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U.S. NUCLEAR REGULATORY COMMISSION January 2007

Revision 1

REGULATORY GUIDE

OFFICE OF NUCLEAR REGULATORY RESEARCH

REGULATORY GUIDE1.200

(Draft was issued as DG-1161, dated September 2006)

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 NRC’s deterministic

approach.” Since that time, many uses have been implemented or undertaken, including modification of

the NRC’s 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 NRC’s PRA Policy Statement and subsequent,

more detailed, guidance in Regulatory Guide 1.174 (RG) (Ref. 2). It is also intended to reflect and

endorse guidance provided by standards-setting and nuclear industry organizations.
Rev. 1 of RG 1.200, Page 2

When used in support of an application, this regulatory guide will obviate the need for

an in-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, as in RG 1.174, 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. 3). 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. 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.
In this regulatory guide, a part of a PRA can be understood to be equivalent to that piece of the analysis for which 1

an applicable PRA standard identifies a supporting level requirement.

Rev. 1 of RG 1.200, 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.

One specific regulatory guide and its associated standard review plan (SRP) is RG 1.174

and SRP Section 19 (Ref. 5), which 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 staff’s 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 parts of the PRA 1

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.

In addition, separate regulatory guides provide guidance for such specific applications

as inservice testing (Ref. 6), inservice inspection (Ref. 7), quality assurance (Ref. 8), and technical

specifications (Ref. 9). The NRC has also prepared SRP sections for each of the application-specific

regulatory guides, with the exception of quality assurance.

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 a full-power, internal events (excluding fire)

Level 1 PRA and a limited Level 2 PRA, and subsequently issued Addenda A and B to that

standard on December 5, 2003, and December 30, 2005, respectively (Ref. 10). ASME issued

Addendum B in response to the NRC staff’s position on Addendum A, lessons learned from

the pilots, and other public comments provided to ASME.

• In December 2003, ANS issued a standard for external events (Ref. 11).

• ASME and ANS are developing Level 1 PRA standards for internal fire, external events,

and low-power shutdown operating mode, as well as Level 2 and Level 3 PRA standards.
Rev. 1 of RG 1.200, Page 4

Reactor owners’ groups have been developing and applying a PRA peer review program

for several years. The Nuclear Energy Institute (NEI) issued NEI-00-02 (Ref. 12), which documents

one such process:

• On August 16, 2002, NEI submitted draft industry guidance for self-assessments (Ref. 12)

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. 12), 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-00-02,

to satisfy the peer review requirement(s) of the ASME PRA Standard (ASME-RA-Sa-2003)

as endorsed/modified by the NRC and updated by Addendum B of the ASME PRA Standard

(Ref. 12).

• 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 (Ref. 13).

• In November 2006, NEI updated the self-assessment guidance in Revision 1 of NEI 00-02 to

address the staff objections raised in Appendix B of DG-1161 (Ref. 12).

SECY-00-0162 (Ref. 14) 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.

Regulatory Guide 1.201, “Guidelines for Categorizing Structures, Systems, and Components

in Nuclear Power Plants According to their Safety Significance” (Ref. 15), discusses an approach,

along with References 9 and 12, to support the new rule established as Title 10, Section 50.69,

of the Code of Federal Regulations (10 CFR 50.69), “Risk-informed categorization and treatment

of structures, systems, and components for nuclear power reactors” (Ref. 16).

SECY-04-0118, “Plan for the Implementation of the Commission’s Phased Approach

to PRA Quality” (Ref. 17), presents the staff’s 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.

Purposes of this Regulatory Guide

The purposes of this regulatory guide are to provide guidance to licensees for use in determining

the technical adequacy of the base PRA used in a risk-informed regulatory activity, and to endorse

standards and industry guidance. Toward that end, this regulatory guide provides guidance in four areas:

(1) a definition of a technically acceptable PRA

(2) the NRC’s position on PRA consensus standards and industry PRA 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
Rev. 1 of RG 1.200, Page 5

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 – 8.

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. 

This regulatory guide does not address such alternative methods to the evaluation of risk contributions;

rather, this guide only addresses PRA methods.

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) RG 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, quality assurance,

and technical specifications (Refs. 4–7); 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 and risk-informed activities, applicationspecific guidance, consensus PRA standards, and industry programs (e.g., NEI-00-02).

Figure 1. Relationship of Regulatory Guide 1.200 to Other Risk-Informed Guidance
Some applications may involve the plant at the design certification or combined operating license stage, where the plant 2

is not built or operated. At these stages, the intent is for the PRA model to reflect the as-designed plant.

Rev. 1 of RG 1.200, Page 6

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 nuclear power plant. 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, asoperated plant to the extent needed to support the application 2

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 types of initiating events 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, in accordance with SECY-00-0162:

(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

• types of initiating events that can potentially challenge and disrupt the normal operation

of the plant
CDF and LERF are generally the metrics used in risk-informed decision-making for operating reactors licensed under 3

Part 50. Large release is defined as the frequency of those accidents leading to rapid, unmitigated release of airborne

fission products from the containment to the environment. 

Rev. 1 of RG 1.200, Page 7

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 light-water reactors). These 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 frequency 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.)

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.3

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) 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 full-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 events that have the ability to challenge the condition of the plant.

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). The risk

perspective is based on a consideration of the total risk, which includes events attributable to both

internal and external sources.

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.
Rev. 1 of RG 1.200, Page 8

Table 1. Technical Elements of a PRA

Scope of

Analysis

Technical Element

Level 1 • Initiating event analysis • Parameter estimation analysis

• Success criteria analysis • Human reliability analysis

• Accident sequence analysis • Quantification

• Systems analysis

Level 2 • Plant damage state analysis • Quantification

• Accident progression analysis • 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.

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

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.
Significant accident sequence: A significant sequence is one of the set of sequences, defined at the functional or 4

systemic level that, when ranked, compose 95% of the CDF or the LERF, or that individually contribute more than

~1% to the CDF or LERF.

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.

Rev. 1 of RG 1.200, Page 9

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

system’s 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, operation, and maintenance

of the plant. This CDF is based on the summation of the estimated CDF from each accident sequence for

each initiator class. If 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

contributors are eliminated. Therefore, the truncation limit can vary for each accident sequence. 4

Consequently, the truncation value is selected so that the accident sequence CDF is stable with respect to

further reduction in the truncation value.
Rev. 1 of RG 1.200, Page 10

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 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 fatalities 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 and the probability of a large

late release.
Rev. 1 of RG 1.200, Page 11

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., submerge, 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. These fire

scenarios are developed during the screening analysis, fire initiation analysis, and the fire damage

analysis. The plant response and quantification that is specific to internal fires is similar in nature to

that for the internal events. 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 as necessary.
Rev. 1 of RG 1.200, Page 12

Screening analysis identifies fire areas where fires could result in significant accident

sequences. Fire areas that cannot result in significant accident sequences can be “screened out” from

further consideration in the PRA analysis. Both qualitative and quantitative screening criteria can be

used. The former address whether an unsuppressed fire in the area poses a nuclear safety challenge; the

latter are compared against a bounding assessment of the fire-induced core damage frequency for the

area. Plant walkdowns are performed where possible to verify the accuracy of the information used in

the screening analysis. It is recognized that at the design and initial licensing stage, plant walkdowns are

not possible. Screening analysis assumptions and results [e.g., the area-specific conditional core damage

probabilities (assuming fire-induced loss of all equipment in the area)] are documented.

