aquatic organisms directly by either heat shock or cold shock. In addition, a number of indirect or sublethal stresses are associated with thermal discharges that have the potential to alter aquatic communities (e.g., increased incidence of disease, predation, or parasitism, as well as changes in dissolved gas concentrations).

As stated in Section 2.3.3, all effluent discharges are regulated by the Clean Water Act and standards established by the EPA and the individual states. Conditions of discharge for each plant are specified in the NPDES permit issued for that plant.

2.3.6 Terrestrial Resources

A number of ongoing issues associated with terrestrial resources can arise in the immediate area around the plant or its power transmission lines. Most power lines are located on easements (or ROWs) that the utility purchased from the landowner. Land uses on the easements are limited to activities compatible with power-line operation. In areas with rapidly growing vegetation, utilities must periodically cut or spray the vegetation to prevent it from growing so close to the conductors that it causes short circuits and endangers power line operation. Other terrestrial resource issues can result from changes in local hydrology. Such changes can occur from altered contouring of the land, reduced tree cover, and increased paving. These changes can reduce the value of land and contribute to local erosion and flooding. Additional impacts can include the effects of cooling tower effluent drift, reduced habitat for plants and animals, disruption of animal transit routes, and bird collisions with cooling towers and transmission lines.

Each plant planning to apply for license renewal will need to consult with the appropriate agency administering the Endangered Species Act of 1973 about the presence of threatened or endangered species. Compliance with the Endangered Species Act will be a necessary part of each plant's environmental documentation at the time of license renewal.

2.3.7 Radiological Impacts

2.3.7.1 Occupational Exposures

Plant workers conducting activities involving radioactively contaminated systems or working in radiation areas can be exposed to radiation. Most of the occupational radiation dose to nuclear plant workers results from external radiation exposure rather than from internal exposure from inhaled or ingested radioactive materials. Experience has shown that the dose to nuclear plant workers varies from reactor to reactor and from year to year. Since the early 1980s, when NRC regulatory requirements and guidance placed increased emphasis on maintaining nuclear power plant occupational radiation exposures as low as reasonably achievable, there has been a decreasing trend in the average annual dose per nuclear plant worker.

The effect of plant refurbishment on occupational doses is evaluated in Sections 3.8.2 and in Appendix B. Similarly, the effect of continued operation associated with license renewal on occupational doses is evaluated in Section 4.6.3.

2.3.7.2 Public Radiation Exposures

Commercial nuclear power reactors, under controlled conditions, release small amounts of radioactive materials to the environment during normal operation. These releases result in radiation doses to humans that are small relative to doses from natural radioactivity. Nuclear power plant licensees must comply with NRC regulations (e.g., 10 CFR Part 20, Appendix I to 10 CFR Part 50, 10 CFR Part 50.36a, and 40 CFR Part 190) and conditions specified in the operating license.

Potential environmental pathways through which persons may be exposed to radiation originating in a nuclear power reactor include atmospheric and aquatic pathways. Radioactive materials released under controlled conditions include fission products and activation products. Fission product releases consist primarily of the noble gases and some of the more volatile materials like tritium, isotopes of iodine, and cesium. These materials are monitored carefully before release to determine whether the limits on releases can be met. Releases to the aquatic pathways are similarly monitored. Radioactive materials in the liquid effluents are processed in radioactive waste treatment systems (Section 2.2.4). The major radionuclides released to the aquatic systems are tritium, isotopes of cobalt, and cesium.

When an individual is exposed through one of these pathways, the dose is determined in part by the exposure time, and in part by the amount of time that the radioactivity inhaled or ingested is retained in the individual's body. The major exposure pathways include the following:

inhalation of contaminated air,
drinking milk or eating meat from animals that graze on open pasture on which radioactive contamination may be deposited,
eating vegetables grown near the site, and
drinking (untreated) water or eating fish caught near the point of discharge of liquid effluents.

Other less important exposure pathways include external irradiation from surface deposition; consumption of animals that drink irrigation water that may contain liquid effluents; consumption of crops grown near the site using irrigation water that may contain liquid effluents; shoreline, boating, and swimming activities; and direct off-site irradiation from radiation coming from the plant.

Radiation doses to the public are calculated in two ways. The first is for the maximally exposed person (that is, the real or hypothetical individual potentially subject to maximum exposure). The second is for average individual and population doses. Doses are calculated using site-specific data where available. For those cases in which site-specific data are not readily available, conservative (overestimating) assumptions are used to estimate doses to the public.

2.3.7.3 Solid Waste

Both nonradioactive and radioactive wastes are generated at nuclear power plants. The nonradioactive waste is generally not of concern unless it is classified as Resource Conservation and Recovery Act (RCRA) waste. All waste that is hazardous, that is, classified as RCRA waste, is packaged and disposed of in a licensed landfill consistent with the provisions of RCRA.

Hazardous chemicals, properly handled and controlled, do not present a major health risk to personnel at nuclear power plants, but they must be understood and treated carefully. Hazardous chemicals may be encountered in the work environment during adjustments to the chemistry of the primary and secondary coolant systems, during biocide application for fouling of heat removal equipment, during repair and replacement of equipment containing hazardous oils or other chemicals, in solvent cleaning, and in the repair of equipment. Exposures to hazardous chemicals are minimized by observing good industrial hygiene practices. Disposal of essentially all of the hazardous chemicals used at nuclear power plants is regulated by RCRA or NPDES permits.

Solid radioactive waste consists of LLW, mixed waste, and spent fuel. LLW is generated by removal of radionuclides from liquid waste streams, filtration of airborne gaseous emissions, and removal of contaminated material from the reactor environment.

Mixed waste is LLW that contains chemically hazardous components as defined under RCRA. Mixed waste consists primarily of decontamination wastes and ion exchange resins. The volume of mixed wastes produced at nuclear power plants is typically a small fraction of their overall waste stream, accounting for less than 3 percent by volume of the annual LLW discharged.

Spent fuel is produced during reactor operations. The buildup of fission products and actinides during normal operation prevents the continued use of the fuel assembly. Spent fuel is stored at the reactor site. Uncertainty exists as to when an MRS or permanent spent-fuel repository may become available. However, NRC has examined this issue and determined that licensees may, without significant impact on the environment, store spent fuel on-site for 80 years after ceasing reactor operation (55 FR 38474).

Four major considerations must be addressed when managing solid radioactive waste: (1) the adequacy of interim storage on-site in lieu of permanent off-site disposal, (2) transport of the radiological wastes to disposal sites over the nation's highways and railways, (3) worker and public radiation exposure resulting from handling and processing operations and transportation, and (4) final disposal.

LLW is normally temporarily stored on-site before being shipped to licensed LLW disposal facilities. Previously these facilities were at Barnwell, South Carolina; Beatty, Nevada; and Hanford, Washington. Under the Low Level Radioactive Waste Policy Act of 1980 and the LLRWPAA of 1985, states must secure their own disposal capacity for LLW generated within their boundaries after 1992 by forming waste compacts that are responsible for siting regional disposal facilities, or by siting their own disposal facilities.

For disposal purposes, mixed waste is principally regulated by NRC (10 CFR Part 61). Although the LLRWPAA of 1985 required states to certify they are capable of providing storage and disposal of mixed wastes in an NRC/EPA-licensed facility by 1992, there are currently no licensed disposal facilities accepting commercially generated mixed waste. Because these facilities are not yet available, mixed waste is currently stored on-site.

Originally, disposal of spent fuel in a deep-geological repository was contemplated. However, because of delays in siting a permanent repository on the part of the Department of Energy and delays in developing an interim MRS facility, as required by the Nuclear Waste Policy Act of 1982, nuclear power plants are storing their spent fuel on-site.

LLW is compacted and packaged, typically in 55-gal drums, then transported via truck or railcar. The packaging and transportation of both LLW and mixed waste must comply with EPA requirements. NRC specifications for reviewing the environmental effects of the transport of spent fuel are contained in the Table S-4 Rule (54 FR 187; 10 CFR Part 51.52). States and communities along transportation corridors may impose additional restrictions on the transport of nuclear waste.

