Generic Environmental Impact Statement for License Renewal of Nuclear Plants (NUREG-1437 Vol. 1)
Decommissioning is defined as the safe removal of a nuclear facility from service and the reduction of residual radioactivity to a level that permits release of the property for unrestricted use and termination of the license (10 CFR Part 50.82). Decommissioning must occur because a licensee is not permitted to abandon a facility after ceasing operation. Decommissioning activities do not include the removal of spent fuel, which is considered to be an operational activity; the storage of spent fuel, which is addressed in the Waste Confidence Rule (10 CFR Part 51.23); or the removal and disposal of nonradioactive structures and materials beyond that necessary to terminate the U.S. Nuclear Regulatory Commission (NRC) license. Disposal of the nonradioactive hazardous waste that is not necessary for NRC license termination is not considered part of the decommissioning process for which NRC is responsible.
The purpose of this chapter is to determine whether license renewal of nuclear power plants would change the impacts of decommissioning to such an extent that those impacts would need to be assessed and mitigative measures considered as part of the environmental review for license renewal. Current licenses allow nuclear power plants to operate for as long as 40 years. License renewal would extend the period of operation by as much as 20 years. This chapter addresses incremental impacts of decommissioning after a 20-year license renewal compared with operating for 40 years.
The following potential impacts are addressed: radiation exposures to workers and the public, socioeconomic effects, waste management impacts, air and water quality impacts, and ecological impacts. The principal impacts of decommissioning are expected to result from radiation exposures to workers and from disposal of radioactive materials. Decommissioning is expected to have only minor radiological impacts on the public (primarily as a result of transporting radioactive waste). Socioeconomic impacts of decommissioning would result from the demands on, and contributions to, the community by the workers employed to decommission a power plant. As shown in this chapter, the air quality, water quality, and ecological impacts of decommissioning are all expected to be substantially smaller than those of power plant construction or operation because the level of activity and the releases to the environment are all expected to be smaller during decommissioning than during construction and operation. The effect of license renewal on the costs of decommissioning are also examined because the costs of decommissioning continues to be a public concern; however, no category conclusion is reached because the impact of license renewal on decommissioning cost is not a consideration in the environmental review and the decision to renew a license.
The impacts resulting from decommissioning at 40 years (baseline) are taken from NUREG-0586, the two source documents NUREG/CR-0130 and NUREG/CR-0672, and updates to those source documents such as draft reports NUREG/CR-5884 and NUREG/CR-6174. The same methods used in those documents were used to project the impacts of decommissioning after 60 years of operation. Where the source documents did not address a potential impact, other available data and staff members' professional judgments were used to assess the potential for impacts to change as a result of extended operation. The analysis in this chapter is based on large "reference" pressurized-water reactor (PWR) and boiling-water reactor (BWR) nuclear power plants; consequently, the impacts of decommissioning all U.S. nuclear power plants that reach the end of their operating lives without a serious accident should be encompassed by those described here. The changes in impacts resulting from the extended operation and in the environment at the time of decommissioning were considered. [The discussion is built around a "reference" PWR identified by NUREG/CR-0130, the 1175-MW Trojan Nuclear Plant at Rainier, Oregon, and a "reference" BWR, the 1155-MW(e) Washington Public Power Supply System Nuclear Project 2, which was being built near Richland, Washington (NUREG/CR-0672).]
7.2 The Decommissioning Process
This section describes the locations of radioactive materials in nuclear power plants, notes the three commonly discussed decommissioning methods, summarizes experience to date with decommissioning nuclear power plants, and provides information on the wastes generated during decommissioning. Except as noted, the information for this section is from NUREG-0586.
7.2.1 Nuclear Power Plants
Nuclear power plants in the United States use two types of nuclear reactors (Chapter 2); the most common type is the PWR. Most of the 118 licensed power reactors in the United States are PWRs. The other type is the BWR. The locations of radioactive components in these two types of power plants are described briefly to aid the reader's understanding of decommissioning.
220.127.116.11 Pressurized-Water Reactors
Buildings or structures associated with a typical large PWR (Figure 7.1) include (1) the heavily reinforced concrete containment building, which houses the pressure vessel, the steam generators, and the pressurizer system; (2) the turbine building, which contains the turbines and the generator; (3) the cooling water system, which may include the cooling tower and other components; (4) the fuel building, which contains fresh and spent fuel, fuel handling facilities, the spent-fuel storage pool and its cooling system, and the solid radioactive waste system; (5) the auxiliary building, which contains the liquid radioactive waste treatment systems, the filter and ion exchanger vaults, the gaseous radioactive waste treatment system, and the ventilation systems for the containment, fuel, and auxiliary buildings; (6) the control building, which houses the reactor control room and personnel facilities; (7) water intake structures; (8) the administration building; and (9) other structures such as warehouses and nonradioactive shops.
The major radioactive components encountered in decommissioning are associated with the reactor itself--the primary coolant loop, the steam generators, the radioactive waste handling systems, and the concrete biological shield that surrounds the pressure vessel. The reactor core, pressure vessel, steam generators, and piping between the reactor and steam generators are highly radioactive. Because some primary-to-secondary leakage is
Figure 7.1 Typical pressurized-water reactor generating station layout. Adapted from NUREG/CR-0130.
impossible to avoid, the secondary loop, including the turbines, is slightly contaminated. Because of leakage and blowdown, the cooling water is very slightly contaminated. Much equipment in the auxiliary building is contaminated, as is the spent-fuel storage pool and its associated equipment.
18.104.22.168 Boiling-Water Reactors
Buildings and structures associated with a typical large BWR (Figure 7.2) include (1) the reactor building, which houses the reactor pressure vessel, the containment structure, the biological shield, the spent-fuel pool, and fuel handling equipment; (2) the turbine building, which houses the
Figure 7.2 Site layout on a typical boiling-water reactor power plant. Adapted from NUREG-0672.
turbine and electric generator; (3) the radioactive waste and control building, which houses the solid, liquid, and gaseous radioactive waste treatment systems and the main control room; (4) the cooling system; (5) water intake structures and pump houses; (6) the service building, which houses the makeup water treatment system, machine shops, and offices; and (7) other minor structures.
The major sources of radiation in decommissioning a BWR are associated with the reactor itself, the containment structure, the concrete biological shield, the primary coolant loop, the turbines, and the radioactive waste handling systems. The reactor building, the turbine generator building, and the radioactive waste building are the only buildings containing radioactive materials. The reactor core and its pressure vessel are highly contaminated, as is the piping to the turbines. The turbines are also contaminated, but the cooling towers and associated piping are not. Much equipment in the radioactive waste building is contaminated, as is the spent-fuel pool in the reactor building.
