ML24150A080

Method for Graded Risk-Informed Performance-Based Control Room Design Criteria

ML24150A080
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Issue date:	08/31/2024
From:	Elijah Dickson NRC/NRR/DRA
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Method for Graded Risk-Informed Performance-Based Control Room Design Criteria Framework September 2024 Elijah Dickson, Ph.D.

Division of Risk Analysis Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission

iii TABLE OF CONTENTS TABLE OF CONTENTS........................................................................................................... iii LIST OF TABLES...................................................................................................................... v EXECUTIVE

SUMMARY

......................................................................................................... vii ABBREVIATIONS.................................................................................................................... ix

1. INTRODUCTION...............................................................................................................11

2. BACKGROUND.................................................................................................................12 2.1 Role and Basis of CR Design Criteria............................................................................12 2.2 Severe Accident Policy and Significant Safety Improvements at US Plants...................13 2.3 PRA Policy and Staff Obligation to Address Unnecessary Burdens...............................13 2.4 Risk-Informed Performance-Based Regulations and Staff Obligation to Reward Improved Outcomes.................................................................................................................14

3. METHOD...........................................................................................................................15 3.1 Literature Review of Scientific Recommendations for Radiation Protection for Workers Under Emergency and Accident Conditions and Modern Health Physics and Epidemiology Knowledge..........................................................................................16 3.1.1 Review of Recommendations of Acceptable Exposures................................... 16 3.1.2 Review of Modern Health Physics and Radiation Epidemiology Knowledge..... 19 3.2 Literature Review of the State of Probabilistic Risk Assessment, Risk Metrics, and Suitability Assessment..............................................................................................21 3.2.1 Overview of Advancements in Probabilistic Risk Assessments........................ 21 3.2.2 Overview of Risk Metrics.................................................................................. 24 3.2.3 Assessment of Risk Metric Suitability for Grading Control Room Design Criteria

................................................................................................................... 26 3.2.3.1 Analysis Methods Supporting Compliance with the Control Room Design Criteria.......................................................................... 26 3.2.3.2 Summary of the Suitability of the CDF Risk Metric for Risk-Informed Grading................................................................................ 29

4. GRADED, RISK-INFORMED, AND PERFORMANCE-BASED FRAMEWORK.................30

iv 4.1 Analysis and Results.....................................................................................................31 Example 1Flexibility that encourages further safety improvements..................................31 Example 2Examining modeling fidelity, quality, and inclusion of external hazards............33 Example 3Examining differences among reactor designs................................................35 CONCLUSION........................................................................................................................37 ACKNOWLEDGEMENTS.......................................................................................................38 REFERENCES.......................................................................................................................39 APPENDIX A..........................................................................................................................44 APPENDIX B..........................................................................................................................49

v LIST OF TABLES Table 1: Graded, Risk-Informed, and Performance-Based Approach for Control Room Design Criteria Using CDF....................................................................................................................30 Table 2: Analysis Result Comparing 1980s to Contemporary CDF Risk Profiles.......................32 Table 3: Example 2Analysis Result Comparing License Renewal SAMA to TSTF-505 CDF Risk Profiles..............................................................................................................................34 Table 4: Example 3 Analysis Results Examining Differences Between Reactor Designs..........35 Figure 1: Method for Developing a Graded, Risk-Informed, and Performance-Based Framework

.................................................................................................................................................15 Figure 2: Depiction of Reviewed Source Material from Organizations Responsible for Making Radiation Protection Recommendations...................................................................................17 Figure 3: The NRCs Concept of PRA Acceptability..................................................................23 Figure 4: The NRCs Concept of PRA Acceptability for Different Licensing Applications...........24 Figure 5: Histogram of Graded, Risk-Informed, and Performance-Based Approach for Control Room Design Criteria Using CDF..............................................................................................31 Figure 6: Example 1Analysis Result Histogram Comparing 1980s to Contemporary CDF Risk Profiles......................................................................................................................................33 Figure 7: Example 2Analysis Results Histogram Comparing License Renewal SAMA to TSTF-505 CDF Risk Profiles.....................................................................................................34 Figure 8: Example 3Analysis Results Examining Difference Between Reactor Designs........36 Figure 9: Control Room Design Criteria of 10 CFR 50.67 and General Design Criterion 19, Control room: Typical Role of Accident Management Guidelines............................................50

vii EXECUTIVE

SUMMARY

This white paper presents a method to develop a framework for a graded, risk-informed and performance-based, control room design criterion for General Design Criterion 19, Control room, in Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, and 10 CFR 50.67(b)(2)(iii) on accident source term. In developing the method, several key Commission probabilistic risk assessment policies were considered and executed to propose a way to transition these rules to a risk-informed regulatory framework. The purpose of this method is to develop a framework that enables a performance based evaluation using traditional deterministic radiological consequence analysis methods within defined risk informed boundaries. These boundaries are defined by acceptable radiation exposure guidelines for radiation workers during accident and emergency conditions and acceptable contemporary nuclear facility risk profiles using modern probabilistic risk assessment methods.

Such a framework is intended to provide flexibility when determining how to meet an established acceptance criterion in a way that encourages and rewards safety of the facility. In practice, the method produces a framework that leverages in part, its safe design and operations to justify a higher control room design criterion with a lower plant-specific risk metric.

ix ABBREVIATIONS 10 CFR Title 10 of the U.S. Code of Federal Regulations ADAMS Agencywide Documents Access and Management System ALARA as low as is reasonably achievable ANSI American National Standards Institute ANS American Nuclear Society BWR Boiling Water Reactor DBA design-basis accident CDF Core Damage Frequency EPA U.S. Environmental Protection Agency FEMA Federal Emergency Management Agency FR Federal Register GDC general design criterion/criteria GEIS Generic Environmental Impact Statement Gy gray(s)

IAR informal assistance request IAEA International Atomic Energy Agency ICRP International Commission on Radiological Protection IPE Individual Plant Examination IPEE Individual Plant Examination for External Events IEFR Individual Early Fatality Risk LCFR Latent Cancer Fatality Risk LD 50/60 lethal dose to 50 percent of the people within 60 days without medical treatment LERF Large Early Release Frequency LPZ low population zone LWR light-water reactor LRF Large Release Frequency NCRP National Council on Radiation Protection and Measurements NEI Nuclear Energy Institute NRC U.S. Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation PWR Pressurized Water Reactor rem roentgen equivalent man RES Office of Nuclear Regulatory Research SRM staff requirements memorandum SSC structure, system, and component Sv sievert(s)

TSTF Technical Specifications Task Force TEDE total effective dose equivalent

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1. INTRODUCTION This white paper presents a method to develop a framework for a graded, risk-informed and performance-based control room design criterion for General Design Criterion 19, Control room, in Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, and 10 CFR 50.67(b)(2)(iii) on accident source term. In developing the method, several key Commission probabilistic risk assessment policies were considered and executed to propose a way to transition these rules to a risk-informed regulatory framework. The purpose of this method is to develop a framework that enables a performance based evaluation using traditional deterministic radiological consequence analysis methods within defined risk informed boundaries. These boundaries are defined by acceptable radiation exposure guidelines for radiation workers during accident and emergency conditions and acceptable contemporary nuclear facility risk profiles using modern probabilistic risk assessment methods. Such a framework is intended to provide flexibility when determining how to meet an established acceptance criterion in a way that encourages and rewards safety of the facility. In practice, the method produces a framework that leverages in part, its safe design and operations to justify a higher control room design criterion with a lower plant-specific risk metric.

Furthermore, the approach supports consistency within the Commission's comprehensive radiation protection framework. This framework is comprehensive in that it provides adequate protection for workers and the public during normal and abnormal situations, including accident conditions. The regulations in 10 CFR Part 20, Standards for Protection Against Radiationwhich are based, in part, on the recommendations of the International Commission on Radiological Protectionserve as the foundation of this radiation protection framework. However, when revising that regulation in 1991, the Commission stated its intent that the regulation be observed to the extent practicable during emergencies, but that conformance with the regulation should not hinder any actions that are necessary to protect public health and safety, such as saving lives or maintaining confinement of radioactive materials. In order to ensure that adequate protective measures can and will be taken in the event of a radiological emergency, the Commission requires that onsite and offsite facility emergency response plans meet regulatory standards that include means for controlling radiological exposures to emergency workers consistent with the U.S.

Environmental Protection Agency (EPA) Emergency Worker and Lifesaving Activity Protection Action Guides. Aside from providing dose-based guidelines for different functions during emergency response (e.g., protection of critical infrastructure and lifesaving), the EPA guidelines apply the fundamental radiation protection operating principle of maintaining doses as low as reasonably achievable, which means making every reasonable effort to maintain exposures to radiation as far below the dose limits as is practical, consistent with the purpose for which the activity is undertaken. The combination of the system of dose limits and requirements provided by 10 CFR Part 20 and the guidelines for emergency response that must be included within licensee emergency response plans ensures adequate protection when considering exposures to workers and the public. One aim of this comprehensive radiation protection framework is to ensure that practices involving radiation exposure are justified with regard to the benefit of the activity.

Inherent in this effort is a consideration of the risks incurred by the workers who provide the benefit.

Therefore, limiting the control room design criterion within an acceptable range of numerical values while leveraging facility-specific risk information to determine facility specific criteria ensures the design criterion aligns with the intentions of the radiation protection framework.

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2. BACKGROUND This section provides a synopsis of the NRCs role of establishing stringent design criteria for the control room and how three key elements of PRA-related policies were considered in formulating the method proposed. First, a discussion is provided of the role and bases of the control room design criterion. Then, Section 2.2 summarizes how various initiative such as the issuance of GL 88-20 prompted significant improvements to plant safety. Section 2.3 highlights NRC staff obligation to remove unnecessary conservatisms. Finally, Section 2.4 provides an excerpt from SRM-SECY 98-144 highlights the role that risk-informed performance-based regulation may be used to reward improved safety.

2.1 Role and Basis of CR Design Criteria General Design Criterion (GDC) 19, Control room, in Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, was developed and issued to establish the minimum necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components (SSCs) that provide reasonable assurance that the facility can be operated without undue risk to public health and safety. Both GDC 19 and 10 CFR 50.67(b)(2)(iii) provide a dose-based criterion of 50 millisieverts (mSv)

(5 rem total effective dose equivalent (TEDE)) for demonstrating the acceptability of the control room design. These regulatory requirements represent a distinct layer of defense in depth required to ensure safety in the unlikely event of a major accident that results in substantial meltdown of the reactor core with subsequent release of appreciable quantities of fission products. These requirements ensure that a licensee or applicant provides adequate radiation protection to permit access to and occupancy of the control room under accident conditions, using traditional deterministic radiological consequence analyses methods to judge the acceptability of the SSC design (e.g., control room habitability system).