Workers receive radiation exposure during the storage and handling of radioactive waste and during the inspection of stored radioactive waste. However, this source of exposure is small compared with other sources of exposure at operating nuclear plants. Members of the general public are also exposed when the LLW is shipped to a disposal site. No other type of radioactive waste is currently being transported from the reactor sites. The public radiation exposures from radioactive material transportation have been addressed rically in Table S-4 of 10 CFR Part 51. Table S-4 indicates that the cumulative dose to the exposed public from the transport of both LLW and spent fuel is estimated to be about 0.03 person-sievert (3 person-rem) per reactor year.

2.3.8 Socioeconomic Factors

2.3.8.1 Work Force

Although the size of the work force varies considerably among U.S. nuclear power plants, the on-site staff responsible for operational activities generally consists of 600 to 800 personnel per reactor unit. The average permanent staff size at a nuclear power plant site ranges from 800 to 2400 people, depending on the number of operating reactors at the site. In rural or low population communities, this number of permanent jobs can provide employment for a substantial portion of the local work force. Table 2.3 depicts mean employment during normal operations in the 1975-1990 period, grouped by the number of reactors.

In addition to the work force needed for normal operations, many nonpermanent personnel are required for various tasks that occur during outages, for example, refueling outages, ISIs, or major refurbishments. Between 200 and 900 additional workers may be employed during these outages to perform the normal outage maintenance work. These are work force personnel who will be in the local community only a short time, but during these periods of extensive maintenance activities, the additional personnel will have a substantial effect on the locality. Table 2.4 indicates the levels of additional personnel typically required for different types of outages.

A substantial portion of the regular plant work force is normally involved in many of the efforts listed in Table 2.4, supplemented as needed by contractor personnel for support during specialized projects. Peak crew sizes are greatly affected by the specific requirements at each plant, utility decisions to make major repairs to systems and components to improve or sustain plant performance, and the relative phasing (schedule overlap) of these activities. Exact crew sizes can, therefore, vary widely from plant to plant.

2.3.8.2 Community

Typically, the immediate environment in which a nuclear power plant is located is rural, but the population density of the larger area surrounding the plant and the distance from a medium- or large-sized metropolitan center varies substantially across sites. Most sites, however, are not extremely remote [i.e., not more than about 30 km (20 miles) from a community of 25,000 or 80 km (50 miles) from a community of 100,000]. The significance of any given nuclear power plant to its host area will depend to a large degree on its location, with the effects generally being most concentrated in those communities closest to the plant. Major influences on the local communities include the plant's effects on employment, taxes, housing, off-site land use, economic structure, and public services.

Table 2.3 Changes in mean operations-period employment at nuclear power plants over time
Operations period	One-unit plants^a	Two-unit plants^a	Three-unit plants^a
Current^b	832 (34)	1247 (28)	2404 (4)
1985-1989	841 (30)	1094 (26)	2095 (4)
1980-1984	447 (19)	946 (21)	1078 (3)
1975-1979	233 (17)	515 (16)	699 (3)

^aNumber in parentheses indicates number of plants providing data.^bApproximately half the respondents reported data for 1989 and half for 1990.

Table 2.4 Mean additional employment per reactor unit associated with three outage types at nuclear power plants
Outage type^a	Number of workers
Typical planned (58)	783
In-service inspection (23)	734
Largest single (45)	1148

^aNumber in parentheses indicates number of plants providing data for the survey (NUMARC).

As noted in Section 2.3.8.1, the average nuclear power plant directly employs 800 to 2400 people. Many hundreds of additional jobs are provided through plant subcontractors and service industries in the area. In rural communities, industries that provide this number of jobs at relatively high wages are major contributors to the local economy. In addition to the beneficial effect of the jobs that are created, local plant purchasing and worker spending can generate considerable income for local businesses.

Nuclear power plants represent an investment of several billion dollars. Such an asset on the tax rolls is extraordinary for rural communities and can constitute the major source of local revenues for small or remote taxing jurisdictions. Often, this revenue can allow local communities to provide higher quality and more extensive public services with lower tax rates. In general, capital expenditures and large changes in public services are seldom necessitated by the presence of the plant and its operating workers, particularly after local communities have adapted to greater and more dynamic changes experienced during plant construction.

As this discussion indicates, nuclear power plants can have a significant positive effect on their community environment. These effects are stable and long term. Because these socioeconomic effects generally enhance the economic structure of the local community, nuclear power plants are accepted by the community, and indeed, become a major positive contributor to the local environs.

2.4 License Renewal--The Proposed Federal Action

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This section provides a brief overview of the most significant requirements of the proposed revision to 10 CFR Part 54, "Nuclear Power Plant License Renewal" (FR 59, no. 174, p. 46574).

Under the license renewal rule (10 CFR Part 54), nuclear power plant licensees would be allowed to operate their plants for a maximum of 20 years past the terms of their original 40-year operating licenses provided that certain requirements are met (Section 1.1). The rule requires licensees submitting license renewal applications to perform specified types of evaluations and assessments of their facilities, and to provide sufficient information for the NRC to determine whether continued operation of the facility during the renewal term would endanger public safety or the environment.

License renewal will be based on ensuring plant compliance with its current licensing basis (i.e., the original plant licensing basis as amended during the initial license term). In addition, licensees will be required to demonstrate for certain important systems, structures, and components (SSCs) that the effects of aging will be managed in the renewal period in a manner so 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 readily monitorable performance or condition characteristics and include most passive, long-lived plant SSCs. Therefore, the NRC's license renewal rule requires a systematic review of, at least, passive, long-lived SSCs that support safety or other critical functions of a nuclear power plant (as delineated in the rule). To make these determinations regarding these SSCs, it is expected that licensees will implement aging management activities for SSCs for which current programs may not be adequate to ensure continued functionality in the renewal term. These aging management activities are expected to include surveillance, on-line monitoring, inspections, testing, trending, repair, refurbishment, replacement, and recordkeeping, as appropriate.

The license renewal rule seeks to ensure that the effects of aging in the period of extended operation are adequately managed. The rule allows credit for existing programs and regulatory requirements that continue to be applicable in the period of extended operation and that provide adequate management of the effects of aging for SSCs. This provision includes credit for rules or requirements, such as those incorporated in the maintenance rule, which could impact license renewal activities performed to detect and mitigate age-related functionality degradation.

The rule requires an integrated plant assessment (IPA). License renewal applicants must perform an IPA to determine which SSCs will be subject to additional review. The IPA would then determine whether additional programs, over and above the current operational and maintenance programs, are required to manage the effects of aging so that equipment function is maintained.

In addition, the license renewal rule requires licensees submitting an application for license renewal to provide the following:

information noting any changes in the current licensing basis that occur during NRC's review of the submittal; and
an evaluation of time-limited aging analyses (i.e., issues such as fatigue, equipment qualification, and reactor-vessel neutron embrittlement which have inherent time limits associated with them).

Key aspects of 10 CFR Part 54 could result in environmental impacts because of the requirements imposed. These key aspects are (1) the enhanced surveillance, on-line monitoring, inspections, testing, trending, and recordkeeping (SMITTR) on SSCs identified in the IPA and (2) the resulting actions taken to ensure that aging would be effectively managed and that the functionality of these SSCs would be maintained throughout the term that the new license would be in effect.

Note that the license renewal rule does not require any specific repairs, refurbishments, or modifications to nuclear facilities, but only that appropriate actions be taken to ensure the continued functionality of SSCs in the scope of the rule.

2.5 Baseline Environmental Impact Initiators Associated with Continued Operation of Nuclear Power Plants

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The previous sections identified the various types of environmental impacts associated with current nuclear power plant operation. Before discussing incremental impacts associated with license renewal, it is useful to first establish a baseline from which to evaluate incremental effects. This baseline is provided by current experience with nuclear power plant operation and the related interactions with the environment. This section presents quantitative information on selected environmental "impact initiators." The term "impact initiators" is defined, followed by estimates of the quantities of each initiator currently generated by typical nuclear power plant operation.