7.2.2 Decommissioning Methods
In the NRC's original decommissioning studies (NUREG/CR-0130 for PWRs and NUREG/CR-0672 for BWRs), three alternatives were defined: DECON (decontamination/dismantlement as rapidly after reactor shutdown as possible to achieve termination of the nuclear license); SAFSTOR (a period of safe storage of the stabilized and defueled facility followed by final decontamination/dismantlement and license termination); and ENTOMB (immediate removal of the highly activated reactor vessel internals for disposal and relocation of the remainder of the radioactively contaminated materials to the reactor containment building, which is then sealed. With sufficient time, the radioactivity on the entombed materials will have decayed to levels that permit termination of the nuclear license). However, because current regulations require decommissioning to be complete within 60 years, ENTOMB may not be a viable option.
Changes in the industrial and regulatory situation in the United States since the late 1970s have forced revisions to the scenarios of the NRC's original decommissioning alternatives. The most recently revised decommissioning scenarios are described for PWRs in NUREG/CR-5884 and for BWRs in NUREG/CR-6174. There are two principal changes in the revised scenarios. One is the delay of major decommissioning actions for at least 5 to 7 years following reactor shutdown because of a Department of Energy (DOE) requirement to cool the spent fuel in the reactor pool to avoid cladding failures in dry storage. The other is the assumption that decommissioning will be complete within 60 years, as required by current regulations. This delay results in an increase in decommissioning costs during the short safe storage period while the spent fuel pool continues to operate. Changes in cumulative occupational radiation doses also result from the decommissioning scenario changes.
The basic concept of the three alternatives remains unchanged. However, because of the accumulated inventory of spent fuel in the reactor storage pool and the requirement for at least 5 years of storage for the spent fuel before transfer to DOE for disposal, the timing and steps in the process for each alternative have been adjusted to reflect present conditions and possibilities. For the DECON alternative, it is assumed that the owner has strong incentives to decontaminate and dismantle the retired reactor facility as promptly as possible [i.e., future availability and cost of low-level radioactive waste (LLW) disposal and the need to reuse or dispose of the site, necessitating transfer of the stored spent fuel from the pool to a dry storage facility on the reactor site]. Although continued storage of spent fuel in the pool would be acceptable, the modified Part 50 license could not be terminated until the pool was emptied. It is also assumed that an acceptable dry transfer system would be available to remove the spent fuel from the dry storage facility and place it into licensed transport casks when the time came for DOE to accept the spent fuel for disposal. Similar assumptions are made for the SAFSTOR and ENTOMB alternatives for convenience of analysis, even though extended use of the spent fuel pool might be more cost-effective for SAFSTOR.
DECON is the decommissioning method in which the equipment, structures, and portions of the facility and site containing radioactive contaminants are removed or decontaminated to a level that permits the property to be released for unrestricted use shortly after cessation of operations. It is the only decommissioning alternative that leads to termination of the facility license and release of the facility and site for unrestricted use shortly after cessation of facility operations. DECON activities are expected to require about 9 years for large light-water reactors; less time should be required for smaller facilities.
Because DECON operations are expected to be completed within a few years following shutdown, radiation exposures to workers generally are higher than for decommissioning methods that allow for radioactive decay by delaying or extending the work over a longer period. DECON also requires larger commitments of money and commercial waste disposal site space than do other decommissioning methods. The principal advantage of DECON is that the site is available for unrestricted use promptly.
Nonradioactive equipment and structures need not be dismantled or removed for termination of the NRC license and release for unrestricted use. Once the facility's radioactive structures are decontaminated to levels permitting unrestricted use of the facility, nonradioactive facilities may either be put to some other use or demolished at the owner's discretion. [NRC has issued proposed amendments to 10 CFR Part 20 containing radiological criteria for decommissioning of NRC-licensed nuclear facilities (FR 59, 43200, August 22, 1994). Currently, NRC uses, on a case-by-case basis, criteria and practices contained in Regulatory Guide 1.86 and in a letter to Stanford University from J. Miller, Office of Nuclear Reactor Regulation, NRC, dated April 21, 1982.]
DECON, as defined by NUREG/CR-5884 and NUREG/CR-6174, comprises four distinct periods of effort: (1) preshutdown planning/engineering and regulatory reviews, (2) plant deactivation and preparation for storage (no dismantling activities are conducted during this period that would affect the safe operation of the spent fuel pool), (3) plant safe storage with concurrent operations in the spent-fuel pool until the pool inventory is zero, and (4) decontamination and dismantlement of the radioactive portions of the plant, leading to license termination. Because of the delays in development of the federal waste management system, it may be necessary to continue operation of a dry fuel storage facility on the reactor site after the reactor systems have been dismantled and the reactor nuclear license terminated. However, these latter storage costs are considered operations costs under 10 CFR 50.54(b)(b) and are not considered part of decommissioning.
SAFSTOR is the decommissioning method in which the nuclear facility is placed and maintained in a condition that allows the safe storage of radioactive components of the nuclear plant and subsequent decontamination to levels that permit release for unrestricted use. SAFSTOR was initially conceived of as having three successive stages: (1) a short period of preparation for safe storage (expected to be up to 2 years after final reactor shutdown); (2) a variable safe storage period of continuing care consisting of security, surveillance, and maintenance during which much of the reactor's radioactivity decays; and finally, (3) a relatively short period of decontamination (NUREG-0586). In NUREG/CR-5884 and NUREG/CR-6174, SAFSTOR is described as five distinct periods of effort, with the initial three periods identical to those of DECON. The fourth period is extended safe storage (50 years) with no fuel in the reactor storage pool, and the fifth period is decontamination and dismantlement of the radioactive portions of the plant.
The radioactive or contaminated material must be decontaminated or removed, packaged, and disposed of at a regulated disposal facility. After it has been determined that residual radioactivity is at acceptable levels, the license will be terminated and the facility can be released for unrestricted use. After termination of the NRC license, disassembly or demolition of nonradioactive facilities would be performed at the owner's discretion.
SAFSTOR may be used as a means of satisfying requirements for protection of the public while minimizing the initial commitments of time, money, radiation exposure, and waste disposal capacity. SAFSTOR may also have some advantage where there are other operational nuclear facilities at the same site or where a shortage of radioactive waste disposal capacity occurs. The disadvantages of SAFSTOR are that the site is unavailable for other uses for an extended time; maintenance, security, and surveillance are required until the final decontamination is complete; and few, if any, personnel familiar with the facility are available at the time of decontamination (up to 60 years after plant shutdown).
ENTOMB is the alternative in which radioactive contaminants are encased in a long-lasting material, such as concrete. The entombed structure is maintained and surveillance is performed until the radioactivity decays to a level permitting release of the property for unrestricted use. ENTOMB also comprises five distinct periods of effort, with the initial three periods identical to those of DECON (NUREG/CR-5884 and NUREG/CR-6174). The fourth period is preparation for entombment, when all of the radioactive materials are consolidated within the containment building and entombed. The fifth period is entombed storage for an extended time, between 60 and 300 years.