Fire initiation analysis determines the frequency and physical characteristics of the detailed

(within-area) fire scenarios analyzed for the unscreened fire areas. The analysis identifies a range of

scenarios that will be used to represent all possible scenarios in the area. The possibility of seismically

induced fires is considered. The scenario frequencies reflect plant-specific experience, to the extent

available and supplemented with industry fire information, and quantified in a manner that is consistent

with its use in the subsequent fire damage analysis (discussed below). Each scenario is physically

characterized in terms that will support the fire damage analysis (especially with respect to fire

modeling).

Fire damage analysis determines the conditional probability that sets of potentially significant

contributors (i.e., components including cables) will be damaged in a particular mode, given a specified

fire scenario. The analysis addresses components whose failure will cause an initiating event, affect the

plant’s ability to mitigate an initiating event, or affect potentially significant contributors (i.e.,

equipment), such as through suppression system actuation. Damage from heat, smoke, and exposure to

suppressants is considered. If fire models are used to predict fire-induced damage, compartment-specific

features (e.g., ventilation, geometry) and target-specific features (e.g., cable location relative to the fire)

are addressed. The fire suppression analysis accounts for the scenario-specific time to detect, respond to,

and suppress the fire. The models and data used to analyze fire growth, fire suppression, and fireinduced component damage are consistent with experience from actual nuclear power plant fires, as well

as experiments.

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 defined in the fire damage analysis. All potentially fire-induced initiating events

that can result in significant accident sequences, including such “special” events as loss of plant support

systems and interactions between multiple nuclear units during a fire event, are addressed. The analysis

addresses the availability of non-fire-affected equipment (including control) and any required manual

actions. The human reliability analysis of operator actions addresses fire effects on operators (e.g., heat,

smoke, loss of lighting, effect on instrumentation) and fire-specific operational issues (e.g., fire response

operating procedures, training on these procedures, potential complications in coordinating activities).

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.
Rev. 1 of RG 1.200, Page 13

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 /year and that the conditional core-damage -5

probability is less than 10 , given the occurrence of the design-basis hazard. An external event that -1

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 sitespecific 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.

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 and large late release probability for both individual sequences and the model

as a total, using such tools as importance measure calculations (e.g., Fussell-Vesely Importance, risk

achievement worth, risk reduction worth, and Birnbaum Importance).
Rev. 1 of RG 1.200, Page 14

An important aspect in understanding the PRA results is understanding the associated

uncertainties. Sources of uncertainty are identified and their impact on the results analyzed. The

potential conservatism associated with the successive screening approach used for the analysis of specific

scope items such as fire, flooding, or seismic initiating events is assessed. 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.

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 (RCP) 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 or LERF 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, as described above, 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.
Rev. 1 of RG 1.200, Page 15

Table 2. Summary of Technical Characteristics and Attributes of a PRA

Element Technical Characteristics and Attributes (see Note 1)

PRA Full-Power, Low-Power, and Shutdown

Level 1 PRA (internal events — transients and LOCAs)

Initiating

Event

Analysis

• sufficiently detailed identification and characterization of initiators

• grouping of individual events according to plant response and mitigating requirements

• proper screening of any individual or grouped initiating events 

Success

Criteria

Analysis

• based on best-estimate engineering analyses applicable to the actual plant design and

operation

• codes developed, validated, and verified in sufficient detail

< 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 (PDSs)]

• 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-built, as-operated plant including how it has performed during the plant

history

• 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 and including appropriate dependencies

(both pre- and post-accident)

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
Table 2. Summary of Technical Characteristics and Attributes of a PRA

Element Technical Characteristics and Attributes (see Note 1)

Rev. 1 of RG 1.200, Page 16

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 verified, validated 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 met. It is recognized that for reactors in the design stage, such items as performing plant

walkdowns, using plant-specific data, are not possible. The attribute as stated may need to be revised in order

to meet the intent.

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.
Rev. 1 of RG 1.200, Page 17

Table 3. Summary of Technical Characteristics and Attributes 

of an Internal Flood and Fire Analysis and External Hazards Analysis

Areas of Analysis 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

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

Internal Fire Analysis

Screening Analysis • fire areas are identified and addressed that can result in significant accident

sequences

• all credited mitigating components and their cables in each fire area are

identified

• screening criteria are defined and justified

• necessary walkdowns are performed to confirm the screening decisions

• screening process and results are documented

• unscreened events areas are subjected to appropriate level of evaluations

(including detailed fire PRA evaluations as described below)

Initiation Analysis • fire scenarios in each unscreened area are addressed that can result in a

significant accident sequence

• fire scenario frequencies reflect plant-specific features

• fire scenario physical characteristics are defined

• bases are provided for screening fire initiators

Damage Analysis • damage to significant contributors (i.e., components) is addressed, considering

all potential component failure modes

• all potentially significant contributors (i.e., damage mechanisms) are identified

and addressed, and damage criteria are specified

• analysis addresses scenario-specific factors affecting fire growth, suppression,

and component damage

• models and data are consistent with experience from actual fires, as well as

experiments

• includes evaluation of propagation of fire and fire effects (e.g., smoke) between

fire compartments
Table 3. Summary of Technical Characteristics and Attributes 

of an Internal Flood and Fire Analysis and External Hazards Analysis

Areas of Analysis Technical Characteristics and Attributes

Rev. 1 of RG 1.200, Page 18

Plant Response Analysis • fire-induced initiating events that can result in significant accident sequences

are addressed so that their bases are included in the model

• includes fire scenario impacts on core damage mitigation and containment

systems, including fire-induced failures

• analysis reflects plant-specific safe shutdown strategy

• potential circuit interactions that can interfere with safe shutdown are addressed

• human reliability analysis addresses effect of fire scenario-specific conditions

on operator performance

Quantification • estimation of fire CDF for chosen fire scenarios

• identification of sources of uncertainty and their impact on the results

• understanding of the impact of the assumptions on the CDF

• all fire-significant sequences are traceable and reproducible

External Hazards Analysis

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 are performed

• non-screened events are subjected to an appropriate level of evaluations

Hazard Analysis • the hazard analysis is site- and plant-specific

• 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

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-built, as-operated plant conditions

• 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

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.
It is recognized that at the design certification or combined operating license stage where the plant is not built or 5

operated, the term “as-built, as-operated” is meant to reflect the as-designed plant assuming site and operational

conditions for the given design.

Rev. 1 of RG 1.200, Page 19

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 key contributors to CDF (initiating events, accident sequences,

equipment failures and human errors)

• identification of sources of uncertainty and their impact on the results

• 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 results

• 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 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-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-built, as-operated plant.

5

The information sources include the applicable design, operation, maintenance, and engineering

characteristics of the plant.

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
Rev. 1 of RG 1.200, Page 20

Further, plant walkdowns are conducted to ensure that information sources being used actually

reflects the plant’s 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 SSC’s 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

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 stage, operational and

maintenance items may not possible. The attribute as stated may need to be revised in order to meet the

intent.
Rev. 1 of RG 1.200, Page 21

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 plant’s 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 decisionmaking. An alternative acceptable approach to using an industry consensus PRA standard is to use an

industry-developed peer review program.

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 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:

(1) The scope and level of detail that reflects the plant design, operation, and maintenance may vary.

(2) Plant-specific information versus generic information is used, such that the as-built and asoperated plant is addressed.