2.5.1 Definition of Environmental Impact Initiators

The terms "environmental impact initiators" and "impact initiators" as used here refer to the precursors to possible environmental impacts. For example, the incremental work force needed to accomplish license renewal activities is not an environmental impact, but the associated effects on housing, transportation, schools, etc., are environmental or socioeconomic impacts. The environmental impact initiators that need to be quantified to estimate overall environmental effects resulting from license renewal are as follows:

Labor hours and work force size associated with on-site craft workers, engineering and administrative personnel, and health physics personnel are needed to estimate socioeconomic impacts to communities affected by personnel employed temporarily at nuclear plants.
Labor costs are used to estimate both economic impacts to affected communities and economic viability of extended plant operation through license renewal.
Occupational radiation exposure is used to estimate radiation-related impacts to workers.
Capital costs of hardware, materials, and equipment are used both to estimate tax-base-related impacts to affected communities and to provide information related to the overall economics of license renewal.
Radioactive waste types, volumes, and disposal costs are used to estimate environmental impacts related to the disposal of such wastes.

These impact initiators are the key elements expected to change, relative to current nuclear plant operation, as a result of actions taken to support license renewal. Other environmental considerations, including water usage, land usage, chemical usage/discharges, and air quality, are not anticipated to change significantly as a result of license renewal activities.

The impact initiators assessed--labor force, labor costs, capital costs, occupational radiation exposure, and radioactive waste volumes--help determine most of the potential changes in environmental impacts resulting from license renewal. For example, estimates of refurbishment labor and capital cost, together with a description of the types of refurbishment activities that might be undertaken, help define potential environmental impacts related to refurbishment period land use, water use, air quality, socioeconomics, nonradiological solid wastes, etc. The impact initiators assessed form a sufficient set from which to assess most license renewal-related environmental impacts. 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).

2.5.2 Baseline Environmental Impact Initiator Estimates

The following discussions provide estimates of the baseline quantities for each of the foregoing impact initiators. These baseline quantities are typical of current nuclear plant operation.

2.5.2.1 Baseline Work Force Size and Expenditures for Labor

Table 2.3 indicates that the current work force at nuclear plant sites is typically in the range of 830 to 2400 permanent staff, depending on the number of operating reactors at a site. On-site personnel responsible for operational activities generally number between 600 and 800 per reactor unit. The average number of permanent staff per reactor unit is estimated to be about 700 people, and this number is approximately the same for both BWRs and PWRs. Assuming a normal 40-hour work week for most on-site staff, this staffing translates into an annual labor effort of about 1.5 million labor hours per unit. The permanent staff is augmented by temporary workers called in to assist with outage activities and special projects. The associated expenditures for labor, including an allowance of roughly 20 percent for temporary staff to support outages and special projects, is estimated to be about $77,000,000 annually per unit.

2.5.2.2 Baseline Capital Expenditures

Nuclear power plants incur expenditures for three major types of capital additions. There are (1) major plant retrofits needed to satisfy NRC requirements to ensure safe plant operation (e.g., changes required as a result of resolution of a generic safety issue), (2) major repairs needed to keep the plant operational (such as main turbine-generator repairs), and (3) discretionary activities undertaken to improve plant performance and labor productivity (DOE/EIA-0547). Expenditures for capital additions have varied widely from plant to plant and from one year to another. In 1989, the average expenditure for capital additions was about $24 per kilowatt, or roughly $24 million for a 1000-MW(e) plant (1989 dollars). These expenditures equate to about $28 million per year per 1000-MW(e) plant in 1994 dollars.

2.5.2.3 Baseline Occupational Radiation Exposure

Occupational radiation exposures vary considerably from plant to plant and from year to year at a given plant. The long-term trends indicate that overall worker exposure has been decreasing on a per-plant basis. The average occupational exposure for the year 1989 was roughly 4.4 person-sievert (440 person-rem) per plant at BWRs and about 3 person-sievert (300 person-rem) per plant at PWRs. For the years 1991 to 1993, the average exposure for all U.S. nuclear plants was about 2.5 person-sievert (250 person-rem) per plant (NUREG-1350, v.6). Significant deviations from these averages are routinely experienced, depending largely on whether a given plant had an outage during a given year and the nature and extent of refurbishment or repair activities undertaken during outages.

2.5.2.4 Baseline Radioactive Waste Generation

Section 2.2.4.3 discussed the different types of radioactive wastes typically generated at nuclear power plants. The type of waste generated in the greatest volumes is LLW. The volume of LLW disposed of annually has shown a decreasing trend over the past several years. Most recently, the amount of LLW disposed of at PWRs has been about 250 m³/year (8800 ft³/year); in contrast, the amount disposed of at BWRs has been about 560 m³/year (19,700 ft³/year).

Small volumes of mixed wastes are also generated by nuclear plant operation. However, any such waste that cannot be treated to eliminate the chemical hazards is currently stored on-site at the nuclear plants and not shipped for disposal.

U.S. reactors generate high-level wastes, primarily in the form of spent fuel. The quantities of spent fuel generated on a per-reactor-year basis is not expected to change with license renewal.

2.6 Environmental Impact Initiators Associated with License Renewal and Continued Operation

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2.6.1 Scope and Objectives of Section 2.6

A major objective of the GEIS is to support the proposed changes to 10 CFR Part 51 by defining the issues that need to be addressed by the NRC and the applicants in plant-specific license renewal proceedings. First, the environmental issues are defined by characterizing and evaluating the actions and activities that may be undertaken by licensees in pursuit of license renewal and extended plant life. These actions and activities are then used to characterize their associated potential environmental impacts.

This section discusses potential actions nuclear power plant licensees may undertake to achieve license renewal and an extended plant life. This section also estimates the extent of the environmental initiators associated with these actions during license renewal and the extended term of operation.

The preceding section noted that the license renewal rule requires that the functionality of important SSCs be maintained throughout the period of the renewed license. To provide this assurance, licensees will likely undertake enhanced SMITTR activities on SSCs identified in the IPA and, based on the findings of these efforts, take appropriate action to ensure that aging is effectively managed and that the functionality of these SSCs is maintained. Incremental repair, refurbishment, and/or replacement of SSCs, as well as related changes to plant operations and maintenance, may be performed to ensure that this objective is achieved. These actions, either directly or indirectly, will produce incremental impacts to the local environment. These incremental effects are over and above those expected if plants were simply to continue to operate as at present.

Licensees may also choose to undertake various refurbishment and upgrade activities at their nuclear facilities to better maintain or improve reliability, performance, and economics of power plant operation during the extended period of operation. These are activities which would be performed at the option of the licensee and which are in addition to those performed to satisfy the license renewal rule requirements.

The set of activities undertaken is expected to vary widely from plant to plant. Some plants may require little refurbishment and upgrading. Other plants may require considerable refurbishment and upgrading. For purposes of the GEIS, two types of license renewal programs were considered for which the environmental impact initiators were developed:

a "typical" or "mid-stream" license renewal program, intended to be representative of the type of program that many plants seeking license renewal might implement, and
a "conservative" or "bounding" program encompassing considerably more activities by licensees, intended to characterize an upper bound, or near upper bound, of the impacts that could be generated at a nuclear power plant.

Each program applies to both BWRs and PWRs. Thus, there are four separate cases or scenarios considered: a typical BWR, an upper bound or conservative BWR, a typical PWR, and a conservative PWR.

The typical scenarios can be used to estimate environmental impacts from an "average" license renewal program and to estimate the nationwide impacts of the total nuclear power plant population. The bounding license renewal scenarios, being much more conservative, are intended to address what might occur for those plants whose impacts will be considerably greater than is typical of the nuclear power reactor population as a whole.

Section 2.6.2 presents the bases and assumptions used in developing the different license renewal scenarios. Section 2.6.3 describes and characterizes the typical license renewal scenarios and the resulting environmental impact initiators. The conservative scenario program is described in Section 2.6.4.