ENTOMB is intended for use where the residual radioactivity will decay to levels permitting unrestricted release of the facility within reasonable time periods (100 years). However, a few radioactive isotopes produced in nuclear reactors have long half-life periods (Section 7.3.1) that prevent the release of the facilities for unrestricted use within the foreseeable lifetime of any man-made structure. ENTOMB would be a viable alternative only for facilities where radioactive isotopes would be expected to decay to safe levels within the expected lifetime of the entombment structure. This condition likely would not pertain to nuclear power reactors. In addition, the use of the ENTOMB alternative contributes to problems associated with increased numbers of sites dedicated to "interim" storage of radioactive materials for long periods of time.
7.2.3 Decommissioning Experience
U.S. commercial nuclear power reactors that have been shut down through 1992 are listed in Table 7.1. An additional 24 reactors have been or are being decommissioned in France, West Germany, Canada, the United Kingdom, Sweden, and Japan (Gaunt et al. 1990).
7.2.4 Inventory and Disposition of Radioactive Materials
Table 7.1 U.S. commercial nuclear power reactors formerly licensed to operate
|Unit/location||Construction typea/MW(t)||Operating license issued/shut down||Decommissioning alternative selected/current status|
Punta Higuera, PR
|Carolina Virginia Tube
Elk River, MN
Lagoona Beach, MI
|Fort St. Vrain
DECON in progress
|GE Vallecitos Boiling Water Reactor
|Humboldt Bay 3
|Indian Point 1
Sioux Falls, SD
DECON in progress
|Peach Bottom 1
Peach Bottom, PA
|San Onofre 1
San Clemente, CA
Wading River, NY
DECON in progress
|Three Mile Island 2
Londonderry Township, PA
Franklin County, MA
aBWR = boiling-water reactor; HTG =
high-temperature gas-cooled; OCM = organically cooled and moderated; PTHW =
pressure tube, heavy water cooled and moderated; PWR =
pressurized-water reactor; SCF = sodium cooled, fast; SCGM = sodium cooled, graphite
bAtomic Energy Commission/Department of Energy owned; not regulated by the Nuclear Regulatory Commission.
cHolds by-product license from state of South Carolina.
dSan Onofre 1 decommissioning plan was due to the Nuclear Regulatory Commission in November 1994.
eThree Mile Island 2 has been placed in a monitored storage mode. The licensee plans to maintain the facility in monitored storage until Three Mile Island 1 permanently ceases operation, at which time both units are to be decommissioned simultaneously.
fTrojan received a possession-only license on 05/05/93. The license is evaluating SAFSTOR and DECON decommissioning alternatives. A decommissioning plan was due to the Nuclear Regulatory Commission in January 1995.
gYankee Rowe received a possession-only license on 08/05/92. The licensee submitted a decommissioning plan on 12/20/93. Decommissioning alternative depends on the availability of low-level waste disposal facilities.
Source: DOE/RW-0006, rev. 6.
Radioactive materials can be classified as activated or radioactively contaminated materials. Materials become activated when they have been exposed to (irradiated by) high levels of neutron radiation (such as in a reactor). When normal (stable) atoms in a material absorb neutrons, they become unstable (radioactive) and subsequently emit energy in the form of radiation. Radioactive contamination is radioactive material in the form of fine particles, liquids, or gases that are deposited on the surface of, or mixed with, materials that otherwise are not radioactive. Contaminated materials can generally be decontaminated to various degrees by several techniques. These techniques range from simply washing with soap and water to sandblasting contaminated surfaces. Decontamination techniques for liquids and gases include filtration and chemical ion exchange. Activated materials cannot be decontaminated; they remain radioactive until the radioactive constituents decay to stable isotopes.
Reactor components are generally both activated and contaminated. The principal activated components of a power plant are the reactor internals and the biological shield. Other reactor system components, such as the primary and possibly the secondary coolant loops, the turbines in BWRs, and the radioactive waste handling systems, are not activated but are highly contaminated by the contaminated fluids they contain. The major source of contamination in reactor coolant is the plant corrosion and wear material suspended in the coolant that becomes activated as it passes through the reactor core. Surface contamination can also be found in areas of the plant where leaks from contaminated systems have occurred.
The inventory of radionuclides for PWRs and BWRs is slightly different. A typical large PWR would have a radioactivity level of about 4.8 million Ci (1Ci = 3.7 x 1010 Bq) in the major reactor components, 4800 Ci of radioactive corrosion products in the primary coolant system, and 1200 Ci of radioactivity in the concrete biological shield at the time of shutdown (NUREG/CR-0130). A typical large BWR would have a radioactivity level of about 6.3 million Ci in the major reactor components, 8600 Ci of radioactive corrosion products in the primary coolant system, and 1000 Ci of radioactivity in the concrete biological shield at the time of shutdown (NUREG/CR-0672).
The principal radioactive isotopes from irradiated steel and concrete, with their modes of decay and their half-lives, are listed in Table 7.2. By the end of 40 years of operation, the radionuclides with half-lives of less than about 5 years are at equilibrium, because their rates of decay equal their rates of generation. No matter how much longer a power plant is operated, the concentration of short-half-life radionuclides will not increase. The longer-lived radionuclides are generated much faster than they decay; thus their concentrations increase approximately in proportion to the reactor operating time. Figure 7.3 illustrates the buildup of some important radionuclides as a function of nuclear plant operating life.
Radioactive isotopes that are mainly beta emitters or that have very short half-lives do not contribute significantly to the personnel radiation dose associated with decommissioning. Because beta radiation is weakly penetrating, it can be shielded easily and presents a hazard mainly if ingested or inhaled by operations personnel. Isotopes with very short half-life periods can be allowed to decay to insignificant levels before decommissioning operations begin.
At the time of decommissioning, radioactive materials are found in the reactor building, the auxiliary building, and the fuel building (Section 7.2.1). Immediately after operations are terminated, these parts of the plant are highly radioactive because of short-lived activation products. The highest levels of radioactivity subside very quickly as short-lived radionuclides decay and progressively longer-lived radionuclides dominate the overall radioactivity. After about a year, 60Co dominates the radiation dose to workers. After about 100 years, 94Nb dominates the radiation dose to workers or persons in the vicinity (Figure 7.4). For all practical purposes, the radiation dose to workers will not decrease further because 94Nb has a 20,000-year half-life. Because 60Co and 94Nb dominate the radiation dose during the time of decommissioning, their characteristics affect the decommissioning process.