It is important to state that GDC 19 and 10 CFR 50.67(b)(2)(iii) are design criteria and not operational limits. While the design criteria are computed in terms of dose, they are figures of merit used to characterize the minimum necessary design, fabrication, construction, testing, and performance of the requirements for SSCs. The design criteria do not represent actual occupational exposures received during normal and emergency conditions, which are primarily controlled by 10 CFR Part 20, Standards for Protection Against Radiation. The statements of consideration for the 10 CFR 50.67, Accident source term, rulemaking (64 FR 71997; December 23, 1999) explain this concept as follows:

the use of 0.05 Sv [sievert] (5 rem) TEDE as the control room criterion does not imply that this would be an acceptable exposure during emergency conditions, or that other radiation protection standards of Part 20, including individual organ dose limits, might not apply. This criterion is provided only to assess the acceptability of design provisions for protecting control room operators under postulated DBA [design-basis accident] conditions.

With the exception of the 1999 rulemaking efforts (10 CFR 50.67(b)(2)(iii)), the original record of decision for the selection of the numerical value of the design criteria is not readily available.

13 Although the author has located early drafts of the criteria, some correspondence, and some meeting minutes, they do not directly address the bases for the numerical criteria. The statements of consideration for the 1971 rule, which first published the GDC, addressed the criteria only in the aggregate; the individual criteria were not discussed. There is, however, a record of a change made to the proposed GDC 11, which became the final GDC 19 (32 FR 10213, 1967) (32 FR 10213; July 11, 1967). The proposed GDC 11 referred to the occupational exposure limits of 10 CFR Part 20 rather than specifying a numerical dose criterion. Industry comments generally recommended deletion of the reference to 10 CFR Part 20. These comments were resolved by deleting the reference to 10 CFR Part 20 occupational exposure limits and providing the current 5 rem whole-body, or its equivalent to any part of the body, for the duration of the accident in its place (see SECY-R-143, Amendment to 10 CFR 50 General Design Criteria for Nuclear Power Plants, dated January 28, 1971 (USNRC, 1971)).

The statements of consideration for the 10 CFR 50.67 rulemaking (64 FR 71990, 1999) included the staffs rationale for establishing 5 rem (0.05 Sv) TEDE as the GDC 19 numeric design criterion for licensees using an alternative source term as follows:

The bases for the NRCs decision are: first, that the criteria in GDC 19 and that in the final rule are based on a primary occupational exposure limit.

In summary, the selection of the numerical value of the control room design criterion, developed in the late 1950s, was based on occupational exposure limits at that time. Furthermore, during the development of 10 CFR 50.67 in the 1990s, selection of the numerical value of the control room design criterion maintained its basis as being the primary occupational exposure limit of 10 CFR Part 20.

2.2 Severe Accident Policy and Significant Safety Improvements at US Plants When developing the method, the U.S. Nuclear Regulatory Commission (NRC) staff considered several key Commission-directed probabilistic risk assessment (PRA) policies that advocate certain changes to the development and implementation of its regulations using risk-informed and, ultimately, performance-based approaches. First, the author reviewed the Severe Reactor Accident Policy Statement (50 (60 FR 42622, August 16, 1995) FR 32138; August 8, 1985),

which describes the policy related to accidents more severe than the design-basis accidents.

The Severe Reactor Accident Policy Statement recognizes the usefulness of PRAs in identifying severe accident vulnerabilities and providing additional insights to ensure that nuclear power plants do not result in an undue risk to public health and safety and ultimately prompted significant improvements in plant safety. For example, as will be demonstrated later in this report, GL 88-20 and its supplements issued since the publication of the Severe Accident Policy (also referred to as IPEs and IPEEEs) resulted in significant reductions in US plants core damage frequencies (CDF). These significant reductions CDF, a manifestation of significant improvements, is one of the several reasons which enables the staff to consider relaxing dose acceptance criteria for the control room design.

2.3 PRA Policy and Staff Obligation to Address Unnecessary Burdens Next, the author reviewed the PRA Policy Statement (60 FR 42622, August 16, 1995), which formalized the Commissions commitment to risk-informed regulation through the expanded use of PRA. The PRA Policy Statement states, in part, the following:

14 The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data, and in a manner that complements the NRCs deterministic approach and supports the NRCs traditional defense-in-depth philosophy.

The Policy Statement discusses how it should be used to reduce unnecessary conservatisms. It states the following regarding use of PRA to reduce unnecessary conservatisms.

PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatisms associated with current regulatory requirements, regulatory guides, license commitments, and staff practices.

Recent experiences and pending needs of operating plants are confronting significant burdens to operate their facilities safely and economically. The need to consider whether there are unnecessary burdens that can be using PRA insights is another reason that prompted the proposed method.

2.4 Risk-Informed Performance-Based Regulations and Staff Obligation to Reward Improved Outcomes Then, the author reviewed Staff Requirements Memorandum (SRM)-SECY-98-144, Staff RequirementsSECY-98-144White Paper on Risk-Informed and Performance-Based Regulation, dated March 1, 1999, which defines the terms and Commission expectations for risk-informed and performance-based regulation (USNRC, 1998). The methodology described in this paper uses the terms and concepts of SRM-SECY-98-144, Item 8, Risk-Informed, Performance-Based Approach, which reads as follows:

risk-informed, performance-based approach to regulatory decision-making combines the risk-informed and performance-based elements discussed in Items 5 and 7, above, and applies these concepts to NRC rulemaking, licensing, inspection, assessment, enforcement, and other decision making. Stated succinctly, a risk-informed, performance-based regulation is an approach in which risk insights, engineering analysis and judgment including the principle of defense-in-depth and the incorporation of safety margins, and performance history are used, to (1) focus attention on the most important activities, (2) establish objective criteria for evaluating performance, (3) develop measurable or calculable parameters for monitoring system and licensee performance, (4) provide flexibility to determine how to meet the established performance criteria in a way that will encourage and reward improved outcomes, and (5) focus on the results as the primary basis for regulatory decision-making.

In accordance with the subject policy, when formulating risk-informed performance-based rules, staff could consider means to encourage and reward improved outcomes.

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3. METHOD Over the decades, little work has been done to develop a control room design criterion basis that considers other risk information and contemporary probabilistic risk insights. When assessing the numerical selection and comparing it to recommendations from national and international organizations responsible for radiation protection standards, PRA technology, and regulatory precedents, the 5 rem TEDE control room design criterion value is on the lower side of a range of recommended values for emergency response planning to protect against actual incurred radiation exposure during an event. Given these other sources, considerable information is now available.

In application, the presented methodology produces a safe, reasonable, graded, risk-informed, and performance-based control room design criterion and is consistent with the Commissions policies. By doing so, the methodology enables a clear deterministic evaluation within defined risk-informed boundaries. The criterion itself would be risk-informed based on plant-specific risk information and contemporary understanding of radiological health effects and radiation protection recommendations. The performance-based aspects would continue to be supported by engineering analysis and judgment, including the principle of defense in depth and the incorporation of safety margins.

The proposed method involves systematically mapping a predetermined range of acceptable dose-based control room design values onto a specified risk-metric criteria range.

The method is illustrated in Figure 1, where the y-axis represents a predetermined range of acceptable control room design values, and the x-axis represents a predetermined risk-metric range. In practice, a lower plant-specific risk metric would justify a higher control room design criterion.

Figure 1: Method for Developing a Graded, Risk-Informed, and Performance-Based Framework

16 To define the axis of the framework, the author performed a literature review to assess (1) the contemporary recommendations for radiation protection for workers under emergency and accident conditions and (2) the state of PRA, various risk metrics, and the suitability of these risk metrics for use in this framework.

Section 3.1 provides results of a literature review of scientific recommendations for radiation protection for workers under emergency and accident conditions as well as for modern health physics-and epidemiology knowledge. Section 3.2 provides results of a literature review of the state of probabilistic risk assessment, risk metrics, and a suitability assessment. Following each section describing an element of the literature review, bordered text boxes summarize the bases for defining the x-and y-axes of the framework. Using the framework, the author performed a series of analyses to understand how the nuclear reactor fleet would align within each control room design criterion bin. Since the CDF was selected as the most appropriate risk-metric, a CDF dataset was developed from extracting results from various regulatory initiatives. This dataset is provided in Appendix A. Results from the analyses are then also presented.

3.1 Literature Review of Scientific Recommendations for Radiation Protection for Workers Under Emergency and Accident Conditions and Modern Health Physics and Epidemiology Knowledge Sections 3.1.1 and 3.1.2 gives the results of a review of a number of source materials to understand the current recommendations from national and international organizations responsible for making recommendations for radiation protection standards. To provide context to these recommendations, this paper includes a brief discussion of knowledge related to modern health physics and radiation epidemiology.

3.1.1 Review of Recommendations of Acceptable Exposures The purpose of this review was to establish a range of acceptable control room design values. Figure 2 depicts some of the reviewed source material.

17 Figure 2: Depiction of Reviewed Source Material from Organizations Responsible for Making Radiation Protection Recommendations Highlights of this review include the following:

Regulations:

At the time that GDC 19 was published in 1971, 10 CFR Part 20 limited occupational radiation exposure to 0.03 Sv (3 rem) whole-body dose per calendar quarter, provided the total lifetime dose was verified not to exceed 0.05 Sv (5 rem) times the individuals age in years minus 18 (i.e., 5(N18)). It was possible to receive a radiation exposure of up to 0.12 Sv (12 rem) in a given year.

The current annual limit on occupational radiation dose exposure in 10 CFR 20.1201, Occupational dose limits for adults, is 0.05 Sv (5 rem) TEDE. Under 10 CFR 20.1201, it is possible to receive occupational radiation exposure of up to 0.10 Sv (10 rem) TEDE over a 12-month period straddling two calendar years.

The current 10 CFR 20.1206, Planned special exposures, permits an adult worker to receive doses in addition to and accounted for separately from the doses received under the limits specified in 10 CFR 20.1201 of five times the annual dose limits during the individuals lifetime, not to accumulate faster than 0.05 Sv (5 rem) TEDE in any one year.