(3) 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.
Rev. 1 of RG 1.200, Page 22

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 as reflected in publicly available documents. (see Note below)

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-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.]
The use of the word “requirement” is standards language (e.g., in a standard, it states the standards “sets forth 6

requirements”) and is not meant to imply a regulatory requirement.

Rev. 1 of RG 1.200, Page 23

The standards are written in terms of “requirements.” These requirements will be either 6

“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. As a general rule, a requirement of a standard is met

when it is demonstrated that there is clear evidence of an intent to meet the requirement.

For process requirements, the intent, is generally straightforward and the requirement is either

met or not met. For the technical requirements, it s 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

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 standard, with a scope for a

PRA for Level 1 and limited Level 2 (LERF) for full-power operation and internal events (excluding

internal fires). 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 standard, when supplemented to account for the staff’s 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 caused by external hazards or internal fire or

caused during the low-power and shutdown modes of operation.
Rev. 1 of RG 1.200, Page 24

2.2 Industry Peer Review Program

An acceptable approach that can be used to ensure technical adequacy is to perform a peer

review of the PRA. A peer review can be used to identify the strengths and weaknesses in the PRA and

their importance to the confidence in the PRA results.

If a peer review process is used to demonstrate conformance with Regulatory Position 1, an

acceptable peer review approach is one that is performed according to an established process that

compares the PRA against the characteristics and attributes 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-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 reviewer’s 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

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. 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.
Rev. 1 of RG 1.200, Page 25

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)

The ASME standard requires a peer review to be performed. The peer review, per ASME,

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 industrydeveloped peer review program (i.e., NEI-00-02, Ref. 11), and noted in the ASME standard as an

acceptable process. The staff regulatory position on the peer review requirements in the ASME PRA

Standard and the peer review process in NEI-00-02 is provided in Appendices A and B, respectively, to

this regulatory guide. When the staff’s regulatory positions contained in Appendices A and B are taken

into account, use of a peer review can be used to demonstrate that the PRA is adequate to support a riskinformed application.

As stated earlier, the peer review is to be performed against established standards (e.g., ASME

PRA Standard). If different criteria are used than 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 resolution of the differences between the established standards, as endorsed by the NRC (i.e., ASME

PRA Standard and Appendix A to this guide) and its peer review criteria. The staff position on this

guidance (referred to as the “Licensee Self-Assessment Guidance”), is provided in Appendix B to this

guide. When the staff’s regulatory positions contained in Appendix B are taken into account, use of the

peer reviews performed using NEI-00-02 can be used to demonstrate that the PRA is adequate to support

a risk-informed application, with regard to a Level 1/LERF PRA for full-power for internal events

(excluding internal fires and external events).
Rev. 1 of RG 1.200, Page 26

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 licensee’s 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:

< 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 most PRAs do not

address this full scope, the decision-makers must make allowances for these omissions. Examples of

approaches to making allowances 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. This regulatory guide

does not address this aspect of decision-making, but it is focused specifically on the quality of the PRA

information used.
Rev. 1 of RG 1.200, Page 27

The PRA standards and industry PRA programs that have been, or are in the process of being,

developed address a specific scope. For example, the ASME PRA Standard addresses internal events

(excluding internal fire) at full-power for a limited Level 2 PRA analysis. Similarly NEI-00-02 is a peer

review process for the same scope. Note that the internal flooding is only addressed in the selfassessment 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-built and as-operated plant, which, in turn, implies that the PRA

is up to date and reflects the current design and operating practices, (2) the 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. This is recognized in RG 1.174,

which gives guidance on how to address the uncertainties. 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 staff’s regulatory position on each of these standards or

programs.
A key assumption is one that is made in response to a key source of uncertainty in the knowledge that a different 7

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

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.

Rev. 1 of RG 1.200, Page 28

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 NRC’s 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 assumptions and 7

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 staff’s review of the risk-informed submittal.
Rev. 1 of RG 1.200, Page 29

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-assessment, 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 of the

maintenance and upgrade activities are maintained, and include 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

• 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-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.
Rev. 1 of RG 1.200, Page 30

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 findings and observations

that are applicable to the parts of the PRA required for the application. This may take the

following forms:

< 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

licensee’s 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 3 of ASME

RA-Sb-2005), to determine whether they lead to limitations on the implementation of the

licensing change.
Rev. 1 of RG 1.200, Page 31

D. IMPLEMENTATION

The purpose of this section is to provide information to applicants and licensees regarding

the NRC staff’s plans for using this regulatory guide. No backfitting is intended or approved

in connection with its issuance.

Except in those cases in which an applicant or licensee proposes or has previously established

an acceptable alternative method for complying with specified portions of the NRC’s regulations, the

methods described in this guide reflect public comments and will be used in evaluating (1) submittals

in connection with applications for construction permits, standard plant design certifications, operating

licenses, early site permits, and combined licenses, and (2) submittals from operating reactor licensees

who voluntarily propose to initiate system modifications if there is a clear nexus between the proposed

modifications and the subject for which guidance is provided herein.

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 1

of Regulatory Guide 1.200. A copy of the draft regulatory analysis is available for inspection or copying for

a fee in the NRC’s Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDR’s mailing

address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800)-397-4209; fax (301)

415-3548; email PDR@nrc.gov.
All Federal Register notices listed herein were issued by the U.S. Nuclear Regulatory Commission, and are available 8

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 NRC’s Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDR’s

mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209;

fax (301) 415-3548; email PDR@nrc.gov.

All regulatory guides listed herein were published by the U.S. Nuclear Regulatory Commission. Many Regulatory 9

Guides are available electronically through the Public Electronic Reading Room on the NRC’s 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 NRC’s 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 NRC’s Public Document Room (PDR), which is located at 11555

Rockville Pike, Rockville, Maryland; the PDR’s 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.

All NUREG-series reports listed herein were published by the U.S. Nuclear Regulatory Commission, and are available 10

electronically through the Public Electronic Reading Room on the NRC’s 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 NRC’s

Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDR’s 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.

Rev. 1 of RG 1.200, Page 32

REFERENCES

1. U.S. Nuclear Regulatory Commission, “Use of Probabilistic Risk Assessment Methods in

Nuclear Activities: Final Policy Statement,” Federal Register, Vol. 60, August 16, 1995,

p. 42622 (60 FR 42622).8

2. Regulatory Guide 1.174, “An Approach for Using Probabilistic Risk Assessment in RiskInformed Decisions on Plant-Specific Changes to the Licensing Basis,” Revision 1, U.S. Nuclear

Regulatory Commission, Washington, DC, November 2002.9

3. U.S. Nuclear Regulatory Commission, Letter from M. Tschiltz to D. Lew, “Resolution of the

Regulatory Guide (RG) 1.200 Implementation Pilot Program,” June 8, 2005.

4. U.S. Nuclear Regulatory Commission, Letter from M. Drouin to J. Monninger, “Resolution of

Stakeholder Comments on DG-1161,” November 30, 2006 (available in ADAMS under

Accession #ML070040474).

5. 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, November 2002.10
Copies of ASME standards and documents may be obtained from the American Society of Mechanical Engineers, 11

Three Park Avenue, New York, NY 10016-5990; phone (212)591-8500.

Copies may be obtained from the American Nuclear Society, 555 N. Kensington Avenue, La Grange, Illinois 60526; 12

phone (708)352-6611.

