ML081200566
| ML081200566 | |
| Person / Time | |
|---|---|
| Issue date: | 06/01/2008 |
| From: | Office of Nuclear Regulatory Research |
| To: | |
| Ridgely J, RES/DE,301-415-6555 | |
| Shared Package | |
| ML081200560 | List: |
| References | |
| DG-1200 RG-1.200 | |
| Download: ML081200566 (200) | |
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U.S. NUCLEAR REGULATORY COMMISSION June 2008 OFFICE OF NUCLEAR REGULATORY RESEARCH Division 1 DRAFT REGULATORY GUIDE
Contact:
M. Drouin (301) 415-6675 DRAFT REGULATORY GUIDE DG-1200 (Proposed Revision 2 of Regulatory Guide 1.200, dated January 2007)
AN APPROACH FOR DETERMINING THE TECHNICAL ADEQUACY OF PROBABILISTIC RISK ASSESSMENT RESULTS FOR RISK-INFORMED ACTIVITIES A. INTRODUCTION In 1995, the U.S. Nuclear Regulatory Commission (NRC) issued a Policy Statement (Ref. 1) on the use of probabilistic risk analysis (PRA), encouraging its use in all regulatory matters. That Policy Statement states that the use of PRA technology should be increased 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. Since that time, many uses have been implemented or undertaken, including modification of the NRCs reactor safety inspection program and initiation of work to modify reactor safety regulations.
Consequently, confidence in the information derived from a PRA is an important issue, in that the accuracy of the technical content must be sufficient to justify the specific results and insights that are used to support the decision under consideration.
This regulatory guide describes one acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water reactors. This guidance is intended to be consistent with the NRCs PRA Policy Statement. It is also intended to reflect and endorse guidance provided by standards-setting and nuclear industry organizations.
When used in support of an application, this regulatory guide will obviate the need for an in-This regulatory guide is being issued in draft form to involve the public in the early stages of the development of a regulatory position in this area. It has not received final staff review or approval and does not represent an official NRC final staff position.
Public comments are being solicited on this draft guide (including any implementation schedule) and its associated regulatory analysis or value/impact statement. Comments should be accompanied by appropriate supporting data. Written comments may be submitted to the Rulemaking, Directives, and Editing Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; emailed to NRCREP@nrc.gov; submitted through the NRCs interactive rulemaking Web page at http://www.nrc.gov; faxed to (301) 415-5144; or hand-delivered to Rulemaking, Directives, and Editing Branch, Office of Administration, US NRC, 11555 Rockville Pike, Rockville, Maryland 20852, between 7:30 a.m. and 4:15 p.m. on Federal workdays. Copies of comments received may be examined at the NRCs Public Document Room, 11555 Rockville Pike, Rockville, MD. Comments will be most helpful if received by August 25, 2008.
Electronic copies of this draft regulatory guide are available through the NRCs interactive rulemaking Web page (see above); the NRCs public Web site under Draft Regulatory Guides in the Regulatory Guides document collection of the NRCs Electronic Reading Room at http://www.nrc.gov/reading-rm/doc-collections/; and the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML081200566.
DG 1200, Page 2 depth review of the base PRA by NRC reviewers, allowing them to focus their review on key assumptions and areas identified by peer reviewers as being of concern and relevant to the application.
Consequently, this guide will provide for a more focused and consistent review process. In this regulatory guide, the quality of a PRA analysis used to support an application is measured in terms of its appropriateness with respect to scope, level of detail, and technical acceptability.
The NRC issued this regulatory guide for trial use in February 2004, and five trial applications were conducted. The NRC subsequently revised this guide to incorporate the lessons learned from those pilot applications (Ref. 2). The NRC also revised the appendices to this regulatory guide to address the changes made in the professional society PRA standards and industry PRA guidance documents. The NRC then issued the revised guide (including its associated appendices) for public review and comment as Draft Guide-1161 in September 2006 (Ref. 3). The staff subsequently reviewed the stakeholder comments and, where appropriate, revised the guide accordingly. (See Ref. 4 for a list of stakeholder comments and the related staff resolutions).
This regulatory guide contains information collections that are covered by the requirements of 10 CFR Part 50 which the Office of Management and Budget (OMB) approved under OMB control number 3150-0011. The NRC may neither conduct nor sponsor, and a person is not required to respond to, an information collection request or requirement unless the requesting document displays a currently valid OMB control number.
DG 1200, Page 3 B. DISCUSSION Existing Guidance Related to the Use of PRA in Reactor Regulatory Activities Since the NRC issued its PRA Policy Statement, a number of risk-informed regulatory activities have been implemented and the necessary technical documents are being developed to provide guidance on the use of PRA information.
Regulatory Guide 1.174 (Ref. 5) and its associated standard review plan (SRP), Section 19 (Ref.
6), provide general guidance on applications that address changes to the licensing basis. Key aspects of this document include the following:
It describes a risk-informed integrated decision-making process that characterizes how risk information is used and, more specifically, it clarifies that such information is one element of the decision-making process. That is, decisions are expected to be reached in an integrated fashion, considering traditional engineering and risk information, and may be based on qualitative factors as well as quantitative analyses and information.
It reflects the staffs recognition that the PRA needed to support regulatory decisions can vary (i.e., that the scope, level of detail, and quality of the PRA is to be commensurate with the application for which it is intended and the role the PRA results play in the integrated decision process). For some applications and decisions, only particular parts1 of the PRA need to be used. In other applications, a full-scope PRA is needed. General guidance regarding scope, level of detail, and quality for a PRA is provided in the application-specific documents.
While this document is written in the context of one reactor regulatory activity (license amendments), the underlying philosophy and principles are applicable to a broad spectrum of reactor regulatory activities.
Regulatory Guide 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance (Ref. 7), discusses an approach to support the new rule established as Title 10, Section 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors, of the Code of Federal Regulations (10 CFR 50.69) (Ref. 8).
Regulatory Guide 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants (Ref. 9), provides guidance for use in complying with requirements that the NRC has promulgated for risk-informed and performance-based fire protection progress that meet the requirements of 10 CFR 50.48(c) (Ref. 10) and National Fire Protection Association 805, Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generating Plants, 2001 Edition (Ref. 11).
Section C.I.19 of Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants (LWR Edition) (Ref. 12), discusses the requirements in 10 CFR Part 52, Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Plants (Ref. 13), for a combined license (COL) applicant to conduct a plant-specific PRA and to describe the plant-specific PRA and its results within its final safety analysis report. The revision to 10 CFR Part 50 included a 1
In this regulatory guide, a part of a PRA can be understood to be equivalent to that piece of the analysis for which an applicable PRA standard identifies a supporting level requirement.
DG 1200, Page 4 requirement for the COL holder to maintain and upgrade the PRA periodically throughout the life of the plant, and a requirement to demonstrate PRA technical adequacy.
In addition, there are other regulatory guides that provide guidance for such specific applications as inservice testing (Ref. 14), inservice inspection (Ref. 15), and technical specifications (Ref. 16). The NRC has also prepared SRP sections for each of the application-specific regulatory guides.
PRA standards have also been under development by the American Society of Mechanical Engineers (ASME) and the American Nuclear Society (ANS):
On April 5, 2002, ASME issued a standard for an at-power, internal events (excluding fire) Level 1 PRA and a limited Level 2 PRA for operating light water reactors (LWRs). ASME subsequently issued Addenda A, B, and C to that standard on December 5, 2003, December 30, 2005, and July 6, 2007, respectively (Ref. 17).
In December 2003, ANS issued a standard for external events and issued Revision 1 on March 1, 2007 for operating LWRs (Ref 18).
In November 2007, ANS issued a standard for internal fire events for operating LWRs (Ref. 19).
On April 9, 2008, ASME and ANS jointly issued a Level 1 and limited Level 2 PRA standard for internal and external events for at-power conditions (requirements for low power shutdown conditions to be added) (Ref 20).
ASME is developing PRA standards for new LWRs applying for design certification (DC) and COLs, for future advanced non-LWRs. ANS is developing a Level 1 and limited Level 2 PRA standard for low-power shutdown operating mode (to be incorporated into the ASME/ANS joint standard), and is also developing Level 2 and Level 3 PRA standards.
Reactor owners groups have been developing and applying a PRA peer review program for several years. The Nuclear Energy Institute (NEI) has issued several peer review guidance documents:
On August 16, 2002, NEI submitted draft industry guidance for self-assessments (Ref. 21) to address the use of industry peer review results in demonstrating conformance with the ASME PRA Standard. This additional guidance, which is intended to be incorporated into a revision of NEI 00-02 (per NEI) (Ref. 22), contains the following:
- Self-assessment guidance document
- Appendix 1 actions for industry self-assessment
- Appendix 2 industry peer review subtier criteria On May 19, 2006, NEI issued a revision to the self-assessment guidance incorporated in NEI 02 (Ref. 23), to satisfy the peer review requirement(s) of the ASME PRA Standard (ASME-RA-Sa-2003 (Ref. 24)) as endorsed/modified by the NRC and updated by Addendum B of the ASME PRA Standard.
In August 2006, NEI issued NEI 05-04, Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard. This document provides guidance for conducting and documenting a follow-on peer review for PRAs using the ASME PRA Standard.
DG 1200, Page 5 In November 2006, NEI updated the self-assessment guidance in Revision 1 of NEI 00-02 (Ref.
- 25) to address the staff objections raised in Appendix B of DG-1161.
In December 2007, NEI updated the process for follow-on peer reviews in Revision 1 of NEI 05-04 (Ref. 26).
In December 2007, NEI issued Revision 0 of NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, (Ref. 27).
SECY-00-0162 (Ref. 28) describes an approach for addressing PRA quality in risk-informed activities, including identification of the scope and minimal functional attributes of a technically acceptable PRA.
SECY-04-0118, Plan for the Implementation of the Commissions Phased Approach to PRA Quality (Ref. 29), presents the staffs approach to defining the needed PRA quality for current or anticipated applications, as well as the process for achieving this quality, while allowing risk-informed decisions to be made using currently available methods until all of the necessary guidance documents are developed and implemented. SECY-07-0042, Status of the Plan for the Implementation of the Commission's Phased Approach to Probabilistic Risk Assessment Quality (Ref. 30), provides an update to the staff plan.
Purposes of this Regulatory Guide The purpose of this regulatory guide is: a) to provide guidance to licensees for use in determining the technical adequacy of the base PRA used in a risk-informed regulatory activity, and b) to endorse standards and industry peer review guidance. Toward that end, this regulatory guide provides guidance in four areas:
(1) a definition of a technically acceptable PRA (2) the NRCs position on PRA consensus standards and industry PRA peer review program documents (3) demonstration that the baseline PRA (in total or specific parts) used in regulatory applications is of sufficient technical adequacy (4) documentation to support a regulatory submittal This regulatory guide provides guidance on PRA technical adequacy needed for the base PRA that is used in a risk-informed integrated decision-making process. It does not provide guidance on how the base PRA is revised for a specific application or how the PRA results are used in application-specific decision-making processes; that guidance is provided in such documents as References 5, 7, 9, and 12.
The regulatory guides that address specific applications, such as RG 1.201, allow for the use of PRAs that are not full-scope (e.g., do not include contributions from external initiating events or low-power and shutdown modes of operation). Those regulatory guides do, however, state that the missing scope items are to be addressed in some way, such as by using bounding analyses, or limiting the scope of the application. This regulatory guide does not address such alternative methods to the evaluation of risk contributions; rather, this guide only addresses PRA methods. NUREG-1855 provides guidance on acceptable bounding analyses (Ref. 31).
DG 1200, Page 6 Relationship to Other Guidance Documents This regulatory guide is a supporting document to other NRC regulatory guides that address risk-informed activities. At a minimum, these guides include (1) Regulatory Guide 1.174 and SRP Section 19, which provide general guidance on applications that address changes to the licensing basis; (2) the regulatory guides for specific applications such as for inservice testing, inservice inspection, and technical specifications (Refs. 14-16); and (3) regulatory guides associated with implementation of certain regulations, particularly those that rely on a plant-specific PRA to implement the rule (e.g., 10 CFR Part 52). In addition, the NRC has prepared corresponding SRP chapters for the application-specific guides.
Figure 1 shows the relationship of this new regulatory guide to risk-informed activities, application-specific guidance, consensus PRA standards, and industry programs (e.g., NEI 00-02, 05-04, 07-12).
Licensing Risk-informed licensing changes 50.48(c)
Fire Protection, National Fire Protection Association Standard NFPA 805 50-69 Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors 10 CFR Part 52 Licenses, certifications, and approvals for nuclear power plants Regulatory Guide 1.174 Regulatory Guide 1.205 Regulatory Guide 1.201 Regulatory Guide 1.206 Regulatory Guide 1.200 PRA Standards and Related Industry Programs APPLICATION APPLICATION-SPECIFIC REGULATORY GUIDE GENERIC SUPPORTING REGULATORY GUIDE Figure 1. Relationship of Regulatory Guide 1.200 to Other Risk-Informed Guidance
DG 1200, Page 7 C. REGULATORY POSITION
- 1.
A Technically Acceptable PRA This section describes one acceptable approach for defining the technical adequacy of an acceptable base PRA of a commercial light water reactor nuclear power plant. However, the term PRA needs to be defined. For a method or approach to be considered a PRA, the method or approach (1) provides a quantitative assessment of the identified risk, for example, a quantification (e.g., frequency) of core damage or a large early release, and (2) is comprised of specific technical elements in performing the quantification. A method that does not provide a quantified assessment or does not include the specified technical elements is not considered to be a PRA.
PRAs used in risk-informed activities may vary in scope and level of detail, depending on the specific application. However, the PRA results used to support an application must be derived from a baseline PRA model that represents the as-built, as-operated plant2 to the extent needed to support the application.
In this section, the guidance provided is for a full-scope Level 1 and Level 2 PRA. The scope is defined in terms of (1) the metrics used to characterize risk, (2) the plant operating states for which the risk is to be evaluated, and (3) the causes of initiating events (hazard groups) that can potentially challenge and disrupt the normal operation of the plant and, if not prevented or mitigated, would eventually result in core damage and/or a large release.
The level of detail required of the PRA model is ultimately determined by its intended use.
Nonetheless, a minimal level of detail is necessary to ensure that the impacts of designed-in dependencies (e.g., support system dependencies, functional dependencies, and dependencies on operator actions) are correctly captured and the PRA represents the as-built, as-operated plant. This minimal level of detail is implicit in the technical characteristics and attributes discussed in this section. Consequently, this section provides guidance in four areas:
(1) definition of the scope of a PRA (2) technical elements of a full-scope PRA (3) attributes and characteristics for technical elements of a PRA (4) development, maintenance, and upgrade of a PRA 1.1 Scope of PRA The scope of a PRA is defined by the challenges included in the analysis and the level of analysis performed. Specifically, the scope is defined in the following terms:
metrics used in characterizing the risk plant operating states for which the risk is to be evaluated causes of initiating events (hazard groups) that can potentially challenge and disrupt the normal operation of the plant 2
Some applications may involve the plant at the DC or COL application stage, at which point the plant is neither built nor operated. At these stages, the intent is for the PRA model to reflect the as-designed plant.
DG 1200, Page 8 Risk characterization is typically expressed by metrics of core damage frequency (CDF) and large early release frequency (LERF) (as surrogates for latent and early fatality risks, respectively, for operating light-water reactors). Large release frequency (LRF) is used as a risk metric for LWR DC and COL applicants. These metrics are defined in a functional sense as follows:
Core damage frequency is defined as the sum of the frequencies of those accidents that result in uncovery and heatup of the reactor core to the point at which prolonged oxidation and severe fuel damage involving a large fraction of the core (i.e., sufficient, if released from containment, to have the potential for causing offsite health effects) is anticipated.
Large early release frequency is defined as the sum of the frequencies of those accidents leading to rapid, unmitigated release of airborne fission products from the containment to the environment occurring before the effective implementation of offsite emergency response and protective actions such that there is the potential for early health effects. (Such accidents generally include unscrubbed releases associated with early containment failure shortly after vessel breach, containment bypass events, and loss of containment isolation.)
Large release frequency is defined as the sum of the frequencies of those accidents where the characteristics of the release (1) have the potential for early health effects (independent of timing or evacuation), or (2) may lead to a statistically significant increase in latent heath effects.
Issues related to the reliability of barriers (in particular, containment integrity and consequence mitigation) are addressed through other parts of the decision-making process, such as consideration of defense-in-depth. To provide the risk perspective for use in decision-making, a Level 1 PRA is required to provide CDF. A limited Level 2 PRA is needed to address LERF and a full Level 2 to address LRF.
Plant operating states (POSs) are used to subdivide the plant operating cycle into unique states, such that the plant response can be assumed to be the same within the given POS for a given initiating event. Operational characteristics (such as reactor power level; in-vessel temperature, pressure, and coolant level; equipment operability; and changes in decay heat load or plant conditions that allow new success criteria or reactor coolant system or containment configuration) are examined to identify those relevant to defining POSs. These characteristics are used to define the states, and the fraction of time spent in each state is estimated using plant-specific information. The risk perspective is based on the total risk associated with the operation of the reactor, which includes not only at-power operation, but also low-power and shutdown conditions. For some applications, the risk impact may affect some modes of operation, but not others.
Initiating events are the plant system perturbations to the steady state of the plant that challenge plant control and safety systems whose failure could lead to core damage and or radioactivity release. These initiating events include failure of equipment from either internal plant causes (such as hardware faults, operator actions, floods, or fires), or external plant causes (such as earthquakes or high winds). These are sometimes referred to as internal initiating events and external initiating events respectively. The internal initiating events can further be classified as either internal hardware events (such as transients and loss of coolant accidents) or internal area events (such as floods and fires). The risk perspective is based on a consideration of the total risk, which includes contributions from initiating events whose causes are attributable to both internal and external sources.
A hazard group is a group of similar causes of initiating events that are assessed in a PRA using a common approach, methods, and likelihood data for characterizing the effect on the
DG 1200, Page 9 plant. Typical hazard groups for a nuclear power plant PRA include internal hardware events (e.g., transients and LOCAs), seismic events, internal fires, internal floods, and high winds.
1.2 Technical Elements of PRA Table 1 provides the list of general technical elements that are necessary for a PRA. A PRA that is missing one or more of these elements would not be considered a complete PRA. The following briefly discusses the objective of each element.
Table 1. Technical Elements of a PRA SCOPE OF ANALYSIS TECHNICAL ELEMENT Level 1 Initiating event analysis Success criteria analysis Accident sequence analysis Systems analysis Parameter estimation analysis Human reliability analysis Quantification Level 2 Plant damage state analysis Accident progression analysis Quantification Source term analysis Interpretation of results and documentation are elements of both Level 1 and Level 2 PRAs.
These technical elements are equally applicable to the PRA models constructed to address each of the contributors to risk (i.e., internal and external initiating events) for each of the POSs. Because additional analyses are required to characterize their impact on the plant in terms of causing initiating events and mitigating equipment failures, internal floods, internal fires, and external hazards are discussed separately in Regulatory Positions 1.2.3, 1.2.4, and 1.2.5, respectively. Further, to understand the results, it is important to examine the different contributors on both an individual and relative basis. Therefore, this element, interpretation of results, is discussed separately in Regulatory Position 1.2.6. Another major element that is common to all of the technical elements is documentation; it is also discussed separately in Regulatory Position 1.2.7. While the technical elements are the same for each POS, within a technical element, other considerations may need to be addressed for low-power and shutdown conditions. These considerations are discussed under Regulatory Position 1.3.
1.2.1 Level 1 Technical Elements Initiating event analysis identifies and characterizes the events that both challenge normal plant operation during power or shutdown conditions and require successful mitigation by plant equipment and personnel to prevent core damage from occurring. Events that have occurred at the plant and those that have a reasonable probability of occurring are identified and characterized. An understanding of the nature of the events is performed such that a grouping of the events into event classes, with the classes defined by similarity of system and plant responses (based on the success criteria), may be performed to manage the large number of potential events that can challenge the plant.
Success criteria analysis determines the minimum requirements for each function (and ultimately the systems used to perform the functions) to prevent core damage (or to mitigate a release) given an initiating event. The requirements defining the success criteria are based on acceptable engineering analyses that represent the design and operation of the plant under consideration. For a function to be successful, the criteria are dependent on the initiator and the conditions created by the initiator. The computer codes used to perform the analyses for developing the success criteria are validated and verified for both technical integrity and suitability to assess plant conditions for the reactor
DG 1200, Page 10 pressure, temperature, and flow range of interest, and they accurately analyze the phenomena of interest.
Calculations are performed by personnel who are qualified to perform the types of analyses of interest and are well trained in the use of the codes.
Accident sequence analysis models, chronologically (to the extent practical), the different possible progressions of events (i.e., accident sequences) that can occur from the start of the initiating event to either successful mitigation or core damage. The accident sequences account for the systems that are used (and available) and operator actions performed to mitigate the initiator based on the defined success criteria and plant operating procedures (e.g., plant emergency and abnormal operating procedures) and training. The availability of a system includes consideration of the functional, phenomenological, and operational dependencies and interfaces between the various systems and operator actions during the course of the accident progression.
Systems analysis identifies the various combinations of failures that can prevent the system from performing its function as defined by the success criteria. The model representing the various failure combinations includes, from an as-built and as-operated perspective, the system hardware and instrumentation (and their associated failure modes) and human failure events that would prevent the system from performing its defined function. The basic events representing equipment and human failures are developed in sufficient detail in the model to account for dependencies among the various systems and to distinguish the specific equipment or human events that have a major impact on the systems ability to perform its function.
Parameter estimation analysis quantifies the frequencies of the initiating events, as well as the equipment failure probabilities and equipment unavailabilities of the modeled systems. The estimation process includes a mechanism for addressing uncertainties and has the ability to combine different sources of data in a coherent manner, including the actual operating history and experience of the plant when it is of sufficient quality, as well as applicable generic experience.
Human reliability analysis identifies and provides probabilities for the human failure events that can negatively impact normal or emergency plant operations. The human failure events associated with normal plant operation include the events that leave the system (as defined by the success criteria) in an unrevealed, unavailable state. The human failure events associated with emergency plant operation represent those human actions that, if not performed, do not allow the needed system to function.
Quantification of the probabilities of these human failure events is based on plant-and accident-specific conditions, where applicable, including any dependencies among actions and conditions.
Quantification provides an estimation of the CDF given the design and/or operation the plant (depending whether the plant is in the design or operating stage). Regardless of the plant stage, the CDF is based on the summation of the estimated CDF from each accident sequence for each initiator class. If
DG 1200, Page 11 truncation of accident sequences and cutsets is applied, truncation limits are set so that the overall model results are not impacted in such a way that significant accident sequences or contributors3 are eliminated.
Therefore, the truncation limit can vary for each accident sequence. Consequently, the truncation value is selected so that the accident sequence CDF is stable with respect to further reduction in the truncation value.
1.2.2 Level 2 Technical Elements Plant damage state analysis groups similar core damage scenarios together to allow a practical assessment of the severe accident progression and containment response resulting from the full spectrum of core damage accidents identified in the Level 1 analysis. The plant damage state analysis defines the attributes of the core damage scenarios that represent boundary conditions to the assessment of severe accidents progression and containment response that ultimately affect the resulting radionuclide releases.
The attributes address the dependencies between the containment systems modeled in the Level 2 analysis with the core damage accident sequence models to fully account for mutual dependencies. Core damage scenarios with similar attributes are grouped together to allow for efficient evaluation of the Level 2 response.
Accident progression analysis models the different series of events that challenge containment integrity for the core damage scenarios represented in the plant damage states. The accident progressions account for interactions among severe accident phenomena and system and human responses to identify credible containment failure modes, including failure to isolate the containment. The timing of major accident events and the subsequent loadings produced on the containment are evaluated against the capacity of the containment to withstand the potential challenges. The containment performance during the severe accident is characterized by the timing (e.g., early versus late), size (e.g., catastrophic versus bypass), and location of any containment failures. The codes used to perform the analysis are validated and verified for both technical integrity and suitability. Calculations are performed by personnel qualified to perform the types of analyses of interest and well-trained in the use of the codes.
Source term analysis characterizes the radiological release to the environment resulting from each severe accident sequence leading to containment failure or bypass. The characterization includes the time, elevation, and energy of the release and the amount, form, and size of the radioactive material that is released to the environment. The source term analysis is sufficient to determine whether a large early release or a large late release occurs. A large early release is one involving the rapid, unmitigated release of airborne fission products from the containment to the environment occurring before the effective implementation of offsite emergency response and protective actions such that there is a potential for early health effects. Such accidents generally include unscrubbed releases associated with early 3
The determination of significance is a function of how the PRA is being, or is intended to be, used. When a PRA is being used to support an application, the significance of an accident sequence or contributor is measured with respect to whether its consideration has an impact on the decision being made. For the base PRA model, significance can be measured with respect to the contribution to the total CDF or LERF, or it can be measured with respect to the contribution to the CDF or LERF for a specific hazard group or POS, depending on the context. For example, for the purposes of defining capability categories, the ASME/ANS PRA Standard, defines significance at the hazard group level. Whatever the context, the following numerical criteria are recommended:
Significant accident sequence: A significant sequence is one of the set of sequences, defined at the functional or systemic level that, when ranked, compose 95% of the CDF or the LERF/LRF, or that individually contribute more than ~1% to the CDF or LERF/LRF.
Significant basic event/contributor: The basic events (i.e., equipment unavailabilities and human failure events) that have a Fussell-Vesely importance greater than 0.005 or a risk-achievement worth greater than 2.
DG 1200, Page 12 containment failure at or shortly after vessel breach, containment bypass events, and loss of containment isolation. With large late release, unmitigated release from containment occurs in a time frame that allows effective evacuation of the close-in population such that early health effects are unlikely.
Quantification integrates the accident progression models and source term evaluation to provide estimates of the frequency of radionuclide releases that could be expected following the identified core damage accidents. This quantitative evaluation reflects the different magnitudes and timing of radionuclide releases and specifically allows for identification of the LERF or LRF.
1.2.3 Internal Floods Technical Elements PRA models of internal floods are based on the internal events PRA model, modified to include the impact of the identified flood scenarios in terms of causing initiating events, and failing equipment used to respond to initiating events. These flood scenarios are developed during the flood identification analysis and the flood evaluation analysis. The quantification task specific to internal floods is similar in nature to that for the internal events. Because of its dependence on the internal events model, the flooding analysis incorporates the elements of Sections 1.2.1 and 1.2.2, as necessary.
Flood identification analysis identifies the plant areas where flooding could result in significant accident sequences. Flooding areas are defined on the basis of physical barriers, mitigation features, and propagation pathways. For each flooding area, flood sources that are attributable to equipment (e.g.,
piping, valves, pumps) and other sources internal to the plant (e.g., tanks) are identified along with the affected structures, systems, and components (SSCs). Flooding mechanisms examined include failure modes of components, human-induced mechanisms, and other water-releasing events. Flooding types (e.g., leak, rupture, spray) and flood sizes are determined. Plant walkdowns are performed to verify the accuracy of the information. It is recognized that at the design and initial licensing stage, plant walkdowns are not possible.
Flood evaluation analysis identifies the potential flooding scenarios for each flood source by identifying flood propagation paths of water from the flood source to its accumulation point (e.g., pipe and cable penetrations, doors, stairwells, failure of doors or walls). Plant design features or operator actions that have the ability to terminate the flood are identified. The susceptibility of each SSC in a flood area to flood-induced mechanisms is examined (e.g., submergence, spray, pipe whip, and jet impingement). Flood scenarios are developed by examining the potential for propagation and giving credit for flood mitigation. Flood scenarios can be eliminated on the basis of screening criteria. The screening criteria used are well-defined and justified.
Quantification provides an estimation of the CDF of the plant that includes internal floods. The frequency of flooding-induced initiating events that represent the design, operation, and experience of the plant are quantified. The Level 1 models are modified and the internal flood accident sequences quantified to (1) modify accident sequence models to address flooding phenomena, (2) perform necessary calculations to determine success criteria for flooding mitigation, (3) perform parameter estimation analysis to include flooding as a failure mode, (4) perform human reliability analysis to account for performance shaping factors that are attributable to flooding, and (5) quantify internal flood accident sequence CDF.
1.2.4 Internal Fire Technical Elements PRA models of internal fires are based on the internal events PRA model, modified to include the impact of the identified fire scenarios in terms of causing initiating events (plant transients and loss-of-coolant accidents (LOCAs)) and failing equipment used to respond to initiating events. The incorporation
DG 1200, Page 13 of the set of fire scenarios into a fire PRA model is performed using a number of technical elements discussed below. Because of its dependence on the internal events model, the internal fire analysis incorporates the elements of Sections 1.2.1 and 1.2.2 of this guide as necessary.
Plant boundary definition and partitioning establishes the overall boundaries of the Fire PRA, and divides the area within that boundary into smaller regions (i.e., physical analysis units), commonly known as fire areas or compartments. The entire fire PRA is generally organized according to these physical analysis units.
Equipment selection identifies the equipment to be included in the fire PRA model. This equipment is selected from the equipment included in the internal events PRA and in the plants fire protection program and analysis (i.e., the postfire safe-shutdown analysis) that, if failed by a fire, could produce a plant initiator or affect the plant response. Fire-induced spurious actuations are of particular interest. The selected equipment is mapped to the physical analysis units.
Cable selection identifies those cables associated with the equipment identified in the Equipment Selection technical element. The selected cables are mapped to the physical analysis units.
Qualitative screening is an optional element that may be used to eliminate certain physical analysis units defined in the plant boundary definition and partitioning element that can be shown to be unimportant to fire risk. General, qualitative criteria are typically applied. Those physical analysis units screened out in this technical element play no role in the more detailed quantitative assessment.
Fire PRA plant response model develops a logic model that represents the plant response following a fire. This model is based upon the internal events PRA model which is modified to account for fire effects. These modifications include system, structure, and component failures that specifically result from fires and consider of fire-specific procedures. The latter are processed through the human reliability analysis technical element.
Fire scenario selection and analysis defines and analyzes fire event scenarios that capture the plant fire risk associated with each physical analysis unit. Fire scenarios are defined in terms of ignition sources, fire growth and propagation, fire detection, fire suppression, and cables and equipment (targets) damaged by the fire. Main control room fire scenarios, including control room abandonment, are analyzed explicitly. Multicompartment fire propagation scenarios, including scenarios from all screened physical analysis units, are also assessed.
Fire ignition frequencies are estimated for the ignition sources postulated for the fire scenarios.
Ignition sources consist of in situ sources, such as electrical cabinets or batteries, and other sources such as transient fires. U.S. nuclear power industry fire event frequencies, possibly augmented with plant-specific experience, are used where available to establish the fire ignition frequencies. Other sources are generally used only for cases when the U.S. nuclear power industry does not provide the representative frequency.
Quantitative screening involves eliminating physical analysis units from further quantitative analysis based on their quantitative contribution to fire risk. Quantitative screening criteria are established in terms of fire-induced CDF and LERF/LRF. This element is not required, although it is expected to be used in most applications. Note that, unlike the physical analysis units screened during qualitative screening, the CDF and LERF/LRF contributions of each of these quantitatively screened units are retained and reported as a part of the total plant fire risk in the fire risk quantification element. All physical analysis units are reconsidered as a part of the multicompartment fire scenario analysis, regardless of the quantitative screening results.
DG 1200, Page 14 Circuit failure analysis treats the impact of fire-induced circuit failures upon the plant response.
In particular, spurious actuations from hot shorts (inter-cable and intra-cable) are analyzed. The conditional probability of the particular circuit failure is identified and assigned.