2.6.2 Bases, Assumptions, and Approach

2.6.2.1 Structures, Systems, and Components of Interest

The SSCs of interest for assessing license renewal-related environmental impacts are those that are critical to the safe operation of the plant and that traditionally do not have readily monitorable performance characteristics, which means that the effects of aging may go undetected and lead to the loss of SSC functionality. Many structures and components in currently-licensed LWRs are subject to programs such as the maintenance rule, periodic surveillances, and periodic replacement and refurbishment and have readily monitorable performance or condition characteristics so that these programs can reveal the effects of aging in sufficient time to prevent loss of SSC functionality. However, many other nuclear plant components, such as passive, long-lived structures and components, may not be subject to programs which reveal the effects of aging in sufficient time to ensure their functionality. Therefore, these passive, long-lived structures and components are the items that may need new or incremental aging management activities. The SSCs used in the current evaluation are discussed in Sections 2.6.3.1 and 2.6.4.1 for the typical and conservative programs, respectively.

2.6.2.2 Definition of Candidate Aging Management 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, that is, 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 ranged from enhanced inspection programs to component replacement. In turn, the potential environmental impacts of each identified activity were examined and analyzed.

Following the identification of candidate SSCs and the related aging management activities for each of the different license renewal programs, quantitative estimates of potential environmental impact initiators were developed. The estimates apply to a particular approach to aging management.

The data needed to characterize aging management activities were developed in the context of the four major license renewal programs previously identified: a typical BWR, a conservative BWR, a typical PWR, and a conservative PWR. Each program consisted of the following:

lists of SSCs for which incremental activities would be performed to ensure that safe and economical operation could be achieved throughout the extended life of the plant;
lists of the activities performed on each SSC to manage aging;
the number of times each activity would be performed, accounting for repetitive actions on individual SSCs and the number of similar items in the plant subject to these activities; and
the specific times during which each activity is performed.

The generic license renewal programs utilized in this evaluation were based on similar schedules for carrying out the selected aging management activities. Any major refurbishment work called for by the programs was assumed to start shortly after a renewed license had been granted. In these example programs, this would occur in roughly year 30 of the original 40-year license term. This work was assumed to be completed over several successive outages, including one at the end of the 40th year of plant operation. Incremental SMITTR actions, and the installation of enhanced or additional surveillance and monitoring equipment and systems, were also assumed to be initiated at this time. The SMITTR actions continue throughout the remaining life of the plants. This is true for both the typical and conservative case scenarios.

2.6.2.3 Incremental Effects Only

All aging management programs of interest to the current effort deliberately omit, to the extent possible, current practice as it has evolved and is expected to evolve in the license renewal period. The programs also exclude any changes in the basic design or technology of the plant. Rather, they include only those activities that would constitute a discrete change in the plant's operation and maintenance program and would be implemented only after issuance of the renewal license. In particular, all normal repair activities, as well as any activities undertaken to satisfy recently enacted requirements such as the Maintenance Rule, are considered to fall within the scope of current practice and were excluded from consideration. Therefore, the impact initiators considered here are incremental to those resulting from the extension of current practice.

2.6.2.4 Reference Plant Size and Characteristics

All assessments presented here reflect design features and quantities consistent with 1000-MW(e) plant designs. For the PWRs, the features and sizing chosen were consistent with those for a four-loop Westinghouse plant design with a large dry containment. The BWR features used were representative of designs utilizing internal jet pumps and two recirculation loops. Mark III containment features were used.

2.6.2.5 Reference SMITTR Program

The generic BWR and PWR aging management programs used in the present evaluations for both the typical and conservative scenarios were based on the safety-centered SMITTR programs that were used in the regulatory analysis for 10 CFR Part 54 (NUREG-1362). These basic SMITTR programs were supplemented by activities planned for the Lead Plant programs (Sciacca 1/3/93 and Sciacca 1/13/93). In addition, the aging management programs used as the basis for the current impact initiator estimates included actions anticipated for non-safety-related systems and equipment, but which licensees may undertake to maintain or enhance plant availability and performance. The conservative case scenarios, in particular, assumed considerable expansion of the basic Part 54 programs to include actions on many balance-of-plant SSCs. The inclusion of activities directed toward non-safety-related SSCs considerably expanded the number of times given activities would be performed and significantly increased the variety of activities performed, compared with those considered for the 10 CFR Part 54 Regulatory Analysis. The inclusion of aging management activities beyond those characterized for safety-centered SMITTR programs enhances the comprehensiveness and conservatism of the estimates used in the preparation of the GEIS conservative cases. The typical license renewal program scenarios also include more SMITTR actions than those used for the 10 CFR Part 54 assessments, but to a lesser degree than the conservative case scenarios. The typical program SMITTR activities incremental to those anticipated under Part 54 were included to allow for voluntary actions on the part of licensees to better manage aging of balance-of-plant SSCs. All typical program activities were reviewed for possible overlap with the Maintenance Rule activities; any activities perceived to fall within the scope of the Maintenance Rule or other rules were eliminated from the programs.

2.6.2.6 Major Refurbishments and Replacements

The major refurbishment/replacement class of activities included in the license renewal programs characterized here is intended to encompass actions which typically take place only once in the life of a nuclear plant, if at all. Replacement of BWR recirculation piping and PWR steam generators falls into this category of activities. Many such activities were included in the conservative case license renewal scenarios. The items making up this category include both activities which have already been performed at some operating LWRs and activities which have not yet been performed, at least not to the extent assumed for the purpose of defining potential environmental impacts. The inclusion of activities which have already been performed on some existing nuclear plants is based on the premise that there are certain plants in the reactor population that will not have to perform these activities during the current license term, but that would elect to perform these major activities to enable safe and economic operation for the incremental term allowed with license renewal. In addition, major refurbishment activities included in these example license renewal programs encompass all areas of a nuclear power plant (e.g., structures, mechanical and electrical systems, fluid systems). This approach further ensures that the impacts characterized for the conservative case scenarios have a high probability of bounding the impacts likely to accrue to any individual plant seeking license renewal and extended plant operation.

The typical scenarios, in contrast, included fewer major refurbishment activities of this type. For these scenarios the assumption was made that most plants will have ongoing effective maintenance and refurbishment programs that preclude the need for refurbishment/replacement of all but a few components and structures.

2.6.2.7 Prototypic License Renewal Schedule

Figure 2.3 shows representative timelines for the license renewal process of a nuclear plant. The timelines shown were judged to be reasonable by the NRC staff. The schedule is applicable to both the typical and conservative license renewal scenarios. The upper timeline shows the relationship of the new license period to the initial license period. The lower line indicates the various outage types and their assumed timing over the period covered by a renewed license. The key underlying assumption for the timelines is that the licensee should be assured by the NRC 10 years before the expiration of its current operating license that the plant in question is suitable for license renewal. These 10 years are required for the licensee to arrange for alternative sources of power should a renewed license not be granted. The license renewal process is presumed to start with the licensee initiating a number of studies and analyses to support the license renewal application 3 years before submitting the application to the NRC. The NRC would then perform a detailed review of the application and, in the successful cases, issue a new license (with conditions) within 2 years after the application is received. The new license would go into effect at that point, covering the balance of the original 40-year term, as well as the additional 20-year term.

It was assumed that licensees would initiate incremental aging detection and management activities as soon as the new license was granted, as called for by 10 CFR Part 54. Discretionary major refurbishment activities might also be undertaken early into the license renewal term.

2.6.2.8 Schedule for Performing Major Refurbishment Activities

The reference schedule assumes that major refurbishment activities associated with

Figure 2.3 License renewal schedule and outage periods considered for environmental impact initiator definition.

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. 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, the 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 the typical programs was reduced compared with that deemed necessary for the conservative case scenarios.