|Table 7.2 Principal activated radioactive isotopes found in operating nuclear power plants (excluding fuel)|
|Element||Isotope||Decay modea||Half-life (years)|
|Hydrogen||3H||1.23 x 101|
|Carbon||14C||5.73 x 103|
|Phosphorus||33P||6.9 x 10-2|
|Silicon||35S||2.4 x 10-1|
|Chlorine||36Cl||,||3.01 x 105|
|Argon||37Ar||9.5 x 10-2|
|Argon||39Ar||2.99 x 102|
|Potassium||40K||,||1.28 x 109|
|Calcium||41Ca||8.0 x 104|
|Calcium||45Ca||4.5 x 10-1|
|Scandium||46Sc||2.3 x 10-1|
|Chromium||46Cr||7.6 x 10-2|
|Manganese||54Mn||8. x 10-1|
|Iron||55Fe||2.7 x 100|
|Iron||59Fe||,||1.2 x 10-1|
|Cobalt||58Co||2.1 x 10-1|
|Cobalt||60Co||,||5.27 x 100|
|Nickel||59Ni||8.0 x 104|
|Nickel||63Ni||9.2 x 101|
|Zinc||65Zn||6.7 x 10-1|
|Niobium||93mNb||1.36 x 101|
|Niobium||94Nb||,||2.03 x 104|
|Niobium||95Nb||,||9.6 x 10-2|
|Molybdenum||93Mo||3.5 x 103|
|Zirconium||95Zr||,||1.8 x 10-1|
|Technetium||99Tc||2.13 x 105|
|Silver||108mAg||,||1.27 x 102|
|Silver||110mAg||,||6.8 x 10-1|
|Cadmium||109Cd||1.3 x 100|
|Samarium||151Sm||,||9.0 x 101|
|Europium||152Eu||,||1.33 x 101|
|Europium||154Eu||,||8.8 x 100|
|Holmium||166mHo||1.2 x 103|
a = beta, = gamma (including x-rays).
Source: R. C. Weast, ed. Handbook of Chemistry and Physics, 53rd ed. 1972-73, Chemical Rubber Company, Cleveland, 1972.
Figure 7.3 Buildup of activation products in pressurized-water reactor internal components as a function of effective full-power years.
Figure 7.4 Time dependence of radioactivity and dose rate in a boiling-water reactor core shroud after 40 years of operation. Source: NUREG/CR-0672.
Both 60Co and 94Nb are activation products--isotopes created when neutrons from nuclear fission convert nonradioactive elements (59Co and 93Nb) in the structural components of the plant into radioactiveisotopes. An important difference is that 94Nb in the steel reactor vessel and components, formed by activation of 93Nb, is not subject to corrosion and movement throughout the primary system to the extent that 60Co is. Consequently, equipment in the reactor containment building that is not exposed to high neutron fluxes and parts of the fuel and auxiliary buildings may be highly contaminated with 60Co but only slightly so with 94Nb.
Extending operations to 60 years would not increase the shutdown radioactivity level of either a PWR or BWR to any appreciable extent. This is because most of the radioactivity at shutdown results from short-half-life radionuclides, such as 60Co, that are already in equilibrium by the time 40 years of operations have transpired. The only change in radioactive inventory resulting from the additional 20 years of operations is the further accumulation of long-half-life radionuclides such as 63Ni and 94Nb, but these long-half-life radionuclides produce only a small fraction of the total radioactivity at shutdown. Of the long-half-life radionuclides, 63Ni contributes most at shutdown but composes less than 3 percent of the total radioactivity. Twenty additional years of operation would increase its contribution to about 4 percent of total shutdown radioactivity. Because 63Ni is a beta emitter, it contributes only a very small part of the dose to workers or the public. Gamma-emitting 94Nb is the most important long-half-life radionuclide with regard to producing external radiation exposure. Based on Figure 7.4, it can be determined that at shutdown 94Nb contributes less than 0.001 percent of the total potential dose. Even though 20 additional years of operation would increase the amount of 94Nb by 50 percent, it would not increase its contribution to the dose much above 0.001 percent.
7.2.5 Waste Generated During Decommissioning
This section summarizes the quantities and types of radioactive waste and emissions generated in decommissioning after 40 and 60 years of operation, respectively. Because the demolition and disposal of nonradioactive parts of nuclear facilities are not considered part of decommissioning, almost all waste generated during decommissioning is radioactive. Although the demolition and disposal of the nonradioactive parts may continue during and after decommissioning, these activities are not regulated by NRC. The impacts of radioactive wastes and emissions are described in Section 7.3. This section does not take into account volume reduction or aggressive processing that could allow release for unrestricted use.
22.214.171.124 Atmospheric Emissions
As shown in Table 7.3, the total atmospheric releases for decommissioning are less than 100 mCi, whereas normal operations average about 3000 Ci/year. Atmospheric releases are expected to consist largely of dust, aerosols, and smokelike particulates produced during the dismantling and handling of reactor components. These releases were estimated by assuming that the airborne concentrations of radionuclides will be a fraction of the contamination level on and in the radioactive components (NUREG/CR-0130 and NUREG/CR-0672). Because the radioactive inventory would be nearly unchanged by operations during a 20-year license renewal term, no difference exists between the base case and 20 years of additional operation.
Table 7.3 Airborne radioactive releases resulting from decommissioning typical pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) with normal operating releases, base case (40 years of operation) a
aDecommissioning releases are for
40 years of operation. Releases for 60 years of operation would be essentially the
bSource: NUREG/CR-0130, Table 11.2-2.
eSource: NUREG/CR-0672, Tables N.2-12, N.3-4, N.4-4, E.2-11. Decommissioning is assumed to last 5 years.
126.96.36.199 Liquid Effluents
No estimates of liquid waste releases are available for decommissioning nuclear power plants. However, liquids will be produced by decontamination procedures (e.g., some cutting operations and possibly some chemical decontamination procedures) and by disposal of plant fluids (e.g., cooling water and water from fuel storage pools). Filtration and ion exchange methods will be used to decontaminate liquids, as would be done during normal operations. Some liquid effluents may be contaminated with chelating agents and may require further processing. These methods are expected to keep waterborne effluents of most radionuclides within the values of normal operations. Tritium (3H) is the only radioactive isotope that cannot be removed from waste water by these means.
Tritium is found principally in the primary coolant-loop water. Tritium cannot be removed from water except by extraordinary means and is normally discharged to a surface water body. Normal 3H discharges from PWRs range from a few hundred to a few thousand curies per year. BWR 3H discharges are generally only about 10 percent as high as 3H discharges from PWRs. About 500 Ci of 3H can be found in PWR primary coolant-loop water. Discharge of the entire volume of primary coolant-loop water over a period of 1 to 5 years after shutdown would be feasible without exceeding normal operating period discharge rates. The amounts or characteristics of liquid effluents discharged during decommissioning would not be changed by operation during a 20-year license renewal term. Discharge of primary coolant water during normal operations limits the accumulation of 3H in the primary coolant loop; thus 3H is in equilibrium in the primary coolant water well before 40 years of operation.
188.8.131.52 Solid Waste
Table 7.4 summarizes the quantities of LLW generated by decommissioning of large PWRs and BWRs. The table shows that the largest amount of LLW is generated by the DECON method and the least is generated by the SAFSTOR method. The quantities listed for the ENTOMB method do not include the volume of the entombing structure or the wastes within.
|Table 7.4 Estimated burial volume of low-level waste and rubble for large pressurized-water reactor (PWR) and boiling-water reactor (BWR) decommissioning, base case (40 years of operation)|
a1 m3 _ 35.3 ft3
Source: NUREG/CR-5884, Table ES.1 and NUREG/CR-6174, Table ES.1.