Post-fire human reliability analysis is conducted to identify operator actions and related human failure events (HFEs), both within and outside the main control room, for inclusion in the plant response model. This element also includes quantification of human error probabilities for the modeled actions.
Modeled operator actions include those introduced into the plant response model resulting strictly from fire-related emergency procedures and those actions retained from the internal events PRA. The latter HFEs are modified to account for fire effects.
Fire risk quantification calculates the fire-induced CDF and LERF/LRF contributions to plant risk and identifies significant contributors to each. In this element, the plant response model is quantified for the set of fire scenarios to produce a conditional core damage probability and conditional large early release probability (CLERP) or conditional large release probability (CLRP) values. The conditional core damage probability and CLERP/CLRP values are mathematically combined with the corresponding fire ignition frequencies and the conditional probabilities of fire damage for the appropriate fire scenario to yield fire-induced CDF and LERF/LRF.
Seismic/fire interactions is a qualitative review of the plant fire risk caused by a potential earthquake. This element seeks to ensure that such seismic/fire interactions have been considered and their impacts assessed.
Uncertainty and sensitivity analysis identifies and characterizes sources of uncertainty as well as the potential sensitivities of the results to related assumptions and modeling approximations. The impact of parameter uncertainties on the quantitative results is assessed.
1.2.5 External Hazards Technical Elements PRA models of external hazards, when required, are based on the internal events PRA model, which are modified to include the impact of the identified external event scenarios in terms of causing initiating events (plant transients and LOCAs), and failing equipment used to respond to initiating events.
However, it is prudent to perform a screening and bounding analysis to screen out those external events that have an insignificant impact on risk. When external events are modeled in detail, the external event scenarios are developed during the hazard analysis and the fragility analysis as discussed below. The quantification task specific to external events is similar in nature to that for the internal events. Because of its dependence on the internal events model, the external events analysis incorporates the elements of Sections 1.2.1 and 1.2.2, as necessary.
Screening and bounding analysis identifies external events other than earthquakes (such as river-induced flooding) that may challenge plant operations and require successful mitigation by plant equipment and personnel to prevent core damage from occurring. The term screening out is used here for the process whereby an external event is excluded from further consideration in the PRA analysis.
There are two fundamental screening criteria embedded here. An event can be screened out if either (1) it meets the design criteria, or (2) it can be shown using an analysis that the mean value of the design-basis hazard used in the plant design is less than 10-5/year and that the conditional core-damage probability is less than 10-1, given the occurrence of the design-basis hazard.4 Alternatively, high confidence that the site hazard is below 10-6/yr could allow the event also to be screened out independent of CCDP. An 4
It is recognized that for those new reactor designs with substantially lower risk profiles (e.g., internal events CDF below 10-6/year), the quantitative screening value should be adjusted according to the relative baseline risk value.
DG 1200, Page 15 external event that cannot be screened out using either of these criteria is subjected to the detailed analysis.
Hazard analysis characterizes non-screened external events and seismic events, generally, as frequencies of occurrence of different sizes of events (e.g., earthquakes with various peak ground accelerations, hurricanes with various maximum wind speeds) at the site. The external events are site-specific and the hazard characterization addresses both aleatory and epistemic uncertainties.
Fragility analysis characterizes conditional probability of failure of SSCs whose failure may lead to unacceptable damage to the plant (e.g., core damage) given occurrence of an external event. For significant contributors (i.e., SSCs), the fragility analysis is realistic and plant-specific. The fragility analysis is based on extensive plant walkdowns reflecting as-built, as-operated conditions. It is recognized that at the design and initial licensing stage, plant walkdowns are not possible; however, the fragility analysis should reflect the as-designed plant.
Plant response analysis and quantification involves the modification of appropriate plant transient and LOCA PRA models to determine the conditional core damage probability, given damage to the sets of components identified. The external events PRA model includes initiating events resulting from the external events, external-event-induced SSC failures, non-external-event-induced failures (random failures), and human errors. The system analysis is well-coordinated with the fragility analysis and is based on plant walkdowns. It is recognized that at the design and initial licensing stage, plant walkdowns are not possible. The results of the external event hazard analysis, fragility analysis, and system models are assembled to estimate frequencies of core damage and large early release.
1.2.6 Interpretation of Results The results of the Level 1 PRA are examined to identify the contributors sorted by initiating events, accident sequences, equipment failures, and human errors. Methods such as importance measure calculations (e.g., Fussell-Vesely Importance, risk achievement worth, risk reduction worth, and Birnbaum Importance) are used to identify the contributions of various events to the estimation of CDF for both individual sequences and the total CDF [that is, both contributors to the total CDF, including the contribution from the different initiators (i.e., internal and external events) and different operating modes (i.e., full-and low-power and shutdown) and contributors to each contributing sequence are identified].
The results of the Level 2 PRA are examined to identify the contributions of various events to the model estimation of LERF or LRF for both individual sequences and the model as a whole, using such tools as importance measure calculations (e.g., Fussell-Vesely Importance, risk achievement worth, risk reduction worth, and Birnbaum Importance).
For many applications, it is necessary to combine the PRA results from different hazard groups (e.g., from internal events, internal fires, and external initiating events). For this reason, an important aspect in interpreting the PRA results is understanding both the level of detail associated with the modeling of each of the hazard groups, and the hazard group-specific model uncertainties. With respect to the level of detail, for example, the analysis of specific scope items such as internal fire, internal flooding, or seismic initiating events typically involves a successive screening approach, so that more detailed analysis can focus on the more significant contributions. The potential conservatism associated with the evaluation of the less significant contributors using this approach is assessed for each hazard group. In addition, each of the hazard groups has unique sources of model uncertainty. The assumptions made in response to these sources of model uncertainty and any conservatisms introduced by the analysis approaches can bias the assessment of importance measures with respect to the combined risk assessment and the relative contributions of the hazard groups to the various risk metrics. Therefore, the sources of
DG 1200, Page 16 model uncertainty are identified and their impact on the results analyzed for each hazard group individually, so that, when it is necessary to combine the PRA results, the overall results can be characterized appropriately. The sensitivity of the model results to model boundary conditions and other assumptions is evaluated, using sensitivity analyses to look at assumptions both individually or in logical combinations. The combinations analyzed are chosen to account for interactions among the variables.
NUREG-1855 provides guidance on the treatment of uncertainties associated with PRA. The understanding gained from these analyses is used to appropriately characterize the relative significance of the contributions from each hazard group.
1.2.7 Documentation Traceability and defensibility provide the necessary information such that the results can easily be reproduced and justified. The sources of information used in the PRA are both referenced and retrievable.
The methodology used to perform each aspect of the work is described either through documenting the actual process or through reference to existing methodology documents. Sources of uncertainty are identified and their impact on the results assessed. Assumptions made in performing the analyses are identified and documented along with their justification to the extent that the context of the assumption is understood. The results (e.g., products and outcomes) from the various analyses are documented. A source of uncertainty is one that is related to an issue where there is no consensus approach or model (e.g., choice of data source, success criteria, reactor coolant pressure seal LOCA model, human reliability model) and where the choice of approach or model is known to have an impact on the PRA results in terms of introducing new accident sequences, changing the relative importance of sequences, or significantly affecting the overall CDF, LERF, or LRF estimates that might have an impact on the use of the PRA in decision-making.
1.3 Attributes and Characteristics of the PRA Technical Elements Tables 2 and 3 describe, for each technical element of a PRA, the technical characteristics and attributes that provide one acceptable approach for determining the technical adequacy of the PRA such that the goals and purposes, defined in Regulatory Position 1.2, are accomplished.
For each given technical element, the level of detail may vary. The detail may vary from the degree to which (1) plant design and operation is modeled, (2) specific plant experience is incorporated into the model, and (3) realism is incorporated into the analyses that reflect the expected plant response.
Regardless of the level of detail developed in the PRA, the characteristics and attributes provided below are included. That is, each characteristic and attribute is always included, but the degree to which it is included may vary.
The level of detail needed is dependent on the application. The application may involve using the PRA during different plant stages (i.e., design, construction, and operation). Consequently, a PRA used to support a design certification will not have the same level of detail as a PRA of a plant that has years of operating experience. While it is recognized that the same level of detail is not needed, each of the technical elements and its attributes has to be addressed.
Table 2. Summary of Technical Characteristics and Attributes of a PRA ELEMENT TECHNICAL CHARACTERISTICS AND ATTRIBUTES (SEE NOTE 1)
PRA At-Power Level 1 PRA (internal events transients and LOCAs)
DG 1200, Page 17 Table 2. Summary of Technical Characteristics and Attributes of a PRA ELEMENT TECHNICAL CHARACTERISTICS AND ATTRIBUTES (SEE NOTE 1)
Initiating Event Analysis sufficiently detailed identification and characterization of initiating events grouping of individual events according to plant response and mitigating requirements proper screening of any individual or grouped initiating events Note: It is recognized that for those new reactor designs with substantially lower risk profiles (e.g., internal events CDF below 10-6/year) that the quantitative screening value should be adjusted according to the relative baseline risk value.
Success Criteria Analysis based on best-estimate engineering analyses applicable to the actual plant design and operation, as available codes developed in sufficient detail to:
analyze the phenomena of interest be applicable in the pressure, temperature, and flow range of interest Accident Sequence Development Analysis defined in terms of hardware, operator action, and timing requirements and desired end states (e.g., core damage or plant damage states) includes necessary and sufficient equipment (safety and non-safety) reasonably expected to be used to mitigate initiators includes functional, phenomenological, and operational dependencies and interfaces Systems Analysis models developed in sufficient detail to achieve the following purposes:
reflect the as-designed, as-built, as-operated plant (as applicable) including how it has performed during the plant history for operating plants reflect the success criteria for the systems to mitigate each identified accident sequence capture impact of dependencies, including support systems and harsh environmental impacts include both active and passive components and failure modes that impact the function of the system include common-cause failures, human errors, unavailability resulting from test and maintenance, etc.
Parameter Estimation Analysis estimation of parameters associated with initiating event, basic event probability models, recovery actions, and unavailability events using plant-specific and generic data as applicable consistent with component boundaries estimation includes a characterization of the uncertainty Human Reliability Analysis identification and definition of the human failure events that would result in initiating events or pre-and post-accident human failure events that would impact the mitigation of initiating events quantification of the associated human error probabilities taking into account scenario (where applicable) and plant-specific factors (as available) and including appropriate dependencies (both pre-and post-accident)
NUREG-1792 (Ref. 32) and NUREG-1842 (Ref. 33) provide good practices for meeting the above attribute and characteristics
DG 1200, Page 18 Table 2. Summary of Technical Characteristics and Attributes of a PRA ELEMENT TECHNICAL CHARACTERISTICS AND ATTRIBUTES (SEE NOTE 1)
Quantification estimation of the CDF for modeled sequences that are not screened as a result of truncation, given as a mean value estimation of the accident sequence CDFs for each initiating event group truncation values set relative to the total plant CDF such that the CDF is stable with respect to further reduction in the truncation value Level 2 PRA Plant Damage State Analysis identification of the attributes of the core damage scenarios that influence severe accident progression, containment performance, and any subsequent radionuclide releases grouping of core damage scenarios with similar attributes into plant damage states carryover of relevant information from Level 1 to Level 2 Severe Accident Progression Analysis use of appropriate codes by qualified trained users with an understanding of the code limitations and the means for addressing the limitations assessment of the credible severe accident phenomena via a structured process assessment of containment system performance including linkage with failure modes on non-containment systems establishment of the capacity of the containment to withstand severe accident environments assessment of accident progression timing, including timing of loss of containment failure integrity Quantification estimation of the frequency of different containment failure modes and resulting radionuclide source terms Source Term Analysis assessment of radionuclide releases including appreciation of timing, location, amount, and form of release grouping of radionuclide releases into smaller subsets of representative source terms with emphasis on large early release and large late release Note 1:
While each technical element has to be met and the associated characteristics addressed, it is the intent of the attribute that needs to be met.
PRA Low Power and Shutdown Level 1 PRA (internal eventstransients and LOCAs)
DG 1200, Page 19 Table 2. Summary of Technical Characteristics and Attributes of a PRA ELEMENT TECHNICAL CHARACTERISTICS AND ATTRIBUTES (SEE NOTE 1)
Plant Operating States The Level 1 PRA involves identification and characterization of a set of plant operational states during low-power and shutdown operations that are representative of all the plant states not covered in the full-power PRA.
The low-power and shutdown evolution is divided into POSs based on the unique impact on plant response to facilitate the practicality and efficiency of the PRA.
Each low-power and shutdown POS required to be considered for the specific application is identified and characterized as to all important conditions affecting the delineation and evaluation of core damage and large early release. These conditions include decay heat level, reactor coolant system configuration, reactor level, pressure and temperature, containment configuration, and the assumed representative plant system configurations within the POS.
Low-power and shutdown POSs that are subsumed into each other are shown to be represented by the characteristics of the subsuming group.
The duration and number of entries into each POS are determined.
The development, grouping, and quantification of the POSs are documented in a manner that facilitates PRA applications, updates, and peer review.
Initiating Event Analysis The initiating event analysis includes the same attributes and characteristics as for at-power, as well as the following:
examination of human-induced initiating events, including different types of LOCAs (e.g., drain-down events as opposed to pipe breaks) and temporary system alignments review of plant operational practices in grouping of events Success Criteria Analysis The success criteria analysis includes the same attributes and characteristics as for at-power, as well as the following:
analysis appropriate to the POS definition and characterization Accident Sequence Development Analysis The accident sequence development analysis includes the same attributes and characteristics as for at-power, as well as the following:
accounting for changing plant conditions within a POS Systems Analysis The systems analysis includes the same attributes and characteristics as for at-power, as well as the following:
identification of conditions varying from POS to POS for spatial and environmental hazards, systems actuation signals, system inventories (e.g., air)
Parameter Estimation Analysis The parameter estimation analysis includes the same attributes and characteristics as for at-power, as well as the following:
performance of estimation on a POS-specific basis, when necessary consideration of plant-specific data unique to POS (i.e., not applicable to at-power)
DG 1200, Page 20 Table 2. Summary of Technical Characteristics and Attributes of a PRA ELEMENT TECHNICAL CHARACTERISTICS AND ATTRIBUTES (SEE NOTE 1)
Human Reliability Analysis The human reliability analysis includes the same attributes and characteristics as for at-power, as well as the following:
differentiation between calibration errors that may impact equipment performance at-power versus low-power and shutdown POSs increased emphasis on contributions to initiating events performance of the analysis on a POS basis identification of dependent human failure events, particularly between those resulting in initiating events and those associated with responses to the initiating events justification for credit of operator actions credited for recovery in slowly developing scenarios (e.g., recovery times greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)
Quantification Quantification includes the same attributes and characteristics for at-power, as well as the following:
estimation of CDF and LERF for each POS Level 2 PRA Plant Damage State Analysis The plant damage state analysis includes the same attributes and characteristics as for at-power.
Severe Accident Progression Analysis The severe accident progression analysis includes the same attributes and characteristics as for at-power, as well as the following:
estimation of containment capacity based on the capacity of temporary closure, although for some POSs, containment may be open or have a reduced pressure capability assessment of the feasibility of the ability of operators to close containment before adverse environmental conditions (e.g., temperature, radiation, humidity, noise) prevent closure Quantification Quantification includes the same attributes and characteristics as for at-power.
Source Term Analysis The source term analysis includes the same attributes and characteristics as for at-power.
In addressing the above elements, because of the nature and impact of internal flood and fire and external hazards, their attributes are discussed separately in Table 3. This is because flood, fire, and external hazards analyses are spatial in nature and have the ability to cause initiating events but also have the capability to impact the availability of mitigating systems. Therefore, regarding the PRA model, the impact of flood, fire, and external hazards is to be considered in each of the above technical elements.
However, the characteristics and attributes provided in Table 3 are only for a Level 2 and Level 2 PRA for at-power conditions. The characteristics for internal flood, internal fire, and external events may be different for a Level 1 and Level 2 PRA for LPSD conditions.
Table 3. Summary of Technical Characteristics and Attributes of an Internal Flood and Fire Analysis and External Hazards Analysis ELEMENT TECHNICAL CHARACTERISTICS AND ATTRIBUTES
DG 1200, Page 21 Table 3. Summary of Technical Characteristics and Attributes of an Internal Flood and Fire Analysis and External Hazards Analysis ELEMENT TECHNICAL CHARACTERISTICS AND ATTRIBUTES Internal Flood Analysis Flood Identification Analysis sufficiently detailed identification and characterization of the following:
flood areas and SSCs located within each area flood sources and flood mechanisms type of water release and capacity structures functioning as drains and sumps verification of the information through plant walkdowns for built plants Flood Evaluation Analysis identification and evaluation of the following:
flood propagation paths flood mitigating plant design features and operator actions the susceptibility of SSCs in each flood area to the different types of floods elimination of flood scenarios uses well-defined and justified screening criteria Quantification identification of flooding-induced initiating events on the basis of a structured and systematic process estimation of flooding initiating event frequencies estimation of CDF for chosen flood sequences modification of the Level 1 models to account for flooding effects including uncertainties NOTE: For low-power and shutdown conditions, the following attributes and characteristics are also needed:
verification of temporary alignments for the specific outage or average modeled outage for data collection identification of previously existing flood barriers that may be impaired or disabled that could impact the flood zone consideration of automatic responses that may differ from at-power conditions
DG 1200, Page 22 Table 3. Summary of Technical Characteristics and Attributes of an Internal Flood and Fire Analysis and External Hazards Analysis ELEMENT TECHNICAL CHARACTERISTICS AND ATTRIBUTES Internal Fire Analysis Plant Boundary Definition and Partitioning Global analysis boundary captures all plant locations relevant to the fire PRA.
Physical analysis units are identified by credited partitioning elements that are capable of substantially confining fire damage behaviors.
Equipment Selection Equipment is selected for inclusion in the plant response model that will lead to a fire-induced plant initiator, or that is needed to respond to such an initiator (including equipment subject to fire-induced spurious actuation that affects the plant response).
The number of spurious actuations to be addressed increases according to the significance of the consequence (e.g., interfacing systems LOCA).
Instrumentation and support equipment are included.
Cable Selection Cables that are required to support the operation of fire PRA equipment (defined in the equipment selection element) are identified and located.
Qualitative Screening (Optional Element)
Screened out physical analysis units represent negligible contributions to risk and are considered no further.
Fire PRA Plant Response Model Based upon the internal events PRA, the logic model is adjusted to add new fire-induced initiating events and modified or new accident sequences, operator actions, and accident progressions (in particular those from spurious actuations).
Inapplicable aspects of the internal events PRA model are bypassed.
Fire Scenario Selection and Analysis Fire scenarios are defined in terms of ignition sources, fire growth and propagation, fire detection, fire suppression, and cables and equipment (targets) damaged by fire.
The effectiveness of various fire protection features and systems is assessed (e.g., fixed suppression systems).
Appropriate fire modeling tools are applied.
The technical basis is established for statistical and empirical models in the context of the fire scenarios (e.g., fire brigade response).
Scenarios involving the fire-induced failure of structural steel are identified and assessed (at least qualitatively).
Fire Ignition Frequencies Frequencies are established for ignition sources and consequently for physical analysis units.
Transient fires should be postulated for all physical analysis units regardless of administrative controls.
Appropriate justification must be provided to use nonnuclear experience to determine fire ignition frequency.
DG 1200, Page 23 Table 3. Summary of Technical Characteristics and Attributes of an Internal Flood and Fire Analysis and External Hazards Analysis ELEMENT TECHNICAL CHARACTERISTICS AND ATTRIBUTES Quantitative Screening Physical analysis units that are screened out from more refined quantitative analysis are retained to establish CDF and LERF/LRF.
Typically, those fire PRA contributions to CDF and LERF/LRF that are established in the quantitative screening phase are conservatively characterized.
Circuit Failure Analysis The conditional probability of occurrence of various circuit failure modes given cable damage from a fire is based upon cable and circuit features.
Postfire Human Reliability Analysis Operator actions and related postinitiator human failure events, conducted both within and outside of the main control room, are addressed.
The effects of fire-specific procedures are identified and incorporated into the plant response model.
Plausible and feasible recovery actions, assessed for the effects of fire, are identified and quantified.
Incorrect responses are identified and assessed for the fire scenarios.
Operator actions from the internal events PRA that are retained in the fire PRA are assessed for fire effects.
Fire Risk Quantification For each fire scenario, the fire risk results are quantified by combining the fire ignition frequency, the probability of fire damage and the conditional core damage probability (and CLRP/CLERP) from the fire PRA plant response model.
Total fire-induced CDF and LERF/LRF are calculated for the plant and significant contributors identified.
The contribution of quantitatively screened scenarios (from the quantitative screening element) is added to yield the total risk values.
Seismic Fire Interactions Potential interactions resulting from an earthquake and a resulting fire that might contribute to plant risk are reviewed qualitatively.
Qualitative assessment verifies that such interactions have been considered and that steps are taken to ensure that the potential risk contributions are mitigated.
Uncertainty and Sensitivity Uncertainty in quantitative fire PRA results because of parameter uncertainties are evaluated.
Model uncertainties as well as the potential sensitivities of the results to associated assumptions are identified and characterized.
DG 1200, Page 24 Table 3. Summary of Technical Characteristics and Attributes of an Internal Flood and Fire Analysis and External Hazards Analysis ELEMENT TECHNICAL CHARACTERISTICS AND ATTRIBUTES External Hazards Analysis (see NOTE)
Screening and Bounding Analysis credible external events (natural and man-made) that may affect the site are addressed screening and bounding criteria are defined and results are documented necessary walkdowns, for built plants, are performed non-screened events are subjected to an appropriate level of evaluations Hazard Analysis the hazard analysis is site-and plant-specific, for built plants the hazard analysis addresses uncertainties Fragility Analysis fragility estimates are plant-specific for significant contributors (i.e.,
SSCs) walkdowns are conducted to identify plant-unique conditions, failure modes, and as-built conditions, for built plants Plant response analysis and quantification external event caused initiating events that can lead to significant core damage and large release sequences are included external event-related unique failures and failure modes are incorporated equipment failures from other causes and human errors are included.
When necessary, human error data are modified to reflect unique circumstances related to the external event under consideration unique aspects of common causes, correlations, and dependencies are included the systems model reflects as-designed, as-built, as-operated plant conditions, as applicable the integration/quantification accounts for the uncertainties in each of the inputs (i.e., hazard, fragility, system modeling) and final quantitative results such as CDF and LERF the integration/quantification accounts for all dependencies and correlations that affect the results Note: In meeting the attributes and characteristics for the seismic portion of an External Events PRA, a seismic margins method is not an acceptable approach in the base PRA for the seismic contributors.
In understanding the results from a PRA, the different initiators and operating states need to be considered, in an integrated manner, when examining the results. The attributes for interpretation of the results are discussed separately in Table 4.
DG 1200, Page 25 Table 4. Summary of Technical Characteristics and Attributes for Interpretation of Results ELEMENT TECHNICAL CHARACTERISTICS AND ATTRIBUTES Level 1 PRA Interpretation of Results identification of the significant contributors to CDF (initiating events, accident sequences, equipment failures and human errors) identification of sources of uncertainty and their potential impact on the PRA model understanding of the impact of the assumptions on the CDF and the identification of the accident sequence and their contributors Level 2 PRA Interpretation of Results identification of the contributors to containment failure and resulting source terms identification of sources of uncertainty and their impact on the PRA model understanding of the impact of the assumptions on Level 2 results A significant aspect of the technical acceptability of the PRA is documentation. The attributes for documentation are discussed separately in Table 5.
Table 5. Summary of Technical Characteristics and Attributes for Documentation ELEMENT TECHNICAL CHARACTERISTICS AND ATTRIBUTES Traceability and defensibility the documentation is sufficient to facilitate independent peer reviews the documentation describes the interim results (sufficient to provide traceability and defensibility of the final results) and the final results, insights, and sources of uncertainties walkdown process, where applicable, and results are fully described 1.4 PRA Development, Maintenance, and Upgrade The PRA results used to support an application are derived from a PRA model that represents the as-designed, as-built, as-operated plant to the extent needed to support the application. Therefore, a process for developing, maintaining, and upgrading a PRA is established. This process involves identifying and using plant information to develop the original PRA and to modify the PRA. The process is performed such that the plant information identified and used in the PRA reflects the as-designed, as-built, as-operated plant, as applicable.5 The information sources include the applicable design, operation, maintenance, and engineering characteristics of the plant.
5 It is recognized that at the design certification or combined operating license stage where the plant is not built or operated, the term as-built, as-operated is meant to reflect the as-designed plant assuming site and operational conditions for the given design.
DG 1200, Page 26 For those SSCs and human actions used in the development of the PRA, the following information is identified, integrated, and used in the PRA:
plant design information reflecting the normal and emergency configurations of the plant plant operational information with regard to plant procedures and practices plant test and maintenance procedures and practices engineering aspects of the plant design Further, plant walkdowns are conducted to ensure that information sources being used actually reflects the plants as-built, as-operated condition. In some cases, corroborating information obtained from the documented information sources for the plant and other information may only be gained by direct observations. It is recognized that at the design and initial licensing stages, plant walkdowns are not possible.
Table 6 describes the characteristics and attributes that need to be included for the above types of information.
Table 6. Summary of Attributes and Characteristics for Information Sources Used in PRA Development TYPE OF INFORMATION ATTRIBUTES AND CHARACTERISTICS (SEE NOTE 1)
Design the safety functions required to maintain the plant in a safe stable state and prevent core or containment damage identification of those SSCs that are credited in the PRA to perform the above functions the functional relationships among the SSCs including both functional and hardware dependencies the normal and emergency configurations of the SSCs the automatic and manual (human interface) aspects of equipment initiation, actuation, operation, as well as isolation and termination the SSCs capabilities (flows, pressures, actuation timing, environmental operating limits) spatial layout, sizing, and accessibility information related to the credited SSCs other design information needed to support the PRA modeling of the plant Operational that information needed to reflect the actual operating procedures and practices used at the plant including when and how operators interface with plant equipment as well as how plant staff monitor equipment operation and status that information needed to reflect the operating history of the plant as well as any events involving significant human interaction Maintenance that information needed to reflect planned and typical unplanned tests and maintenance activities and their relationship to the status, timing, and duration of the availability of equipment historical information related to the maintenance practices and experience at the plant
DG 1200, Page 27 Table 6. Summary of Attributes and Characteristics for Information Sources Used in PRA Development TYPE OF INFORMATION ATTRIBUTES AND CHARACTERISTICS (SEE NOTE 1)
Engineering the design margins in the capabilities of the SSCs operating environmental limits of the equipment expected thermal hydraulic plant response to different states of equipment (such as for establishing success criteria) other engineering information needed to support the PRA modeling of the plant Note 1:
While each source of information needs to be used and the associated characteristics addressed, it is the intent of the attribute that needs to met. It is recognized that for reactors in the design or construction stage, the level of operational and maintenance information may vary.
As a plant operates over time, its associated risk may change. This change may occur for the following reasons:
The PRA model may change as a result of improved methods or techniques.
Operating data may change the availability or reliability of the plants structures, systems, and components.
Plant design or operation may change.
Therefore, to ensure that the PRA represents the risk of the current as-built and as-operated plant, the PRA needs to be maintained and upgraded over time. Table 7 provides the attributes and characteristics of an acceptable process.
Table 7. Summary of Characteristics and Attributes for PRA Maintenance and Upgrade CHARACTERISTICS AND ATTRIBUTES Monitor PRA inputs and collect new information Ensure cumulative impact of pending plant changes are considered Maintain configuration control of the computer codes used in the PRA Identify when PRA needs to be updated based on new information or new models/techniques/tools Ensure peer review is performed on PRA upgrades
- 2.
Consensus PRA Standards and Industry PRA Programs One acceptable approach to demonstrate conformance with Regulatory Position 1 is to use an industry consensus PRA standard or standards that address the scope of the PRA used in the decision-making. ASME and ANS have issued a PRA standard that provides technical requirements for a Level 1 and limited Level 2 PRA for internal and external events for at-power conditions (Ref. 20). This standard is not prescriptive in that it only establishes what a technically acceptable PRA needs to include,
DG 1200, Page 28 but it does not detail the requirements for performing a technically acceptable PRA.6 A peer review is needed to determine if the intent of the requirements in the standard is met.
2.1 Consensus PRA Standards In general, if a PRA standard is used to demonstrate conformance with Regulatory Position 1, the standard should be based on a set of principles and objectives. Table 8 provides an acceptable set of principles and objectives that were established and used by ASME/ANS in development of their Level 1/LERF PRA standard. Principle 3 recognizes that the various parts of a PRA can be, and generally are, performed to different capabilities. In developing the various models in the PRA, the different capabilities are distinguished by three attributes, determined by the degree to which the following criteria are met:
(5)
The scope and level of detail that reflects the plant design, operation, and maintenance may vary.
(6)
Plant-specific information versus generic information is used, such that the as-designed, as-built and as-operated plant is addressed.
(7)
Realism is incorporated, such that the expected response of the plant is addressed.
It is recognized that the various parts of a PRA will not be to the same capability category.
Which part of the PRA meets what capability category is dependent on the specific application.
These requirements will be either process in nature, or technical in nature. The process type requirements address the process for application, development, maintenance and upgrade, and peer review. The technical requirements address the technical elements of the PRA and what is necessary to adequately perform that element. Therefore, when a standard is used to demonstrate conformance with Regulatory Position 1, the requirements in the standard will need to be met.
6 The standards are written in terms of requirements. Therefore, the use of this work in this regulatory guide is standards language (e.g., in a standard, it states the standards sets forth requirements) and is not meant to imply a regulatory requirement.
DG 1200, Page 29 Table 8. Principles and Objectives of a Standard
- 1. The PRA standard provides well-defined criteria against which the strengths and weaknesses of the PRA may be judged so that decision-makers can determine the degree of reliance that can be placed on the PRA results of interest.
- 2. The standard is based on current good practices(see Note below) as reflected in publicly available documents. The need for the documentation to be publicly available follows from the fact that the standard may be used to support safety decisions.
- 3. To facilitate the use of the standard for a wide range of applications, categories can be defined to aid in determining the applicability of the PRA for various types of applications.
- 4. The standard thoroughly and completely defines what is technically required and should, where appropriate, identify one or more acceptable methods.
- 5. The standard requires a peer review process that identifies and assesses where the technical requirements of the standard are not met. The standard needs to ensure that the peer review process meets the following criteria:
determines whether methods identified in the standard have been used appropriately determines that, when acceptable methods are not specified in the standard, or when alternative methods are used in lieu of those identified in the standard, the methods used are adequate to meet the requirements of the standard assesses the significance of the results and insights gained from the PRA of not meeting the technical requirements in the standard highlights assumptions that may significantly impact the results and provides an assessment of the reasonableness of the assumptions is flexible and accommodates alternative peer review approaches includes a peer review team that is composed of members who are knowledgeable in the technical elements of a PRA, are familiar with the plant design and operation, and are independent with no conflicts of interest that may influence the outcome of the peer review
[this clause was not in the ASME definition]
- 6. The standard addresses the maintenance and update of the PRA to incorporate changes that can substantially impact the risk profile so that the PRA adequately represents the current as-designed
[added], as-built and as-operated plant.
- 7. The standard is a living document. Consequently, it should not impede research. It is structured so that, when improvements in the state of knowledge occur, the standard can easily be updated.
Note: Current good practices are those practices that are generally accepted throughout the industry and have shown to be technically acceptable in documented analyses or engineering assessments. [No definition was provided for these terms by ASME.]
