United States Nuclear Regulatory Commission - Protecting People and the Environment

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


5. Environmental Impacts of Postulated Accidents

[ Prev | Next | Table of Contents ]


5.1 Introduction

[ Prev | Next | Table of Contents ]

This section discusses each aspect of postulated accidents that is normally treated in the final environmental statements (FESs) for the operation of nuclear power plants. Methodologies that estimate future risks at each existing nuclear power plant site in the United States are developed in Section 5.3.3, considering an additional 20-year period of operation beyond the current license term.

The characteristics of nuclear power plant accidents are discussed (Section 5.2.1) to acquaint the reader with (1) the sources of radiation from postulated accidents, (2) the potential pathways of radiation to the environment, and (3) the possible health effects of exposure to such accidental releases. Historical experience and observed impacts of nuclear power plant accidents are discussed next (Section 5.2.2), followed by a description of the various measures taken in the design and operation of a power plant to reduce the likelihood or consequences of an accident (Section 5.2.3).

The impacts of accident risks during a license renewal period are discussed in Section 5.3. A brief discussion of the primary concern arising from extending the operational life of nuclear power plants is provided (Section 5.3.1). This concern centers on the effects that plant aging and increasing population can have on the probability and consequences of accidents. Calculation of the estimated environmental impacts and risks due to postulated accidents during the license extension period is discussed in Sections 5.3.2 and 5.3.3. Consequences of design-basis and severe accidents are reviewed. The potential pathways for radiation release examined are (1) direct release to the atmosphere, (2) fallout on open bodies of water, and (3) groundwater. Existing severe accident analyses were reviewed and used to predict consequences at all sites. The potential economic impacts of accidents during the renewal period were also reviewed (Section 5.3.4). To maintain a perspective on the results of this analysis, a discussion of the uncertainties associated with the types of consequence analyses used in this evaluation is provided (Section 5.3.5). Finally, a discussion is given on the role of severe accident mitigation design alternatives (SAMDAs) in the license renewal process (Section 5.4).


5.2 Plant Accidents

[ Prev | Next | Table of Contents ]

5.2.1 General Characteristics of Accidents

The term "accident" refers to any unintentional event outside the normal plant operational envelope that results in a release or the potential for release of radioactive materials into the environment. Generally, the U.S. Nuclear Regulatory Commission (NRC) categorizes accidents as "design basis" (i.e., the plant is designed specifically to accommodate these) or "severe" (i.e., those involving multiple failures of equipment or function and, therefore, whose likelihood is generally lower than design-basis accidents but where consequences may be higher), for which plants are analyzed to determine their response. The predominant focus in environmental assessments is on events that can lead to releases substantially in excess of permissible limits for normal operation. Normal release limits are specified in the NRC's regulations (10 CFR Part 20 and 10 CFR Part 50, Appendix I).

Many features combine to minimize the risk of accidents at nuclear power plants. These features include high-quality reactivity control and reactor cooling systems and containment and backup safety systems to respond to equipment failure. The incorporation of safety into design, construction, and operation is to a very large extent devoted to minimizing the possibility of the release of radioactive materials from their normal places of confinement within the plant. Descriptions of these safety features are provided in each licensee's final safety analysis report (FSAR) and in the NRC's safety evaluation report.

The plant design, including the types and number of safety systems, takes into consideration the specific locations of radioactive materials within the plant; their amounts; their nuclear, physical, and chemical properties; and their potential for transport into the environment and for creating health hazards.

5.2.1.1 Fission Product Characteristics

By far the largest inventory of radioactive material in a nuclear power plant is produced as a by-product of the fission process and is located in the uranium oxide fuel pellets in the reactor core in the form of fission products. During periodic refueling shutdowns, the assemblies containing these fuel pellets are transferred to a spent-fuel storage pool; the second largest inventory of radioactive material is located in this storage area. Much smaller inventories of radioactive materials are also normally present in the water that circulates in the reactor coolant system and in the systems used to process gaseous and liquid radioactive wastes in the plant.

Radioactive materials in power plants exist in a variety of physical and chemical forms. Their potential for dispersal into the environment depends not only on mechanical forces that might physically transport them, but also on their inherent properties, particularly their volatility. The majority of these materials exist as nonvolatile solids over a wide range of temperatures. Some, however, are relatively volatile solids, and a few are gaseous at normal temperatures and pressures. These characteristics have a significant bearing on the assessment of the environmental radiological impacts of accidents.

The gaseous materials include radioactive forms of the chemically inert noble gases krypton and xenon. These two gases have the highest potential for release into the atmosphere. If a reactor accident involving degradation of the fuel cladding were to occur, the release of substantial quantities of these radioactive gases from the fuel into the reactor cooling system would be virtually certain. Such accidents are low-frequency, but credible, events. For this reason, the safety analysis of each nuclear power plant incorporates a hypothetical design-basis accident that postulates the release of the entire contained inventory of radioactive noble gases from the fuel in the reactor into the containment structure. If the noble gases were further released to the environment as a result of failure to maintain the containment boundary, the hazard to individuals from these noble gases would arise predominantly through the external gamma radiation from the airborne plume. The reactor containment structure and containment support systems are designed to minimize the possibility of this type of release.

Radioactive forms of iodine are produced in substantial quantities in the fuel by the fission process, and in some chemical forms they may be quite volatile. For these reasons, iodine has traditionally been regarded as having a relatively high potential for release from the fuel into the containment during certain design-basis accidents. Because iodine concentrates in the thyroid gland, the release of radioiodines to the atmosphere is controlled by containment and by the use of special systems (i.e., filters) designed to retain the iodine.

The chemical forms in which the fission product radioiodines are found are generally solid materials at room temperatures; hence, they have a strong tendency to condense (or "plate out") on cooler surfaces. In addition, most of the iodine compounds are quite soluble in, or are chemically reactive with, water. Although these properties do not prevent the release of radioiodines from degraded fuel, they would act to inhibit release from the containment structure that has large internal surface areas and may contain large quantities of water as a result of an accident. The same properties affect the behavior of radioiodines that may "escape" into the atmosphere. Thus, if it rains during a release, or if there is moisture on exposed surfaces (for example, dew), the radioiodines will show a strong tendency to be absorbed by the moisture.

Other radioactive materials formed during the operation of a nuclear power plant are less volatile and have a much smaller tendency to escape from degraded fuel unless the temperature of the fuel becomes very high. Such materials tend to condense quite rapidly when they are transported to a lower temperature region or to dissolve in water when it is present. This mechanism can result in production of very small particles that can be carried some distance by a moving stream of gas or air. If such particulate materials are dispersed into the atmosphere as a result of containment failure, they tend to be carried downwind and deposited on surfaces by gravitational settling (fallout) or by precipitation (washout or rainout), where they will become "contamination" hazards in the environment.

All of these radioactive materials exhibit the property of radioactive decay with half-life periods ranging from fractions of a second to many days or years. Many of the radioactive materials decay through a sequence of decay processes, and all eventually become stable (nonradioactive). The radiation emitted during these decay processes renders the radioactive materials hazardous.

5.2.1.2 Meteorological Considerations

Two separate analyses of accident sequences are performed during the licensing process for a nuclear power plant. The first analysis is the determination of the consequences for design-basis accidents and is performed for the NRC's safety evaluation report. This analysis is performed to ensure that the doses to any individual at the exclusion area boundary over a period of 2 hours, or at the outer boundary of the low population zone (LPZ) during the entire period of plume passage, will not exceed the siting dose guidelines of 25 rem to the whole body or 300 rem to the thyroid, pursuant to 10 CFR Part 100. This analysis is used to examine site suitability (10 CFR Part 100) and the mitigative capability of certain plant safety features (10 CFR Part 50). The atmospheric dispersion model for this evaluation, as described in the NRC Regulatory Guide 1.145, uses on-site meteorological data (typically, a multiyear record) considered representative of the site and vicinity to calculate relative dilutions that will be exceeded no more than 0.5 percent of the time in any one sector (22.5° ) and no more than 5 percent of the time for all sectors (360° ) at the exclusion area boundary and LPZ. These dilution factors, because they provide little plume spread, ensure site-specific calculated doses that could be exceeded only 5 percent of the time.

The second analysis of accident consequences is generally found in the environmental documentation for the most recently licensed nuclear plants and considers a spectrum of releases, including those for severe accidents. Actual meteorological conditions from a representative 1-year period of record of on-site data are used in this environmental analysis. A detailed description of the atmospheric dispersion model used to estimate the environmental impacts of such accidents is contained in NUREG-75/014 (formerly WASH-1400), Appendix VI.

5.2.1.3 Exposure Pathways

The radiation exposure (hazard) to individuals is determined by the individual's proximity to the radioactive materials; the duration, intensity, and type (external versus internal) of exposure; and factors that act to shield the individual from the radiation. Many of the pathways for radiation and the transport of radioactive materials that lead to radiation exposure hazards to humans are the same for accidental as for "normal" releases. These pathways are depicted in Figure 5.1. Two additional possible pathways that could be significant for accident releases are not shown in Figure 5.1. One of these pathways is the fallout of radioactivity onto open water or onto land with runoff into open water bodies. The second pathway would be unique to an accident involving temperatures high enough to cause melting of the reactor core and subsequent penetration of the reactor vessel and underlying base mat by the molten core debris. Such an occurrence would create the potential for the release of radioactive material into the hydrosphere via groundwater beneath the plant. These pathways may lead to external exposure to radiation and also to internal exposure if radioactive material is contacted, inhaled, or ingested from contaminated food or water.

It is characteristic of all these pathways that during the transport of radioactive material by wind or water, the material tends to spread and disperse--like a plume of smoke from a smokestack--becoming less concentrated in larger volumes of air or water. The result of these natural processes is to lessen the intensity of exposure to individuals downwind or downstream of the point of release, but the processes also tend to increase the number who may be exposed. For a release into the atmosphere, the degree to which dispersion reduces the concentration in the plume at any downwind point is governed by the turbulence characteristics of the atmosphere, which vary considerably with time and from place to place. This fact, taken in conjunction with the variability of wind direction and the presence or absence of precipitation, means that accident consequences depend largely upon the

Figure 5.1 Potential exposure pathways to individuals.

weather conditions existing at the time of the accident.

5.2.1.4 Adverse Health Effects

The cause-and-effect relationships between radiation exposure and adverse health effects are quite complex. Whole-body radiation exposure resulting in a dose greater than about 25 rem over a short period of time (hours) is necessary before any physiological effects to an individual are clinically detectable1 shortly thereafter. Doses about 10 to 20 times larger, also received over a relatively short period of time (hours to a few days), can be expected to cause some fatalities. At the severe (but extremely low probability) end of the accident spectrum, very high exposures of these magnitudes are theoretically possible for persons in proximity to the plant if measures are not or cannot be taken to provide protection, such as by sheltering or evacuation.

Lower levels of exposures may constitute a longer-term health risk. The effects of such exposures may include randomly occurring cancer in the exposed population and genetic changes in future generations after exposure of a prospective parent. Relating a given health effect to a known exposure to radiation is most often not possible because of the many other possible causes for such effects. For this reason, it is necessary to assess radiation-induced cancer effects on a statistical basis.

Occurrences of cancer in the exposed population may begin to develop only after a lapse of 2 to 15 years (latent period) from the time of exposure and continue over a period of over 40 years (plateau period). However, in the case of exposure of fetuses (in utero), occurrences of cancer may begin to develop at birth (no latent period) and end at age 10 (that is, the plateau period is 10 years). The occurrence of cancer itself is not necessarily indicative of a fatality because the ratio of mortality to incidence of cancer depends upon the cancer type and advances in medical treatment.

The estimates of health consequences used for latent fatalities in this document are the same estimators used in the development of the revised 10 CFR 20 regulations. A discussion is provided in Sections 3.8 and 4.6, and a more detailed discussion is provided in Section E.4, which details the recent developments in radiation risk estimation that lead to the health-consequence risk estimates in this section. The discussion in Section E.4 includes background information about epidemiology as well as health-effects information compiled by the United Nations Scientific Committee on the Effects of Atomic Radiation, by the National Academy of Sciences (reports of Advisory Committees on the Biological Effects of Ionizing Radiation--BEIR-I, BEIR-III, BEIR-V), and by the International Commission on Radiological Protection. The risk estimates for fatal cancers are considered to be nominally 500 per million person-rem, consistent with the risk factors described earlier (Section 3.8.1.3 and Appendix E.4). In addition, approximately 100 genetic disorders per million person-rem are projected for the succeeding generations.

5.2.1.5 Avoiding Adverse Health Effects

Radiation hazards in the environment disappear by the natural process of radioactive decay. Where the decay process is a slow one, however, and where the material becomes relatively fixed in its location as an environmental contaminant (such as in soil), the hazard can continue to exist for a long period of time--months, years, or even decades. Thus, a possible environmental societal reaction to severe accidents is avoidance of the potential health hazards by restrictions on the use of the contaminated property or contaminated foodstuffs, milk, and drinking water.

5.2.2 Accident Experience and Observed Impacts

A limited number of accidents have been recorded in the experience data of the world's nuclear programs. The United States, Great Britain, and the Soviet Union have all experienced accidents of varying magnitudes and consequences. The following paragraphs will discuss first the United States experience, followed by the British and then the Soviet accident experience.

As of September 1990, 112 commercial nuclear power reactor units were licensed for operation in the United States (Table 2.1) with power-generating capacities ranging from 72 to 1270 MW(e). The combined experience with these operating units represents approximately 1300 reactor-years (RYs) of operation over an elapsed time of about 28 years. [An additional 6 plants, with individual generating capacities of up to 1314 MW(e), are expected to receive an operating license within the next 10 years.] Several of these facilities have experienced accidents (ORNL 1980; NUREG-0651; Thompson and Beckerley 1964), some of which have resulted in small releases of radioactive material to the environment. None is known to have caused any radiation injury or fatality to any member of the public, nor any significant contamination of the environment. Although the experience base with light-water reactors (LWRs) having containments such as those licensed in the United States is not large enough to permit reliable statistical prediction of accident probabilities, it does, however, suggest that significant environmental impacts caused by accidents are not at all likely to occur over time periods of a few decades.

Melting or severe degradation of reactor fuel has occurred in only one U.S.-licensed commercial LWR--during the accident at Three Mile Island Unit 2 (TMI-2) on March 28, 1979. It has been estimated that about 2.5 million Ci of noble gases (about 0.9 percent of the core inventory) and 15 Ci of radioiodine (about 0.00003 percent of the core inventory) were released to the environment at TMI-2 (NUREG/CR-1250).2 No other radioactive fission products were released in measurable quantities. It has been estimated that the maximum cumulative off-site radiation dose to an individual was less than 100 mrem (NUREG/CR-1250; President's Commission 1979). The total population exposure has been estimated to be in the range from about 1000 to 5000 person-rem. (This range is discussed on page 2 of NUREG-0558.) This exposure could statistically produce between zero and one additional fatal cancer over the lifetime of the exposed population of approximately 2 million in the site area. The same population receives about 240,000 person-rem each year from natural background radiation, and approximately a half million cancers are expected to develop in this group over the lifetime of the population (NUREG/CR-1250; President's Commission 1979), primarily from causes other than radiation. Trace quantities (barely above the limit of detectability, below allowable limits, and less than that from fallout due to nuclear tests) of radioiodine were found in a few samples of milk produced in the area. No other food or water supplies were affected.

Accidents at U.S. nuclear power plants have also caused occupational injuries and a few occupational fatalities, but these were not due to radiation exposure. Individual worker exposures have ranged up to about 4 rem as a direct consequence of reactor accidents (although there have been higher exposures to individual workers as a result of other unusual occurrences). However, the collective worker exposure levels (person-rem) from accidents are a small fraction of the exposures experienced during routine operation; during the 1982-1986 time period, routine exposures ranged from 23 to 2880 person-rem/year in pressurized-water reactors (PWR) and 84 to 4080 person-rem/year in boiling-water reactors (BWR) per RY (NUREG-0713).

Accidents have also occurred at other nuclear facilities in the United States and in other countries (ORNL 1980; Bertini 1980). Reactor fuel melted in at least seven of these accidents: Fermi 1 (Lagoona Beach, Michigan), St. Laurent (France), NRX Reactor (Chalk River, Canada), Experimental Breeder Reactor 1 (Idaho), Heat Transfer Reactor Experiment Facility (Idaho), Westinghouse Reactor (Waltz Mills, Pennsylvania), and Oak Ridge Research Reactor (Tennessee). Because of inherent differences in design, construction, operation, and purpose of these other facilities, their accident record has only indirect relevance to current nuclear power plants. The most relevant accident was the one in 1966 at Enrico Fermi Atomic Power Plant Unit 1. Fermi Unit 1 was a sodium-cooled fast breeder demonstration reactor designed to generate 61 MW(e). The damages were repaired and the reactor reached full power 4 years after the accident. It operated successfully and completed its mission in 1973. The Fermi accident did not release any measurable radioactivity to the environment.

A reactor accident in 1957 at Windscale, England (renamed Seascale), released a significant quantity of radioiodine, approximately 20,000 Ci, to the environment and minor quantities of 137Cs, 89Sr, and 90Sr (Eisenbud 1987) and 240Po (Crick and Linsley 1983). This reactor, which was not operated to generate electricity, used a graphite core design and circulated air rather than water to cool the uranium fuel. During a special operation to heat the large amount of graphite in this reactor (an operation normal for this graphite-moderated reactor), the fuel overheated and radioiodine and noble gases were released directly to the atmosphere from a 123-m (405-ft) stack. Milk produced in a 518-km2 (200-mile2) area around the facility was impounded for up to 44 days. The United Kingdom National Radiological Protection Board (1957) estimates that the releases may have caused as many as 260 cases of thyroid cancer, about 13 of them fatal, and as many as 7 deaths from other cancers or hereditary diseases. This kind of accident cannot occur in a water-moderated and -cooled reactor like those in the U.S. nuclear power program.

On April 26, 1986, a major accident occurred at reactor 4 of the Chernobyl Nuclear Power Station in the Soviet Union. This reactor differs substantially from LWRs licensed to operate in the United States. The initiating event, a reactivity insertion, was recognized as a potential problem early in U.S. power reactor design; consequently, licensed U.S. power reactors are designed to prevent or accommodate occurrences of reactivity insertions. A major difference in safety between the U.S. designs and Chernobyl is that the Chernobyl reactor did not have a containment similar to those found on U.S. reactors. Also, the Chernobyl plant, which used graphite as a neutron moderator rather than water as with U.S. designs, had a positive power coefficient for the off-normal plant conditions that were present at the time of the accident. Thus, the accident has only indirect relevance to LWRs. The release of radioactive material from the accident was initially reported by the Soviets to be about 100 million Ci of fission products, but (except for the noble gases) that estimate included only material deposited within the European part of the Soviet Union. As a result of the accident, radionuclides were deposited throughout the Northern Hemisphere.

Of the almost 3 billion people in the Northern Hemisphere receiving Chernobyl radiation, about 800 million people account for 97 percent of the total risk increment. The remaining 3 percent of the dose commitment in Asia and North America represents a minuscule risk increment. Outside of the 30-km (19-mile) zone surrounding Chernobyl, the incremental increase in fatal cancer risk is a fraction of a percent and is not likely ever to be detected epidemiologically (DOE/ER-0332; Goldman 1987).

5.2.3 Mitigation of Accident Consequences

5.2.3.1 Design Features

All U.S. power reactors contain system features designed to prevent accidental release of radioactive fission products from the fuel and to lessen the consequences should such a release occur. Many of the design and operating specifications of these features are derived from the analysis of postulated events known as design-basis accidents. These accident-preventing and -mitigating system features are collectively referred to as "engineered safety features." Safety injection systems are incorporated to provide cooling water to the reactor core during a loss-of- coolant accident to prevent or minimize fuel damage. Heat-removal capability is provided inside the containment to prevent containment failure from overpressure. Long-term decay heat removal systems are also provided to remove decay heat from the core. All the mechanical systems mentioned above are supplied with emergency power from on-site diesel generators in the event that normal off-site station power is interrupted.

Containment structures are used as a mitigating system to provide a nearly leaktight barrier to minimize the escape of fission products to the environment in the event of a fission product release inside containment. These structures are designed to withstand the internal pressure and temperature associated with design-basis accidents.

The fuel-handling structures also have accident-mitigating systems. Spent fuel is handled and stored under water, which would tend to greatly reduce the amount of radioactive material released to the building environment in the event of fuel failure. A safety-grade exhaust air ventilation subsystem contains both charcoal and high-efficiency particulate filters. The ventilation systems are also designed to keep the area around the spent-fuel pool below the prevailing barometric pressure during fuel-handling operations to minimize the outleakage through building openings. Upon detection of high radiation, exhaust air is routed through the filter units, and radioactive iodine and particulate fission products which escaped from the spent fuel pool would be removed from the flow stream before exhausting to the atmosphere.

Much more extensive discussions of the safety features and characteristics of a particular plant may be found in the FSAR for that plant. In addition, the implementation of the lessons learned from the TMI-2 accident--in the form of improvements in design, procedures, and operator training--has significantly reduced the likelihood of a degraded-core accident that could result in large releases of fission products to the containment. These TMI-2-related requirements are specified in NUREG-0737.

5.2.3.2 Site Features

The NRC's site criteria, found in 10 CFR Part 100, require that every power reactor site have certain characteristics that tend to reduce the risk and potential impact of accidents. First, the site must have an exclusion area around the reactor. A typical exclusion area radius is about 0.8 km (0.5 mile). No residents are allowed in the exclusion area. Public transportation routes and other public activities are allowed within the exclusion area, but these routes and activities must be demonstrated to be controllable in the event of an emergency. Second, beyond and surrounding the exclusion area is an LPZ. A typical LPZ radius is about 5 km (3 miles). Within this zone, the licensee must ensure that there is a reasonable probability that appropriate protective measures could be taken on behalf of the residents and other members of the public in the event of a serious accident. Third, 10 CFR Part 100 requires that the distance from the reactor to the nearest boundary of a densely populated area containing more than 25,000 residents be at least one and one-third times the distance from the reactor to the outer bound of the LPZ.

5.2.3.3 Emergency Preparedness

Each licensee is required to establish emergency preparedness plans to be implemented in the event of an accident, including protective action measures for the public. The NRC, as well as other federal and state regulatory agencies, review the subject plans to ensure that the condition of on- and off-site emergency preparedness provides reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. Among the standards that must be met by these plans are provisions for two emergency planning zones (EPZs). A plume exposure pathway EPZ (requiring preplanned evacuation procedures) of about 16 km (10 miles) in radius and an ingestion exposure pathway EPZ (where interdiction of foodstuffs is planned) of about 80 km (50 miles) in radius are required. Other standards include appropriate ranges of protective actions for each of these zones; provisions for dissemination to the public of basic emergency planning information; provisions for rapid notification of the public during a serious reactor emergency; and methods, systems, and equipment for assessing and monitoring actual or potential off-site consequences in the event of a radiological emergency condition.


5.3 Accident Risk and Impact Assessment for License Renewal Period

[ Prev | Next | Table of Contents ]

5.3.1 Regulatory Interface Between License Renewal and Accident Impacts

In general, the safety and environmental issues associated with license renewal fall into two general categories: (1) effects of aging on the physical plant itself and the associated impact of these effects on accident frequency and radiological releases and (2) effects on accident consequences due to the changing environment in which the plant exists.

The potential effects of age on the physical plant are addressed through engineering and research programs. Potential deterioration of plant components and structures due to physical processes such as corrosion, erosion, mechanical wear, and embrittlement could result in the increased likelihood of component or structure failure. These increased failure rates, in turn, could lead to a higher frequency of accidents or more severe consequences. Therefore, control of these effects is necessary if the plant is to be assured of continuing to operate in a safe manner. As a result, NRC has developed the license renewal process within the context of the aging issue. The process will provide assurance that these age-related impacts are controlled and adequate protection of the health and safety of the public is maintained during the 20-year license renewal period. To supplement the control that the normal regulatory process has over the aging effects on the plant, the NRC requires that the renewal applicant specifically address the issue of age-related degradation by identifying, in an integrated plant assessment process, those structures and components which are susceptible to age-related degradation and whose functions are necessary to ensure that the facility's licensing basis is maintained. The licensee will further be required to demonstrate that the effects of aging will be adequately managed so that the intended functions of these structures and components will be maintained for the period of extended operation. The combined impact of these actions should be to provide high confidence that significant incremental increases in public risk will not result from aging effects related to the plant. A comprehensive discussion of the NRC rule, programs, and activities to provide this assurance is found in 55 FR 29043, dated July 17, 1990.

In assessing the impact on the environment from postulated accidents during the license renewal period, the assumption has been made that the license renewal process will ensure that aging effects on the plant are controlled and that the probability of any radioactive releases from accidents will not increase over the license renewal period.

The effects due to the changing environment around the plant during the license renewal period are less predictable, are generally not subject to regulatory controls, and could cause an increase in public risk as the plant continues to operate. These effects manifest themselves primarily by increasing the consequences of a given accident. For example, as the general population in the vicinity of a nuclear plant increases, the number of persons that could be affected by an accidental release also increases. Because these impacts are "noncontrollable," their potential for increasing risk as well as the magnitude of any such increase in risk must be specifically examined. Such an examination is presented in the following sections, which will discuss and assess the potential adverse impacts to the environment from postulated accidents during the license renewal period.

5.3.2 Design-Basis Accidents

Two classes of accidents are evaluated. The first class of accidents, design-basis accidents, is discussed in this section. The second, severe accidents, is discussed in Section 5.3.3. As noted previously, design-basis accidents are those that both the licensee and the NRC staff evaluate to ensure that the plant meets acceptable design and performance criteria. The environmental impacts of design-basis accidents are evaluated during the initial license process, and the ability of the plant to accommodate these accidents is demonstrated to be acceptable before issuance of the license. The results of these evaluations are found in license documentation such as FESs and FSARs. The licensee is required to maintain these acceptable design and performance criteria throughout the life of the plant, including any extended-life operation. The consequences for these events are evaluated for the hypothetical maximum exposed individual; as such, changes in the plant environment will not affect these evaluations. Because of the requirements that continuous acceptability of the consequences and aging management programs be in effect for license renewal, the environmental impacts as calculated for design-basis accidents should not differ significantly from initial licensing assessments over the life of the plant, including the license renewal period. In addition, any refurbishment necessary to prepare for license renewal would be done in a fashion consistent with the limits set for design-basis accidents and would not alter their consequences. Accordingly, the design of the plant relative to design-basis accidents during the extended license period is considered to remain acceptable and the environmental impacts of those accidents will not be examined further in this section.

5.3.3 Probabilistic Assessment of Severe Accidents

This section presents the staff's assessment of impacts of severe accidents during the license renewal period. Methodologies were developed to evaluate each of the dose pathways by which a severe accident may result in adverse environmental impacts and to estimate off-site costs of severe accidents.

Three pathways for release of radioactive material to the environment from severe accidents are evaluated in this section for each plant site for the license renewal period. These pathways are (1) air, (2) air to surface water, and (3) groundwater to surface water. For most plants, the air pathway represents the most likely pathway for significant dose to the public. The air to surface water pathway is significant for only a few sites that are close to large but confined bodies of water. The third pathway represents a less significant potential for dose because of reduction in radioactivity due to retention in the ground and greater flexibility and time to implement interdiction measures. Separate methodologies were developed for quantifying the potential consequences resulting from each pathway for each site. Economic impacts from severe accidents during the license renewal period are also described in this chapter.

Section 5.3.3.1 reviews the existing analyses available for use in this Generic Environmental Impact Statement (GEIS) study; Section 5.3.3.2 examines the effects of atmospheric releases, including vegetation pathways; Section 5.3.3.3 examines the effects from direct fallout onto open bodies of water; Section 5.3.3.4 reviews effects from releases to groundwater; and Section 5.3.3.5 examines the economic impacts of severe accidents. All analyses will adhere to a process that uses the results of existing analyses and site-specific information to conservatively predict the environmental impacts of severe accidents for each plant during the license renewal period.

5.3.3.1 Review of Existing Impact Assessments

The public risk due to nuclear power accidents has a range of values. The staff believes that current regulatory practices ensure that the basic statutory requirement, adequate protection of the public, is met (51 FR 28044). These risk estimates are representative of the magnitude of risk associated with current regulatory practices. Since the early 1970s, there have been increasing efforts to determine severe accident risks more precisely and on a plant-specific basis. The first comprehensive plant-specific examination of risk was the Reactor Safety Study (RSS), published in 1975 (NUREG 75/014, formerly WASH-1400). The risk values calculated in RSS were later updated in NUREG-0773 and used in NRC FESs published after 1980. Later, more complex and more intensive plant-specific risk studies were developed, both by NRC and the industry. The most recent NRC studies of severe accident consequences are found in the NUREG-1150 analyses. To date, about 40 percent of the 118 operating plants and plants under construction have had some level of plant-specific risk analysis reviewed by NRC. This body of knowledge was used in the prediction of environmental impacts of severe accidents for all plants. Further details of these studies are provided in the following paragraphs.