The decommissioning waste volumes for all three methods of decommissioning also would not be affected by extending the volume of radioactive materials would not increase. (Operational waste quantities would continue, but they do not affect the amount of decommissioning waste.) An additional 20 years of operation would slightly affect the waste characteristics. As discussed in Section 7.2.4, the quantity of long-lived activation products such as 94Nb would continue to increase, essentially in proportion to the additional operational time. As a result, the long-half-life radionuclides in the waste would increase by 50 percent if the plants were operated an additional 20 years. However, as explained earlier, these long-lived radionuclides contribute only a small fraction of the shutdown radioactivity level.
7.3 Decommissioning Impacts and Changes Resulting from Life Extension
Estimated decommissioning impacts for 40 years of operation--the base case (taken primarily from NUREG-0586, NUREG/CR-0130, and NUREG/CR-0672)--and the change in impacts caused by continued operations for an additional 20 years under license renewal are reported for each impact area in the following sections. These impacts are estimated for PWRs and BWRs. The per-reactor impacts of decommissioning at multiple-reactor sites are not expected to be significantly different from those at single-reactor sites. [The impacts would be smaller at multiple reactor sites if the reactor decommissionings were staggered and if LLW were stored on the site (NUREG-0586)].
7.3.1 Radiation Dose
The estimated occupational and public radiation doses resulting from the three decommissioning methods after 40 years of operation (base case) are summarized in this section. Occupational dose estimates were presented in draft reports NUREG/CR-5884 and NUREG/CR-6174. These reports do not provide estimates of doses to the public. The Atomic Energy Act requires the Nuclear Regulatory Commission to promulgate, inspect, and enforce standards that provide an adequate level of protection of the public health and safety and the environment. These responsibilities, singly and in the aggregate, provide a margin of safety. For the purposes of assessing radiological impacts, the Commission has concluded that impacts are of small significance if doses and releases do not exceed permissible levels in the Commission's regulations.
184.108.40.206 Occupational Dose
For both PWRs and BWRs, there are substantial differences among the occupational radiation doses for the decommissioning methods (Table 7.5). The DECON method has the highest doses, followed by ENTOMB and then SAFSTOR. Although extending operations 20 years would increase the doses from 94Nb and other less-important long-half-life radionuclides, these doses would not have any appreciable effect on the occupational dose because short-lived radionuclides (primarily 60Co) are the principal sources of worker exposure. For each decommissioning method, the bulk of the dose comes during activities in the first few years after termination of plant operations (period four begins less than 5 years after terminating operations for DECON), when the radioactivity level of 60Co is still significant. At the end of 60 years of SAFSTOR, the dose rate would have decayed to about 0.01 percent of the dose rate at the end of operations, at which time 94Nb would contribute only about 2 percent of the total (Figure 7.4).
An additional 20 years of operation before 60 years of SAFSTOR would increase the amount of 94Nb by approximately 50 percent. During period 5, occupational exposures from SAFSTOR activities would be no more than 10 person-rem. (Section E.A.3 of Appendix E discusses the International System units used in measuring radioactivity and radiation dose. The contribution from 94Nb would be less than 0.2 person-rem. The increase in dose during decommissioning after 20 additional years of operation would be no more than about 0.1 person-rem.
Although total doses to the decommissioning workforce may increase slightly as a result of an additional 20 years of plant operation, the exposure of individual workers will be maintained well below the existing regulatory limits of 10 CRF Part 20. Accordingly, the Commission concludes that radiological impacts to the decontamination workforce as a result of license renewal is of small significance.
The potential increase in total dose to the decommissioning work force may be mitigated by programs that are responsive to 10 CFR 20.1101(b), which requires that "The licensee shall use, to the extent practicable, procedures and engineering controls based upon sound radiation protection principles to achieve occupational doses and doses to members of the public that are as low as is reasonably achievable (ALARA)." The ongoing ALARA programs within the industry already employ measures that would be considered for mitigating the generation or the accumulation of long-lived activation products during 20 additional years of operation. Two examples of mitigation measures that are already in use are (1) replacing components using cobalt alloys with those using low-cobalt or cobalt-free alloys and (2) full system decontamination (e.g., see Moore 1995). No additional mitigation measures warranted. This is a Category 1 issue.
Table 7.5 Estimated occupational radiation doses for decommissioning a large reactor (person-rem), base case (40 years of operation)a.
|Decommissioning period b||DECONc,d||SAFSTORc,e||ENTOMBc,f|
|Boiling-water reactor i|
aOccupational radiation exposures are for
decommissioning after 40 years of operations.
bDecommissioning periods are defined in NUREG/CR-6174 and NUREG/CR-5884.
cDECON, SAFSTOR, and ENTOMB are defined differently by NUREG/CR-5884 and NUREG/CR-6174 than by previous analyses.
hTotals may not equal sum of entries because of rounding.
220.127.116.11 Dose to the Public
For both PWRs and BWRs, the radiation dose to the public results primarily from waste shipment (Table 7.6). Furthermore, the dose is almost exclusively caused by shipment of 60Co and shorter-lived radionuclides; for truck shipments, the SAFSTOR 100-years alternative shows negligible dose to the public. Because only the quantities of long-lived radionuclides would increase if plants were operated an additional 20 years, only the dose caused by the long-lived radionuclides would increase. Because the dose to the public from long-lived radionuclides after 40 years of operation is negligible (see the SAFSTOR 100-years alternative in Table 7.6), an increase of 50 percent of this negligible amount would still remain a negligible dose (less than 0.1 person-rem).
Table 7.6 Estimated radiation dose to the public for decommissioning a largen-rem), base case (40 years of operation)a,b
|DECON||30 years||100 years||ENTOMB|
|SAFSTOR preparation truck shipments||NA||2||2||NA|
|Decontamination truck shipments||21c||0.4c||negc||NA|
|Entombment truck shipments||NA||NA||NA||4|
|SAFSTOR preparation truck shipments||NA||2||2||NA|
|Decontamination truck shipments||10c||negc||negc||NA|
|Entombment truck shipments||NA||NA||NA||5-7d|
aPublic radiation exposures are for decommissioning after 40 years of operation (NUREG-0586).
Decommissioning exposures after 60 years would be identical, except as noted. Draft reports NUREG/CR-5884 and
NUREG/CR-6174 do not provide updates for this information.
bNA means not applicable and neg means negligible.
cDecommissioning after 60 years of operation would increase occupational and public exposure during
(1) decontamination and (2) decontamination truck shipments by only negligible amounts.
dRanges are for removing or leaving internal components or leaving them in place. The higher exposures are associated with removing the internals.
Note: To convert person-rem to person-sievert, multiply by 0.01.
The negligible public radiation exposures for SAFSTOR preparation, continuing care, and decontamination (Table 7.6) include exposures from atmospheric and liquid releases during routine decommissioning operations. There are no historical records of significant releases during decommissioning, and no reliable estimates can be made of the probability and consequences of such events. However, the probability and consequences of such releases are not expected to be different for decommissioning a base case facility versus decommissioning a facility that has had 20 years of additional operation.