All NEI documents may be obtained from the Nuclear Energy Institute, Attn: Mr. Biff Bradley, Suite 400, 1776 I

13

Street, NW, Washington, DC 20006-3708; phone (202) 739-8083.

Rev. 1 of RG 1.200, Page 33

6. Regulatory Guide 1.175, “An Approach for Plant-Specific, Risk-Informed Decisionmaking: 

Inservice Testing,” U.S. Nuclear Regulatory Commission, Washington, DC, August 1998.

7. Regulatory Guide 1.178, “An Approach for Plant-Specific, Risk-Informed Decisionmaking: 

Inservice Inspection of Piping,” Revision 1, U.S. Nuclear Regulatory Commission, Washington,

DC, September 2003.

8. Regulatory Guide 1.176, “An Approach for Plant-Specific, Risk-Informed Decisionmaking: 

Graded Quality Assurance,” U.S. Nuclear Regulatory Commission, Washington, DC, August

1998.

9. Regulatory Guide 1.177, “An Approach for Plant-Specific, Risk-Informed Decisionmaking: 

Technical Specifications,” U.S. Nuclear Regulatory Commission, Washington, DC, August 1998.

10. ASME RA-S-2002, “Standard for Probabilistic Risk Assessment for Nuclear Power Plant

Applications,” ASME, New York, New York, April 5, 2002.11

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.

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.

11. ANSI/ANS-58.21-2003, “American National Standard External-Events PRA Methodology,”

American Nuclear Society, La Grange Park, Illinois, December 2003.12

12. NEI-00-02, “Probabilistic Risk Assessment Peer Review Process Guidance,” Revision A3,

Nuclear Energy Institute, Washington, DC, March 20, 2000.13

Nuclear Energy Institute, Letter from Anthony Pietrangelo, Director of Risk- and PerformanceBased 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.

Nuclear Energy Institute, Letter from Anthony Pietrangelo, Director of Risk- and PerformanceBased 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, 2002.
All Commission papers (SECYs) listed herein were published by the U.S. Nuclear Regulatory Commission, and are 14

available electronically through the Public Electronic Reading Room on the NRC’s 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 NRC’s Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDR’s

mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209;

fax (301) 415-3548; email PDR@nrc.gov.

Rev. 1 of RG 1.200, Page 34

Nuclear Energy Institute, Letter from Anthony Pietrangelo, Director of Risk- and PerformanceBased 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.

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.

13. NEI-05-04, “Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA

Standard,” Nuclear Energy Institute, Washington, DC, January 2005.

14. SECY-00-0162, “Addressing PRA Quality In Risk-Informed Activities,” U.S. Nuclear

Regulatory Commission, Washington, DC, July 28, 2000.14

15. Regulatory Guide 1.201, “Guidelines for Categorizing Structures, Systems, and Components in

Nuclear Power Plants According to Their Safety Significance,” U.S. Nuclear Regulatory

Commission, Washington, DC, May 2006.

16. SECY-02-0176, “Proposed Rulemaking to Add New Section 10 CFR 50.69, ‘Risk-Informed

Categorization and Treatment of Structures, Systems, and Components’,” WITS 199900061,

U.S. Nuclear Regulatory Commission, Washington, DC, September 30, 2002.

17. SECY-04-0118, “Plan for the Implementation of the Commission’s Phased Approach to PRA

Quality,” U.S. Nuclear Regulatory Commission, Washington, DC, July 13, 2004.
This guide provides the regulatory position for a full-scope Level 1 and Level 2 PRA for operating LWRs. ASME RA- 15

Sb-2005 is a standard for CDF and LERF for full-power conditions and internal events (excluding internal fire). 

Consequently, the staff position on ASME RA-Sb-2005 does not include a full Level 2, low power shutdown

conditions, internal fire or external events.

Appendix A to Rev. 1 of RG 1.200, Page A-1

APPENDIX A

NRC REGULATORY POSITION ON ASME PRA STANDARD

Introduction

The American Society of Mechanical Engineers (ASME) has published ASME RA-S-2002,

“Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications” (April 5, 2002),

Addendum A to this standard (ASME RA-Sa-2003, December 5, 2003), and Addendum B to this

standard (ASME RA-Sb-2005, December 30, 2005). 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 RA-S-2002, RA-Sb-2003, and RA-Sb-2005 against the characteristics

and attributes for a technically acceptable PRA as discussed in Regulatory Positions 1 and 2 of this 15

regulatory guide. The staff’s position on each requirement (referred to in the standard as a requirement, a

high-level requirement, or a supporting requirement) in ASME RA-S-2002, RA-Sb-2003, and RA-Sb2005 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.

Table A-1 provides the staff’s position on each requirement in ASME RA-S-2002, ASME RASa-2003, and ASME RA-Sb-2005. A discussion of the staff’s 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 standard) for the staff to

have no objection are provided.
Appendix A to Rev. 1 of RG 1.200, Page A-2

Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

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 or

disapproves of information contained in

the referenced document.

Chapter 1

1.1 The standard is only for current

generation of operating lightwater reactors; the

requirements may not be

sufficient or adequate for

advanced LWRs (particularly

in the design stage), or other

types of reactors.

Clarification This Standard sets forth requirements for

Probabilistic Risk Assessments (PRAs) used

to support risk-informed decisions for

operating commercial light-water reactor

nuclear power plants, and prescribes a

method for applying these requirements for

specific applications (additional or revised

requirements may be needed for advanced

LWRs, other reactor designs, or for

reactors in the design stage).

1.2 The standard is only for

operating light-water reactors;

the requirements may not be

sufficient, adequate, or

applicable for reactors in the

design or construction stage.

Internal events includes internal

fire which is not part of this

standard.

Clarification This standard applies to PRAs used to

support applications of risk-informed

decision making related to design, licensing,

procurement, construction, operation, and

maintenance. This Standard established

requirements for ..... internal events

(excluding internal fires) while at power.

1.3 – 1.7 ----------------- No objection ----------------------------

Chapter 2

2.1 ----------------- No objection ----------------------------

2.2

Core damage The use of the term “a large

fraction of the core” should be

consistent with the definition of

“large” used in the LERF

definition.

Clarification core damage: ...involving a large fraction

of the core (i.e., sufficient, if released from

containment, has the potential to cause

offsite health effects) is anticipated.

Extremely

rare event

A frequency cutoff should be

provided as part of this

definition.

Clarification extremely rare event: one that would not be

expected to occur even once throughout the

world nuclear industry over many years (e.g.,

<1E-6/rx yr).
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-3

Internal event Internal fire is an internal

event, and not an external

event.

Qualification internal event: ...By convention, loss of

offsite power is considered to be an internal

event, and internal fire is considered to be an

external event.

Key

assumption

A assumption in the base PRA

is independent of the decision

under consideration. A key

assumption is dependent on the

decision under consideration. 

This distinction is not clear as

written. See discussion on

Regulatory Position 1.2.7 and

3.3.2.

Clarification key assumption: an assumption made in

response to a key source of uncertainty.

An assumption is made in response to a key

source of uncertainty in the knowledge ..... is

at least as sound as that of the assumptions

being challenged. 

Key source

of uncertainty

A source of uncertainty in the

base PRA is independent of the

decision under consideration. 

A key source of uncertainty is

dependent on the decision

under consideration. This

distinction is not clear as

written. See discussion on

Regulatory Position 1.2.7 and

3.3.2.

Clarification key source of uncertainty: a 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 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. A key

source of uncertainty is a source of uncertainty 

that influences or 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. 