FES reports since 1980 (Table 5.1) have provided assessments of impacts resulting from postulated severe accidents. Both the frequency and magnitude of the source terms ("source term" is a descriptive name for the releases of radioactive material to the environment under various accident conditions) for such assessments were usually taken from the rebaselined RSS (NUREG-0773). [These values are the result of updating the original RSS (NUREG-75/014, formerly WASH-1400) results using improved methods relative to the original WASH-1400 methodology.] Table 5.2 provides more information on the source-term data used in the FES analyses. These rebaselined source terms were then used with site-specific meteorological and demographic data to calculate off-site risk. A separate rebaselined set of source terms was provided for each of the two types of reactor designs, BWRs and PWRs. In most FES assessments, these same sets of data, without change, were used to evaluate off-site risks. Accordingly, the risk values provided in these FESs are based upon the plant designs analyzed in WASH-1400. As such, they do not represent plant-specific analyses for the FES plants but are sufficient to illustrate the general magnitude and types of risks that may occur from reactor accidents. There were some exceptions in that several studies included some further modification of the rebaselined RSS frequency estimates to better account for plant-specific design differences from the RSS plants. When available, other studies used plant-specific information on severe accident risks [such as probabilistic risk assessments (PRAs)]. Once the source-term data were established, all plants used the Calculation of Reactor Accident Consequences (CRAC) code to determine environmental consequences. Site-specific information regarding meteorology, population, and evacuation was used. Assumptions regarding exposure pathway, exposure limits, and plume behavior remained largely unchanged for all analyses.

The NUREG-1150 study is an NRC-sponsored risk examination of five U.S. nuclear power plants.3 These analyses used state-of-the-art technology in evaluation of source-term release frequency, source-term characteristics, and consequence evaluation. Efforts were made to explore uncertainties in accident frequency, containment behavior, and radioactive material release and transport so that from this distribution of results, mean values of risk could be determined. Source terms and frequencies specific to the plant were determined. Advanced computer codes were used. For example, the MELCOR Accident Consequence Code System (MACCS) computer code for consequence evaluation was used instead of CRAC.

Table 5.1 Available risk analyses associated with final environmental statements
Plant NSSSa vendor Plant size
[MW(e)]
Containment type NUREG document number NUREG date
Beaver Valley 2 W 836 Subatmospheric 1094 9-85
Braidwood 1, 2 W 1120 Large dry 1026 6-84
Byron 1, 2 W 1120 Large dry 0848 4-82
Callaway 1 W 1171 Large dry 0813 1-82
Catawba 1, 2 W 1145 Ice condenser 0921 1-83
Clinton 1 GE 933 Mark III 0854 5-82
Comanche Peak 1, 2 W 1150 Large dry 0775 9-81
Fermi 2 GE 1093 Mark I 0769 8-81
Grand Gulf 1, 2 GE 1250 Mark III 0777 9-81
Shearon Harris 1, 2 W 900 Large dry 0972 10-83
Hope Creek GE 1067 Mark I 1074 6-84
Indian Point 2, 3 W 873/965 Large dry b b
Limerick 1, 2 GE 1055 Mark II 0974 12-83
Millstone 3 W 1154 Subatmospheric 1064 12-84
Nine Mile Point 2 GE 1091 Mark II 1085 5-85
Palo Verde 1, 2, 3 CE 1270 Large dry 0841 2-82
Perry 1, 2 GE 1191 Mark III 0884 8-82
River Bend GE 936 Mark III 1073 1-85
San Onofre 2, 3 CE 1070/1080 Large dry 0490 4-81
Seabrook 1, 2 W 1198 Large dry 0895 12-82
South Texas 1, 2 W 1250/1251 Large dry 1171 8-86
St. Lucie 2 CE 830 Large dry 0842 4-82
Summer 1 W 900 Large dry 0719 5-81
Susquehanna 1, 2 GE 1050 Mark II 0564 6-81
Vogtle 1, 2 W 1101 Large dry 1087 3-85
Waterford 3 CE 1104 Large dry 0779 9-81
Wolf Creek 1 W 1170 Large dry 0878 6-82
WNP-2c GE 1100 Mark II 0812 12-81

aNSSS = nuclear steam supply system, W = Westinghouse, GE = General Electric, CE = Combustion Engineering.
bIndian Point 2 and 3 consequence information was obtained from Atomic Safety and Licensing Board testimony.
cWNP-2 = Washington Nuclear Project 2.

Table 5.2 Source term information used for final environmental statement severe accident analyses
Plant Source term used Comments
Beaver Valley 2 Rebaselined Reactor Safety Study (RSS) Pressurized-water reactor (PWR) source terms and frequencies from NUREG-0773 used
Braidwood 1, 2 Rebaselined RSS modified PWR source terms and frequencies from NUREG-0773 modified for specific Braidwood design features
Byron 1, 2 Rebaselined RSS Same as Beaver Valley
Callaway 1 Rebaselined RSS Same as Beaver Valley
Catawba 1, 2 Rebaselined RSS Same as Beaver Valley
Clinton 1 Rebaselined RSS Boiling-water reactor (BWR) source terms and frequencies from NUREG-0773 used
Comanche Peak 1, 2 Rebaselined RSS Same as Beaver Valley
Fermi 2 Rebaselined RSS Same as Clinton
Grand Gulf 1, 2 Rebaselined RSS Same as Clinton
Shearon Harris 1, 2 Rebaselined RSS Same as Beaver Valley
Hope Creek Rebaselined RSS Same as Clinton
Indian Point 2, 3 Plant specific  
Limerick 1, 2 Rebaselined RSS (modified) BWR source terms and frequencies from NUREG-0733 modified for specific Limerick design features. External events also included
Millstone 3 Plant-specific probabilistic risk analysis (PRA) Source terms and frequencies from plant specific PRA used
Nine Mile Point 2 Limerick PRA (modified) Source terms and frequencies from Limerick PRA modified for specific Nine Mile Point Unit 2 design features
Palo Verde 1, 2, 3 Rebaselined RSS Same as Beaver Valley
Perry 1, 2
See footnote at end of table.
Rebaselined RSS Same as Clinton
River Bend Grand Gulf RSS Methodologies Applications Program (MAP) Source terms and frequencies from Grand Gulf RSS MAP (NUREG/CR-1659) with no modification
San Onofre 2, 3 Rebaselined RSS Same as Beaver Valley
Seabrook 1, 2 Rebaselined RSS Same as Beaver Valley
South Texas 1, 2 Rebaselined RSS (modified) PWR source terms and frequencies from NUREG-0773 modified for specific South Texas design features
St. Lucie 2 Rebaselined RSS Same as Beaver Valley
Summer 1 Rebaselined RSS Same as Beaver Valley
Susquehanna 1, 2 Rebaselined RSS Same as Clinton
Vogtle 1, 2 Rebaselined RSS (modified) PWR source terms and frequencies from NUREG-0773 modified for specific Vogtle design features
Waterford 3 Rebaselined RSS Same as Beaver Valley
Wolf Creek 1, 2 Rebaselined RSS Same as Beaver Valley
WNP-2a Rebaselined RSS Same as Clinton

aWashington Nuclear Project 2.

The industry-sponsored risk assessments (e.g., Oconee 3, Seabrook, and Millstone 3) are similar in that efforts are made to reduce the degree of conservatism and to use the best information available. For these studies, source-term levels and frequencies specific to the plant are calculated.

Finally, studies exist that provide a detailed assessment of the risk due to specific types of accidents. For example, two such studies are NUREG-0440, which is a generic study of the radiological risks that could result from a severe accident that releases significant contamination into the groundwater, and NUREG-0769 (Addendum 1), which estimates the risks from direct contamination of the Great Lakes due to fallout from a severe accident at the Enrico Fermi 2 power plant. These two as well as other specific risk studies are used in this GEIS to provide a projection of risk during the license renewal period.

Severe accidents initiated by external phenomena such as tornadoes, floods, earthquakes, fires, and sabotage have not traditionally been discussed in quantitative terms in FESs. With the exception of sabotage, the NRC staff has, however, reviewed or performed detailed probabilistic assessments of external events for Zion Units 1 and 2, Indian Point Units 2 and 3, Limerick Units 1 and 2, Surry Unit 1, Peach Bottom Unit 2, and Millstone Unit 3. In most cases, the external event risks were determined to be comparable to internal event risks. However, for Zion and Limerick, the licensee's PRAs indicated that external events could be significant contributors to risk. For the Indian Point units, NRC staff evaluations also indicated that external events could significantly contribute to severe accident risk. The most recent NRC analysis of external events has been the NUREG-1150 external events assessment for Surry Unit 1 and Peach Bottom Unit 1. This analysis examined a broad range of external events and found that they could range from negligible to significant contributors to risk when compared with internal initiators. It should be noted, however, that in cases where external event risk was shown to be a significant contributor to the overall risk, the majority of the estimated risk arose from large beyond design basis earthquakes; but in all cases, the total risk (internal and external) is still small.

NRC's earthquake design standards have been conservatively developed to ensure protection of the public health and safety from earthquakes whose magnitudes are well above the most likely earthquake magnitude when considering the collective earthquakes history for specific plant sites in the United States. Therefore, earthquakes exceeding NRC seismic design standards are extremely unlikely. However, in the unlikely event of such an earthquake, there would be substantial damage to older residential structures, commercial structures, and high-hazard facilities such as dams whose seismic design standards are below nuclear seismic design standards. The societal impact due to the non-nuclear losses alone from an earthquake larger than the design basis of a nuclear plant, including property damage, injuries, and fatalities, would be major. The technology for assessing losses from such large earthquakes is a developing one, and there are several ongoing studies of this technology, including work at the United States Geological Survey. Presently there is no agreed-upon method for performing such assessments, although a recent report of the National Academy of Sciences suggests some broad guidelines (NAS 1989). The NRC has not developed a method for assessing the societal losses from large earthquakes such that the reactor contribution to accident consequences can be quantitatively compared with the non-nuclear losses. However, as supported by at least one study (Lee et al. 1979), the commission expects that the reactor accident contribution to the losses from large beyond design basis earthquakes would be small relative to the non-nuclear losses. While this in itself does not mean the reactor consequences from such an earthquake would be small, the commission concludes that even with potentially high consequences from a beyond design basis earthquake, the extremely low probability of such earthquake yields a small risk from beyond design basis earthquakes at existing nuclear power plants.

With regard to sabotage, quantitative estimates of risk from sabotage are not made in external event analyses because such estimates are beyond the current state of the art for performing risk assessments. The commission has long used deterministic criteria to establish a set of regulatory requirements for the physical protection of nuclear power plants from the threat of sabotage, 10 CFR Part 73, "Physical Protection of Plants and Materials", delineates these regulatory requirements. In addition, as a result of the World Trade Center bombing, the Commission amended 10 CFR Part 73 to provide protection against malevolent use of vehicles, including land vehicle bombs. This amendment requires licenses to establish vehicle control measures, including vehicle barrier systems to protect against vehicular sabotage. The regulatory requirements under 10 CFR part 73 provide reasonable assurance that the risk from sabotage is small. Although the threat of sabotage events cannot be accurately quantified, the commission believes that acts of sabotage are not reasonably expected. Nonetheless, if such events were to occur, the commission would expect that resultant core damage and radiological releases would be no worse than those expected from internally initiated events.

Based on the above, the commission concludes that the risk from sabotage and beyond design basis earthquakes at existing nuclear power plants is small and additionally, that the risks form other external events, are adequately addressed by a generic consideration of internally initiated severe accidents.

Although external events are not discussed in further detail in this chapter, it should be noted that the NRC is continuing to evaluate ways to reduce the risk from nuclear power plants from external events. For example, each licensee is performing an individual plant examination to look for plant vulnerabilities to internally and externally initiated events and considering potential improvements to reduce the frequency or consequences of such events. Additionally, as discussed in Section 5.4.1.2, as part of the review of individual license renewal applications, a site-specific consideration of alternatives to mitigate severe accidents will be performed in order to determine if improvements to further reduce severe accident risk or consequences are warranted.

5.3.3.2 Dose and Adverse Health Effects from Atmospheric Releases

5.3.3.2.1 Methodology for Predicting Future Risk Summary of methodology

The assessment of environmental impacts due to the atmospheric release pathway are described in this section. This pathway includes the exposure of individuals directly from the passage of the cloud of radioactive material released from an accident and from material deposited on the ground, as well as the longer-term effects from other terrestrial pathways such as the ingestion of crops. Doses and the resulting health effects (early and latent fatalities) will be estimated for the middle year of relicense (MYR) population. The MYR is the estimated midpoint of the renewal period for a given plant rounded upward to the next year of available population data. Predictions of MYR risk were generated by taking the results of existing risk calculations (i.e., plant-specific estimates of early fatalities, latent fatalities, and dose) and regressing those values against a composite site-specific variable called the exposure index (EI). EI is a function of population surrounding the plant weighted by the site-specific wind direction frequency and, thus, is a site-specific parameter. Because meteorological patterns, including wind direction frequency, tend to remain constant over time, EI changes as populations change or become redistributed.

A straight-line regression of the total risk values (taken from FES analyses) for each plant listed in Table 5.1 versus the EI for that plant (at the date assumed in the FES analyses) was calculated. Average and 95 percent upper confidence bound values of total risk were estimated. Risks for individual plants for their license renewal period were then estimated from the upper confidence bound values based on MYR population data converted to MYR EI. In the prediction of risk using EI (discussed in the preceding paragraph), the assumption was made that future plant risk is primarily a function of population and wind direction. Secondary factors--such as terrain, rainfall, and wind stability--also have some effect on risk, but their impact was judged to be much smaller than the effects of population and wind direction.

Selection of appropriate existing analyses for use in regression

Currently, 118 nuclear plants are in operation or under active construction in the United States. These 118 plants represent 72 sites for the evaluation of air pathway consequences bsp;sites are used for the other two pathway evaluations).4 As noted previously, only a portion of these nuclear plants have severe accident analyses available for review.

The data selected for use in this GEIS analysis were taken from the FESs published since 1981. As discussed previously, these FES analyses are based upon source terms resulting from the RSS (NUREG-75/014, formerly WASH-1400) rebaselined in NUREG-0773. As such, these source terms (and the resulting risk and environmental impacts calculated using them) reflect the plant designs used in WASH-1400. However, this approach is considered conservative because the source terms developed in WASH-1400 generally reflect an "as found" (late 1970s) and, as such, do not reflect the improvements that have been made in nuclear industry plant design and operations since the early 1980s. Accordingly, the use of WASH-1400 source terms in the FESs may, in many cases, tend to overestimate the actual environmental consequences and risks.

Since the RSS study was completed, the NRC has implemented several industry-wide risk-reduction programs. These programs, such as station blackout (10 CFR 50.63), anticipated transient without scram (10 CFR 50.62), resolution of other generic safety issues, improvements resulting from the extensive reviews of the accident at Three Mile Island (NUREG-0737), and the individual plant examination and containment performance improvement programs, have served to lower the overall average values of nuclear plant risk relative to their values prior to the changes. Because the programs are implemented on an industry-wide basis, risk values should be smaller at all plants. No quantification of the overall risk reduction has been performed, but it is believed that the risk reduction is significant. As a result of the changes, the staff believes that the spectrum of risk for the entire nuclear industry shifted downward to lower overall risk values, and the average total risk for all nuclear plants is smaller than the risk estimated in the original RSS analyses. Thus, RSS risk estimates are more representative of the upper end of the total nuclear plant risk spectrum than the actual current risks.

The preceding discussion shows that the use of the FES risk values provides reasonable estimates of the actual average risk of the general nuclear plant population and that the use of the FES values in this analysis results in appropriate risk values in the GEIS. Where there were choices of methodology and the best method was not obvious, the staff chose the method that would lead to higher predicted values. The use of the 95 percent upper prediction confidence bounds from the regression in this analysis (discussed later) provides even greater assurance that the GEIS does not underestimate potential future environmental impacts.

As for use of detailed PRA analyses in the GEIS, particularly the NUREG-1150 studies, the plants represented in these detailed PRAs have had the benefit of considerable risk reduction feedback and improvement; consequently, their predicted risk values are not considered to be representative of the absolute values of the general plant population risk. However, these studies do provide significant risk information on the relative risks to the public as a function of distance from the plant. Because of the much more advanced computational tools available during the NUREG-1150 studies (which could better model secondary effects such as rainfall pattern), as well as more than 10 years of additional knowledge about severe accidents, the information on the distribution of risk at a specific plant, as estimated by the NUREG-1150 reports, is considered more realistic and representative of the actual environmental impacts due to the air pathway for severe accidents. The GEIS uses this relative risk information in its analysis process.

Enveloping of all plants with FES analyses

Many factors could potentially increase the consequences to the general public resulting from a severe-accident release. A comprehensive listing and description of factors that influence consequences are provided in the PRA Procedures Guide (NUREG/CR-2300). The purpose of this section is to use, to the extent possible, the available severe accident results (Table 5.3), in conjunction with those factors that are important to risk and that change with time to estimate the consequences of nuclear plant accidents for all plants for a time period that exceeds the time frame of existing analyses. This estimation process was completed by predicting increases or decreases in consequences as the plant lifetime is extended past the normal license period by considering the projected changes in the risk factors. The primary assumption in this analysis is that regulatory controls will ensure that the physical plant condition (i.e., the predicted probability of and radioactive releases from an accident) will be maintained at a constant level during the renewal period; therefore, the frequency and magnitude of a release will remain relatively constant. In other words, significant changes in consequences will result only from changes in the plant's external environment. The most logical approach, then, would be to incorporate the most significant environmental factors into calculations of consequences for subsequent correlation with existing analyses (which use the consequence computer codes). The two parameters selected for this analysis are population and wind direction, as discussed in the following paragraphs.

Table 5.3 Comparison of general site characteristics. Italics indicate that the final environmental statement contains severe accident evaluations

Plant MYR evaluation datea MYR 50-mile populationb MYR 50-mile population in high-frequency wind direction Rainc Snowc General terraind
Arkansas 1 2030 245,476 20,471 51 5 3
Arkansas 2 2030 245,476 20,471 51 5 3
Beaver Valley 1 2030 4,039,282 1,177,194 36 46 3
Beaver Valley 2 2050 4,169,673 1,202,284 36 46 3
Bellefonte 1 2050 1,473,597 60,836 56 3 4
Bellefonte 2 2050 1,473,597 60,836 56 3 4
Big Rock Point 2030 228,199 61 31 111 2
Braidwood 1 2050 5,092,832 1,534,979 30 28 2
Braidwood 2 2050 5,092,832 1,534,979 30 28 2
Browns Ferry 1 2030 926,918 27,791 47 3 4
Browns Ferry 2 2030 926,918 27,791 47 3 4
Browns Ferry 3 2030 926,918 27,791 47 3 4
Brunswick 1 2030 304,703 7,703 51 2 1
Brunswick 2 2030 304,703 7,703 51 2 1
Byron 1 2050 1,141,541 29,618 18 34 2
Byron 2 2050 1,141,541 29,618 18 34 2
Callaway 1 2030 463,360 17,712 37 19 3
Calvert Cliffs 1 2030 3,481,008 256,881 41 21 1
Calvert Cliffs 2 2030 3,481,008 256,881 41 21 1
Catawba 1 2050 2,337,775 139,401 42 5 4
Catawba 2 2050 2,337,775 139,401 42 5 4
Clinton 1 2050 869,226 27,294 35 23 2
Comanche Peak 1 2030 1,654,378 54,431 31 3 1
Comanche Peak 2 2050 1,875,643 61,419 31 3 1
Cooper 2030 217,516 19,745 28 28 2
Crystal River 3 2030 655,382 0 42 0 1
D.C. Cook 1 2030 1,440,998 15 36 69 2
D.C. Cook 2 2030 1,440,998 15 36 69 2
Davis Besse 2030 2,169,925 20 32 37 2
Diablo Canyon 1 2050 419,046 4 32 0 6
Diablo Canyon 2 2050 419,046 4 32 0 6
Dresden 2 2030 7,453,539 143,593 33 30 2
Dresden 3 2030 7,453,539 143,593 33 30 2
Duane Arnold 1 2030 754,825 26,445 33 31 2
Farley 1 2030 488,464 21,412 54 0 1
Farley 2 2050 542,746 25,242 54 0 1
Fermi 2 2050 6,647,763 0 31 31 2
FitzPatrick 2030 804,876 12,128 34 88 2
Fort Calhoun 1 2030 887,478 14,526 30 32 2
Ginna 2030 1,112,686 0 33 86 2
Grand Gulf 1 2050 504,930 15,143 50 2 1
Haddam Neck
(Connecticut Yankee)
2030 4,136,066 120,354 43 53 5
Hatch 1 2030 416,412 43,798 44 1 1
Hatch 2 2030 416,412 43,798 44 1 1
Hope Creek 2050 5,424,373 54,596 40 23 1
Indian Point 2e 2030 15,195,541 602,427 43 6 3
Indian Point 3e 2030 15,195,541 602,427 43 26 3
Kewanee 1 2030 733,618 0 28 45 2
La Salle 1 2050 1,366,307 61,875 35 28 2
La Salle 2 2050 1,366,307 61,875 35 28 2
Limerick 1 2050 7,615,980 794,765 59 20 1
Limerick 2 2050 7,615,980 794,765 59 20 1
Maine Yankee 2030 830,737 19,668 43 71 5
McGuire 1 2050 2,543,485 134,597 43 6 4
McGuire 2 2050 2,543,485 134,597 43 6 4
Millstone 1 2030 3,138,820 1,419 39 26 5
Millstone 2 2030 3,137,820 1,419 39 26 5
Millstone 3 2050 3,325,582 1,462 39 26 5
Monticello 1 2030 2,815,967 1,587,694 24 42 2
Nine Mile Point 1 2030 802,759 12,239 34 88 2
Nine Mile Point 2 2050 811,475 12,478 34 88 2
North Anna 1 2030 1,478,490 41,700 44 16 4
North Anna 2 2030 1,478,490 41,700 44 16 4
Oconee 1 2030 1,311,318 53,947 53 6 4
Oconee 2 2030 1,311,318 53,947 53 6 4
Oconee 3 2030 1,311,318 53,947 53 6 4
Oyster Creek 1 2030 4,561,213 929 41 16 1
Palisades 2030 1,337,910 9,582 36 69 2
Palo Verde 1 2050 1,974,946 2,700 13 0 3
Palo Verde 2 2050 1,974,946 2,700 13 0 3
Palo Verde 3 2050 1,974,946 2,700 13 0 3
Peach Bottom 2 2030 5,283,198 122,770 38 35 4
Peach Bottom 3 2030 5,283,198 122,770 38 35 4
Perry 1 2050 2,767,417 0 34 52 2
Pilgrim 1 2030 4,881,755 0 42 42 1
Point Beach 1 2030 700,257 13,275 24 45 2
Point Beach 2 2030 700,257 13,275 24 45 2
Prairie Island 1 2030 2,961,583 29,124 24 44 2
Prairie Island 2 2030 2,961,583 29,124 24 44 2
Quad Cities 1 2030 810,640 13,191 36 29 2
Quad Cities 2 2030 810,640 13,191 36 29 2
Rancho Seco 1 2030 2,589,992 303,556 17 0 6
River Bend 2050 1,105,994 15,770 54 0 1
Robinson 2 2030 991,450 30,941 45 2 4
Salem 1 2030 5,180,877 49,873 40 23 1
Salem 2 2050 5,372,611 54,002 40 23 1
San Onofre 1 2030 7,048,438 0 12 0 1
San Onofre 2 2050 7,764,644 0 12 0 1
San Onofre 3 2050 7,764,644 0 12 0 1
Seabrook 1 2050 4,452,452 344 43 75 5
Sequoyah 1 2030 1,208,316 205,182 58 4 3
Sequoyah 2 2050 1,334,579 226,371 58 4 3
Shearon Harris 1 2050 2,122,597 75,055 45 7 4
Shoreham 2050 5,692,690 170,058 47 34 1
South Texas 1 2050 382,195 29,850 42 0 1
South Texas 2 2050 382,195 29,850 42 0 1
St. Lucie 1 2030 1,036,446 41,401 32 0 1
St. Lucie 2 2050 1,245,868 49,375 32 0 1
Summer 1 2050 1,385,612 83,181 45 2 4
Surry 1 2030 2,506,022 36,210 45 10 1
Surry 2 2030 2,506,022 36,210 45 10 1
Susquehanna 1 2050 1,575,680 34,206 35 50 4
Susquehanna 2 2050 1,575,680 34,206 35 50 4
Three Mile Island 1 2030 2,294,045 263,028 38 37 3
Trojan 1 2030 2,822,894 116,369 42 7 6
Turkey Point 3 2030 4,156,261 93,491 54 0 1
Turkey Point 4 2030 4,156,261 93,491 54 0 1
Vermont Yankee 2030 1,709,869 58,938 43 60 5
Vogtle 1 2050 932,240 17,480 42 1 1
Vogtle 2 2050 932,240 17,480 42 1 1
Waterford 3 2050 2,778,959 45,309 54 0 1
Watts Bar 1 2050 1,367,537 56,133 53 9 3
Watts Bar 2 2050 1,367,537 56,133 53 9 3
WNP-2 f 2050 405,235 23,692 5 18 3
Wolf Creek 1 2050 273,225 26,641 31 15 2
Yankee Rowe 2010 1,796,823 471,262 37 66 5
Zion 1 2030 8,199,956 0 32 58 2
Zion 2 2030 8,199,956 0 32 58 2

aMYR = Middle year of license renewal period rounded up to the next year for which population forecasts were available.
b50 miles = 80 km.
cAnnual average in inches.
dTerrain categories: 1 = coastal plain, 2 = central lowlands, 3 = plateaus, 4 = parallel valleys and ridges, 5 = rolling hills to high mountains, 6 = steep mountains.
e
Severe accident information obtained from Atomic Safety and Licensing Board testimony and not from the final environmental statement.
fWNP-2 = Washington Nuclear Project 2.

Many factors can affect the amount of radiation to which the public is exposed. The magnitude of impact varies for any individual factor and generally is very specific to a particular plant or site. If the FES risk results are to be used to predict future risk values for all plants, it should be demonstrated that the FES plants provide a reasonable envelope of the more significant risk factors for all plants. Such factors include population density, meteorology, evacuation, and interdiction. Studies have shown that some factors have a greater degree of influence than others; for example, population has a very strong influence over risk (NUREG/CR-2239, NUREG-1150). Evacuation can have a significant influence on early fatality risk but a much more limited impact on latent fatality risk. Interdiction primarily reduces latent fatality risk. While particular aspects of meteorology, such as rainfall, can have a significant impact on peak risk values, mean health effect values are relatively insensitive to meteorology. When the basic reasons for the risk influence of each factor are examined, these factors can generally be reduced to three issues: (1) the number of people exposed to the severe accident release, (2) the likelihood that any given individual receives an exposure, and (3) the amount of radiation the individual receives. Consequently, site population (which reflects the number of people potentially at risk to severe accident exposure) and wind direction frequency (which reflects the likelihood of exposure) have been chosen as the primary factors affecting risks.

Although there are other secondary factors (e.g., source term and dose response relationship) that can influence risk and were not specifically analyzed on a plant-specific basis in this GEIS, these factors were not ignored as their impact is included in the FES analyses whose results are the basis for the GEIS analyses. Consequently, their effects are indirectly considered in the prediction of future risks and are reflected within the uncertainty bounds generated by the regression of the FES risk values. To ensure that the existing FES analyses cover a range of secondary factors representative of the total population of plants, the more significant secondary factors were examined as discussed below. The secondary factors examined are as follows:

  • average annual precipitation,
  • residential population within a 50-mile (80-km( radius of the plant,
  • population [50 miles (90 km)] in highest frequency wind direction,
  • general terrain, and
  • emergency planning.

Average annual precipitation. After an atmospheric release caused by a severe accident, the fallout rate of the released radionuclides is generally the result of gravitational settling and, consequently, is not a rapid process. This slow fallout allows a given release to be suspended for sufficient time to allow for some radioactive decay of the shorter-lived radionuclides, resulting in lower individual doses to the public. In addition, releases are distributed over a wide area, resulting in relatively low individual doses (although the overall total population dose is not greatly affected). However, precipitation counteracts both of these effects by washing the radionuclides out of the atmosphere and not allowing time for extensive dispersion or decay. Thus, plant sites with higher levels of annual precipitation may indicate higher levels of risk for those measures that are based on individual doses.

Residential population within a 50-mile radius of the plant. This factor is a rather understandable selection in that plant sites with larger resident populations will have a larger number of persons at risk for a given severe accident release. Population projections were made based on the 1980 census data and projected growth (decline) factors derived from the U.S. Bureau of Census evaluations. A radius of 50 miles was selected for comparison purposes because existing analyses indicate a large majority (although not all) of early health effects from a severe accident release occur within 50 miles of the plant site.

Population (50 miles) in highest-frequency wind direction. This factor highlights a "higher risk" sector of the overall population around a specific plant site. The sector is 22.5° and the population is 0 to 50 miles from the plant in that sector. Higher populations combined with higher frequency of wind in that direction may indicate higher risks in that sector.

General terrain. This factor is chosen because the dispersion behavior of the plume may be influenced by the general terrain surrounding the plant (e.g., plains versus mountains). Six terrain classifications were selected as described in footnote c to Table 5.3.