Extending reactor operating life from 40 to 60 years is expected to increase the concentration of long-half-life radionuclides in the facility by up to 50 percent. By the end of the initial 40 years of operation, the radionuclides with half-lives of less than about 5 years are at equilibrium because their rates of decay equal their rates of generation. The release of radioactivity to the atmosphere during decontamination is negligibly small and primarily involves short-lived nuclides. Public exposure even with the increased concentration of long-lived nuclides would remain negligible. The exposure of individual members of the public will be maintained well below existing regulatory limits. Accordingly, the staff concludes that the contribution of license renewal to radiological impacts from decontamination is of small significance. As discussed in Section 18.104.22.168, measures that can reduce possible dose levels to the public are available and are being employed in pursuit of ALARA.
Radiation doses (public and occupational) from decommissioning that are attributable to license renewal are a Category 1 issue.
7.3.2 Waste Management Impacts
An operating 1000-MW(e) reactor generates about 38 m3 (1300 ft3) of spent fuel and about 52,000 m3 (1,800,000 ft3) of LLW over its 40-year life (NUREG-0586, pp. 2-21). (LLW is defined in Chapter 6.) The reference PWR and BWR are about 15 percent larger, so they would be expected to generate about 15 percent more waste than a 1000-MW(e) plant. As shown by Table 7.4, decommissioning either type of plant after 40 years of operation (base case) would generate less than 15,000 m3 (530,000 ft3) of LLW for DECON or short-term SAFSTOR and less than 1,200 m3 (42,000 ft3) of LLW for SAFSTOR of 50 years or longer. These waste volumes include spent chelating agent used to decontaminate liquids. The 15,000 m3 (530,000 ft3) of decommissioning LLW is about 25 percent, and 1,200 m3 (42,000 ft3) is only about 2 percent, of the LLW generated by 40 years of operations. None of these estimates of waste volume includes waste generation during refurbishment.
Extending operations by 20 years would not increase decommissioning waste volumes, so the ratio of decommissioning waste volume to operating waste volume would be even lower. After 60 years of operation, decommissioning LLW would be less than about 20 percent of the operational LLW. If SAFSTOR were used, the decommissioning LLW would be only about 1 percent of the LLW generated by operations.
While the volume of decommissioning waste will not increase with 20 years of additional operating time, the concentration of long-half-life radionuclides will increase. LLW is classified by 10 CFR Part 61 into three waste classes denoted A, B, and C and a category of LLW designated "greater than Class C" (GTCC). Classes A and B are wastes that are contaminated with relatively short-half-life radionuclides and may be safely disposed of near the earth's surface because they will decay to a nonhazardous condition within about 100 years. Class C waste can be disposed of at a moderate depth or near the earth's surface with engineered barriers to prevent inadvertent intrusion into the wastes. GTCC waste cannot safely be disposed of near the earth's surface (Section 22.214.171.124; 10 CFR Part 61.7).
Table 7.7 gives the estimated decommissioning LLW breakdown (DECON scenario) for the base case by waste class per 10 CFR Part 61. Items classified as C and GTCC consist of highly activated metal located in the high-flux neutron field. For the PWR, the GTCC items include the lower core barrel, the thermal shields, the core shroud, and the lower grid plate. The class C items are the upper grid plate and the lower support column. The class B wastes consist of spent resins used during decommissioning, part of the combustible contaminated wastes, and part of the cylindrical pressure vessel wall. The only GTCC wastes from a BWR are the core shroud and top fuel guide. BWR class C wastes are from the control rods and in-core instrumentation, jet pump assemblies, and the top fuel guide. The class B wastes are from the steam separator assembly, the reactor vessel wall, and portions of the clean-up wastes.
|Table 7.7 Decommissioning waste volumes for reference pressurized-water reactor (PWR) and boiling-water reactor (BWR) after 40 years of operationa|
|Class A||Class B/C||GTCCb|
|PWR||6,797 m3||184 m3||11 m3|
|BWR||13,903 m3||372 m3||6.9 m3|
aDECON decommissioning method.
Other methods would have smaller volumes of Class A and B wastes; Class C and GTCC
wastes volumes would not change for other methods. A plant that has operated 60 years would have essentially the same waste volumes and classifications.
bGTCC = greater than Class C.
Source: NUREG/CR-5884 and NUREG/CR-6174.
Note: 1 m3 _ 35.3 ft3.
The radionuclides of most importance for determining the classification of these LLWs are those that have relatively long half-life periods, such as 59Ni and 94Nb. These are also the radionuclides that accumulate in proportion with the length of reactor operation. The estimates in Table 7.7 are made for a plant that has operated 40 years. A plant that has operated 60 years would have essentially the same decommissioning waste volumes and classifications. Because the radionuclide concentration differences among waste classes are large (factors of 10 or more) and because the concentrations of radionuclides increase by no more than 50 percent, few components would be advanced to a higher classification by an additional 20 years of operations. Because the decommissioning waste volumes and classifications are essentially unchanged by an additional 20 years of plant operation, the Commission finds that the environmental impacts of decommissioning waste due to license renewal are of small significance. Measures employed within the context of ALARA, as discussed in Section 126.96.36.199, have the potential to reduce slightly the volume of LLW generated by decommissioning. The impact on decommissioning waste management attributable to license renewal is a Category 1 issue.
7.3.3 Air Quality Impacts
Air quality impacts of decommissioning are expected to be negligible. No major land disturbance for construction laydown or temporary waste storage areas is anticipated. The principal air quality impacts would result from motor vehicles operated by workers for transportation on-site and for movement of people and materials to and from the site. Most decommissioning activities would be conducted inside the containment, the auxiliary building, and the fuel-handling buildings. Because there would be a possibility of airborne releases of radioactivity within these buildings during decommissioning, releases to the ambient environment would be controlled. These impacts would be much smaller than those associated with construction or demolition of the facilities on-site and would not change with 20 additional years of operation. License renewal and an additional 20 years of reactor operation will have no impact on air quality during decommissioning; thus the impact of license renewal on decommissioning air quality impacts is of small significance for all plants. Because license renewal does not affect the level of air pollution during decommissioning, there is no need for the consideration of mitigation as part of the license renewal environmental review. The impact of decommissioning on air quality attributable to license renewal is a Category 1 issue.
7.3.4 Water Quality Impacts
The principal water quality impacts expected from decommissioning are those associated with sanitary sewer operations. Because the decommissioning work force is likely to be smaller than those of construction and certain operational activities (see Section 7.3.7), no increase in water quality impacts is expected. Soil erosion and chemical spills associated with increased site activities during decommissioning have the potential to degrade water quality, but such effects are readily controllable. The potential for significant water quality impacts from erosion or spills is no greater if decommissioning occurs after a 20-year license renewal instead of after the original 40 years of operation. Measures to minimize occupational and public radiation exposure will also protect water quality. License renewal and an additional 20 years of reactor operation will have no impact on water quality during decommissioning; thus the impact is of small significance. Because license renewal does not affect water quality impacts during decommissioning, there is no need for the consideration of mitigation as part of the license renewal environmental review. The impact of decommissioning on water quality impacts attributable to license renewal is a Category 1 issue.