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....
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-4

Rare event A frequency cutoff should be

provided as part of this

definition.

Clarification rare event: one that might be expected to

occur only a few times throughout the world

nuclear industry over many years (e.g., <1E4/rx yr).

Reactor-year This term references the wrong

footnote and could more

accurately reference the right

table in Section 4.5.

Clarification reactor year: a calender year in the

operating life of one reactor, regardless of

power level. See Note 2 3 in Table 4.5.1-2

(c).

Reactoroperatingstate-yearThis term references the wrong

footnote and could more

accurately reference the right

table in Section 4.5.

Clarification ...See Note 2 3 in Table 4.5.1-2 (c).

Resource

expert

See the issue discussed on

definition of “Accident

sequence, dominant.”

Clarification resource expert: A technical expert with

knowledge of a particular technical areas of

importance to a PRA.

Significant

contributor

This term is used in the

standard and a definition is

necessary.

Clarification significant contributor: (a) in the context

of an accident sequence, a significant basic

event or an initiating event that

contributes to a significant sequence; (b)

in the context of an accident progression

sequence, a contributor which is an

essential characteristic (e.g., containment

failure mode, physical phenomena) of a

significant accident progression sequence,

and if not modeled would lead to the

omission of the sequence; for example, not

modeling hydrogen detonation in an ice

condenser plant would result in a

significant LERF sequence not being

modeled.

Other

Definitions

----------------- No objection ----------------------------
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-5

Chapter 3

3.1 – 3.4 ----------------- No objection ----------------------------

3.5 Use of the word “significant”

should match definitions

provided in Section 2.2.

Clarification 2 paragraph: nd

If the PRA does not satisfy a SR for the

appropriate Capability Category, then

determine if the difference is relevant or

significant.... Acceptable requirements for

determining the significance of this

difference differences include the following:

(a) The difference is not relevant if it is not

applicable or does not affect the

quantification....

(b) The difference is not significant if the

mModeled accident sequences accounting

for at least 90% of CDF/LERF, as

applicable....

These determinations Determination of

significance will depend....

If the difference is not relevant or

significant, then the PRA is acceptable for

the application. If the difference is relevant

or significant, then....

3.6 Use of the word “safety” is not

needed.

Clarification -----------------

Chapter 4

4.1,

4.2

----------------- No objection ----------------------------

4.3

4.3.1, 

4.3.2

----------------- No objection ----------------------------

4.3.3 The use of the word “should”

does not provide a minimum

requirement.

Clarification ...The PRA analysis team shall should use

outside experts, even when....

4.3.4 thru

4.3.7

----------------- No objection ---------------------------

4.4,

4.5

----------------- No objection ----------------------------

4.5.1 - IE

4.5.1.1 ----------------- No objection ----------------------------

Table 4.5.1-1 ----------------- No objection ----------------------------

Tables 4.5.1-2(a) thru 4.5.1-2(d)
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-6

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 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 SUBSUM E

scenarios into a group....

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-5750 [Note (1)].
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-7

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 values

valves and their relevant failure modes, and

the existence, size, and positioning of relief

valves

IE-C13 ----------------- No objection --------------------------

Footnote 3 to

Table 4.5.1-

2c)

The first example makes an

assumption that the hourly

failure rate is applicable for all

operating conditions.

Clarification ...Thus,

bus at power f = 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,

IE-D2

----------------- No objection ----------------------------

IE-D3 All the sources of uncertainty

and assumptions that can

impact the risk profile of the

base PRA need to be

documented; see definition of

key source of uncertainty for

definition of source of

uncertainty.

Clarification DOCUMENT the key assumptions and key

sources of uncertainty associated ....

4.5.2 - AS

4.5.2.1 The HLR and associated SRs

are written for CDF and not

LERF; therefore, references to

LERF are not appropriate.

Clarification 4.5.2.1 Objectives. The objectives...

reflected in the assessment of CDF and

LERF is such a way that....
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-8

Table 4.5.2-1 ----------------- No objection ---------------------------

Tables 4.5.2-2(a) thru 4.5.2-2©)

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 ----------------------------

AS-C1,

AS-C2

----------------- No objection ----------------------------

AS-C3 All the sources of uncertainty

and assumptions that can

impact the risk profile of the

base PRA need to be

documented; see definition of

key source of uncertainty for

definition of source of

uncertainty.

Clarification DOCUMENT the key assumptions and key

sources of uncertainty associated ....

4.5.3 - SC

4.5.3.1 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 4.5.3-1 ----------------- No objection ----------------------------

Tables 4.5.3-2(a) thru 4.5.3-2©)

SC-A1,

SC-A2

-----------------

Note: SC-A3 was deleted in

Addendum B.

No objection ----------------------------

SC-A4 thru

SC-A6

----------------- No objection ----------------------------
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-9

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,

SC-C2

----------------- No objection ----------------------------

SC-C3 All the sources of uncertainty

and assumptions that can

impact the risk profile of the

base PRA need to be

documented; see definition of

key source of uncertainty for

definition of source of

uncertainty.

Clarification DOCUMENT the key assumptions and key

sources of uncertainty associated ....

4.5.4 - SY

4.5.4.1 ----------------- No objection ----------------------------

Table 4.5.4-1 ----------------- No objection ----------------------------

Tables 4.5.4-2(a) thru 4.5.4-2©)

SY-A1 thru

SY-A21

----------------- No objection ----------------------------

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. Clarification ...required mission time (see also ASY-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 ----------------------------
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-10

SY-C1

SY-C2

----------------- No objection ----------------------------

SY-C3 All the sources of uncertainty

and assumptions that can

impact the risk profile of the

base PRA need to be

documented; see definition of

key source of uncertainty for

definition of source of

uncertainty.

Clarification DOCUMENT the key assumptions and key

sources of uncertainty associated ....

4.5.5 - HR

4.5.5.1 ----------------- No objection ----------------------------

Table 4.5.5-1 ----------------- No objection ----------------------------

Tables 4.5.5-2(a) thru 4.5.5-2(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 preinitiator 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 ----------------------------
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-11

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

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

HR-D4 thru

HR-D7

----------------- No objection ----------------------------

Notes to

Table 4.5.5-

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. O’Hara, 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

----------------- Clarification ----------------------------

HR-G1,

HR-G2

----------------- No objection ----------------------------
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-12

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-G9

----------------- No objection ----------------------------

HR-H1 thru 

HR-H3

----------------- No objection ----------------------------

HR-I1,

HR-I3

----------------- No objection ----------------------------

HR-I3 All the sources of uncertainty

and assumptions that can

impact the risk profile of the

base PRA need to be

documented; see definition of

key source of uncertainty for

definition of source of

uncertainty.

Clarification DOCUMENT the key assumptions and key

sources of uncertainty associated ....

4.5.6 - DA

4.5.6.1 ----------------- No objection ----------------------------

Table 4.5.6-1 ----------------- No objection ----------------------------

Tables 4.5.6-2(a) thru 4.5.6-2(e)

DA-A1 thru

DA-A3

----------------- No objection ----------------------------

DA-B1, 

DA-B2

----------------- No objection ----------------------------
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-13

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/CR4550 [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, plantspecific 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 4.5.6-

2©)

Additional references cited in

the clarification to DA-C.

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.