2.6.2.9 Outage Types and Durations

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: normal refuelings, 5-year ISI outages, 10-year ISI outages, current term refurbishment outages, and major refurbishment outages. Figure 2.3 illustrates when these outages are assumed to occur. The current term outages fall within the 40-year period initially covered by the plant's current license, but with license renewal they occur during the period covered by the new license.

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. For example, the number of workers required at a site for a given outage is dependent 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.

Table 2.5 summarizes the different outage types and durations for both reactor types and for both the typical and conservative license renewal scenarios. Additional discussion of the basis used in selecting outage durations is provided in Appendix B.

2.6.3 Typical License Renewal Scenario

The characteristics of the typical license renewal program are discussed briefly in Section 2.6.3.1. Listings of the SSCs likely to be subject to incremental aging management activities are provided. Listings of the types of SMITTR actions and major refurbishment activities that may be performed as part of a typical license renewal program are reviewed and discussed in Appendix B. Section 2.6.3.2 summarizes the impact initiator quantities expected to be generated by such a program. Section 2.6.3.2 compares the impact initiator quantities for the typical program scenarios with the impactor initiator quantities currently produced from routine reactor operation.

2.6.3.1 Characterization of Typical License Renewal Programs

The characterization of license renewal programs required that three key types of information be developed: (1) identification of the SSCs likely to be subject to incremental aging management activities, (2) candidate lists of the activities to be performed on these systems and components to suitably manage aging effects that could have potential environmental consequences, and (3) identification of environmental attributes (impact initiators) associated with those activities. The typical programs are intended to be representative of the typical or "average" plant's activities in support of license renewal. However, the typical programs are still somewhat conservative; that is, some plants will not require all of the actions identified in the typical programs. The typical license renewal scenarios were based on the following.

The Monticello and Yankee Rowe lead plant life extension (PLEX) programs were carefully reviewed. Activities included in either program were, with some exceptions, incorporated into the typical license renewal scenarios. The information obtained from the lead plants was also used to establish both the numbers of SSCs subject to a given activity and the schedule for performing such activities.
All activities included in the Part 54 Regulatory Analysis which were pertinent to passive, long-lived SSCs and which were not likely to be implemented because of other rules or regulations were retained as incremental actions. The Part 54 activities were retained both to maintain consistency with the updated Part 54 Regulatory Analysis and to allow for a modest amount of conservatism in the typical scenarios.
As noted previously, recently enacted rules and regulations, in particular the Maintenance Rule, were taken into account in developing typical license renewal or PLEX-related activities.
Surveys were made to help establish the likelihood that certain major activities would be performed by typical licensees seeking license renewal. In particular, assessments were made relative to steam

generator replacement and reactor vessel annealing for PWRs, and for recirculation piping replacement for BWRs. These assessments reviewed the fraction of the affected reactor population that has already performed these refurbishment/replacement activities and ascertained whether such activities might need to be repeated for extended plant life. Based on the results of these reviews, it was assumed that typical license renewal programs will not need to include many such major activities.

Table 2.5 Outage duration summary
	Outage duration (months)
Outage type	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

Typical program structures, systems, and components subject to incremental activities

Tables 2.6 and 2.7 list the SSCs used in the typical program evaluations for which incremental activities are assumed to be onducted during license renewal and extended life. Table 2.6 lists the items subject to incremental SMITTR actions; Table 2.7 lists items subject to major refurbishment/replacement activities. Table 2.6 includes SSCs subject to the addition of new or improved condition monitoring systems, as well as those subject to incremental SMITTR activities. Most of the items in these tables are common to both BWRs and PWRs.

Although the specific numbers of components and design features may be different for these two reactor types, they are similar enough that the environmental impacts resulting from aging management activities on these items will be reasonably similar for both reactor types. Differences in the numbers of like items employed in each plant design were taken into account in assessing impacts.

Table 2.6 Typical program structures and components subject to incremental SMITTR^a activities in support of license renewal
Item	BWR/PWR^b
AC or DC busses	Both
Actuation and instrumentation channels	Both
Bellows	BWR
Building cranes and hoists	Both
BWR control rod drive mechanisms	BWR
BWR recirculation pumps and motors	BWR
Check valves	Both
Compressed air system	Both
Containment	Both
Emergency diesel generators	Both
Fan coolers	Both
Fuel pool	Both
Heat exchangers	Both
Heating, ventilation, and air conditioning	Both
Hydraulic or air operated valves	Both
Main condensor	Both
Main generator	Both
Main turbine	Both
Metal containment, including suppression chamber	BWR
Motor-operated valves	Both
Motor-driven pumps and motors	Both
Nuclear steam supply system supports	Both
PWR critical concrete structure--containment	PWR
PWR reactor coolant pump	PWR
Reactor pressure vessel	Both
Reactor pressure vessel internals	Both
Turbine-driven pumps and turbines	Both

^aSMITTR = surveillance, on-line monitoring, inspections, testing, trending, and recordkeeping.

^bBWR = boiling-watert reactor; PWR = pressurized-water reactor.

Table 2.7 Typical program systems, structures, and components subject to major refurbishment or replacement activities
Item	BWR/PWR^a
BWR safe ends and recirculation and feedwater piping inside containment	BWR
Compressed air system	Both
Containment	Both
Emergency diesel generators	Both
Main generator	Both
Major structures, including buildings and pipe enclosures	Both
Motor-operated valves	Both
Piping sections	Both
Reactor containment building	Both
Reactor pressure vessel	Both
Reactor pressure vessel internals	Both
Steam generators	PWR
Storage tanks	Both

^aBWR = boiling-water reactor; PWR = pressurized-water reactor

Certain SSCs such as the reactor recirculation piping for BWRs and steam generators for PWRs are unique to the plant design type. Potential impacts from aging management activities on such items were treated separately for the two major plant categories.

Definition of aging management activities

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: (1) SMITTR actions, most of which are repeated at regular intervals, and (2) major refurbishment or replacement actions, which usually occur fairly infrequently and possibly only once in the life of the plant for any given item.

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 for the current efforts it has been assumed that 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 a SMITTR activity could involve replacement or refurbishment of the SSC being addressed. Any such SMITTR replacement/refurbishment activities for a particular item typically occur more than once in the extended life of the plant.

Table B.1 of Appendix B 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. These activities include some which are undertaken only to improve reliability or economic performance; thus, Table B.1 includes several active components in addition to the passive, long-lived SSCs that are the focus of 10 CFR Part 54.

Table B.2 of Appendix B lists the major refurbishment or replacement activities used to estimate environmental impacts. 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 and from reported existing industry experience on major refurbishments (Sciacca 1/3/93 and 1/13/93). In other cases engineering judgment provided the basis for the portions of the systems or structures being replaced or refurbished. The extent of major refurbishments envisioned for typical license renewal programs is fairly modest.

2.6.3.2 Typical Program Incremental Initiator Quantities

Table 2.8 summarizes the typical program impact initiator quantities resulting from the incremental SMITTR and major refurbishment/replacement activities assumed to be carried out in support of license renewal and extended plant life. Estimates of the amounts generated are shown for each of the outage types previously discussed, during which these impact initiators are expected to be generated from license renewal activities. Separate estimates are provided for BWRs and PWRs. All figures are shown on a per-plant basis (i.e., for a single nuclear plant).

A comparison of the figures shown in Table 2.8 with current reactor experience as discussed in Section 2.5.2 indicates that, for the typical license renewal scenario, incremental license renewal effects are expected to be relatively modest. For example, with current nuclear plant operation, roughly 1.5 million person-hours are expended each year for on-site operations and maintenance activities. The incremental efforts associated with license renewal-related activities are estimated to add between 500,000 and 700,000 person-hours for all such activities over the remaining life of a typical plant. Thus, the license renewal activities would add roughly 20,000 person-hours per year, which is a small increment compared to the 1.5 million person-hours per year typical of current reactor operation.