7.3.5 Ecological Impacts
Terrestrial biota impacts, if any, would be associated with land disturbance for laydown or temporary waste storage areas, and no such land disturbance is anticipated. No direct impacts to aquatic biota are expected from routine decommissioning activities. Measures employed to protect water quality will also prevent toxic effects to aquatic organisms from liquid effluents. Therefore, the ecological impacts associated with decommissioning are not expected to vary with the length of time the plant is operated. Decommissioning after a 20-year license renewal would have the same ecological impacts, if any, as decommissioning after 40 years of operation; thus the impact is of small significance. Because license renewal does not affect ecological impacts during decommissioning, there is no need for the consideration of mitigation as part of the license renewal environmental review. The impact of decommissioning on ecological resources attributable to license renewal is a Category 1 issue.
7.3.6 Economic Impacts
In general, the nature of the activities and the elements of the costs associated with decommissioning are well understood, and the necessary skills and equipment should be readily available when needed. Table 7.8 lists percentage estimates of total costs for decommissioning large PWR and BWR reactors by the DECON method.
A 1991 national survey had estimates that averaged $218 million per 1000 MW for a PWR reactor and $283 million per 1000 MW for a BWR. The standard deviation was $74 million for PWRs and $144 million for BWRs. For both types of reactors, the range for plus and minus one standard deviation was $131 million to $350 million (OTA-E-575). These varying estimates reflect the uncertainty of projecting costs well into the future. Additionally, the unique aspects of a plant's design and operating history can affect decommissioning costs (e.g., Three Mile Island Unit 2 and Fort St. Vrain).
The largest cost category is "undistributed"; the largest component of this cost is utility support staff. The timing of decommissioning could influence disposal costs depending on the price of disposal services. The current trend is steeply increasing cost per units of radioactive waste disposal. If this trend continues over the long run, then one effect of license renewal could be to increase decommissioning costs. However, disposal costs should stabilize by the time that most existing plants would be eligible for license renewal. If this is the case, license renewal would have a minimal effect on the undiscounted costs of decommissioning after a 20-year extended operation period, compared with after 40 years of operation.
For the cost estimates included in Table 7.8, doubling the cost per cubic foot of waste disposal would increase total decommissioning costs by about 13 percent for PWRs and 20 percent for BWRs. The assumed rate charged for disposal would have to increase by a factor of about 6 to double the total cost of decommissioning. If the rate of disposal costs turns out to be significantly more than has been assumed in decommissioning cost estimates, there would tend to be significantly more attention devoted to volume reduction; thus, total cost of disposal would tend to increase less than the proportional increase in the rate charged per cubic foot (NUREG/CR-5884, vol. 1, pp. 3.56, 3.57, and NUREG/CR-6174, vol. 1, p. 3.55).
The timing of decommissioning could also affect costs if progress in robotics technology reduces costs and worker radiation exposure. This progress would affect a relatively small part of the decommissioning process and thus is unlikely to reduce the total cost of decommissioning significantly; however, it could result in substantial dose reductions.
Table 7.8 Summary and distribution of decommissioning costs for large pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) (thousands of 1993 dollars)
of total cost
($ x 103)
|Present valueh of
savingsi for license renewal|
($ x 103)
aPreshutdown period not included in duration
bIncludes direct decommissioning labor and materials for chemical decontamination of systems, cleaning of surfaces, and waste water treatment.
cIncludes direct labor and materials costs of removal.
dIncludes direct costs of waste disposal packages.
eIncludes cask rental costs and transportation costs.
fIncludes all costs of disposal at the LLW disposal facility.
gIncludes all costs that are period-dependent--e.g., commissioning operations contractor (DOC) mobilization/demobilization, utility and DOC overhead staff, nuclear insurance, regulatory costs, plant power usage, taxes, laundry services, environmental monitoring. Most of the undistributed costs are for staffing.
hAt 3 percent discount rate.
iThe decommissioning costs have been discounted at a rate of 3 percent real (assumes no inflation). At this rate, delaying decommissioning by the 20-year period of license renewal saves about 45 percent of the decommissioning cost; however, present value total costs have been figured at 2.5 years from final plant shutdown, resulting in savings from license renewal of about 40 percent.
Source: Tables 3.1 and 4.1 and pp. 3.59, 4.13, and 5.13 of NUREG/CR-5884, Vol. 1; Tables 3.1 and 4.1 and pp. 3.58, 4.12, and 5.11 of NUREG/CR-6174, Vol. 1.
The preceding sections show that there is no reason to expect the physical requirements of decommissioning to be materially different when comparing the base case to a 20-year extended operation period. The undiscounted economic costs, although uncertain, should also be relatively stable and thus unaffected by license renewal. However, because of financial considerations, the timing of decommissioning costs is important. To compare costs of activities that occur at different times, it is necessary to discount these costs to a common point in time. This is accomplished through present worth calculations, which account for the real opportunity cost or time value of money. Delaying decommissioning will allow any funds accumulated for this purpose to earn a return over the additional 20 years of license renewal and thus to reduce the present value of the decommissioning costs. The reduction in the present value is a function of the delay (license renewal period) and the time value of money, so the present value would be reduced by the same amount even if no fund were established and decommissioning were financed with borrowed money at the end of the plant operations. Regardless of how it is financed, the present value of delaying decommissioning costs will result in significant financial cost savings if a positive real discount rate is assumed.
Because total decommissioning costs are uncertain, the amount of financial savings that results from delaying decommissioning is also uncertain. Higher-than-expected decommissioning costs would result in higher cost savings resulting from delaying these costs, and vice versa. At a 3 percent real (i.e., above general inflation) discount rate, the present value savings associated with license renewal is about 40 percent of decommissioning costs (Table 7.8). Real cost increases, which might occur for waste disposal costs, could reduce the cost advantage of license renewal, but waste disposal costs are expected to stabilize before the current licenses of most plants expire. The impact of license renewal on decommissioning costs is not a consideration in the environmental review and decision whether to renew a license.
7.3.7 Socioeconomic Impacts
Socioeconomic impacts associated with decommissioning will be induced by the net change in the labor force as incoming decommissioning workers replace emigrating operations workers. The nature of these impacts will depend on the vitality of local economic activity at the time of decommissioning.
One of the difficulties of attempting to evaluate the socioeconomic impacts of decommissioning in year 40 of a plant's life compared with decommissioning in year 60 relates to the uncertainties about the size of the work force required. The largest nuclear power plant decommissioned to date has been the 150-MW(e) Shippingport Station (Section 7.2.3), which required an average work force during the peak year of approximately 230 workers (DOE/SSDP-0081); this work force was larger than the estimated work forces for very large power plants examined in studies prepared before the Shippingport experience (NUREG/CR-0130, Table 9.1-1; NUREG/CR-0672, Table 9.1-3). Because more-recent manpower estimates for large nuclear power plants are not available, the actual work force required in the future might be substantially larger than currently expected.