Table 5.3 shows the values for these four factors for all nuclear plant sites. As can be seen, the existing severe accident analysis as provided in those FESs containing a severe accident evaluation provides a reasonable envelope for precipitation (rainfall and snowfall), 50-mile population, and 50-mile population in the direction of highest wind frequency. All six terrain classifications are also covered by referenced FES analyses. From review of these data, it is concluded that the FES plants sufficiently envelop these factors. Likewise, any plant risk projections that are developed from the FES severe accident results will reasonably account for secondary effects from these factors if the upper confidence bounding values from the projections are used to estimate the risk from atmospheric releases for plants during their license renewal period.

Emergency planning. Even in the event of a release of radioactive material from a plant, protective actions can be taken to move or shelter members of the public in the projected path of the radioactive cloud. The success of these actions in preventing exposure of members of the public to the radioactive material is dependent upon the warning time available prior to the release and the time it takes to carry out the protective actions. In general, this latter item (the time to carry out the protective action) is mostly influenced by the size of the population around the plant. Each FES that addresses severe accidents considers the effects of site-specific emergency planning in calculating exposures and risks to the public. Since the FES plants include sites with populations that reasonably cover the range of populations at all 74 sites, a range of emergency planning is considered in the data used for the predictions of early and latent fatalities during the license renewal period. Thus, this GEIS analysis should reasonably account for the effects of emergency planning.

Projections of estimates of risk

Detailed severe accident consequence (early and latent fatalities and total dose) evaluations are not available for all plants. Therefore, a predictor for these consequences was developed using correlations based upon the calculated results from the existing FES severe accident analyses. This predictor was then used to infer the future consequence level of all individual nuclear plants. Correlations were developed using two environmental parameters that are available for all plants. This correlation process is described below.

Discussion of exposure index

Population, which changes over time, defines the number of people within a given distance from the plant. Wind direction, which is assumed not to change from year to year, helps determine what proportion of the population is at risk in a given direction, because radionuclides are carried by the wind. Therefore, an EI relationship was developed by multiplying the wind direction frequency (fraction of the time per year) for each of 16 (22.5° ) compass sectors times the population in that sector for a given distance from the plant and summing all products. An example calculation for an EI value for 1990 at 10 miles (16 km) is shown in Table 5.4. The EI value, as calculated in Table 5.4, can be considered to be the expected population at risk for the year 1990 out to a distance of 10 miles from the nuclear power plant. Population varies with population growth and movement, and with the distance from any given plant. As the population changes for that plant, the EI also changes (the larger the EI, the larger the number of people at risk). Thus, EI is proportional to risk and an EI for a site for a future year can be used to predict the risk to the population around that site in that future year.

Regression of FES values

Several relationships of EI versus risk were developed by regressing total early fatality, normalized total latent fatality, and normalized total dose values on various EI values for the FES plants (see Appendix G). The EI values at 10 miles were found to best correlate with early fatalities, which is to be expected because, in the FES analyses, early fatalities tend to be clustered close to the plant. The EI values at 150 miles (241 km) were found to best correlate with latent fatalities and total dose. This finding is to be expected because the magnitudes of these risk values are largely influenced by the exposure of large populations around the plant.

Table 5.4 Example calculation for exposure index (EI) value with 1990 populationat 10-mile radius from plant
Direction segment A (wind frequency in given direction) B (1990 population within 10 miles of Plant X)a C (product)
N 0.06 100 6.0
NNE 0.06 105 6.3
NE 0.02 55 1.1
ENE 0.10 20 2.0
E 0.08 25 2.0
ESE 0.08 24 1.92
SE 0.09 75 6.75
SSE 0.10 125 12.5
S 0.06 400 24.0
SSW 0.05 275 13.75
SW 0.07 100 7.0
WSW 0.06 78 4.68
W 0.06 72 4.32
WNW 0.06 40 2.40
NW 0.02 80 1.6
NNW 0.03 78 2.34
  1.00 1652 EI = 98.66

a10 miles = 16 km.

Note: To calculate EI value: A x B = C; EI = sum of C.

Because the magnitude of the source term is generally proportional to plant power for a given accident sequence, the FES estimates for total latent fatalities used in the latent fatality regression were first normalized to 1000 MW(t) to minimize the regression variance due to the differing plant sizes. A linear dose response function is used in the FES analyses, and because of the assumptions of downwind and crosswind spread, radioactive material is predicted to be widely dispersed. Thus, the larger the amount of radioactive material released, the larger the predicted latent fatality level (slightly reduced from strict linearity by the interdiction assumptions). Similar logic is applicable to normalization of total dose. Normalization was not used for early fatalities because of the highly nonlinear dose response function used in the FES analyses and the use of a threshold effect (that is, there is a dose level below which no early fatality is predicted). Nonetheless, early fatalities are also highly influenced by the amount of radioactive material released (i.e., plant size), and to help ensure that early fatality data from smaller plants do not distort the regression results for the larger plants, only the early fatality data for plants greater than 3025 MW(t) were used in the regression of early fatalities (Table 5.5, footnote f). The inability to correct fully for the effects of plant size and the dose-early fatality relationship leads to a higher dispersion in the regression estimates, which influences the upper confidence bound (UCB) as will be seen in subsequent sections.

Also, in several of the FES documents, two sets of early fatality values were provided, one set which assumed minimal medical support was available to aid the exposed population and a second set which assumed normal and expected levels of medical support were available. The regression used those early fatality values associated with expected medical support levels. The assumption there would be only minimal or no medical support after an accident was considered to be unrealistic. A detailed discussion of the regression analyses and UCB is provided in Appendix G.

5.3.3.2.2 Results

The data in Table 5.5 summarize the information for 28 nuclear plant sites that were used to develop the relationship between EI and consequences of severe accidents analysis for both PWRs and BWRs. Because of fundamental design differences between PWRs and BWRs, separate regression analyses were performed for each to better account for the BWR-PWR differences in plant failure modes and source terms. Accordingly, the PWR regression was used to determine the best fit relationship for PWR risk values and the BWR regression was used to determine the best fit relationship for FES BWR risk values. As can be seen in Figures 5.2-5.7, two lines (representing average and UCB values) result from the regression analyses for total early fatalities, total latent fatalities, and total dose. The 95 percent UCB (dashed line) was developed based on the scatter in the data. Two points need to be made about the UCB. First, two UCB values were calculated: one value assuming that the data points (i.e., early and latent fatalities and population dose) had a normal distribution about some mean and the second value assuming that the data points did not have a normal distribution about the mean. The larger of the two UCB values was then used in making plant risk projections. The second point to be noted is that because of the small number of data used in the regressions (18 PWR data points and 10 BWR data points), the scatter in the data (expressed as residuals) for all 28 data points was used in determining the UCB for both the PWR and BWR regressions.

Table 5.5 Information used for regression analyses for expected early, latent, and total dose at 28 nuclear plant sites for the license renewal period

Normalized values are obtained by converting nonnormalized values to the equivalent of a 1000-MW(t) plant

Plant FES analysisa date of population EI valuesb
(10 miles)
EI valuesb
(150 miles)
Expected earlyc,d fatalities
(persons/reactor year)
Expected latent fatalities
(persons/reactor year)
Expected total dose
(person-rem/reactor year)
          Nonnormalizedc Normalizede Nonnormalizedc Normalizede
Beaver Valley 2 2010 9,195 958,330 0.002 f 0.022 0.0083 230 86.73
Braidwood 1, 2 2000 1,916 1,435,347 0.00038 0.0138 0.004 180 52.77
Byron 1, 2 2000 1,391 1,084,499 0.00026 0.016 0.0047 218 63.91
Callaway 1 2000 508 343,991 0.0001 0.0077 0.0022 126 35.34
Catawba 1, 2 2000 5,414 678,486 0.0011 0.0124 0.0036 170 49.84
Clinton 1 2000 658 1,272,955 0.000009 f 0.0191 0.0066 320 110.57
Comanche Peak 1, 2 2000 1,251 292,169 0.0001 0.0046 0.0014 58 17.00
Fermi 2 2000 4,165 1,112,272 0.00074 0.04 0.012 520 157.96
Grand Gulf 1, 2 2000 437 297,829 0.00006 0.0055 0.0014 100 26.09
Shearon Harris 1, 2 2010 1,415 550,951 0.00018 f 0.0088 0.0032 114 41.08
Hope Creek 2010 1,541 1,822,818 0.0003 0.07 0.021 1000 303.67
Indian Point 2, 3g 1990 18,325 2,743,032 0.0115 h 0.826i 0.299 10,400 i 3770.85
Limerick 1, 2 2000 10,307 2,455,497 0.00914 0.0957 0.029 1360 413.00
Millstone 3 2010 8,751 1,397,683 0.0008 0.05 0.015 1000 293.17
Nine Mile Point 2 2000 1,500 269,042 0.0004 0.023 0.007 300 90.28
Palo Verde 1, 2, 3 2000 67 194,928 0.0000021 0.00456 0.0012 67 17.63
Perry 1, 2 2000 4,465 920,212 0.000016 0.0285 0.008 470 131.32
River Bend 2000 1,485 334,565 0.0004 f 0.047 0.016 700 241.88
San Onofre 2, 3 2000 3,950 978,306 0.001 0.033 0.0097 380 112.09
Seabrook 1, 2 2000 4,090 448,066 0.0006 0.0075 0.0022 105 30.78
South Texas 1, 2 2010 236 461,241 0.0000007 0.0108 0.0028 250 65.79
St. Lucie 2 2000 8,739 540,442 0.00007 f 0.0064 0.0024 78 28.89
Summer 1 2000 647 627,969 0.00017 f 0.0094 0.0034 130 46.85
Susquehanna 1, 2 2000 3,760 1,995,580 0.00077 0.0227 0.0069 360 109.32
Vogtle 1, 2 2010 117 469,641 0.00001 0.024 0.007 310 90.88
Waterford 3 2000 4,745 285,560 0.00057 0.0059 0.0017 69 20.35
Wolf Creek 1 2000 311 289,260 0 j 0.00559 0.0016 99 29.02
WNP-2k 2000 108 100,055 0.00032 0.00487 0.0015 77 23.17

aThe population estimates for the indicated year were used to evaluate the consequences to the public for the final environmental statement (FES).
bExposure index (EI) values are given for FES analysis date of population (see note a).
cValues obtained from FES for the respective plant with the exception of Indian Point (See note g).
dDue to threshold dose effects, these estimates cannot be normalized (i.e., effects are not linear until an exposure threshold is reached).
eNormalized to 1000 MW(t) (see Appendix G).
fPlant thermal power < 3025 MW(t) and was not used in the regression for expected early fatalities.
gExpected risk values obtained from Atomic Safety and Licensing Board testimony and not from FES.
hRisk values for Indian Point 3 are listed.
iRisk values for Indian Point 2 are listed.
jBecause values of zero have no meaning on log scales, this data point was not used in the regression for early fatalities.
kWNP-2 = Washington Nuclear Project 2.

Note: Multiply person-rem by 0.01 to find person-sieverts; multiply miles by 1.609 to find kilometers.

Figure 5.2 Log plot of early fatalities (average deaths per reactor-year) for final environmental statement boiling-water reactor plants, fitted regression line (solid curve), and 95 percent normal-theory upper prediction confidence bounds (dotted curve).

Figure 5.3 Log plot of early fatalities (average deaths per reactor-year) for final environmental statement pressurized-water reactor plants, fitted regression line (solid curve), and 95 percent normal-theory upper prediction confidence bounds (dotted curve).

Figure 5.4 Log plot of normalized latent fatalities (average deaths per 1000 MW reactor-year) for final environmental statement boiling-water reactor plants, fitted regression line (solid curve), and 95 percent distribution-free upper prediction confidence bounds (dotted curve).

Figure 5.5 Log plot of normalized latent fatalities (average deaths per 1000 MW reactor-year) for final environmental statement pressurized-water reactor plants, fitted regression line (solid curve), and 95 percent distribution-free upper prediction confidence bounds (dotted curve).

Figure 5.6 Log plot of normalized total dose (person-rem per 1000 MW reactor-year) for final environmental statement boiling-water reactor plants, fitted regression line (solid curve), and 95 percent distribution-free upper prediction confidence bounds (dotted curve).

Figure 5.7 Log plot of normalized total dose (person-rem per 1000 MW reactor-year) for final environmental statement pressurized-water reactor plants, fitted regression line (solid curve), and 95 percent distribution-free upper prediction confidence bounds (dotted curve).

Using the UCB results of the regression analysis, the values for total early fatalities, total latent fatalities, and total dose were then predicted for each site at their MYRs, rounded up to the nearest year for which projected population data are available (2010, 2030, 2050). The results of the UCB projections for early fatalities, latent fatalities, and total dose are shown in Table 5.6. The EI values corresponding to the MYR for each site, which were used to make these predictions, are shown in ;Tables 5.7 and 5.8. Data for the Millstone plant provide a good example of the process by which these projections were made. The EI at 10 miles for Millstone (an FES plant) is 9420 at its MYR (2050) (Table 5.7). An EI of 9420 results in a projected early fatality UCB of 0.025 fatalities/RY. This value is higher than that reported in the Millstone FES for the year 2010 (0.0008 fatalities/RY, as shown in Table 5.5) and represents a conservative projection of the increase in early fatalities that could occur as a result of increased population around the Millstone site. The effects on risk due to factors such as emergency planning, meteorology (other than the frequency of wind direction--e.g., rainfall), and topography were accounted for in the FES analyses of severe accidents and are consequently incorporated into the FES risk values. Any variation in risk resulting from variation of these secondary parameters among FES plants will be reflected in the UCB calculated by the regressions. As discussed in Section 5.3.3.2.1, the FES plants reasonably envelop these secondary effects. If the future risks for all plants are then estimated using the appropriate (BWR or PWR) regression and the MYR EI, the resulting UCB values are estimated future risks that are not expected to be exceeded.

It should be noted that the risk values for latent fatalities provided in the FESs were calculated using the CRAC computer code which used a linear-quadratic cancer model based on older, low-level radiation exposure data (BEIR-III). Recent evaluations of the EI methodology (see Section 5.3.3.2.3) have been conducted using MACCS, the current, state-of-the-art computer code for assessing risks associated with postulated severe reactor accidents. Unlike CRAC, MACCS uses a linear cancer model based on the newer BEIR-V report. These evaluations suggest that latent fatality values in the FESs are an order of magnitude too low. Therefore, to account for this, the latent fatality results predicted from the FES values have been multiplied by a factor of 10 to obtain the final predicted latent fatality results in this GEIS.

Total population dose for an accident during each plant's relicensing period was also estimated by regression analysis. This dose includes the contribution from direct exposure to the radioactive cloud at the time of release as well as the longer-term effects from ground contamination. Table 5.9 shows the results of this (in person-rem/RY) along with an estimate of the respective average individual dose, in rem/RY, for each plant. Average individual doses were estimated by distributing the UCB total dose estimates from the regression analysis over the population within 150 miles (240 km) of the plant. Because it is virtually certain that people beyond this 150-mile radius would receive some incrementally small dose from an accident, attributing the total dose to the population within 150 miles will provide a conservative average individual dose estimate. For perspective, the annual average background dose to an individual from all other causes, including radon, is estimated as 3 x 10-1 rem per year.

Table 5.6 Predicted early and latent fatalities and dose estimates per reactor-year (RY) for all sites at their middle year of license renewal period, prior to incorporation of benchmark data.

Plant Predicted UCB total early fatalities/RY
(95% UCB)a
Nonnormalized predicted latent total fatalities/RY
(95% UCB)
Nonnormalized predicted total dose
(person-rem/RY)
(95% UCB)
Arkansas 3.3 x 10-3 1.7 x 10-2 238
Beaver Valley 2.5 x 10-2 1.3 x 10-1 1720
Bellefonte 4.0 x 10-3 1.0 x 10-1 1335
Big Rock Point 2.7 x 10-3 3.2 x 10-3 48
Braidwood 3.6 x 10-3 3.3 x 10-1 4418
Browns Ferry 4.3 x 10-3 9.7 x 10-2 1446
Brunswick 3.5 x 10-3 4.7 x 10-2 704
Byron 2.3 x 10-3 2.2 x 10-1 2867
Callaway 6.9 x 10-4 3.6 x 10-2 509
Calvert Cliffs 1.8 x 10-3 2.3 x 10-1 2995
Catawba 1.7 x 10-2 1.4 x 10-1 1880
Clinton 3.0 x 10-3 1.8 x 10-1 2549
Comanche Peak 2.3 x 10-3 3.3 x 10-2 466
Cooper 2.6 x 10-3 6.3 x 10-2 955
Crystal River 1.5 x 10-3 5.0 x 10-2 700
D. C. Cook 8.4 x 10-3 1.8 x 10-1 2311
Davis Besse 1.4 x 10-3 1.5 x 10-1 2021
Diablo Canyon 1.5 x 10-3 2.5 x 10-2 346
Dresden 4.6 x 10-3 1.4 x 10-1 1991
Duane Arnold 8.0 x 10-3 3.7 x 10-2 561
Farley 1.5 x 10-3 2.4 x 10-2 334
Fermi 2 6.8 x 10-3 1.9 x 10-1 2722
FitzPatrick 3.8 x 10-3 5.0 x 10-2 728
Fort Calhoun 1.7 x 10-3 8.0 x 10-3 111
Ginna 3.9 x 10-3 1.5 x 10-2 203
Grand Gulf 2.8 x 10-3 9.7 x 10-2 1441
Haddam Neck
(Connecticut Yankee)
1.2 x 10-2 2.0 x 10-1 2618
Hatch 2.6 x 10-3 5.7 x 10-2 855
Hope Creek 4.1 x 10-3 2.5 x 10-1 3604
Indian Point 6.5 x 10-2 7.7 x 10-1 9727
Kewanee 8.9 x 10-4 2.2 x 10-2 303
La Salle 3.6 x 10-3 2.0 x 10-1 2898
Limerick 1.1 x 10-2 3.1 x 10-1 4461
Maine Yankee 1.8 x 10-3 3.0 x 10-2 414
McGuire 1.0 x 10-2 1.4 x 10-1 1806
Millstone 2.5 x 10-2 3.1 x 10-1 3988
Monticello 4.1 x 10-3 5.0 x 10-2 730
Nine Mile Point 3.8 x 10-3 6.7 x 10-2 996
North Anna 9.4 x 10-4 1.1 x 10-1 1496
Oconee 1.1 x 10-2 1.0 x 10-1 1311
Oyster Creek 7.4 x 10-3 1.5 x 10-1 2125
Palisades 4.2 x 10-3 1.3 x 10-1 1691
Palo Verde 1.1 x 10-4 2.6 x 10-2 369
Peach Bottom 4.2 x 10-5 2.0 x 10-1 2950
Perry 6.9 x 10-3 1.7 x 10-1 2544
Pilgrim 3.7 x 10-3 6.0 x 10-2 873
Point Beach 2.5 x 10-3 2.3 x 10-2 309
Prairie Island 3.7 x 10-3 1.7 x 10-2 237
Quad Cities 4.5 x 10-3 1.1 x 10-1 1588
Rancho Seco 1.1 x 10-3 1.3 x 10-1 1723
River Bend 4.1 x 10-3 8.0 x 10-2 1168
Robinson 3.1 x 10-3 7.0 x 10-2 926
Salem 2.9 x 10-3 5.0 x 10-1 6059
San Onofre 1.1 x 10-2 2.4 x 10-1 3099
Seabrook 1.1 x 10-2 6.0 x 10-2 819
Sequoyah 6.6 x 10-3 1.1 x 10-1 1474
Shearon Harris 2.8 x 10-3 7.3 x 10-2 1001
South Texas 3.3 x 10-4 8.0 x 10-2 1065
Saint Lucie 3.2 x 10-2 8.0 x 10-2 1063
Shoreham 7.7 x 10-3 6.3 x 10-2 2724
Summer 1.3 x 10-3 1.0 x 10-1 1381
Surry 1.6 x 10-2 9.0 x 10-2 1200
Susquehanna 6.0 x 10-3 2.8 x 10-1 4010
Three Mile Island 2.8 x 10-2 3.3 x 10-1 4381
Trojan 3.7 x 10-2 1.5 x 10-1 1971
Turkey Point 6.0 x 10-2 2.0 x 10-2 278
Vermont Yankee 4.6 x 10-3 9.0 x 10-2 1314
Vogtle 1.6 x 10-4 7.3 x 10-2 983
WNP-2b 2.3 x 10-3 4.3 x 10-2 649
Waterford 1.4 x 10-2 3.3 x 10-2 477
Watts Bar 1.8 x 10-3 1.2 x 10-1 1540
Wolf Creek 4.7 x 10-4 3.3 x 10-2 466
Yankee Rowe 3.3 x 10-3 6.7 x 10-2 872
Zion 5.6 x 10-2 1.8 x 10-1 2379

aUCB = upper confidence bound. For description and explanation of these values, see Appendix G.
bWNP-2 = Washington Nuclear Project 2.Note: Multiply person-rem by 0.01 to find person-sieverts.

Table 5.7 Middle year of license renewal period (MYR) evaluation date and 10-mile exposure index (EI) for each licensed nuclear plant in the U.S.

Values are given in descending order

Plant MYR evaluation datea EIb
(10 miles)
Indian Point 2030 18,959
Turkey Point 2030 17,852
Zion 2030 16,913
Trojan 2030 12,556
St. Lucie 2030 11,447
Limerick 2050 10,709
Three Mile Island 2030 10,327
Beaver Valley 2050 9,535
Millstone 2050 9,420
Catawba 2050 7,219
Surry 2030 6,796
Duane Arnold 2030 6,283
Waterford 2050 6,163
Shoreham 2050 5,915
Oyster Creek 2030 5,584
Haddam Neck
(Connecticut Yankee)
2030 5,476
Seabrook 2050 5,234
Oconee 2030 5,184
San Onofre 2050 5,179
Perry 2050 5,020
Fermi 2 2050 4,919
McGuire 2050 4,919
D. C. Cook 2030 4,163
Susquehanna 2050 3,976
Sequoyah 2050 3,471
Palisades 2030 2,421
Vermont Yankee 2030 2,408
Dresden 2030 2,345
Bellefonte 2050 2,317
Ginna 2030 2,291
Quad Cities 2030 2,228
Prairie Island 2030 2,188
Braidwood 2050 2,126
Browns Ferry 2030 2,019
Yankee Rowe 2010 1,998
Arkansas 2030 1,993
Peach Bottom 2030 1,972
River Bend 2050 1,857
Salem 2050 1,808
Robinson 2030 1,889
Monticello 2030 1,832
Hope Creek 2050 1,807
Shearon Harris 2050 1,773
Point Beach 2030 1,612
Nine Mile Point 2050 1,568
FitzPatrick 2030 1,532
Comanche Peak 2030 1,518
Byron 2050 1,468
Pilgrim 2030 1,435
La Salle 2050 1,307
Maine Yankee 2030 1,246
Watts Bar 2050 1,241
Calvert Cliffs 2030 1,232
Brunswick 2030 1,195
Fort Calhoun 2030 1,155
Crystal River 2030 1,064
Farley 2050 1,021
Diablo Canyon 2050 1,020
Davis-Besse 2030 979
Summer 2050 902
Rancho Seco 2030 835
Clinton 2050 760
North Anna 2030 704
Kewanee 2030 671
Grand Gulf 2050 562
Callaway 2030 541
Big Rock Point 2030 476
Cooper 2030 411
Wolf Creek 2050 381
Hatch 2030 372
South Texas 2050 278
Vogtle 2050 141
WNP-2c 2050 134
Palo Verde 2050 96

aThe renewal period evaluation year is the estimated midpoint of the renewal period for that plant conservatively rounded upward to the next year of available population data (MYR). Dates of license expiration were obtained from Table2.1. The maximum renewal period of 20 years was assumed.
bValue obtained by multiplying wind frequency in each of 16 compass sectors by population 0 to 10 miles (16 km) from plant in that compass sector, then summing all products.
cWNP-2 = Washington Nuclear Project 2.

Table 5.8 Middle year of license renewal period (MYR) evaluation date and 150-mile exposure index (EI) for each licensed nuclear plant in the U.S.

Values are given in descending order.

Plant MYR evaluation datea EIb
(150 miles)
Indian Point 2030 2,863,844
Limerick 2050 2,647,224
Susquehanna 2050 2,279,528
Shoreham 2050 2,014,947
Salem 2050 1,979,840
Oyster Creek 2030 1,970,098
Hope Creek 2050 1,955,878
Three Mile Island 2030 1,928,285
Yankee Rowe 2010 1,739,663
Haddam Neck
(Connecticut Yankee)
2030 1,722,399
Braidwood 2050 1,615,088
Millstone 2050 1,510,698
Calvert Cliffs 2030 1,459,323
Peach Bottom 2030 1,453,860
Clinton 2050 1,418,383
La Salle 2050 1,396,350
Fermi 2 2050 1,287,935
Vermont Yankee 2030 1,286,085
San Onofre 2050 1,284,282
Byron 2050 1,214,624
Dresden 2030 1,193,394
Zion 2030 1,107,448
Davis-Besse 2030 1,104,797
D. C. Cook 2030 1,051,654
Palisades 2030 1,041,961
Beaver Valley 2050 1,021,547
Perry 2050 1,021,049
Rancho Seco 2030 992,605
Trojan 2030 944,628
Catawba 2050 914,688
McGuire 2050 890,305
North Anna 2030 876,587
Oconee 2030 867,675
Quad Cities 2030 854,803
Summer 2050 852,405
Surry 2030 846,246
Watts Bar 2050 798,733
Sequoyah 2050 769,140
Robinson 2030 738,770
Saint Lucie 2030 727,763
Shearon Harris 2050 688,554
Bellefonte 2050 678,549
Vogtle 2050 590,283
South Texas 2050 579,617
Crystal River 2030 573,211
Seabrook 2050 523,715
Browns Ferry 2030 491,751
Monticello 2030 487,606
Pilgrim 2030 486,154
Point Beach 2030 469,985
Kewanee 2030 440,217
River Bend 2050 432,680
Cooper 2030 428,471
Maine Yankee 2030 391,929
Grand Gulf 2050 388,245
Prairie Island 2030 375,227
Callaway 2030 373,564
Waterford 2050 370,569
Comanche Peak 2030 363,530
Wolf Creek 2050 363,380
Ginna 2030 357,773
Hatch 2030 347,873
Turkey Point 2030 345,115
Farley 2050 344,405
Duane Arnold 2030 329,426
Diablo Canyon 2050 302,887
Palo Verde 2050 290,395
Nine Mile Point 2050 273,322
FitzPatrick 2030 270,532
Arkansas 2030 265,479
Brunswick 2030 256,923
Fort Calhoun 2030 242,370
Big Rock Point 2030 136,942
WNP-2c 2050 132,195

a The renewal period evaluation year is the estimated midpoint of the renewal period for that plant conservatively rounded up to the next year of available population data (MYR). Dates of license expiration were obtained from Table2.1. The maximum renewal period of 20 years was assumed.
bValue obtained by multiplying wind frequency in each of 16 compass sectors by population 0 to 150 miles (240 km) from plant in that compass sector, then summing all products.
cWNP-2 = Washington Nuclear Project 2.