For process requirements, the intent is generally straightforward and the requirement is either met or not met. For the technical requirements, it is not always as straightforward. Many of the technical requirements in a standard apply to several parts of the PRA model. For example, the requirements for systems analysis apply to all systems modeled, and certain of the data requirements apply to all parameters for which estimates are provided. If among these systems or parameter estimates there are a few examples in which a specific requirement has not been met, it is not necessarily indicative that this requirement has not been met. If the requirement has been met for the majority of the systems or
DG 1200, Page 30 parameter estimates, and the few examples can be put down to mistakes or oversights, the requirement would be considered to be met. If, however, there is a systematic failure to address the requirement (e.g.,
component boundaries have not been defined anywhere), then the requirement has not been complied with. In either case, the examples of noncompliance are to be (1) rectified or demonstrated not to be relevant to the application, and (2) documented.
Further, the technical requirements may be defined at two different levels: (1) high-level requirements, and (2) supporting requirements. High-level requirements are defined for each technical element and capture the objective of the technical element. These high-level requirements are defined in general terms, need to be met regardless of the capability category, and accommodate different approaches. Supporting requirements are defined for each high-level requirement. These supporting requirements are those minimal requirements needed to satisfy the high-level requirement. Consequently, determination of whether a high-level requirement is met, is based on whether the associated supporting requirements are met. Whether or not every supporting requirement is needed for a high-level requirement is application-dependent and is determined by the application process requirements.
One example of an industry consensus PRA standard is the ASME/ANS standard, with a scope for a PRA for Level 1 and limited Level 2 (LERF) for at-power operation and internal and external events. The staff regulatory position regarding this document is provided in Appendix A to this regulatory guide. If it is demonstrated that the parts of a PRA that are used to support an application comply with the ASME/ANS standard, when supplemented to account for the staffs regulatory positions contained in Appendix A, it is considered that the PRA is adequate to support that risk-informed regulatory application.
Additional staff positions will be added in future updates to this regulatory guide to address requirements for other risk contributors, such as accidents occurring during the low-power and shutdown modes of operation.
2.2 Industry Peer Review Program A peer review of the PRA is performed to determine whether the requirements established in the standard (as endorsed by the NRC in the appendices to this guide) have been met. An acceptable peer review approach is one that is performed according to an established process and by qualified personnel, and one that documents the results and identifies both strengths and weaknesses of the PRA.
The peer review process includes a documented procedure used to direct the team in evaluating the adequacy of a PRA. The review process compares the PRA against established criteria (e.g., technical requirements defined in a PRA standard that conforms to the PRA characteristics and attributes such as those provided in Regulatory Position 1.3). In addition to reviewing the methods used in the PRA, the peer review determines whether the methods were applied correctly. The PRA models are compared against the plant design and procedures to validate that they reflect the as-designed, or the as-built and as-operated plant. Assumptions are reviewed to determine if they are appropriate and to assess their impact on the PRA results. The PRA results are checked for fidelity with the model structure and for consistency with the results from PRAs for similar plants based on the peer reviewers knowledge. Finally, the peer review process examines the procedures or guidelines in place for updating the PRA to reflect changes in plant design, operation, or experience.
The team qualifications determine the credibility and adequacy of the peer reviewers. To avoid any perception of a technical conflict of interest, the peer reviewers will not have performed any actual work on the PRA. Each member of the peer review team must have technical expertise in the PRA elements he or she reviews, including experience in the specific methods that are used to perform the
DG 1200, Page 31 PRA elements. This technical expertise includes experience in performing (not just reviewing) the work in the element assigned for review. Knowledge of the key features specific to the plant design and operation is essential.7 Finally, each member of the peer review team must be knowledgeable in the peer review process, including the desired characteristics and attributes used to assess the adequacy of the PRA.
Documentation provides the necessary information such that the peer review process and the findings are both traceable and defensible. Descriptions of the qualifications of the peer review team members and the peer review process are documented. The results of the peer review for each technical element and the PRA update process are described, including the areas in which the PRA does not meet or exceed the desired characteristics and attributes used in the review process. This includes an assessment of the importance of any identified deficiencies on the PRA results and potential uses and how these deficiencies were addressed and resolved.
Table 9 provides a summary of the characteristics and attributes of a peer review.
Table 9. Summary of the Characteristics and Attributes of a Peer Review ELEMENT CHARACTERISTICS AND ATTRIBUTES Peer Review Process uses documented process uses as a basis for review a set of desired PRA characteristics and attributes uses a minimum list of review topics to ensure coverage, consistency, and uniformity reviews PRA methods reviews application of methods reviews assumptions and assesses their validity and appropriateness determines if PRA represents as-built and as-operated plant reviews results of each PRA technical element for reasonableness reviews PRA maintenance and update process reviews PRA modification attributable to use of different model, techniques, or tools Team Qualifications independent with no conflicts of interest collectively represent expertise in all the technical elements of a PRA including integration expertise in the technical element assigned to review knowledge of the plant design and operation knowledge of the peer review process Documentation describes the peer review team qualifications describes the peer review process documents where PRA does not meet desired characteristics and attributes assesses and documents significance of deficiencies describes the scope of the peer review performed (i.e., what was reviewed by the peer review team) 7 For new reactor designs that have not yet gone into commercial operation, it is recognized that a peer reviewer will not have knowledge of plant operation, and familiarity with some plant features (e.g., passive mitigation systems) may be limited. This is not to be construed as a limitation for performing a peer review using personnel who are otherwise qualified and generally familiar with the design and operation of similar plant types (e.g., pressurized-water reactors).
DG 1200, Page 32 The ASME/ANS standard requires a peer review to be performed. The peer review, per ASME/ANS, requires that (1) a peer review process be established, and (2) provides requirements for team qualifications and documentation. A peer review methodology (i.e., process) is provided in the industry-developed peer review programs (i.e., NEI 00-02, NEI 05-04, NEI 07-12 Ref. 27), and noted in the ASME/ANS standard as an acceptable process. The staff regulatory position on the peer review requirements in the ASME/ANS PRA Standard and the peer review process in NEI 00-02, 05-04 and 07-12 is provided in Appendices B, C and D, respectively, to this regulatory guide. When the staffs regulatory positions contained in the appendices are taken into account, use of a peer review can be used to demonstrate that the PRA [with regard to a Level 1/LERF PRA for at-power for internal events (excluding external events)] is adequate to support a risk-informed application.
As stated earlier, the peer review is to be performed against established standards (e.g., the ASME/ANS PRA standard). If different criteria are used than those in the established standard, then it needs to be demonstrated that these different criteria are consistent with the established standards, as endorsed by the NRC. NEI 00-02 provides separate criteria for a peer review of a Level 1 LERF PRA at full power for internal events, excluding internal flood and fire and external events. NEI 00-02 also provides guidance for resolving the differences between the established standards, as endorsed by the NRC (i.e., the ASME/ANS PRA standard and Appendix A to this guide), and its peer review criteria.
Appendix B to this guide provides the staff position on this guidance (referred to as the Licensee Self-Assessment Guidance). However, this self-assessment was developed because an established national consensus standard was not available at the time of the peer review; consequently, the criteria in NEI 00-02 were used to judge the technical acceptability of the PRA. It is the NRCs expectation that future peer reviews should be performed against the established standards, as endorsed in this guide.
- 3.
Demonstrating the Technical Adequacy of a PRA Used to Support a Regulatory Application This section of the regulatory guide addresses the third purpose identified above, namely, to provide guidance to licensees on an approach acceptable to the NRC staff to demonstrate that the quality of the PRA used, in total or the parts that are used to support a regulatory application, is sufficient to support the analysis.
The application-specific regulatory guides identify the specific PRA results to support the decision-making and the analysis needed to provide those results. The parts of the PRA to support that analysis must be identified, and it is for these elements that the guidance in this regulatory guide is applied. Regulatory Positions 3.1 and 3.2 summarize the expected outcome of the application of the application-specific regulatory guides in determining the scope of application of this regulatory guide.
3.1 Identification of Parts of a PRA Used To Support the Application When using this regulatory guide, it is anticipated that the licensees description of the application will include the following:
SSCs, operator actions, and plant operational characteristics affected by the application a description of the cause-effect relationships among the change and the above SSCs, operator actions, and plant operational characteristics mapping of the cause-effect relationships onto PRA model elements a definition of the acceptance criteria:
DG 1200, Page 33
- identification of the PRA results that will be used to compare against the acceptance criteria or guidelines and how the comparison is to be made
- the scope of risk contributors to support the decision Based on an understanding of how the PRA model is to be used to achieve the desired results, the licensee will have identified the parts of the PRA required to support a specific application. These include (1) the logic model events onto which the cause-effect relationships are mapped (i.e., those directly affected by the application), (2) all the events that appear in the accident sequences in which the first group of elements appear, and (3) the parts of the analysis required to evaluate the necessary results.
For some applications, this may be a limited set, but for others (e.g., risk-informing the scope of special treatment requirements), all parts of the PRA model are relevant.
3.2 Scope of Risk Contributors Addressed by the PRA Model Based on the definition of the application, and in particular the acceptance criteria or guidelines, the scope of risk contributors (internal and external initiating events and modes of plant operation) for the PRA is identified. For example, if the application is designed around using the acceptance guidelines of RG 1.174, the evaluations of CDF, CDF, LERF, and LERF should be performed with a full-scope PRA, including external initiating events and all modes of operation. However, since many PRAs do not address this full scope, the decision-makers must make allowances for these omissions. Examples of approaches to making allowances may in some cases include the introduction of compensatory measures, restriction of the implementation of the proposed change to those aspects of the plant covered by the risk model, and use of bounding arguments to cover the risk contributions not addressed by the model.
However, it should be noted, that consistent with the Commission endorsed phased PRA Quality Initiative, all risk contributors that cannot be shown as insignificant, should be assessed through quantitative risk assessment methods to support risk-informed licensing actions. This regulatory guide does not address this aspect of decision-making, but it is focused specifically on the quality of the PRA information used. As noted elsewhere in this guide, a PRA is considered a quantitative risk assessment method.
The PRA standards and industry PRA programs that have been, or are in the process of being, developed to address a specific scope. For example, the ASME/ANS PRA Standard addresses internal and external events, at-power for a limited Level 2 PRA analysis. NEI 00-02 is a peer review process for internal events (excluding internal fire) [Note that the internal flooding is only addressed in the self-assessment portion of NEI 00-02 (Appendix D).] Neither addresses external (including internal fire) initiating events or the low-power and shutdown modes of operation. The different PRA standards or industry PRA programs are addressed separately in appendices to this regulatory guide. In using this regulatory guide, the applicant will identify which of these appendices is applicable to the PRA analysis.
3.3 Demonstration of Technical Adequacy of the PRA There are two aspects to demonstrating the technical adequacy of the parts of the PRA to support an application. The first aspect is the assurance that the parts of the PRA used in the application have been performed in a technically correct manner, and the second aspect is the assurance that the assumptions and approximations used in developing the PRA are appropriate.
For the first, assurance that the parts of the PRA used in the application have been performed in a technically correct manner implies that (1) the PRA model, or those parts of the model required to support the application, represents the as-designed or as-built and as-operated plant, which, in turn, implies that the PRA is up to date and reflects the current design and operating practices, where appropriate, (2) the
DG 1200, Page 34 PRA logic model has been developed in a manner consistent with industry good practice (see footnote to Table 8) and that it correctly reflects the dependencies of systems and components on one another and on operator actions, and (3) the probabilities and frequencies used are estimated consistently with the definitions of the corresponding events of the logic model.
For the second, the current state-of-the-art in PRA technology is that there are issues for which there is no consensus on methods of analysis. Furthermore, PRAs are models, and in that sense the developers of those models rely on certain approximations to make the models tractable and on certain assumptions to address uncertainties as to how to model specific issues. Regulatory Guide 1.174, and, in more detail, NUREG-1855 provide guidance on how to address and treat the uncertainties associated with a PRA. In accordance with that guidance, the impact of these assumptions and approximations on the results of interest to the application needs to be understood.
3.3.1 Assessment that the PRA Model is Technically Correct When using risk insights based on a PRA model, the applicant must ensure that the PRA model, or at least those parts of it needed to provide the results, is technically correct as discussed above.
The licensee is to demonstrate that the model is up-to-date in that it represents the current plant design and configuration and represents current operating practices to the extent required to support the application. This demonstration can be achieved through a PRA maintenance plan that includes a commitment to update the model periodically to reflect changes that impact the significant accident sequences.
The various consensus PRA standards and industry PRA programs that provide guidance on the performance of, or reviews of, PRAs are addressed individually in the appendices to this regulatory guide.
These appendices document the staffs regulatory position on each of these standards or programs.
When the issues raised by the staff are taken into account, the standard or program in question may be interpreted to be adequate for the purpose for which it was intended. If the parts of the PRA can be shown to have met the requirements of these documents, with attention paid to the NRCs objections, it can be assumed that the analysis is technically correct. Therefore, other than an audit, a detailed review by NRC staff of the base model PRA will not be necessary. When deviations from these documents exist, the applicant must demonstrate either that its approach is equivalent or that the influence on the results used in the application are such that no changes occur in the significant accident sequences or contributors.
3.3.2 Assessment of Assumptions and Approximations Since the standards and industry PRA programs are not (or are not expected to be) prescriptive, there is some freedom on how to model certain phenomena or processes in the PRA; different analysts may make different assumptions and still be consistent with the requirements of the standard or the assumptions may be acceptable under the guidelines of the peer review process. The choice of a specific assumption or a particular approximation may, however, influence the results of the PRA. For each application that calls upon this regulatory guide, the applicant identifies the key assumptions8 and 8
A key assumption is one that is made in response to a key source of uncertainty in the knowledge that a different reasonable alternative assumption would produce different results, or an assumption that results in an approximation made for modeling convenience in the knowledge that a more detailed model would produce different results. For the base PRA, the term different results refers to a change in the risk profile (e.g., total CDF and total LERF, the set of initiating events and accident sequences that contribute most to CDF and to LERF) and the associated changes in insights derived from the changes in the risk profile. A reasonable alternative assumption is one that has broad
DG 1200, Page 35 approximations relevant to that application. This will be used to identify sensitivity studies as input to the decision-making associated with the application. Each of the documents addressed in the appendices either requires, or represents (in the case of the industry peer review program) a peer review. One of the functions of the peer review is to address the assumptions and make judgments as to their appropriateness.
- 4.
Documentation to Support a Regulatory Submittal The licensee develops documentation of the PRA model and the analyses performed to support the risk-informed regulatory activity. This documentation comprises both archival (i.e., available for audit) and submittal (i.e., submitted as part of the risk-informed request) documentation. The former may be required on an as needed basis to facilitate the NRC staffs review of the risk-informed submittal.
4.1 Archival Documentation Archival documentation associated with the base PRA includes the following:
A detailed description of the process used to determine the adequacy of the PRA.
The results of the peer review and/or self-assessment9, and a description of the resolution of all the peer review or self-assessment findings and observations. The results are documented in such a manner that it is clear why each requirement is considered to have been met. This can be done, for example, by providing a reference to the appropriate section of the PRA model documentation.
The complete documentation of the PRA model. If the staff elects to perform an audit on all or any parts of the PRA used in the risk-informed application, the documentation maintained by the licensee must be legible, retrievable (i.e., traceable), and of sufficient detail that the staff can comprehend the bases supporting the results used in the application. Regulatory Position 1.3 of this guide provides the attributes and characteristics of archival documentation associated with the base PRA.
A description of the process for maintenance and upgrade of the PRA. The history is maintained of the maintenance and upgrade activities, and the history includes the results of any peer reviews that were performed as a result of an upgrade.
The archival documentation associated with a specific application is expected to include enough information to demonstrate that the scope of the review of the base PRA is sufficient to support the application. This includes the following information:
the impact of the application on the plant design, configuration, or operational practices acceptance within the technical community and for which the technical basis for consideration is at least as sound as that of the assumption being challenged.
A key source of uncertainty is one that is related to an issue in which there is no consensus approach or model and where the choice of approach or model is known to have an impact on the risk profile (e.g., total CDF and total LERF, the set of initiating events and accident sequences that contribute most to CDF and to LERF) such that it influences a decision being made using the PRA. Such an impact might occur, for example, by introducing a new functional accident sequence or a change to the overall CDF or LERF estimates significant enough to affect insights gained from the PRA.
9 When referring to self-assessment, this term is meant to refer to the self-assessment process in NEI 00-02 for Level 1/LERF PRA for at-power internal events (excluding internal fire).
DG 1200, Page 36 the risk assessment, including a description of the methodology used to assess the risk of the application, how the base PRA model was modified to appropriately model the risk impact of the application, and details of quantification and the results the acceptance guidelines and method of comparison the scope of the risk assessment in terms of initiating events and operating modes modeled the parts of the PRA required to provide the results needed to support comparison with the acceptance guidelines 4.2 Licensee Submittal Documentation To demonstrate that the technical adequacy of the PRA used in an application is of sufficient quality, the staff expects the following information will be submitted to the NRC. Previously submitted documentation may be referenced if it is adequate for the subject submittal:
To address the need for the PRA model to represent the as-designed or as-built, as-operated plant, Identification of permanent plant changes (such as design or operational practices) that have an impact on those things modeled in the PRA but have not been incorporated in the baseline PRA model.
If a plant change has not been incorporated, the licensee provides a justification of why the change does not impact the PRA results used to support the application. This justification can be in the form of a sensitivity study that demonstrates the accident sequences or contributors significant to the application were not impacted (remained the same).
Documentation that the parts of the PRA required to produce the results used in the decision are performed consistently with the standard as endorsed in the appendices of this regulatory guide.
If a requirement of the standard (as endorsed in the appendix to this guide) has not been met, the licensee is to provide a justification of why it is acceptable that the requirement has not been met.
This justification should be in the form of a sensitivity study that demonstrates the accident sequences or contributors significant to the application were not impacted (remained the same).
A summary of the risk assessment methodology used to assess the risk of the application, including how the base PRA model was modified to appropriately model the risk impact of the application and results. (Note that this is the same as that required in the application-specific regulatory guides.)
Identification of the key assumptions and approximations relevant to the results used in the decision-making process. Also, include the peer reviewers assessment of those assumptions.
These assessments provide information to the NRC staff in their determination of whether the use of these assumptions and approximations is appropriate for the application, or whether sensitivity studies performed to support the decision are appropriate.
A discussion of the resolution of the peer review (or self-assessment, for peer reviews performed using the criteria in NEI 00-02) findings and observations that are applicable to the parts of the PRA required for the application. This may take the following forms:
DG 1200, Page 37
- a discussion of how the PRA model has been changed
- a justification in the form of a sensitivity study that demonstrates the accident sequences or contributors significant to the application were not impacted (remained the same) by the particular issue The standards or peer review process documents may recognize different capability categories or grades that are related to level of detail, degree of plant specificity, and degree of realism. The licensees documentation is to identify the use of the parts of the PRA that conform to capability categories or grades lower than deemed required for the given application (Section 1-3 of ASME/ANS RA-S-2008).
DG 1200, Page 38 D. IMPLEMENTATION The purpose of this section is to provide information to applicants and licensees regarding the NRCs plans for using this draft regulatory guide. The NRC does not intend or approve any imposition or backfit in connection with its issuance.
The NRC has issued this draft guide to encourage public participation in its development. The NRC will consider all public comments received in development of the final guidance document. In some cases, applicants or licensees may propose an alternative or use a previously established acceptable alternative method for complying with specified portions of the NRCs regulations. Otherwise, the methods described in this guide will be used in evaluating compliance with the applicable regulations for license applications, license amendment applications, and amendment requests.
REGULATORY ANALYSIS A draft regulatory analysis was published with the draft of this guide when it was originally published for public comment as Draft Regulatory Guide DG-1122. That draft regulatory analysis required no changes, so the NRC staff did not prepare a separate analysis for this proposed Revision 2 of Regulatory Guide 1.200. A copy of the draft regulatory analysis is available for inspection or copying for a fee in the NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDRs mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548; email PDR@nrc.gov.
DG 1200, Page 39 REFERENCES
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Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.12 10 All Federal Register notices listed herein were issued by the U.S. Nuclear Regulatory Commission, and are available electronically through the Federal Register Main Page of the public GPOAccess Web site, which the U.S. Government Printing Office maintains at http://www.gpoaccess.gov/fr/index.html. Copies are also available for inspection or copying for a fee from the NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDRs mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548; email PDR@nrc.gov.
11 Available in the Nuclear Regulatory Commissions Agencywide Documents Access and Management System (ADAMS) under ADAMS Accession No. ML051590519.
12 All regulatory guides listed herein were published by the U.S. Nuclear Regulatory Commission. Many Regulatory Guides are available electronically through the Public Electronic Reading Room on the NRCs public Web site, at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/. In addition, where an ADAMS accession number is identified, the specified regulatory guide is available electronically through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. Single copies of regulatory guides may also be obtained free of charge by writing the Reproduction and Distribution Services Section, ADM, USNRC, Washington, DC 20555-0001, or by fax to (301) 415-2289, or by email to DISTRIBUTION@nrc.gov. Active guides may also be purchased from the National Technical Information Service (NTIS) on a standing order basis. Details on this service may be obtained by contacting NTIS at 5285 Port Royal Road, Springfield, Virginia 22161, online at http://www.ntis.gov, by telephone at (800) 553-6847 or (703) 605-6000, or by fax to (703) 605-6900. Copies are also available for inspection or copying for a fee from the NRCs Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville, Maryland; the PDRs mailing address is USNRC PDR, Washington, DC 20555-0001. The PDR can also be reached by telephone at (301) 415-4737 or (800) 397-4205, by fax at (301) 415-3548, and by email to PDR@nrc.gov.
DG 1200, Page 40
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NUREG-0800, Standard Review Plan for the Review of the Safety Analysis Reports for Nuclear Power Plants, Section 19, Use of Probabilistic Risk Assessment in Plant-Specific, Risk-Informed Decisionmaking: General Guidance, U.S. Nuclear Regulatory Commission, Washington, DC, various dates and revisions.13
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Regulatory Guide 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance.
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10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, § 50.48(c), Fire Protection -- National Fire Protection Association Standard NFPA 805, U.S. Nuclear Regulatory Commission, Washington, DC.
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Regulatory Guide 1.175, An Approach for Plant-Specific, Risk-Informed Decisionmaking:
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13 All NUREG-series reports listed herein were published by the U.S. Nuclear Regulatory Commission, and are available electronically through the Public Electronic Reading Room on the NRCs public Web site, at http://www.nrc.gov/reading-rm/doc-collections/nuregs/. Copies are also available for inspection or copying for a fee from the NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDRs mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548; email PDR@nrc.gov. In addition, copies are available at current rates from the U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402-9328, telephone (202) 512-1800; or from the NTIS at 5285 Port Royal Road, Springfield, Virginia 22161, online at http://www.ntis.gov, by telephone at (800) 553-6847 or (703) 605-6000, or by fax to (703) 605-6900.
14 All NRC regulations listed herein are available electronically through the Electronic Reading Room on the NRC=s public Web site, at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Copies are also available for inspection or copying for a fee from the NRC=s Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD; the mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548; and e-mail PDR@nrc.gov.
DG 1200, Page 41
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- 19.
ANSI/ANS-58.23-2007, Fire PRA Methodology, American Nuclear Society, La Grange Park, Illinois, November 20, 2007.
- 20.
ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME, New York, New York, American Nuclear Society, La Grange Park, Illinois, April 9, 2008.
- 21.
Nuclear Energy Institute, Letter from Anthony Pietrangelo, Director of Risk-and Performance-Based Regulation Nuclear Generation, Nuclear Energy Institute, to Scott Newberry, Director of Risk Analysis and Applications Division, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC, August 16, 200217.
- 22.
NEI 00-02, Probabilistic Risk Assessment Peer Review Process Guidance, Revision A3, Nuclear Energy Institute, Washington, DC, March 20, 2000.
- 23.
Nuclear Energy Institute, Letter from Anthony Pietrangelo, Director of Risk-and Performance-Based Regulation Nuclear Generation, Nuclear Energy Institute, to Mary Drouin, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC, NEI 00-02, Probabilistic Risk Assessment Peer Review Process Guidance, Revision 1, May 19, 2006.
- 24.
ASME RA-Sa-2003, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to ASME RA-S-2002, ASME, New York, New York, December 5, 2003.
- 25.
Nuclear Energy Institute, Letter from Biff Bradley, Manager of Risk Assessment, Nuclear Energy Institute, to Mary Drouin, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC, Update of Appendix D to Revision 1 of NEI-00-02, Probabilistic Risk Assessment Peer Review Process Guidance, November 15, 2006.
- 26.
NEI 05-04, Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard, Revision 1, Nuclear Energy Institute, Washington, DC, December 2007.
- 27.
NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, Draft Version F, Revision 0, Nuclear Energy Institute, Washington, DC, December 2007.
15 Copies of ASME standards and documents may be obtained from the American Society of Mechanical Engineers, Three Park Avenue, New York, NY 10016-5990; phone (212) 591-8500.
16 Copies may be obtained from the American Nuclear Society, 555 N. Kensington Avenue, La Grange, Illinois 60526; phone (708) 352-6611.
17 All NEI documents may be obtained from the Nuclear Energy Institute, Attn: Mr. Biff Bradley, Suite 400, 1776 I Street, NW, Washington, DC 20006-3708; phone (202) 739-8083.
DG 1200, Page 42
- 28.
SECY-00-0162, Addressing PRA Quality In Risk-Informed Activities, U.S. Nuclear Regulatory Commission, Washington, DC, July 28, 2000.18
- 29.
SECY-04-0118, Plan for the Implementation of the Commissions Phased Approach to PRA Quality, U.S. Nuclear Regulatory Commission, Washington, DC, July 13, 2004.
- 30.
SECY-07-0042, Status of the Plan for the Implementation of the Commission's Phased Approach to Probabilistic Risk Assessment Quality, U.S. Nuclear Regulatory Commission, Washington, DC, March 7, 2007.
- 31.
NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Draft Report for Comment, U.S. Nuclear Regulatory Commission, Washington, DC, November 2007.
- 32.
NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA), U.S.
Nuclear Regulatory Commission, Washington, DC, April 2005.
- 33.
NUREG-1842, Evaluation of Human Reliability Analysis Methods Against Good Practices, U.S. Nuclear Regulatory Commission, Washington, DC, September 2006.
18 All Commission papers (SECYs) listed herein were published by the U.S. Nuclear Regulatory Commission, and are available electronically through the Public Electronic Reading Room on the NRCs public Web site, at http://www.nrc.gov/reading-rm/doc-collections/commission/secys/. Copies are also available for inspection or copying for a fee from the NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDRs mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548; email PDR@nrc.gov.
DG 1200, Page 43 BIBLIOGRAPHY ASME RA-Sb-2005, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum B to ASME RA-S-2002, ASME, New York, New York, December 30, 2005.
ASME RA-Sc-2007, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum C to ASME RA-S-2002, ASME, New York, New York, July 6, 2007.
Nuclear Energy Institute, Letter from Anthony Pietrangelo, Director of Risk-and Performance-Based Regulation Nuclear Generation, Nuclear Energy Institute, to Ashok Thadani, Director of Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC, December 18, 2001.
Appendix A to DG 1200, Page A-1 APPENDIX A NRC REGULATORY POSITION ON ASME/ANS PRA STANDARD Introduction The American Society of Mechanical Engineers (ASME) and the American Nuclear Society (ANS) has published ASME RA-S-2008, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications (April, 2008). The standard states that it sets forth requirements for probabilistic risk assessments (PRAs) used to support risk-informed decision for commercial nuclear power plants, and describes a method for applying these requirements for specific applications. The NRC staff has reviewed ASME/ANS RA-S-2008 against the characteristics and attributes for a technically acceptable PRA as discussed in Regulatory Positions 1 and 2 of this regulatory guide. The staffs position on each requirement (referred to in the standard as a requirement, a high-level requirement, or a supporting requirement) in ASME/ANS RA-S-2008 is categorized as no objection, no objection with clarification, or no objection subject to the following qualification, and defined as follows:
No objection. The staff has no objection to the requirement.
No objection with clarification. The staff has no objection to the requirement. However, certain requirements, as written, are either unclear or ambiguous, and therefore the staff has provided its understanding of these requirements.
No objection subject to the following qualification. The staff has a technical concern with the requirement and has provided a qualification to resolve the concern.
ASME/ANS RA-S-2008 PRA standard is divided into four parts; Part 1 includes general requirements, Part 2 technical requirements for internal events (excluding internal fire), Part 3 internal fire events, and Part 4 external events. Tables A-1 through A-4 provides the staffs position on each requirement in Part1 thru 4, respectively. A discussion of the staffs concern (issue) and the staff proposed resolution is provided. In the proposed staff resolution, the staff clarification or qualification to the requirement is indicated in either bolded text (i.e., bold) or strikeout text (i.e., strikeout); that is, the necessary additions or deletions to the requirement (as written in the ASME/ANS standard) for the staff to have no objection are provided.
Appendix A to DG 1200, Page A-2 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution Global Use of references, the various references, may be acceptable, in general; however, there may be aspects that are not applicable or not acceptable.
Clarification For every reference (except NEI 00-02):
No staff position is provided on this reference. The staff neither approves nor disapproves of information contained in the referenced document.
Section 1-1 1-1.1 thru 1-1.2 No objection
Appendix A to DG 1200, Page A-3 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution 1-1.3 The standard has two objectives: (1) specifying the requirements for a
[baseline] PRA, and (2) prescribing a method for applying the requirements for specific applications.
The discussion in this section mixes the two objectives. Further, the capability categories do not apply to the HLRs; with the discussion on capability categories moved to 1-1.4.3, there is not any need to mention it here. This section can easily be interpreted to mean that the entire PRA model or the PRA model for a specific hazard group can be performed to a single capability category. Even the title of the section suggests this interpretation. This inference is not correct.
The capability category concept was developed to recognize that the degree of detail, plant-specificity and realism within a PRA model does vary.
Consequently, the Capability Categories were meant to allow a supporting requirement to have this variance. This confusion could be clarified by provided the discussion on capability categories with the discussion on supporting requirements (i.e., Section 1-1.4.3).
Clarification 1-1.3 PRA Capability Categories This Standard is intended for a wide range of applications....
When a specific application is undertaken, judgment is needed to determine which Capability Category is needed for each portion of the PRA, and hence which SRs apply to the applications. (See Section 1-3).
Section deleted.
Appendix A to DG 1200, Page A-4 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution 1-1.4.2 Changes suggested to accommodate moving the salient parts of Section 1-1.3 moved to Section 1-1.4.3, Clarification 1-1.4.2 High Level Requirements. A set of Objectives and HLRs is provided for each PRA Element in the Technical Requirements Section of each respective Part of this Standard. All PRA using this Standard shall satisfy each of these HLRs as explained in para. 1-1.3. The HLRs set forth the minimum requirements for a technically acceptable baseline PRA, independent of an application. meeting this Standard The HLRs are defined in general terms and present the top level logic for the derivation of more detailed SRs for each of the PRA Capability Categories. The HLRs reflect not only the diversity of approaches that have been used to develop the existing PRAs, but also the need to accommodate future technological innovations.
1-1.4.3 See Issue Discussion under 1-1.3.
Clarification 1-1.4.3 Supporting Requirements. A set of SRs is provided for each HLR (that is provided for each PRA Element) in the Technical Requirements Section of each respective Part of this Standard. The SRs for the Technical Elements are presented.... whose action statements spans multiple categories is stated in Table 1-1.4-1. It is intended that, by meeting all the SRs under a given HLR, a PRA will meet that HLR.