DA-D1 Other approved statistical

processes for combining plantspecific 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....
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-14

DA-D6a,

DA-D7

----------------- No objection ----------------------------

DA-D8 New requirement needed, DAC14 was incomplete, only

provided 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,

DA-E2

----------------- No objection ----------------------------

DA-E3 All the sources of uncertainty

and assumptions that can

impact the risk profile of the

base PRA need to be

documented; see definition of

key source of uncertainty for

definition of source of

uncertainty.

Clarification DOCUMENT the key assumptions and key

sources of uncertainty associated ....

4.5.7 - IF

4.5.7.1 ----------------- No objection ----------------------------

Table 4.5.7-1 ----------------- No objection ----------------------------

Tables 4.5.7-2(a) thru 4.5.7-2(f)

IF-A1 thru 

IF-A4

----------------- No objection ----------------------------

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, ...fire protection

system....

IF-B1a thru

IF-B2

----------------- No objection ----------------------------
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-15

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 of water

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., IFB1b and IFC5),....
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-16

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:

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:

INCLUDE failure by submergence and

spray in the identification process.

ASSESS qualitatively the impact of floodinduced 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 maintenanceinduced 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 IF-D1 incorrectly references

Table 4.5.7-1 when it should

cite Table 4.5.1-2(b).

Note that IF-D2 was deleted in

Addendum B.

Clarification ...IDENTIFY the corresponding plant

initiating event group identified per Table

4.5.7-1 4.5.1-2(b)....

IF-D3 The action verb AVOID is

ambiguous.

Clarification Cat II:

AVOID subsuming DO NOT SUBSUM E

scenarios into a group....

IF-D3a thru

IF-D7

----------------- No objection ----------------------------
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-17

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,... 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,

IF-F2

----------------- No objection ----------------------------

IF-F3 All the sources of uncertainty

and assumptions that can

impact the risk profile of the

base PRA need to be

documented; see definition of

key source of uncertainty for

definition of source of

uncertainty.

Clarification DOCUMENT the key assumptions and key

sources of uncertainty associated ....

4.5.8 - QU

4.5.8.1 SRs for LERF quantification

reference the SRs in 4.5.8, and

therefore, need to be

acknowledged in 4.5.8.

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 4.5.8-1

HLR-QU-A

thru

HLR-QU-C

----------------- No objection ----------------------------

Table 4.5.8-1

HLR-QU-D

SRs for LERF quantification

reference the SRs in 4.5.8 and,

therefore, need to be

acknowledged in 4.5.8.

Clarification ...significant contributors to CDF (and

LERF), such as initiating events, accident

sequences....

Table 4.5.8-1

HLR-QU-E

All the sources of uncertainty

and assumptions that can

impact the risk profile of the

base PRA need to be identified;

see definition of key source of

uncertainty for definition of

source of uncertainty.

Clarification Uncertainties in the PRA results shall be

characterized. Key s Sources of model

uncertainty and key assumptions shall be

identified, and their potential impact on the

results understood.
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-18

Table 4.5.8-1

HLR-QU-F

----------------- No objection ----------------------------

Tables 4.5.8-2(a) thru 4.5.8-2(f)

QU-A1,

QU-A2a

----------------- No objection ----------------------------

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.

Clarification Cat II:

ESTIMATE the mean CDF from internal

events, accounting for the “state-ofknowledge” correlation between event

probabilities when significant (see NOTE 1).

QU-A3, 

QU-A4

----------------- No objection ----------------------------

QU-B1 thru

QU-B9

----------------- No objection ----------------------------

QU-C1 thru

QU-C3

----------------- No objection ----------------------------

Table 4.5.8-

2(d)

HLR-QU-D and Table 4.5.8-

2(d) objective statement just

before table need to agree; SRs

for LERF quantification

reference the SRs in 4.5.8 and,

therefore, need to be

acknowledged in 4.5.8.

Clarification ...significant contributors to CDF (and

LERF), such as initiating events, accident

sequences....

QU-D1a thru

QU-D5b

----------------- No objection ----------------------------

QU-E1 All the sources of uncertainty

that can impact the risk profile

of the base PRA need to be

assessed; see definition of key

source of uncertainty for

definition of source of

uncertainty.

Clarification IDENTIFY key sources of uncertainty.

QU-E2 All the assumptions that can

impact the risk profile of the

base PRA need to be assessed;

see definition of key

assumption for definition of

assumption.

Clarification IDENTIFY key assumptions made in the

development of the PRA model.

QU-E3 ----------------- No objection ----------------------------
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-19

QU-E4 Understanding of the model

uncertainties and assumptions

is an essential aspect of

uncertainty analysis. In

addition, all the sources of

uncertainty and assumptions

that can impact the risk profile

of the base PRA need to be

assessed; see definition of key

source of uncertainty for

definition of source of

uncertainty.

Clarification Cat I:

PROVIDE an assessment of the impact of

the key model uncertainties and

assumptions on the results of the PRA.

Cat II:

EVALUATE the sensitivity of the results to

key model uncertainties....

Cat III:

EVALUATE the sensitivity of the results to

uncertain model boundary conditions and

other key assumptions using...

QU-F1 ----------------- No objection ----------------------------

QE-F2 “QE” should be “QU”

SR needs to use defined term

“significant” instead of

“dominant.”

Clarification QEQU-F2

(g) equipment or human actions that are the

key factors in causing the accidents

sequences to be non-dominant nonsignificant.

QU-F3,

QU-F5,

QU-F6

----------------- No objection ----------------------------

QU-F4 All the sources of uncertainty

and assumptions that can

impact the risk profile of the

base PRA need to be

documented; see definition of

key source of uncertainty for

definition of source of

uncertainty. What assumptions

and uncertainties need to be

documented for the other

elements are stated in the

documentation SR for that

element.

Clarification DOCUMENT the key assumptions and key

sources of uncertainty associated with the

quantification analysis. , such as: possible

optimistic or conservative success criteria,

suitability of the reliability data, possible

modeling uncertainties (modeling limitations

due to the method selected), degree of

completeness in the selection of initiating

events, possible spatial dependencies, etc.

4.5.9 - LE

4.5.9.1 ----------------- No objection ----------------------------

Table 4.5.9-1 ----------------- No objection ----------------------------

Tables 4.5.9-2(a) thru 4.5.9-2(g)

LE-A1 thru

LE-A5

----------------- No objection ----------------------------

LE-B1 thru

LE-B3

----------------- No objection ----------------------------
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-20

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 NUREG/CR-6595, Appendix A 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 thru

LE-G3,

LE-G5,

LE-G6

----------------- No objection ----------------------------

LE-G4 All the sources of uncertainty

and assumptions that can

impact the risk profile of the

base PRA need to be

documented; see definition of

key source of uncertainty for

definition of source of

uncertainty.

Clarification DOCUMENT the key assumptions and key

sources of uncertainty associated ....

Chapter 5

5.1 ----------------- No objection ----------------------------

5.2 ----------------- No objection ----------------------------

5.3 ----------------- No objection ----------------------------

5.4 See the issue discussed on

definition of “Accident

sequence, dominant.”

Clarification 2 para: ...Changes that would impact risk- nd

informed decisions should be prioritized to

ensure that the most significant changes are

incorporated as soon as practical.

5.5, 5.6 ----------------- No objection ----------------------------

5.7 ----------------- No objection ----------------------------

5.8 (a)–(d) ----------------- No objection ----------------------------
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-21

5.8 (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 6.6)”

5.8 (f), 5.8(g) ----------------- No objection ----------------------------

Chapter 6

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 requirements in

Section 4, the “quality” (i.e.,

strengths and weaknesses) of

the PRA; this goal is to be

clearly understood by the peer

review team.