As discussed in the 1991 revision to the NRCs standards for protection against radiation (56 FR 23360; May 21, 1991), specifically at 56 FR 23372, the amended 10 CFR Part 20 retained planned special exposures in part to address the fact that under the new system of dose limitation, workers would no longer have a lifetime dose limit, or dose bank, equaling five times the quantity of the age of the worker

18 minus 18, or 5(N18). While the staff predicted that planned special exposures would be used infrequently, it expected that they would be applied if the elimination of the lifetime dose limits might create a severe handicap to the licensees operations. Therefore, the Commission concluded that an infrequent exposure of workers up to twice the occupational dose limit was adequately protective of radiation workers.

The regulations in 10 CFR 50.47(b)(11) address control of radiological exposures in an emergency, stating the following:

(11) Means for controlling radiological exposures, in an emergency, are established for emergency workers. The means for controlling radiological exposures shall include exposure guidelines consistent with EPA Emergency Worker and Lifesaving Activity Protective Action Guides.

The events that could lead to control room radiation exposures comparable to the 10 CFR Part 20 normal occupational exposure limit of 5 rem (50 mSv) TEDE would result in the activation of the facilitys emergency response plan and the emergency response organization. The U.S. Environmental Protection Agency (EPA) exposure guidelines found in EPA-400/R-17/001, PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents, issued January 2017 (EPA, 2017),

recommend that doses received under emergency conditions should be maintained as low as reasonably achievable (ALARA) and, to the extent practicable, limited to 0.05 Sv (5 rem). The guideline for actions to protect valuable property is 0.10 Sv (10 rem) where a lower dose is not practicable, the guideline for actions to save a life or to protect large populations is 0.25 Sv (25 rem) where a lower dose is not practicable, and exposures greater than 0.25 Sv (25 rem) may be appropriate for lifesaving or protecting large populations if the workers are volunteers who are fully aware of the risks involved.

Organizations Responsible for Radiation Protection Recommendations Health Physics Society (HPS) Position Statement PS010-4, Radiation Risk in Perspective, issued January 2020 (HPS, 2020), states that substantial and convincing scientific data show evidence of health effects following high-dose exposures (i.e., many multiples of natural background). However, below levels of about 100 mSv (10 rem) above background from all sources combined, the observed radiation effects in people are not statistically different from zero.

International Commission of Radiological Protection (ICRP) Publication 109, Application of the Commissions Recommendations for the Protection of People in Emergency Exposure Situations, issued 2009 (ICRP, 2009), specifies a range of 0.02-0.10 Sv (2-10 rem) acute for emergency exposure situations. For doses above 0.10 Sv (10 rem), protective measures should be justified.

19 International Atomic Energy Agency (IAEA) guidance (IAEA, 2024) for emergency workers specifies a range of 0.05-1 Sv (5-100 rem), depending on the severity of the actions needed.

National Council on Radiation Protection and Measurements (NCRP) Report No. 180, Management of Exposure to Ionizing Radiation: Radiation Protection Guidance for the United States, issued 2018 (NCRP, 2018), specifies that (1) during lifesaving activities or actions to prevent a catastrophic situation, which includes other urgent rescue activities, 0.5 gray (Gy) cumulative whole-body absorbed dose (50 rad) should be implemented at the command level, and (2) for other emergency activities, including extended activities following initial lifesaving, rescue, and damage control response, an effective dose to emergency workers should not exceed 0.10 Sv (10 rem).

3.1.2 Review of Modern Health Physics and Radiation Epidemiology Knowledge A review of modern health physics and radiation epidemiology knowledge was provided by Brock, et. Al, (USNRC, 2023a) in the context of the Commissions radiation protection framework. Brocks review is support by Regulatory Guide 8.29, Revision 1, Instruction Concerning Risks from Occupational Radiation Exposure, issued February 1996 (USNRC, 1996), which contains more information on risk from ionizing radiation. Both sources of information are discussed to provide technical background for the recommendations provided in Section 3.1.1 The source materials present in Section 3.1.1 provide their recommendations as two radiation units that assess either stochastic effects or deterministic radiation health effects. The first is the Sv (or rem TEDE), which is the sum of the effective dose equivalent for external exposures and the committed effective dose equivalent for internal exposures. The rem TEDE adjusts the dose equivalent radiation exposure using a tissue weighting factor that represents the proportion of the risk of stochastic effects resulting from irradiation of an organ, or tissue, to the total risk of stochastic effects when the whole body is irradiated uniformly. The committed effective dose equivalent is a 50-year committed dose based on an initial intake of radioactive material used to estimate the stochastic health effect of cancer mortality. In other words, the dose is assigned to the first year of intake to estimate the increased probability of cancer-induced fatality after 50 years. The second radiation unit is the Gy, or rad, which is used to measure the amount of energy deposited in tissue. Typically, radiation exposures that cause deterministic health effects are measured in Gy (rad).

Exposure of the whole body to a large dose over a short period of time (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) may cause effects due to the sensitivity of cells in the body. Acute radiation syndrome can result following significant whole-body exposures in a short time. Effects may include blood changes, nausea, vomiting, diarrhea, and central nervous system damage. Hematopoietic syndrome is observed as a decrease in blood cell count at doses of about 1 Gy (100 rad). Gastrointestinal syndrome from a dose of about 5 Gy (500 rad) will result in nausea, vomiting, and diarrhea.

Central nervous system syndrome, observed at about a dose of 20 Gy (2,000 rad), will affect the muscle and brain function of the central nervous system. The dose that is lethal to

20 50 percent of the people within 60 days if medical treatment is not provided is called the LD 50/60. The LD 50/60 dose is approximately 3-5 Gy (300-500 rad); this is about 90 times the annual dose limit for routine occupational exposure in an hour to an average adult. If received within a short time period (e.g., a few hours), an LD 50/60 dose will cause vomiting and diarrhea within a few hours and loss of hair, fever, and weight loss within a few weeks. These effects would not occur if the same dose were accumulated gradually over many weeks or months, such as during radiation therapy treatments.

The radiation risk of cancer mortality is assessed by first quantifying the baseline risk of cancer death within a given population. According to the National Research Council, Health Effects of Exposure to Low Levels of Ionizing Radiation, Report of the Committee on the Biological Effects of Ionizing Radiation, published in 1990, approximately one in five adults normally will die from cancer from all possible causes such as smoking, food, alcohol, drugs, air pollutants, natural background radiation, and inherited traits. (NRC, 1990) Thus, in any group of 10,000 workers, about 2,000 (20%) will die from cancer without any occupational radiation exposure. In Regulatory Guide 8.29, Revision 1, Instruction Concerning Risks from Occupational Radiation Exposure, the NRC adopted a risk value for an occupational dose of 0.01 Sv (1 rem) TEDE of 4 in 10,000 of developing a fatal cancer, or approximately 1 chance in 2,500 of fatal cancer per rem of TEDE received. (USNRC, 1996) The uncertainty associated with this risk estimate does not rule out the possibility of higher risk, or the possibility that the risk may even be zero at low occupational doses and dose rates. The radiation risk incurred by a worker depends on the amount of dose received. Under the linear no threshold model, a worker who receives 0.05 Sv (5 rem) in a year incurs 10 times as much risk as another worker who receives only 0.005 Sv (0.5 rem).

Thus, in a group of 10,000 people, each exposed to 0.01 Sv (1 rem) of ionizing radiation, and using the risk factor of 4 effects per 10,000 rem of dose, 4 of the 10,000 people might die from delayed cancer because of that 0.01 Sv (1 rem) dose in addition to the 2,000 normal cancer fatalities expected to occur in that group from all other causes. However, the natural incidence of fatal cancers is not precisely 2,000, and it is not possible to unequivocally distinguish these additional cases from those occurring naturally. From an individual perspective, a 0.01 Sv (1 rem) dose may increase an individual workers chances of dying from cancer from 20 percent to 20.04 percent. If ones lifetime occupational dose is 0.1 Sv (10 rem), the estimate would increase to 20.4 percent. A lifetime dose of 1 Sv (100 rem) may increase chances of dying from cancer from 20 to 24 percent. This small increase in cancer risk could be inferred over the lifetime, however it is unlikely that an increased incidence of cancer due to irradiation would be discernible. This is because the normal variability in baseline rates of cancer incidence is much larger than the inferred radiation-associated cancer rates. As a point of reference, according to NUREG-0713, Volume 43, Occupational Radiation Exposure at Commercial Nuclear Power Reactors and other Facilities, published in 2021, the average measurable dose for radiation workers reported to the NRC was 0.0016 Sv (0.16 rem) for 2021. (USNRC, 2021)

21 The control room design criterion radiation unit of rem TEDE does not technically match the expected measured deterministic health effects expected from a reactor accident. However, the 10 CFR Part 20 annual occupational exposure limit of 0.05 Sv (5 rem) TEDE is set sufficiently low that no deterministic threshold dose would be reached. This occupational exposure limit is applicable to both normal and emergency conditions. The limitation of stochastic effects is achieved by keeping all justifiable exposures ALARA, with economic and social factors considered, subject always to the boundary condition that the appropriate dose limit is not exceeded. Deterministic effects have a threshold dose that must be exceeded for the effects to occur, and the severity of these effects also increases with dose. Deterministic effects are prevented by setting dose equivalent limits at sufficiently low values so that no threshold dose would be reached, even following exposure for the whole of a lifetime or for the total period of an individuals working life.

3.2 Literature Review of the State of Probabilistic Risk Assessment, Risk Metrics, and Suitability Assessment This section presents advancements in PRA methodologies and technology that have propelled the field to a level of maturity at which it now can complement and enhance the demonstration of compliance with the 10 CFR 50.67 and GDC 19 acceptance criteria in a risk-informed manner. First, this section gives a brief overview of advancements in PRAs and how risk information has enhanced nuclear safety over several decades. Next, it assesses common risk metrics used in nuclear risk analyses. Lastly, it explains the selection and rationale for using the CDF risk metric for applying the method to develop a graded, risk-informed, and performance-based framework.