In developing the different portions of the PRA model (e.g., system model), it is recognized that not every, for example, system model, will be or need be developed to the same level of detail, same degree of plant-specificity, or the same degree of realism. Three levels of degree to which three attributes can be developed are defined and labeled either Capability Category I, II or III. Table 1-1.3-1 describes the bases for defining the capability categories. This table was used to develop the SRs for each HLR.
[insert Table 1-1.3-1 here]
Appendix A to DG 1200, Page A-5 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution The intent of the capability categories is that, generally in developing the SRs, from Capability Category I to Capability Category III, the degree of scope and level of detail, the degree of plant-specificity, and the degree of realism increases. However, the Capability Categories are not based on the level of conservatism. [4th, 5th,
and 6th paragraphs from 1-1.3]. and hence which SRs apply to the applications. (See Section 1-3).
For each Capability Category, the SRs define the minimum requirements necessary to meet that Capability Category. Some SRs apply to only one Capability Category and some extend across two or three Capability Category. When a SR spans multiple categories, it applies equally to each Capability Category. When necessary, the differentiation between Capability Categories is made in other associated SRs. The interpretation of a SR that spans multiple categories is stated in Table 1-1.4-2.
[insert Table 1-1.4-1 here, replace Column heading 1, Action Statement Spans with SR Spans]
The Technical Requirements. to facilitate PRA applications, upgrades and peer review.
The SRs specify what to do rather than the peer review process described in Section 1-6.
1-1.5 thru 1-1.7 No objection
Appendix A to DG 1200, Page A-6 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution 1-1.8, 3rd paragraph The standard states that a full understanding of the differences in conservatism and level of detail is needed. The word full could be misinterpreted to mean an analysis way beyond the intent. These words could be interpreted to mean an analysis either way beyond the intent, or not meeting the intent Clarification.While there is a need in some applications to assess the significance with respect to the total CDF or LERF, this assessment has to be done with an full understanding of the differences in conservatism and level of detail introduced by the modeling approaches for the different hazard groups, as well as within each hazard group.
1-1.8, 6th paragraph The standard states that each hazard group are
.treated with an equivalent level of realism This statement only addresses one of the characteristics that distinguish Capability Category II.
Clarification.In Capability Category II, this Standard strives to ensure that the more significant contributors to each hazard group are understood and treated with an equivalent level of resolution, plant-specificity, and realism so as to not skew the results for that hazard group.
Section 1-2 1-2.1 Acronyms COL Acronym is needed Clarification COL: Combined License Other acronyms No objection 1-2.2 Definitions As-built, as-operated plant As written in the standard, these words are meant to address operating plants.
However, this standard is currently supporting design certifications and combined licenses.
Clarification As-built, as-operated plant: the technical elements of the PRA model reflects the plant design and operation as it is currently built and operated at the time of the application (use) of the PRA.
NOTE: At the design certification and combined license application stage, the plant is not built or operated. For these situations, the intent of the PRA model is to reflect the as-designed, as-to-be-built, and as-to-be-operated plant.
Appendix A to DG 1200, Page A-7 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution External event Internal fire from sources inside the plant is an internal event, and not an external event.
Qualification external event: an IE event originating outside a nuclear power plant that causes safety system failures, operator errors, or both, that in turn may lead to core damage or large early release. Events such as earthquakes, tornadoes, and floods from sources outside the plant and fires from sources inside or outside the plant are considered external events (see also internal event). By convention, LOSP not caused by another external event is considered to be an internal event Hazard (as used in Part 4),
Hazard (as used in probabilistic hazard assessment in Part 4), and Hazard Group These definitions could cause confusion.
Clarification hazard (as used in Part 4):.
hazard (as used in probabilistic hazard assessment in Part 4):.
hazard group:. to treat internal flooding as a separate hazard group.
Guidance on interpretation of the definitions for hazard and hazard group: In the context of Part 4, the term hazard when used as a noun, is intended to refer to a natural phenomenon. When used as a qualifier to assessment or analysis (e.g., hazard assessment) in Part 4, it identifies which of the hazards (i.e., natural phenomena) is being addressed. In the term hazard group the word hazard is generalized to represent the nature of the challenge to the plant which can include both natural phenomena and man-made challenges, as addressed by a specific Part of the standard.
Appendix A to DG 1200, Page A-8 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution Initiating event The definition includes the phrase any event either internal or external to the plant. This could imply, for example, that an earthquake is an initiating event.
However, in the context of this standard, in Parts 3 and 4, a fire is a cause of an initiating event and an earthquake is a cause of an initiating event. The initiating event is a description of the results of the disturbance to the plant, e.g., transient, LOCA.
Qualification initiating event: an event that perturbs the steady state operation of the plant by challenging plant control and safety systems whose failure could potentially lead to core damage and or radioactivity release. These events include failure of equipment from either internal plant causes (such as hardware faults, operator actions, floods, or fires),
or external plant causes (such as earthquakes or high winds). any event either internal or external. could potentially lead to core damage or large early release.
Internal event Internal fire from sources inside the plant is an internal event, and not an external event.
Qualification internal event: an event originating within a nuclear power plant that includes failure of equipment from internal plant causes (such as hardware faults, operator actions, floods, or fires), and that, in combination with safety system failures and/or operator errors, can affect the operability of plant systems and may lead to core damage or large early release.
Internal events can be classified as either internal hardware events (such as transients and loss of coolant accidents) or internal area events (such as floods and fires). By convention, loss of offsite power is considered to be an internal event, and internal fire is considered to be an external event.
PRA While a PRA may contain qualitative aspects, the overall risk assessment is quantitative, the definition is confusing as written Clarification Probabilistic risk assessment (PRA): a qualitative and quantitative assessment of of the risk....
Appendix A to DG 1200, Page A-9 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution PRA upgrade See the issue discussed on definition of Accident sequence, dominant.
Clarification PRA upgrade: The incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant sequences.
This could.
Other Definitions No objection Section 1-3 1-3.1 Global comment: It is assumed in the standard, as it is written, that every risk-informed activity is a licensee application and one that involves an actual change to the plant.
Every risk-informed activity is not necessarily a licensee application that involves a change to the plant.
Clarification Change, where appropriate, application to activity, and change, where appropriate, proposed change to proposed decision.
1-3.1, Stage A As noted above in the Issue discussion for 1-1.3, a Capability Category is not applied across the entire PRA model or across a portion of the PRA model. Capability Category distinction is applied at the SR level.
At this stage in the decision process, the analyst is determining the relative importance of the portion of the PRA model, such as for a hazard group.
Clarification (a) Stage A. An application is defined in terms of. the proposed change, the Capability Categories relative importance for each portion of the PRA necessary to support the application are is determined. Different portions. May be irrelevant.
Appendix A to DG 1200, Page A-10 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution 1-3.2.2 See Issue discussion above on 1-3.1, Stage A Clarification 1-3.2.2 Determination of Capability Categories. The Technical Requirements Section. are described in Subsection 1-1.3. For the application, determine the Capability Category relative importance for each portion of the PRA. (Box 4 of Figure 1-3.1-1). Depending on the relative importance, it will dictate which Capability Category is used for each SR for This determination dictates which SRs are used to evaluate the capabilities of each portion of the PRA to support the application. To determine this relative importance these capabilities, an evaluation shall be performed. the following application attributes shall be considered:....
1-3.3 3.4 No objection 1-3.5, 2nd paragraph Use of the word significant should match definitions provided in Section 2.2.
Clarification (b) The difference is not significant if the modeled accident sequences accounting for at least 90% 95% of CDF/LERF for the hazard group.
1-3.6 No objection
Appendix A to DG 1200, Page A-11 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution Figure 1-3.1-1 See staff proposed resolution for Section 1-1.4.2, text in Box 4 of Figure 1-3.1-1 needs to be modified be consistent with the text.
Clarification Section 1-4 1-4.1, 1-4.2 No objection 1-4.3 1-4.3.1, 1-4.3.2 No objection 1-4.3.3, 2nd paragraph The intent of this statement/requirement is for the use of outside expert, as such the use of the word should does not provide a minimum requirement.
Clarification The PRA analysis team shall should use outside experts, even when.
1-4.3.4 thru 1-4.3.7 No objection For each relevant Hazard Group, 4 Determine the relative importance to the application, identify the portions of the HG PRA relevant to the application, and for each relevant portion of the hazard group, determine the Capability Category for each SR needed for each potion of PRA to support application 5 PRA scope and risk metrics sufficient to evaluate plant change?
Appendix A to DG 1200, Page A-12 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution 1-4.4, 1-4.5 No objection Section 1-5 1-5.1 thru 1-5.6 No objection 1-5.7 (a)-(d)
No objection 1-5.7 (e)
It is unclear what is to be documented from the peer review.
Clarification (e) record of the performance and results of the appropriated PRA reviews (consistent with the requirements of Section 1-6.6) 1-5.7 (f),
1-5.7(g)
No objection Section 1-6 1-6.1 The purpose, as written, implies that it is solely an audit against the requirements of Section 4.
A key objective of the peer review is to ensure when evaluating the PRA against the technical requirements, the quality (i.e., strengths and weaknesses) of the PRA; this goal is to be clearly understood by the peer review team.
Further, the statement that the peer review need not assess all aspects of the PRA against all requirements could be taken to imply that some of the requirements could be skipped.
Clarification another purpose of the peer review is to determine the strengths and weaknesses in the PRA. Therefore, the peer review shall also assess the appropriateness of the assumptions. The peer review need not assess all aspects of the PRA against all requirements in the Technical Requirements Section of each respective Part of this Standard; however, enough aspects of the PRA shall be reviewed for the reviewers to achieve consensus on the assessment of each applicable supporting requirement, as well as on the adequacy of methodologies and their implementation for each PRA Element.
1-6.1.1, 1-6.1.2 No objection 1-6.2 1-6.2.1 thru 1-6.2.3 No objection
Appendix A to DG 1200, Page A-13 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution 1-6.3 As written, there does not appear to be a minimum set. The requirement as written provides suggestions. A minimal set of items is to be provided; the peer reviewers have flexibility in deciding on the scope and level of detail for each of the minimal items.
Clarification The peer review team shall use the requirements of this Standard. For each PRA element, a set of review topics required for the peer review team are provided in the subparagraphs of para.
6.3. Additional material for those Elements may be reviewed depending on the results obtained. These suggestions are not intended to be a minimum or comprehensive list of requirements. The judgment of the reviewer shall be used to determine the specific scope and depth of the review in each of each review topic for each PRA element.
1-6.4, 1-6.5 No objection
Appendix A to DG 1200, Page A-14 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution 1-6.6 1-6.6.1 The specific SRs addressed in the peer review need to be documented. As written it is not clear whether certain essential items are included in the documentation requirements that are necessary to accomplish the goal of the peer revew.
Clarification (e) a discussion of the extent to which each PRA Element was reviewed, including a list of SRs that were reviewed 1-6.6.2 No objection --------------------
Appendix 1-A Global The word significant is used in many places throughout the Appendix.
For example, the term significant changes in scope or capability is used to classify a change as a PRA upgrade, rather than a PRA maintenance.
The term significant change in risk insights is used to indicate when a focused peer review is suggested even for what is nominally classified as a PRA maintenance.
While what is meant by the former is clarified in the examples, what constitutes a significant change in risk insights needs to be defined and added to the defined terms in Section 1-2.
Clarification Add to list of definitions --
Significant change in risk insights:
Whether a change is considered significant is dependent on the context in which the insights are used. A change in the risk insights is considered significant when it has the potential to change a decision being made using the PRA.
Appendix A to DG 1200, Page A-15 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution 1-A.1.2. new paragraph An internal review is recommended in several places. This recommendation is made instead of an outside peer review. It needs to be made clear that this internal review is a type of peer review and should follow the process and requirements for the peer review requirements.
Clarification E.
When performing an internal review, the objective is to assess that the change to the PRA was correctly performed. In performing this assessment, the reviewer should use as guidance those applicable requirements in the standard.
1-A.1, 4th paragraph As written, it could be inferred that a newly developed method would not be considered an upgrade.
Clarification... new should be interpreted as new to the subject PRA even though the methodology in question has been applied in other PRAs and includes newly developed methods that have been used in the base PRA by the analyst. It is not intended to imply a newly developed method. This interpretation...
Table 1-A.1 This table is confusing and does not appear to add any value. How this table is to be used is not clear.
Clarification Delete table 1-A.3, Examples 8, 10, 17 It is assumed that a change to the base PRA that involves a calculation using the same computer code is a PRA maintenance type change rather than a PRA upgrade type change.
This assumption would only be valid if the calculation does not involve any new assumptions and the same analyst is performing the calculation.
Clarification Change:.... using the same computer code that was used for the prior calculations, given the calculation does not involve any new assumptions and the calculation is performed using the same guidance.
NOTE: the words that was used for the prior calculations do not appear in Example #8, staff clarification includes these words in Example #8.
Appendix A to DG 1200, Page A-16 Table A-1. Staff Position on ASME/ANS RA-S-2008 Part 1 Index No Issue Position Resolution 1-A.3, Example 18 Changing the definition of core damage without changing the thermal-hydraulic methodology may result in changed success criteria which could change the accident progression delineated by the accident sequences. It is not a foregone conclusion that this is a simple change to the PRA model. It needs to be reviewed to ensure that the resulting changes are appropriate. Further, what would be a significant change is open to interpretation, and would be prudent is not as strong as should.
Clarification Discussion and/or Alternative Recommendation: While this change may not be a new methodology, it could result in changing the success criteria with implications for the development of accident sequences, and potentially on the HRA (through timing), data, and quantification. If such changes have occurred, a focused peer review should be performed. If this change leads to a significant change in risk insights, a focused peer review would be prudent.
1-A.3, Example 21 This assumes that the important human actions are of the same nature as the new ones being added and utilize the ASEP method in the exact same manner. This can not be assumed.
Clarification Rationale: If it can be shown that the previous important human actions fully utilized the ASEP method, and that any deficiencies by the analyst were corrected, then, if there is no significant impact on risk insights, this change falls into.......
1-A.3, Example 23 This assumes that the development of the new human error probabilities invokes the HRA method used previously in the same manner. This can not be assumed.
Clarification Discussion and/or Alternative Recommendation:.... impact on human error probabilities. A focused peer review is advisable to ensure that the development of the new probabilities did not invoke a different application of the HRA method than previously used, and if so, it was appropriately performed and meets the applicable HRA requirements.
Appendix A to DG 1200, Page A-17 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution Section 2-1 2-1.1, thru 2-1.3 No objection 2-1.4 - IE No objection 2-1.4.1.1 No objection Table 2-1.4.1-1 No objection Tables 2-14.1-2(a) thru 2-1.4.1-2(d)
IE-A1 thru IE-A3a No objection IE-A4 The search for initiators should go down to the subsystem/train level.
Capability Category III should consider the use of other systematic processes.
Clarification Cat I and II:
PERFORM a systematic evaluation of each system where necessary (e.g., down to the subsystem or train level),
including support systems.
Cat III:
PERFORM a systematic evaluation of each system down to the subsystem or train level, including support systems.
PERFORM an FMEA (failure modes and effects analysis) or other systematic process to assess.
IE-A4a Initiating events from common cause or from both routine and non-routine system alignments should be considered.
Clarification Cat II and III:
resulting from multiple failures, if the equipment failures result from a common cause, and or from routine system alignments resulting from preventive and corrective maintenance.
IE-A5 thru IE-A10 No objection IE-B1 thru IE-B2 No objection IE-B3 The action verb AVOID is ambiguous.
Clarification Cat II:
AVOID subsuming DO NOT SUBSUME scenarios into a group.
Appendix A to DG 1200, Page A-18 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution IE-B4 thru IE-B5 No objection IE-C1 thru IE-C9 No objection IE-C10 Providing a list of generic data sources would be consistent with other SRs related to data.
Clarification COMPARE results and EXPLAIN differences in the initiating event analysis with generic data sources to provide a reasonable check of the results.
An example of an acceptable generic data sources is NUREG/CR-6928 [Note (1)].
IE-C11 Definitions of rare and extremely rare events can be deleted from this SR since they have been added to Chapter 2.
How plant-specific features are included in the use of generic data for establishing rare event frequencies requires clarification.
Clarification CC I and II:
For rare initiating events, USE industry generic data and INCLUDE plant-specific functions features in deciding which generic data is most applicable.
IE-C12 The size of relief valves is an important consideration when evaluating ISLOCAs.
Clarification CC I and II:
(a) configuration of potential pathways including numbers and types of valves and their relevant failure modes, and the existence, size, and positioning of relief valves IE-C13 No objection
Appendix A to DG 1200, Page A-19 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution Footnote 3 to Table 2-1.4.1-2(c)
The first example makes an assumption that the hourly failure rate is applicable for all operating conditions.
Clarification Thus, fbus at power = 1x10-7/hr
- 8760 hrs/yr
- 0.90
= 7.9x10-4/reactor year.
In the above example, it is assumed the bus failure rate is applicable for at-power conditions. It should be noted that initiating event frequencies may be variable from one operating state to another due to various factors. In such cases, the contribution from events occurring only during at-power conditions should be utilized.
IE-D1, thru IE-D3 No objection 2-1.4.2 - AS 2-1.4.2.1 The HLR and associated SRs are written for CDF and not LERF; therefore, references to LERF are not appropriate.
Clarification 2-1.4.2.1 Objectives. The objectives reflected in the assessment of CDF and LERF is such a way that.
Table 2-1.4.2-1 No objection Tables 2-1.4.2-2(a) thru 2-1.4.2-2(c)
AS-A1 thru AS-A8 No objection AS-A9 The code requirements for acceptability need to be stated.
Clarification Cat II and III:
affect the operability of the mitigating systems. (See SC-B4.)
AS-A10 The modifier significant does not have a clear definition.
Examples provide a clear understanding.
Clarification Cat II:
INCLUDE for each modeled initiating event, sufficient detail that significant differences in requirements on systems and required operator responses interactions (e.g., systems initiations or valve alignments) are captured.
AS-A11 No objection AS-B1 thru AS-B6 No objection
Appendix A to DG 1200, Page A-20 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution AS-C1 thru AS-C3 No objection 2-1.4.3 - SC 2-1.4.3.1 The HLR and associated SRs are written for CDF and not LERF; therefore, references to LERF are not appropriate.
Clarification (a) overall success criteria are defined (i.e.,
core damage and large early release)
Table 2-1.4.3-1 No objection Tables 2-1.4.3-2(a) thru 2-1.4.3-2(c)
SC-A1, SC-A2 Note: SC-A3 was deleted in Addendum B.
No objection SC-A4 thru SC-A6 No objection SC-B1 Requirements concerning the use of thermal/hydraulic codes should be cross-referenced.
Clarification Cat II and III:
for thermal/hydraulic, requiring detailed computer modeling. (See SC-B4.)
SC-B2 thru SC-B5 No objection SC-C1 thru SC-C3 No objection 2-1.4.4 - SY 2-1.4.4.1 No objection Table 2-1.4.4-1 No objection Tables 2-1.4.4-2(a) thru 2-1.4.4-2(c)
SY-A1 thru SY-A21 No objection
Appendix A to DG 1200, Page A-21 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution SY-A22 There are no commonly used analysis methods for recovery in the sense of repair, other than use of actuarial data.
Clarification is justified through an adequate analysis or examination of data collected in accordance with DA-C14 and estimated in accordance with DA-D8. (See DA-C14.)
SY-B1 thru SY-B8 Note: SY-B9 was deleted in Addendum B No objection SY-B10 References wrong SR; the parenthetical is referring to the requirement of include appropriate interfaces wit the support system, it is not referring to the mission time.
Clarification..INCLUDE appropriate interfaces with the support systems (see also SY-A6) required for successful operation of the system for a required mission time (see also AS-A6).
Examples of support systems include:
SY-B11 thru SY-B14 No objection SY-B15 Containment vent and failure can cause more than NPSH problems (e.g., harsh environments).
Clarification Examples of degraded environments include:
(h) harsh environments induced by containment venting or failure that may occur prior to the onset of core damage SY-B16 No objection SY-C1 thru SY-C3 No objection 2-1.4.5 - HR 2-1.4.5.1 No objection Table 2-1.4.5-1 No objection
Appendix A to DG 1200, Page A-22 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution Tables 2-1.4.5-2(a) thru 2-1.4.5-2(i)
HR-A1 Inspection may implicitly be included using test and maintenance, but explicit use of inspection term may eliminate interpretation errors (e.g.,
inspection may require actions to gain access to equipment, which could inadvertently cause a pre-initiator problem).
Clarification For equipment modeled in the PRA, IDENTIFY, through a review of procedures and practices, those test, inspection, and maintenance activities that require realignment of equipment outside its normal operational or standby status.
HR-A2, HR-A3 No objection HR-B1, HR-B2 No objection HR-C1 thru HR-C3 No objection HR-D1, HR-D2 No objection HR-D3 Add examples for what is meant by quality in items (a) and (b) of Cat II, III.
Clarification Cat II, III:
(a) the quality (including format, logical structure, ease of use, clarity, and comprehensiveness) of written procedures (for performing tasks) and the quality (e.g., configuration control process, technical review process, training processes, and management emphasis on adherence to procedures) of administrative controls (for independent review)
(b) the quality (e.g., adherence to human factors guidelines [Note (3)] and results of any quantitative evaluations of performance per functional requirements) of the human-machine interface, including both the equipment configuration, and instrumentation and control layout
Appendix A to DG 1200, Page A-23 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution HR-D4, HR-D5 No objection HR-D6 This SR should be written similarly to HR-G9 Clarification PROVIDE an assessment of the uncertainty in the. point estimates of HEPs. CHARACTERIZE the uncertainty in the estimates of the HEPs consistent with the quantification approach, and PROVIDE mean values for use in the quantification of the PRA results.
HR-D7 No objection Notes to Table 2-1.4.5-2(d)
Additional references cited in clarification to HR-D3.
Clarification NOTES:
(3) NUREG-0700, Rev. 2, Human-System Interface Design Review Guidelines; J.M. OHara, W.S. Brown, P.M. Lewis, and J.J. Persensky, May 2002.
HR-E1 No objection HR-E2 Need to explicitly state the need for some level of diagnosis in identifying the failure(s).
Clarification (b) those actions performed by the control room staff either in response to procedural direction or as skill-of-the-craft to diagnose and then recover a failed function, system or component that is used in the performance of a response action as identified in HR-H1.
HR-E3, HR-E4 No objection HR-F1, HR-F2 No objection HR-G1, HR-G2 No objection
Appendix A to DG 1200, Page A-24 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution HR-G3 In item (d) of CC II, III, clarify that clarity refers the meaning of the cues, etc.
In item (a) of CC I and item (g) of CC II, III, clarify that complexity refers to both determining the need for and executing the required response.
Clarification Cat I, II, and III:
(d) degree of clarity of the meaning of cues/indications (g) complexity of detection, diagnosis and decision-making, and executing the required response.
HR-G4 Requirements concerning the use of thermal/hydraulic codes should be cross-referenced.
Clarification Cat I, II, and III:
BASE. (See SC-B4.) SPECIFY the point in time.
HR-G5 thru HR-G8 No objection HR-G9 Action verb should be capitalized Clarification CHARACTERIZE Characterize the uncertainty..
HR-H1 thru HR-H3 No objection HR-I1 thru HR-I3 No objection 2-1.4.6 - DA 2-1.4.6.1 No objection Table 2-1.4.6-1 No objection Tables 2-1.4.6-2(a) thru 2-1.4.6-2(e)
DA-A1 thru DA-A3 No objection DA-B1, DA-B2 No objection
Appendix A to DG 1200, Page A-25 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution DA-C1 The list of data sources needs to be updated.
Clarification Examples of parameter estimates and associated sources include:
(a) component failure rates and probabilities: NUREG/CR-4639 [Note (1)], NUREG/CR-4550 [Note (2)],
NUREG-1715 [Note 7]
See NUREG/CR-6823 [Note 8] for lists of additional data sources.
DA-C2 thru DA-C13 No objection DA-C14 This SR provides a justification for crediting equipment repair (SY-A22). As written, it could be interpreted as allowing plant-specific data to be discounted in favor of industry data. In reality, for such components as pumps, plant-specific data is likely to be insufficient and a broader base is necessary.
Qualification IDENTIFY instances of plant-specific experience or and, when that is insufficient to estimate failure to repair consistent with DA-D8, applicable industry experience and for each repair, COLLECT.
DA-C15 No objection Notes to Table 2-1.4.6-2(c)
Additional references cited in the clarification to DA-C1.
Clarification NOTES:
(7) NUREG-1715, Component performance study, 1987-1998, Vols. 1-4.
(8) NUREG/CR-6823, Handbook of Parameter Estimation for Probabilistic Risk Assessment, USNRC, September 2003.
Appendix A to DG 1200, Page A-26 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution DA-D1 Other approved statistical processes for combining plant-specific and generic data are not available.
Clarification CC II and III:
USE a Bayes update process or equivalent statistical process that assigns that assigns appropriate weight to the statistical significance of the generic and plant specific evidence and provides an appropriate characterization of the uncertainty. CHOOSE.
DA-D2 thru DA-D5 No objection DA-D6 For consistency with Table 1.3-1 and DA-D1, the Cat III requirement is to apply to all common-cause events.
Clarification Cat III:
USE realistic common-cause failure probabilities for significant common-cause basic events. An example.
DA-D6a, DA-D7 No objection DA-D8 New requirement needed, DA-C14 was incomplete, only provied for data collection, not quantification of repair.
(See SY-A22.)
Qualification Cat I, II, and III:
For each SSC for which repair is to be modeled, ESTIMATE, based on the data collected in DA-C14, the probability of failure to repair the SSC in time to prevent core damage as a function of the accident sequence in which the SSC failure appears.
DA-E1 thru DA-E3 No objection 2-1.4.7 - IF 2-1.4.7.1 No objection Table 2-1.4.7-1 No objection Tables 2-1.4.7-2(a) thru 2-1.4.7-2(f)
IF-A1 thru IF-A4 No objection
Appendix A to DG 1200, Page A-27 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution IF-B1 The list of fluid systems should be expanded to include fire protection systems.
Clarification For each flood area. INCLUDE: (a) equipment (e.g., piping, valves, pumps) located in the area that are connected to fluid systems (e.g., circulating water system, service water system, and reactor coolant system, and fire protection system).
IF-B1a thru IF-B2 No objection IF-B3 It is necessary to consider a range of flow rates for identified flooding sources, each having a unique frequency of occurrence. For example, small leaks that only cause spray are more likely than large leaks that may cause equipment submergence.
Clarification (b) range of flow rates IF-B3a Note: IF-B4 was deleted in Addendum B No objection IF-C1 For a given flood source, there may be multiple propagation paths and areas of accumulation.
Clarification For each defined flood area and each flood source, IDENTIFY the propagation paths from the flood source area to the areas of accumulation.
IF-C2 thru IF-C2b No objection IF-C2c There is circular logic between this SR and IF-C5. This SR requires identifying SSCs for flood areas not screened out in IF-C5. A listed reason for screening a flood area in IF-C5 is that it does not contain SSCs.
Clarification For each flood area not screened out using the requirements under other Internal Flooding supporting requirements (e.g.,
IF-B1b and IFC5),.
Appendix A to DG 1200, Page A-28 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution IF-C3 For Cat II, it is not acceptable to just note that a flood-induced failure mechanism is not included in the scope of the internal flooding analysis. Some level of assessment is required.
Qualification Cat I:
For the SSCs identified in IF-C2c, IDENTIFY the susceptibility of each SSC in a flood area to flood-induced failure mechanisms. INCLUDE failure by submergence and spray in the identification process.
EITHER:
(a) ASSESS by using conservative assumptions; OR (b) NOTE that these mechanisms are not included in the scope of the evaluation.
Cat II:
For the SSCs identified in IF-C2c, IDENTIFY the susceptibility of each SSC in a flood area to flood-induced failure mechanisms. INCLUDE failure by submergence and spray in the identification process.
ASSESS qualitatively the impact of flood-induced mechanisms that are not formally addressed (e.g., using the mechanisms listed under Capability Category III of this requirement), by using conservative assumptions.
IF-C3a No objection IF-C3b Both a Capability Category II and III PRA should include the potential for maintenance-induced unavailability of barriers.
Qualification Cat II, III:
IDENTIFY inter-area.
INCLUDE potential for structural failure (e.g., of doors or walls) due to flooding loads and the potential for barrier unavailability, including maintenance activities.
IF-C3c thru IF-C9 No objection IF-D1 No objection IF-D2 Note that IF-D2 was deleted in Addendum B.
No objection
Appendix A to DG 1200, Page A-29 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution IF-D3 The action verb AVOID is ambiguous, it is a permissive, not a minimum requirement.
Clarification Cat II:
AVOID subsuming DO NOT SUBSUME scenarios into a group.
IF-D3a thru IF-D7 No objection IF-E1 thru IF-E6 No objection IF-E6a This supporting requirement should indicate the need to adjust the definition of common-cause failure groups while doing the internal flooding analysis.
Clarification INCLUDE, in the quantification, the combined effects of including equipment failures, due to causes independent of the flooding including unavailability due to maintenance, common-cause failures and other credible causes.
IF-E6b thru IF-E8 No objection IF-F1 thru IF-F23 No objection 2-1.4.8 - QU 2-1.4.8.1 SRs for LERF quantification reference the SRs in 2-1.4.8, and therefore, need to be acknowledged in 2-1.4.8.
Clarification The objectives of the quantification element are to provide an estimate of CDF (and support the quantification of LERF) based upon the plant-specific.
(b) significant contributors to CDF (and LERF) are identified such as initiating events.
Table 2-1.4.8-1 HLR-QU-A, HLR-QU-B, HLR-QU-C No objection Table 2-1.4.8-1 HLR-QU-D SRs for LERF quantification reference the SRs in 2-1.4.8 and, therefore, need to be acknowledged in 2-1.4.8.
Clarification significant contributors to CDF (and LERF), such as initiating events, accident sequences.
Appendix A to DG 1200, Page A-30 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution Table 2-1.4.8-1 HLR-QU-E, HLR-QU-F No objection Tables 2-1.4.8-2(a) thru 2-1.4.8-2(f)
QU-A1 No objection QU-A2a Need to acknowledge LERF quantification Clarification.consistent with the estimation of total CDF (and LERF) to identify significant accident.
QU-A2b The state-of-knowledge correlation should be accounted for all event probabilities. Left to the analyst to determine the extent of the events to be correlated. Need to also acknowledge LERF quantification Clarification Cat I:
ESTIMATE the point estimate CDF (and LERF)
Cat II:
ESTIMATE the mean CDF (and LERF),
accounting for the state-of-knowledge correlation between event probabilities when significant (see NOTE 1).
Cat III:
CALCULATE the mean CDF (and LERF) by.
QU-A3, QU-A4 No objection QU-B1 thru,,
QU-B5, QU-B7 thru QU-B9 No objection QU-B6 Need to acknowledge LERF quantification Clarification ACCOUNT for. realistic estimation of CDF or LERF. This accounting.
QU-C1 thru QU-C3 No objection
Appendix A to DG 1200, Page A-31 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution Table 2-1.4.8-2(d)
HLR-QU-D and Table 2-1.4.8-2(d) objective statement just before table need to agree; SRs for LERF quantification reference the SRs in 2-1.4.8 and, therefore, need to be acknowledged in 2-1.4.8.
Clarification significant contributors to CDF (and LERF), such as initiating events, accident sequences.
QU-D1a thru QU-D5b No objection QU-E1, QU-E2 No objection QU-E3 Need to acknowledge LERF quantification Clarification Cat I and II:
ESTIMATE the uncertainty interval of the CDF (and LERF) results.