See the issue discussed on

definition of “Accident

sequence, dominant.”

Clarification “...The peer review shall assess the PRA to

the extent necessary to determine if the

methodology and its implementation meet

the requirements of this Standard 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....”

6.1.1 ----------------- No objection ----------------------------

6.1.2 ----------------- No objection ----------------------------

6.2

6.2.1, 6.2.2,

6.2.3

----------------- No objection ----------------------------

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. Some subparagraphs of para. 6.3

contain specific suggestions for the review

team to consider during the review. 

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.”

6.3.1 thru

6.3.9

----------------- No objection ----------------------------
Table A-1. Staff Position on ASME RA-S-2002, ASME RA-Sa-2003, and ASME RA-Sb-2005

Index No Issue Position Resolution

Appendix A to Rev. 1 of RG 1.200, Page A-22

6.3.9.1 ----------------- No objection ----------------------------

6.3.9.2 See the issue discussed on

definition of “Accident

sequence, dominant.”

Clarification (I) the containment response calculations,

performed specifically for the PRA, for the

dominant significant plant damage states

6.4 ----------------- No objection ----------------------------

6.5 ----------------- No objection ----------------------------

6.6

6.6.1 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

review.

Clarification “(g) identification and significance of

exceptions... including an assessment of PRA

assumptions that the reviewers have

determined to be relevant

(h) ...

(I) ...

(j) identification of the strengths and

weaknesses that have a significant impact on

the PRA

(k) an assessment of the capability

category of the SRs (i.e., identification of

what capability category is met for the

SRs)”

6.6.2 ----------------- No objection ----------------------------
The staff position is not on the historical application of the process in NEI 00-02, but on future applications of the 16

process.

Appendix B to Rev. 1 of RG 1.200, 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.

This appendix provides the staff’s position on the NEI Peer Review Process (i.e., NEI 00-02), the

proposed self-assessment process, and the self-assessment actions. The staff’s positions are categorized 16

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. The stated positions are based on the historical use of NEI 00-02 and on the performance of

a self-assessment to address those 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.
Appendix B to Rev. 1 of RG 1.200, Page B-2

Table B-1. NRC Regulatory Position on NEI 00-02

Report

Section

Regulatory

Position

Commentary/Resolution

Section 1. INTRODUCTION

1.1 Overview

and Purpose

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.

The ASME PRA Standard (with the staff’s position provided in Appendix A to

this regulatory guide) can provide an adequate basis for a peer review of an atpower, 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 the ASME PRA Standard, the staff’s 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.

1.1 Scope 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 5 of the ASME

PRA Standard which requires a peer review for PRA upgrades (PRA

methodology changes).

1.2 Historical

Perspective

No objection -------------------------------------

1.3 Process 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 staff’s position is that a peer review based on the

checklists and supplemental subtier criteria is incomplete. The NEI selfassessment process, as endorsed by the staff in this appendix, is needed.

1.4 PRA Peer

Review

Criteria and

Grades

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.

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.
Table B-1. NRC Regulatory Position on NEI 00-02

Report

Section

Regulatory

Position

Commentary/Resolution

Appendix B to Rev. 1 of RG 1.200, Page B-3

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 Objectives Clarification See comment for Section 1.1.

2.2 Process

Description

Clarification The ASME PRA Standard (with the staff’s position provided in Appendix A to

this regulatory guide) can provide an adequate basis for a peer review of an atpower, 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 the ASME PRA Standard, the staff’s 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.

Steps 4, 7, & 8 Clarification See previous comment.

2.3 PRA Peer

Review Team

Clarification The peer reviewer qualifications do not appear to be consistent with the

following requirements specified in Section 6.2 of the ASME 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
Table B-1. NRC Regulatory Position on NEI 00-02

Report

Section

Regulatory

Position

Commentary/Resolution

Appendix B to Rev. 1 of RG 1.200, Page B-4

Section 3. PRA PEER REVIEW PROCESS ELEMENTS AND GUIDANCE

3.1 No objection -------------------------------------

3.2 Criteria

and

3.3 Grading

Clarification See comment for Section 1.1.

3.3 Grading Clarification The NEI peer review process grades each PRA element from 1 to 4, while the

ASME 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.

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 Additional

Guidance on

the Technical

Elements

Review

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.
Table B-1. NRC Regulatory Position on NEI 00-02

Report

Section

Regulatory

Position

Commentary/Resolution

Appendix B to Rev. 1 of RG 1.200, Page B-5

Section 4. PEER REVIEW PROCESS RESULTS AND DOCUMENTATION

4.1 Report 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. The peer

review requirements in the ASME PRA Standard requires analysis of key

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 key 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 need to address all key

assumptions, not just those that are unique or unusual. The NEI selfassessment process needs to address those assumptions not reviewed in the

NEI peer review process. See staff position on definition of key assumption.

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 and 4.3 No objection -------------------------------------

Appendix A. PREPARATION MATERIAL FOR THE PEER TEAM REVIEW

A.1 through

A.6

No objection -------------------------------------

A.7 Sensitivity

Calculations

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 through

A.10

No objection -------------------------------------

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 PRA Standard is documented in Table B-3.
Table B-1. NRC Regulatory Position on NEI 00-02

Report

Section

Regulatory

Position

Commentary/Resolution

The NEI comparison between NEI 00-02 criteria and the ASME requirements utilized the original standard as modified

17

by subsequent addenda (A and B).

Appendix B to Rev. 1 of RG 1.200, Page B-6

Appendix C. GUIDANCE FOR THE PEER REVIEW TEAM

C.1 Purpose No objection -------------------------------------

C.2 Peer

Review Team

Mode of

Operation

No objection -------------------------------------

C.3

Recommended

Approach to

Completing

the Review

Clarification See comment for Section 4.1.

C.4 Grading Clarification/

Qualification

See the two comments on Section 3.3.

C.5 Peer

Review Team

Good Practice

List

No objection -------------------------------------

C.6 Output Qualification See the comments on Section 4.1.

C.7 Forms 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 selfassessment process and the specific actions recommended by NEI to address missing ASME standard

requirements are addressed.

17

Table B-2 provides the NRC position on the NEI self-assessment process documented in

Appendix D1 to NEI 00-02, Revision 1.
Appendix B to Rev. 1 of RG 1.200, Page B-7

Table B-2. NRC Regulatory Position on NEI Self-Assessment Process

Report Section Regulatory

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 PRA Standard and Addenda A and

B, 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 staff’s focus is on ensuring

that the self-assessment addresses important aspects of a PRA that are not

explicitly addressed in the NEI subtier criteria.

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 staff’s position is

provided in Table B-5 of this appendix.

General Notes for Self-Assessment Process

1 No objection -------------------------------------

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 needs to be addressed in the NEI selfassessment process.

3 Clarification The self-assessment process should consider the clarifications and

qualifications on Addendum B that will be provided Appendix A to RG

1.200, Rev. 1.

SelfAssessment

Process

Attributes

No objection -------------------------------------
Table B-2. NRC Regulatory Position on NEI Self-Assessment Process

Report Section Regulatory

Position

Commentary/Resolution

The NEI self-assessment process was revised to address the requirements in Addendum B of the ASME standard. 18

Appendix B to Rev. 1 of RG 1.200, Page B-8

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.