3.2.1 Overview of Advancements in Probabilistic Risk Assessments The NRC first developed the concept of nuclear power plant PRA in the 1970s. Since then, the agency has refined its methods and developed new risk insights. The NRC combines these insights with traditional engineering methods to make regulatory decisions about power plants, medical uses of nuclear materials, and the handling of nuclear waste. This risk-informed approach to regulation has been applied in several areas. Licensees use PRA for integrated plant evaluations that discover and correct vulnerabilities, resulting in significant improvements to reactor safety. Inspections use PRA insights to focus on plant systems, operations, and human performance that are most important to safety. The significance determination process in the NRCs Reactor Oversight Process uses plant-specific PRA models to assess the significance of inspection findings. The NRC increases its inspection and oversight as nuclear The review of recommendations for radiation protection standards and the knowledge of modern health physics and radiation epidemiology has identified a range of values within 0.1 to 0.25 Sv (10 to 25 rem). A criterion in this range would continue to provide reasonable assurance that the facility can be operated during an emergency without undue risk to public health and safety, as it would be set sufficiently low to protect against deterministic health effects that would cause operator impartment. This range of values is used to define the y-axis of the framework.

22 plant problems rise in risk importance. The NRC often uses PRA to confirm that new or revised rules are rigorous enough to cover uncertainties and to justify new requirements. The nuclear industry also uses PRA to enhance existing plant designs by reducing vulnerabilities and to analyze and enhance new reactor designs before asking the NRC to certify them. There is also a tremendous amount of licensing experience to leverage a licensees facility-specific PRA to reduce risk when multiple systems are being maintained. Often, PRA methods have also focused licensee resources on the most safety-significant systems and components, through risk-informed technical specifications, in-service inspection programs, and categorization of SSCs.

In any regulatory decision, the goal is to make a sound safety decision based on technically defensible information. Therefore, to make a regulatory decision relying upon risk insights as one source of information, the regulator must have confidence in the PRA results from which the insights are derived. Consequently, the PRA needs to have the proper scope and technical attributes to give an appropriate level of confidence in the results used in regulatory decision-making.

The NRC recognizes that these PRA aspects can vary depending on the specific decision under consideration. In SECY-00-0162, Addressing PRA Quality in Risk-Informed Activities, dated July 28, 2000 (USNRC, 2000), the staff described its approach to addressing the issue of PRA scope and quality in risk-informed activities in response to the Commissions SRM dated April 18, 2000 (USNRC, 2000a). This document was prepared in response to the U.S. General Accounting Office (GAO) report GAO/RCED-99-95, Nuclear Regulation: Strategy Needed to Regulate Safety Using Information on Risk, issued March 1999 (GAO, 1999). The GAO report identified a number of issues that GAO believed required resolution for the NRC to successfully implement a risk-informed regulatory approach. Among these, the GAO indicated that more guidance was needed to develop standards on the scope and detail of risk assessments needed for utilities to determine that changes to their plants design will not negatively affect safety.

Since the publication of the GAO report, the scope, depth, and technical content of the PRA have increased along with their purposes in risk-informed regulations. This is due, in part, to PRA standards development by the American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS). In addition, the National Fire Protection Association (NFPA) developed a fire protection standard with an appendix on fire PRA. ASME has issued a standard for a full-power, internal events Level 1 PRA and a limited Level 2 PRA (ASME, 2024).

The ANS also issued a standard for external events risk (ANSI/ANS, 2007). Also, the reactor owners groups have developed and applied a PRA peer review program for several years (NEI, 2006a).

The NRC has reviewed and approved the results of many of these initiatives through various regulatory processes. For example, Regulatory Guide 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, issued December 2020 (USNRC, 2020), provides an approach for determining whether the base PRA, in total or the parts that are used to support an application, is acceptable for use in regulatory decision-making for light-water reactors. Regulatory Guide 1.200 currently endorses the previous PRA standard ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic

23 Risk Assessment for Nuclear Power Plant Applications. (ASME/ANS, 2009) This standard addresses PRA for CDF and large early release frequency for internal and external hazard groups during at-power operations. Regulatory Guide 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance, issued May 2006 (USNRC, 2006), provides guidance on an acceptable method for use in complying with the Commissions requirements in 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, with respect to the categorization of SSCs that are considered in risk-informing special treatment requirements. Then, NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, issued March 2017 (USNRC, 2017), provides guidance on how to treat uncertainties associated with PRA in risk-informed decision-making. This guidance is intended to foster an understanding of the uncertainties associated with PRA and their impact on the results. Regulatory Guide 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, issued January 2018, (USNRC, 2018),

provides an approach for developing risk-informed applications for a licensing basis change that considers engineering issues and applies risk insights. It contains general guidance concerning analysis of the risk associated with proposed changes in plant design and operation. Lastly, the methodology based on Nuclear Energy Institute (NEI) Topical Report 06-09, Revision 0-A (NEI 06-09-A), Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, issued November 2006 (NEI, 2006) provides a methodology to evaluate and extend the completion times for required actions in the technical specification limiting condition for operations. These various regulatory risk-informed applications are consistent with the philosophy of Regulatory Guide 1.174 and the technical adequacy expectations for the PRA model explained in Regulatory Guide 1.200. As such, the NRC has increasingly used the results from PRAs in the regulatory process, starting with generic safety issue prioritization and progressing to regulatory analysis in support of rulemaking and backfits and currently risk-informed regulation, which has opened up the possibility of using PRA information in many new ways.

Figure 3 visualizes the NRCs concept of the acceptability of a PRA model in which each element of staff positions, national consensus standards, and peer review depends on the other (USNRC, 2020).

Figure 3: The NRCs Concept of PRA Acceptability

24 Figure 4 illustrates the NRCs concept of PRA acceptability for different licensing applications.

These licensing applications consider the required scope, level of detail, technical robustness, and plant representation modeled. Depending on the application, a greater reliance on the model to provide more operational flexibilities also comes with more complex staff reviews. The models associated with applications under Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion TimesRITSTF Initiative 4b, dated July 2, 2018 (TSTF, 2018), require the greatest level of detail, technical robustness, and plant representation.

Figure 4: The NRCs Concept of PRA Acceptability for Different Licensing Applications 3.2.2 Overview of Risk Metrics The NRC Safety Goal Policy Statement published 1986 (51 FR 30028; August 21, 1986),

established qualitative safety objectives (QSOs) and provided quantitative health objectives (QHOs) that the staff could use to determine whether QSOs have been achieved to limit the level of radiological risk from accidents at operating reactors. This policy statement focuses on the risks to the public from nuclear power plant operation. Its objective is to establish goals that broadly define an acceptable level of radiological risk. These two supporting objectives are based on the principle that nuclear risks should not be a significant addition to other societal This section presents advancements in PRA methodologies and technology that have propelled the field to a level of maturity at which it now can complement and enhance the demonstration of compliance with the 10 CFR 50.67 and GDC 19 acceptance criteria in a risk-informed manner. This method will leverage PRA results from approved license amendment requests under TSTF-505. Acceptability of the model is determined with respect to scope, level of detail, conformance to consensus standard technical elements (i.e., technical robustness), and plant representation. Quantification of risk due to internal fire and other significant external events is also necessary for this application, through PRA or bounding methods. For each facility, an all hazard risk metric (CDF) is computed by summing all CDF results (e.g., internal, flood, fires, seismic, high winds).

25 risks. The Commission recommended the Large Release Frequency (LRF) of a facility be used as implementing guidance. However, the Commission did not define the LRF. Attempt by the staff to define an LRF that can be used as a risk metric was discontinued since various definitions postulated for LRF by the staff were determined to be more conservative than the QHOs. As discussed in SRM SECY 97-077, (USNRC, 1997), based on proposals made by the staff, the Commission directed staff to abandon efforts to define LRF and endorsed the use of CDF as a surrogate for latent cancer fatality risk (LCFR) and large early release frequency (LERF) as a surrogate for individual early fatality risk (IEFR) as the appropriate risk metrics for operating reactors.

The analyses to compute CDF and large early release frequency are data driven, leveraging operational experience and probabilistic models to understand how the engineered system responds to diverse hazards and events. By assessing these data though PRA modeling techniques, simulations of a wide range of hazards are possible. Risk insights are drawn from these analyses to provide valuable understanding of the systems performance and vulnerabilities, complex interplay between subsystems, unique operational factors, and response to various external threats. Both CDF and LERF risk metrics have been used historically to varying degrees to assess and report on predicted individual reactor performance.

Despite lacking a Commission endorsed definition of LRF however, it is also used as a risk-metric to assess the adequacy of plant safety during design reviews. CDF represents the likelihood of a severe nuclear reactor accident resulting in damage to the reactor core. It quantifies the probability of a core meltdown, which can lead to significant radioactive release into the containment. There can be a small fraction of CDF sequences that my bypass the containment such as in the case a steam generator tube rupture. In contrast, large early release frequency assesses the likelihood of a severe accident leading to a significant and rapid release of radioactive material into the environment before emergency planning measures can be implemented, often associated with early containment failure. In addition to these risk metrics, others have been used to directly assess offsite human health consequences.

Aside from CDF and LERF, the nuclear industry has used other more direct human-health risk metrics. For example, NUREG-1150, Severe Accident Risks: An Assessment for Five U.S.

Nuclear Power Plants, issued December 1990 (USNRC, 1990), assessed several offsite human health consequence measures. These measures include early fatalities, latent cancer fatalities, total population dose (within 50 miles and entire site region), and two measures for comparison with the NRCs safety goals. These two measures were (1) the average individual early fatality probability within 1 mile and (2) average individual latent cancer fatality probability within 10 miles of the site boundary. A later NRC assessment, NUREG/CR-7110, Volume 1, Revision 1, State-of-the-Art Reactor Consequence Analyses Project: Peach Bottom Integrated Analysis, issued May 2013 (USNRC, 2013) (SOARCA), presented offsite consequences in terms of probabilities and risks to members of the public for a few selected accident scenarios.

The report tabulated both conditional probabilities and scenario-specific risks (the product of accident frequency and conditional probability) of health effects for individuals. The conditional probabilities assume that the accident occurs and present the probabilities of health effects occurring to individuals as a result of the accident. The scenario-specific risks are the product of the accident frequency and the conditional probabilities. These scenario-specific risks represent the likelihood of receiving a fatal cancer or early fatality for an average individual living within a specified radius of the plant per year of plant operation. In other words, SOARCA assessed

26 latent cancer fatality risk and early fatality risk to residents in circular regions surrounding the nuclear facility (USNRC, 2014).