QU-E4 The note has no relevance to the base model and could cause confusion; it should be deleted.
Clarification For each source of model uncertainty introduction of a new initiating event)
[Note (1)].
NOTE: For specific applications,. And in logical combinations.
QU-F1 No objection QU-F2 SR needs to use defined term significant instead of dominant. In addition, there is no requirement to perform sensitivity studies, and therefore, requirement is not needed for documentation.
Clarification (g) equipment or human actions that are the key factors in causing the accidents sequences to be non-dominant nonsignificant.
(h) the results of all sensitivity studies QU-F3 thru QU-F6 No objection 2-1.4.9 - LE 2-1.4.9.1 No objection Table 2-1.4.9-1 No objection
Appendix A to DG 1200, Page A-32 Table A-2. Staff Position on ASME/ANS RA-S-2008 Part 2 Index No Issue Position Resolution Tables 2-1.4.9-2(a) thru 2-1.4.9-2(g)
LE-A1 thru LE-A5 No objection LE-B1 thru LE-B3 No objection LE-C1 The SR for Capability Category II contains the statement:
NUREG/CR-6595, Appendix A provides an acceptable definition of LERF source terms. In fact, the appendix contains three possible definitions of LERF.
Clarification Cat II:
NUREG/CR-6595, App. A [Note (1)]provides a discussion and examples an acceptable definition of LERF source terms.
LE-C2a thru LE-C10 No objection LE-D1 thru LE-D6 No objection LE-E1 thru LE-E4 No objection LE-F1a thru LE-F3 No objection LE-G1, LE-G3 thru LE-G6 No objection LE-G2 There is no requirement to perform sensitivity studies.
Clarification (h) the model integration quantification including uncertainty and sensitivity analyses, as appropriate for the level of analysis Section 2-2.1 2-2.1 thru 2-2.2.9.2 No objection
Appendix A to DG 1200, Page A-33 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution Section 3-1 Risk Assessment Technical Requirements for Internal Fire Events At Power 3-1.1 thru 3-1.3 No objection 3-1.3.1 This paragraph (Subsection) is both unnecessary and confusing. Should be deleted Clarification 3-1.3.1 Scope: Screening.
The scope of this Part.to the extent the requirements are relevant.
3-1.3.2 thru 3-1.6 No objection 3-1.7, 3rd paragraph Need to clarify what is meant by the equipment to be identified.
Clarification (b) Fire PRA equipment selection (ES).
This element identifies. (b) equipment (including alarms, indicators, and controls) required to respond to each of the initiating events identified, and..
3-1.7, last paragraph Statement is made that two examples are provided; however, no examples are provided Clarification Tables of HLRs and SRs for the 13 FPRA elements. To each Capability Category.
When necessary, the differentiation. is made in other associated SRs; two examples are stated below. The interpretation of..
3-1.7.1 No objection Table 3-1.7.1-1 No objection Tables 3-1.7.1-2(a) thru 3-1.7.1-2(c)
PP-A1 No objection
Appendix A to DG 1200, Page A-34 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution PP-B1 The requirement for CC I is one of the options for CC II/III, and therefore, redundant.
This SR makes a distinction between the use of the fire protection program fire areas, and a subdivision of those areas. Making the recommended adjustment together with the suggested changes to PP-B2, 3, and 5 below is adequate to characterize the plant partitioning.
Qualification Cat II and III:
DEFINE Fire PRA physical units based on one of the following approaches:
Use the fire areas or Use using a combination of fire areas and physical analysis units where each physical analysis unit.
and If a fire area (PP-B2 through PP-B7)
PP-B2, PP-B3, and PP-B5 PP-B2 through PP-B7 apply only to capability categories II and III Clarification Cat I:
DO NOT CREDIT.
No requirement because not applicable PP-B4, PP-B6, PP-B7 No objection PP-C1 thru PP-C4 No objection 3-1.7.2 No objection Table 3-1.7.2-1 HLR-ES-A Grammatical change for clarity Clarification.identify equipment whose failure, including spurious operation, caused by an initiating fire, including spurious operation will would contribute.
Tables 3-1.7.2-2(a) thru 3-1.7.2.2(d)
Table 3-1.7.2-2(a) HLR-ES-A Conforming change to HLR-ES-A Clarification.identify equipment whose failure, including spurious operation, caused by an initiating fire, including spurious operation will would contribute.
ES-A1 Conforming change to HLR-ES-A Clarification IDENTIFY equipment whose failure, including spurious operation, caused by an initiating fire, including spurious operation would contribute.
Appendix A to DG 1200, Page A-35 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution ES-A2, ES-A3 No objection ES-A4 The parenthetical associated with the affected equipment does not appear to add anything since the scope of the fire damaged equipment which could cause an initiator is specified Clarification INCLUDE additional equipment based on the consideration with other fire-induced loss of function failures could cause an initiating event associated with the affected equipment considering:
ES-A5, ES-A6 No objection ES-B1 The notes states this requirement is a starting point for selection of mitigating equipment, and that an iterative process will provide the completeness with respect to Table 1-1.3-1, which specifies that the significant contributors be included in the model.
The requirement should represent the end result, not the beginning point.
Although the definition of failure mode in Part 1 includes spurious operation, it is worth explicitly including since it is an important issue.
Qualification Cat II:
IDENTIFY Fire.. and INCLUDE fire risk significant equipment from the internal events PRA.
NOTE-ES-B1-7: The gradation across.
the Fire PRA (other equipment can be assumed failed in the worst possible failure mode, including spurious operation). This will tend ES-B1 thru ES-B6 No objection
Appendix A to DG 1200, Page A-36 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution ES-C1 There is a concern with the way in which the term significant has been used. It is ambiguous as to whether the reference is to the total CDF, the internal events CDF, or the fire CDF. In order to avoid ambiguity, it is necessary to have a definition of the term significant. The terms significant accident sequence, significant accident progression sequence, significant basic event, significant cutset, and significant contributor are defined in Part 1 within the context of the hazard group, so that in Part 3, they should be interpreted as being measured with respect to the fire risk.
Clarification NOTE-ES-C1-3:. is not a significant contributor (as defined in Part 1),.
ES-C2 In CCI, need to clarify that spurious operation is a failure mode.
Clarification Cat I:
. the consequences of other selected equipment whose failures, including or spurious operations, will be included ES-D1 No objection 3-1.7.3 No objection Table 3-1.7.3-1 No objection Tables 3-1.7.3-2(a) thru 3-1.7.3-2(c)
CS-A1 thru CS-A9, CS-A11 No objection
Appendix A to DG 1200, Page A-37 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution CS-A10 PP-B1 already allows physical analysis units to be defined in terms of fire areas. As such the distinction between CCI and CCII is unnecessary.
Clarification Cat I:
IDENTIFY the fire areas. and CONFIRM. terminal end locations.
Cat II:
IDENTIFY. and CONFIRM.
terminal end locations.
Cat I and II:
IDENTIFY the physical analysis units, consistent with the plant partitioning analysis, through which each cable associated with a credited Fire PRA function passes and CONFIRM that the information includes treatment of cable terminal end locations.
CS-B1 No objection CS-C1 thru CS-C4 No objection 3-1.7.4 No objection Table 3-1.7.4-1 No objection Tables 3-1.7.4-2(a) thru 3-1.7.4-2(b)
QLS-A1, QLS-A4 No objection QLS-B1 thru QLS-B3 No objection 3-1.7.5 No objection Table 3-1.7.5-1 No objection Tables 3-1.7.5-2(a) thru 3-1.7.5-2(d)
PRM-A1 thru PRM-A5 No objection PRM-A6 Need to clarify that spurious operation is a failure mode.
Clarification Note-PRM-A6-1: This treatment is to include modeling of the equipment failure modes, including spurious operation, attributable to.
Appendix A to DG 1200, Page A-38 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution PRM-B1 thru PRM-B8 No objection PRM-B9 Need to clarify that spurious operation is a failure mode.
Clarification MODIFY the Fire PRA plant, are failed in the worst possible failure mode, including spurious operation.
PRM-B10 thru PRM-B14 No objection PRM-C1 NOTE PRM-C1-2 acknowledges that there is no current approach to addressing this requirement. It also appears to be open-ended, with no way of judging if it has been met.
Qualification Cat I:
No Requirement Cat II and III:
APPLY a systematic approach.
NOTE PRM-C1-1: The above..
NOTE PRM-C1-2: It is acknowledged NOTE PRM-C1-3: An example. for injection.
PRM-D1 No objection 3-1.7.6 No objection Table 3-1.7.6-1 No objection Tables 3-1.7.6-2(a) thru 3-1.7.6-2(h)
FSS-A1 No objection FSS-A2 Need to clarify that spurious operation is a failure mode.
Clarification..For each target set, SPECIFY
..including specification of the failure modes, including spurious operation.
FSS-A3 No objection FSS-A4 Use of language, one or more, is problematic, since it does not specify a minimum requirement.
Clarification IDENTIFY sufficient one or more combinations of target sets.... has been represented.
Appendix A to DG 1200, Page A-39 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution FSS-A5 The number of individual fire scenarios and level of detail should be commensurate with the relative risk importance of the physical analysis unit.
Clarification Cat I and II:
For each unscreened can be characterized commensurate with its risk significance.
NOTE FSS-A5-2: It is expected will be commensurate with the capability category and the fire relative risk importance FSS-A6 No objection FSS-B1 No objection
Appendix A to DG 1200, Page A-40 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution FSS-B2 There is a problem with the phrase can either be bounded or accurately characterized. This is not appropriate for a standard which should express the minimum requirement.
Clarification Cat I and II:
SELECT one. accurately characterized.
Cat I:
SELECT a sufficient number of fire scenarios, either in the MCR or elsewhere, leading to MCR abandonment and/or a reliance on ex-control room operator actions including remote and/or alternate shutdown actions, consisting of a combination of an ignition source (or group of ignition sources), such that the selected scenarios provide reasonable assurance that the MCR abandonment fire risk contribution can be bounded.
Cat II:
SELECT a sufficient number of fire scenarios, either in the MCR or elsewhere, leading to MCR abandonment and/or a reliance on ex-control room operator actions including remoter and/or alternate shutdown actions, consisting of a combination of an ignition source (or group of ignition sources), such that the selected scenarios provide reasonable assurance that the MCR abandonment fire risk contribution can be accurately characterized.
Cat III:
SELECT one. that the fire risk contribution of the MCR abandonment can be either bounded or accurately characterized and such.
FSS-C1 No objection
Appendix A to DG 1200, Page A-41 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution FSS-C2 See Issue for ES-C1 Clarification Cat II and III:
For those scenarios that represent significant contributors to a physical analysis units fire risk, CHARACTERIZE NOTE FSS-C3-1:. are not significant contributors (as defined in Part 1),.
FSS-C3 No objection FSS-C4 CCI has no requirement that the severity factor reflect the conditions and assumptions of the specified fire scenarios, as in CCII and CCIII.
Qualification Cat I:
If a severity.
the severity factor reflects the fire event set used to estimate fire frequency, and a technical basis supporting the severity factors determination is provided, and the severity factor bounds the conditions and assumptions of the specified fire scenarios under analysis.
FSS-C5 thru FSS-C8 No objection FSS-D1, FSS-D2 No objection FSS-D3 Again the either bounded or accurately characterized issue for CC II and CC III.
Clarification Cat I:
..in the analysis of each fire scenario such that the fire risk contribution of each unscreened physical analysis unit is bounded.
Cat II:
..the fire risk contribution of each unscreened physical analysis unit can be either bounded or accurately characterized.
Cat III:
..the fire risk contribution of each unscreened physical analysis unit can be either bounded or accurately characterized and such that the risk.
Appendix A to DG 1200, Page A-42 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution FSS-D4 thru FSS-D11 No objection FSS-E1 thru FSS-E4 No objection FSS-F1 Use of the term SELECT one or more Clarification Cat II and II:
..SELECT one or more fire scenarios(s) a sufficient number of fire scenarios to characterize could damage, including collapse, of the exposed structural steel..
FSS-F2, FSS-F3 No objection FSS-G1 thru FSS-G6 No objection FSS-H1 thru FSS-H10 No objection 3-1.7.7 No objection Table 3-1.7.7-1 No objection Tables 3-1.7.7-2(a) thru 3-1.7.7-2(b)
IGN-A1 The note, IGN-A1-1, appears to be more relevant to IGN-A2 than it is for IGN-A1. Item (e) only makes sense when there is equivalent nuclear experience.
Clarification NOTE IGN-A1-1:.(e) if being used as a supplement to, rather than in lieu of, nuclear data, that the fire frequencies calculated are consistent with those derived from nuclear experience ;..
IGN-A2 thru IGN-A10 No objection IGN-B1 thru IGN-B5 No objection 3-1.7.8 No objection Table 3-1.7.8-1 No objection Tables 3-1.7.8-2(a) thru 3-1.7.8-2(d)
QNS-A1 No objection
Appendix A to DG 1200, Page A-43 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution QNS-B1, QNS-B2 No objection QNS-C1 The screening criteria in Capability Categories II and III should relate to the total CDF and LERF for the fire risk, not the internal events risk.
See Issue for 3-1.7.2-2(c).
NOTE ES-C1 Clarification Cat II:
. and the sum of the CDF contribution for all screened fire compartments is
<10% of the estimated total CDF for internal fire events and the sum of the LERF contributions for all screened fire compartments is
<10% of the estimated total LERF for internal fire events Cat III:
.. and the sum of the CDF contributions for all screened fire compartments is
<1% of the estimated total CDF for internal fire events and the sum of the LERF contributions for all screened fire compartments is
<1% of the estimated total LERF for internal fire events 3-1.7.9 No objection Table 3-1.7.9-1 No objection Tables 3-1.7.9-2(a) thru 3-1.7.9-2(d)
CF-A1 See Issue for ES-C1 Clarification NOTE CF-A1-1:. for non-risk significant contributors (as defined in Part 1),.
CF-A2 No objection CF-B1 No objection
Appendix A to DG 1200, Page A-44 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution 3-1.7-10, 3-1.7-11 No objection 3-1.7-12 No objection Table 3-1.7.12-1 HLR-FQ-E See Issue for ES-C1 Clarification HLR-FQ-E:. and significant contributors (as defined in Part 1) to CDF and LERF.
Tables 3-1.7.12-2(a) thru 3-1.7.12-2(f)
FQ-A1 thru A1-A4 No objection FQ-B1 No objection FQ-C1 No objection FQ-D1 No objection FQ-E1 See Issue for ES-C1 Clarification IDENTIFY significant contributors (as defined in Part 1).
FQ-F1 See Issue for ES-C1 Clarification DOCUMENT the CDF and LERF SRs QU-F2 and QU-F3. are significant contributors (as defined in Part 1);.
FQ-F2 No objection
Appendix A to DG 1200, Page A-45 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution 3-1.7.13 A stated in Part 2, the objective of the uncertainty analysis is to identify and characterize the sources of uncertainty, so that when the PRA is used in an application, their impact on the results can be assessed.
Qualification The objectives of the Uncertainty and Sensitivity Analysis (UNC)are:
To identify key sources of analysis uncertainty To characterize these uncertainties, and To assess their potential impact of these uncertainties on the CDF and LERF estimates.
This section provides the ensuring that key the sources of uncertainties, i.e.,
those sources of uncertaintiesy that.
characterized with their potential impacts on the results Fire PRA understood.
Table 3-1.7.13-1 UNC-A The concepts of key sources of uncertainty and key assumptions make more sense in the context of an application and not a base PRA Qualification The Fire PRA shall identify key sources of CDF and LERF uncertainties and related including key assumptions and modeling approximations. These uncertainties shall be characterized such that their potential impacts on the results are understood.
Table 3-1.7.13-2(a)
The concepts of key sources of uncertainty and key assumptions make more sense in the context of an application and not a base PRA Qualification The Fire PRA shall identify key sources of CDF and LERF uncertainties and related including key assumptions and modeling approximations. These uncertainties shall be characterized such that their potential impacts on the results are understood.
UNC-A1 No objection UNC-A2 IGN-A8 is erroneously referenced for the treatment of uncertainty Clarification INCLUDE the treatment. As called out in SRs PRM-A4, FQ-F1, IGN-A810, IGN-B5, FSS-E3, FSS-E4, FSS-H5, FSS-H9, and CF-A2.
Appendix A to DG 1200, Page A-46 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution UNC-A3 An evaluation of the sensitivity of the results to sources of uncertainty is not needed for the base case. This should be performed on an as needed basis for an application Qualification INCLUDE an evaluation. and PROVIDE a basis for the evaluation.
No objection
Appendix A to DG 1200, Page A-47 Table A-3. Staff Position on ASME/ANS RA-S-2007, Part 3 Index No Issue Position Resolution Section 3-2 Peer Review for Internal Fire Events At Power 3-2.1 No objection 3-2.2 Expertise in Fire HRA is needed for the peer review Clarification
.fire modeling, and fire protection programs and their elements, and Fire HRA.
3.2.3, 2nd paragraph The paragraph states that:
Prior to performing the initial Fire PRA peer review, the peer review team should verify that the Internal Events PRA has been reviewed against Part 2. This should be a requirement.
Qualification Prior to performing the initial Fire PRA peer review, the peer review team should shall verify that the Internal Events PRA has been reviewed against Part 2. The results of the Internal Events PRA peer review should shall be reviewed as a part of the Fire PRA peer review.
3-2.3.1 thru 3-2.3.5 No objection 3-2.3.6, 2nd bullet Need to clarify that spurious operation is a failure mode.
Clarification., including specification of failure modes, such as spurious operation, given the nature..
3-2.3.7 thru 3-2.3.13 No objection Appendix 3-A FPRA Methodology (Nonmandatory)
The staff does not endorse the material in this appendix, and as such, does not have a position (i.e., no objections, no objection with clarification, or no objection with qualification) on any of the material contained in this appendix. However, it should be noted, that consistent with the Commission endorsed phase PRA Quality Initiative, all risk contributors that cannot be shown as insignificant, should be assessed through quantitative risk assessment methods to support risk informed licensing actions.
Appendix A to DG 1200, Page A-48 Table A-4. Staff Position on ASME/ANS RA-S-2008, Part 4 Index No Issue Position Resolution Section 4-1 4-1.1 thru 4-1.2 No objection 4-1.3.1 thru 4-1.3.2 No objection 4-1.3.3, 1st paragraph Incorrect section reference Clarification As discussed in 4-1.3 4-1.1.
4-1.4 thru 4-1.7 No objection 4-1.8 No objection 4-1.8.1 thru 4-1.8.2 No objection 4-1.8.3, 7th paragraph Incorrect section reference Clarification As discussed in 4-1.3 4-1.4.
4-1.8.3.1 No objection 4-1.8.4 No objection 4-1.8.4.1 4-1.8.4.1.1 No objection 4-1.8.4.1.2 Table 4-1.8.4.1.2-1 HLR-HA-G The reference to NUREG/CR-0098 broad band spectrum shape should be made in a supporting requirement.
Further, NURGE/CR-0098 spectral shapes are not always appropriate, particularly for CEUS sites.
Clarification For further use in the SPRA, the spectral shape SHALL be based on a site-specific evaluation taking into account the contributions of deaggregated magnitude-distance results of the probabilistic seismic hazard analysis. Broad-band, smooth spectral shapes, such as those presented in NUREG/CR-0098 [6]
(for lower-seismicity sites such as most of those east of the U.S. Rocky Mountains) are also acceptable if they are shown to be appropriate for the site. The use of uniform hazard response spectra is also acceptable unless evidence comes to light that would challenge these uniform hazard spectral shapes is acceptable if it reflects the site-specific shape.
Tables 4-1.8.4.1.2-2(a) to 4-1.8.4.1.2.2(j)
HA-A1 thru HA-A5 No objection
Appendix A to DG 1200, Page A-49 Table A-4. Staff Position on ASME/ANS RA-S-2008, Part 4 Index No Issue Position Resolution HA-B1 thru HA-B3 No objection HA-C1 thru HA-C4 No objection HA-D1 thru HA-D4 No objection HA-E1, HA-E2 No objection HA-F1 thru HA-F3 No objection 4-1.8.4.1.2 Table 4-1.8.4.1.2-2(g)
The reference to NUREG/CR-0098 broad band spectrum shape should be made in a supporting requirement.
Further, NURGE/CR-0098 spectral shapes are not always appropriate, particularly for CEUS sites.
Clarification For further use in the SPRA, the spectral shape SHALL be based on a site-specific evaluation taking into account the contributions of deaggregated magnitude-distance results of the probabilistic seismic hazard analysis. Broad-band, smooth spectral shapes, such as those presented in NUREG/CR-0098 [6]
(for lower-seismicity sites such as most of those east of the U.S. Rocky Mountains) are also acceptable if they are shown to be appropriate for the site. The use of uniform hazard response spectra is also acceptable unless evidence comes to light that would challenge these uniform hazard spectral shapes is acceptable if it reflects the site-specific shape.
HA-G1 Spectral shapes used to evaluate in-structure SSCs must include the effects of amplification from both local site conditions and SSI.
Based on IPEEE reviews, certain UHS shapes used for CEUS were not appropriate for the screening purpose.
Clarification NOTE HA-G1: The issue of which spectral shape should be used in the screening of structures, systems, and components (SSCs) and in quantification of SPRA results requires careful consideration. For screening purposes, the spectral shape used should have amplification factors, including effects from both local site conditions as well as soil-structure interaction, such that the demand resulting from the use of this shape is higher than that based on the design spectra. This will preclude premature screening of components and will avoid anomalies such as the screened components (e.g., surrogate elements) being the dominant
Appendix A to DG 1200, Page A-50 Table A-4. Staff Position on ASME/ANS RA-S-2008, Part 4 Index No Issue Position Resolution significant risk contributing components. Additional discussion on this issue can be found in Ref. 22.
In the quantification of fragilities and of final risk results, it is important to use as realistic a shape as possible.
Semi-site specific shapes, such as those given in NUREG-0098, have been used in the past and are considered may be adequate for this purpose, provided that they are shown to be reasonably appropriate for the site [42]. The uniform hazard response spectrum (UHS) is acceptable for this purpose if it can be shown that the UHS shape is appropriate for the site. unless evidence comes to light (e.g., within the technical literature) that these UHS do not reflect the spectral shape of the site-specific events. Recent developments [42] indicate that these spectral shapes are not appropriate for CEUS sites where high frequency content is dominant at hard rock sites.
HA-H It is not clear which requirement, CCI or CCIII, applies to CCII.
Clarification Cat I and II:
Use of existing studies allowed HA-I No objection HA-J1, HA-J2 No objection 4-1.8.4.2 4-1.8.4.2.1 No objection 4-1.8.4.2.2 Table 4-1.8.4.2.2-1 No objection Tables 1-8.4.2.2-2(a) thru 1-8.4.2.2-2(f)
SA-A1 thru SA-A3 No objection SA-B1, SA-B2, SA-B2a No objection SA-B3 Format is incorrect and not clear what is the requirement for CCII.
Clarification Cat I and II:
PERFORM an analysis of seismic-caused dependencies
Appendix A to DG 1200, Page A-51 Table A-4. Staff Position on ASME/ANS RA-S-2008, Part 4 Index No Issue Position Resolution USE bounding and PROVIDE the basis for such use.
SA-B4 thru SA-B10 No objection SA-C1 No objection SA-D1 No objection SA-E1 thru SA-E6 No objection SA-F1 thru SA-F3 No objection 4-1.8.4.3 No objection 4-1.8.4.3.1 Table 4-1.8.4.3.1-1 No objection Table 1-8.4.2.2-2(a) thru 1-8.4.2.2-2(g)
FR-A1, FR-A2 No objection FR-B1, FR-B2 No objection FR-C1 thru FR-C6 No objection FR-D1, FR-D2 No objection FR-E1 thru FR-E5 No objection FR-F1 thru FR-F4 No objection FR-G1 thru FR-G4 No objection 4-1.8.5 4-1.8.5.1 No objection 4-1.8.5.2 Table 4-1.8.5.2-1 No objection Table 1-8.5.2-2(a) thru 1-8.5.2-2(d)
WIND-A1 The six elements described in NOTE WIND-A1 provide the details required for the tornado wind hazard analysis and should be included in WIND-A1 as requirements.
Qualification Cat II and III:
In the tornado wind hazard analysis, USE. a mean hazard curve can be derived.
INCLUDE the following elements in the tornado wind hazard analysis:
(1) Variation of tornado intensity with occurrence frequency (The
Appendix A to DG 1200, Page A-52 Table A-4. Staff Position on ASME/ANS RA-S-2008, Part 4 Index No Issue Position Resolution frequency of tornado occurrence decreases rapidly with increased Intensity);
(2) Correlation of tornado width and length of damage area; longer tornadoes are usually wider; (3) Correlation of tornado area and intensity; stronger tornadoes are usually larger than weaker tornadoes; (4) Variation in tornado intensity along the damage path length; tornado intensity varies throughout its life cycle; (5) Variation of tornado intensity across the tornado path width.
(6) Variation of tornado differential pressure across the tornado path width.
NOTE WIND-A1: State-of-the-art methodologies are given... can be found in Refs. 13, 56, and 57.
Tornado wind hazard analysis SHOULD include the following elements:
(a) variation of tornado intensity with occurrence..
(f) variation of tornado differential pressure across the tornado path width.
WIND-A2 thru WIND-A4a No objection WIND-B1, WIND-B1a No objection Table 4-1.8.5.2-2(c)
The word significant has been correctly added in this HLR in Table 4-1.8.5.2-1 but is missing from the HLR statement in Table 4-1.8.5.2-2(c)
Clarification The wind-PRA systems model shall include significant wind-caused initiating events.
WIND-C1, WIND-C2 No objection
Appendix A to DG 1200, Page A-53 Table A-4. Staff Position on ASME/ANS RA-S-2008, Part 4 Index No Issue Position Resolution WIND-D1 thru WIND-D7 No objection 4-1.8.6 4-1.8.6.1 No objection 4-1.8.6.2 Table 4-1.8.6.2-1 No objection Table 1-8.6.2-2(a) thru 1-8.6.2-2(d)
FLOOD-A1 thru FLOOD-A6 No objection FLOOD-B1, FLOOD-B1a No objection 4-1.8.6.2 Table 4-1.8.6.2-2(c)
The word significant has been correctly added in this HLR in Table 4-1.8.6.2-1 but is missing from the HLR statement in Table 4-1.8.6.2-2(c)
Clarification The external-flooding-PRA systems model shall include all significant flood-caused initiating events.
FLOOD-C1, FLOOD-C2 No objection FLOOD-D1 thru FLOOD-D7 No objection Section 4-2 Peer Review for External Events At Power 4-2.1 thru 4-2.6 No objection Section 4-3Documentation Requirements for External Events At Power 4-3.1 thru 4-3.2 No objection Appendices 4-A No objection 4-B No objection 4-C The staff does not endorse the material in this appendix, and as such, does not have a position (i.e., no objections, no objection with clarification, or no objection with qualification) on any of the material contained in this appendix. However, it should be noted, that consistent with the Commission endorsed phase PRA Quality Initiative, all risk contributors that cannot be shown as insignificant, should be assessed through quantitative risk assessment methods to support risk informed licensing actions.
4-D The staff does not endorse the material in this appendix, and as such, does not have a position (i.e., no objections, no objection with clarification, or no objection with qualification) on any of the material contained in this appendix. However, it should be noted, that consistent with the Commission endorsed phase PRA Quality Initiative, all risk contributors that cannot be shown as insignificant,
Appendix A to DG 1200, Page A-54 Table A-4. Staff Position on ASME/ANS RA-S-2008, Part 4 Index No Issue Position Resolution should be assessed through quantitative risk assessment methods to support risk informed licensing actions.
Appendix B to DG 1200, Page B-1 APPENDIX B NRC POSITION ON THE NEI PEER REVIEW PROCESS (NEI 00-02)
Introduction The Nuclear Energy Institute (NEI) Peer Review Process is documented in NEI 00-02, Revision 1. It provides guidance for the peer review of probabilistic risk assessments (PRAs) and the grading of the PRA subelements into one of four capability categories. This document includes the NEI subtier criteria for assigning a grade to each PRA subelement. The NEI subtier criteria for a Grade 3 PRA have been compared by NEI to the requirements in the American Society of Mechanical Engineers (ASME) PRA Standard (ASME RA-Sb-2005) listed for a Capability Category II PRA. A comparison of the criteria for other grades/categories of PRAs was not performed since NEI contends that the results of the peer review process generally indicate the reviewed PRAs are consistent with the Grade 3 criteria in NEI 00-02. However, the PRAs reviewed have contained a number of Grade 2, and even Grade 4 elements. The comparison of the NEI subtier criteria with the ASME PRA Standard has indicated that some of the Capability Category II ASME PRA Standard requirements are not addressed in the NEI Grade 3 PRA subtier criteria. Thus, NEI has provided guidance to the licensees to perform a self-assessment of their PRAs against the criteria in the ASME PRA Standard that were not addressed during the NEI peer review of their PRA. A self-assessment is likely to be performed in support of risk-informed applications. This self-assessment guidance is also included in NEI 00-02, Revision 1.
However, since the issuance of ASME RA-Sb-2005, both an addendum has been issued (RA-Sc-2007) and a revision (ASME/ANS RA-S-2008). These documents contain requirements that were either revised or added, as compared to RA-Sb-2005. Consequently, the comparison of the NEI subtier criteria is not complete because there may still exist requirements in ASME/ANS RA-S-2008 not addressed by the subtier criteria.
This appendix provides the staffs position on the NEI Peer Review Process (i.e., NEI 00-02), the proposed self-assessment process, and the self-assessment actions. The staffs positions are categorized as following:
No objection. The staff has no objection to the requirement.
No objection with clarification. The staff has no objection to the requirement. However, certain requirements, as written, are either unclear or ambiguous, and therefore the staff has provided its understanding of these requirements.
No objection subject to the following qualification. The staff has a technical concern with the requirement and has provided a qualification to resolve the concern.
In the proposed staff resolution, the staff clarification or qualification that is needed for the staff to have no objection are provided.
NRC Position on NEI 00-02 Table B-1 provides the NRC position on the NEI Peer Review Process documented in NEI 00-02, Revision 1. With regard to the guidance in NEI 00-02 on the peer review process, the staff position is for future applications of the process.
Appendix B to DG 1200, Page B-2 Table B-1. NRC Regulatory Position on NEI 00-02 Section Position Commentary/Resolution Global Qualification The peer review process and self-assessment process in NEI 00-02 is based on Addendum B to the ASME PRA standard (RA-Sb-2005). The staff position on NEI 00-02 in Appendix B of Revision 1 of Regulatory Guide 1.200 is based on the staff position of RA-Sb-2005 as documented in Appendix A of Revision 1 of Regulatory Guide 1.200. However, since that time, ASME has issued Addendum C (RA-Sc-2007) and ASME and ANS has issued a revision (ASME/ANS RA-S-2008) to RA-S-2002 which incorporates the changes in RA-Sc-2007.
ASME/ANS RA-S-2008 (and ASME RA-Sc-2007) contains requirements that were revised or new requirements that were added (as compared to RA-Sb-2005). NEI 00-02, particularly the self-assessment process, needs to be based on this revised standard (RA-S-2008) taking into account the staff position on RA-S-2008 as endorsed in Appendix A in this revision of this regulatory guide. That is, for example, RA-S-2008 may have requirements either revised from RA-Sb-2005 or not included in RA-Sb-2005, and now need to be self-assessed in the context of the staff endorsement.