Table 5.9 Predicted dose estimate (total and average individual) per reactor-year (RY) for all sites at their middle year of license renewal (MYR)
Plant Predicted UCB total dose 95% UCBa
(person-rem/RY)
150-mile MYR population
(in millions)
Predicted UCB average individual dose 95% UCBb
(person-rem/RY)
Arkansas 238 4.1 6 x 10-5
Beaver Valley 1720 15.6 1 x 10-4
Bellefonte 1335 12.3 1 x 10-4
Big Rock Point 48 2.3 2 x 10-5
Braidwood 4418 20.5 2 x 10-4
Browns Ferry 1446 8.6 2 x 10-4
Brunswick 704 5.2 1 x 10-4
Byron 2867 17.8 2 x 10-4
Callaway 509 6.6 8 x 10-5
Calvert Cliffs 2995 20.8 1 x 10-4
Catawba 1880 13.8 1 x 10-4
Clinton 2549 18.6 1 x 10-4
Comanche Peak 466 8.8 5 x 10-5
Cooper 955 5.4 2 x 10-4
Crystal River 700 10.6 7 x 10-5
D.C. Cook 2311 20.1 1 x 10-4
Davis-Besse 2021 20.6 1 x 10-4
Diablo Canyon 346 5.7 6 x 10-5
Dresden 1991 18.9 1 x 10-4
Duane Arnold 561 6.0 9 x 10-5
Farley 334 5.7 6 x 10-5
Fermi 2 2722 21.3 1 x 10-4
FitzPatrick 728 6.1 1 x 10-4
Fort Calhoun 111 3.6 3 x 10-5
Ginna 203 5.8 4 x 10-5
Grand Gulf 1441 6.2 2 x 10-4
Haddam Neck
(Connecticut Yankee)
2618 32.4 8 x 10-5
Hatch 855 5.8 1 x 10-4
Hope Creek 3604 35.5 1 x 10-4
Indian Point 9727 35.7 3 x 10-4
Kewanee 303 7.4 4 x 10-5
La Salle 2898 19.1 2 x 10-4
Limerick 4461 39.3 1 x 10-4
Maine Yankee 414 7.6 5 x 10-5
McGuire 1806 14.3 1 x 10-4
Millstone 3988 32.6 1 x 10-4
Monticello 730 5.9 1 x 10-4
Nine Mile Point 996 6.2 2 x 10-4
North Anna 1496 14.7 1 x 10-4
Oconee 1311 14.1 1 x 10-4
Oyster Creek 2125 34.0 6 x 10-5
Palisades 1691 20.4 8 x 10-5
Palo Verde 369 4.9 8 x 10-5
Peach Bottom 2950 33.1 9 x 10-5
Perry 2544 19.7 1 x 10-4
Pilgrim 873 13.9 6 x 10-5
Point Beach 309 7.4 4 x 10-5
Prairie Island 237 6.5 4 x 10-5
Quad Cities 1588 15.0 1 x 10-4
Rancho Seco 1723 16.5 1 x 10-4
River Bend 1168 7.9 1 x 10-4
Robinson 926 11.9 8 x 10-5
Salem 6059 36.1 2 x 10-4
San Onofre 3099 23.6 1 x 10-4
Seabrook 819 14.7 6 x 10-5
Sequoyah 1474 12.9 1 x 10-4
Shearon Harris 1001 11.8 8 x 10-5
Shoreham 2724 36.0 8 x 10-5
South Texas 1065 10.2 1 x 10-4
Saint Lucie 1063 12.0 9 x 10-5
Summer 1381 12.6 1 x 10-4
Surry 1200 12.9 9 x 10-5
Susquehanna 4010 36.0 1 x 10-4
Three Mile Island 4381 29.3 1 x 10-4
Trojan 1971 9.4 2 x 10-4
Turkey Point 278 6.9 4 x 10-5
Vermont Yankee 1314 20.9 6 x 10-5
Vogtle 983 9.4 1 x 10-4
WNP-2c 649 2.5 3 x 10-4
Waterford 477 6.8 7 x 10-5
Watts Bar 1540 13.4 1 x 10-4
Wolf Creek 466 6.3 7 x 10-5
Yankee Rowe 872 25.0 3 x 10-5
Zion 2379 18.1 1 x 10-4

aUCB = upper confidence bound. For description and explanation of these values see Appendix G.
bObtained by dividing total fatalities by 150-mile population.
cWNP-2 = Washington Nuclear Project 2.

Note: Multiply person-rem by 0.01 to find person-sieverts; 150 miles = 240 km.

5.3.3.2.3 Benchmark Evaluations of the EI Methodology

Values for the consequences associated with nuclear power plant severe accidents have been taken from the FESs and used to establish the regressions and corresponding 95 percent UCBs presented in this chapter. As described previously, the FES values were calculated using the CRAC computer model. Using these regressions and the site-specific EI, UCB estimates for early and normalized latent fatalities and normalized total dose were obtained. As a means of assessing the performance of the EI methodology, two additional studies have been performed (Yambert and Linn 1992 and Tingle 1993). The primary goal of these studies was to demonstrate the accuracy of the EI methodology in predicting consequences associated with severe nuclear power plant accidents. In addition, insight gained from these evaluations was used to adjust values estimated using the EI regressions to reflect current, state-of-the-art calculation techniques.

The most direct means to perform this benchmarking would be to compare the outputs from CRAC used in the FESs with the output of the MACCS code, given identical inputs, for each of the FES plants. However, CRAC has undergone numerous revisions and a working version of the code used for the FES calculations is not available. Also, because the original CRAC input files for the FES plants were no longer available, detailed MACCS input files reflecting the FES plant inputs could not be created. Consequently, a direct comparison of the FES and MACCS output could not be made.

To bridge the gap between the FES values upon which the GEIS results were based and the MACCS code, CRAC2S was used to evaluate 72 hypothetical nuclear power plants. The first effort was to determine if CRAC2S would be an adequate surrogate for CRAC. Then CRAC2S and MACCS would be used to evaluate the 72 hypothetical sites and the results would be compared.

Benchmarking the CRAC2S code

A benchmark run was made using CRAC2S in order to verify its ability to satisfactorily reproduce the CRAC results found in the FESs. This was done by trying to match the input data sets for Indian Point Units 1 and 2 (Acharya and Blond n.d.). The data used in the CRAC2S main input file were identical to those used for the Indian Point CRAC main input file. However, the on-site meteorological data file used in the original Indian Point CRAC calculations was not available. In its place, meteorological data from the nearest monitoring station location, New York City, were used. The results of the benchmark found the early fatalities at Indian Point as calculated by the CRAC2S code were almost five times lower than the values calculated by CRAC. The values for latent cancers and total dose were almost identical. From these results, the staff concluded that the CRAC2S code could be used as a reasonable surrogate for CRAC in benchmarking it against the MACCS code.

Comparison of CRAC2S results to EI results for the hypothetical sites

Yambert and Linn (1992) created 72 hypothetical reactor sites. These sites were constructed by combining projected year 2030 population data for 9 existing reactor sites (5 PWR and 4 BWR) with actual meteorological data taken from 8 stations located across the U.S. The meteorological data were independent of the population data sites. Reactor locations for which population data were selected were chosen such that areas with high, medium, and low populations were represented. Similarly, the meteorological data were selected to represent a wide range of weather patterns (i.e., wet site, dry site, calm site, windy site, etc.). Values for early and normalized latent fatalities, and normalized total dose were then calculated for each of these hypothetical PWR and BWR sites using the CRAC2S computer code.5 These values were then compared to estimates obtained by applying the EI methodology to the 72 hypothetical sites.

Comparison of the two sets of estimates showed that in all cases, the consequence values calculated using CRAC2S fell below the corresponding 95 percent UCB limit predicted using the EI methodology. In the case of early fatalities, CRAC2S calculated values for PWRs averaged about 2 to 3 times lower than expected values predicted using the fitted EI regression line. This difference was greater for hypothetical BWR sites where CRAC2S calculated values averaged an order of magnitude less than the corresponding expected values from the EI regression. In addition, for a hypothetical site with a very low EI value (less than 100), the CRAC2S predictions were 4 orders of magnitude lower than the EI regression line. This large variability was attributed to the sensitivity of the CRAC2S code results to the number of persons located near the site, particularly in the 0 to 2 mile radius from the facility. The CRAC2S values for both normalized latent fatalities and normalized total dose were nearly identical to the EI fitted regression line for BWRs and were slightly below the regression line for PWRs.

The preceding paragraphs showed that the CRAC code can be adequately represented by the CRAC2S code and that the EI methodology (which is derived from values calculated by the CRAC code) predicts higher or equal consequences for all combinations of population and meteorology compared with the CRAC2S results. The final step was to compare the CRAC2S computations with the latest consequence code to determine if the CRAC2S values, and by inference, the EI methodology values, conservatively overpredict consequences when compared to the state-of-the-art consequence models and computation techniques.

A study was conducted at Brookhaven National Laboratory (Tingle 1993) which compared predictions made by the CRAC2S code to those of the MACCS code. MACCS is the consequence code currently supported by NRC for estimating consequences associated with severe reactor accidents. The Brookhaven study used MACCS to analyze the 72 hypothetical reactor sites (Yambert and Linn 1992).

Early fatality values calculated using MACCS for PWR sites were about a factor of 2 higher than those calculated by CRAC2S. For BWR sites the MACCS values were about a factor of 10 to 20 higher. Consequently, CRAC2S underpredicted MACCS early fatality values by a factor of 2 for PWR sites and a factor of 10 to 20 for BWR sites. However, the EI regression values overpredict the values for early fatality estimates from the CRAC2S code (which are based on CRAC analyses performed for the plants' FES) by factors of 3 for PWRs and 10 for BWRs. These results show that the early fatalities values estimated using the most advanced consequence computer code, MACCS, can be adequately predicted using the fitted EI regression methodology and are well within the 95 percent UCB determined by the EI regression. Consequently, the early fatality regression values shown in Table 5.10 are conservative estimates of this potential impact.

The values for total dose calculated by MACCS and CRAC2S were nearly identical, differing by no more than a factor of 2. This was the same result in comparing the CRAC2S and EI regression values. Thus, the EI regression can be used as an adequate predictor of population total dose due to a severe accident release.

For latent fatalities, the study showed some significant differences between the values predicted by the two codes. MACCS estimates for latent fatalities were consistently factors of 5-15 higher than estimates from the CRAC2S code. Since the CRAC2S values for latent fatalities were very close to the expected EI regression values, the EI regressions underestimate the current best estimates for latent fatalities by approximately a factor of 10. In order to enable the EI methodology to be an adequately conservative predictor of latent fatalities, this information was incorporated by taking the 95 percent UCB values as estimated from the EI regressions and increasing the values by a factor of 10. It is these increased values which are used in the GEIS. The adjusted latent fatality estimates are shown in Table 5.11.

5.3.3.2.4 Conclusion

As can be seen from the data in Tables 5.10 and 5.11, the risk of early and latent fatalities from individual nuclear power plants is small. It represents only a small fraction of the risk to which the public is exposed from other sources. Even if the predicted early and latent fatalities from all 118 plants were considered (that is, the risk to the population of the United States from all 118 nuclear power plants), this would only result in a predicted risk of approximately one additional early fatality per year and approximately 30 additional latent fatalities per year, which is still a small fraction of the approximately 100,000 early and 500,000 latent cancer fatalities per year from other sources.

Table 5.10 Predicted early fatality estimates per reactor-year (RY) for all sites at their middle year of license renewal
Plant Predicted UCB total early fatalities/RY (95% UCB)a
Arkansas 3.3 x 10-3
Beaver Valley 2.5 x 10-2
Bellefonte 4.0 x 10-3
Big Rock Point 2.7 x 10-3
Braidwood 3.6 x 10-3
Browns Ferry 4.3 x 10-3
Brunswick 3.5 x 10-3
Byron 2.3 x 10-3
Callaway 6.9 x 10-4
Calvert Cliffs 1.8 x 10-3
Catawba 1.7 x 10-2
Clinton 3.0 x 10-3
Comanche Peak 2.3 x 10-3
Cooper 2.6 x 10-3
Crystal River 1.5 x 10-3
D. C. Cook 8.4 x 10-3
Davis-Besse 1.4 x 10-3
Diablo Canyon 1.5 x 10-3
Dresden 4.6 x 10-3
Duane Arnold 8.0 x 10-3
Farley 1.5 x 10-3
Fermi 2 6.8 x 10-3
FitzPatrick 3.8 x 10-3
Fort Calhoun 1.7 x 10-3
Ginna 3.9 x 10-3
Grand Gulf 2.8 x 10-3
Haddam Neck 1.2 x 10-2
Hatch 2.6 x 10-3
Hope Creek 4.1 x 10-3
Indian Point 6.5 x 10-2
Kewanee 8.9 x 10-4
La Salle 3.6 x 10-3
Limerick 1.1 x 10-2
Maine Yankee 1.8 x 10-3
McGuire 1.0 x 10-2
Millstone 2.5 x 10-2
Monticello 4.1 x 10-3
Nine Mile Point 3.8 x 10-3
North Anna 9.4 x 10-4
Oconee 1.1 x 10-2
Oyster Creek 7.4 x 10-3
Palisades 4.2 x 10-3
Palo Verde 1.1 x 10-4
Peach Bottom 4.2 x 10-3
Perry 6.9 x 10-3
Pilgrim 3.7 x 10-3
Point Beach 2.5 x 10-3
Prairie Island 3.7 x 10-3
Quad Cities 4.5 x 10-3
Rancho Seco 1.1 x 10-3
River Bend 4.1 x 10-3
Robinson 3.1 x 10-3
Salem 2.9 x 10-3
San Onofre 1.1 x 10-2
Seabrook 1.1 x 10-2
Sequoyah 6.6 x 10-3
Shearon Harris 2.8 x 10-3
Shoreham 3.3 x 10-3
South Texas 3.2 x 10-3
Saint Lucie 7.7 x 10-2
Summer 1.3 x 10-4
Surry 1.6 x 10-2
Susquehanna 6.0 x 10-3
Three Mile Island 2.8 x 10-2
Trojan 3.7 x 10-2
Turkey Point 6.0 x 10-2
Vermont Yankee 4.6 x 10-3
Vogtle 1.6 x 10-4
WNP-2b 2.3 x 10-3
Waterford 1.4 x 10-2
Watts Bar 1.8 x 10-3
Wolf Creek 4.7 x 10-4
Yankee Rowe 3.3 x 10-3
Zion 5.6 x 10-2

aUCB = upper confidence bound. For description and explanation of these values, see Appendix G.
bWNP-2 = Washington Nuclear Project 2.

Table  5.11 Predicted latent fatality estimates per reactor-year (RY) for all sites at their middle year of license renewal (MYR)
Plant Nonnormalized predicted UCB total latent fatalities/RY
(95% UCB)a
Arkansas 1.7 x 10-1
Beaver Valley 1.3 x 100
Bellefonte 1.0 x 100
Big Rock Point 3.2 x 10-2
Braidwood 3.3 x 100
Browns Ferry 9.7 x 10-1
Brunswick 4.7 x 10-1
Byron 2.2 x 100
Callaway 3.6 x 10-1
Calvert Cliffs 2.3 x 100
Catawba 1.4 x 100
Clinton 1.8 x 100
Comanche Peak 3.3 x 10-1
Cooper 6.3 x 10-1
Crystal River 5.0 x 10-1
D. C. Cook 1.8 x 100
Davis-Besse 1.5 x 100
Diablo Canyon 2.5 x 10-1
Dresden 1.4 x 100
Duane Arnold 3.7 x 10-1
Farley 2.4 x 10-1
Fermi 2 1.9 x 100
FitzPatrick 5.0 x 10-1
Fort Calhoun 8.0 x 10-2
Ginna 1.5 x 10-1
Grand Gulf 9.7 x 10-1
Haddam Neck
(Connecticut Yankee)
2.0 x 100
Hatch 5.7 x 10-1
Hope Creek 2.5 x 100
Indian Point 7.7 x 100
Kewanee 2.2 x 10-1
La Salle 2.0 x 100
Limerick 3.1 x 100
Maine Yankee 3.0 x 10-1
McGuire 1.4 x 100
Millstone 3.1 x 100
Monticello 5.0 x 10-1
Nine Mile Point 6.7 x 10-1
North Anna 1.1 x 100
Oconee 1.0 x 100
Oyster Creek 1.5 x 100
Palisades 1.3 x 100
Palo Verde 2.6 x 10-1
Peach Bottom 2.0 x 100
Perry 1.7 x 100
Pilgrim 6.0 x 10-1
Point Beach 2.3 x 10-1
Prairie Island 1.7 x 10-1
Quad Cities 1.1 x 100
Rancho Seco 1.3 x 100
River Bend 8.0 x 10-1
Robinson 7.0 x 10-1
Salem 5.0 x 100
San Onofre 2.4 x 100
Seabrook 6.0 x 10-1
Sequoyah 1.1 x 100
Shearon Harris 7.3 x 10-1
Shoreham 8.0 x 10-1
South Texas 8.0 x 10-1
St. Lucie 6.3 x 10-1
Summer 1.0 x 100
Surry 9.0 x 10-1
Susquehanna 2.8 x 100
Three Mile Island 3.3 x 100
Trojan 1.5 x 100
Turkey Point 2.0 x 10-1
Vermont Yankee 9.0 x 10-1
Vogtle 7.3 x 10-1
WNP-2b 4.3 x 10-1
Waterford 3.3 x 10-1
Watts Bar 1.2 x 100
Wolf Creek 3.3 x 10-1
Yankee Rowe 6.7 x 10-1
Zion 1.8 x 100

aUCB = upper confidence bound. For description and explanation of these values, see Appendix G.

bWNP-2 = Washington Nuclear Project 2.

In addition, the prediction technique used was designed to overestimate the risk from reactor accidents. Table 5.12 illustrates this point by comparing--for the five NUREG-1150 plants--the early and latent risk values obtained from Tables 5.10 and 5.11 versus those from the NUREG-1150 analyses. In all cases the NUREG-1150 analyses predict lower risk values (one to five orders of magnitude) than the GEIS

prediction technique. Although some of the difference can be attributed to the fact that the NUREG-1150 results incorporated plant modifications discovered and corrected as a result of the NUREG-1150 analyses, some can also be attributed to the conservatism of the prediction technique used versus the more recent detailed analyses used for NUREG-1150.

Table 5.12 Comparison of predicted early and latent fatality estimates to NUREG-1150 findings
  Early fatalities Latent fatalities  
Plant Table 5.10 NUREG-1150 Table 5.11 NUREG-1150
Grand Gulf 2.8 x 10-3/RY 1 x 10-8/RY 9.7 x 10-1/RY 9 x 10-4/RY
Peach Bottom 4.2 x 10-3/RY 3 x 10-8/RY 2.0 x 100/RY 4 x 10-3/RY
Sequoyah 6.6 x 10-3/RY 3 x 10-5/RY 1.1 x 100/RY 1 x 10-2/RY
Surry 1.6 x 10-2/RY 2 x 10-6/RY 9.0 x 10-1/RY 5 x 10-3/RY
Zion 5.6 x 10-2/RY 3 x 10-5/RY 1.8 x 100/RY 8 x 10-3/RY
Note: RY = reactor-year.

5.3.3.3 Dose and Adverse Health Effects from Fallout onto Open Bodies of Water

5.3.3.3.1 Methodology

Following a severe accident, a radiation hazard may exist from the deposition of airborne, radioactive fallout onto open bodies of water. Depending on the type of water body, this hazard may lead to internal exposure from the ingestion of contaminated water or from consuming contaminated aquatic fauna. External exposure may result from swimming in the contaminated water or from recreational activities on the shoreline. The extent of the hazard is largely determined by the proximity of individuals to the reactor, the areal extent of contamination, and the ability for interdiction to reduce the exposure hazard. The risk from this exposure at plants sited on all types of water bodies is compared with that of the Fermi plant, located on Lake Erie, for which an analysis has been completed for an uninterdicted dose (NUREG-0769, Addendum 1). The potential risk is also discussed for a dose with interdiction.

This section examines such radiation exposure risk at nuclear power reactors in the event of a severe reactor accident in which radioactive contaminants are released into the atmosphere and subsequently deposited onto open bodies of water. The drinking-water pathway is treated separately from the aquatic food, swimming, and shoreline pathways. The latter three pathways are addressed collectively, and the rationale for selecting only the aquatic food pathway for analysis is presented. In the case of the drinking-water pathway, environmental parameters at representative sites are compared with such parameters at the Enrico Fermi Atomic Power Plant, Unit 2, to arrive at some indication of comparative, uninterdicted hazard.

For the aquatic food pathway, the methodology in the Fermi analysis was used with site-specific data to arrive at a comparative population dose. The Fermi analysis applied the completely mixed lake model bottom sedimentation, so that sedimentation processes are accounted for in the residence times. Population dose estimates for both the drinking-water and aquatic food pathways are compared with estimates from the atmospheric pathway. Analysis of the drinking-water pathway precedes that of the aquatic food pathway.

For the drinking-water pathway, sites adjacent to bodies of fresh water that can be used as a source of drinking water are considered. One estuarine site, which is not used as a source of drinking water, is examined for comparison purposes only. Direct deposition onto the surface water is the only pathway evaluated. The contamination of surface water bodies by the land erosion of atmospherically deposited radionuclides is not considered. One study concludes that risk from such a pathway is small compared with that of the atmospheric and terrestrial pathways (Helton et al. 1985). The study indicates that the contribution to latent facility from runoff to a great lake is less than 15 percent of what would be expected by direct deposit onto the lake. For both a great lake and a river, the expected latent facilities are only a small fraction of the latent fatalities predicted from land contamination. (Terrestrial pathways, including ingestion of crops, are considered in the atmospheric pathway in Section 5.3.3.2.)

Radioactive material released to the atmosphere tends to spread and disperse in air and dilute in water. The concentration of the contaminated material is thus related to the volume of contaminated air and water and meteorological and hydrological conditions at the time of release. These dilution processes reduce the intensity of the hazard downwind and downstream from the point of release but tend to increase the areal extent of the exposure hazard.

Several studies provide partial benchmarks that can be used to comparatively evaluate the surface water ingestion pathway at reactors located adjacent to bodies of fresh water. The Liquid Pathway Generic Study (LPGS) (NUREG-0440) examines surface water contamination via groundwater transport following a severe accident at a generic small river site, large river site, Great Lakes site, estuary site, coastal site, and dry site. Transport via groundwater to surface water bodies, however, is not directly applicable to the direct deposition pathway examined here. Results of the LPGS study indicate that the maximum individual total body dose associated with a severe core-melt accident for the small river site was one to two orders of magnitude higher than for the large river and Great Lakes site. The high values for the small river site were related to lower flow rates. Uninterdicted population drinking-water dose estimates calculated in the LPGS are as follows: large river site, 1.08 x  105 person-rem; small river site, 8.87 x 106 person-rem; and Great Lakes site, 2.34 x  106 person-rem.

Two analyses establish precedent for the direct-deposition, surface-water ingestion pathway. One (NUREG-0769, Addendum 1) is an estimate of risk performed for the Enrico Fermi Atomic Power Plant Unit 2. This assessment indicates that estimated individual and societal uninterdicted doses from the surface water pathway are of the same order of magnitude as interdicted doses from the airborne pathway. A whole body dose to an individual after a 3-year period of exposure was estimated at 0.8 rem. Interdiction comparable to that for the terrestrial pathway could substantially reduce this dose estimate and is equally likely. A second study for the Indian Point reactors (Codell 1985) developed empirical models based on considerations of radionuclide data associated with fallout from atmospheric weapons tests. A maximum 194 person-rem/RY whole-body uninterdicted dose via drinking water was estimated and compared with a maximum 2610 person-rem/RY whole-body dose from the direct airborne pathways. Although both latent and early fatality risks are associated with direct airborne pathways, only latent risks were found to be associated with the liquid (drinking-water) pathway because doses were well below the predicted rate and threshold for early fatalities or radiation illness.

Analyses in environmental documents prepared subsequent to the Fermi and Indian Point studies and the LPGS used results of these three studies as benchmarks. Representative conclusions from these documents are summarized here. Using site-specific parameters for the Perry plant (NUREG-0884), individual and societal latent cancer fatality risks from unrestricted use of Lake Erie were found to be about twice the risks from the Fermi reactor but the same order of magnitude as the air and ground pathways. For the Limerick plant (NUREG-0974), the small surface area of the nearby receiving water body (the Schuylkill River) relative to the total area of fallout is cited as the basis for concluding that the surface water pathway would be of small importance compared with the land pathway. The analysis for the Vogtle plant (NUREG-1087) qualitatively compares site-specific characteristics with those from both the Fermi and Indian Point studies and concludes that the surface water pathway would be of little importance compared with the results from atmospheric fallout onto land.

Environmental parameters important for input in performing the above analyses, and for use in analyses of additional sites, are the surface area of the receiving body, the volume of water in the body, and the flow rate. In the absence of a rigorous site-specific analyses, these data can provide estimates of the extent of contamination in the receiving water body and the residence time of the contaminant in the affected water body. Comparing these estimates and site environmental parameters with those for Fermi can provide some indication of the comparative hazard associated with drinking contaminated surface water among sites and the need for site-specific analyses. Accounting for population and meteorological data in the comparison can provide further indication of relative risk among sites.

The method used for evaluating the direct-deposition surface water ingestion pathway compares water body surface area, volume, and flow rate data at plants for which analyses have not been performed with similar data used in the Fermi 2 analysis. Table 5.13 lists all plants by adjacent water body category. Type of plant site categories have been assigned consistent with the LPGS analysis. Plants were selected for analysis to represent the spectrum of environmental characteristics found at all plants; those not evaluated are considered to possess environmental characteristics within the range of those evaluated.

Table 5.13 Nuclear power plants by water body category
Estuary or coastal Great lakes Small river or impoundment Large river
Diablo Canyon
Crystal River
Maine Yankee
Seabrook
Salem
South Texas
San Onofre
St. Lucie
Calvert Cliffs
Hope Creek
Shoreham
Surry
Millstone
Turkey Point
Pilgrim
Oyster Creek
Brunswick
Indian Point
Shoreham
Zion
Fermi
Ginna
Perry
Big Rock Point
Palisades
Nine Mile Point
Kewaunee
Cook
FitzPatrick
Davis-Besse
Point Beach
Bellefonte
Haddam Neck
Braidwood
Dresden
Duane Arnold
Waterford
Prairie Island
Fort Calhoun
McGuire
Peach Bottom
Browns Ferry
Arkansas
Hatch
Byron Station
La Salle
Wolf Creek
Yankee Rowe
Callaway
Beaver Valley
Susquehanna
Farley
Vogtle
Clinton
Quad Cities
Monticello
Cooper
Shearon Harris
Limerick
Three Mile
Island
Catawba
Summer
Oconee
Sequoyah
Vermont
Yankee
Connecticut
Yankee
Robinson
Watts Bar
North Anna
Grand Gulf
Trojan
WNP-2a
River Bend

aWNP-2 = Washington Nuclear Project 2.

Nine Mile Point and Zion were selected to include Great Lakes other than Lake Erie. Zion was selected because, of those plants located near Lake Michigan, Zion's location near the southwestern shore of the lake would enable a large portion of a plume to be deposited onto the lake near a highly populated area. Trojan and Grand Gulf were selected to represent each of the two large rivers adjacent to plants. Because the LPGS analysis indicates higher population dose estimates for small rivers (NUREG-0440), a larger number of small river sites have been evaluated. Small river sites were selected to represent (1) a range of flow rates, (2) proximity to small rivers that are the only affected water body, and (3) proximity to small rivers where other water bodies are also affected. An estuary site, in which the principal water body is not used as a source of drinking water, is included for comparison purposes only.

Great Lakes data as presented in the Fermi analysis are used in this evaluation. The assumptions used for determining river width, depth, and flow rate throughout the affected river reach are as follows: (1) large rivers and small rivers are uniformly 6- and 3-m (20- and 10-ft) deep, respectively, (2) river width at the reactor site is the same throughout the affected area, and (3) reported flow rate at the site is assumed throughout. Surface area and volume data for small lakes and reservoirs were obtained from federal and state agencies. In those cases in which part of a small lake is included in the affected area, the entire surface area and volume of the lake are included. As in the Fermi analysis, contaminant is assumed to be thoroughly mixed in the water body.

For the purposes of this analysis, it is assumed that essentially all atmospheric fallout occurs within 80 km (50 miles) of the reactor. For river sites, the "potentially" affected area includes all surface water bodies within 80 km (50 miles) of the reactor while the "likely" affected area assumes that only a limited portion of the potentially affected area is affected. (The likely affected area includes water body surface areas and volumes within 80 km of the reactor site and within 6 of the 22.5° compass sectors toward which the wind blows the greatest percentage of time.) All major surface water bodies are assumed to be sources of drinking water at the evaluation year (MYR). For Great Lakes and the estuary site, it is assumed that the adjacent water body is both the potentially and likely affected area. The potentially and likely affected populations at the MYR are obtained as above for the affected area. Data are presented in Tables 5.14a and 5.14b.

To facilitate comparison of environmental parameters and analysis those parameters among sites, selected data in Tables 5.14a and 5.14b are presented in histograms in Figures 5.8 through 5.11. The data included in the figures and a brief description of es of information provided by the figures follow. Figure 5.8 compares surface areas and water volumes of potentially affected areas. In addition to illustrating the smaller surface area of rivers available to receive fallout compared with the Great Lakes, these data provide some indication of relative dilution capacity (water volume) of the bodies of water. Figure 5.9 compares surface areas and water volumes of likely affected areas. These data further illustrate the smaller affected surface area and dilution capacity of rivers compared with the Great Lakes. Figure 5.10 depicts the three-order-of-magnitude spread in water body flow rate that contributes to additional dilution over longer time periods. Figure 5.11 compares estimated contaminant residence mes in the likely affected water bodies and the surface-area-to-volume ratios to provide some indication of the relative level of contamination per unit of water. The data in this figure (obtained by simple computation from data presented in previous figures) are the principal basis for comparison with the Fermi plant.

In addition to examining the drinking-water pathway, NUREG-0769 (1981) considers the aquatic food, shoreline, and swimming exposure pathways for the Fermi reactor. Since the principal uninterdicted, whole-body population dose in the Fermi analysis is derived from aquatic food (8 x 107 person-rem), as compared to drinking water (4 x 106 person-rem), shoreline (2 x  106 person-rem), and swimming (6 x 103 person-rem), the uninterdicted aquatic food pathway is examined. Particularly in the case of estuaries, aquatic food consumption constitutes the principal pathway of exposure.