3.2.3 Assessment of Risk Metric Suitability for Grading Control Room Design Criteria Section 3.2.3 provides an assessment of the suitability for grading the control room design criteria with the risk metrics discussed in Section 3.2.2. First, Section 3.2.3.1 provides a brief summary of traditional deterministic radiological consequence analysis. This summary includes the analyzed accident sequences for deriving the appliable accident source terms and examining attributes of the dose-receptor locations. Then, Section 3.2.3.2 provides the summary of the rationale for selecting CDF as the appropriate risk metrics for graded the control room design criteria.

3.2.3.1 Analysis Methods Supporting Compliance with the Control Room Design Criteria As stated in 10 CFR 50, Appendix A, the GDCs establish minimum necessary design, fabrication, construction, testing, and performance requirements for SSCs that provide reasonable assurance that the facility can be operated without undue risk to public health and safety. Nuclear steam supply system engineers, architect-engineers, and utility engineers use these criteria and other regulatory requirements in establishing the design basis for the facility to be constructed. Although some design criteria may be reflected in technical specifications, the GDC are not operational limits in themselves. The criteria require that adequate radiation protection be provided to allow access and occupancy during normal and accident conditions.

Architect-engineers and utility engineering personnel design the control room SSCs so that, during an accident, personnel can access and occupy the control room. In evaluating the adequacy of the design, designers consider the performance of the control room design by performing a series of pre-selected sequences that are characterized as design basis accidents.

During its review of the license application, the NRC staff reviews the design and the applicants design-basis accident analyses, performs any needed confirmatory analyses, and accepts or rejects the applicants control room design.

These control room design-basis accident analyses are not intended to be actual event sequences but, rather, are surrogates to enable deterministic evaluation of the performance of plant engineered safety features, such as the control room habitability systems. These analyses are intentionally conservative in order to address uncertainties in accident progression, fission product transport, and atmospheric dispersion and to provide desirable defense in depth. They do not consider many other layers of defense in depth, including the NRCs comprehensive radiation protection and emergency planning regulatory framework. As a result, the actual doses are expected to be significantly lower than the computed doses in most accident scenarios, since no credit is taken for radiation protection and emergency response programs such as ALARA practices and issuance of potassium-iodine tablets. In rare situations of an actual event, the accident sequence may not progress as modeled as the sequence used to assess the control room design adequacy (e.g., may involve multiple failures), resulting in a greater challenge to control room systems. If the challenge exceeds the design basis, radiation exposures in the control room may exceed those envisioned in the design bases of the control room habitability features. In such an event, the facility radiation protection and emergency response programs implement measures to minimize radiation exposures.

27 The analysis to demonstrate compliance is a traditional deterministic design-basis accident radiological consequence analysis. The regulation requires the analysis to assume a radionuclide release to the reactor containment associated with a substantial meltdown of the reactor core. The use of this postulated accidental release of radionuclides is an important feature of the regulatory practices and policies adopted by the NRC in pursuit of a defense-in-depth safety philosophy. This radionuclide mix, known as the accident source term, refers to the magnitude and mix of the radionuclides released from the fuel, expressed as fractions of the fission product inventory in the fuel, as well as their physical and chemical form and the timing of their release. This accident source term is often referred to as the in-containment source term developed from the maximum hypothetical accident. It is used to demonstrate compliance with various regulations such as 10 CFR 50.67; 10 CFR Part 100, Reactor Site Criteria; 10 CFR 50.34, Contents of application; technical information; and 10 CFR Part 52, Licenses, Certification, and Approvals for Nuclear Power Plants. The radiological consequence analyses used to demonstrate compliance divorce themselves from the severe accident sequences and frequencies from which they were derived and assume the containment remains intact and leaks at the design-basis leak rate. The source term that leaks from the containment is termed the radiological release to the environment. The magnitude of the radiological release to the environment can be estimated from the containment leak rate and the radionuclide inventory suspended in the containment atmosphere as a function of time. The radionuclide inventory suspended in the containment atmosphere depends on the amount released to the containment as well as the effectiveness of natural and engineered processes that lead to radionuclide deposition within containment.1 The accident source term for currently licensed plants was originally set forth in Regulatory Guide 1.3, Revision 2, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors, issued June 1974 (USNRC, 1974a), and Regulatory Guide 1.4, Revision 2, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors, issued June 1974 (USNRC, 1974b). The accident source term for these regulatory guides had been derived from the supporting technical basis document for 10 CFR Part 100, TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites, issued 1962, (USAEC, 1962). This accident source term was assumed to be deterministically released into containment without explanation. Shortly after the accident at the Three Mile Island Nuclear Generating Station in 1979, accident source term estimates under severe accident conditions became of great interest when it was observed that only relatively small amounts of iodine were released to the environment when compared with the amount predicted using the TID-14844 accident source term. This led several observers to claim that severe accident releases were much lower than previously estimated. The NRC, with support from the U.S. Department of Energy national laboratories, began major research efforts in the early 1980s to obtain a better understanding of fission-product transport and release mechanisms in light-water reactors under severe accident conditions. These efforts resulted in a number of updated accident source terms to ensure that various regulatory applications remain updated and reflect the latest 1

The use of this accident source term has not been confined to the evaluation of site suitability. Other regulatory applications include the basis for the (1) postaccident radiation environment for which safety-related equipment should be qualified, (2) postaccident habitability requirements for the control room and technical support center, and (3) postaccident sampling systems and accessibility.

28 scientific understandings as new information and technologies emerge (see (USNRC, 1995),

(SNL, 2011), and (SNL, 2023)). As discussed in NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, issued February 1995 (USNRC, 1995), in contrast to the TID-14844 accident source term and containment leakage release used for design-basis accidents, severe accident releases to the environment first arose in PRA. An example is the Reactor Safety Study (Rasmussen, 1975), which examined accident sequences that involved core melt and containments that could fail. These severe accident releases represent mechanistically determined best estimate releases to the environment, including estimates of failures of containment integrity. This is very different from the combination of the nonmechanistic release to containment postulated by TID-14844, coupled with the assumption of very limited containment leakage used for 10 CFR Part 100 siting calculations for design-basis accidents. The bases of these revised accident source terms stem from a broad range of postulated accident sequences thought to be significant risk contributors at the time of their development. The scope of these severe accident sequences accounts for a diverse set of postulated accident sequences that include principle contributors to CDF. Every scenario produces an accident progression characterized by a substantial core-melt accident and radionuclide releases to the containment structure commensurate with each accident release phase. The accident sequences comprise an accident initiating event, defined system availability/functionality, and prescribed component failures to address uncertainties in component failure.

The dose-receptor calculation currently used in traditional deterministic design-basis accident radiological consequence analyses computes human health effects in terms of rem TEDE. The calculation requires modeling a dose receptor with various attributes such as health effects from radiation, human performance, and available protective measures. The rem TEDE is the sum of the effective dose equivalent for external exposures and the committed effective dose equivalent for internal exposures. It adjusts the dose equivalent radiation exposure using a tissue weighting factor that represents the proportion of the risk of stochastic effects resulting from irradiation of an organ or tissue to the total risk of stochastic effects when the whole body is irradiated uniformly. The committed effective dose equivalent is a 50-year committed dose based on an initial intake of radioactive material used to estimate the stochastic health effect of cancer mortality. In other words, the dose is assigned to the first year of intake to estimate the increased probability of cancer-induced fatality after 50 years. The assume dose-receptor model for assessing control room habitability is based on the ICRP Task Group on Reference Man model (ICRP, 1975). The Reference Man model is a well-defined characterization of a hypothetical average man in terms of both anatomical and physiological parameters. This model was used to derive dose coefficients that convert an exposure to ionizing radiation to a committed effective dose equivalent (see (EPA, 1988) and (EPA, 1993)). The human performance of the control room dose receptor is modeled as an occupancy factor. The occupancy factor is used to estimate the amount of time an individual is present at a particular location, typically expressed as a fraction per day. The origin of this assumption is referenced from Murphy and Campe (Murphy, 1974). It is based on a summary of lessons learned following a review of over 50 control room designs that describe typical time-stepped occupancy factors.

As discussed above, the radiological consequence analyses do not consider many layers of defense in depth, including the NRCs comprehensive radiation protection and emergency planning regulatory framework. The philosophy is that the design of the control room and its

29 habitability systems provides for a short-sleeved, comfortable environment for the control room operators. Such an environment was perceived to facilitate operator response to normal and accident conditions and would minimize errors of omission or commission. The design of the control room would therefore not rely on limiting exposure through pre-assumed protective actions, such as evacuation, the use of respiratory equipment, or the issuing of potassium-iodine tablets.

3.2.3.2 Summary of the Suitability of the CDF Risk Metric for Risk-Informed Grading Of the practical alternatives available to perform risk-informed grading, CDF is the most appropriate for the purposes of developing a graded, risk-informed and performance-based framework. This is because CDF accounts for a broad range of accident scenarios and can encompass the risk relevant to the sequences considered for control room habitability. Since LERF is used to assess early fatalities, and deterministic radiological consequence analyses must consider the potential of exposure for the complete duration of the accident, early as well as late, LERF is not appropriate. One could argue that LRF is a more suitable risk metric if it is defined to capture all releases from the containment to the atmosphere. However, LRF was not selected as the metric as the Commission has not endorsed definition its definition for use and is not widely utilized in the nuclear industry. Even though the use of the CDF, as opposed to LRF, appears to overlook containment performance, many activities the licensees have implemented improve containment performance would likely mitigate that impact. Regulations governing nuclear power plant emergency planning (10 CFR 50.47, Emergency plans) and operator training (10 CFR Part 55, Operators Licenses) are designed to ensure readiness for a wide range of beyond-design-basis accident conditions. These regulations mandate comprehensive training programs and emergency response drills to equip plant operators and staff with the skills and knowledge to effectively manage the event while maintaining actual incurred exposure to ionizing radiation ALARA. (Appendix B reviews the integrated use of emergency response procedures and guidelines.) Furthermore, when granting relief beyond the 10-rem, the staff can consider the containment performance on a plant specific basis.

Furthermore, the accident source term is derived from a range of severe accident sequences that are the dominant contributors to CDF. The CDF risk metric, used as a surrogate for latent cancer fatality risk, is consistent with the rem TEDE used to estimate the stochastic health effect of cancer mortality. Additionally, control room habitability design primarily concerns the ability of personnel to maintain reactor safety during and after accidents, which aligns closely with the overarching goal of preventing core damage and mitigating radiological releases, captured by the CDF risk metric.