Section 1. Introduction Clarification The NEI process uses a set of checklists as a framework within which to evaluate the scope, comprehensiveness, completeness, and fidelity of the PRA being reviewed. The checklists by themselves are insufficient to provide the basis for a peer review since they do not provide the criteria that differentiate the different grades of PRA. The NEI subtier criteria provide a means to differentiate between grades of PRA.
1.1 Clarification Part 2 of the ASME/ANS PRA Standard (with the staffs position provided in Appendix A to this regulatory guide) can provide an adequate basis for a peer review of an at-power, internal events PRA (including internal flooding) that would be acceptable to the staff. Since the NEI subtier criteria do not address all of the requirements in Part 2 of the ASME/ANS PRA Standard, the staffs position is that a peer review based on these criteria is incomplete. The PRA standard requirements that are not included in the NEI subtier criteria (identified for a Grade 3 PRA in Table B-3) need to be addressed in the NEI self-assessment process as endorsed by the staff in this appendix.
Appendix B to DG 1200, Page B-3 Table B-1. NRC Regulatory Position on NEI 00-02 Section Position Commentary/Resolution 1.1 Clarification This section states that the NEI peer review process is a one-time evaluation process but indicates that additional peer review may be required if substantial changes are made to the PRA models or methodology. The staff position on additional peer reviews is to follow the guidance in Section 1-5 of Part 1 of the ASME/ANS PRA Standard which requires a peer review for PRA upgrades (PRA methodology changes).
1.2 No objection 1.3 Clarification Figure 1-3 indicates in several locations that the checklists included in NEI 00-02 are used in the peer review process. As indicated in the comment on Section 1.1 of NEI 00-02, the staffs position is that a peer review based on the checklists and supplemental subtier criteria is incomplete. The NEI self-assessment process, as endorsed by the staff in this appendix, is needed.
Clarification The NEI peer review process provides a summary grade for each PRA element. The use of a PRA for risk-informed applications needs to be determined at the subelement level. The staff does not agree with the use of an overall PRA element grade in the assessment of a PRA.
1.4 Clarification This section indicates that the process requires that the existing PRA meet the process criteria or that enhancements necessary to meet the criteria have been specifically identified by the peer reviewers and committed to by the host utility. Thus, the assigned grade for a subelement can be contingent on the utility performing the prescribed enhancement. An application submittal that utilizes the NEI peer review results needs to identify any of the prescribed enhancements that were not performed.
Clarification The staff believes that the use of PRA in a specific application should be of sufficient quality to support its use by the decision-makers for that application. The NEI peer review process does not require the documentation of the basis for assigning a grade for each specific subtier criterion. However, the staff position is that assignment of a grade for a specific PRA subelement implies that all of the requirements listed in the NEI subtier criteria have been met.
1.5 No Objection Section 2. Peer Review Process 2.1 Clarification See comment for Section 1.1.
Appendix B to DG 1200, Page B-4 Table B-1. NRC Regulatory Position on NEI 00-02 Section Position Commentary/Resolution 2.2 Clarification Part 2 of the ASME/ANS PRA Standard (with the staffs position provided in Appendix A to this regulatory guide) can provide an adequate basis for a peer review of an at-power, internal events PRA (including internal flooding) that would be acceptable to the staff. Since the NEI subtier criteria do not address all of the requirements in Part 2 of the ASME/ANS PRA Standard, the staffs position is that a peer review based on these criteria is incomplete. The PRA standard requirements that are not included in the NEI subtier criteria (identified for a Grade 3 PRA in Table B-3) need to be addressed in the NEI self-assessment process as endorsed by the staff in this appendix.
2.2 Steps 4, 7, &
8 Clarification See previous comment.
2.3 Clarification The peer reviewer qualifications do not appear to be consistent with the following requirements specified in Part 1, Section 1-6.2 of the ASME/ANS PRA Standard:
the need for familiarity with the plant design and operation the need for each person to have knowledge of the specific areas assigned for review the need for each person to have knowledge of the specific methods, codes, and approaches used in the PRA element assigned for review The NEI self-assessment process needs to address the peer reviewer qualifications with regard to these factors.
2.4 and 2.5 No objection Section 3. Pra Peer Review Process Elements and Guidance 3.1 No objection 3.2, 3.3 Clarification See comment for Section 1.1.
Appendix B to DG 1200, Page B-5 Table B-1. NRC Regulatory Position on NEI 00-02 Section Position Commentary/Resolution Clarification The NEI peer review process grades each PRA element from 1 to 4, while the ASME/ANS PRA Standard uses Capability Categories I, II, and III. The staff interpretation of Grades 2, 3, and 4 is that they correspond broadly to Capability Categories I, II, and III, respectively. This statement is not meant to imply that the supporting requirements, for example, for Category I are equally addressed by Grade 2 of NEI-00-02. The review of the supporting requirement for Category II against Grade 3 of NEI-00-02 indicated discrepancies and consequently the need for a self-assessment. The existence of these discrepancies would indicate that it would not be appropriate to assume that there are not discrepancies between Category I and Grade 2. A comparison between the other grades and categories has not been performed.
The implications of this are addressed in item 7a on Table B-2.
3.3 Qualification The staff believes that different applications of a PRA can require different PRA subelement grades. The NEI peer review process is performed at the subelement level and does not provide an overall PRA grade. Therefore, it is inappropriate to suggest an overall PRA grade for the specific applications listed in this section. The staff does not agree with the assigned overall PRA grades provided for the example applications listed in this section of NEI 00-02.
3.4 Clarification The general use and interpretation of the checklists in the grading of PRA subelements is addressed in this section. The subtier criteria provide a more substantial documentation of the interpretations of the criteria listed in the checklists. However, as previously indicated, the subtier criteria do not fully address all of the PRA standard requirements. The PRA standard requirements that are not included in the NEI subtier criteria (identified for a Grade 3 PRA in Table B-3) need to be addressed in the NEI self-assessment process as endorsed by the staff in this appendix.
Appendix B to DG 1200, Page B-6 Table B-1. NRC Regulatory Position on NEI 00-02 Section Position Commentary/Resolution Section 4. Peer Review Process Results and Documentation 4.1 Clarification A primary function of a peer review is to identify those assumptions and models that have a significant impact on the results of a PRA and to pass judgment on the validity and appropriateness of the assumptions. A review of the NEI 00-02 and the subtier criteria section on quantification and results interpretation failed to identify specific wording in any requirements to review the impact of assumptions on the results.
However, there are requirements to identify unique or unusual sources of uncertainty not present in typical or generic plant analyses. Since the evaluation of the impact of assumptions is critical to the evaluation of a PRA and its potential uses, the NEI peer review process needs to address assumptions, not just those that are unique or unusual. The NEI self-assessment process needs to address those assumptions not reviewed in the NEI peer review process. See staff position in Appendix A on Section 1-6.1 of Part 1 of the ASME/ANS PRA standard.
Qualification The NEI peer review report provides a summary grade for each PRA element. The use of a PRA for risk-informed applications needs to be determined at the subelement level. The staff does not agree with the use of an overall PRA element grade in the assessment of a PRA.
4.2, 4.3 No objection Appendix A. Preparation Material for the Peer Team Review A.1 thu A.6 No objection A.7 Clarification A list of sensitivity calculations that a utility can perform prior to the peer review is provided. Additional or alternative sensitivities can be identified by the utility. Sensitivity calculations that address key assumptions that may significantly impact the risk-informed applications results need to be considered in the NEI self-assessment process.
A.8 thu A.10 No objection
Appendix B to DG 1200, Page B-7 Table B-1. NRC Regulatory Position on NEI 00-02 Section Position Commentary/Resolution Appendix B. Technical Element Checklists Checklist tables No objection As previously stated, the staff position is that the checklists by themselves are insufficient to provide the basis for a peer review.
(See the comment for Section 1.1.) Because of this, the staff has not reviewed the contents or the assigned grades in these checklists. However, the staff position on the comparison of the Grade 3 NEI subtier criteria to the Capability Category II requirements in the ASME/ANS PRA Standard is documented in Table B-3.
Appendix C. Guidance for the Peer Review Team C.1 No objection C.2 No objection C.3 Clarification See comment for Section 4.1.
C.4 Clarification/
Qualification See the two comments on Section 3.3.
C.5 No objection C.6 Qualification See the comments on Section 4.1.
C.7 Clarification The staff does not agree with the use of an overall PRA element grade (documented in Tables C.7-5 & C.7-6) in the assessment of a PRA.
NRC Position on the Self-Assessment Process The staff position on the self-assessment process proposed by NEI to address the requirements in the ASME PRA Standard and Addenda A and B (ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005) that are not included in the NEI subtier criteria are addressed in this section. Both the self-assessment process and the specific actions recommended by NEI to address missing ASME standard requirements are addressed.19 However, since the issuance of ASME RA-Sb-2005, both an addendum has been issued (RA-Sc-2007) and a revision (ASME/ANS RA-S-2008). These documents contain requirements that were either revised or added, as compared to RA-Sb-2005. Consequently, the comparison of the NEI subtier criteria is not complete because there may still exist requirements in ASME/ANS RA-S-2008 not addressed by the subtier criteria.
19 The NEI comparison between NEI 00-02 criteria and the ASME requirements utilized the original standard as modified by subsequent addenda (A and B).
Appendix B to DG 1200, Page B-8 Table B-2 provides the NRC position on the NEI self-assessment process documented in Appendix D1 to NEI 00-02, Revision 1.
Table B-2. NRC Regulatory Position on NEI Self-Assessment Process Section Position Commentary/Resolution Summary No objection Regulatory Framework No objection Industry PRA Peer Review Process Clarification See the staff comments on the NEI peer review process provided in Table B-1.
ASME PRA Standard Clarification See the staff comments on the ASME/ANS PRA Standard provided in Appendix A to this regulatory guide.
Comparison of NEI 00-02 and ASME Standard Clarification The NRC position is that the performance of the existing peer reviews as supplemented by the NEI self-assessment process, as clarified in Regulatory Guide 1.200, meets the NRC requirements for a peer review.
The staff does not agree or disagree with the number of supporting requirements of the ASME PRA Standard that are addressed (completely or partially) in the NEI subtier criteria. The staffs focus is on ensuring that the self-assessment addresses important aspects of a PRA that are not explicitly addressed in the NEI subtier criteria. [See Note (1) at end of Table B-2.]
Clarification It is stated that If, the PRA is upgraded, new peer reviews may be required to meet paragraph 5.4 of the ASME standard. NEI-05-04, Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard, provides guidance in this regard. NRC has not endorsed NEI-05-04. The staff has reviewed NEI-05-04, and the staffs position is provided in Appendix C of this regulatory guide. [See Note (1) at end of Table B-2.].
General Notes for Self-Assessment Process 1
No objection
Appendix B to DG 1200, Page B-9 Table B-2. NRC Regulatory Position on NEI Self-Assessment Process Section Position Commentary/Resolution 2
Clarification Certain ASME PRA Standard requirements, although not explicitly listed in the NEI subtier criteria, may generally be included as good PRA practice. Credit may be taken for meeting these ASME requirements subject to confirmation in the self-assessment that the requirements were in fact addressed by the peer review. Table B-4 identifies the ASME PRA Standard requirements not explicitly addressed in the NEI subtier criteria that the staff believes need to be addressed in the NEI self-assessment process. [See Note (1) at end of Table B-2.].
3 Clarification The self-assessment process should consider the clarifications and qualifications on Addendum B that will be provided in Appendix A. [See Note (1) at end of Table B-2.]
Self-Assessment Process Attributes No objection Overall Peer Review Process and Decision No objection Self-Assessment Process Steps
- 1. thru 6.
No objection 7.a Clarification For the PRA subelements assigned a grade other than a Grade 3 in the NEI peer review (i.e., Grade 1, 2, or 4), a self-assessment of those PRA subelements required for the application against the Capability Category requirements (of the ASME PRA Standard as qualified in Appendix A to this regulatory guide) determined to be applicable for the application needs to be performed and documented. However, it is reasonable to assign an SR that requires that no Appendix B self-assessment received an NEI Grade 4 for Capability Category II without further review. [See Note (1) at end of Table B-2.].
7.b thru 8.
No objection
- 9.
No objection
- 10. thru 13.
No objection
Appendix B to DG 1200, Page B-10 Table B-2. NRC Regulatory Position on NEI Self-Assessment Process Section Position Commentary/Resolution
- 14.
Clarification The staffs comments on which ASME PRA requirements need to be addressed in the self-assessment, and on the suggested actions (Appendix D2 to NEI 00-02, Rev. 1) are provided in Table B-3.
In addition, the staffs position on the ASME PRA Standard, as documented in Appendix A to this regulatory guide, needs to be included in the self-assessment of the PRA subelements. [See Note (1) at end of Table B-2.].
Note (1):
The self-assessment in NEI 00-02 was performed against RA-S-2002 including Addenda A and B; the staff position is also based on RA-S-2002 (also including Addenda A and B). However, since the issuance of ASME RA-Sb-2005, both an addendum has been issued (RA-Sc-2007) and a revision (ASME/ANS RA-S-2008). These documents contain requirements that were either revised or added, as compared to RA-Sb-2005. Consequently, the comparison of the NEI subtier criteria is not complete because there may still exist requirements in ASME/ANS RA-S-2008 not addressed by the subtier criteria. Consequently, the staff has a global qualification in that the differences between RA-Sb-2005 and the applicable requirements in Parts 1 and 2 of the ASME/ANS PRA standard need to be addressed as part of the self-assessment.
Tables B-3 and B-4 provide the staff position on the NEI comparison of NEI 00-02 (including the subtier criteria) to the ASME PRA Standard Addendum B and the self-assessment actions provided in Appendix D2 to NEI 00-02, Revision 1.20 The staffs position on the ASME PRA Standard (Addendum B) documented in Appendix A to this regulatory guide was considered in the comparison. The review of the NEI comparison and proposed actions was performed under the assumption that all of the requirements in the NEI subtier criteria were treated as mandatory. Thus, the staff position is predicated on the requirement that all of the requirements in the NEI subtier criteria are interpreted as shall being required. See Note (1) to Table B-2.
Table B-3 provides the staff position of the explanatory table preceding the comparison and self-assessment actions table provided in Appendix D2. The first two columns are taken directly from the table in Appendix D2.
20 The NEI self-assessment process was revised to address the requirements in Addendum B of the ASME standard.
Appendix B to DG 1200, Page B-11 Table B-3. NRC Regulatory Positions on Actions Utilities Need to Take in Self-Assessment Actions Text Utility Actions Regulatory Position Comment/Resolution YES and NONE in Action column None No objection YES and clarifications included in Action column Take action(s) specified in the comments column.
No objection PARTIAL Take action(s) specified in Comments column.
No objection NO Take action(s) specified in Comments column.
No objection In Table B-4, the NEI Assessment includes, for each supporting requirement in the ASME standard (column heading: ASME STD SR)21:
whether NEIs assessment of each SR is addressed in NEI 00-02 (column heading: Addressed by NEI 00-02?)
if it is addressed in NEI 00-02, then where it is addressed is identified (column heading:
Applicable NEI 00-02 Elements) whether NEI recommends any self-assessment by the licensee (column heading: Industry Self-Assessment Actions)
In summary, following completion of the industry self-assessment actions, as augmented by the regulatory position for all applicable NEI Grade 3 sub-elements (and Grade 4 if no self-assessment specified), the corresponding SR may be considered to have met Capability Category II requirements of the standard except as noted by the staffs global qualification (see global qualification to Table B-1 and Note (1) to Table B-2). For NEI sub-elements receiving other grades, a self-assessment against the capability category requirements of the ASME standard (with Appendix A modifications) will determine the capability category for the corresponding SR.
21 See Note (1) to Table B-2.
Appendix B to DG 1200, Page B-12 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position Global In performing the self-assessment action, the action has to conform with the staff position in Appendix A of this document for the action to be acceptable the self-assessment has to account for the differences between the NEI subties criteria with the requirements in Part 2 of the ASME/ANS PRA standard (as endorsed in Appendix A of this document) as opposed to the ASME standard (RA-Sb-2005)
Initiating Events IE-A1 Yes IE-7, IE-8, IE-9, IE-10 None No objection IE-A2 Yes IE-5, IE-7, IE-9, IE-10 Confirm that the initiators [including human-induced initiators, and steam generator tube rupture (PWRs)] were included. This can be done by citing either peer review documentation/conclu sions or examples from your model.
NEI 00-02 does not explicitly mention human-induced initiators; however, in practice, peer reviews have addressed this; the definition of active component provided in the Addendum B of the ASME standard needs to be used when verifying ISLOCAs were modeled.
No objection IE-A3 Yes IE-8, IE-9 None No objection IE-A3a(1)
Yes IE-8, IE-9 None No objection
Appendix B to DG 1200, Page B-13 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IE-A4 Partial IE-5, IE-7, IE-9, IE-10 Check for initiating events that can be caused by a train failure or a system failure.
No objection IE-A4a(1)
Partial IE-5, IE-7, IE-9, IE-10 Check for initiating events that can be caused by multiple failures, if the equipment failures result from a common cause or from routine system alignments.
No objection IE-A5 Yes IE-8 Confirm requirement met. Identification of low-power and shutdown events not explicitly addressed in NEI 00-02, but in practice, the peer reviews have addressed events resulting in a controlled shutdown that include a scram prior to reaching low power.
No objection
Appendix B to DG 1200, Page B-14 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IE-A6 No Confirm requirement met. Specifying plant operations (etc.)
review and participation is not explicitly addressed in NEI 00-02, but in practice, the peer reviews have addressed the need for examination of plant experience (e.g.,
LERs), and input from knowledgeable plant personnel. Interviews conducted at similar plants are not acceptable.
No objection IE-A7 Yes IE-16, IE-10 None No objection IE-A8 Deleted from ASME PRA Standard IE-A9 Deleted from ASME PRA Standard IE-A10 Yes IE-6 None No objection IE-B1 Yes AS-4, IE-4 None No objection IE-B2 Yes IE-4, IE-7 None No objection IE-B3 Yes IE-4, IE-12 Confirm that the grouping does not impact significant accident sequences.
No objection IE-B4 Yes IE-4 None No objection IE-B5(3)
Yes IE-6 None No objection IE-C1 Yes IE-13, IE-15, IE-16, IE-17 None No objection; IE-16 is the applicable NEI 00-02 element.
Appendix B to DG 1200, Page B-15 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IE-C1a(1)
Yes IE-13, IE-15, IE-16, IE-17 None No objection; IE-16 is the applicable NEI 00-02 element.
IE-C1b(1)
Yes IE-13, IE-15, IE-16, IE-17 Justify recovery credit as evidenced by procedures or training.
No objection IE-C2 Yes IE-13, IE-16 Justify informative priors used in Bayesian update.
No objection IE-C3 No Document that the ASME standard requirements were met. NEI 00-02 does not address this supporting requirement.
No objection IE-C4 No Document that the ASME standard requirements were met. Specific screening criteria were not used in NEI 00-02, but bases for screening of events were examined in the peer reviews. The text of the ASME standard needs to be assessed.
Acceptable criteria for dismissing IEs are listed in IE-C4 in the ASME PRA Standard.
No objection IE-C5 No requirement for Category II N/A No objection; the ASME PRA Standard only requires time trend analysis for a Category III PRA.
Appendix B to DG 1200, Page B-16 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IE-C6 Yes IE-15, IE-17 Check that fault tree analysis, when used to quantify IEs, meets the appropriate systems analysis requirements.
No objection IE-C7 No Document that the ASME standard requirements were met. NEI 00-02 does not address this supporting requirement.
No objection IE-C8 No Document that the ASME standard requirements were met. NEI 00-02 does not address this supporting requirement.
No objection IE-C9 Yes IE-15, IE-16 Check that the recovery events included in the IE fault trees meet the appropriate recovery analysis requirements.
This can be done by citing either peer review documentation/conclu sions or examples from your model.
No objection IE-C10 Yes IE-13 None No objection IE-C11 Yes IE-12, IE-13, IE-15 Check that the expert elicitation requirements in the ASME PRA Standard were used when expert judgment was applied to quantifying extremely rare events.
No objection
Appendix B to DG 1200, Page B-17 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IE-C12 Yes IE-14 Confirm that secondary pipe system capability and isolation capability under high flow or differential pressures are included.
No objection IE-C13(3)
No None Confirm IE-C13 is met.
No objection IE-D1 Partial IE-9, IE-18, IE-19, IE-20 Action is to confirm availability of documentation. In general, specified documentation items not explicitly addressed in NEI 00-02 checklists were addressed by the peer review teams. If not available, documentation may need to be generated to support particular applications or respond to NRC requests for additional information (RAIs) regarding applications.
No objection
Appendix B to DG 1200, Page B-18 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IE-D2 Partial IE-9, IE-18, IE-19, IE-20 Action is to confirm availability of documentation. In general, specified documentation items not explicitly addressed in NEI 00-02 checklists were addressed by the peer review teams. If not available, documentation may need to be generated to support particular applications or respond to NRC RAIs regarding applications.
No objection IE-D3 Partial QU-27, QU-28, QU-29, QU-34 Confirm that the key assumptions and key sources of uncertainty consistent with the definitions of the ASME PRA Standard are documented.
No objection with Clarification: See staff position on definition of key assumption and key source of uncertainty in Appendix A.
IE-D4 Deleted from ASME PRA Standard Accident Sequence Analysis AS-A1 Yes AS-4, AS-8 None No objection AS-A2 Yes AS-6, AS-7, AS-8, AS-9, AS-17 None No objection AS-A3 Yes AS-7, SY-17, AS-17 None No objection AS-A4 Yes AS-19, SY-5 None No objection AS-A5 Yes AS-5, AS-18, AS-19, SY-5 None No objection AS-A6 Yes AS-8, AS-13, AS-4 None No objection
Appendix B to DG 1200, Page B-19 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position AS-A7 Yes AS-4, AS-5, AS-6, AS-7, AS-8, AS-9 None No objection AS-A8 Partial AS-20, AS-21, AS-22, AS-23 Since there is no explicit requirement for steady-state condition for end state in NEI 00-02 checklists, this should be evaluated even though this was an identified issue in some reviews. This can also be done by citing either peer review documentation/conclu sions or examples from your model.
Refer to SC-A5.
No objection AS-A9 Yes AS-18, TH-4 Verify AS-A9 is met.
Note that AS-A9 is related to the environmental conditions challenging the equipment during the accident sequence, AS-18 and TH-4 are focused on the initial success criteria.
No objection AS-A10 Yes AS-4, AS-5, AS-6, AS-7, AS-8, AS-9, AS-19, SY-5, SY-8, HR-23 None No objection
Appendix B to DG 1200, Page B-20 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position AS-A11 Yes AS-8, AS-10, AS-15, DE-6, AS Checklist Note 8 The guidance in AS-15 must be followed.
AS-8 states that transfers may be treated quantitatively or qualitatively while AS-15 states that transfers between event trees should be explicitly treated in the quantification.
No objection AS-B1 Yes IE-4, IE-5, IE10, AS-4, AS-5, AS-6, AS-7, AS-8, AS-9, AS-10, AS-11, DE-5 None No objection AS-B2 Yes AS-10, AS-11, DE-4, DE-5, DE-6 None No objection; AS-10 and AS-11 are the applicable NEI 00-02 elements.
AS-B3 Yes DE-10, SY-11, TH-8, AS-10 None No objection; AS-10 and SY-11 are the applicable NEI 00-02 elements.
AS-B4 Yes AS-8, AS-9, AS-10, AS-11 Confirm requirement met.
No objection AS-B5 Yes DE-4, DE-5, DE-6, AS-10, AS-11, QU-25 None No objection elements.
AS-B5a(1) Yes DE-4, DE-5, DE-6, AS-10, AS-11, QU-25 Confirm that system alignments that may affect dependencies among systems or functions are modeled.
No objection AS-B6 Yes AS-13 None No objection AS-C1(2)
Yes AS-11, AS-24, AS-25, AS-26 None No objection
Appendix B to DG 1200, Page B-21 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position AS-C2(2)
Partial AS-11, AS-24, AS-25, AS-26 Action is to confirm availability of documentation. In general, specified documentation items not explicitly addressed in NEI 00-02 checklists were addressed by the peer review teams. If not available, documentation may need to be generated to support particular applications or respond to NRC RAIs regarding applications.
No objection AS-C3(2)
Partial QU-27, QU-28, QU-29, QU-34 Confirm that the key assumptions and key sources of uncertainty consistent with the definitions of the ASME PRA Standard are documented.
No objection with Clarification: See staff position on definition of key assumption and key source of uncertainty in Appendix A.
AS-C4 Deleted from ASME PRA Standard Success Criteria SC-A1 Yes AS-20, AS-22, AS Footnote 4 None No objection SC-A2 Yes TH-4, TH-5, TH-7, AS-22, AS Footnote 4 None No objection SC-A3 Deleted from ASME PRA Standard SC-A4 Yes AS-7, AS-17, AS-18, SY-17, TH-9, IE-6, DE-5, SY-8 None No objection
Appendix B to DG 1200, Page B-22 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position SC-A4a(1) Yes IE-6, DE-5 Confirm that this requirement is met.
This can be done by citing either peer review documentation conclusions or examples from your model. Although there is no explicit requirement in NEI 00-02 that mitigating systems shared between units be identified, in practice, review teams have evaluated this.
No objection SC-A5 Partial AS-21, AS-23, AS-20 Ensure mission times are adequately discussed as per the ASME PRA Standard.
Since there are no explicit requirements for steady-state condition for end state, refer to the ASME PRA Standard for requirements or cite peer review documentation/
conclusions or examples from your model. Refer to AS-A8.
No objection SC-A6 Yes AS-5, AS-18, AS-19, TH-4, TH-5, TH-6, TH-8, ST-4, ST-5, ST-7, ST-9, SY-5 None No objection SC-B1 Yes AS-18, SY-17, TH-4, TH-6, TH-7 None No objection
Appendix B to DG 1200, Page B-23 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position SC-B2 No TH-4, TH-8 NEI 00-02 does not address this supporting requirement. Use the ASME standard for requirements. Refer to SC-C2.
No objection SC-B3 Yes AS-18, TH-4, TH-5, TH-6, TH-7 None No objection SC-B4 Yes AS-18, TH-4, TH-6, TH-7 None No objection SC-B5 Yes TH-9, TH-7 None No objection SC-B6 Deleted from ASME PRA Standard SC-C1(2)
Yes ST-13, SY-10, SY-17, SY-27, TH-8, TH-9, TH-10, AS-17, AS-18, AS-24, HR-30 None No objection SC-C2(2)
Partial ST-13, SY-10, SY-17, SY-27, TH-8, TH-9, TH-10, AS-17, AS-18, AS-24, HR-30 Action is to confirm availability of documentation. In general, specified documentation items not explicitly addressed in NEI 00-02 checklists were addressed by the peer review teams. If not available, documentation may need to be generated to support particular applications or respond to NRC RAIs regarding applications.
No objection
Appendix B to DG 1200, Page B-24 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position SC-C3(2)
Partial QU-27, QU-28, QU-29, QU-34 Confirm that the key assumptions and key sources of uncertainty consistent with the definitions of the ASME PRA Standard are documented.
No objection with Clarification: See staff position on definition of key assumption and key source of uncertainty in Appendix A.
SC-C4 Deleted from ASME PRA Standard Systems Analysis SY-A1 Yes SY-4, SY-19 None No objection SY-A2 Yes AS-19, SY-5, SY-13, SY-16 None No objection SY-A3 Yes SY-5, SY-6, SY-8, SY-12, SY-14 None. Although there are no explicit requirements in NEI 00-02 that match SY-A3, performance of the systems analysis would require a review of plant-specific information sources.
No objection SY-A4 Partial DE-11, SY-10, SY Footnote 5 Confirm that this requirement is met.
This can be done by citing either peer review results or example documentation. NEI 00-02 does not address interviews with system engineers and plant operators to confirm that the model reflects the as-built, as-operated plant.
No objection
Appendix B to DG 1200, Page B-25 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position SY-A5 Partial QU-12, QU-13, SY-8, SY-11 Confirm this requirement is met, and that the PRA considered both normal and abnormal system alignments.
This can be done by citing either peer review results or example documentation.
Although NEI 00-02 does not explicitly address both normal and abnormal alignments, their impacts are generally captured in the peer review of the listed elements.
No objection SY-A6 Yes SY-7, SY-8, SY-12, SY-13, SY-14 None No objection SY-A7 Yes SY-6, SY-7, SY-8, SY-9, SY-19 Check for simplified system modeling as addressed in SY-A7.
No objection
Appendix B to DG 1200, Page B-26 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position SY-A8 Partial SY-6, SY-9 Check to ensure boundaries are properly established.
This can be done by citing either peer review results or example documentation. NEI 00-02 does not address component boundaries except for EDGs. There is no explicit requirement that addresses modeling shared portions of a component boundary.
In practice, the peer reviews have examined consistency of component and data analysis boundaries.
No objection SY-A9 Deleted from ASME PRA Standard SY-A10 Partial SY-9 Action is to determine if the requirements of the ASME standard are met. NEI 00-02 does not address all aspects of modularization.
No objection SY-A11 Yes AS-10, AS-13, AS-16, AS-17, AS-18, SY-12, SY-13, SY-17, SY-23 None No objection
Appendix B to DG 1200, Page B-27 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position SY-A12 Partial SY-6, SY-7, SY-8, SY-9, SY-12, SY-13, SY-14 Document that modeling is consistent with exclusions provided in SY-A14.
Consistent with subelement SY-A12 of the ASME PRA Standard, critical passive components whose failure affects system operability should be included in system models.
No objection SY-A12a(1) Partial SY-6, SY-7, SY-8, SY-9, SY-12, SY-13, SY-14 Document that modeling is consistent with exclusions provided in SY-A12a.
No objection SY-A12b(3)
Partial SY-15, SY-17 Document that modeling incorporates flow diversion failure modes.
No objection SY-A13 Yes DA-4, SY-15, SY-16 None No objection SY-A14 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection SY-A15 Yes SY-8, HR-4, HR-5, HR-7 None No objection SY-A16 Yes SY-8, HR-8, HR-9, HR-10 None No objection
Appendix B to DG 1200, Page B-28 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position SY-A17 Yes AS-13, SY-10, SY-11, SY-13, SY-17 None. SY-A17 is evaluated in the NEI 00-02 PRA peer review as follows:
SY-10 Failures or system termination (trip) due to spatial or environmental effects.
SY-11 Failure modes induced by accident conditions.
SY-13 System Termination (failure or trip) due to exhaustion of inventory (water, air).
SY-17 Success Criteria evaluation determined by plant-specific analysis that includes system trips or isolations on plant parameters.
AS-13 Failure of systems due to time phased effects such as loss of battery voltage.
No objection SY-A18 Yes DA-7, SY-8, SY-22 None No objection SY-A18a(3) No Confirm this is accounted for in the PRA. NEI 00-02 does not explicitly identify the criteria for tracking and modeling of coincident maintenance actions that may lead to unavailability of multiple redundant trains or systems.
No objection
Appendix B to DG 1200, Page B-29 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position SY-A19 Yes AS-18, DE-10, SY-11, SY-13, SY-17, TH-8 Verify SY-A19 has been met. Ensure there is a documented basis (engineering calculations are not necessary) for modeling of the conditions addressed.
NEI 00-02 focuses on environmental limitations.
No objection SY-A20 Partial AS-19, SY-5, SY-11, SY-13, SY-22, TH-8 Document component capabilities where applicable. NEI 00-02 does not explicitly require a check for crediting components beyond their design basis.
No objection SY-A21 Yes SY-18 None. Comment:
Footnote to SY-18 explains lack of Grade provision for this sub-element.