Table  5.14a Comparison of Fermi 2 site data with data from other representative nuclear plants
Plant  Type of sitea Potentially affected surface area
(m2)
Potentially affected water volume
(m3)
Likely affected surface area
(m2)b
Likely affected water volume
(m3)b
Average flow rate
(m3/year)
Fermi Lake 2.57 x 1010 4.58 x 1011 2.57 x 1010 4.58 x 1011 1.75 x 1011
Beaver Valley Small river 9.44 x 107 2.83 x 108 6.74 x 107 2.02 x 108 3.31 x 1010
Braidwood Small river 2.27 x 107 6.82 x 107 1.28 x 107 3.58 x 107 3.47 x 109
Browns Ferry Small river 5.19 x 108 2.45 x 109 2.38 x 108 1.16 x 109 3.82 x 1010
Byron Station Small river 2.12 x 107 6.36 x 107 4.85 x 106 1.46 x 107 4.42 x 109
Callaway Small river 1.24 x 108 3.73 x 108 6.41 x 106 1.92 x 107 1.17 x 1010
Catawba Small river 6.99 x 107 4.25 x 108 6.79 x 107 3.65 x 108 3.47 x 109
Clinton Small river 1.03 x 108 4.71 x 108 3.20 x 107 1.33 x 108 2.21 x 108
Dresden Small river 2.27 x 107 6.82 x 107 6.41 x 106 1.92 x 107 3.78 x 109
Duane Arnold Small river 4.27 x 107 1.18 x 108 4.27 x 107 1.18 x 108 2.68 x 109
Grand Gulf Large river 1.19 x 108 7.15 x 108 1.19 x 108 7.15 x 108 6.02 x 1011
Hope Creek Estuary 2.07 x 109 3.45 x 1010 2.07 x 109 3.45 x 1010 3.71 x 1011
Limerick Small river 6.03 x 107 1.81 x 108 1.28 x 107 3.85 x 107 1.70 x 109
McGuire Small river 3.80 x 108 2.15 x 109 2.72 x 108 1.48 x 109 2.37 x 109
Monticello Small river 5.46 x 108 3.45 x 109 5.46 x 108 3.45 x 109 4.10 x 109
Nine Mile Point Lake 1.97 x 1010 1.64 x 1012 1.97 x 1010 1.64 x 1012 2.09 x 1011
North Anna Small river 6.59 x 107 6.81 x 108 6.08 x 107 6.66 x 108 3.15 x 108
Oconee Small river 3.46 x 108 5.86 x 109 3.46 x 108 5.86 x 109 9.46 x 108
Prairie Island Small river 5.70 x 108 3.53 x 109 5.70 x 108 3.53 x 109 1.34 x 1010
Quad Cities Small river 6.24 x 107 1.87 x 108 4.33 x 107 1.30 x 108 4.23 x 1010
Robinson Small river 9.40 x 107 5.08 x 108 1.51 x 107 5.63 x 107 1.58 x 108
Shearon Harris Small river 1.17 x 108 1.11 x 109 8.64 x 107 1.01 x 109 2.78 x 109
Summer Small river 3.33 x 108 3.35 x 109 6.79 x 107 4.19 x 108 5.36 x 109
Three Mile Island Small river 4.71 x 107 1.41 x 108 4.71 x 107 1.41 x 108 3.04 x 1010
Trojan Large river 8.73 x 107 5.24 x 108 8.73 x 107 5.24 x 108 3.85 x 1011
Vermont Yankee Small river 6.20 x 107 1.86 x 108 6.20 x 107 1.86 x 108 7.73 x 109
Wolf Creek Small river 1.50 x 108 6.41 x 108 7.04 x 107 2.05 x 108 1.42 x 109
Yankee Rowe Small river 1.97 x 108 1.87 x 109 1.48 x 107 4.43 x 107 6.62 x 108
Zion Lake 5.80 x 1010 4.87 x 1012 5.80 x 1010 4.87 x 1012 1.58 x 1011

aAs designated in Liquid Pathway Generic Study analysis (NUREG-0440).
bIn the likely affected water body.

Note: Multiply square meters by 1.20 to find square yards; multiply cubic meters by 1.307 to find cubic yards.

Table 5.14b Comparison of Fermi 2 site data with data from other representative nuclear plants
Plant Type of sitea Residence time
(years)b
Surface area to volume ratiob Potentially affected populationc Percentage of population likely to be affected Average annual wind velocity
(mph)
Average annual precipitation
(inches)
              Rain Snow
Fermi Lake 2.6 x 100 5.6 x 10-2 6,647,763 41 8.9 31 31
Beaver Valley Small river 6.1 x 10-3 3.3 x 10-1 4,169,673 48 9.3 36 46
Braidwood Small river 1.1 x 10-2 3.3 x 10-1 5,092,832 43 10.3 30 24
Browns Ferry Small river 3.0 x 10-2 2.1 x 10-1 926,918 13 8-12 47 3
Byron Station Small river 3.3 x 10-3 3.3 x 10-1 1,141,541 38 9.9 18 34
Callaway Small river 1.6 x 10-3 3.3 x 10-1 463,360 11 10.3 37 19
Catawba Small river 1.1 x 10-1 1.9 x 10-1 2,337,775 13 6.9 42 56
Clinton Small river 6.0 x 10-1 2.4 x 10-1 869,226 6 11.4 35 23
Dresden Small river 5.1 x 10-3 3.3 x 10-1 7,453,539 17 9.7 33 37
Duane Arnold Small river 4.4 x 10-2 3.6 x 10-1 754,825 46 8.0 33 31
Grand Gulf Large river 1.2 x 10-3 1.7 x 10-1 504,930 18 7.7 50 2
Hope Creek Estuary 9.3 x 10-2 6.0 x 10-2 5,424,373 26 8.9 40 23
Limerick Small river 2.3 x 10-2 3.3 x 10-1 7,615,980 37 9.1 59 20
McGuire Small river 6.2 x 10-1 1.8 x 10-1 2,543,485 23 6.9 43 6
Monticello Small river 8.4 x 10-1 1.6 x 10-1 2,815,967 82 NA 24 42
Nine Mile Point Lake 7.8 x 100 1.2 x 10-2 811,475 9 10.0 34 88
North Anna Small river 2.1 x 100 9.1 x 10-2 1,478,490 61 6.8 44 16
Oconee Small river 6.0 x 100 5.9 x 10-2 1,311,318 20 7.6 53 NA
Prairie Island Small river 2.6 x 10-1 1.6 x 10-1 2,961,583 79 6.3 25 44
Robinson Small river 3.6 x 10-1 2.7 x 10-1 991,450 28 6.2 33 NA
Shearon Harris Small river 3.6 x 10-1 8.6 x 10-2 2,122,597 49 4.6 36 NA
Summer Small river 7.8 x 10-2 1.6 x 10-1 1,385,612 27 NA 45 2
Trojan Large river 1.4 x 10-3 1.7 x 10-1 2,822,894 91 8.2 42 7
Vermont Yankee Small river 2.4 x 10-2 3.3 x 10-1 1,709,869 45 7.8 43 60
Wolf Creek Small river 1.4 x 10-1 3.4 x 10-1 273,225 35 10.3 31 15
Yankee Rowe Small river 6.7 x 10-2 3.3 x 10-1 1,796,823 38 NA NA 100
Zion Lake 3.1 x 101 1.2 x 10-2 8,199,956 10 NA 32 58

a As designated in Liquid Pathway Generic Study analysis (NUREG-0440).
b In the likely affected water body.
c Population projected for the middle year of license renewal (Table 5.5); 80-km (50-mile) radius from the site.

NA = Data not available.

Note: To convert mph to kph, multiply by 1.61; to convert inches to centimeters, multiply by 2.54.

Figure 5.8 Water body surface areas and volumes within 80 km (50 miles) of representative nuclear power plant sites (potentially affected water bodies).

Figure 5.9 Water body surface areas and volumes within 80 km (50 miles) of the reactor site and within six of the 22.5° compass sectors that exhibit the greatest percentage of time for which the wind blows toward that compass direction (likely affected water bodies).

Figure 5.10 Water body flow rate at representative nuclear power plant sites.

Figure 5.11 Contaminant residence time (flushing rate) and surface area-volume ratios for water bodies within an 80-km (50-mile) radius of selected nuclear power plants.

The process for examining the aquatic food pathway began with a comparison of edible aquatic food harvest data from major eastern U.S. and Gulf Coast estuaries, three Great Lakes, and generic large and small rivers. Sites were selected to include major aquatic food producing water bodies adjacent to plant sites. Data for East Coast estuaries, the eastern Gulf Coast, and the Great Lakes sites were obtained from the U.S. Department of Commerce (1980) and NUREG-0056. In the absence of site-specific data for river sites, generic aquatic harvest data from LPGS (NUREG-0440) were applied to a representative large and small river site. It is recognized that the quantity of aquatic food harvested from a water body varies temporally with the environmental quality of the water body. As in LPGS, it is assumed that 53 percent of the round weight of marine fish, 26 percent of the crustacea, and 28 percent of the freshwater fish are edible by humans. Mollusc data are reported in weight of edible meat. Recreational harvest data for molluscs and crustacea are unavailable and not included. While aquacultural harvests may be large locally [as much as about 3000 kg/ha in 1971 (2700 lbs/acre)], potential dose from this aquatic food source, like commercial harvests, is readily interdicted, and aquaculture harvest data are not included.

Many commercially and recreationally important marine fauna depend on the estuarine waters of the eastern Atlantic and Gulf Coasts for some portion of their life cycle. These include summer and winter flounder, striped bass, bluefish, alewives, black and rock sea basses, butterfish, croaker, weakfish, kingfish, shad, spot, menhaden, blackfish, mackerel, and shrimp. The presence of these fauna in an affected estuary for a few months and their subsequent migration from the estuary for later harvest elsewhere is acknowledged, although their contribution to the food pathway cannot be reasonably quantified and accommodated in the present analysis. However, compared with those organisms that are harvested in the estuary, the contribution of migratory fauna to estimates of dose and population risk is considered to be relatively small.

The modeling methodology in the Fermi analysis accounts for changes in radioactive nuclide concentrations in both sediment and surface water. The sediment model accounts for both the removal of radionuclides through sedimentation, as well as leaching back of the radionculides from the sediment into the water column. Surface water transport models are used to determine dispersing waterborne concentration functions, resulting in time-dependent water concentrations. The bioaccumulation approach is considered to be appropriate when the organisms have been in a reasonably constant concentration field for a period of sufficient duration for trophic and biological exchange processes to approach equilibrium. Since the time frame of interest for aquatic food concentrations extends up to 1 year, utilization of the various time-dependent waterborne radionuclide concentrations, when divided into periods of reasonably constant concentration, will provide reliable determinations of aquatic food concentrations of radioactive nuclides. A detailed discussion of the use of the bioaccumulation factor is provided in Appendix C of NUREG-0440, Liquid Pathway Generic Study.

Annual aquatic harvest data are compared with similar data used in the Fermi analysis to arrive at comparative estimates of whole-body population dose (assuming a constant source term among sites). From these data, estimates of population exposure and individual latent cancer fatality risk per RY are calculated as in the Fermi analysis. It is assumed that all of the edible aquatic harvest is consumed by humans, and a linear relationship between edible aquatic food harvest and population dose is assumed using data in the Fermi analysis for comparison.

Population exposure values are computed as the product of the projected population dose (scaled using the Fermi analysis), an assigned probability for an atmospheric release (about 2 x 10-5; consistent with the Fermi analysis), and the probability of the wind blowing toward the water body. Probability estimates of deposition onto the water body are obtained from site-specific meteorological data, which provide the fraction of time the wind is blowing toward the water body.

5.3.3.3.2 Results

Results for the aquatic food pathway follow those for the drinking-water pathway. In the case of the drinking-water pathway, comparisons of each type of environmental data with those at Fermi and the implications of those comparisons with respect to surface water contamination are presented. Great Lakes and the estuarine site comparisons precede those for river sites.

For all sites evaluated, average annual precipitation and wind data are similar to those at Fermi, which suggests no appreciably different effects from meteorological conditions on fallout among sites. The surface areas of the estuarine and Great Lakes sites evaluated are the same order of magnitude or less than those of the Fermi plant, which indicates that these sites would receive less than or essentially the same proportion of contaminant as Fermi. For the other Great Lakes sites evaluated, water volume values nearly one order of magnitude greater than those of Fermi would result in significantly greater dilution. Because water volume is one order of magnitude lower for the estuarine site, less dilution would occur. Similar flow rates for the Great Lakes sites would not alter the comparative dilution capacity. A flow rate that is greater by an apparent factor of two for the estuarine site would suggest an increase in dilution. While it is acknowledged that the contribution of tidal flow to the overall flow rate is large, tidal flow may incompletely remove contaminants, and a reasonable means of accommodating that phenomenon in the analysis is not available. Therefore, tidal flow is simply included with other flow data. Nine Mile Point exhibits a contaminant resident time that is a factor of 3 longer than that of Fermi; Zion has a residence time that is an order of magnitude greater than that of Fermi, and Hope Creek's is nearly 2 orders of magnitude less. Surface-area-to-volume ratios about a factor of 5 lower for Nine Mile Point and Zion suggest lower contaminant concentrations for these sites. Essentially the same surface-area-to-volume ratio for Hope Creek indicates a contaminant concentration similar to that of Fermi.

Comparisons of the estuary and Great Lakes site data with Fermi data indicate that contamination of these sites is of the same order of magnitude as Fermi. In the case of the Great Lakes sites, effects from longer comparative residence times are countered by lower surface-area-to-volume ratios (lower contaminant concentrations). Comparisons of those parameters for the estuarine site indicates much lower residence times than for Fermi and essentially the same contaminant concentration.

In the case of the river sites, the data corroborate the assumption that the surface area of the water body available to receive fallout is small compared with Fermi. Compared to Fermi, however, 2- to 3-orders-of-magnitude-lower water volumes for both small and large rivers result in a significantly reduced capacity for dilution in that portion of the river that receives fallout. Factors of 2 to 3 higher comparative flow rates for large rivers increase dilution capacity, whereas comparative 1- to 3-orders-of-magnitude-lower flow rates for small rivers contribute to further reduced dilution capacity. The ability of many river sites to remove contaminants rapidly is evident in the residence time data, the values of which are measured in days or weeks rather than years, as for the Great Lakes sites. Longer residence times for some small rivers are attributed to low flow rates proximity to small lakes. Comparatively higher surface-area-to-volume ratios for river sites indicate higher concentrations of contaminant in the affected portion of the river.

In contrast to the estuarine and Great Lakes sites evaluated--in which the data are amenable to direct interpretative comparison with the Fermi analysis--a direct comparison between river and Great Lakes systems does not yield viable results. Therefore, combinations of characteristics are utilized in arriving at data interpretations for river sites. While comparatively small portions of rivers may receive fallout, the concentration of contaminant in all river sites evaluated is essentially the same as or exceeds that of Fermi by as much as a factor of 6. Residence times for large rivers are more than 3 orders of magnitude less than for Fermi, while small river values vary from nearly 3 orders of magnitude less to more than a factor of 2 greater than for Fermi. Examination of the data in Table 5.14b indicates that certain small river sites could result in a combination of residence time and concentration that would exceed Fermi by a factor of 2 to 3; however, the potentially affected population is much smaller than at Fermi.

River sites that may receive relatively high concentrations of contaminant but which exhibit flow rates sufficient to enable the removal of contaminants within short periods of time (hours to several days) would reduce potential contaminant exposure time such that risk at these sites is likely to be bound by the Fermi analysis. However, such is not the case at all river sites, and they may not be bound by the Fermi analysis. These small river sites include the 13 (Table 5.15) with the following combined characteristics: (1) low on-site average annual flow rates, 2) comparatively long residence times, and (3) comparatively large surface-area-to-volume ratios.

However, particularly for these 13 sites, an estimate of the uninterdicted population dose per RY from drinking water may be made considering the Fermi value (4 x 106 person-rem). Using an estimated value of 2 x 10-5/RY for the likelihood of release and a 0.445 probability of the wind blowing over the lake results in an estimated 36 person-rem/RY for Fermi. Assuming a 25 percent MYR increase in population, the uninterdicted person-rem per RY value would be approximately 45, which is less than 2 percent of the value from the atmospheric pathway (Table 5.6). Because combined residence time and surface-area-to-volume ratios for the 13 small river sites in Table 5.15 exceed values at Fermi by less than a factor of 3, and these sites have populations lower than Fermi by at least a factor of 2, the population dose at these sites would be expected to remain a small fraction of the value estimated for the atmospheric pathway. In addition, these sites are considered to be at least as amenable to interdictive measures as Fermi, which would further reduce population dose.

Table 5.15 Reactor sites that may not be bound by the Fermi 2 surface water analysis
Catawba Clinton North Anna
McGuire Monticello Robinson
Oconee Prairie Island Wolf Creek
Shearon Harris Summer  
Yankee Rowe Duane Arnold  

Results of the uninterdicted aquatic food pathway analysis are presented in Table 5.16, which compares estimated annual aquatic food harvest, population dose, and population exposure among sites. Because of conservative assumptions in several steps in the analysis, these values are considered to constitute upper bound value estimates. It is also assumed that the entire harvest is consumed by humans, which results in maximum population dose to those sites with the greatest harvest, independent of population. This assumption implies a linear relationship between harvest and population dose.

It can be seen in Table 5.16 that those sites with the greatest aquatic harvest result in the highest values of population exposure per RY. For most sites, population exposure estimates are well below those estimated for the atmospheric pathway (Table 5.6). For those with values that exceed the atmospheric pathway value, it is reasonable to expect that dose reduction would occur as a result of interdiction. Interdiction has the potential to reduce the dose by factors of from 2 to 10 (NUREG-0769, Addendum I; NUREG-0440, Table 7.3.2); accordingly, values of population exposure for all sites would be essentially the same as or significantly less than values from the atmospheric pathway.

Interdiction could consist of preventing use of the water or making contaminated food difficult to obtain. Thus, limiting people's contact with contamination through such measures as preventing or confiscating catches of recreational and commercial fish and shellfish, prohibiting water-based recreation, and eliminating surface water as a drinking-water source would have to be employed.

This type of interdiction might have to be long term because the residence times could be long in certain situations. The food pathway of ocean and estuarine sites would be the hardest in which to effect interdiction. Not only are the physical transport mechanisms of these systems complex, but many of the important recreational and commercial organisms are highly mobile. Thus, the ability of humans to obtain these organisms would need to be controlled.

5.3.3.3.3 Conclusion

Analyses for both the drinking water and aquatic food pathways have been performed with and without considering interdiction. In the case of the drinking-water pathway, the Great Lakes and the estuarine sites are bound by the Fermi analysis while small river sites with relatively low annual flow rates, long residence times, and large surface-area-to-volume ratios may potentially not be bound by the Fermi analysis. In all cases, however, interdiction can reduce relative risk to levels at or below that of Fermi and significantly below that for the atmospheric pathway. River sites that may have relatively high concentrations of contaminants but which remove contaminants within short periods of time (hours to several days) are amenable to short-term interdiction. A similar level of reduced risk can be achieved at those sites with longer residence times (months) by more extensive interdictive measures.

Table 5.16 Comparison of aquatic food harvest, uninterdicted population dose and exposure among representative sites
Plant Water body Annual edible aquatic food harvest
(kg)a
Estimated population dose--whole body
(person-rem)b
Population exposure per reactor-year
(person-rem)c
Calvert Cliffs Chesapeake Bay 3.0 x 108 7.1 x 108 5500
Crystal River Gulf Coast 6.4 x 107 1.5 x 108 1400
Fermi Lake Erie 6.7 x 107 1.6 x 108 1400
Hope Creek Delaware Bay 8.1 x 107 2.0 x 108 270
Millstone Long Island Sound 5.8 x 107 1.4 x 108 500
Nine Mile Point Lake Ontario 1.8 x 107 4.2 x 107 300
Seabrook Gulf of Maine 7.4 x 107 1.7 x 108 2100
Zion Lake Michigan 2.7 x 107 6.4 x 107 650
Small river Generic 2.2 x 104 5.2 x 104 0.4
Large river Generic 2.0 x 104 4.7 x 104 0.4

aIncludes combined commercial and recreational harvest estimates.
bAssumes linear relationship between aquatic harvest and population dose using data in Fermi analysis (NUREG-0769) as basis for comparison.
cDerived as in the Fermi analysis (NUREG-0769) using site-specific data to obtain wind probability values. For the river sites, meteorological data from those sites in the drinking-water pathway analysis with the highest likely/potentially affected surface area ratios were used.

Note: Multiply by 2.2 to convert kilograms to pounds; multiply by 0.01 to convert person-rem to person-sieverts.

For the aquatic food pathway, population dose and population exposure per RY are directly related to aquatic food harvest. For river sites, uninterdicted population exposure is orders of magnitude lower than that for the atmospheric pathway. For Great Lakes sites, the uninterdicted population exposure is a substantial fraction of that predicted for the atmospheric pathway but is reduced significantly by interdiction. For estuarine sites with large annual aquatic food harvests, dose reduction of a factor of 2 to 10 through interdiction provides essentially the same population exposure estimates as the atmospheric pathway.

For these reasons, population dose for the drinking-water pathway is found to be a small fraction of that for the atmospheric pathway. Risk associated with the aquatic food pathway is found to be small relative to the atmospheric pathway for most sites and essentially the same as the atmospheric pathway for the few sites with large annual aquatic food harvests.

5.3.3.4 Possible Releases to Groundwater

5.3.3.4.1 Methodology

This section discusses the potential for radiation exposure from the groundwater pathway as the result of postulated severe accidents at a nuclear reactor during the license-renewal period. Severe accidents are the only accidents capable of producing significant groundwater contamination.

For this pathway, the core is postulated to "melt down," breach the reactor vessel, and fall onto the reactor building floor. As a result of chemical energy and decay heat, the melted fuel reacts with the concrete floor. The basemat of the containment building is eventually breached, and molten core debris and radioactive water penetrate strata beneath the plant. The soluble radionuclides in the debris can be leached and transported with groundwater and contaminated reactor water to downgradient domestic wells used for drinking water or to surface water bodies used for drinking water, aquatic food, and recreation. In reality, the probability of such an accident is small. In general, the probable frequency of core melt is less than 10-4/RY; however, some plants may have core damage frequencies that slightly exceed this value. From NUREG-1150, the conditional probability of basemat melt-through ranges from 0.05 to 0.24 occurrences per core melt. Therefore, it is reasonable and conservative to assume a 10-4 probability of occurrence of basemat melt-through per reactor-year for this analysis.

In this analysis, site-specific information on groundwater travel time; retention-adsorption coefficients; distance to surface water; and soil, sediment, and rock characteristics is compared with previous groundwater contamination analyses. Previous analyses are contained in LPGS and FESs.

First, uninterdicted doses received through the groundwater pathway are compared; however, the effects of interdiction are discussed later in this section.

Groundwater contamination due to severe accidents has been evaluated generically in LPGS (NUREG-0440). LPGS assumes that core melt and subsequent basemat melt-through occur and evaluates the consequences. LPGS examines six generic sites using typical or comparative assumptions on geology, adsorption factors, etc. Twenty-seven sites (hereafter called current sites) of the 74 nuclear power plant sites performed groundwater pathways analyses for FES and compared the results with the conclusions in LPGS. These comparisons indicate whether the current plant sites present significantly larger population doses than those calculated in LPGS. For the other 47 sites (hereafter called earlier sites) for which no groundwater pathway analyses were performed, this study compares the physical characteristics of each site with both the generic sites used in the LPGS study and the current sites.

The LPGS results are believed to provide generally conservative uninterdicted population dose estimates in the six generic plant-site categories. Five of these categories are site groupings in common locations adjacent to small rivers, large rivers, the Great Lakes, oceans, and estuaries. In a severe accident, contaminated groundwater could reach nearby surface water bodies and the population could be exposed to this source of contamination through drinking of surface water, ingestion of finfish and shellfish, and shoreline contact. Exposure by drinking contaminated groundwater is considered to be minor or nonexistent in these five categories because of a limited number of drinking-water wells. The sixth category is a "dry" site located either at a considerable distance from surface water bodies or where groundwater flow is away from a nearby surface water body. In this case, the only population exposure results from drinking contaminated groundwater. In each LPGS category, the generic site is a PWR that produces 1150 MW(e) and is located 457 m (1500 ft) from the nearest surface water (or from the boundary of the exclusion area in the dry site case).

In LPGS, five of the site categories (the dry site is the one exception) have the same generic groundwater characteristics. The groundwater velocity is 2.04 m/day (6.7 ft/day) and travel time to the nearest surface water is 0.61 year. The adsorption-retention factors (the products of these factors and the groundwater travel time are travel times of each isotope) for 90Sr and 137Cs are 9.2 and 83, respectively, and the corresponding amounts of each isotope reaching surface water (taking into account their radioactive decay rates) are 88 percent and 31 percent of the core-melt inventory, respectively. The groundwater velocity and travel time to the exclusion boundary of the dry site are 1.32 m/day (4.35 ft/day) and 0.95 year, respectively. In this case, the adsorption-retention factors

(retardation coefficients) for 90Sr and 137Cs are 28 and 253, respectively. All LPGS parameters were taken from the WASH-1400 study (NUREG-75/014).

A summary of uninterdicted population doses for LPGS generic wet sites is provided in Table 5.17. The largest LPGS drinking-water dose to the population is attributed to the small river site (8.9 x 106 person-rem). The largest total population dose is attributed to the estuarine site (1.8 x 108 person-rem), which is more than an order of magnitude greater than the next largest total population dose (9.9 x  106 person-rem) for the small river site.

In the following comparisons, current FES results are tabulated separately and by generic category for ease of comparison. A major objective of these comparisons is to establish whether the generic LPGS or current FES severe accident liquid pathway analyses provide conservative uninterdicted population dose estimates in each site category. According to LPGS (NUREG-0440), the generic liquid pathway uninterdicted dose estimates are one or more orders of magnitude lower than those attributed to the atmospheric pathway. Therefore, if the 27 current site FES dose estimates do not significantly exceed those of LPGS, the liquid pathway may also be considered an insignificant contributor to the population dose that could result from a severe accident for the plants. The remaining 47 earlier sites are then placed into the appropriate categories and their physical characteristics are compared with those of the selected largest dose estimate site to determine if they also represent comparatively insignificant contributors to population dose.

Table 5.17 Summary of surrogate uninterdicted population doses for Liquid Pathway Generic Study base cases
Generic sitea Drinking-water dose
(person-rem)b
Seafood ingestion dose
(person-rem)
Shoreline exposure
(person-rem)
Total
(person-rem)
Large river 1.08 x 105 6.83 x 103 7.457 x 103 1.228 x 105
Small river 8.865 x 106 6.563 x 105 3.577 x 105 9.88 x 106
Great Lakes 2.34 x 106 6.369 x 105 4.066 x 105 3.540 x 106
Estuary 0 1.463 x 107 1.626 x 108 1.772 x 108
Coastal 0 5.348 x 105 2.36 x 103 5.372 x 105

a Data for the dry site are not provided.
b Multiply person-rem by 0.01 to find person-sieverts.

Source: NUREG-1054.

Note: These doses should not be accepted at face value, but should be used only for comparison with other sites.

5.3.3.4.2 Small River Sites

Table 5.18 compares results of current small-river plant sites (i.e., those with groundwater pathway analyses in their FESs) with the LPGS results. Beneath the name of each plant is its location, a brief description of the groundwater pathway, surface water bodies affected, and average stream flow rates past each plant site. Numerical tabulations include the estimated percentages of radionuclides 90Sr and 137Cs reaching the nearest downgradient surface water body as well as groundwater travel times and radionuclide adsorption-retention factors, which--together with radionuclide decay rates--were used to calculate these percentages. Also included are estimates of the magnitude of three potential uninterdicted population dose sources: drinking-water, finfish- and shellfish-ingestion, and shoreline-swimming exposure.

Population dose-estimate ratios (plant/LPGS) for drinking water, ingestion, and shoreline exposures are presented in the right-hand column of Table 5.18. These dose-estimate ratios are also based on the assumption of no interdiction. At face value, the majority of these dose-estimate ratios are several orders of magnitude less than 1. However, these dose-estimate ratios are sometimes based on parameters that may be nonconservative. There was also a lack of population dose information in FESs. Therefore, dose estimate ratios must be inferred from the percentage of radionuclides reaching surface water. At several sites, these ratios are near unity (i.e., the site and LPGS dose estimates are the same order of magnitude); at two sites (Byron Station and Clinton), ingestion dose ratios are significantly larger than unity (i.e., the site seafood ingestion dose is more than an order of magnitude greater than the corresponding LPGS dose).