Based on the review of the state of PRA, various risk metrics, and the suitability assessment, the CDF risk metric has been used to define the x-axis of the framework.

The risk metric range is based on those presented in Regulatory Guide 1.174, as they are well established.

30

4. GRADED, RISK-INFORMED, AND PERFORMANCE-BASED FRAMEWORK The author applied the methodology to develop an example of a graded, risk-informed, performance-based control room design criterion. To do so, findings from the Background and Method sections were applied as follows:

design criteria: range from 10 to 25 rem TEDE risk metric: CDF range from 1E-4 to 1E-5 per reactor year The control room design criterion range of 10 to 25 rem TEDE was divided into four bins. The CDF risk-metric range is based on those presented Regulatory Guide 1.174. The CDF risk-metric criteria define the lower and upper boundaries for the four bins. Table 1 lists of the binned results. Figure 5 illustrates how each CDF bin is mapped to the range of dose-based criteria.

Table 1: Graded, Risk-Informed, and Performance-Based Approach for Control Room Design Criteria Using CDF Regulatory Guide 1.174 CDF Ranges Design Criteria (rem TEDE CDF 1E-5 25 1E-5 < CDF 5E-5 20 5E-5 < CDF 1E-4 15 CDF > 1E-4 (or no PRA) 10

31 Figure 5: Histogram of Graded, Risk-Informed, and Performance-Based Approach for Control Room Design Criteria Using CDF 4.1 Analysis and Results Using the framework presented above, the author performed a series of analyses to understand how the nuclear reactor fleet would align within each control room design criterion bin. This was done by developing a CDF dataset extracted from various regulatory initiatives. For the purposes of this analysis, the dataset represents PRAs of varying quality and understood risk profiles over several decades. Initiatives examined include the Individual Plant Examination (IPE) and Individual Plant Examination for External Events, representing understood risk profiles from the 1980s; the NRC Standardized Plant Analysis Risk models; several risk-informed applications; and license renewals with severe accident mitigation alternatives analyses (SAMAs).

Insights into the frameworks usefulness and flexibility are shown by using the information from the CDF dataset. This is done through three examples, each which has varying constraints:

(1)

Example 1Flexibility that encourages further safety improvements (2)

Example 2Examining modeling fidelity, quality, and inclusion of external hazards (3)

Example 3Examining differences among reactor designs Example 1Flexibility that encourages further safety improvements Using the framework, the author performed an analysis to understand how the operating nuclear reactor fleet would align within each control room design criterion bin over a period of time. The purpose was to assess whether the framework could provide flexibility by encouraging further safety improvements. The premise is that with safety enhancements made to the facility over an operating lifetime, flexibility with an assigned control room design criterion would be achieved

32 and rewarded.2 Example 1 included the following constraints:

Models assess internal events only.

IPE results represent the base-case model.

License renewal SAMA results represent major facility changes improving safety since the IPE submission.

The PRA quality of the SAMA model results is consistent with the approved guidance in NEI 05-01, Severe Accident Mitigation Alternatives (SAMA) Analysis, issued November 2005 (NEI, 2005).

The model would be the most recent internal events risk model that considers all plant changes implemented up to the date of the license renewal submittal, uses failure and unavailability data to the same date, and resolves industry peer review comments on previous revisions to the model.

The model includes major contributors, initiators, or accident classes. It includes contributions to CDF from station blackout (single unit and dual unit) and anticipated transient without scram events.

Example 1 filters both the IPE and license renewal SAMA CDF data into the graded, risk-informed, and performance-based framework. The IPE data represent the understood risk profiles from the 1980s. The SAMA analysis data represent contemporary plant risk profiles from more than 40 years of operations with continuous facility safety improvements. Table 2 gives the example 1 analysis results where the number of facilities are binned into the control room design criterion framework. Figure 6 illustrates the binning of these results.

Table 2: Analysis Result Comparing 1980s to Contemporary CDF Risk Profiles CDF Risk-Metric Bins IPE CDF Data 1980s SAMA CDF Data Contemporary Binned Percentile Binned Percentile CDF 1E-5 15 16%

47 51%

1E-5 < CDF 5E-5 42 46%

40 43%

5E-5 < CDF 1E-4 22 24%

3 3%

CDF > 1E-4 (or no PRA) 13 14%

2 2%

2 The safety enhancements made to the facilities over their operating lifetime was evident in that each plant reviewed had lowered its CDF from that of the IPE to what was seen in the SAMA.

33 Figure 6: Example 1Analysis Result Histogram Comparing 1980s to Contemporary CDF Risk Profiles The analysis results from example 1 reflect how licensees would benefit from the framework that credits facility safety improvements and results in their overall risk reduction over time.

Analysis results show that 51 percent of the fleet would now have a control room design criterion of 25 rem TEDE based on the contemporary CDF data as opposed to 16 percent of the fleet from the IPE data from the 1980s.

The NRC has documented many of the facility-specific improvements in the document series NUREG-1437, Generic Environmental Impact Statement for License Renewal of Nuclear Plants (LR GEIS). The agency published the latest LR GEIS (Revision 2) in draft form in February 2023 (USNRC, 2023d). It provides updated information from the NUREG-1437 supplements about the risk and environmental impacts of severe accidents caused by internal events. This information indicates that CDFs are significantly less than those forming the basis of the original 1996 LR GEIS. On average, internal event CDFs for pressurized-water reactors (PWRs) have decreased by about a factor of 4 and CDFs for boiling-water reactors (BWRs) have decreased by about a factor of 6 compared to the CDFs used in the 1996 LR GEIS.

Furthermore, the internal event accident frequencies have further decreased, as reported in recent risk-informed license amendment requests to the NRC.

Example 2Examining modeling fidelity, quality, and inclusion of external hazards Using the framework, the author performed an analysis to understand how the operating nuclear reactor fleet would align within each control room design criterion bin due to increased modeling fidelity and the inclusion of external event risk contributors. The purpose was to assess how the framework would provide flexibilities when also considering external hazards. The premise is that the Commission undertakes certain regulatory initiatives to enhance safety in response to unforeseen events, and that certain licensing initiatives are required to be consistent with specific risk-informed philosophies and meet a certain level of technical adequacy. Example 2 included the following constraints:

34 License renewal SAMA results represent the base-case model.

The TSTF-505 results are consistent with approved guidance in NEI 06-09-A.

Example 2 filters license renewal SAMA and TSTF-505 CDF data into the graded, risk-informed, and performance-based framework. The TSTF-505 PRA offers a higher quality assessment compared to a license renewal SAMA due to its specific focus on addressing a facilitys technical specifications and other regulatory requirements. Following the guidance for a TSTF-505 submittal results in a more comprehensive and rigorous analysis. Those licensees that do not have an approved TSTF-505 program would use the lowest control room design criterion bin of 10 rem TEDE. Table 3 provides the results of the example 2 analysis where the number of facilities are binned into the control room design criterion framework. Figure 7 illustrates the binning of these results.

Table 3: Example 2Analysis Result Comparing License Renewal SAMA to TSTF-505 CDF Risk Profiles Risk-Metric CDF Bins TSTF-505 Data SAMA CDF Data Binned Percentile Binned Percentile CDF 1E-5 0

0%

47 51%

1E-5 < CDF 5E-5 18 21%

40 43%

5E-5 < CDF 1E-4 40 47%

3 3%

CDF > 1E-4 (or no TSTF-505) 27 32%

2 2%

Figure 7: Example 2Analysis Results Histogram Comparing License Renewal SAMA to TSTF-505 CDF Risk Profiles

35 The analysis results from example 2 reflect differences in how licensees would benefit from varying degrees of PRA modeling fidelity, quality, and inclusion of external hazards. The results show that none of the fleet would fall within the 25 rem TEDE control room design criterion bin based on the TSTF-505 CDF data as opposed to 51 percent based on the license renewal SAMA data.

Example 3Examining differences among reactor designs Using the framework, the author performed an analysis to understand how the operating nuclear reactor fleet would align within each control room design criterion bin based on reactor design type. The purpose was to assess how the framework would provide flexibilities between the BWR and PWR designs. Example 3 included the following constraints:

This example used the TSTF-505 results.

This example filtered the TSTF-505 CDF by BWR and PWR categories into the framework.

Table 4 gives the results of the example 3 analysis in which the TSTF-505 CDF results are binned into the control room design criterion framework by reactor type. Figure 8 illustrates the binning of these results.

Table 4: Example 3 Analysis Results Examining Differences Between Reactor Designs Risk-Metric CDF Bins BWR TSTF-505 Data PWR TSTF-505 Data Binned Percentile Binned Percentile CDF 1E-5 0

0%

0 0%

1E-5 < CDF 5E-5 14 45%

4 7%

5E-5 < CDF 1E-4 11 35%

29 48%

CDF > 1E-4 (or no TSTF-505) 6 19%

28 46%

36 Figure 8: Example 3Analysis Results Examining Difference Between Reactor Designs The analysis results from example 3 reflect differences between reactor types and the number of licensees that have chosen to adopt a TSTF-505 program. First, relatively more licensees in the BWR fleet have adopted TSTF-505 than those in the PWR fleet. No facility would achieve a control room design criterion of 25 rem TEDE, as none had an all-hazard CDF less than 1E-5 per reactor. For those that have adopted TSTF-505, more BWRs than PWRs have all-hazard CDFs between 5E-5 and 1E-4, resulting in a control room design criterion of 25 rem TEDE.

37 CONCLUSION The methodology, analysis, and results presented are intended to demonstrate the viability of developing a graded, risk-informed framework for a traditionally prescriptive rule. In developing the method, the author considered and applied several key Commission PRA policies to propose a transition of these rules to a risk-informed regulatory framework. The purpose of this methodology is to develop a framework that enables a performance-based evaluation using traditional deterministic radiological consequence analysis methods within defined risk-informed boundaries. These boundaries are defined by acceptable radiation exposure guidelines for radiation workers during accident and emergency conditions and acceptable contemporary nuclear facility risk profiles using modern PRA methods. The intent of such a framework is to provide flexibility when determining how to meet an established acceptance criterion in a way that encourages and rewards improved outcomes. In practice, the method produces a framework that justifies a higher control room design criterion with a lower plant-specific risk metric.