Table 2.8 Typical license renewal program environmental impact initiators
Outage type	Labor hours	Additional on-site personnel	Waste volumes (as-shipped) (m³)	Occupational rad exps (person-sieverts)	Waste disposal costs (1994$)^a	Labor costs (1994$)^a	Capital costs (1994$)^a	Total on-site costs (1994$)^a	Off-site costs (1994$)^a	Total costs (1994$)^a
Boiling-water reactors
Full power operation (20 yrs)	0	0	0	0.00	0	0	0	0	0	0
Normal refueling^b	4,148	10	2	0.04	23,000	196,940	215,460	435,400	47,751	483,151
5-yr ISI^c refueling^d	38,675	63	17	0.71	244,000	1,789,900	314,100	2,348,000	0	2,348,000
10-yr ISI refueling^e	62,208	110	30	0.91	424,000	3,082,450	589,550	4,096,000	0	4,096,000
Current term refurbishments^f	45,294	71	17	0.10	245,000	1,715,040	579,360	2,539,400	177,347	2,716,747
Major refurbishment outage^g	298,375	361	69	1.53	976,000	12,585,040	57,589,360	71,150,400	13,804,688	84,955,088
Total all occurrences	660,000	--	220	4.57	3,052,000	27,700,000	62,800,000	93,600,000	14,900,000	108,500,000
Pressurized-water reactors
Full power operation (20 yrs)	0	0	0	0.00	0	0	0	0	0	0
Normal refueling^b	3,488	8	1	0.03	18,000	166,265	145,635	329,900	27,179	357,079
5-yr ISI refueling^d	20,935	33	11	0.30	153,000	953,750	185,250	1,292,000	13,886	1,305,886
10-yr ISI refueling^e	37,482	60	22	0.51	313,000	1,691,600	309,400	2,314,000	831	2,314,831
Current term refurbishments^f	45,924	72	18	0.11	272,000	1,741,880	580,920	2,594,800	176,530	2,771,330
Major refurbishment outage^g	219,018	264	44	0.79	1,631,000	9,108,830	49,380,970	60,120,800	12,068,028	72,188,828
Total all occurrences	510,000	--	170	2.61	3,482,000	21,000,000	53,500,000	78,000,000	13,000,000	91,000,000

Notes:

^aAll cost figures are undiscounted 1994 dollars
^b8 occurrences, 2-month duration each
^cISI = in-service inspection
^d2 occurrences, 3-month duration each
^e1 occurrence, 4-month duration
^f4 occurrences, 4-month duration each
^g1 occurrence, 9-month duration

To convert m³ to ft³, multiply by 35.32.
To convert person-sievert to person-rem, multiply by 100.

Source: Science and Engineering Associates, Inc., January 1995.

Table 2.8 indicates that the number of additional on-site personnel needed to accomplish license renewal-related activities is quite modest for most periods when such activities will be performed. The exception is the major refurbishment outage, when an average of between 200 and 400 additional personnel may be needed. Note that these personnel are in addition to the 700- to 800-person temporary work force typically called in to assist with current outages at nuclear power plants (see Table 2.4). The estimates of additional personnel presented in Table 2.8 are based on the assumption that the incremental work efforts are spread uniformly over the entire duration of the associated outages. In reality, some peaking of staffing requirements will occur during each outage. Additional analyses were performed to evaluate the extent of such peaking, and these analyses are discussed in Appendix B. For the typical BWR license renewal scenario, these analyses indicated that the on-site temporary work force would peak at about 1000 personnel. This peak occurs during the major refurbishment outage, and it includes the temporary work force needed to accomplish refueling and routine outage activities (e.g. routine maintenance and ISI activities) as well as license renewal-related activities. For the PWR, the corresponding temporary worker requirements reach a peak at about 900 additional staff. This peak requirement occurs during the current term outages.

The incremental occupational radiation exposure estimated to accrue because of license renewal activities is between 2.5 and 5 person-sievert (250 and 500 person-rem). On an annualized basis, this represents an increase in annual exposures of about 3 to 4 percent relative to current reactor operation experience.

LLW generation resulting from license renewal activities is projected to be between 185 and 220 m³ (6,000 and 8,000 ft³) of as-shipped LLW over the remaining life of the plants. Currently, PWRs typically generate about 250 m³/year (8800 ft³/year); the amount disposed of at BWRs has been about 560 m³/year (19,700 ft³/year). Thus, the amount of LLW expected to be added because of license renewal activities is roughly the equivalent of one-half to one year's production of waste under current operating conditions. This represents an increment over the remaining life of the plants of about 1 to 3 percent relative to what would be produced with continued present-basis plant operation.

Table 2.8 presents several types of costs associated with license renewal and extended plant life. These include incremental costs associated with additional labor, waste disposal, capital costs, and off-site costs (off-site engineering and administrative support). For the typical BWR license renewal program, the total incremental costs are estimated to be almost $110 million; those for the typical PWR program are estimated to be about $90 million. Although these costs will be incurred over the remaining life of a plant, more than half of these costs might well be incurred in the first few years after a renewed license is granted. For comparison purposes, recent non-fuel operations and maintenance (O&M) costs at U.S. nuclear plants have averaged about $75 million per year for a 1000-MW(e) plant, and capital additions have averaged about $28 million per year (1994 dollars). Thus, the estimated labor and capital expenditures associated with incremental license renewal activities over the remaining life of a plant with a renewed license are the equivalent of roughly a year's expenditures for O&M and capital additions currently experienced by LWRs, or less than a 5 percent increase for such expenditures on an annualized basis.

2.6.4 Conservative License Renewal Scenario

The characteristics of the conservative case license renewal programs are discussed briefly in Section 2.6.4.1. As was done in Section 2.6.3.1 for the typical programs, listings are provided of the SSCs likely to be subject to incremental aging management activities. Listings of the types of SMITTR actions and major refurbishment activities that may be performed as part of a conservative license renewal program are reviewed and discussed in Appendix B. Section 2.6.4.2 summarizes the impact initiator quantities expected to be generated by such programs and compares the impact initiator quantities for the conservative program scenarios with the impactor initiator quantities currently produced in routine reactor operation.

2.6.4.1 Characterization of the Conservative Program

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, or bounding, programs are quite comprehensive, they subsume impacts from more atypical plants.

The conservative case license renewal scenario uses a conservative basis for projecting activities and impacts. The primary bases and assumptions are as follows.

In contrast with the typical programs, the recently enacted rules and regulations, in particular the Maintenance Rule, were not taken into account in revising license renewal or PLEX-related activities. This simplified approach was taken because accounting for such effects would have a negligible impact on the estimates of environmental impact initiator quantities.
All activities included in the Part 54 Regulatory Analysis were retained as incremental actions. In many instances, the number of SSCs subjected to particular SMITTR activities was increased to reflect optional actions on the part of licensees to better ensure reliable and economical service for balance-of-plant systems and components.
The major refurbishment and replacement activities included in the programs are quite expansive and encompass all aspects of the plant designs (e.g., structural, mechanical, and electrical). Similarly, the extent of such activities for particular SSCs is considerable in most cases and is more extensive than that anticipated for the average plant seeking license renewal.
As was previously noted, several of the major refurbishment activities included in the present estimates have already occurred at many nuclear plants. These are activities such as steam generator replacement in PWRs and recirculation piping replacement in BWRs. These activities are included in the conservative case scenarios to encompass those plants that must perform such activities to achieve the desired extended plant life and efficiency, but that have not already done so or that might have to repeat such actions.

License renewal program definition

Conservative program SSCs subject to incremental activities. The conservative program SSCs assumed to be subject to incremental SMITTR activities included all of the SSCs identified in Table 2.6 for the typical program. In addition, the conservative program included the items listed in Table 2.9. The conservative program, in most instances, also included a greater number of a given type of SSC subject to SMITTR actions than did the typical programs. For example, the conservative programs included roughly twice the number of motor-operated valves subject to incremental aging detection and mitigation actions as did the typical programs. This approach was taken with the conservative programs to encompass what might occur at outlier plants.