Table 5.18 Current small river site severe accident liquid pathway analyses compared with Liquid Pathway Generic Study (LPGS) results

Average on-site flow rates less than 2830 m 3 /s (100,000 ft 3 /s)

Plant Location and ground-water pathway Distance from reactor to nearest downgradient surface water
(m) a
Groundwater velocity
(m/d) a
Groundwater travel time from reactor to surface water
(years)
Adsorption- retention factor
(Sr/Cs)
% radio-nuclide reaching surface water
(Sr/Cs)
Drinking-water population
(x 106)
Annual aquatic food catch
(x 106 kg) b
Annual shoreline user rate
(user-h x 106)
Dose-estimate ratios: drinking water, ingestion, direct contact
LPGS 20 km SW of Oak Ridge, Tenn.--soil and weathered limestone to Clinch River, then to Tennessee, Ohio, and Mississippi rivers. Average flow rate is 50 m3/second. c 457 2.04 0.6 9.2/83 88/31 0.62 1.2 110 1, 1, 1
Beaver Valley 40 km d NW of Pittsburgh, Pa.--terrace alluvial aquifer to Ohio River, then to Mississippi River. Average on-site flow rate of river is 1,050 m3/second. ~137 0.03 12.3 9.2/83 e 6/0 NP f NP NP ~0.01 (combined)
Braidwood 38 km SSW of Joliet, Ill.--pleistocene till and Pennsylvanian sandstone to Mazon River, then to Kankakee, Illinois, and Mississippi rivers. Average on-site flow rate of river is 110m3/second. 5940 (Strip-mine area) 0.01 1780 1/1 g 0/0 NP NP NP ~0 (combined)
Byron Station 27 km SW of Rockford, Ill.--through limestone to springs discharging to tributaries of Rock River, then to the Mississippi River. Average on-site flow rate of river is 140 m3/second. 1100 0.12 24.7 1/1 g 56/57 2.1 9.6 110 1.3, 24, 3
Callaway 16 km SE of Fulton, Mo.--shallow limestone-sandstone aquifer to tributary of Mud Creek, then to the Missouri and Mississippi rivers. Average on-site flow rate of river is 2000 m3/second. ~760 0.03 68.5 7.1/14.5 <1/~0 NP NP NP << 1 (combined)
Catawba 10 km NNW of Rock Hill, S.C.--through shallow fractures in granite to Lake Wylie, then to Catawba River and a set of lakes near Charleston, S.C. Average on-site flow rate of river is 110 m3/second. 210 0.61 1.0 6/560 88/~0 0.43 3.0 NP 0.7, 1.8, 0
Clinton 10 km E of Clinton, Ill.--sand lenses in glacial till to Lake Clinton, then to Salt Creek, Sangamon, Illinois, and Mississippi rivers. Average on-site flow rate of river is 7 m3/second. NP NP 0.5 17/211 h

68/960 i

82/10

42/~0

2.1 7.0 NP 0.6, 23 h NP

0.3, 1.3 i NP

Harris 32 km SW of Raleigh, N.C.--fractures in diabase (volcanic) rocks to cooling water reservoir, then to Cape Fear River and Atlantic Ocean. Average on-site flow rate of river is 88 m3/second. 730 0.30 6.6 49/480 0.1/~0 NP NP NP << 1 (combined)
Limerick 34 km NW of Philadelphia, Pa.--shallow fractures in sandstone/siltstone to Shuylkill River, then to the Delaware River, Delaware Bay and Atlantic Ocean. Average on-site flow rate of river is 54 m3/second. 240 0.20 3.3 20/193 18/~0 1.9 NP NP ~1, NP, NP
South Texas 19 km S of Bay City, Tex.--wetlands to the Colorado River and the Gulf of Mexico (could have been classified as an estuary site). 4900 (Wetlands) 0.21 62.6 9.2/83 c ~0/~0 NP NP NP << 1 (combined)
Summer 42 km NW of Columbia, S.C.--shallow fractures in igneous and metamorphic rocks to the Broad River, then to the Congaree River and Lakes Marion and Moultrie, then marshes and estuaries to the Atlantic Ocean. Average on-site flow rate of river is 170 m3/second. NP NP 7.4 8.6/154 19/~0 ~0.62 3.5 NP 0.9, 2.0, ~0
Susquehanna 11 km NE of Berwick, Pa.--lateral flow in fractured shale and Pleistocene-Holocene alluvium to a tributary of Lake Took-A-While, then to the Susquehanna and Delaware rivers to Delaware Bay and the Atlantic Ocean. Average on-site flow rate of river is 380 m3/second. NP NP 9.2 35/500 0.2/~0 NP NP NP << 1 (combined)
Vogtle 42 km SE of Augusta, Ga.--construction backfill to shallow limestone, discharge to springs feeding Mathes Pond, then to the Savannah River and Atlantic Ocean. Average on-site flow rate of river is 340 m3/second. 850 0.15 15.3 21.5/165 0.05/~0 0.06 < 1.2 NP 10-5, 10-4, 0
Wolf Creek 6 km NE of Burlington, Kans.--shallow limestone to cooling water reservoir, then to Neosho River (presumably), then through a series of Lakes to the Arkansas and Mississippi rivers. Average on-site flow rate is 45 m3/second. 790 0.006 356. 9.2/83 c ~0/~0 NP NP NP << 1 (combined)

aMultiply by 3.28 to convert to ft or ft/d.
bMultiply by 2.20 to convert to pounds.
cMultiply by 35.3 to convert to ft3/second.
dMultiply by 0.625 to convert to miles.
eAssumed same value as used in the LPGS.
fNP = not provided.
gHighly conservative estimate (no adsorption).
hConservative estimate.
iRealistic estimate.

However, the seafood ingestion dose at the generic small river site is only about 6 percent of the total generic population dose as shown in Table 5.17, and Byron Station's total population dose is about three times that of the LPGS generic site for small rivers. The dose-estimate ratios in Table 5.18 suggest that the Byron Station FES severe accident liquid pathway analysis provides the highest population dose groundwater pathway for sites located along small rivers. However, Byron Station's groundwater pathway population dose is less than an order of magnitude greater than the LPGS dose. Therefore, the FES groundwater pathway population dose for the small river category does not exceed that of the atmospheric pathway.

At 11 of the 27 current sites, the population doses were found to be essentially zero. In these cases, percentages of 90Sr and 137Cs reaching surface water are generally low, based on long groundwater travel times, large adsorption-retention factors, or both. In some cases, the liquid pathway analysis was terminated without calculating population dose estimates. Most current FESs refer to Isherwood (1977), NUREG/CR-1596, or Parsons (1962) for representative adsorption data instead of acquiring site-specific data.

In contrast, the Byron Station liquid pathway analysis (citing a lack of site-specific adsorption data) used the highly conservative assumption that neither 90Sr nor 137Cs was adsorbed and that these isotopes would be transported at the same velocity as groundwater. As a result, the analysis shows that more than half the Byron Station severe accident inventory of 90Sr and 137Cs would reach surface water. Consequently, dose sources and estimates were found to be high.

In NUREG-1054, Codell recommends caution in using adsorption data to characterize the groundwater pathway through fractured rock. In cases involving groundwater pathways in open fractures that were not accounted for, the adsorption-retention factors and groundwater velocity may have been significantly overestimated and underestimated, respectively, leading to nonconservative (low) population dose estimates. Current FESs for Callaway, Harris, Limerick, Summer, Vogtle, and Wolf Creek all cite low groundwater velocities (site-specific data) or large adsorption-retention factors (literature values as cited previously) as well as groundwater pathways in fractured rock. Therefore, because it was not clear whether these sites were bound by the Byron Station population dose estimates, they were investigated further. However, the Summer and Vogtle sites have much smaller drinking-water populations and seafood catches than Byron Station. So, even if no adsorption were assumed for these two sites, their dose estimates would not be likely to exceed those of Byron Station.

The Callaway, Harris, Limerick, Summer, Vogtle, and Wolf Creek FSARs contain liquid pathway analyses for a postulated rupture of a liquid radioactive waste tank. Each of these plants is discussed further in the following paragraphs.

In four cases (Callaway, Harris, Limerick, and Vogtle), the FSAR findings are compatible with those of corresponding FESs. Furthermore, it is clear from the Limerick FSAR that the fracture pathway is not a significant liquid conduit (runoff water from Hurricane Agnes had to be pumped from the Limerick open excavation because it did not drain through the fractured rock). It is assumed in the Vogtle FSAR that fractured limestone drains freely, but it has been adequately demonstrated that radionuclides are sufficiently retained by an extensive construction fill between the radioactive waste tank and the limestone. Fractured volcanic rocks have been identified as groundwater conduits in the Harris FSAR, but the groundwater velocity and radionuclide retardation values for these rocks are not significantly different from those listed in the FES; and in the Callaway FSAR, unfractured sandstone, rather than a fractured limestone, has been identified as the primary groundwater pathway.

In one case (Summer), the FSAR analysis is inconsistent with that of FES. The Summer FSAR identifies the groundwater pathway as weathered rock without clay minerals and gives no credit for cation exchange (adsorption-retention factor = 1); however, in FES, adsorption-retention factors are 8.6 and 154 for strontium and cesium, respectively. Based on FSAR's conservative adsorption estimate, 77 percent of the strontium and cesium would reach surface water (about 50 percent more radionuclides than at Byron Station), rather than 19 percent strontium and < 1 percent cesium as estimated in the FES. Based on FSAR data, Summer's plant-LPGS drinking-water dose ratio would be between 3 and 4, rather than 0.9 as listed in Table 5.18. However, the drinking water population and aquatic catch estimates are less than those of Byron Station by about a factor of 3. Therefore, the conservative analysis for Summer and LPGS result in roughly similar population doses that are less than that at Byron Station.

In one case (Wolf Creek), the FSAR analysis is not sufficiently detailed to permit a direct comparison with FES. The Wolf Creek FES groundwater velocity estimate is extremely low compared with those of other small river sites [0.006 m/day (0.0065 yd/day) compared with 0.01 to 0.6 m/day (0.012 to 0.65 yd/day) for other sites]. All but one site reported groundwater velocities at least an order of magnitude greater than that at Wolf Creek. Therefore, Wolf Creek's estimated groundwater velocity may be unreasonably low. Based on this low groundwater velocity, FES concludes that less than 1 percent of the radionuclide inventory in a core-melt accident would reach surface water.

The Wolf Creek liquid pathway dose estimates have been recalculated, based on a higher groundwater velocity. It was assumed that the groundwater velocity at Wolf Creek may be similar to that at South Texas [0.21 m/day (0.23 yd/day)] where the terrain is similarly flat. If one assumes no retardation of radionuclides at Wolf Creek, 80 percent of the radioactive strontium would reach surface water before decaying--about 1.5 times more than at Byron Station. However, Byron Station has a larger downstream population dose (St. Louis and Memphis for Byron Station compared with Fort Smith and Little Rock, Arkansas, for Wolf Creek). The Mississippi River between St. Louis and its confluence with the Arkansas River has a greater fish catch. Therefore, for this Wolf Creek analysis, population dose is considered to be comparable to that at Byron Station (the largest groundwater pathway population dose estimate for small rivers).

In general, sites that are located on a floodplain or on glacial till may be expected to produce dose estimates that are lower than those for the LPGS case, assuming that the melted reactor core does not reach bedrock beneath the site (resulting in a groundwater pathway through fractured rock). Low groundwater gradients on floodplains and low hydraulic conductivity in glacial till (with the exception of glacial outwash deposits) generally result in low groundwater velocities. Furthermore, significant percentages of clay minerals are generally available for radionuclide adsorption in both floodplain deposits and glacial till. Beaver Valley, Braidwood, Clinton, South Texas, and possibly Susquehanna are representative of sites on floodplains or glacial till.

Table 5.19 compares earlier small river sites (i.e., sites without FES groundwater pathway analyses) with the Byron Station and LPGS generic small river sites. Because no severe accident pathway analyses were provided for the earlier sites, reactor size, distance from the reactor to the nearest downgradient surface water, and river flow rates are the only parameters directly comparable to the current site analysis. The total downstream population at risk (all municipalities along the river from the plant-site to the sea) is an indicator of potential drinking-water, ingestion, and shoreline exposures for these sites. No groundwater travel times or adsorption-retention factors are available for these sites.

Cooper, Farley, Fort Calhoun, Hatch, and Quad cities are sites listed in Table 5.19 that most likely would not exceed the Byron Station or LPGS dose estimates because of their locations on thick floodplain or coastal plain sediments. Groundwater pathways through fractured rock or Pleistocene outwash deposits are unlikely at these sites. Floodplains and coastal plains are expected to have low groundwater velocities (because of low groundwater gradients) or relatively high adsorption-retention factors (because of high clay content), or both.

The other sites in Table 5.19 have large populations downstream at risk and are located on either fractured rock or a Pleistocene aquifer (suggesting the presence of outwash deposits) or have a groundwater pathway that is not well known. Groundwater velocities may be higher than those at Byron Station, and adsorption cannot be relied upon to delay entry of 90Sr and 137Cs into nearby surface water. Therefore, uninterdicted doses at some of these sites may significantly exceed those at Byron Station. It is uncertain whether all small river site groundwater pathway population doses without interdiction would be less than that of the atmospheric pathway.

5.3.3.4.3 Large River Sites

Table 5.20 compares current large river plant sites with the LPGS generic large river site. The format for this table is the same as that for Table 5.18. The Grand Gulf and River Bend plants are located far from the Mississippi River shoreline but on its floodplain where the groundwater velocity is expected to be low and floodplain sediments would be expected to adsorb radionuclides to some extent. The River Bend site analysis was based on conservative estimates of adsorption-retention factors, and the resulting plant-LPGS population dose ratio is 0.39. The adsorption-retention factors for Grand Gulf are higher; however, using the same adsorption-retention factors for Grand Gulf as for River Bend would not have produced significantly higher doses. Therefore, the dose-estimate ratios

in Table 5.20 suggest that the LPGS generic site analysis provides the largest uninterdicted population dose estimate for large rivers, at least for sites with locations similar to those of Grand Gulf and River Bend.

The Washington Nuclear Project 2 (WNP-2) site in Table 5.20 differs from the other two sites in that the assumed groundwater velocity is higher than that used in LPGS. However, the adsorption-retention factors for WNP-2 appear to be nonconservative. If conservative adsorption-retention factors similar to those of Grand Gulf and River Bend are used for WNP-2, an estimate of the population dose can be made (Portland, Oregon, is downstream of the WNP-2 site). The staff's conservative analysis for WNP-2 (using adsorption and drinking-water dose

from River Bend and aquatic catch from Grand Gulf) yields a total population dose of 1.5 times that of the LPGS for large rivers. The difference in the conservative WNP-2 and LPGS groundwater pathway population doses is small in comparison with the order of magnitude greater atmospheric pathway population dose.

Table 5.21 compares the only earlier large river plant site (Trojan) with the LPGS generic large river site. This site is located much closer to the nearest downgradient surface water (the Columbia River) than is the generic site, and fractured rock underlies the site, suggesting that the groundwater travel time may be less than that of the LPGS. Therefore, the uninterdicted dose from the Trojan Plant would probably be less than that of the LPGS study.

Table 5.21 Earlier large river sites without severe accident liquid pathway analyses compared to the Liquid Pathway Generic Study (LPGS) results

Average on-site flow rates greater than 2830 m 3 /second (100,000 ft 3 /second)

Plant Location and ground-water pathway Reactor size
[MW(e)]
Distance from reactor to nearest downgradient surface water (m)a Average/river flow rate (m3/second)b Downstream population at risk x 106
(1988)
LPGS On the lower Mississippi River. 1150 457 13,900 1.9
Trojan 48 kmc NW of Portland, Oreg.--soil and shallow fractured rock to the Columbia River, then to the Pacific Ocean. 1130 ~90 12,200 0.1

aMultiply by 3.28 to convert to ft.
bMultiply by 35.3 to convert to ft3/second.
cMultiply by 0.625 to convert to miles.

5.3.3.4.4 Great Lakes Sites

Table 5.22 compares current Great Lakes plant sites with the LPGS generic Great Lakes site. These sites are all located on or adjacent to flat Pleistocene lake bed sediments which underlie modern lake sediments and shorelines. These sediments generally have a high clay and silt content. Groundwater passing through fractured rock must also pass through these lake-bed sediments before reaching the lake. Therefore, groundwater gradients and groundwater velocities are expected to be low, and adsorption-retention factors are expected to be high relative to those of the generic site. However, the current sites have larger populations at risk (the generic site is on Lake Ontario, the farthest downstream of the string of Great Lakes) and are closer to the shoreline than the Great Lakes generic site. Taking all these factors into account yields dose-estimate ratios between 0 for Nine Mile Point and 1.4 for Fermi (the latter site has, by far, the largest drinking water population). Therefore, the severe accident liquid pathway analysis for the Fermi site provides the largest uninterdicted population doses for current FES sites adjacent to the Great Lakes. The differences between groundwater population doses for the Fermi and LPGS sites are small in comparison with differences in atmospheric pathway doses for the sites.

Table 5.23 compares earlier Great Lakes sites with the Fermi and LPGS generic Great Lakes sites. Populations at risk at sites near standing bodies of water (lakes, estuaries, and oceans) are defined as all people living within 80 km (50 miles) of the site, rather than as all people living downstream from a river site. Geologic conditions at these sites are similar to those of the current plant sites described in Table 5.22. Although some of these sites have groundwater pathways through Pleistocene outwash and fractured rock, groundwater must also pass through lake-bed sediments before reaching the lake. The Zion site is comparable to Fermi in size, distance from shoreline, and population within 80 km. Therefore, Zion's population doses would probably be similar to those of Fermi. All other sites would have population doses lower than those of Fermi, based on smaller reactor sizes, greater distances to shoreline, and lower populations within 80 km. Therefore, groundwater pathway population doses at all Great Lakes sites are expected to be less than or equal to that of the Fermi site.

5.3.3.4.5 Ocean Sites

Table 5.24 compares current ocean plant sites with the LPGS generic ocean site. The Seabrook severe accident liquid pathway analysis has the largest estimated uninterdicted population doses for sites adjacent to the ocean. Based on short groundwater travel time and low adsorption-retention factors, nearly all the strontium inventory (94 percent) and more than half the cesium inventory (58 percent) reaches the Gulf of Maine, compared with LPGS generic estimates of 88 percent and 31 percent, respectively. These percentage comparisons suggest that a severe accident at Seabrook has the potential for producing a larger maximum individual dose than that of the LPGS generic ocean site. In consideration of the large annual seafood catch and shoreline user rates, the uninterdicted total population dose estimate for Seabrook is 6 times that of the LPGS generic ocean site. Seabrook's estimated groundwater pathway population dose is still below that of the atmospheric

pathway but at a reduced level of confidence.

Table 5.25 compares earlier ocean plant sites with both the LPGS generic and Seabrook ocean sites. The Seabrook reactor is the largest in Table 5.25. This reactor is also closest to the shoreline and has a large nearby population comparable to that of the Pilgrim site. However, the Pilgrim reactor is little more than half the size of Seabrook and, thus, may have a population dose roughly half that of Seabrook. The Diablo Canyon reactor is roughly comparable to Seabrook in size and distance from shore but has only one-tenth the population within 80 km. Furthermore, the sandstones and volcanic rocks at Diablo Canyon may have higher adsorption-retention factors than the quartzite and granite at Seabrook. Therefore, Diablo Canyon's potential population dose is expected to be at least 1 order of magnitude less than that of Seabrook and also less than that of the LPGS generic ocean site. Turkey Point is located on a flat coastal plain where the groundwater gradient is expected to be low; hence, groundwater velocity and travel time are expected to be correspondingly low and high, respectively, with respect to Seabrook. The Turkey Point reactor is located about the same distance from the shoreline as is the LPGS generic site and four times farther inland than Seabrook. However, a barge canal is less than 50 m (164 ft) from Unit 3. Interdiction at Turkey Point could be accomplished by closing off or filling in the barge canal. Thus, based on the above site-specific assumptions, it can be concluded that Seabrook represents the largest uninterdicted population dose at ocean sites other than Turkey Point.

5.3.3.4.6 Estuarine Sites

Table 5.26 compares current FESs for which groundwater pathway analysis is available with the LPGS generic estuarine site. There is only one estuarine site (Hope Creek) for which a current FES is available. However, a detailed severe accident liquid pathway analysis is also available for the Indian Point site (ConEd 1982).

Hope Creek's estimated uninterdicted total population dose is less than 1 percent of the LPGS generic dose for estuaries. The LPGS annual aquatic catch and shoreline use are 3 and 83 times, respectively, as large as Hope Creek's. Even if 100 percent of Hope Creek's strontium inventory reaches surface water, the LPGS population dose would not be exceeded.

Indian Point's estimated uninterdicted population doses vary from insignificant to 0.44 times that of the LPGS, depending upon the magnitude of the assumed strontium and cesium adsorption-retention estimates. The first Indian Point estimates in Table 5.26 are from Consolidated Edison (ConEd) (1982). These dose estimates are very low (from 1.5 x 105 to 4.9 x 105) compared with the LPGS generic estuarine dose estimate (1.8 x 108) in Table 5.26. ConEd's adsorption-retention factors for strontium and cesium are 270 and 1626, respectively, compared with 9.2 and 83 for the LPGS case. The very large adsorption-retention factors at Indian Point are based on the assumption that groundwater flow is through intergranular pore spaces in rock with very low porosity (0.5 percent). ConEd's groundwater flow assumption may be nonconservative because flow is more likely to occur through open fractures rather than intergranular pore spaces. Other

parameters (reactor size, groundwater travel time, aquatic catch, and shoreline user-hours) are roughly comparable for Indian Point and the LPGS.

The second set of Indian Point estimates in Table 5.26 is based on a conservative assumption used in this analysis. Indian Point's reactor foundations are located on highly fractured (brecciated) limestone, and some of these fractures are open (ConEd). If the primary groundwater flow is through open fractures, the effective porosity of the fractured rock may range between 0 percent and 20 percent. Assuming that the effective porosity is 10 percent, adsorption-retention factors for strontium and cesium are about 13.5 and 82, respectively. These Indian Point adsorption-retention factors are comparable to those for the LPGS case and those for Seabrook (Table 5.25), which is also located on fractured rock. The staff analysis also uses ConEd's most conservative value of the hydraulic conductivity (0.122 m/day). Indian Point's aquatic, shoreline, and total population doses are 1.1, 0.38, and 0.44 times the respective LPGS generic estuary doses based on this second analysis.

Table 5.27 compares earlier FESs with the LPGS generic site for estuaries. The Salem reactor adjoins Hope Creek, and its estimated population dose is expected to be similar. All other sites have smaller reactors, have smaller nearby populations, or are located farther from surface water. All but Maine Yankee and Indian Point are located on coastal plains or alluvial sediments having at least some clay minerals in them, and coastal plain sites have low groundwater velocities and relatively high adsorption-retention factors. None of these sites should exceed the LPGS population dose for estuaries.

5.3.3.4.7 Dry Sites

Table 5.28 compares current dry plant sites with the LPGS generic dry site. Only one site (Palo Verde) provides significant information on which a comparison could be based. Palo Verde is located in a desert valley where the groundwater gradient and velocity are expected to be low. Alluvium in the groundwater pathway should have adsorption-retention factors comparable to those of the LPGS (if not greater, as indicated in Table 5.28). In contrast, the LPGS generic site is on the Snake River plain above the Snake River Canyon. Fractured volcanic rocks and Pleistocene glacial and alluvial sediments underlie the LPGS generic site. Accordingly, the groundwater gradient and velocity at the LPGS site are extraordinarily high, and the groundwater travel time is low. Even without adsorption, strontium would require five times as long to reach the Palo Verde site boundary as in the LPGS generic case. Because of its location on the Snake River plain, the LPGS site's uninterdicted population dose is believed to represent the largest dose for dry sites. Therefore, all dry sites are expected to have significantly lower groundwater pathway population doses than those of the atmospheric pathway.

Table 5.29 compares the only earlier dry site (Rancho Seco) with the Palo Verde and LPGS generic dry sites. As seen in the table, the Rancho Seco and Palo Verde sites are strikingly similar. The only significant difference is that the Rancho Seco reactor is only three-fourths as large as those of Palo Verde. Therefore, a severe accident at Rancho Seco is expected to produce a population dose similar to or less than that at Palo Verde.

5.3.3.4.8 Results

Table 5.30 summarizes sites having uncertain groundwater pathway population doses compared with the LPGS study or other FES groundwater analyses. All but two of these sites are along small rivers. Uncertain groundwater pathways were the greatest concern.

Fractured rock, solution cavities in limestone, weathered rock, incompletely described geologic conditions, and the uncertain character of glacial or Pleistocene deposits are important geologic concerns. Several sites have large nearby populations, one is unusually close to surface water, and another is close to a stream with very low average flow rate.

The above liquid pathway analyses can be considered representative of uninterdicted population doses from a severe accident during the initial 40-year operating term. Liquid pathway population dose estimates at MYR would be smaller for a few plants, 10 to 30 percent higher for the majority of plants, and perhaps 50 percent higher for a few plants because of the general increase in population over a 50-year time interval beyond the FES analysis. Assuming such increases in population are representative of liquid dose increases, their effect on the results would be insignificant in relation to other uncertainties in the liquid pathway analysis.

However, it should be recognized that the uncertainty factor for liquid pathway uninterdicted population dose estimates in Tables 5.17 through 5.28 may be 10 or more. Codell (1985) does not recommend that these values be accepted at face-value; rather, they should be used for comparative purposes only (NUREG-1054). As stated previously, several parameters that are needed to perform a liquid pathway analysis (i.e., porosity, hydraulic conductivity, and adsorption coefficient) are not known with sufficient precision to provide better than order-of-magnitude estimates of population doses.

The LPGS and FES liquid pathway analyses (described above) provide uninterdicted population dose estimates based on the assumptions that core meltdown and penetration of the basemat have taken place. Such analyses are deterministic (i.e., they assume that the worst-case accident has occurred). However, the probability of occurrence of such an event is low (estimated to be no more than 10-4/RY). Contamination of groundwater is not likely to occur in the event of a core meltdown unless the basemat is penetrated. Therefore, the deterministic population doses given in Table 5.17 should be multiplied by a factor of about 10-4 to obtain the risk (probability estimates times consequences) of annual uninterdicted population doses for an 1150-MW reactor.

The population doses provided by these analyses are also based on the assumption that contaminated surface water and groundwater are not interdicted. Interdiction would lower the population doses significantly and could consist of preventing the contaminants from reaching the surface water, preventing use of the water, or making it difficult to obtain contaminated food. It is assumed, however, that interdicting the source of contamination once it enters the groundwater is not by itself sufficient because it may be impractical to completely isolate a contaminated aquifer from its surroundings. At best, containment measures such as grout curtains slow the groundwater movement but do not prevent it. However, the increased travel time reduces the rate of groundwater discharge to surface water bodies and reduces the concentration of radionuclides through prolonged radioactive decay. In any event, limiting people's contact with contamination through such measures as preventing or confiscating catches of recreational and commercial fish and shellfish, prohibiting water-based recreation, and eliminating surface water as a drinking-water source may have to be employed.

Table 5.30 Sites having uncertain groundwater pathway population doses with respect to the Liquid Pathway Generic Study and other final environmental statement analyses
Category site Major concern ~Downstream population x 103
Small river    
Arkansas Nuclear Weathered rock 2506
Arnold Pleistocene-holocene aquifer 5780
Bellefonte Fractured limestone 3243
Browns Ferry Limestone 3078
Connecticut Yankee Uncertain pathway < 10
Dresden Pleistocene aquifer 6037
La Salle Uncertain characteristics of glacial deposits 6012
McGuire Weathered rock 1683
Monticello Large nearby population 8690
North Anna Weathered rock < 10
Oconee Fractured rock 752
Peach Bottom Weathered rock < 10
Prairie Island Large nearby population 6302
Robinson Low stream flow 231
Sequoyah Weathered limestone 3681
Three Mile Island Surrounded by Holocene alluvium 20
Vermont Yankee Uncertain characteristics of glacial deposits 1724
Watts Bar Limestone 3681
Yankee Row Uncertain pathway 1724
Large river    
Trojan Fractured rock  
WNP-2a,b Fractured rock  

aWNP-2 = Washington Nuclear Project 2.

bThis site has an existing severe accident liquid pathway analysis. Analytical results for this site may be non-conservative.

Ocean and estuarine sites would be the hardest in which to effect interdiction because of the food pathway. Not only are the physical transport mechanisms of these systems complex, but many of the important recreational and commercial organisms are highly mobile. Thus, total confinement of the contamination would not be likely and controlling the taking of these organisms by man would need to be relied upon. However, it is reasonable to expect that dose reduction would occur as a result of interdiction of the pathways. It is estimated that the dose could be reduced by an order of magnitude (NUREG-0440, Table 7.3.2).

The risk to the population from releases to groundwater can be estimated by considering the information in Tables 5.17 through 5.29.