Concerning PRA quality, development of a graded, risk-informed, and performance-based framework should leverage PRA models of the highest pedigree approved for use with TSTF-505 licensing applications. This would support regulatory decisions in this area where the goal is to make a sound safety decision based on technically defensible information. The models developed for TSTF-505 applications have the proper scope and technical attributes to give an appropriate level of confidence in the results for a wide range of hazards. The methodology is based on NEI 06-09-A, which evaluates and extends the completion times for required actions under technical specification limiting conditions for operation. This methodology is consistent with the philosophy in Regulatory Guide 1.174 and technical adequacy expectations for the PRA model in Regulatory Guide 1.200. Regulatory Guide 1.200 references the ASME PRA standard, RA-S-2009, for internal events at power. Acceptability of the model is determined with respect to scope, level of detail, conformance to consensus standard technical elements (i.e., technical robustness), and plant representation. Quantification of risk due to internal fire and other significant external events is also necessary for this application, through PRA or bounding methods. For each facility, an all hazard risk metric (CDF) is computed by summing all CDF results (e.g., internal, flood, fires, seismic, high winds).

Lastly, this method is designed to meet the intent of the rules for control room habitability. It is not intended to assess other layers of defense in depth, such as the robustness and reliability of emergency core cooling performance requirements, containment fragility, or radiation protection and emergency planning.

38 ACKNOWLEDGEMENTS Special thanks to the NRCs Christopher Hunter, Senior Reliability and Risk Analyst in the Office of Nuclear Regulatory Research, who develops and maintains a number of risk-related datasets and dashboards for research-related activities. This resource allowed for the efficient assessment of the framework. Additional thanks to Sunil Weerakkody, Senior Level Advisor for Probabilistic Risk Assessment, and Kristy Bucholtz, Reliability and Risk Analyst, in the Office of Nuclear Reactor Regulation their guidance, mentorship, and critical review.

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43

44 APPENDIX A Table A-1: Dataset of Individual Plant Examination (IPE), Severe Accident Mitigation Alternatives Analysis (SAMA), and TSTF-505 Core Damage Frequency (CDF)

Facility Type Internal Events IPE Internal Events SAMA TSTF-505 Data Internal Events CDF Internal Floods CDF Internal Fires CDF Seismic CDF High Winds CDF External Flood CDF Other External Events CDF CDF All Hazard Source Arkansas Nuclear One 1 PWR 4.67E-05 1.10E-05 6.50E-06 0.00E+00 4.60E-05 5.50E-06 5.80E-05 ML22356A249 Arkansas Nuclear One 2 PWR 3.40E-05 7.20E-06 8.50E-06 0.00E+00 4.40E-05 5.50E-06 5.80E-05 ML23095A281 Beaver Valley 1 PWR 2.10E-04 3.98E-06 Beaver Valley 2 PWR 1.90E-04 9.53E-06 Braidwood 1 PWR 2.74E-05 0.00E+00 7.40E-05 4.20E-06 7.82E-05 ML20037B221 Braidwood 2 PWR 2.70E-05 0.00E+00 7.70E-05 4.20E-06 8.12E-05 ML20037B221 Browns Ferry 1 BWR 0.00E+00 0.00E+00 3.50E-06 2.75E-05 5.87E-06 3.69E-05 ML22090A287 Browns Ferry 2 BWR 4.80E-05 2.60E-06 2.99E-06 3.19E-05 5.80E-06 4.07E-05 ML22090A287 Browns Ferry 3 BWR 0.00E+00 3.40E-06 3.34E-06 2.69E-05 6.23E-05 9.25E-05 ML22090A287 Brunswick 1 BWR 2.70E-05 4.19E-05 2.23E-06 1.71E-06 3.24E-05 2.81E-06 3.92E-05 ML22082A268 Brunswick 2 BWR 2.70E-05 4.19E-05 2.45E-06 1.38E-06 4.04E-05 2.81E-06 4.70E-05 ML22082A268 Byron 1 PWR 3.09E-05 2.58E-04 7.70E-05 0.00E+00 7.90E-06 8.49E-05 ML20037B221 Byron 2 PWR 3.09E-05 2.58E-04 7.80E-05 0.00E+00 7.90E-06 8.59E-05 ML20037B221 Callaway PWR 4.06E-05 0.00E+00 4.46E-06 6.52E-06 1.09E-05 4.01E-05 5.40E-06 6.74E-05 ML22301A007 Calvert Cliffs 1 PWR 2.40E-04 1.58E-06 9.50E-06 4.20E-05 1.10E-06 3.30E-07 5.29E-05 ML18270A130 Calvert Cliffs 2 PWR 2.40E-04 1.58E-06 9.60E-06 4.00E-05 1.10E-06 5.40E-07 5.12E-05 ML18270A130 Catawba 1 PWR 4.40E-05 4.70E-05

45 Facility Type Internal Events IPE Internal Events SAMA TSTF-505 Data Internal Events CDF Internal Floods CDF Internal Fires CDF Seismic CDF High Winds CDF External Flood CDF Other External Events CDF CDF All Hazard Source Catawba 2 PWR 4.40E-05 4.70E-05 Clinton BWR 2.60E-05 0.00E+00 3.30E-06 7.80E-05 6.40E-06 8.77E-05 ML21132A288 Columbia BWR 1.75E-05 0.00E+00 2.36E-06 4.06E-05 1.73E-05 6.03E-05 ML23013A081 Comanche Peak 1 PWR 5.72E-05 0.00E+00 1.10E-06 1.19E-07 5.62E-05 1.93E-06 3.86E-06 6.32E-05 ML22192A007 Comanche Peak 2 PWR 5.72E-05 0.00E+00 1.00E-06 1.39E-07 4.29E-05 1.93E-06 3.85E-06 4.98E-05 ML22192A007 Cooper BWR 7.97E-05 9.30E-06 D.C. Cook 1 PWR 6.26E-05 5.00E-05 D.C. Cook 2 PWR 6.30E-05 5.00E-05 Davis-Besse PWR 6.60E-05 0.00E+00 Diablo Canyon 1 PWR 8.80E-05 0.00E+00 Diablo Canyon 2 PWR 8.80E-05 0.00E+00 Dresden 2 BWR 1.85E-05 1.90E-06 Dresden 3 BWR 1.90E-05 1.90E-06 Farley 1 PWR 1.30E-04 3.35E-05 8.91E-06 8.35E-05 4.51E-06 9.69E-05 ML19175A243 Farley 2 PWR 1.30E-04 3.35E-05 8.76E-06 7.89E-05 4.51E-06 9.22E-05 ML19175A243 Fermi BWR 5.70E-06 3.35E-05 Fitzpatrick BWR 1.92E-06 2.74E-06 3.20E-06 1.90E-05 2.50E-06 2.47E-05 ML22223A141 Ginna PWR 8.23E-05 2.00E-05 7.50E-06 3.80E-05 3.40E-06 4.89E-05 ML22119A094 Grand Gulf BWR 1.72E-05 0.00E+00 2.12E-06 4.65E-07 2.22E-05 9.31E-07 2.57E-05 ML23158A043 Hatch 1 BWR 2.10E-05 1.60E-05 4.54E-06 3.97E-07 5.65E-05 1.18E-06 6.26E-05 ML22297A146 Hatch 2 BWR 2.20E-05 1.60E-05 4.93E-06 2.13E-07 4.98E-05 1.18E-06 5.61E-05 ML22297A146

46 Facility Type Internal Events IPE Internal Events SAMA TSTF-505 Data Internal Events CDF Internal Floods CDF Internal Fires CDF Seismic CDF High Winds CDF External Flood CDF Other External Events CDF CDF All Hazard Source Hope Creek BWR 4.58E-05 5.10E-06 LaSalle 1 BWR 4.40E-05 0.00E+00 1.30E-06 1.00E-05 1.10E-05 2.23E-05 ML21162A069 LaSalle 2 BWR 4.40E-05 0.00E+00 1.30E-06 7.80E-06 1.10E-05 2.01E-05 ML21162A069 Limerick 1 BWR 4.30E-06 0.00E+00 3.20E-06 5.20E-06 3.70E-06 1.21E-05 ML20034F637 Limerick 2 BWR 4.30E-06 0.00E+00 3.20E-06 5.20E-06 3.70E-06 1.21E-05 ML20034F637 McGuire 1 PWR 4.00E-05 2.80E-05 3.14E-06 4.86E-06 3.37E-05 2.85E-05 3.02E-06 7.32E-05 ML23047A465 McGuire 2 PWR 4.00E-05 2.80E-05 3.16E-06 6.38E-06 4.06E-05 2.85E-05 3.13E-06 8.18E-05 ML23047A465 Millstone 2 PWR 3.40E-05 7.17E-05 Millstone 3 PWR 5.65E-07 2.57E-05 Monticello BWR 2.60E-05 4.50E-05 6.54E-06 5.75E-05 3.00E-05 9.40E-05 ML21148A274 Nine Mile Point 1 BWR 5.50E-06 1.30E-05 1.20E-06 2.60E-05 1.60E-06 5.00E-06 3.38E-05 ML22349A108 Nine Mile Point 2 BWR 3.10E-05 5.80E-05 1.80E-06 3.10E-05 6.40E-07 0.00E+00 3.34E-05 ML21082A221 North Anna 1 PWR 6.80E-05 3.50E-05 North Anna 2 PWR 6.80E-05 3.50E-05 Oconee 1 PWR 2.30E-05 2.60E-05 Oconee 2 PWR 2.30E-05 2.60E-05 Oconee 3 PWR 2.30E-05 2.60E-05 Palo Verde 1 PWR 9.00E-05 5.07E-06 1.36E-06 4.10E-07 2.80E-05 2.80E-05 5.78E-05 ML19085A525 Palo Verde 2 PWR 9.00E-05 5.07E-06 1.36E-06 4.10E-07 2.80E-05 2.80E-05 5.78E-05 ML19085A525 Palo Verde 3 PWR 9.00E-05 5.07E-06 1.36E-06 4.10E-07 2.80E-05 2.80E-05 5.78E-05 ML19085A525 Peach Bottom 2 BWR 5.53E-06 4.50E-06 3.40E-06 2.80E-05 2.20E-05 5.34E-05 ML21074A411