No objection SY-A22 Yes SY-24, DA-15, QU-18, SY-12 None No objection SY-A23 Deleted from ASME PRA Standard SY-B1 Yes DA-8, DA-14, DE-8, DE-9, SY-8 None No objection SY-B2 Not required for Capability Category II None No objection SY-B3 Yes DE-8, DE-9, DA-10, DA-12 None No objection
Appendix B to DG 1200, Page B-30 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position SY-B4 Yes DA-8, DA-10, DA-11, DA-12, DA-13, DA-14, DE-8, DE-9, QU-9, SY-8 None No objection SY-B5 Yes DE-4, DE-5, DE-6, SY-12, None No objection SY-B6 Yes SY-12, SY-13 Self-assessment needs to confirm that the support system success criteria reflect the variability in the conditions that may be present during postulated accidents.
No objection SY-B7 Yes AS-18, SY-13, SY-17, TH-7, TH-8 None No objection SY-B8 Yes DE-11, SY-10 None No objection SY-B9 Deleted from ASME PRA Standard SY-B10 Yes SY-12, SY-13 None No objection SY-B11 Yes SY-8, SY-12, SY-13 Confirm by citing either peer review documentation/conclu sions or examples from your model.
NEI 00-02 does not explicitly address permissives and control logic. In practice, the items in SY-B11 have generally been examined in the peer reviews.
No objection SY-B12 Yes SY-13 None No objection
Appendix B to DG 1200, Page B-31 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position SY-B13 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection SY-B14 Partial DE-6, AS-6 Confirm by citing either peer review documentation/conclu sions or examples from your model.
Ensure that modeling includes situations where one component can disable more than one system.
No objection SY-B15 Yes SY-11 None No objection SY-B16 Yes SY-8 None No objection SY-C1(2)
Yes SY-5, SY-6, SY-9, SY-18, SY-23, SY-25, SY-26, SY-27 None No objection
Appendix B to DG 1200, Page B-32 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position SY-C2(2)
Partial SY-5, SY-6, SY-9, SY-18, SY-23, SY-25, SY-26, SY-27 Action is to confirm availability of documentation. In general, specified documentation items not explicitly addressed in NEI 00-02 checklists were addressed by the peer review teams. If not available, documentation may need to be generated to support particular applications or respond to NRC RAIs regarding applications.
Comment: Footnote to SY-18 explains lack of Grade provision for this sub-element.
No objection SY-C3(2)
Partial QU-27, QU-28, QU-29, QU-34 Confirm that the key assumptions and key sources of uncertainty consistent with the definitions of the ASME PRA Standard are documented.
No objection with Clarification: See staff position on definition of key assumption and key source of uncertainty in Appendix A.
Human Reliability Analysis HR-A1 Yes HR-4, HR-5 Determine if analysis has included and documented failure to restore equipment following test or maintenance.
No objection HR-A2 Yes HR-4, HR-5 None No objection HR-A3 Yes DE-7, HR-5 None No objection HR-B1 Yes HR-5, HR-6 None No objection
Appendix B to DG 1200, Page B-33 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position HR-B2 Partial HR-5, HR-6, HR-7, HR-26, DA-5, DA-6 Ensure single actions with multiple train consequences are evaluated in pre-initiators, since the screening rules in HR-6 do not preclude screening of activities that can affect multiple trains of a system.
No objection HR-C1 Yes HR-27, SY-8, SY-9 None No objection HR-C2 Yes HR-7, HR-27, SY-8, SY-9 Confirm that this requirement is met.
The specific list of impacts in HR-C2 is not included in NEI 00-02; however, in practice, the peer reviewers (in reviewing sub-elements HR-7 and related sub-elements) addressed these items.
No objection HR-C3 Yes HR-5, HR-27, SY-8, SY-9 None No objection HR-D1 Yes HR-6 None No objection HR-D2 Yes HR-6 None No objection
Appendix B to DG 1200, Page B-34 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position HR-D3 No Action is to confirm that HR-D3 is met.
This item is implicitly included in the peer review of HRA by virtue of the assessment of the crews ability to implement the procedure in an effective and controlled manner.
The pre-initiator HRA adequacy is determined reasonable and representative considering the procedure quality.
No objection HR-D4 Partial HR-6 Use the ASME standard for requirements. NEI 00-02 does not explicitly cite the treatment of recovery actions for pre-initiators. PRA implementation varied among utilities with some using screening values and others incorporating recovery. The peer review team examines this treatment.
No objection HR-D5 Yes DE-7, HR-26, HR-27 None No objection
Appendix B to DG 1200, Page B-35 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position HR-D6 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection HR-D7 Not required for Capability Category II None No objection HR-E1 Yes AS-19, HR-9, HR-10, HR-16, SY-5 None No objection; the example process in HR-9 for a Grade 3 PRA (i.e., identify those operator actions identified by others) is not good practice and contrary to HR-10, which is the process recommended in HR-E1.
HR-E2 Yes HR-8, HR-9, HR-10, HR-21, HR-22, HR-23, HR-25 None No objection (HR-9 and HR-10 do not appear to match subject matter but HR-8 does).
HR-E3 Partial HR-10, HR-14, HR-20 The ASME standard supporting requirements are to be used during the self-assessment to confirm that the ASME intent is met for this requirement. NEI 00-02 does not explicitly specify the same level of detail that is included in the ASME standard. The peer review team experience is relied upon to investigate the PRA given general guidance and criteria.
No objection
Appendix B to DG 1200, Page B-36 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position HR-E4 Partial HR-14, HR-16 The ASME standard supporting requirements are to be used during the self-assessment to confirm that the ASME intent is met for this requirement. NEI 00-02 does not explicitly specify the same level of detail that is included in the ASME standard. The peer review team experience is relied upon to investigate the PRA given general guidance and criteria.
No objection HR-F1 Yes AS-19, HR-16, SY-5 None No objection HR-F2 Partial AS-19, HR-11, HR-16, HR-17, HR-19, HR-20, SY-5 Determine whether the requirements of the ASME standard are met. HR-F2 is generally addressed by NEI 00-02 and the PRA Peer Review.
One additional item is highlighted to be checked. NEI 00-02 does not explicitly cite indication for detection and evaluation. However, by invoking the standard HRA methodologies the treatment of cues and other indications for detecting the need for action are included.
No objection
Appendix B to DG 1200, Page B-37 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position HR-G1 Yes HR-15, HR-17, HR-18 None No objection HR-G2 Yes HR-2, HR-11 None. NEI 00-02 criteria for Grade 3 require a methodology that is consistent with industry practice.
This includes the incorporation of both the cognitive and execution (human error probabilities) in the HEP assessment.
HR-11 provides further criteria to ensure that the cognitive portion of the HEP uses the correct symptoms to formulate the crews response. Self-assessment needs to document if both cognitive and execution errors are included in the evaluation of HEPs.
No objection
Appendix B to DG 1200, Page B-38 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position HR-G3 Partial HR-17, HR-18 The ASME standard supporting requirements are to be used during the self-assessment to confirm that the ASME intent is met for this requirement. NEI 00-02 does not explicitly enumerate the same level of detail that is included in the ASME standard. However, by invoking the standard HRA methodologies the performance shape factors are necessarily evaluated. The peer review team experience is relied upon to investigate the PRA given general guidance and criteria.
No objection
Appendix B to DG 1200, Page B-39 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position HR-G4 Partial AS-13, HR-18, HR-19, HR-20 The ASME standard supporting requirements are to be used during the self-assessment to confirm that the ASME intent is met for this requirement. NEI 00-02 does not explicitly cite the necessity to define the time at which operators are expected to receive indications. However, invoking the standard HRA methods leads to the necessity for the analysts to define this input to the HRA.
The peer review team experience is relied upon to investigate the PRA given general guidance and criteria.
No objection HR-G5 Partial HR-16, HR-18, HR-20 Evaluate proper inputs per the ASME standard or cite peer review documentation/conclu sions or examples from your model.
NEI 00-02 explicitly addresses observations and operations staff input for time required. ASME PRA Standard requires time measurements.
No objection
Appendix B to DG 1200, Page B-40 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position HR-G6 Yes HR-12 Check to ensure they are met by citing peer review documentation/conclu sions or examples from your model.
HR-12 does not explicitly address all the items of the ASME standard list.
In practice, peer reviews addressed these items.
No objection
Appendix B to DG 1200, Page B-41 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position HR-G7 Partial DE-7, HR-26 Check to see if factors that are typically assumed to lead to dependence were included (e.g., use of common indications and/or cues to alert control room staff to need for action), and a common procedural direction that leads to the actions. This can also be done by citing either peer review documentation/
conclusions or examples from your model. NEI 00-02 does not provide explicit criteria that address the degree of dependence between HFEs that appear in the same accident sequence cutset.
However, invoking the standard HRA methods leads to the necessity for the analysts to define this input to the HRA. In general, the peer reviews addressed this. See also QU-C2.
No objection HR-G8 Not required for Capability Category II
Appendix B to DG 1200, Page B-42 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position HR-G9 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection HR-H1 Yes HR-21, HR-22, HR-23 The self-assessment needs to confirm that the requirements in HR-H1 in the ASME standard were addressed in the HRA.
No objection HR-H2 Yes HR-22, HR-23 The self-assessment needs to confirm that all the requirements of HR-H2 in the ASME standard were included in the HRA.
No objection HR-H3 Yes HR-26 None No objection HR-I1(2)
Partial HR-28, HR-30 None No objection HR-I2(2)
Partial HR-28, HR-30 Action is to confirm availability of documentation. In general, specified documentation items not explicitly addressed in NEI 00-02 checklists were addressed by the peer review teams. If not available, documentation may need to be generated to support particular applications or respond to NRC RAIs regarding applications.
No objection
Appendix B to DG 1200, Page B-43 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position HR-I3(2)
Partial QU-27, QU-28, QU-29, QU-34 Confirm that the key assumptions and key sources of uncertainty consistent with the definitions of the ASME PRA Standard are documented.
No objection with Clarification: See staff position on definition of key assumption and key source of uncertainty in Appendix A.
Data Analysis DA-A1 Yes DA-4, DA-5, DA-15, SY-8, SY-14 None No objection DA-A1a(1) No Confirm that the component boundary is consistent with the data applied.
No objection DA-A2 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection DA-A3 Yes DA-4, DA-5, DA-6, DA-7, SY-8 None No objection with Qualification: The subject matter in DA-A3 is not explicitly addressed in NEI 00-02 (not a critical requirement since identification of the needed parameters would be a natural part of the data analysis).
DA-B1 Yes DA-5 None No objection
Appendix B to DG 1200, Page B-44 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position DA-B2 Yes DA-5, DA-6 Confirm that this requirement is met.
NRC comment:
Grouping criteria listed in DA-5 should be supplemented with a caution to look for unique components and/or operating conditions and to avoid grouping them.
Peer review teams were careful to assess plant-specific data evaluations to identify cases where outlier data values or components were not properly accounted for.
No objection DA-C1 Yes DA-4, DA-7, DA-9, DA-19, DA-20 None No objection DA-C2 Yes DA-4, DA-5, DA-6, DA-7, DA-14, DA-15, DA-19, DA-20, MU-5 None No objection DA-C3 Partial DA-4, DA-5, DA-6, DA-7, MU-5 Use the ASME standard for requirements. NEI 00-02 does not enumerate the items considered appropriate in a plant-specific data analysis.
No objection DA-C4 No NEI 00-02 does not explicitly cite this definition of failure and degraded state.
Use the ASME standard for requirements.
No objection
Appendix B to DG 1200, Page B-45 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position DA-C5 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection DA-C6 Yes DA-6, DA-7 Confirm that this requirement is met.
NEI 00-02 addresses data needs when the standby failure rate model is used for demands. There are no stated criteria for the demand failure model; however, in practice, this was addressed during peer reviews.
No objection DA-C7 Yes DA-6, DA-7 None No objection DA-C8 Yes Confirm that this requirement is met.
Although there are no specific criteria for determining operational time of components in operation or in standby, the development needs to include these times.
These issues were addressed during peer reviews.
No objection
Appendix B to DG 1200, Page B-46 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position DA-C9 Yes DA-4, DA-6, DA-7 Confirm that this requirement is met.
Although there are no specific criteria for determining operational time of components in operation or in standby, the development needs to include these times.
These issues were addressed during peer reviews.
No objection DA-C10 No NEI 00-02 does not address this supporting requirement. Use the ASME standard for requirements.
No objection DA-C11 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection DA-C11a(3)
No Use the ASME PRA Standard for requirements. PRA peer review teams found that support system unavailabilities are treated within the support system and not within the associated frontline system.
No objection
Appendix B to DG 1200, Page B-47 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position DA-C12 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection DA-C13 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection DA-C14 Yes DA-15, AS-16, SY-24 None No objection DA-C15 Yes IE-13, IE-15, IE-16, AS-16, DA-15, SY-24, QU-18 Confirm that this requirement is met.
Although it is relatively rare to see credit taken for repair of failed equipment in PRAs (except in modeling of support system initiating events), any credit taken for repair should be well-justified, based on ease of diagnosis, the feasibility of repair, ease of repair, and availability of resources, time to repair and actual data.
This can be done by citing either peer review results or example documentation.
No objection
Appendix B to DG 1200, Page B-48 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position DA-D1 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection DA-D2 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection DA-D3 Partial QU-30 Verify that SR DA-D3 has been met. A requirement for establishing the parameter distributions is not in the data analysis section but could be inferred from QU-30.
QU-30 does not provide guidance on which events to include in the uncertainty analysis.
No objection DA-D4 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement. This was performed as part of the peer review team implementation of NEI 00-02. (See DE-9.)
No objection
Appendix B to DG 1200, Page B-49 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position DA-D5 Partial DE-9, DA-8, DA-9, DA-10, DA-11, DA-12, DA-13, DA-14 Check for acceptable common-cause failure models. This can be done by citing either peer review documentation/conclu sions or example documentation. This was performed as part of the peer review team implementation of NEI 00-02. (See DE-9.) The criteria for NEI 00-02 elements DA-13 and DA-14 only apply to Grade 4.
No objection DA-D6 Partial DE-9, DA-8, DA-9, DA-10, DA-11, DA-12, DA-13, DA-14 None No objection DA-D6a(3) Partial (see Self-Assessment Action)
DA-14 Plant-specific screening and mapping of industry-wide data is not required for Capability Category II.
However, if this approach is used, DA-D6a should be confirmed to be met.
If it is performed, see DE-9 from NEI 00-02.
No objection DA-D7 No Use the ASME standard for requirements. NEI 00-02 does not specifically address how to deal with data for equipment that has been changed.
No objection
Appendix B to DG 1200, Page B-50 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position DA-E1(2)
Partial DA-1, DA-19, DA-20, DE-9 None No objection DA-E2(2)
Partial DA-1, DA-19, DA-20, DE-9 Action is to confirm availability of documentation. In general, specified documentation items not explicitly addressed in NEI 00-02 checklists were addressed by the peer review teams. If not available, documentation may need to be generated to support particular applications or respond to NRC RAIs regarding applications.
No objection DA-E3(2)
Partial QU-27, QU-28, QU-29, QU-34 Confirm that the key assumptions and key sources of uncertainty consistent with the definitions of the ASME PRA Standard are documented.
No objection with Clarification: See staff position on definition of key assumption and key source of uncertainty in Appendix A.
Internal Flooding IF-A1 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-A1a(1)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection
Appendix B to DG 1200, Page B-51 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IF-A1b(1)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-A2 ASME PRA Deleted from Standard IF-A3 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-A4 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-B1 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-B1a(4)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection
Appendix B to DG 1200, Page B-52 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IF-B1b(3)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-B2 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-B3 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-B3a(3)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-B4 Deleted from ASME PRA Standard IF-C1 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection
Appendix B to DG 1200, Page B-53 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IF-C2 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-C2a(1)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-C2b(2)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-C2c(5)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-C3 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection
Appendix B to DG 1200, Page B-54 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IF-C3a(1)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-C3b(3)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-C3c(6)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-C4 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-C4a(4)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection
Appendix B to DG 1200, Page B-55 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IF-C5 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-C5a(1)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-C6 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-C7(3)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-C8(3)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection
Appendix B to DG 1200, Page B-56 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IF-C9(3)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-D1 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-D2 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-D3 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-D3a(3)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection
Appendix B to DG 1200, Page B-57 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IF-D4 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-D5 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-D5a(1)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-D6(3)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-D7(3)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection
Appendix B to DG 1200, Page B-58 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IF-E1 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-E2 Deleted from ASME PRA Standard IF-E3 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-E3a(3)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-E4 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-E5 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection
Appendix B to DG 1200, Page B-59 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IF-E5a(1)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-E6 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-E6a(1)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-E6b(1)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-E7 No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection
Appendix B to DG 1200, Page B-60 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position IF-E8(3)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-F1(2)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-F2(2)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection IF-F3(2)
No Use the ASME standard for requirements. NEI 00-02 does not address this supporting requirement.
No objection Quantification Analysis QU-A1 Yes AS-4, AS-5, AS-6, AS-7, AS-8, AS-9, AS-10, AS-19 The requirement in QU-A1 is not explicitly stated in any element, but is achieved through compliance with the identified NEI 00-02 elements and others that support complying with those elements.
No objection
Appendix B to DG 1200, Page B-61 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position QU-A2a Yes QU-8 None No objection QU-A2b(1) No ASME PRA Standard SR should be addressed. State of knowledge correlation is not explicitly cited in NEI 00-02 to be checked.
No objection QU-A3 Yes QU-4, QU-8, QU-9, QU-10, QU-11, QU-12, QU-13 The requirement in QU-A3 is not explicitly stated in any element, but is achieved through compliance with the identified NEI 00-02 elements and others that support complying with those elements.
No objection QU-A4 Yes QU-18, QU-19 None No objection QU-B1 Yes QU-6 None No objection QU-B2 Yes QU-21, QU-22, QU-23, QU-24 Confirm that this requirement is met. In practice, the industry peer reviews have generally used the stated guidance as a check on the final cutset level quantification truncation limit applied in the PRA.
No objection; QU-21 and QU-23 are the relevant elements that address the requirements in QU-B2 while the remaining NEI 00-02 elements provide additional guidance on truncation. It is not clear what events and failure modes are being addressed in QU-22. If the element is referring to a cutset truncation limit, then the values presented are reasonable.
Appendix B to DG 1200, Page B-62 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position QU-B3 Partial QU-21, QU-22, QU-23, QU-24 The self-assessment should confirm that the final truncation limit is such that convergence toward a stable CDF is achieved.
No objection QU-B4 Yes QU-4 None No objection. Although the stated purpose of the criterion for QU-4 is to verify that the base computer code and its inputs have been tested and demonstrated to produce reasonable results, the subtier criteria do not address this criterion, but instead provide some dos and donts for quantification.
QU-B5 Yes QU-14 None No objection QU-B6 Yes AS-8, AS-9, QU-4, QU-20, QU-25 Check for proper accounting of success terms. The NEI 00-02 guidance adequately addresses this requirement, but QU-25 should not be restricted to addressing just delete terms.
No objection QU-B7a Yes QU-26 None No objection QU-B7b(1) Yes QU-26 None No objection
Appendix B to DG 1200, Page B-63 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position QU-B8 No Use the ASME standard for requirements. NEI 00-02 does not explicitly cite the details of Boolean logic code implementation.
No objection QU-B9 Partial SY-9 The warnings in SY-A10 must be considered in the modularization process. SYSA addresses the traceability of basic events in modules but does not address the correct formulation of modules that are truly independent.
No objection QU-C1 Yes QU-10, QU-17, HR-26, HR-27 None No objection QU-C2 Yes QU-10, QU-17 Verify dependencies in cutsets/sequences are assessed. Verify that dependence between the HFEs in a cutset or sequence is assessed in accordance with ASME SRs HR-D5 and HR-G7.
No objection QU-C3 Yes QU-20 Confirm that this requirement is met.
QU-20 does not explicitly require that the critical characteristic, not just the frequency, be transferred; however, in practice, this was addressed during peer reviews.
No objection
Appendix B to DG 1200, Page B-64 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position QU-D1a Yes QU-8, QU-9, QU-10, QU-11, QU-12, QU-13, QU-14, QU-15, QU-16, QU-17 None No objection; the requirements in QU-D1 are addressed primarily in QU-
- 8. The requirements in QU-9, QU-10, QU-14, QU-16, and QU-17 appear to be focused on modeling and not interpretation of results. As such, they are redundant to elements in the data, dependent failure, and HRA sections.
QU-D1b(1) Yes QU-8, QU-9, QU-10, QU-11, QU-12, QU-13, QU-14, QU-15, QU-16, QU-17, QU-23 None No objection; the requirements in QU-D1 are addressed primarily in QU-
- 8. The requirements in QU-9, QU-10, QU-14, QU-16, and QU-17 appear to be focused on modeling and not interpretation of results. As such, they are redundant to elements in the data, dependent failure, and HRA sections.
QU-D1c(1) Yes QU-8, QU-9, QU-10, QU-11, QU-12, QU-13, QU-14, QU-15, QU-16, QU-17 None No objection; the requirements in QU-D1 are addressed primarily in QU-
- 8. The requirements in QU-9, QU-10, QU-14, QU-16, and QU-17 appear to be focused on modeling and not interpretation of results. As such, they are redundant to elements in the data, dependent failure, and HRA sections.
QU-D2 Deleted from ASME PRA Standard
Appendix B to DG 1200, Page B-65 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position QU-D3 Yes QU-8, QU-11, QU-31 None No objection; consistency with other PRA results is addressed in QU-11 and QU-31.
QU-D4 Yes QU-15 None No objection QU-D5a Yes QU-8, QU-31 Confirm that this requirement is met.
The subject matter in QU-D5a is partially addressed in NEI 00-02 in element QU-31 (QU-8 checks the reasonableness of the results). The contributions from IEs, component failures, common-cause failures, and human errors are not addressed. In practice, these were addressed during peer reviews.
No objection QU-D5b(5) No Confirm that this requirement is met.
No objection
Appendix B to DG 1200, Page B-66 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position QU-E1 Yes QU-27, QU-28, QU-30 Confirm that QU-E1 is addressed. The definition of the sources of model uncertainty is provided by the ASME PRA Standard Addendum B. This nomenclature was not available when NEI 00-02 was implemented. The PRA Peer Review did examine the PRAs to see if modeling uncertainties were addressed appropriately.
No objection with Clarification: QU-30 does not provide guidance on sources of uncertainty.
See staff position on definition of key assumption and key source of uncertainty in Appendix A.
Appendix B to DG 1200, Page B-67 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position QU-E2 Yes QU-27, QU-28, QU-30 Confirm that this requirement is met.
QU-27 and QU-28 focus on the assumptions and unusual sources of uncertainty.
Assumptions and unusual sources of uncertainty correspond to plant-specific hardware, procedural, or environmental issues that would significantly alter the degree of uncertainty relative to plants that have previously been assessed, such as NUREG-1150 or the Risk Methodology Integration and Evaluation Program (RMIEP). Unusual sources of uncertainty could also be introduced by the PRA methods and assumptions. In practice, when applying NEI 00-02 sub-elements QU-27 and QU-28, the reviewers considered the appropriateness of the assumptions.
No objection.
Appendix B to DG 1200, Page B-68 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position QU-E3 Partial QU-30 The uncertainty band associated with each risk metric is to be estimated. The parametric uncertainty band is to be estimated taking into account the state of knowledge correlation. This was to be checked by the peer review team.
No objection QU-E4 Partial QU-28, QU-29, QU-30 Use the ASME standard for requirements. NEI 00-02 does not explicitly specify that sensitivity studies of logical combinations of assumptions and parameters be evaluated.
No objection QU-F1(2)
Partial QU-31, QU-32, QU-34 None No objection QU-F2(2)
Yes MU-7, QU-4, QU-12, QU-13, QU-27, QU-28, QU-31, QU-32 No action required for (m). Normal industry practice requires documentation of computer code capabilities. Confirm availability of documentation, or generate as necessary to support applications. Also needed to confirm computer code has been sufficiently verified such that there is confidence in the results.
No objection
Appendix B to DG 1200, Page B-69 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position QU-F3(2)
Partial QU-31 Use the ASME standard for requirements at the time of doing an application.
No objection QU-F4(2)
No QU-27, QU-28, QU-32 Use the ASME standard for requirements at the time of doing an application. NEI 00-02 does not address this supporting requirement.
No objection QU-F5(2)
No Use the ASME standard for requirements at the time of doing an application. NEI 00-02 does not address this supporting requirement.
No objection QU-F6(3)
No Use the ASME standard for requirements at the time of doing an application. NEI 00-02 does not address this supporting requirement.
No objection
Appendix B to DG 1200, Page B-70 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position LERF Analysis LE-A1 Partial AS-14, AS-21, AS-23, L2-7 Confirm that the specifics identified in LE-A1 are included in the PRA. NUREG/CR-6595 methodology is not adequate for Capability Category II and III. It is further noted that NEI 00-02 does not address criteria for the grouping into plant damage states (PDSs) (i.e.,
there are no criteria provided as to what information has to be transferred from the Level 1 to the Level 2 analysis). L2-7 states the transfer from Level 1 to Level 2 should be done to maximize the transfer of relevant information, but does not specifically identify the type of information that must be transferred. L2-7 does refer to grouping sequences with similar characteristics and cautions care in transferring dependencies on accident conditions, equipment status and operator errors. In practice, this step included review of the process for developing and binning the PDSs and ensuring consistency between the PDSs and the plant state.
No objection
Appendix B to DG 1200, Page B-71 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position LE-A2 Partial L2-7, L2-8, AS-21 Confirm that the specifics identified in LE-A2 are included in the PRA.
NUREG/CR-6595 methodology is not adequate for Capability Category II and III. It is noted that NEI 00-02 does not address criteria for the grouping into PDSs (i.e., there are no criteria provided as to what information has to be transferred from the Level 1 to the Level 2 analysis).
L2-7 states the transfer from Level 1 to Level 2 should be done to maximize the transfer of relevant information, but does not identify the type of information that must be transferred.
No objection
Appendix B to DG 1200, Page B-72 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position LE-A3 Partial L2-7, L2-8 Confirm that the specifics identified in LE-A3 are included in the PRA.
NUREG/CR-6595 methodology is not adequate for Capability Category II and III. It is further noted that NEI 00-02 does not address criteria for the grouping into PDSs (i.e., there are no criteria provided as to what information has to be transferred from the Level 1 to the Level 2 analysis). L2-7 states the transfer from Level 1 to Level 2 should be done to maximize the transfer of relevant information, but does not identify the type of information that must be transferred.
No objection
Appendix B to DG 1200, Page B-73 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position LE-A4 Partial L2-7,L2-8, L29, L2-24, L2-25 Confirm that the specifics identified in LE-A4 are included in the PRA.
NUREG/CR-6595 methodology is not adequate for Capability Category II and III. It is further noted that NEI 00-02 does not address criteria for the grouping into PDSs (i.e., there are no criteria provided as to what information has to be transferred from the Level 1 to the Level 2 analysis). L2-7 states the transfer from Level 1 to Level 2 should be done to maximize the transfer of relevant information, but does not identify the type of information that must be transferred.
No objection
Appendix B to DG 1200, Page B-74 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position LE-A5 Partial L2-7, L2-8, L2-9, L2-24, L2-25 Confirm that the specifics identified in LE-A5 are included in the PRA.
NUREG/CR-6595 methodology is not adequate for Capability Category II and III. It is further noted that NEI 00-02 does not address criteria for the grouping into PDSs (i.e., there are no criteria provided as to what information has to be transferred from the Level 1 to the Level 2 analysis). L2-7 states the transfer from Level 1 to Level 2 should be done to maximize the transfer of relevant information, but does not identify the type of information that must be transferred.
L2-24 and L2-25 clearly indicate that the dependencies of systems, crew actions, and phenomena in the entire PRA need to be integrated into the model.
No objection LE-B1 Yes L2-8, L2-10, L2-15, L2-16, L2-17, L2-19 None No objection LE-B2 Yes L2-13, L2-14 None No objection
Appendix B to DG 1200, Page B-75 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position LE-B3(3)
No NEI 00-02 does not address this supporting requirement. Use the ASME PRA Standard for requirements.
No objection LE-C1 Yes L2-24, L2-5, L2-8, L2-13, L2-14, L2-15, L2-16, L2-17, L2-19, L2-20 Confirm that the specifics identified in LE-C1 with regard to the basis for assigning sequences to the LERF and non-LERF category meet the intent of LE-C1.
No objection LE-C2a Yes L2-9, L2-12, L2-25 Confirm that the actions credited are supported by AOPs, EOPs, SAMGs, TSC guidance or other procedural or guidance information as noted in LE-C2a.
No objection LE-C2b(1) Partial L2-9, L2-12, L2-25 Confirm that the specifics identified in LE-C2b are included in the PRA. Repair of equipment would be subsumed under recovery actions in L2-9 and L2-5. If credit was taken for repair, actual data and sufficient time must be available and justified.
No objection
Appendix B to DG 1200, Page B-76 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position LE-C3 Partial L2-8, L2-24, L2-25 Confirm that the justification for inclusion of any of the features listed in LE-C3 meets the revised requirements of LE-C3 in Addendum B of the ASME standard.
No objection LE-C4 Partial L2-4, L2-5, L2-6 The self-assessment needs to confirm the revised requirements of LE-C4 in Addendum B of the ASME standard.
No objection LE-C5 Yes AS-20, AS-21, L2-7, L2-11, L2-25 None No objection LE-C6 Yes L2-12, L2-24, L2-25 None No objection LE-C7 Partial L2-7, L2-11, L2-12, L2-24 Confirm that the requirements in LE-C7 are included in the PRA.
No objection LE-C8a Partial L2-11, L2-12 Confirm that the treatment of environmental impacts meets the revised requirements in LE-C8a in Addendum B of the ASME standard.
No objection LE-C8b(1)
Partial L2-11, L2-12 Confirm requirements of LE-C8b are implemented in the PRA.
No objection
Appendix B to DG 1200, Page B-77 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position LE-C9a Partial AS-20, L2-11, L2-12, L2-16, L2-24, L2-25 Confirm that the treatment of environmental impacts meets the revised requirements of LE-C9a in Addendum B of the ASME standard. NEI 00-02 does not differentiate between containment harsh environments and containment failure effects on systems and operators.
This was typically addressed during peer reviews.
No objection LE-C9b(1)
Partial AS-20, L2-11, L2-12, L2-16, L2-24, L2-25 Confirm the treatment of containment failure meets the revised requirements of LE-C9b. NEI 00-02 includes the effects of containment harsh environments and containment failure effects on systems and operators. This was typically verified during peer reviews.
No objection LE-C10 Partial L2-7, L2-8, L2-13, L2-24, L2-25 The revised requirements of LE-C10 in Addendum B of the ASME standard need to be considered in the self-assessment.
Containment bypass is explicitly identified in the failure modes addressed by the LERF analysis.
No objection
Appendix B to DG 1200, Page B-78 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position LE-D1a Partial L2-14, L2-15, L2-16, L2-17, L2-18, L2-19, L2-20, ST-5, ST-6 Confirm that the containment performance analysis meets the revised requirements of LE-D1a in Addendum B of the ASME standard.
No objection LE-D1b(1) Partial L2-14, L2-15, L2-16, L2-17, L2-18, L2-19, L2-20, ST-5, ST-6 Confirm requirements of LE-D1b are implemented.
No objection LE-D2 Partial L2-14, L2-19 Confirm the requirements of LE-D2 are implemented.
NEI 00-02 does not explicitly enumerate this supporting requirement.