  • For large river sites, risk to the population can be estimated from the LPGS analyses as approximately 12 person-rem/RY (assuming the annual probability is 1 x  10-4 for a core melt with penetration of the basemat). For the large river site that has a larger population dose estimate than the LPGS (WNP-2) it is estimated that WNP-2 exceeds the LPGS by 50 percent (Table 5.20) for a risk of approximately 18 person-rem/RY. Pathway interdiction can reduce this dose by an order of magnitude; thus, the predicted annual population dose for large river sites is only a small fraction of that from the atmospheric pathway.
  • For small river sites, the risk to the population can be estimated from the LPGS analyses as approximately 1000 person-rem/RY with drinking-water risk contributing 890 person-rem/RY, ingestion contributing 70 person-rem/RY, and shoreline exposure contributing 40 person-rem/RY. Table 5.18 shows that the Byron Station FES-predicted population doses are higher than the LPGS small river site and would result in an annual population risk of approximately 3000 person-rem/RY at MYR. However, pathway interdiction could reduce this figure by a factor of 10, thus making the risk from groundwater releases only a small fraction of that from the atmospheric pathway for Byron Station. All other plants listed in Table 5.18 have much lower risk from groundwater releases than Byron Station.
    From Table 5.19, there may be as many as 19 small river sites that could exceed the Byron Station dose estimate. However, conservatively assuming that all of the radionuclides would reach the river and considering the potentially greater population that could be exposed, it is estimated that in several cases the Byron Station population doses could be exceeded by up to a factor of 10; but in most cases the population doses would be similar to or less than those of Byron Station. Accordingly, the risk from groundwater releases at small river sites is, in most cases, a small fraction of that from atmospheric releases and in several cases may be similar to that from atmospheric releases.
  • For Great Lakes sites, the risk to the population can be estimated from the LPGS analyses as approximately 350 person-rem/RY. However, the Fermi-2 FES analyses estimate a risk to the population of approximately 40 percent higher than the LPGS (Table 5.22), or approximately 500 person-rem/RY, uninterdicted. Pathway interdiction could reduce this by a factor of 10, thus making the annual population risk from groundwater releases only a small fraction of that from atmospheric releases. Since Section 5.3.3.4.4 concludes that the Fermi analysis provides the largest estimated groundwater pathway population dose of all Great Lake sites, the risk from groundwater releases at these sites is only a small fraction of that from atmospheric releases.
  • For estuarine sites, the LPGS analyses predict a high population risk without interdiction (17,700 person-rem/RY). Pathway interdiction could reduce this by a factor of 10. Section 5.3.3.4.6 indicates that the LPGS analyses provide the largest estimated population risk for all estuarine sites. Therefore, the risk from groundwater releases at estuarine sites is lower than or comparable to that from atmospheric releases.
  • For ocean sites, the risk to the population can be estimated from the LPGS analysis as approximately 55 person-rem/RY. From review of the Seabrook FES, it is estimated that the risk to the population may be as much as six times higher for Seabrook (Table 5.24), or approximately 330 person-rem/RY. Since pathway interdiction can reduce this by a factor of 10, it is a small fraction of the predicted risk from atmospheric releases from Seabrook. For other ocean sites, as discussed in Section 5.3.3.4.5, the Seabrook analysis provides a larger groundwater pathway population dose than all but Turkey Point. However, from the data in Table 5.25, assuming all the radionuclides from the reactor reach the groundwater, the population dose from Turkey Point at MYR would not be expected to exceed Seabrook (considering the differences in reactor size and surrounding population). Therefore, it can be concluded that the risk from groundwater releases at ocean sites would be a small fraction of that from atmospheric releases.
  • For dry sites, all predicted releases are orders of magnitude lower than the LPGS. From the LPGS (NUREG-0440, Table 6.2.21), the uninterdicted population risk from drinking water could be as high as 104 person-rem, which would be one person-rem/RY on an annual risk basis. This is much less than the risk from atmospheric releases.

5.3.3.4.9 Conclusion

Based on the above discussion, it is concluded that groundwater generally contributes only a small fraction of that risk attributable to the atmospheric pathway but in a few cases may contribute a comparable risk.

5.3.3.5 Economic Impacts

The purpose of this section is to determine if the economic costs of the severe accidents that have been estimated in the 27 FESs that contain severe accident analyses can be used to predict the future costs of such accidents at all sites. Similar to Section 5.3.3.2, the EI is used as a predictor of cost because the cost should be dependent upon the economic impact in the same way and for the same reason as population dose estimates are dependent.

CRAC was used to calculate off-site severe accident costs for the area contaminated by the accident. The off-site costs that were considered relate to avoidance of adverse health effects and are categorized as follows:

  • evacuation costs,
  • value of crops contaminated and condemned,
  • value of milk contaminated and condemned,
  • costs of decontamination of property where practical, and
  • indirect costs resulting from the loss of use of property and incomes derived therefrom (including interdiction to prevent human injury).

The severe accident analysis for the 27 FES plants uses these five cost category models to estimate an average (annual) expected cost due to a severe accident. These costs are a sum of the costs for a range of accidents multiplied by the probability that each of the accidents will occur. Costs in this section are stated in 1980 dollars to facilitate comparisons among plants. Key cost variables include projected population distributions, habitable land fraction, and statewide land-use statistics that identify land and crop values. The off-site consequence code then computes the off-site mitigation costs described above. For the FES plants that have severe accident analyses, estimated off-site accident costs could reach as high as $6 billion to $8 billion, but the probability of an accident with such high consequences would only be once in one million operating years. Higher costs are estimated for accidents with much lower probabilities. Projected costs of adverse health effects from deaths and illnesses would average about 10-20 percent of off-site mitigation costs. These costs are not considered in the economic cost calculations. One addition to these off-site costs was made in NRC risk analyses beginning in 1984. Recognizing that termination of economic activities in a contaminated area would create adverse economic impacts in wider regional markets and sources of supplies outside the contaminated area, NRC began estimating these additional economic costs in FESs. These costs are calculated only for a 1-year period after an accident and can reach into the billions of dollars.

Because some key variables affecting cost are strongly related to population density, it may be possible to predict mitigation costs for contaminated areas off-site using the EI developed in Section 5.3.3.2.1. To test this possibility, the expected cost of an accident calculated in 27 FESs having severe accident analyses was normalized for a plant size of 1000 MW(t) (Table 5.31) and then regressed against the EI value at 150 miles for that plant (Table 5.4).

Upper bound normalized expected costs of accidents during the MYR period for all plants were then predicted using this regression and the EI for populations for the MYR period. The estimates were then nonnormalized to convert to expected costs (MYR).

MYR 95% projected cost/RYb (1994 dollars)
Table 5.31 Average expected costs during the current license period and predicted expected costs during the middle year of license renewal (MYR) resulting from a severe accident
Plant Average expected cost/RYa
(dollars)
10-mile MYR population
Arkansas -- 33,992 477,750
Beaver Valley 29,000 155,141 1,565,550
Bellefonte -- 35,846 2,278,500
Big Rock Point -- 11,037 73,500
Braidwood 14,000 32,652 6,357,750
Browns Ferry -- 36,400 1,984,500
Brunswick -- 15,348 992,250
Byron 8,400 23,900 4,226,250
Callaway 4,300 6,877 1,528,800
Calvert Cliffs -- 24,564 4,336,500
Catawba 7,100 130,735 1,764,000
Clinton 6,700 16,543 3,344,250
Comanche Peak 3,900 19,400 882,000
Cooper -- 6,768 2,116,800
Crystal River -- 20,368 1,249,500
D.C. Cook -- 63,680 2,094,750
Davis-Besse -- 19,714 3,013,500
Diablo Canyon -- 29,591 661,500
Dresden -- 48,248 2,609,250
Duane Arnold -- 94,461 463,050
Farley -- 16,421 624,750
Fermi 2 23,000 93,010 2,138,850
FitzPatrick -- 34,403 1,029,000
Fort Calhoun -- 17,978 220,500
Ginna -- 39,649 404,250
Grand Gulf 3,060 10,943 1,984,500
Haddam Neck -- 91,760 2,249,100
Hatch -- 6,607 1,881,600
Hope Creek 40,000 32,844 4,704,000
Indian Point -- 247,253 8,246,700
Kewanee -- 12,966 551,250
La Salle -- 20,204 3,785,250
Limerick 62,200 178,626 3,505,950
Maine Yankee -- 41,435 771,750
McGuire -- 72,117 1,697,850
Millstone 3 80,000 130,000 3,461,850
Monticello -- 28,091 992,250
Nine Mile Point 8,000 35,208 1,396,500
North Anna -- 11,668 2,352,000
Oconee -- 77,790 1,234,800
Oyster Creek -- 96,364 1,675,800
Palisades -- 39,720 2,572,500
Palo Verde 2,260 1,378 1,176,000
Peach Bottom -- 34,894 3,858,750
Perry 7,300 89,247 2,028,600
Pilgrim -- 45,921 1,176,000
Point Beach -- 26,447 588,000
Prairie Island -- 28,450 441,000
Quad Cities -- 42,521 2,131,500
Rancho Seco -- 12,489 2,646,000
River Bend 50,000 33,120 1,580,250
Robinson -- 37,681 1,543,500
Salem -- 32,868 8,636,250
San Onofre 19,000 91,940 2,734,200
Seabrook 5,800 130,574 882,000
Sequoyah -- 66,110 1,433,250
Shearon Harris 3,770 26,423 1,690,500
Shoreham -- 113,644 2,138,850
South Texas 2,600 4,149 2,998,800
St. Lucie 4,250 166,860 1,058,400
Summer 4,800 14,997 2,205,000
Surry -- 103,830 1,146,600
Susquehanna 9,000 54,887 3,153,150
Three Mile Island -- 170,142 3,748,500
Trojan -- 21,958 3,050,250
Turkey Point -- 11,136 551,250
Vermont Yankee -- 2,354 2,763,600
Vogtle 16,000 2,648 2,763,600
Waterford 4,500 1,930 3,998,400
Watts Bar -- 95,237 573,300
WNP-2c 2,600 22,878 918,750
Wolf Creek 3,600 7,239 1,411,200
Yankee Row -- 27,263 1,249,500
Zion -- 293,491 2,138,850

aRY = reactor year; estimates presented in the final environmental statements for operation license.
bDistribution free values (nonparametric--see Appendix G). Includes MELCOR Accident Consequence Code System-implied correction factors as well as an inflation multiplier derived from the icit Gross Domestic Price Inflator Index = 125.9/85.7 = 1.47 (from 1980 to 2nd quarter 1994).
cWNP-2 = Washington Nuclear Project 2.

Note: 10 miles = 16 km.

Economic consequences were also benchmarked to the MACCS computer code to ensure the calculated values were based on the most current models and data. The benchmark computations indicated that the CRAC calculations used to estimate the economic impacts for the FES plants did not have a continuous linear relationship with population. The MACCS code predicted higher costs than did the CRAC code; low population sites were underpredicted by substantial margins. The differences were primarily due to the difference in the handling of decontamination costs in the two codes. Results from Tingle (1993) indicate that in order to be comparable to results calculated from MACCS, the regression values should be adjusted though the use of population-dependent correction factors. Table 5.31 reflects average expected cost values that were derived from the regression and then corrected with the following factors:

  • Sites with MYR 10-mile (16-km) populations £ 10,000 multiply cost data by 40.
  • Sites with MYR 10-mile populations > 10,000 and £ 50,000 multiply cost data by 25.
  • Sites with MYR 10-mile populations > 50,000 multiply cost data by 15.

Also, the FES values were in 1980 dollars. To correct for this the average expected cost values were adjusted to 1994 dollars.

In addition to assessing the economic impact of severe accidents, six of the 27 FESs that analyze severe accidents also assess the amount of off-site land that could be contaminated and subject to long-term interdiction as a result of a severe accident.

These plants and their predicted conditional mean values of land contamination are listed below:

  • Hope Creek
7000 m2/year
(8400 yd2/year)
  • Limerick 1 and 2
1500 m2/year
(1800 yd2/year)
  • Millstone 3
4000 m2/year
(4800 yd2/year)
  • Nine Mile Point 2
20,000 m2/year
(24,000 yd2/year)
  • River Bend
40,000 m2/year
(48,000 yd2/year)
  • South Texas 1 and 2
600 m2/year
(720 yd2/year)

These predicted values would not be expected to change for the license renewal period since they are not affected by increases in population.

As can be seen by the values listed above, the predicted conditional land contamination is small (10 acres/year at most). This is also consistent with WASH-1400 (NUREG-75/014) and a 1982 study on siting criteria (NUREG/CR-2239) which predicts small conditional land contamination values. The land contamination values for these six plants can be considered representative of all plants since they cover the major vendor and containment types and include sites at the upper end of annual rainfall. However, even considering that land contamination values can vary at other sites, it is not expected that predicted land contamination from plants at other sites would vary more than 1 or 2 orders of magnitude from the values listed above and would, therefore, still be a small impact.

5.3.4 Uncertainties

FESs referred to in this section have been based mostly upon the methodology presented in RSS, which was published in 1975 (NUREG-75/014).

Although substantial improvements have been made in various facets of the RSS methodology since its publication, large uncertainties in the results of these analyses remain, including uncertainties associated with the likelihood of the accident sequences and containment failure modes leading to the release categories, the source terms for the release categories, and the estimates of environmental consequences. A comprehensive discussion of the uncertainties associated with risk assessments is provided in NUREG-1150. The relatively more important contributors to uncertainties in the results presented in this environmental statement are as follows.

5.3.4.1 Probability of Occurrence of Accident

If the probability of a release category were to change by some percentage, the probabilities of various types of consequences from that release category would also change by the same percentage. Thus, an order of magnitude uncertainty in the probability of a release category would result in a corresponding order of magnitude uncertainty in both societal and individual risks stemming from the release category. In RSS, there are substantial uncertainties in the probabilities of the release categories. This uncertainty is due, in part, to difficulties associated with the quantification of human error and to limitations in the database on failure rates of individual plant components and in the database on external events and their effects on plant systems, structures, and components that are used to calculate the probabilities. However, since the publication of RSS, substantial NRC programs to improve nuclear plant safety have been implemented such as resolution of generic safety issues (NUREG-0933), Station Blackout and Anticipated Transient Without Scram Rulemakings, and improvements resulting from reviews of the TMI accident (NUREG-0737). These programs, as well as others, all served to reduce the average risk of the overall nuclear industry such that in this GEIS, the use of RSS risk values and their associated frequencies of an accident (because they are embodied within the risk calculation) are reasonable upper estimates of risk for the industry. This is true for even those plants that have not had the benefit of a PRA analysis.

5.3.4.2 Quantity and Chemical Form of Radioactivity Released

There are also significant uncertainties associated with the timing, quantity, and chemical form of each radionuclide species that would be released from a reactor unit during a particular accident sequence. Radioactive material originates in the fuel and would be released from any damaged fuel during an accident. Some would be attenuated by physical and chemical processes en route to being released to the environment. Depending on the accident sequence, such factors as attenuation in the reactor vessel, the rest of the cooling system, the containment, and adjacent buildings would influence both the magnitude and chemical form of radioactive releases. Additional radionuclide releases may originate from on-site dry cask storage facilities for those sites which develop the capability, although the radionuclide inventory is much less than that in the reactor core. Information available in NUREG-0956, in NUREG-1150, and from the latest research activities sponsored by NRC and the industry indicates that the uncertainty in radionuclide source terms is large and represents a significant contribution to the uncertainty in the absolute value of risk. In comparison with the RSS source terms (which are used in the FES analyses), source terms in recent studies were in some instances higher and in other instances lower. However, for the early containment failure sequences, which have the greatest impact on risk, the RSS source terms appear to be larger than the mean values estimated from the recent work and are typically at the upper bound of the uncertainty range of estimates for NUREG-1150.

5.3.4.3 Atmospheric Dispersion Modeling for the Radioactive Plume Transport

Uncertainties are involved in modeling the atmospheric transport of radioactivity in gaseous and particulate states and the actual transport, diffusion, and deposition or fallout that would occur during an accident (including the effects of condensation and precipitation). The phenomenon of plume rise from heat associated with the atmospheric release, effects of precipitation on the plume, and fallout of particulate matter from the plume all have considerable impact on the magnitudes of early health consequences along with the distances from the reactors where these consequences would occur. These factors can result in overestimates or underestimates of both early and later effects (health and economic).

Other areas that have effects on uncertainty are as follows:

  • Duration, energy release, and in-plant radionuclide decay time. These areas relate to the differences between assumed release duration, energy of release, and the in-plant radioactivity decay times compared with those that would actually occur during a real accident.
    For an atmospheric release of relatively long duration (greater than a half-hour), the actual cross-wind spread (i.e., the width) of the radioactive plume would likely be larger than the width calculated by the dispersion model in the staff code (CRAC). However, the effective width of the plume is calculated in the code using a plume expansion factor that is determined by the release duration. For a given quantity of radionuclides in a release, the plume and, therefore, the area that would come under its cover would become wider if the release duration were longer. In effect, this would result in lower air and ground concentrations of radioactivity but a greater area of contamination.
    The thermal energy associated with the release affects the plume rise phenomenon; a plume that rises quickly or to a high altitude (as in the Chernobyl accident) results in relatively lower air and ground concentrations in the closer-in regions and relatively higher concentrations in the farther-out regions (because of fallout) than would be predicted for plumes that do not rise. Therefore, if large thermal energy were associated with a release containing a large fraction of core-inventory radionuclides, it could increase the distance from the reactor over which early health effects may occur. If, on the other hand, the release behavior were dominated by the presence of large amounts of condensing steam, very much the reverse could occur because of close-in deposition of radionuclides induced by the falling water condensed from the steam.
    The time from reactor shutdown until the beginning of the release to the environment (atmosphere), known as the time of release, is used to calculate the depletion of radionuclides by radioactive decay within the plant before release. The depletion factor for each radionuclide (determined by the radioactive decay constant and the time of release) multiplied by the release fraction of the radionuclide and its core inventory determines the actual quantity of the radionuclide released to the environment. Later releases would result in the release of fewer curies to the environment for given values of release fractions.
    These parameters can all have significant impacts on accident consequences, particularly early consequences.
  • Meteorological sampling scheme used. There is a possibility that the meteorological sequences used with the selected start times (sampling) in CRAC may not adequately represent all meteorological variations during the year, or that the year of meteorological data may not represent all possible conditions. This factor is judged to produce greater uncertainties for early effects and less for latent effects.
  • Emergency response effectiveness and warning time. This relates to the differences between modeling assumptions regarding the emergency response of the people residing near nuclear facilities compared with what would happen during an actual severe reactor accident. Included in these considerations are such subjects as evacuation effectiveness under different circumstances, possible sheltering and its effectiveness, the effectiveness of population relocation, and the fraction of people assumed not to relocate. The warning time is the interval between the time the plant operating staff recognize plant conditions which would indicate that protective actions should be taken for the general population and the time of the release of radioactive material from the plant. In calculations with CRAC, it is assumed that the protective action taken would always be evacuation. Therefore, in the calculation, the evacuating public could be caught by a radioactive plume and exposed or could evacuate into a passing plume. In reality, there are other protective actions that might be called for by public officials--for instance, sheltering to avoid such a situation. This can affect the simplified assumptions about protective actions in the calculated results and would most likely be in the direction of larger calculated early effects. Longer warning times are always more favorable in reality because they would allow time for consideration of several protective action options. The uncertainties associated with emergency response effectiveness and warning time could cause large uncertainties in early health consequences. The uncertainties in latent health consequences and costs are considered smaller than those for early health consequences.
  • Dose-conversion factors and dose-response relationships for early health consequences. There are uncertainties associated with the conversion of contamination levels to doses, relationships of doses to health effects, and considerations of the availability of what was described in RSS as supportive medical treatment (a specialized medical treatment program, of limited availability in the local area but with additional availability outside the area, that would minimize the early health effect consequences of high levels of radiation exposure following a severe reactor accident). Although all health impacts have not been enumerated in this evaluation, the primary ones have been, and references to other documents such as RSS provide additional insights into the subject.
  • Dose-conversion factors and dose-response relationships for latent health consequences. Estimates of dose and latent (delayed and long-term) health effects on individuals and on their succeeding generations involve uncertainties associated with conversion of contamination levels to doses and of doses to health effects. The staff judges that this category has a large uncertainty. The uncertainty could result in relatively small underestimates of consequences, but also in substantial overestimates of consequences. Previous FES analyses have been based on results that utilized dose-response relationships provided in BEIR-III (or earlier reports). Consequently the results presented in this GEIS have been corrected to account for the more recent dose-response relationships provided in BEIR-V and to reflect models and relationships found in the most current consequence assessment codes.
  • Chronic exposure pathways. Uncertainty arises from the possibility that different protective action guide levels may be used for interdiction or decontamination of the exposure pathways (both the atmospheric pathway and the groundwater pathway) than those assumed in the staff analysis. Furthermore, uncertainty arises because there is a lack of precise knowledge about the fate of the radionuclides in the environment as influenced by natural processes such as runoff and weathering. The staff's qualitative judgment is that the uncertainty from these considerations is substantial.
  • Economic data and modeling. This relates to uncertainties in the economic parameters and economic modeling such as costs of evacuation, relocation, medical treatment, and decontamination of properties and other costs of property damage. Uncertainty in this area could be substantial.

NUREG-1150 contains a state-of-the-art quantification of the uncertainties in core-melt frequency, containment behavior, and source term evaluation. Also included are discussions of the major factors affecting the uncertainty. For further detail on the topics discussed in Sections 5.3.4.1 through 5.3.4.3, refer to the appropriate topics in NUREG-1150.

5.3.4.4 Assumption of Normality for Random Error Components

The predictions of risk values (early and latent fatalities and total dose) were developed statistically by regressing consequence values calculated in recent nuclear plant FESs. A "standard" assumption in the calculation of confidence bounds for these predictions is that the regression errors have a normal distribution. However, without specific evidence of normality, normal-theory confidence bounds for the risk may be too high or low, possibly by a significant margin. Therefore, alternative confidence bounds were considered, which do not rely on the errors having a specified distribution such as the normal, but depend instead on a large-sample approximation. When the normal-theory and alternative bounds differed, the ones leading to higher calculated values were used. (This subject is discussed in Appendix G.)

5.3.4.5 Exposure Index

The concept of using a parameter such as EI to predict future risks is also subject to uncertainty. Such issues are discussed below.

  • Selection of EI parameters. EI is a calculated parameter based on plant-specific information: population surrounding the plant and wind direction frequency data for the plant. The data on population projections used in the calculation of EI values are based on the 1980 census. EI estimates were made for years 1990, 2000, 2010, 2030, and 2050 and for populations at 10 and 150 miles from the plant. Population estimates for these years were obtained from data provided by the Bureau of Economic Analysis. It is estimated that the uncertainty in these population projections is relatively small, certainly less than a factor of two, and, consequently, would not significantly impact the conclusions of this evaluation.
    The wind data were obtained from plant license documentation such as environmental reports or final safety analysis reports; site-specific data are used in this analysis.
    However, other parameters such as exclusion area distance, rainfall, evacuation speed, and terrain can also affect the consequence calculations. The NUREG-1150 study found that for the five plants studied, the fatality magnitudes (early and latent) were driven primarily by the core-damage frequency, the source term releases, site meteorology, population distribution, and the effectiveness of emergency response measures. All these factors were considered in the CRAC analyses done for the FES plants, using site-specific information for meteorology, population, and emergency response actions. The FES plant analyses enveloped a broad range of such site-specific values. Consequently, it is likely that the use of the UCB limit to estimate future environmental impacts would envelop the effects of these parameters for all plants. The FES analyses were usually performed assuming populations representative of the middle year of the normal 40-year license period. Populations would continue to increase as operation continued into a renewal period. Thus, renewal period risks were predicted using population representative of the middle year of the renewal period. Wind direction frequency is very plant-specific and was not considered to be adequately enveloped for the non-FES plants by the FES plant's wind direction frequencies, especially when these frequencies are weighted by the plant-specific population. However, by selection of population and wind direction frequency for the EI and using UCB values to envelop the effects of other parameters, the uncertainty introduced by the selection of EI parameters should be minimized.
  • Selection of distances. Although the selection of 10 miles and 150 miles for computing EI values produces rather strong correlations between the EI values and the reported effects in FESs (Appendix G), other distances could exist whose selection would ult in stronger correlations. Indeed, as shown in Table 5.5, the FES plants showed a range of 7 to 50 miles for occurrence of total acute fatalities whereas the GEIS analysis used only one distance, 10 miles. However, the effect of stronger correlations would serve primarily to reduce the uncertainty of the regression, thus resulting in a general reduction in the UCB values. Consequently, because the correlations that are used in the study are relatively strong and GEIS uses the UCB values to estimate risk, the possibility of under prediction should be small.
  • Regressing early fatalities for only large plants. As described in Section 5.3.3.2.1, the regressions for early fatality estimates were performed using data for FES plants having thermal power levels greater than about 3025 MW(t). Although there is some relationship between plant size and predicted early fatalities (all other factors being held constant), the relationship is not linear because of the threshold effects for early fatalities. Therefore, normalization for plant size for the early fatality regression process was not considered appropriate; rather, early fatalities were predicted based only on the data for large plants. This approach should generally provide overpredictions for plants less than 3025 MW(t) resulting in most of the uncertainty being in the direction of smaller predicted effects. For plants equal to or greater than 3025 MW(t), small uncertainty in the calculated values may be present. However, the use of UCBs for predicting risk values should minimize the possibility of underprediction.
  • Normalization of plants for latent fatalities, costs, and dose. As described in Section 5.3.3.2.1, the regression for latent fatality and dose curves were performed using FES data that had been normalized to 1000 MW(t) in order to reduce the influence of plant size on the fitted parameters. Actual plant size was used for making the predictions and, therefore, the final results reflect nonnormalized values. The regression of latent fatalities, dose, and costs using normalized FES values assumes a linear relationship between power level and source term released. The use of UCBs to predict risk values should minimize the possibility of underprediction.

5.3.4.6 Summary

The state of the art for the quantitative evaluation of the uncertainties for PRA analyses is presented in the NUREG-1150 studies. The NUREG-1150 results indicate that reduction of uncertainty considerations or previously unanalyzed phenomena and sequences and consideration of plant changes have resulted in individual risk components that are both higher and lower than originally provided in RSS. However, NUREG-1150 shows that the cumulative effect is a reduction in risk for those plants studied, and it is also likely to be the case for the industry as a whole. The GEIS results, when reviewed against current data and methodology, have large uncertainties associated with them. The bounds on this uncertainty could be between a factor of 10 and 1000 and could result in the values used being higher or lower.


5.4 Severe Accident Mitigation Design Alternatives (SAMDAs)

[ Prev | Next | Table of Contents ]

In 1980 NRC issued an interim policy statement on the consideration of severe accidents in environmental impact statements (EISs) (45 FR 40101) applicable to Construction Permit and Operating License applications submitted on or after July 1, 1980. That policy statement states that it is "the intent of the Commission that the staff take steps to identify additional cases that might warrant early consideration of either additional features or other actions which would prevent or mitigate the consequences of serious accidents." Recently, these features have become commonly referred to as SAMDAs. The policy statement goes on to say, "cases for such consideration are those for which a Final Environmental Statement has already been issued at the Construction Permit stage but for which the Operating License review stage has not yet been reached." This statement was made in recognition of the fact that changes in plant design features may be more easily incorporated in plants when construction has not yet progressed very far.

In August 1985, NRC issued its policy statement on severe reactor accidents. That policy statement presented NRC's conclusion that existing plants pose no undue risk to public health and safety and that there was no present basis for immediate action on generic rulemaking or other regulatory changes for those plants because of severe accident risk. Nevertheless, it called for each licensee to perform an analysis designed to discover instances of particular vulnerability to core melt or unusually poor containment performance given a core-melt accident. NRC believed that this policy statement was a sufficient basis for not requiring a consideration of SAMDAs at the operating license review stage for previously constructed plants. However, a 1989 court decision ruled that such a policy statement was not sufficient to preclude a consideration of SAMDAs and that such a consideration is required for plant operation, Limerick Ecology Action v. NRC, 869 F.d 719 (3rd Cir. 1989). In order to assess whether SAMDAs can be adequately addressed generically for all plants in this GEIS, it is necessary to consider the level of experience the commission has regarding SAMDAs and the extent to which this experience can reasonably address the SAMDA issue for all plants.

5.4.1 Commission Experience Regarding Severe Accident Mitigation

NRC has gained considerable experience regarding severe accident mitigation during the past several years through implementation of its severe accident policy statement. Specific major actions that have been initiated and, in some cases, completed are (1) evaluation of containment performance and various alternatives for improvement, (2) initiation of individual plant examination, and (3) initiation of an accident management program. Additionally, NRC has performed three site-specific evaluations of SAMDAs pursuant to the 1989 court decision. These SAMDA analyses were included in the final environmental impact statements for Limerick 1 and 2 and Comanche Peak 1 and 2 operating license reviews, and the Watts Bar supplemental final environmental statement for operation. These actions are addressed below.

5.4.1.1 Containment Performance

NRC has examined each of five U.S. reactor containment types (BWR Mark I, II and III; PWR Ice Condenser; and PWR Dry) with the purpose of examining the potential failure modes, potential fixes, and the cost benefit of such fixes. This examination has been called the containment performance improvement (CPI) program and has been documented in a series of reports (NUREG/CR-5225; NUREG/CR-5278; NUREG/CR-5528; NUREG/ CR-5529; NUREG/CR-5565; NUREG/CR-5567; NUREG/CR-5575; NUREG/CR-5586; NUREG/CR-5589; NUREG/CR-5602; NUREG/CR-5623; NUREG/ CR-5630). Tables 5.32 through 5.34 summarize the results of this program. As can be seen from these tables, many potential changes were evaluated but only a few containment improvements were identified for site-specific review. The items evaluated in the CPI program were also included in the list of plant-specific SAMDAs examined in the Limerick, Comanche Peak, and Watts Bar FES supplements, discussed later.