47 Facility Type Internal Events IPE Internal Events SAMA TSTF-505 Data Internal Events CDF Internal Floods CDF Internal Fires CDF Seismic CDF High Winds CDF External Flood CDF Other External Events CDF CDF All Hazard Source Peach Bottom 3 BWR 5.50E-06 4.50E-06 3.90E-06 4.00E-05 2.20E-05 6.59E-05 ML21074A411 Perry BWR 1.32E-05 0.00E+00 Point Beach 1 PWR 1.15E-04 3.59E-05 2.36E-06 4.33E-07 5.77E-05 6.24E-06 1.23E-06 6.80E-05 ML22140A131 Point Beach 2 PWR 1.15E-04 3.59E-05 2.30E-06 4.68E-07 6.84E-05 6.24E-06 1.24E-06 7.86E-05 ML22140A131 Prairie Island 1 PWR 5.00E-05 9.79E-06 1.36E-05 6.65E-05 4.88E-07 8.06E-05 ML20346A020 Prairie Island 2 PWR 5.00E-05 1.21E-05 1.27E-05 6.63E-05 4.88E-07 7.95E-05 ML20346A020 Quad Cities 1 BWR 1.20E-06 2.20E-06 3.90E-06 3.80E-05 4.30E-06 4.62E-05 ML23159A249 Quad Cities 2 BWR 1.20E-06 2.20E-06 3.90E-06 4.30E-05 4.30E-06 5.12E-05 ML23159A249 River Bend BWR 1.87E-05 0.00E+00 1.09E-06 1.94E-06 1.47E-05 9.81E-07 1.87E-05 ML23058A215 Robinson PWR 3.20E-04 4.32E-05 Salem 1 PWR 5.20E-05 4.80E-05 Salem 2 PWR 5.50E-05 4.80E-05 Seabrook PWR 6.05E-05 0.00E+00 Sequoyah 1 PWR 1.70E-04 0.00E+00 4.88E-06 0.00E+00 6.21E-05 4.19E-06 7.12E-05 ML22210A118 Sequoyah 2 PWR 1.70E-04 0.00E+00 5.19E-06 0.00E+00 6.63E-05 3.95E-06 7.54E-05 ML22210A118 Shearon Harris PWR 7.00E-05 9.24E-06 2.86E-06 2.36E-06 3.20E-05 2.14E-06 3.94E-05 ML21047A314 South Texas 1 PWR 4.40E-05 0.00E+00 South Texas 2 PWR 4.40E-05 0.00E+00 St. Lucie 1 PWR 2.30E-05 2.86E-05 5.34E-06 8.58E-07 5.16E-05 3.49E-06 4.01E-06 6.53E-05 ML19113A099 St. Lucie 2 PWR 2.60E-05 2.43E-05 6.77E-06 8.98E-08 6.96E-05 3.49E-06 4.01E-06 8.40E-05 ML19113A099 Summer PWR 2.00E-04 5.60E-05

48 Facility Type Internal Events IPE Internal Events SAMA TSTF-505 Data Internal Events CDF Internal Floods CDF Internal Fires CDF Seismic CDF High Winds CDF External Flood CDF Other External Events CDF CDF All Hazard Source Surry 1 PWR 7.40E-05 3.80E-05 Surry 2 PWR 7.40E-05 3.80E-05 Susquehanna 1 BWR 8.80E-08 2.00E-06 1.20E-06 9.60E-07 5.00E-05 1.70E-06 5.39E-05 ML22200A062 Susquehanna 2 BWR 8.80E-08 2.00E-06 1.20E-06 4.50E-07 6.30E-05 1.70E-06 6.64E-05 ML22200A062 Turkey Point 3 PWR 1.00E-04 1.60E-05 7.18E-07 1.62E-07 8.66E-05 6.98E-07 8.82E-05 ML18270A429 Turkey Point 4 PWR 1.00E-04 1.60E-05 7.13E-07 1.13E-07 7.69E-05 6.98E-07 7.84E-05 ML18270A429 Vogtle 1 PWR 4.90E-05 1.55E-05 2.25E-05 5.22E-05 6.00E-06 8.07E-05 ML15127A669 Vogtle 2 PWR 4.90E-05 1.55E-05 2.25E-05 5.19E-05 6.00E-06 8.04E-05 ML15127A669 Waterford PWR 1.70E-05 0.00E+00 3.03E-06 1.49E-06 2.01E-05 3.22E-06 2.78E-05 ML22322A109 Watts Bar 1 PWR 3.30E-04 1.40E-05 Watts Bar 2 PWR 0.00E+00 1.40E-05 Wolf Creek PWR 4.20E-05 3.00E-05

49 APPENDIX B This appendix reviews the integrated use of emergency response procedures and guidelines.

The U.S. Nuclear Regulatory Commission (NRC) has established emergency planning regulations in Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities, to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, and planning standards for nuclear power reactors in 10 CFR 50.47, Emergency plans, for the purpose of providing reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency. The regulation at 10 CFR 50.47(b)(11) addresses control of radiological exposures in an emergency and states that the means for controlling radiological exposures shall include exposure guidelines consistent with U.S. Environmental Protection Agency (EPA)

Emergency Worker and Lifesaving Activity Protection Action Guides. The events that could result in control room radiation exposures comparable to the normal occupational exposure limit of 5 rem total effective dose equivalent in 10 CFR Part 20, Standards for Protection Against Radiation, would result in the activation of the facilitys emergency response plan and the emergency response organization. Regulatory Guide 1.101, Revision 6, Emergency Response Planning and Preparedness for Nuclear Power Reactors, issued June 2021 (USNRC, 2021),

endorses NUREG-0654/FEMA-REP-1, Revision 2, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, issued December 2019 (USNRC/USFEMA, 2019). NUREG-0654/FEMA-REP-1 provides specific acceptance criteria for complying with the standards set forth in 10 CFR 50.47, which designates an on-shift emergency coordinator. The on-shift emergency coordinator has the authority and responsibility to immediately and unilaterally initiate any emergency actions.

These emergency actions include establishing higher exposure limits for control room operators if necessary to provide public health and safety. The emergency coordinator can also authorize issuance of potassium-iodide tablets for thyroid protection or use of emergency respiratory protection equipment.

Under severe accident conditions, the Commissions framework also addresses the integrated use of emergency response procedures and guidelines in such a way that they work together to implement the best available strategy for preventing or mitigating fuel damage and limiting radiological releases of a beyond-design-basis event and severe accident. The assessment of operator performance under accident conditions falls within a different regulatory area than the assessment of the design of the control room habitability envelope. Specifically, operator performance, training, qualifications, and proficiency maintenance are under the requirements of 10 CFR Part 55, Operators Licenses. This includes 10 CFR 55.41, Written examination:

Operators; 10 CFR 55.43, Written examination: Senior operators; 10 CFR 55.45, Operating tests, which controls the ability to implement the emergency operator procedures (EOPs); and 10 CFR 55.59, Requalification, which requires licensed operators to maintain their knowledge of the EOPs through requalification.

Figure 9 illustrates this framework, in which various aspects have been developed separately

50 over time and subject to varying levels of regulatory requirements and industry commitments, as are the training, drills, and exercises intended to maintain the capability for effective implementation.

Figure 9: Control Room Design Criteria of 10 CFR 50.67 and General Design Criterion 19, Control room: Typical Role of Accident Management Guidelines The left side of Figure 9 presents the various procedures and guidelines, stacked in order of event severity. The right side of Figure 9 illustrates how these procedures and guidelines are implemented during various plant conditions. The procedures are documents written as sequential instructions for performing a function or addressing plant conditions. Operators and plant staff are expected to follow the prescribed instructions in a step-by-step and verbatim manner. The guidelines do not necessarily provide a prescribed set of instructions and may not be followed in a step-by-step manner. Rather, they give suggested strategies and implementing methods that may be used to address an adverse event or condition, typically those beyond a plants design basis.

Abnormal operating procedures direct operator actions for restoring a function, system, or component to normal operating conditions following a transient or event. The abnormal operating procedures may also be used to mitigate an event or condition that is not severe enough to require use of an EOP, such as primary system leakage.

The EOPs direct operator actions for mitigating the consequences of transients and accidents that cause plant parameters to exceed actuation setpoints for the reactor protection system or

51 engineered safety features. These procedures are developed using guidelines set forth by the applicable owners group in response to Item I.C.1, Guidance for the Evaluation and Development of Procedures for Transients and Accidents, in NUREG-0737, Clarification of TMI Action Plan Requirements, issued November 1980 (USNRC, 1980).

The severe accident management guidelines (SAMGs) come from lessons learned from the accident at Three Mile Island Nuclear Generating Station in 1979, and the nuclear industry developed them as a voluntary initiative. The SAMGs describe additional strategies meant to provide operators and the plant staff with the capability to manage accident sequences that progress beyond the capacity of the mitigating strategies contained in the EOPs (e.g., adequate core cooling cannot be maintained). In doing so, the strategys focus changed from preventing fuel damage to mitigating the consequences of fuel damage, including minimizing radiological releases and protecting personnel. As indicated in Generic Letter 88-20, Individual Plant Examination for Severe Accident Vulnerabilities10 CFR 50.54(f), dated November 23, 1988, there is no regulatory requirement to develop, maintain, train, drill, or exercise SAMGs. (NRC U.

, 1988) Following the accident at the Fukushima Daiichi nuclear power plant, both owners groups updated SAMGs/severe accident guidelines, licensees incorporated them, and the NRC updated the Reactor Oversight Process for inspectors to verify that licensees keep them updated.

FLEX support guidelines provide strategies relying upon the use of installed and portable equipment and resources to maintain or restore core cooling, containment function, and spent fuel pool cooling capabilities during beyond-design-basis events. The strategies and capabilities reflected in these guidelines address the requirements of NRC Order EA-12-049, issued 2012, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design Basis External Events. (USNRC, 2012) Guidance concerning the development of FLEX support guidelines is contained in Nuclear Energy Institute (NEI), issued December 2016, 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide. (NEI, 2016)

Following the terrorist attacks of September 11, 2001, the NRC ordered licensees to develop strategies and specific implementing guidance for maintaining or restoring core cooling, containment, and spent fuel pool cooling capabilities under the circumstances associated with the loss of large areas of the plant due to explosions or fire. The agency subsequently imposed these extensive damage mitigation requirements as license conditions for individual licensees and then made them generically applicable under 10 CFR 50.54(hh)(2) through the Power Reactor Security Requirements final rule (74 FR 13926; March 27, 2009). Guidance concerning the development of extensive damage mitigation guidelines is contained in the NEI document, issued December 2006, in NEI 06-12, B.5.b Phase 2 & 3 Submittal Guideline. (NEI, 2006b)