However, the containment failure analysis includes by its nature for Capability Category II the location of the failure mode.
Therefore, both the analysis and the peer review have typically addressed this SR.
No objection LE-D3 Partial IE-14, ST-9 Confirm the requirements of LE-D3 are implemented in accordance with Addendum B. In practice, peer review teams evaluated the ISLOCA frequency calculation. F&Os under IE and AS would be written if this was not adequate.
No objection
Appendix B to DG 1200, Page B-79 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position LE-D4 No NEI 00-02 does not address this supporting requirement. Use the ASME standard for Supporting Requirement LE-D4.
No objection LE-D5 No NEI 00-02 does not address this supporting requirement. Use the ASME standard for Supporting Requirement LE-D5.
No objection LE-D6 Partial L2-16, L2-18, L2-19, L2-24, L2-25 Confirm that the containment isolation treatment meets the revised requirements of LE-D6 in Addendum B of the ASME standard. The guidance provided in NEI 00-02 does not explicitly enumerate the requirements in LE-D6. However, the PRAs were constructed to address the requirements of NUREG1335, which explicitly required containment isolation evaluation. Therefore, the PRAs and the Peer Reviews have typically addressed this SR.
No objection LE-E1 Yes L2-11, L2-12 None No objection LE-E2 Partial DA-4, HR-15, L2-12, L2-13, L2-17, L2-18, L2-19, L2-20 Confirm that the requirements of LE-E2 of Addendum B are met.
No objection
Appendix B to DG 1200, Page B-80 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position LE-E3(3)
No NEI 00-02 does not address this supporting requirement. Use the ASME PRA Standard for Supporting Requirement LE-E3.
No objection LE-E4(7)
Partial QU sub-elements applicable to LERF The self-assessment needs to confirm that the parameter estimation meets the revised requirements of LE-E4 in Addendum B of the ASME standard.
No objection LE-F1a Yes QU-8, QU-9, QU-10, QU-11, QU-31, L2-26 None No objection LE-F1b(1)
Yes L2-26 None No objection LE-F2 No QU-27, L2-26 NEI 00-02 does not address this supporting requirement. Use the ASME standard for Supporting Requirement LE-F2.
No objection LE -F3(3)
No NEI 00-02 does not address this supporting requirement. Use the ASME standard for Supporting Requirement LE-F3 No objection LE-G1(2)
Yes L2-26, L2-27, L2-28 None No objection
Appendix B to DG 1200, Page B-81 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position LE-G2(2)
Partial L2-26, L2-27, L2-28 In general, specified documentation items not explicitly addressed in NEI 00-02 checklists were addressed by the peer review teams. Action is to confirm availability of documentation. If not available, documentation may need to be generated to support particular applications or respond to NRC RAIs regarding applications.
No objection LE-G3(2)
Partial L2-26, L2-27, L2-28 In general, specified documentation items not explicitly addressed in NEI 00-02 checklists were addressed by the peer review teams. Action is to confirm availability of documentation. If not available, documentation may need to be generated to support particular applications or respond to NRC RAIs regarding applications.
No objection LE-G4(2)
Partial QU-27, QU-28, QU-29, QU-34 Confirm that the key assumptions and key sources of uncertainty consistent with the definitions of the ASME PRA Standard are documented.
No objection with Clarification: See staff position on definition of key assumption and key source of uncertainty in Appendix A.
Appendix B to DG 1200, Page B-82 Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions NEI Assessment ASME Std SR Addressed by NEI 00-02?
Applicable NEI 00-02 Elements Industry Self-Assessment Actions Regulatory Position LE-G5(2)
Partial L2-26, L2-27, L2-28 In general, specified documentation items not explicitly addressed in NEI 00-02 checklists were addressed by the peer review teams. Action is to confirm availability of documentation. If not available, documentation may need to be generated to support particular applications or respond to NRC RAIs regarding applications.
No objection LE-G6(3)
No NEI 00-02 does not address this supporting requirement. Use ASME PRA Standard Addendum B SR LE-G6 for requirements.
No objection Notes from NEI 00-02 Appendix D2:
1 Subdivided from a previous SR in Addendum A of the ASME PRA Standard. It is noted that Addendum B of the ASME PRA Standard has subdivided a number of SRs for the purpose of clarifying and separating the assignment of Capability Category of the SR in a clearly delineated fashion.
2 Revised to reflect new format for documentation section and SRs.
3 New SR added.
4 SR added to address multi-unit sites.
5 Formerly IF-A2.
6 Formerly IF-E2.
7 Formerly LE-E3.
Appendix C to DG 1200, Page C-1 APPENDIX C NRC POSITION ON THE NEI PROCESS FOR PERFORMING FOLLOW-ON PRA PEER REVIEWS FOR INTERNAL EVENTS (NEI 05-04)
The Nuclear Energy Institute (NEI) Peer Review Process for performing follow-on probabilistic risk assessments (PRAs) peer reviews is documented in NEI 05-04, Revision 1.
This appendix provides the staffs position on the NEI 05-04. The staffs positions are categorized as following:
No objection. The staff has no objection to the guideline.
No objection with clarification. The staff has no objection to the guideline. However, certain guidelines, as written, are either unclear or ambiguous, and therefore the staff has provided its understanding of these guidelines.
No objection subject to the following qualification. The staff has a technical concern with the guidelines and has provided a qualification to resolve the concern.
Table C-1 provides the NRC position on the NEI Follow-on Peer Review Process documented in NEI 05-04, Revision 1. A discussion of the staffs concern (issue) and the staff proposed resolution is provided. In the proposed staff resolution, the staff clarification or qualification is indicated in either bolded text (i.e., bold) or strikeout text (i.e., strikeout); that is, the necessary additions or deletions to the guidance (as written in NEI 05-04) for the staff to have no objection are provided.
Appendix C to DG 1200, Page C-2 Table C-1. NRC Regulatory Position on NEI 05-04 Section Issue Position Resolution Global The peer review process in NEI 05-04 is based on Addendum B to the ASME PRA standard (RA-Sb-2005).
The staff position on NEI 05-04 in Appendix B of Revision 1 of Regulatory Guide 1.200 is based on the staff position of RA-Sb-2005 as documented in Appendix A of Revision 1 of Regulatory Guide 1.200.
However, since that time, ASME has issued Addendum C (RA-Sc-2007) and ASME and ANS have issued a revision (ASME/ANS RA-S-2008) to RA-S-2002 which incorporates the changes in RA-Sc-2007. ASME/ANS RA-S-2008 (and ASME RA-Sc-2007) contain requirements that were revised or new (as compared to RA-Sb-2005). NEI 05-04 needs to be based on this revised standard (RA-S-2008) taking into account the staff position on RA-S-2008 as endorsed in Appendix A in this revision of this regulatory guide. That is, for example, RA-S-2008 may have requirements either revised from RA-Sb-2005 or not included in RA-Sb-2005, and now need to be assessed in the context of the staff endorsement.
Qualification Replace throughout the guide, ASME standard with the following, as appropriate:
with Parts 1 and 2 of ASME/ANS RA-S-2008 PRA standard Section 1.0. Introduction 1.1 thru 1.3 No objection
Appendix C to DG 1200, Page C-3 Table C-1. NRC Regulatory Position on NEI 05-04 Section Issue Position Resolution Section 2.0. General Overview Of Peer Review Process 1st paragraph A follow-on peer review of an at-power, internal events PRA (including internal flooding) that uses as criteria the supporting requirements of Part 2 of the ASME/ANS PRA Standard needs to address the staffs position provided in Appendix A to this regulatory guide to be acceptable to the staff for a regulatory application.
Clarification
. Follow-on peer review that cover the scope of the ASME/ANS PRA Standard will use the supporting requirements (SRs) in Section 4 Part 2 of the ASME/ANS PRA standard, supplemented, as appropriate, by the results of the original peer review. In addition, the NRCs position on Part 2 as provided in Appendix A to Regulatory Guide 1.200, should also be considered.
4th paragraph Per Section 1-6.3 of the ASME/ANS PRA Standard, the staff position is that, in addition to the results of the PRA, the follow-on peer review must review the PRA models and assumptions related to the PRA upgrade to determine their reasonableness given the design and operation of the plant.
Clarification In general, it is essential. of the PRA.
In addition, the follow-on peer review should review the PRA models and assumptions related to the PRA upgrade to determine their reasonableness given the design and operation of the plant. For example,....
Appendix C to DG 1200, Page C-4 Section 3.0. Grading Process 3.0 1st paragraph NEI 05-04 indicates that one of the outcomes of the follow-on peer review process is the assignment of grades for each SR that are used to indicate the relative capability level of each PRA technical element. However, for any application, a technical element not all the SRs have to be performed to the same capability. What capability is needed for a given SR is application dependent. Further, the next paragraph contradicts. It states that the. PRA Technical Elements are assigned an overall Capability Category.
Clarification One of the outcomes of Capability Categories, which are used to indicate the relative capability level of each technical element based on the SRs as defined in the ASME PRA Standard. For follow-on peer reviews against the ASME/ANS PRA Standard..
3.0 2nd paragraph NEI states that it is essential to focus the peer review on the specific conclusions of the PRA to ensure that the review directly addresses intended plant applications.
The staff position is that the follow-on peer review must also review the PRA models and assumptions related to the PRA upgrade in addition to the results of the PRA in order to ensure the PRA can be used for specific applications.
Clarification In general, it is essential. of the PRA.
In addition, the follow-on peer review should also review the PRA models and assumptions related to the PRA upgrade in addition to the results of the PRA in order to ensure the PRA can be used for specific applications. It is important.
Appendix C to DG 1200, Page C-5 3.1 2nd paragraph A follow-on peer review of an at-power, internal events PRA (including internal flooding) that uses as criteria the supporting requirements of Part 2, and the requirements of Part 1, Section 1-5 of the ASME/ANS PRA Standard needs to address the staffs position provided in Appendix A to this regulatory guide to be acceptable to the staff for a regulatory application.
Clarification For a peer review.. meets for that SR.
In addition, a follow-on peer review should also address the NRCs position on Parts 1 and 2 of the ASME/ANS standard provided in Appendix A to Regulatory Guide 1.200.
3.1 5th paragraph NEI 05-04 indicates that although no grades are assigned to HLRs, a qualitative assessment of the HLRs will be made based on the associated SR grades.
The staffs position is consistent with the ASME/ANS PRA Standard, which indicates that a PRA reviewed against the standard must satisfy all HLRs. To meet an HLR, all SRs under that HLR must meet the requirements of one of the three Capability Categories.
Clarification When the peer review. based on the associated SR Capability Categories, given that all the SRs for the HLR were met.
Appendix C to DG 1200, Page C-6 3.2 Comparison Against Grading Process for NEI 00-02 The NEI 00-02 process uses a set of checklists as a framework within which to evaluate the scope, comprehensiveness, completeness, and fidelity of the PRA being reviewed.
The checklists by themselves are insufficient to provide the basis for a peer review since they do not provide the criteria that differentiate the various grades of PRA. The NEI subtier criteria provide a means to differentiate between grades of PRA.
However, since the NEI subtier criteria do not address all of the requirements in the ASME/ANS PRA Standard (Parts 1 and 2), the staffs position is that a peer review based on these criteria is incomplete. The PRA standard requirements that are not included in the NEI 00-02 subtier criteria (identified for a Grade 3 PRA in Table B-3) need to be addressed in the NEI 00-02 self-assessment process as endorsed by the staff in this appendix. (Staff comment on Section 1.1 on NEI 00-02)
Clarification Under the NEI 00-02 grading process.
These checklists are contained in Appendix B of NEI 00-02. However, the checklists by themselves are insufficient to provide the basis for a peer review.
The requirements in the ASME/ANS PRA standard (Parts 1 and 2) should serve as the basis for the peer review in using the checklists.
Appendix C to DG 1200, Page C-7 The NEI 00-02 peer review process grades each PRA element from 1 to 4, while the ASME/ANS PRA Standard uses Capability Categories I, II, and III. The staff interpretation of Grades 2, 3, and 4 is that they correspond broadly to Capability Categories I, II, and III, respectively. This statement is not meant to imply that the supporting requirements, for example, for Category I are equally addressed by Grade 2 of NEI 00-02. The review of the supporting requirement for Category II against Grade 3 of NEI 00-02 indicated discrepancies and consequently the need for a self-assessment. The existence of these discrepancies would indicate that it would not be appropriate to assume that there are not discrepancies between Category I and Grade 2. A comparison between the other grades and categories has not been performed. The implications of this are addressed in item 7 of Table B-2. (Staff comment on Section 3.3 on NEI 00-02)
Clarification In general, the following approximate correspondence exists between the two grading systems:
NEI 00-02 ASME/ANS PRA Standard Grade 1 No equivalent grade Grade 2 Capability Category I Grade 3 Capability Category II Grade 4 Capability Category III The above comparison is not meant to imply that the supporting requirements, for example, for Category I are equally addressed by Grade 2 of NEI 00-02. It would not be appropriate to assume that there are not discrepancies between Category I and Grade 2, Category II and Grade 3, and Category III and Grade 4.
Appendix C to DG 1200, Page C-8 Section 4.0. Follow-On Peer Review: ASME[/ANS] PRA Standard Scope 4.1 thru 4.5 No objection 4.6 12th and 13th paragraphs Section 1-6.1 in Part 1 of the ASME/ANS PRA Standard indicates that the peer review need not assess all aspects of the PRA against all of the Section 4 requirements. The NEI 05-04 process interpretation of this statement allows for skipping review of selected SRs if the reviewers determine they can achieve consensus on the adequacy of the PRA with respect to the HLR associated with the SRs that are not reviewed. The staffs position is that the statement quoted refers to the scope of the models being reviewed and not the scope of the SRs to be reviewed. The staffs position is that all SRs pertinent to the PRA upgrade must be reviewed against a sufficient number and variety of models in the PRA (e.g.,
selected fault and event trees) to determine the SR capability categories.
Without a review, the capability category for skipped SRs cannot be determined.
Clarification As stated in Section 1-6.1 in Part 1 of the ASME/ANS PRA Standard, The peer review.. for each PRA element..
Must be addressed.
In performing the review of a given technical element, the Lead Reviewer may elect to skip the review of selected SRs must document their basis fro skipping the given SR.
4.7 No objection APPENDICES A, B No objection
Appendix C to DG 1200, Page C-9 C
Slide 3 states the Appendix A of this regulatory guide must be used to clarify the ASME PRA standard but fails to mention that this appendix must be used for clarifications to NEI 05-04.
Clarification
-- NRC clarifications and qualitifications as provided in Appendixces A, B and C of RG 1.200, Rev.1 D
No objection
Appendix D to DG 1200, Page D-1 APPENDIX D NRC POSITION ON THE NEI INTERNAL FIRE PEER REVIEW PROCESS (NEI-07-12)
The Nuclear Energy Institute (NEI) Peer Review Process for a fire probabilistic risk assessment (PRA) is documented in NEI 07-12, Revision 0. It provides guidance for the peer review of probabilistic risk assessments (PRAs) and the grading of the PRA subelements into one of four capability categories.
This appendix provides the staffs position on the NEI Fire PRA Peer Review Process (i.e., NEI 07-12). The staffs positions are categorized as following:
No objection. The staff has no objection to the guideline.
No objection with clarification. The staff has no objection to the guideline. However, certain guidelines, as written, are either unclear or ambiguous, and therefore the staff has provided its understanding of these guidelines.
No objection subject to the following qualification. The staff has a technical concern with the guidelines and has provided a qualification to resolve the concern.
Table D-1 provides the NRC position on the NEI Fire PRA Peer Review Process documented in NEI 07-12, Revision 0. A discussion of the staffs concern (issue) and the staff proposed resolution is provided. In the proposed staff resolution, the staff clarification or qualification is indicated in either bolded text (i.e., bold) or strikeout text (i.e., strikeout); that is, the necessary additions or deletions to the guidance (as written in NEI 07-12) for the staff to have no objection are provided.
Appendix D to DG 1200, Page D-2 Table D-1. NRC Regulatory Position on NEI 07-12 Index No Issue Position Resolution Global The peer review should be performed using the ASME/ANS PRA standard, RA-S-2008.
Although the technical requirements are the same as in the ANS Fire PRA standard, the organization is different (e.g., section number, table numbers) which will cause confusion when attempting to review the results of the peer review against the ASME/ANS standard. In addition, the maintenance of the requirements will be in the ASME/ANS standard.
Further, the ASME standard (ASME-RA-Sc-2007), while it does exists, has been revised, and is the ASME/ANS RA-S-2008 PRA standard.
Qualification Replace throughout the guide with one of the following, as appropriate:
ANS FPRA standard with ASME/ANS RA-S-2008 PRA standard ANS FPRA standard with ASME/ANS RA-S-2008 (Part 3) PRA standard ASME standard with ASME/ANS RA-S-2008 PRA standard Section 1.0. Introduction Section 1.1, 2nd paragraph Editorial Clarification The FPRA Peer Review employs. In the ANS FPRA Standard ASME/ANS PRA standard (ASME/ANS RA-S-2008) [1].
The ASME/ANS PRA standard, in particular Parts 1 and 3, FPRA Standard provides a basis for both an objective review of the FPRA technical elements, and an assessment, based on the Section 1.1, 3rd paragraph, 3rd sentence The findings from a self assessment should also be assessed Clarification Another key aspect of the FPRA. of the fact and observation sheets (F&Os) and the results of any self-assessment that has been performed for relevance to the FPRA.
Appendix D to DG 1200, Page D-3 Table D-1. NRC Regulatory Position on NEI 07-12 Index No Issue Position Resolution 1.1 5th paragraph Reference to now outdated terminology level A/B F&Os Clarification The FPRA Peer Review process discussed
.... at least the level A/B F&Os classified as Findings from that review....
1.2 No objection 1.3 2nd paragraph The WOG also adapted the BWROG peer review process with some slight modification prior to the parallel effort that included NEI. The B&WOG may have done so likewise. The citing solely of the CEOG implies the other two OGs have not performed any work on fire peer reviews.
Clarification The PWR Owners Groups (Combustion Engineering [CEOG], Westinghouse
[WOG] and Babcock & Wilcox
[B&WOG]) adopted...
1.3 4th paragraph Not all the SRs for developing system models and supporting analysis refer to the ASME PRA Standard HLRs and SRs Clarification In November 2007,.Second, many of the SRs for developing...
1.4 No objection 1.5 1st paragraph This states that the FPRA Peer Review uses Capability Categories to assess the relative technical merits and capabilities of each technical element reviewed, including ANS Fire PRA Standard HLRs and each of the SRs. This is contradicted in numerous places in NEI 07-12 (including the paragraph that immediately follows),
where it is made clear that only the SRs are assigned a capability category.
Clarification The FPRA Peer Review uses Capability Categories to assess the relative technical merits and capabilities of each technical element reviewed, including ANS Fire PRA Standard HLRs and each in terms of the Fire PRA SRs in Part 3 of the ASME/ANS PRA standard.
1.5 2nd paragraph This states that a summary of the technical adequacy of each HLR to support a risk-informed application Clarification
. Then, based on the SR Capability Categories, a summary of the technical adequacy to support a risk-informed
Appendix D to DG 1200, Page D-4 Table D-1. NRC Regulatory Position on NEI 07-12 Index No Issue Position Resolution is provided. This implies that a risk-informed application must be contemplated in order to provide the summary for the HLR, which appears contrary to the concept that the base fire PRA model is evaluated using the standard and the technical adequacy, for any specific risk application, is evaluated on a case-by-case basis.
application is provided for each 1.5 3rd paragraph This misstates the major benefit of the peer review process. Although recommendations for improvement and recognizing the strengths of a given fire PRA are important, the purpose of the peer review is to assess (and document) the technical adequacy of the base fire PRA. The F&Os, apart from their resolution, are important when NRC assesses the fire PRA in the context of a risk-informed submittal. (Note that the self-assessment of a licensees fire PRA to the standard, which NEI 07-12 specifies to be completed in advance of the peer review, is another appropriate mechanism to identify areas for improvement in order to meet the fire PRA standard at the desired capability category.)
All peer reviewers would interact with the host utility.
Clarification The major benefits of this review process, therefore, are not the assignments of SR Capability Categories, but rather are the assignments of capability categories to SRs which assess the technical adequacy of the base fire PRA, as well as the recommendations for improvements and the acknowledgement of the strengths of the FPRA which originate from assessment of the fire PRA in the peer review.
Additional beneficial outcomes among the Host Utility and utility industry reviewer personnel,...
Appendix D to DG 1200, Page D-5 Table D-1. NRC Regulatory Position on NEI 07-12 Index No Issue Position Resolution 1.6 No objection Section 2.0. Peer Review Process 2.1 1st paragraph The peer review may be done independent of an application to assess and characterize the technical adequacy of the PRA Clarification The FPRA Peer Review is a requirement of the ASME/ANS RA-S-2008 ANS Fire PRA standard to identify assess the technical adequacy of the FPRA to support risk-informed applications, and is complementary 2.1 3rd paragraph To meet the HLR, all SRs must be addressed, even if it is to say they are NA.
An optional step that is not performed is the only exception.
Clarification
. For example, if the FPRA does not include. the quantitative screening HLRs/SRs do not need to be reviewed are not applicable.
2.1 6th paragraph This guide discusses peer review team member expertise as needed for specific technical reviews, and says that the host utility can request particular expertise for the peer review team.
However, this guide is directed towards a full fire PRA review. The PRA standard requires that the peer review team members have expertise and experience for a full fire PRA. This expertise and experience required in Part 1 and 3 of the standard should not be reduced.
Clarification Selection of a Peer Review Team can alsoFPRA information. As discussed in Section 2.2 below, the Peer Review Team skills can vary for each review, depending on whether the FPRA includes particular fire modeling analysis or detailed circuit analysis should possess sufficient expertise to cover all of the FPRA elements. The utility can request particular expertise beyond the general expertise identified in Section 3-2 of the Standard Sections 1-6.2 and 3-2.2 of the ASME/ANS RA-S-2008 standard (and considering the staffs position in Regulatory Guide 1.200) for the Peer Review Team, should when specific more specialized skills be are needed. The team leader should verify..
Section 2.1, step 7b, 2nd paragraph It states: In general, it is essential to focus the review on the specific conclusions of the FPRA to ensure that the review directly addresses previously defined plant applications of the FPRA.
For example, a plant transitioning to NFPA-805, where local manual actions are analyzed in Qualification for each FPRA SR. In general, it is essential. local manual actions.
Appendix D to DG 1200, Page D-6 Table D-1. NRC Regulatory Position on NEI 07-12 Index No Issue Position Resolution detail, the focus of the review would include the fire scenarios involving local manual actions.
This is contrary to the fire PRA requirements in the ASME/ANS PRA standard, which focuses on assessing the base PRA and covers all elements that are applicable (the only exceptions being, for example, QLS if a plant did not perform qualitative screening).
2.1 Step 8, 2nd paragraph Reference to now outdated terminology level A/B F&Os Clarification Consensus sessions. Similarly, the assignment of A/B F&Os classified as Findings is also based on.
2.2 4th paragraph Reference to experience regarding optimum team size gives the impression that FPRA Peer Reviews have been conducted previously Clarification Experience from internal events PRA peer reviews has indicated that an optimum team size is five or six members. This is believed to be generally optimal for FPRA peer reviews as well. The actual...
2.3, 2.4 No objection Section 3.0. FPRA Peer Review Process Elements and Guidance 3.1 No objection 3.2 4th paragraph This states that the assessment of the fire PRA should be derived from what is in the Fire PRA Standard (and not based on NRC clarification or qualifications). This presents a potential problem, since one benefit of completing a peer review is to allow risk-informed licensing applications with reduced Qualification. However, a reviewers assessment should be derived from what is in the Fire PRA Standard (and not based on NRC clarification or qualifications). If so dersired by the host utility, the The reviewer(s) may should also provide
Appendix D to DG 1200, Page D-7 Table D-1. NRC Regulatory Position on NEI 07-12 Index No Issue Position Resolution NRC review of the technical adequacy of the base fire PRA. If the NRC clarifications and qualifications are not considered when assigning a capability category to a given SR, then additional NRC staff review of the base fire PRA model may be required as part of the review of risk-informed applications.
3.2 9th paragraph The emulation of best practices should not be limited to utilities with findings.
Clarification Originally, the S classification... to the extent that utilities (with findings) would want to emulate...
3.2 13th paragraph Some related requirements from the Part 2 of the ASME/ANS PRA standard, are incorporated by reference.
Section 1-6.3 of the ASME internal events PRA standard states:
The review team shall use the requirements of the Peer Review Section of each respective Part of this Standard for the PRA Elements being reviewed to determine if the methodology and the implementation of the methodology for each PRA Element meet the requirement of this Standard. Further it states: The HLRs and the composite of the SRs of the Technical Requirements Section of each respective Part of this Standard shall be used by the peer review team to assess the completeness of Qualification During the review of a can be excluded with justification. While Section 1-6.1 of the ASME/ANS PRA standard states that not all aspects of the PRA need be assessed, this statement is intended to limit how much of the model needs be considered when determining whether an SR or HLR is met. The SRs form the basis for determining whether the related HLR is met, and every SRs in the HLR should be assessed by the review team.
Appendix D to DG 1200, Page D-8 Table D-1. NRC Regulatory Position on NEI 07-12 Index No Issue Position Resolution a PRA Element.
Contrary to this, NEI 07-12 would allow the peer review team to elect to skip selected SRs.
3.2 14th paragraph The focus of this review should be on the sources of model uncertainty and related assumptions. See the FRN on source of uncertainty.
Clarification The reviewers should specifically address key assumptions and key source of uncertainty sources of model uncertainty and related assumptions in the elements being reviewed. Such assumptions and uncertainties, and their potential impact on the basline PRA results and PRA applications, and the manner.. addresses them, should be reviewed. The reviewers these key assumptions and 3.3 2nd paragraph Could give a false impression that only utility peer reviewers interact Clarification The major benefits... among the Host Utility and utility industry reviewer personnel,...
3.3.1 There are SRs in Part 2 of the ASME/ANS standard that, although they span all three categories, invoke the supporting requirements in Part 2.
Examples are HRA-B1, HRA-B2, HRA-B2 and FQ-A4. The intent in Part 2 is that these supporting requirements be assessed individually. Therefore, the peer reviewer should also assessed these referenced requirements.
Clarification NOTE:
It should be noted that several of the SRs of the fire PRA include statements that invoke high level requirements or specific SRs in Part 2 of the ASME/ANS PRA standard. An example is HRA-B1, which includes the statement and in accordance with HLR-HR-F and its SRs in Part 2. The intent in Part 3 is that each of these supporting requirements be assessed as written in Part 2, which may include subdivision into capability categories. In such a case, each SR that is included by reference needs to be assessed on its own merit. In this case, the SR in Part 3 is like a high level requirement in the sense that there is no need to give it a capability category, although a summary assessment could be helpful.
3.4 1st bullet This appears to be a reference to item MU-14 in Appendix C, and not Appendix B. It is not clear what relevance the independent review of the Clarification The independent review identified the specified FPRA element. This independent review may have been performed as part of the IPEEE process.
The Peer Review Team
Appendix D to DG 1200, Page D-9 Table D-1. NRC Regulatory Position on NEI 07-12 Index No Issue Position Resolution IPEEE has for a newly developed FPRA.
3.4 Last bullet This states, in part: A certain level of subjectivity is expected when determining if an SR is in compliance with the Fire PRA Standard.
Since the SR is part of the fire PRA standard, the sentence does not make sense. It should be re-worded to make it clear that it is the fire PRA model that is being assessed.
Clarification A certain level of subjectivity is expected when determining if an SR has been met is in compliance with the Fire PRA Standard.
3.5 No objection 3.6 3rd paragraph The last sentence, reads:
In general, the assigned Capability Category will be based on the assessed level for the significant contributors, while ensuring the non-significant contributors still meet a lower Capability Category.
This is unclear and potentially problematic.
Until the fire PRA model has been assessed as meeting at least Capability Category I for all applicable SRs, the review team has no way to know whether the significant contributors have been properly identified.
Clarification Many of the SRs In general, the assigned Capability Category will be based on the assessed level for the significant contributors, while ensuring the non-significant contributors still meet a lower Capability Category I.
3.6 2nd paragraph Editorial issue Clarification When reviewing then Capability Category I would likely be assessed assigned even if all other significant contributors were analyzed with detailed fire modeling
Appendix D to DG 1200, Page D-10 Table D-1. NRC Regulatory Position on NEI 07-12 Index No Issue Position Resolution Section 4.0. PRA Process Results and Documentation 4.1 2nd bullet The use of findings in second bullet could be misleading. Assuming the intent is that the report will present all F&Os, not just findings, the use of the term findings here is misleading.
Clarification the findings conclusions of the Peer Review Team...
4.2 thru 4.4 No objection Appendix A: Preparation Material A.1 thru A.6 No objection A.7 1st paragraph 2nd bullet The first bullet seems to have extra words Clarification The fire cutsets or fire sequences that appear to not address dependencies that have not been properly accounted for in the model and quantification process; A.7 2nd paragraph Numerical CDF results from sensitivity cases may be meaningful depending upon the specific item for which the sensitivity is exercised (e.g., a mathematically impossible result may indicate an incorrect sensitivity analysis)
Clarification Note that the actual CDF... and are not necessarily meaningful may be of limited relevance for the peer review.
A.7 3rd paragraph Printout of top 200 cutsets/sequences may be too many or too few, depending on the relative contributions among them, including their total Clarification The sensitivity studies... a printout of the top 200 a sufficient number of cutsets or sequences (at least 100) to illustrate that the conclusions relative to the stated aims are robust, plus importance...
A.8 thru A.10 No objection Table A-1 No objection Exhibit A-1 No objection Appendix B: Sample Summary Tables Tables B-1A For Table B-1A, HLR-PP-Clarification HLR-PP-A:
Appendix D to DG 1200, Page D-11 Table D-1. NRC Regulatory Position on NEI 07-12 Index No Issue Position Resolution and B-1B B, the generic term, physical analysis units, has replaced fire areas and/or compartments.
HLR summaries in these tables should be consistent with Part 3 of the ASME/ANS standard.
Perform a plant partitioning analysis to identify and define the fire areas and/or fire compartments physical analysis units.
HLR-ES-A, B, C:
Identify and locate.or mal-operation spurious operation.
Table B-1 The SRs that are invoked from Part 2 need to be recorded (See comment on 3.3.1)
Clarification For those SRs that invoke HLRs or specific SRs from Part 2 of the ASME/ANS PRA standard, add these HLRs and SRs to the table and address their assessment to the same extent as the SRs from Part 3.
For example, PRM-B8 of Part 3 invokes all the SRs under HLR-SY-A and HLR-SY-B of Part 2. Therefore, SRs SY-A1 thru SY-A22 and SY-B1 thru SY-B16 should be added to the table and their assessment recorded.
Table B-1 ASM, the acronym for an older name of a technical element, is no longer appropriate.
Clarification Replace ASM with PRM. Review table SRs and PRA capability categories for consistency with Part 3 of the ASME/ANS Standard.
Appendix C: Maintenance and update Process review check list Table MU It is noted in Appendix C that the Checklist Criteria were extracted from Table MU in Appendix B of NEI 00-02. In Appendix B of this regulatory guide, the staff position is that the checklists by themselves are insufficient to provide the basis for a peer review.
Clarification See staff comment on checklist tables in Appendix B of NEI 00-02 in Appendix B of this regulatory guide.
Appendix D: Lessons Learned Lessons Learned Input Form No objection Appendix E: Sample Fact and Observation Form Fact/Observation Regarding PRA Technical Elements Form No objection
Appendix D to DG 1200, Page D-12