5.4.1.2 Individual Plant Examinations

In accordance with NRC's policy statement on severe accidents, each licensee has been requested to perform an individual plant examination (IPE) to look for vulnerabilities to both internal and external initiating events (Generic Letter 88-20, Supplements 1-4). This examination will consider potential improvements on a plant-specific basis. In effect, IPE could be considered equivalent to a monitoring program that looks at the severe accident performance of each licensed plant. Detailed guidance has been issued to each licensee regarding the scope and conduct of IPE and the reporting requirements. NRC staff intends to review each submittal and, if plant modifications not proposed by the licensee appear warranted, to pursue the incorporation of such modifications via NRC's backfit rule (10 CFR Part 50.109). To date, 22 IPEs have been reviewed by NRC. These IPEs have resulted in plant procedural and programmatic improvements (i.e., accident management) and, in only a few cases, minor plant modifications, to further reduce the risk and consequences of severe accidents.

5.4.1.3 Accident Management

Accident management involves the development of procedures that promote the most effective use of available plant equipment and staff in the event of an accident. NRC has indicated its intent (Generic Letter 88-20, Supplement 2) to request that licensees develop an accident management framework that will include implementation of accident management procedures, training, and technical guidance. It is expected that insights gained as a result of IPE will be factored into the accident management program. As discussed earlier, the majority of improvements identified from the completed IPEs to date have been in the area of accident management or other procedural and programmatic improvements.

5.4.1.4 SAMDA Analyses

Table 5.32 Potential boiling-water reactor containment improvements considered in the containment performance improvement program
Number Potential improvement Resolution Comments
1 Enhanced ADS, low pressure water supply, and backup power Include in IPE a
2 Hardened vent Implemented for Mark-Is, included in IPE for Mark-II and IIIs b
3 ATWS sized-hardened vent Drop c
4 External filter Drop c
5 Dedicated suppression pool cooling Drop c
6 Alternate decay heat removal Drop c
7 Core debris control Drop c
8 Enhanced drywell spray Drop c
9 Drywell head flood Drop c
10 Enhanced reactor building DF Drop  
11 Backup power for hydrogen ignitors (Mark IIIs) Included in IPE d

Acronyms: ADS = automatic depressurization system, IPE = individual plant examination, ATWS = anticipated transit without scram, DF = decontamination factor.

aAnalysis showed that potential improvement may be cost beneficial.
bCost beneficial for Mark-Is.
cNot cost effective--potential improvement will be too expensive with too little benefit.
dMay be cost beneficial.

Table  5.33 Potential pressurized-water reactor ice condenser improvements considered in the containment performance improvement program
Potential improvement Resolution Comments
Reactor cavity flooding Drop Not cost beneficial. Might cause ex-vessel steam explosion.
Backup water to the containment spray system Drop Not cost beneficial
Backup power to the air return fan system Drop Not cost beneficial. May increase containment pressurization
Reactor depressurization Include in accident management Currently being pursued as a viable accident management strategy
Improved hydrogen ignitor system (backup power) Include in individual plant examination (IPE) Most cost beneficial of all alternatives considered (although it still does not meet the backfit test). To be looked at within the IPE program
Containment inerting Drop Not cost beneficial, may reduce accessibility for maintenance
Filtered vent Drop Not cost beneficial
Ex-vessel core debris curb Drop Large uncertainty as to effectiveness
Steam generator tube rupture improvements--increased testing Further research needed Being examined in separate Nuclear Regulatory Commission program by the Materials Engineering Branch, RES
Containment bypass improvements Included in generic issues program Being examined as part of a separate interfacing system loss of coolant accident generic issue (GSI 105)
Table 5.34 Potential pressurized-water reactor (PWR) large, dry containment improvements considered in the containment performance improvement program
Potential improvement Resolution Comments
Operator depressurization using power-operated relief valve Drop No conclusive findings on its benefit to risk reduction
Addition of a cavity flooding system Drop Not cost beneficial. The effect of a flooded cavity on the direct containment heating threats may be beneficial or detrimental, depending on each plant
Addition of hydrogen control system Assess in individual plant examination (IPE) Recommend all dry PWR containments assess the likelihood of local hydrogen detonation in the IPE

Site specific SAMDA analyses were performed for Limerick, Comanche Peak, and Watts Bar. A listing of the specific SAMDAs reviewed for applicability to Limerick is provided in Table 5.35. The staff examined each SAMDA (individually and, in some cases, in combination) to determine its individual risk reduction potential. This risk reduction was then compared with the cost of implementing the SAMDA to provide cost-benefit evidence of its value. Considering that the estimates of risk at Limerick used by the staff in these evaluations were considered to be high and that the uncertainties associated with the costs, effectiveness, and/or operational disadvantages of some SAMDAs were large, the staff concluded that there was no clear evidence that modifications to Limerick were justified for the purpose of further mitigating severe accident risks.

The staff made a similar assessment of SAMDAs for the Comanche Peak Steam Electric Station. A list of the SAMDAs reviewed in this evaluation is provided in Table 5.36. As with the Limerick evaluation, the staff had no basis for concluding that modifications to Comanche Peak were justified for the purpose of further mitigating environmental concerns as they relate to severe accidents. Recently, the staff evaluated SAMDAs for the Watts Bar Nuclear Plant. As in the Limerick and Comanche Peak analyses, no plant modifications were justified for the purpose of further mitigating severe accident risk and consequences.

Several important items from these analyses should be noted.

  • First, the SAMDAs considered at Limerick, Comanche Peak, and Watts Bar covered a broad range of accident prevention and mitigation features. These features included the items that were evaluated for all containment types as part of the CPI Program.
  • Second, the Limerick analyses were for a plant at a high population site. Since risk to the public is generally proportional to the population surrounding the plant, one would generally expect SAMDAs for plants at high population sites to have the most favorable cost-benefit ratio. Since SAMDAs were found not to be justified at Limerick, it is unlikely that they would be justified for plants at other sites.
  • Third, plant procedural and programmatic improvements (rather than plant modifications) were the only cost-beneficial improvements identified from these analyses.

Table  5.35 Severe accident mitigation design alternatives (SAMDAs) considered for the Limerick Generating Station

1. Installation of alternative means to maintain suppression pool subcooling to improve plant's capability to remove decay heat and prevent containment overpressure challenge

2. Provision of an alternative means of decay heat removal

3a. Installation of containment vent of sufficient size to prevent containment overpressure due to an anticipated transient without scram event

3b. Installation of containment vent and filter of sufficient size to prevent containment overpressure due to an inability to remove decay heat

3c. Installation of containment vent (no filter) of sufficient size to prevent containment overpressure due to an inability to remove decay heata

4. Installation of core debris control devices to prevent core/concrete interaction and remove decay heat from the core debris

5a. Provide enhanced drywell spray capability to increase the reliability for removal of heat from the drywell atmosphere and the core debris, thereby minimizing the threat of containment failure due to overpressure

5b. Provide modification for flooding of the drywell head to help mitigate accidents that result in leakage through the drywell head seal

6. Provide the capability for diesel-driven, low-pressure makeup to the reactor to help in mitigation of core damage resulting from accident sequences in which the reactor vessel is depressurized and all other means of injecting water to the vessel have been lost

7. Improve the reliability of the automatic depressurization system to reduce the probability of vessel failure at high pressure during a severe accident

8. Establish an improved decontamination factor for secondary containment through enhancement to the fire protection system and/or the standby gas treatment system hardware and procedures to improve fission product removal

aThis SAMDA has been implemented for plants having Mark I containments.

Table 5.36 Listing of severe accident mitigation design alternatives considered for the Comanche Peak Steam Electric Station

  1. Additional Instrumentation for Bypass Sequences: Install pressure-monitoring or leak-monitoring instruments (permanent pressure sensors) between the first two pressure isolation valves on low-pressure injection lines, residual heat removal (RHR) suction lines, and high-pressure injection lines. The additional instrumentation would improve the ability to detect valve leakage or open valves, and would decrease the frequency of interfacing system loss-of-coolant accidents (LOCAs).
  2. Deliberate Ignition System: Provide a system to promote ignition of combustible gases (hydrogen and carbon monoxide) at low concentrations. The ignition system would prevent large-scale deflagrations or detonations in events involving gradual releases of combustibles (such as from cladding oxidation or core-concrete interactions) but may be ineffective for rapid releases of hydrogen that could occur coincident with reactor vessel failure at high pressure.
  3. Reactor Coolant System Depressurization: Provide a capability to rapidly depressurize the reactor coolant system. Reactor depressurization would allow injection using low-pressure systems and would reduce the threat of direct containment heating and induced failures of steam generator tubes and primary coolant piping in the event low-pressure injection systems are not available. Depressurization could be achieved by a system specially designed to manually depressurize the reactor vessel or by actuation of existing pressurizer power-operated relief valves, reactor vessel heat vent valves, and secondary system valves.
  4. Independent Containment Spray System: Provide an independent containment spray system, using the existing spray headers if appropriate. The spray system would cool the containment and the core debris, thereby reducing the challenge to containment from overtemperature and long-term overpressure by steam. However, unless the sprays terminate core-concrete interactions, the noncondensable gases released from the concrete are expected to cause the containment to eventually fail by overpressure.
  5. Reactor Cavity Flooding System: Provide a capability to flood the reactor cavity before and after reactor vessel breach. Cavity flooding would promote debris coolability, reduce core-concrete interactions and noncondensable gas production, and provide fission product scrubbing.
  6. Filtered Containment Venting: Provide a capability to vent the containment through a vent path routed to an external filter. The filtered vent would mitigate challenges to containment from long-term overpressure and hydrogen burn (by reducing the baseline containment pressure) but may not be effective for mitigating energetic events such as hydrogen burns coincident with reactor vessel failure.
  7. Additional Diesel Generator: Provide an additional diesel generator with cross-ties to both Class 1E buses. This modification would increase the availability of the AC power system and reduce the frequency of station blackout sequences.
  8. Additional DC Battery Capability: Provide additional DC battery capability to ensure eight hours of instrumentation and control power, as opposed to four in the event of a station blackout. This would extend the time available for recovery and reduce the frequency of long-term station blackout sequences.
  9. Alternative Means of Core Injection: Provide a capability for makeup water to the reactor using a low-pressure, diesel-driven pump of sufficient capacity and associated piping hardware and procedures. The diesel-driven pump would serve as a backup to the front-line, low-pressure injection systems and could also be used to maintain core cooling in the event of a LOCA.
  10. Improved Availability of Recirculation Mode: Provide a system to automatically switch the suction of the safety injection and centrifugal charging pumps to the RHR pump discharge when the refueling water storage tank is depleted. Automatic switchover would reduce the potential for operator error and improve the availability of core cooling in the recirculation mode.
  11. Additional Service Water Pump: Add a third 100 percent service water pump to improve the availability of the station service water system. This would reduce the frequency of sequences involving failure of vital plant equipment due to loss of cooling.

5.4.1.5 Conclusion

Although NRC has gained considerable experience regarding severe accident mitigation improvements, the ongoing regulatory programs related to severe accident mitigation (i.e., individual plant examination/individual plant examination of external events and Accident Management) have not been completed for all plants. Since these programs have identified plant programmatic and procedural improvements (and in a few cases, minor plant modification) as cost effective in reducing severe accident consequence and risk, it would be premature to generically conclude that a consideration of severe accident mitigation is not required for license renewal.

However, based on the experiences discussed above, the NRC expects that a site-specific consideration of severe accident mitigation for license renewal will only identify procedural and programmatic improvements (and perhaps minor hardware changes) as being cost-beneficial in reducing severe accident risk or consequence. Therefore, a site-specific consideration of alternatives to mitigate severe accidents shall be performed for license renewal unless such a consideration has already been included in a previous EIS or related supplement. Staff evaluations of alternatives to mitigate severe accidents have already been completed and included in an EIS or supplement for Limerick, Comanche Peak, and Watts Bar; therefore, severe accident mitigation need not be reassessed for these plants for license renewal.


5.5 Summary and Conclusions

[ Prev | Next | Table of Contents ]

The foregoing discussions have dealt with the environmental impacts of accidents during operation after license renewal. The primary assumption for this evaluation is that the frequency (or likelihood of occurrence) of an accident at a given plant would not increase during the plant lifetime (inclusive of the license renewal period) because regulatory controls ensure the plant's licensing basis is maintained and improved, where warranted. However, it was recognized that the changing environment around the plant is not subject to regulatory controls and introduces the potential for changing risk. Estimation of future severe accident consequences and risk was based upon existing risk and consequence analyses found in FES for recently licensed plants because these include severe accident analyses and constitute a representative set of plants and sites for the United States.

5.5.1 Impacts from Design-Basis Accidents

The environmental impacts of postulated accidents were evaluated for the license renewal period in GEIS Chapter 5. All plants have had a previous evaluation of the environmental impacts of design-basis accidents. In addition, the licensee will be required to maintain acceptable design and performance criteria throughout the renewal period. Therefore, the calculated releases from design-basis accidents would not be expected to change. Since the consequences of these events are evaluated for the hypothetical maximally exposed individual at the time of licensing, changes in the plant environment will not affect these evaluations. Therefore, the staff concludes that the environmental impacts of design-basis accidents are of small significance for all plants. Because the environmental impacts of design basis accidents are of small significance and because additional measures to reduce such impacts would be costly, the staff concludes that no mitigation measures beyond those implemented during the current term license would be warranted. This is a Category 1 issue.

5.5.2 Impacts from Severe Accidents

5.5.2.1 Atmospheric Releases

The evaluation of health and dose effects caused by atmospheric releases used a prediction process to identify those plant sites that are bounded by existing analyses. Existing analyses represent only a subset of operating plants. A particular portion of this subset, specifically those plants having severe accident analyses in their respective FESs, was used in this evaluation. EI (which is a function of population and wind direction), in conjunction with the FES severe accident analyses, was then used to develop a means to predict consequences for all plants. Average values and 95 percent UCB values were estimated. Table 5.6 provides the results of this prediction process.

Results indicate that the predicted effects of a severe accident during MYR at the 74 sites of nuclear power plants in the United States are not expected to exceed a small fraction of that risk to which the population is already exposed. In addition, the dose to individuals was also predicted. Results indicate that the highest e individual dose would be 3 x  10-4 rem/RY. This dose compares to an average of 3 x 10-1 rem/person/year for all other causes, including radon. Therefore, the probability-weighted consequences from atmospheric releases associated with severe accidents is judged to be of small significance for all plants.

5.5.2.2 Fallout onto Open Bodies of Water

The results of comparative analyses for the drinking-water pathway concluded that Great Lakes sites have the same order-of-magnitude risk that was calculated in the Fermi 2 FES, which is only a small fraction of the risk from atmospheric pathway releases. River sites with potentially greater risk than in the Fermi FES are amenable to interdiction, which can significantly reduce risk. In the case of the aquatic food pathway, interdicted population exposures are less than or essentially the same as atmospheric pathway releases. For both the drinking water and aquatic food pathways, the probability-weighted consequences from fallout due to severe accidents is of small significance.

5.5.2.3 Releases from Groundwater

The comparative analyses for this pathway were done by first segregating all sites into six general categories as called out in the NRC LPGS (NUREG-0440) and then estimating if the risk consequences calculated in existing analyses (including the LPGS) bounds the risks for all other plants within each category.

Of the six categories, three are judged to be bound by existing analyses. These categories are Great Lake sites, estuaries, and dry sites.

For the other categories, estimates were made of the degree to which groundwater releases could exceed existing analyses. For all six categories, the staff concluded that the risk to the population was either a small fraction of that for atmospheric releases or, in a few cases, comparable to that from atmospheric releases. Therefore, the probability-weighted consequences from groundwater releases due to severe accidents is judged to be of small significance for all plants.

5.5.2.4 Societal and Economic Risks

The expected costs resulting from a severe accident at nuclear power plants during their renewal periods have been predicted from evaluations presented in 27 FESs. Estimates of the extent of land contamination have also been presented. In both cases, the conditional impacts are judged to be of small significance for all plants.

5.5.2.5 SAMDAs

The staff concluded that the generic analysis summarized above applies to all plants and that the probability-weighted consequences of atmospheric releases, fallout onto open bodies of water, releases to ground water, and societal and economic impacts of severe accidents are of small significance for all plants. However, not all plants have performed a site-specific analysis of measures that could mitigate severe accidents. Consequently, severe accidents are a Category 2 issue for plants that have not performed a site-specific consideration of severe accident mitigation and submitted that analysis for Commission review.


5.6 Endnotes

[ Prev | Next | Table of Contents ]

  1. While a dose as low as 10 rem may cause such observable physiological changes as chromosomal aberrations, these changes are not classified as clinical injury.
  2. Also referred to as the Rogovin report.
  3. Grand Gulf, Sequoyah, Surry, Peach Bottom, and Zion.
  4. The FitzPatrick and Nine Mile Point units are located closely enough to assume that they are located on the same site. A similar observation can be made for the Hope Creek and Salem units.
  5. Because the hypothetical sites were to be modeled as either PWRs or BWRs, those using population data of actual PWR sites utilized updated WASH-1400 source terms taken from the Byron FES (NUREG-0848), while those using population data for BWRs utilized updated WASH-1400 source terms taken from the Clinton FES (NUREG-0854).

5.7 REFERENCES

Acharya, S., and R. M. Blond, Transcript of Testimony Before the Atomic Safety and Licensing Board in the Matter of Consolidated Edison of New York (Indian Point, Unit 2) and Power Authority of the State of New York (Indian Point, Unit 3) (Docket Nos. 50-247-SP and 50-286-SP), Sections III.C.A and III.C.B, n.d.

BEIR-I, The Effects on Populations of Exposure to Low Levels of Ionizing Radiations, National Research Council, Advisory Committee on the Biological Effects of Ionizing Radiation, National Academy of Sciences, Washington, D.C., 1972.

BEIR-III, The Effects on Populations of Exposure to Low Levels of Ionizing Radiation: 1980, National Research Council, Advisory Committee on the Biological Effects of Ionizing Radiation, National Academy of Sciences, Washington, D.C., 1980.

BEIR-V, Health Effects of Exposure to Low Levels of Ionizing Radiation, National Research Council, Advisory Committee on the Biological Effects of Ionizing Radiation, National Academy of Sciences, Washington, D.C., 1990.

Bertini, H. W., Descriptions of Selected Accidents That Have Occurred at Nuclear Reactor Facilities, ORNL/NSIC-176, Oak Ridge National Laboratory, Oak Ridge, Tennessee, April 1980.

Codell, R. B., "Potential Contamination of Surface Water Supplies by Atmospheric Releases from Nuclear Plants," Health Physics 49(5), 713-30, November 1985.

ConEd (Consolidated Edison Company), Indian Point Nuclear Power Plant, Unit 2 Final Safety Analysis Report, June 1982.

Crick, M. J., and G. S. Linsley, An Assessment of the Radiological Impact of the Windscale Reactor Fire, National Radiological Protection Board, Chilton, United Kingdom, 1983.

DOE/ER-0332, M. Goldman et al., Health and Environmental Consequences of the Chernobyl Nuclear Power Accident, U.S. Department of Energy, Washington, D.C., 1987.

Eisenbud, M., Environmental Radioactivity from National, Industrial, and Military Sources, Academic Press, New York, 1987.

Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities," U.S. Nuclear Regulatory Commission, November 23, 1988.

Generic Letter 88-20, Supplement 1, "Initiation of the Individual Plant Examination for Severe Accident Vulnerabilities," U.S. Nuclear Regulatory Commission, August 29, 1989.

Generic Letter 88-20, Supplement 2, "Accident Management Strategies for Consideration in the Individual Plant Examination Process," U.S. Nuclear Regulatory Commission, April 4, 1990.

Generic Letter 88-20, Supplement 3, "Completion of Containment Performance Improvement Program and Forwarding Insights for Use in the Individual Plant Examination for Severe Accident Vulnerabilities," U.S. Nuclear Regulatory Commission, July 6, 1990.

Generic Letter 88-20, Supplement 4, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," U.S. Nuclear Regulatory Commission, June 28, 1991.

Goldman, M. "Chernobyl: A Radiological Perspective," Science, 238, 622-23, 1987.

Helton, J. C., et al. "Contamination of Surface-Water Bodies After Reactor Accidents by the Erosion of Atmospherically Deposited Radionuclides," Health Physics, 48(6), 757-71, 1985.

Isherwood, D., Preliminary Report on Retardation Factors and Radionuclide Migration, Lawrence Livermore Laboratory, Livermore, California, August 1977.

Lee, Y. T., et al., "A Comparison of Background Seismic Risks and the Incremental Seismic Risk Due to Nuclear Power Plants," Nuclear Engineering and Design, 53, 141-54, 1979.

NAS (National Academy of Sciences), Estimating Losses from Future Earthquakes--Panel Report, Panel on Earthquake Loss Estimation Methodology, National Academy Press, 1989.

NRC Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Office of Standards Development, November 1982, Reissued 1983.

NUREG-0056, Final Environmental Statement Related to the Manufacture of Floating Nuclear Power Plants by Offshore Power Systems, Vol. 1, U.S. Nuclear Regulatory Commission, September 1976.

NUREG-0440, Liquid Pathway Generic Study: Impacts of Accidental Radioactive Releases to the Hydrosphere from Floating and Land-Based Nuclear Power Plants, U.S. Nuclear Regulatory Commission, February 1978.

NUREG-0558, Population Dose and Health Impact of the Accident at the Three Mile Island Nuclear Station, U.S. Nuclear Regulatory Commission, May 1979.

NUREG-0651, Evaluation of Steam Generator Tube Rupture Accidents, U.S. Nuclear Regulatory Commission, March 1980.

NUREG-0713, Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities 1986, Vol. 8, U.S. Nuclear Regulatory Commission, August 1989.

NUREG-0737, Clarification of TMI Action Plan Requirements, U.S. Nuclear Regulatory Commission, November 1980.

NUREG-0769, Final Environmental Statement Related to the Operation of Fermi, Unit 2, U.S. Nuclear Regulatory Commission, August 1981.

NUREG-0769, Addendum 1, Final Environmental Statement Related to the Operation of Enrico Fermi Atomic Power Plant, Addendum No. 1, U.S. Nuclear Regulatory Commission, August 1982.

NUREG-0773, The Development of Severe Reactor Accident Source Terms: 1957-1981, U.S. Nuclear Regulatory Commission, November 1981.

NUREG-0775, Supplement, Final Environmental Statement Related to the Operation of Comanche Peak Steam Electric Station, Units 1 and 2, U.S. Nuclear Regulatory Commission, October 1989.

NUREG-0848, Final Environmental Statement Related to the Operation of Byron Station, Units 1 and 2, U.S. Nuclear Regulatory Commission, April 1982.

NUREG-0854, Final Environmental Statement Related to the Operation of Clinton Power Station, Unit 1, U.S. Nuclear Regulatory Commission, May 1982.

NUREG-0884, Final Environmental Statement Related to the Operation of Perry, Units 1 and 2, U.S. Nuclear Regulatory Commission, August 1982.

NUREG-0933, R. Emrst and W. Minners, Prioritization of Generic Safety Issues, U.S. Nuclear Regulatory Commission, Washington, D.C., December 1983.

NUREG-0956, Reassessment of the Technical Bases for Estimating Source Terms: Draft Report for Comment, U.S. Nuclear Regulatory Commission, July 1985.

NUREG-0974, Supplement 1, Draft Environmental Statement Related to the Operation of Limerick, Units 1 and 2, U.S. Nuclear Regulatory Commission, December 1983.

NUREG-0974, Supplement, Final Environmental Statement Related to the Operation of Limerick Generating Station, Units 1 and 2, U.S. Nuclear Regulatory Commission, August 1989.

NUREG-1054, R. Codell, Simplified Analysis for Liquid Pathway Studies, U.S. Nuclear Regulatory Commission, August 1984.

NUREG-1087, Final Environmental Statement Related to the Operation of Vogtle Electric Generating Plant, Units 1 and 2, U.S. Nuclear Regulatory Commission, March 1985.

NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, U.S. Nuclear Regulatory Commission, May 1989.

NUREG-1272, Office of Analysis and Evaluation of Operational Data 1989 Annual Report: Power Reactors, U.S. Nuclear Regulatory Commission, Office of Analysis and Evaluation of Operational Data, Washington, D.C., July 1990.

NUREG-1365, Revised Severe Accident Research Program Plan FY 1990-1992, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Division of Systems Research, Washington, D.C., August 1989.

NUREG-75/014, Reactor Safety Study--An Assessment, U.S. Nuclear Regulatory Commission, October 1975.

NUREG/CR-1250, Three Mile Island--A Report to the Commissioners and the Public, Vol. 1, U.S. Nuclear Regulatory Commission, January 1980.

NUREG/CR-1596, S. J. Niemczyk et al., The Consequences from Liquid Pathways After a Reactor Meltdown Accident, Sandia National Laboratories, U.S. Nuclear Regulatory Commission, June 1981.

NUREG/CR-1659, Reactor Safety Study Methodology Applications Program: Grand Gulf No. 1 BWR Power Plant, Vol. 4, U.S. Nuclear Regulatory Commission, October 1981.

NUREG/CR-2239, Technical Guidance for Siting Criteria Development, U.S. Nuclear Regulatory Commission, December 1982.

NUREG/CR-2300, PRA Procedures Guide, U.S. Nuclear Regulatory Commission, January 1983.

NUREG/CR-4674, J. W. Minarick et al., Precursors to Potential Severe Core Damage Accidents, 1985: A Status Report, prepared by Oak Ridge National Laboratory, Oak Ridge, Tennessee, for U.S. Nuclear Regulatory Commission, December 1986.

NUREG/CR-5225, Addendum 1, An Overview of BWR Mark-I Containment Venting Risk Implications, U.S. Nuclear Regulatory Commission, June 1989.

NUREG/CR-5278, Cost Analysis for Potential BWR Mark-I Containment Improvements, U.S. Nuclear Regulatory Commission, January 1989.

NUREG/CR-5528, An Assessment of BWR Mark-II Containment Challenges, Failure Modes and Potential Improvements in Performance, U.S. Nuclear Regulatory Commission, July 1990.

NUREG/CR-5529, An Assessment of BWR Mark-III Containment Challenges, Failure Modes, and Potential Improvements in Performance, U.S. Nuclear Regulatory Commission, January 1991.

NUREG/CR-5565, S. R. Greene et al., The Response of BWR Mark-II Containments to Station Blackout Severe Accident Sequences, prepared by Oak Ridge National Laboratory, Oak Ridge, Tennessee, for U.S. Nuclear Regulatory Commission, May 1991.

NUREG/CR-5567, PWR Dry Containment Issue Characterization, U.S. Nuclear Regulatory Commission, August 1990.

NUREG/CR-5575, Quantitative Analysis of Potential Performance Improvements for the Dry PWR Containment, U.S. Nuclear Regulatory Commission, August 1990.

NUREG/CR-5586, Mitigation of Direct Containment Heating and Hydrogen Combustion Events in Ice Condenser Plants, U.S. Nuclear Regulatory Commission, October 1990.

NUREG/CR-5589, Assessment of Ice-Condenser Containment Performance Issues, U.S. Nuclear Regulatory Commission, July 1990.

NUREG/CR-5602, Simplified Containment Event Tree Analysis for the Sequoyah Ice Condenser Containment, U.S. Nuclear Regulatory Commission, December 1990.

NUREG/CR-5623, BWR Mark-II Ex-Vessel Corium Interaction Analyses, U.S. Nuclear Regulatory Commission, 1991.

NUREG/CR-5630, R. G. Gido et al., PWR Dry Containment Parametric Studies, prepared by Sandia National Laboratories, for U.S. Nuclear Regulatory Commission, Office of Research, Division of Safety Issue Resolution, April 1991.

ORNL (Oak Ridge National Laboratory), Nuclear Safety Information Center, Description of Selected Accidents That Have Occurred at Nuclear Reactor Facilities, Oak Ridge, Tennessee, April 1980.

Parsons, P., "Underground Movement of Radioactive Wastes at Chalk River," in J. Morgan et al., eds., Proceedings of 2nd Conference on Groundwater Disposal of Radioactive Wastes, Chalk River, Canada, U.S. Atomic Energy Commission, TID-7628, 1962.

Report of the President's Commission on the Accident at Three Mile Island, Washington, D.C., October 1979.

Thompson, T. J., and J. G. Beckerley, The Technology of Nuclear Reactor Safety, Vol. 1, MIT Press, Cambridge, Massachusetts, 1964.

Tingle, A., CRAC2S-MACCS Comparison of Exposure Index Calculations, Brookhaven National Laboratory, Upton, New York, February 1993.

United Kingdom National Radiological Protection Board, An Assessment of the Radiological Impact of the Windscale Reactor Fire, October 1957.

 U.S. Department of Commerce, Fishery Statistics of the United States, 1976, Statistical Digest No. 70, National Oceanic and Atmospheric Administration, 1980.

Yambert, M., and M. Linn, Assessment of the Exposure Index as a Means of Predicting Potential Consequences Associated with Nuclear Power Plant Accidents, ORNL/NRC/LTR-92/10/RI, Oak Ridge National Laboratory, Oak Ridge, Tennessee, June, 1992.

Page Last Reviewed/Updated Thursday, March 29, 2012