Both the SSCs subject to incremental SMITTR activities and those subject to major refurbishment activities for the conservative program are more inclusive than those included in the typical program scenarios. A comparison of Tables 2.6 and 2.7 with Tables 2.9 and 2.10 readily demonstrates the more comprehensive nature of the conservative program compared with the typical program scenarios.

Table 2.9 Conservative program additional structures and components subject to incremental SMITTR^a activities in support of license renewal
Item	BWR/PWR^b
BWR control rod drive mechanism	BWR
Compressed air system	Both
Emergency diesel generator	Both
Fan cooler	Both
Main turbine	Both

^aSMITTR = surveillance, on-line monitoring, inspections, testing, trending, and recordkeeping.
^bBWR = boiling-water reactor; PWR = pressurized-water reactor.

Table 2.10 lists items subject to major refurbishment/replacement activities. Most of the items in these tables are common to both BWRs and PWRs.

Definition of conservative program aging management activities. As for the typical programs, the incremental aging management activities carried out for the conservative license renewal scenarios to allow operation beyond the original 40-year license term will include both SMITTR activities and major refurbishment activities.

The SMITTR activities associated with the conservative programs are quite similar to those developed for the typical programs, except that they cover additional types and numbers of SSCs. The scenarios developed for the conservative programs assumed that many balance-of-plant SSCs would be subject to license renewal-related activities to better ensure reliable and economical operation for the extended life of the plant.

Table 2.10 Conservative program systems, structures, and components subject to major refurbishment or replacement activities
Item	BWR/PWR^a
Building crane	Both
BWR recirculation pump and motor	BWR
BWR safe ends and recirculation and feedwater piping inside containment	BWR
Concrete imbedments	Both
Condensate storage tank	Both
Control room communication systems	Both
Electrical cables in and out of containment	Both
Electrical raceways	Both
Emergency diesel generator	Both
Feedwater heater	Both
Heating, ventilation, and air conditioning	Both
Main generator	Both
Main turbine	Both
Major structures, including buildings and pipe enclosures	Both
Metal containment, including suppression chamber	BWR
Nuclear steam supply system supports	Both
Pressurizer and surge line	PWR
Piping section	Both
PWR coolant and feedwater piping inside containment	PWR
Radioactive waste processing system	Both
Reactor containment building	Both
Reactor pressure vessel	Both
Reactor pressure vessel internals	Both
Steam generator	PWR
Steam valve	Both
Switchyard	Both
Turbine pedestal	Both
Ultimate heat sink structures	Both

^aBWR = boiling-water reactor; PWR = pressurized-water reactor.

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.

The list of major replacement and refurbishment activities included here was derived largely from areas of concern identified in the industry pilot and lead NP-5181M, EPRI NP-5289P, EPRI NP-5002). This is true 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, as a minimum, that 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 be other plants which would undertake such tasks only to allow for extended plant operation. Inclusion of these activities in the conservative scenario evaluations provides for an upper bound estimate of what at least a few plants may undertake for license renewal.

Table B.2 of Appendix B lists the major refurbishment or replacement activities used to estimate environmental impacts for the conservative case scenarios. Unless otherwise noted, 100 percent of an SSC was assumed to be replaced or refurbished.

2.6.4.2 Conservative Program Incremental Initiator Quantities

Table 2.11 summarizes the conservative program impact initiator quantities resulting from the incremental SMITTR and major refurbishment/replacement activities assumed to be carried out in support of license renewal and extended plant life. A comparison with the estimates provided for the typical programs (Table 2.8) indicates that the conservative program scenario estimates of impact initiator quantities are factors of four to six greater than those for the typical programs. The type of information provided in Table 2.11 is identical to that provided in Table 2.8. Separate estimates are provided for BWRs and PWRs, and all figures are shown on a per-plant basis.

Table 2.11 Conservative license renewal program environmental impact initiators
Outage type	Labor hours	Additional on-site personnel	Waste volumes (as-shipped) (m³)	Occupational rad exps (person-sieverts)	Waste disposal costs (1994$)^a	Labor costs (1994$)^a	Capital costs (1994$)^a	Total on-site costs (1994$)^a	Off-site costs (1994$)^a	Total costs (1994$)^a
Boiling-water reactors
Full power operation (20 yrs)	49,900	1	0	0.00	0	2,089,856	0	2,089,856	0	2,089,856
Normal refueling^b	11,352	27	5	0.10	64,182	556,407	612,043	1,232,632	131,856	1,364,488
5-yr ISI^c refueling^d	48,406	78	21	0.27	290,508	2,258,137	712,251	3,260,896	0	3,260,896
10-yr ISI refueling^e	101,308	122	38	1.08	537,102	4,585,522	1,250,536	6,373,160	0	6,373,160
Current term refurbishments^f	732,280	866	233	1.91	3,303,684	28,170,043	10,843,605	42,317,332	3,122,803	45,440,135
Major refurbishment outage^g	1,642,760	867	814	15.61	11,525,736	73,719,268	119,968,099	205,213,104	28,546,104	233,759,207
Total all occurrences	4,910,000	--	1,900	26.66	26,372,000	202,000,000	170,900,000	399,300,000	42,100,000	441,400,000
Pressurized-water reactors
Full power operation (20 yrs)	49,900	1	0	0.00	0	2,089,856	0	2,089,856	0	2,089,856
Normal refueling^b	8,733	21	3	0.07	46,166	406,936	410,540	863,642	79,897	943,539
5-yr ISI refueling^d	28,550	46	13	0.35	185,790	1294,224	451,076	1,931,090	50,734	1,981,824
10-yr ISI refueling^e	62,295	75	29	0.66	416,620	2,867,021	845,401	4,129,042	74,282	4,203,324
Current term refurbishments^f	768,460	909	264	2.00	2,889,204	29,607,382	9,687,766	43,184,352	2,821,826	46,006,178
Major refurbishment outage^g	3,241,260	1,713	1,324	13.80	20,204,944	139,806,842	110,947,895	270,959,681	26,185,773	297,145,454
Total all occurrences	6,550,000	--	2,500	23.74	36,919,300	269,000,000	154,700,000	460,700,000	38,300,000	499,000,000

Notes:

To convert m³ to ft³, multiply by 35.32
To convert person-sievert to person-rem, multiply by 100.

Source: Science and Engineering Associates, Inc., January 1995.

A comparison of the figures shown in Table 2.11 with current reactor experience as discussed in Section 2.5.2 indicates that, for the conservative license renewal scenario, incremental license renewal effects are expected to be fairly significant. The incremental efforts associated with license renewal-related activities are estimated to add between 5 million and 7 million person-hours for all such activities over the remaining life of a conservative plant. These increments for license renewal can be compared with the roughly 1.5 million person-hours expended annually with current reactor operation.

If the license renewal efforts were uniformly spread over the 30-year period that a renewed license would be in effect, they would increase annual labor requirements by 10 to 15 percent. The effect of the incremental license renewal labor will be even more significant for certain periods. For example, the number of additional workers needed to accomplish the major refurbishment activities during the major refurbishment outage could potentially double or triple the number needed during a normally scheduled outage. The projected number of additional workers needed for the BWR major refurbishment outage is almost 900, averaged over the entire outage. For certain periods during this outage, the number of additional workers is estimated to be about 1200. For the PWR, the outage average increment in additional personnel needed for the major refurbishment outage is about 1700, and the number is expected to peak at about 2300 for certain periods during this outage. Note that these estimates of peak incremental personnel include the 700- to 800-person temporary work force typically called in to assist with current outages at nuclear power plants (see Table 2.4). Appendix B provides additional discussion of license renewal-related incremental staffing requirements.

The overall occupational radiation exposure estimated to accrue because of conservative program license renewal activities is between 23 and 24 person-sievert (2300 and 2400 person-rem). The large increase compared with the exposures anticipated for the typical programs is largely a result of the extensive major refurbishment activities expected to be undertaken with the conservative program scenarios. On an annualized basis, this is equivalent to an increase in annual exposures of about 20 to 30 percent relative to current reactor operation experience.