If the decommissioning process requires a smaller work force than the on-site operating staff and if the local economy is stable or declining, the result could be economic hardships, including declining property values and business activity, and problems for local government as it adjusts to lower levels of tax revenues. However, even this reduced work force will tend to mitigate temporarily the full adverse socioeconomic effects of terminating operations.
If there is a net reduction in the community work force but the economy is growing, the adverse impacts of this ongoing growth (e.g., housing shortages and school overcrowding) could be reduced.
If the decommissioning work force were substantially larger than the operational work force, the result could be increased demand for housing and public services but also increased tax revenues and higher real estate values. If the economy is characterized by decline, decommissioning could temporarily reverse the adverse economic effects.
In a stable economy, a net increase in the community work force could lead to some shortages in housing and public services, as well as to the higher tax revenues and real estate values mentioned previously. In a growing economy, decommissioning could act as an exacerbating factor to the ongoing shortages that already might exist.
Although the staff cannot project with certainty either the size of the required decommissioning work force or the state of the local economy at the time of decommissioning, the staff has assumed that the baseline conditions will be negligibly different in year 40, compared with year 60. Therefore, the staff expects that the socioeconomic impacts of decommissioning would be essentially similar whether that action were taken in year 60 or in year 40. The impact of license renewal on the socioeconomic impacts of decommissioning are of small significance. Because license renewal does not affect the socioeconomic impacts that will occur at the time of decommissioning, there is no need for the consideration of mitigation as part of the license renewal environmental review. The impact of decommissioning on socioeconomic resources attributable to license renewal is a Category 1 issue.
The physical requirements and attendant effects of decommissioning nuclear power plants after a 20-year license renewal are not expected to differ from those of decommissioning at the end of 40 years of operation. Decommissioning after a 20-year license renewal would increase the occupational dose no more than 0.1 person-rem (compared with 7,000 to 14,000 person-rem for DECON decommissioning at 40 years) and the public dose by a negligible amount (Section 7.3.1). License renewal would not increase to any appreciable extent the quantity or classification of LLW generated by decommissioning (Section 7.3.2). Air quality, water quality, and ecological impacts of decommissioning would not change as a result of license renewal (Sections 7.3.3, 7.3.4, and 7.3.5). There is considerable uncertainty about the cost of decommissioning; however, while license renewal would not be expected to change the ultimate cost of decommissioning, it would reduce the present value of the cost (Section 7.3.6). The socioeconomic effects of decommissioning will depend on the magnitude of the decommissioning effort, the size of the community, and the other economic activities at the time, but the impacts will not be increased by decommissioning at the end of a 20-year license renewal instead of at the end of 40 years of operation (Section 7.3.7). Incremental radiation doses, waste management, air quality, water quality, ecological, and socioeconomic impacts of decommissioning due to operations during a 20-year license renewal term would be of small significance. No mitigation measures beyond those provided by ALARA are warranted within the context of the license renewal process. The impacts of license renewal on radiation doses, waste management, air quality, water quality, ecological resources, and socioeconomics impacts from decommissioning are Category 1 issues.
DOE/EP-0093, Energy Technology Characterizations Handbook, Environmental Pollution and Control Factors, U.S. Department of Energy, Washington, D.C., March 1983.
DOE/RW-0006, Rev. 5, Vol. 1, Integrated Data Base for 1989: Spent Fuel and Radioactive Waste Inventories, Projections, and Characteristics, prepared by Oak Ridge National Laboratory, Oak Ridge, Tennessee, for U.S. Department of Energy, Washington, D.C., November 1989.
DOE/RW-0006, Rev. 6, Integrated Data Base for 1990: Spent Fuel and Radioactive Waste Inventories, Projections, and Characteristics, prepared by Oak Ridge National Laboratory, Oak Ridge, Tennessee, for U.S. Department of Energy, Washington, D.C., November 1990.
DOE/SSDP-0081, Final Project Report Shippingport Station Decommissioning Project, prepared by Westinghouse Hanford Company, Shippingport Station Decommissioning Project Office, for the U.S. Department of Energy, Richland Operations Office, Richland, Washington, December 1989.
"France," Power in Europe 88, 14, December 6, 1990.
Gaunt, J., et al., Decommissioning of Nuclear Power Facilities, Energy Saves Paper No. 28, The World Bank, Washington, D.C., April 1990.
Moore, T., "Milestone Achieved in Nuclear System Decontamination," EPRI J., 28-32, November/December 1995.
NUREG-0586, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, U.S. Nuclear Regulatory Commission, Office of Regulatory Research, August 1988.
NUREG/CR-0130, Vol. 1, R. I. Smith, et al., Technology, Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station, prepared by Pacific Northwest Laboratory, Richland, Washington, for U.S. Nuclear Regulatory Commission, June 1978.
NUREG/CR-0130, Addendum 4, G. J. Konzek and R. I. Smith, Technology, Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station's Technical Support for Decommissioning Matters Related to Preparation of the Final Decommissioning Rule, prepared by Battelle, Pacific Northwest Laboratory, Richland, Washington, for U.S. Nuclear Regulatory Commission, July 1988.
NUREG/CR-0672, H. D. Oak, et al., Technology, Safety and Costs of Decommissioning a Reference Boiling Water Reactor Power Station, prepared by Pacific Northwest Laboratory, Richland, Washington, for U.S. Nuclear Regulatory Commission, June 1980.
NUREG/CR-0672, Addenda 3 and 4, G. J. Konzek and R. I. Smith, Technology, Safety and Costs of Decommissioning a Reference Boiling Water Reactor Power Station: Comparison of Two Decommissionings Cost Estimates Developed for the Same Commercial Nuclear Reactor Power Station, prepared by Pacific Northwest Laboratory, Richland, Washington, for U.S. Nuclear Regulatory Commission, December 1990.
NUREG/CR-5491 (PNL-7191), R. P. Allen and A. B. Johnson, Shippingport Station Aging Evaluation, U.S. Nuclear Regulatory Commission, Washington, D.C., January 1990.
NUREG/CR-5884 (PNL-8742), J. G. Konzek, et al., Revised Analyses of Decommissioning of the Reference Pressurized Water Reactor Power Station (draft report for comment), Vols. 1 and 2, prepared by Pacific Northwest Laboratory, Richland, Washington, for U.S. Nuclear Regulatory Commission, October 1993.
NUREG/CR-6174 (PNL-9975), R. I. Smith, et al., Revised Analyses of Decommissioning for the Reference Boiling Water Reactor Power Station (draft report for comment), Vols. 1 and 2, prepared by Pacific Northwest Laboratory, Richland, Washington, for U.S. Nuclear Regulatory Commission, September 1994.
OTA-E-575, Aging Nuclear Power Plants: Managing Plant Life and Decommissioning, U.S. Congress, Office of Technology Assessment, Washington, D.C., September 1993.