LLW generation from license renewal activities is projected to be between 1,900 and 2,500 m³ (65,000 and 90,000 ft³) of as-shipped LLW over the remaining life of the plants. Currently, PWRs typically generate about 250 m³/year (8800 ft³/year); the amount disposed of at BWRs has been about 560 m³/year (19,700 ft³/year). Thus, the amount of LLW expected to be added because of conservative program license renewal activities represents several years worth of production of waste under current operating conditions. This represents an increment over the remaining life of the plants of about 11 percent annually for the BWRs and about 30 percent annually for the PWRs relative to what would be produced with present-basis, continued plant operation. The larger percentage of PWR LLW results primarily from the large volume of the steam generators, which it is assumed will be replaced for the conservative program.

Table 2.11 indicates that the overall incremental costs associated with conservative program license renewal activities are projected to be in the range of $450 million to $500 million per plant (1994 dollars). With current nuclear plant operation, annual expenditures for fuel, O&M, and capital costs are in the range of $150 million to $250 million, depending on individual plant conditions. Thus, the license renewal expenditures represent 2 to 4 years of current overall operating costs.

2.6.5 Impact Initiator Estimate Uncertainties

The NRC staff believes that the license renewal scenarios presented in Section 2.6.4 reasonably characterize both the nature and magnitude of licensee activities that may be undertaken in support of license renewal and extended plant life. Both the typical and conservative programs include some discretionary activities that are assumed to be undertaken by licensees to better ensure economical and reliable plant operation, and that are in addition to those activities performed to meet the requirements of 10 CFR Part 54. The licensee actions in response to the 10 CFR Part 54 requirements, believed to be fairly modest, consist of a considerably smaller set of activities than those characterized for the typical license renewal scenarios. Appendix B presents estimates of impact initiator quantities strictly related to meeting the requirements of the license renewal rule. Thus, a broad spectrum of license renewal programs are possible, and the license renewal-related environmental impacts can vary widely from one plant to another, depending on specific plant conditions and on discretionary activities undertaken by each licensee/applicant. This variability in program characteristics, coupled with uncertainties in parameter values used to estimate specific initiator quantities, results in a considerable degree of uncertainty in the estimates presented in Tables 2.8 and 2.11. Although a rigorous uncertainty analysis has not been performed, the estimates of individual impact initiators provided in Table 2.8 for the typical programs are judged to have uncertainties in the range of ± 30 percent. The more bounding assumptions employed for the conservative scenarios reduce the likelihood that the actual impact initiators experienced could be much higher than those presented in Table 2.11. The uncertainty range for the Table 2.11 estimates, therefore, is judged to be on the order of +10 percent to +30 percent.

2.7 Summary

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This chapter described operating U.S. nuclear power plants and described the nature of their interactions with the environment. The basic requirements of the license renewal rule, 10 CFR Part 54, were reviewed with the focus on aspects which may result in incremental environmental impacts. Chapter 2 also described both typical and conservative license renewal programs characterized for the purpose of estimating license renewal-related environmental impacts. Estimates were provided of environmental impact initiators associated with these programs. These impact initiators are used in the balance of this document to identify and quantify anticipated environmental impacts associated with nuclear power plant license renewal.

2.8 Endnotes

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Construction of nuclear units Grand Gulf Unit 2, Perry Unit 2, and Washington Nuclear Project Units 1, 3, 4, and 5 has been suspended; therefore, these units are not considered in this GEIS.
This category is generally discussed as a separate source of liquid waste primarily for PWRs in which the water has a different radionuclide content and chemistry from primary coolant.

2.9 References

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DOE (U.S. Department of Energy), Transporting Radioactive Materials: Answers to Your Questions, U.S. Department of Energy, Washington, D.C., August 1989a.

DOE (U.S. Department of Energy), Annual Report to Congress, U.S. Department of Energy, Office of Civilian Radioactive Waste Management, Washington, D.C., December 1989b.

DOE/RW-0065, Transporting Spent Nuclear Fuel: An Overview, U.S. Department of Energy, Office of Civilian Radioactive Waste Management, Washington, D.C., March 1986.

DOE/RW-0220, Final Version Dry Cask Storage Study, U.S. Department of Energy, Office of Civilian Radioactive Waste Management, Washington, D.C., February 1989.

EPRI CS-3748, Dechlorination Technology Manual, Electric Power Research Institute, Palo Alto, California, 1984.

EPRI NP-3765, Project 2062-11, W. J. Bailey, et al., Surveillance of LWR Spent Fuel in Wet Storage, Final Report, prepared by Battelle, Pacific Northwest Laboratories, Richland, Washington, for the Electric Power Research Institute, Palo Alto, California, October 1984.

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.

EPRI NP-5526-V1, Project 1557-26, Radwaste Generation Survey Update, Volume 1: Boiling Water Reactors: Final Report, prepared by Analytical Resources, Inc., Sinking Spring, Pennsylvania, for the Electric Power Research Institute, Palo Alto, California, February 1988.

EPRI NP-5526-V2, Project 1557-26, Radwaste Generation Survey Update, Volume 2: Pressurized Water Reactors: Final Report, prepared by Analytical Resources, Inc., Sinking Springs, Pennsylvania, for the Electric Power Research Institute, Palo Alto, California, February 1988.

EPRI NP-5983, Project 2412-6, Assessing the Impact of NRC Regulation 10 CFR 61 on the Nuclear Industry: Final Report, prepared by Vance and Associates, Ruidoso, New Mexico, for the Electric Power Research Institute, Palo Alto, California, August 1988.

EPRI NO-6163, Project 2724-3, On-Site Storage of Low-Level Radioactive Waste at Power Reactors: An International Scoping Study, Final Report, prepared by Science Applications International Corporation, Inc., for the Electric Power Research Institute, Palo Alto, California, December 1988.

Gerstberger, C., Jr., "Westinghouse At-Reactor Consolidation Program," Journal of Nuclear Materials Management, 15(3), 30-31, April 1987.

Johnson, E. R., Trip Report--Attendance at IAEA Technical Committee Meeting on Methods for Expanding Spent Fuel Storage Facilities, prepared by E. R. Johnson, Associates, Oak Ridge, Tennessee, for Martin Marietta Energy Systems, Inc., under Contract 41X-SD841V, Task 10, July 3, 1989.

NUMARC (Nuclear Management Resources Council), Survey of U.S. utility-owned nuclear power plants, Oak Ridge National Laboratory, Oak Ridge, Tennessee, and NUMARC, Washington, D.C., June 1990.

NUREG-0170, Vols. 1 and 2, Final Environmental Impact Statement on the Transportation of Radioactive Material by Air and Other Modes, Volumes 1 and 2, U.S. Nuclear Regulatory Commission, Office of Standards Development, December 1977.

NUREG-0945, "Licensing Requirements for Land Disposal of Radioactive Waste," Final Environmental Impact Statement on 10 CFR Part 61, Vol. 1, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguard, November 1982.

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-2907, J. Tichler, et al., Radioactive Materials Released from Nuclear Power Plants, Annual Report 1987, prepared by Brookhaven National Laboratory, Upton, New York, for the U.S. Nuclear Regulatory Commission, 1989.

NUREG/CR-5640, Overview and Comparison of U.S. Commercial Nuclear Power Plant, Nuclear Power Plant System Source Book, U.S. Nuclear Regulatory Commission, September 1990.

ORNL/TM-6472, J. B. Cannon, et al., Fish Protection at Steam-Electric Power Plants: Alternative Screening Devices, Oak Ridge National Laboratory, Oak Ridge, Tennessee, July 1979.

O'Sullivan, R. A., "International Consensus for the Safe Transport of Radioactive Materials: An Experience to Imitate," IAEA Bulletin, 30(3), 31-34, 1988.

OTA-SET-304, Transportation of Hazardous Materials, Office of Technology Assessment, U.S. Congress, Washington, D.C., July 1986.

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.

TVA (Tennessee Valley Authority), Summary of Added Chemicals and Resulting End Product Chemicals, U.S. Nuclear Regulatory Commission, 1978.

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