7.b thru 8. No objection -------------------------------------

9. No objection -------------------------------------

10. thru 13. No objection -------------------------------------

14. Clarification The staff’s 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

staff’s 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.

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. The staff’s position on the ASME PRA Standard (Addendum 18

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.

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.
Appendix B to Rev. 1 of RG 1.200, Page B-9

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 actions(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”):

• whether NEI’s 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 SelfAssessment 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. 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.
Appendix B to Rev. 1 of RG 1.200, Page B-10

Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Global

--- --- --- --- In performing the selfassessment action, the action

has to conform with the staff

position in Appendix A of this

document for the action to be

acceptable. 

INITIATING EVENTS 

IE-A1 Yes IE-7, IE-8, IE9, IE-10None No objection

IE-A2 Yes IE-5, IE-7, IE9, IE-10Confirm 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/conclusions

or examples from your

model. NEI 00-02 does not

explicitly mention humaninduced 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 Yes (1) IE-8, IE-9 None No objection

IE-A4 Partial IE-5, IE-7, IE9, IE-10Check for initiating events

that can be caused by a

train failure or a system

failure.

No objection

IE-A4a Partial IE-5, IE-7, IE- (1)

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-11

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

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 Yes IE-6 None No objection (3)

IE-C1 Yes IE-13, IE-15,

IE-16, IE-17

None No objection; IE-16 is the

applicable NEI 00-02 element.

IE-C1a Yes IE-13, IE-15, (1)

IE-16, IE-17

None No objection; IE-16 is the

applicable NEI 00-02 element.

IE-C1b Yes IE-13, IE-15, (1)

IE-16, IE-17

Justify recovery credit as

evidenced by procedures or

training.

No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-12

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.

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-13

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/conclusions

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

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 No None Confirm IE-C13 is met. No objection (3)

IE-D1 Partial IE-9, IE-18, IE19, IE-20Action 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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-14

IE-D2 Partial IE-9, IE-18, IE19, IE-20Action 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

AS-A7 Yes AS-4, AS-5,

AS-6, AS-7,

AS-8, AS-9

None No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-15

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/conclusions

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

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 AS11 are the applicable NEI 00-

02 elements.

AS-B3 Yes DE-10, SY-11,

TH-8, AS-10

None No objection; AS-10 and SY11 are the applicable NEI 00-

02 elements.
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-16

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 Yes DE-4, DE-5, (1)

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 Yes AS-11, AS-24, (2)

AS-25, AS-26

None No objection

AS-C2 Partial AS-11, AS-24, (2)

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 Partial QU-27, QU-28, (2)

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

-- -- --
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-17

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

SC-A4a Yes IE-6, DE-5 Confirm that this (1)

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, ST9, SY-5

None No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-18

SC-B1 Yes AS-18, SY-17,

TH-4, TH-6,

TH-7

None No objection

SC-B2 No TH-4, TH-8 NEI 00-02 does not address

this supporting

requirement. Use the

ASME standard for

requirements. Refer to SCC2.

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 Yes ST-13, SY-10, (2)

SY-17, SY-27,

TH-8, TH-9,

TH-10, AS-17,

AS-18, AS-24,

HR-30

None No objection

SC-C2 Partial ST-13, SY-10, (2)

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

SC-C3 Partial QU-27, QU-28, (2)

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

-- -- --
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-19

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 SYA3, performance of the

systems analysis would

require a review of plantspecific 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 asbuilt, as-operated plant.

No objection

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-20

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 

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

SYA12a(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

SYA12b(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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-21

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

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

SYA18a(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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-22

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

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-23

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/conclusions

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

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/conclusions

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 Yes SY-5, SY-6, (2)

SY-9, SY-18,

SY-23, SY-25,

SY-26, SY-27

None No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-24

SY-C2 Partial SY-5, SY-6, (2)

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 SY18 explains lack of Grade

provision for this subelement.

No objection

SY-C3 Partial QU-27, QU-28, (2)

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

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 preinitiators, since the

screening rules in HR-6 do

not preclude screening of

activities that can affect

multiple trains of a system.

No objection 
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-25

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 subelements) 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

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 crew’s 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

HR-D6 No Use the ASME standard for

requirements. NEI 00-02

does not address this

supporting requirement.

No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-26

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 HRE1.

HR-E2 Yes HR-8, HR-9,

HR-10, HR-21,

HR-22, HR-23,

HR-25

None No objection (HR-9 and HR10 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

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-27

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

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

crew’s response. Selfassessment needs to

document if both cognitive

and execution errors are

included in the evaluation

of HEPs.

No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-28

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

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-29

HR-G5 Partial HR-16, HR-18,

HR-20

Evaluate proper inputs per

the ASME standard or cite

peer review

documentation/conclusions

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 

HR-G6 Yes HR-12 Check to ensure they are

met by citing peer review

documentation/conclusions

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-30

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

-- -- --

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 Partial HR-28, HR-30 None No objection (2)
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-31

HR-I2 Partial HR-28, HR-30 Action is to confirm (2)

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

HR-I3 Partial QU-27, QU-28, (2)

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

DAA1a(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-

002 (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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-32

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-33

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 plantspecific 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

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-34

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

DAC11a(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

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-35

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

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. QU30 does not provide

guidance on which events

to include in the uncertainty

analysis.

No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-36

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

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/conclusions

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 & 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

DAD6a(3)Partial (see

SelfAssessment

Action)

DA-14 Plant-specific screening

and mapping of industrywide 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

DA-E1 Partial DA-1, DA-19, (2)

DA-20, DE-9

None No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-37

DA-E2 Partial DA-1, DA-19, (2)

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 Partial QU-27, QU-28, (2)

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 No Use the ASME standard for (1)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-A1b No Use the ASME standard for (1)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-A2 Deleted from

ASME PRA

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-38

IF-B1 No Use the ASME standard for

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-B1a No Use the ASME standard for (4)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-B1b No Use the ASME standard for (3)

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 No Use the ASME standard for (3)

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

IF-C2 No Use the ASME standard for

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-C2a No Use the ASME standard for (1)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-C2b No Use the ASME standard for (2)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-C2c No Use the ASME standard for (5)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-39

IF-C3 No Use the ASME standard for

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-C3a No Use the ASME standard for (1)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-C3b No Use the ASME standard for (3)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-C3c No Use the ASME standard for (6)

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 No Use the ASME standard for (4)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-C5 No Use the ASME standard for

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-C5a No Use the ASME standard for (1)

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 No Use the ASME standard for (3)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-40

IF-C8 No Use the ASME standard for (3)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-C9 No Use the ASME standard for (3)

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 No Use the ASME standard for (3)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

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 No Use the ASME standard for (1)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-D6 No Use the ASME standard for (3)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-D7 No Use the ASME standard for (3)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-41

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 No Use the ASME standard for (3)

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

IF-E5a No Use the ASME standard for (1)

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 No Use the ASME standard for (1)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-E6b No Use the ASME standard for (1)

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

IF-E8 No Use the ASME standard for (3)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-42

IF-F1 No Use the ASME standard for (2)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-F2 No Use the ASME standard for (2)

requirements. NEI 00-02

does not address this

supporting requirement.

No objection

IF-F3 No Use the ASME standard for (2)

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

QU-A2a Yes QU-8 None No objection

QUA2b(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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-43

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 QU23 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.

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 provides

some do’s and don’ts 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

QUB7b(1)Yes QU-26 None No objection

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-44

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

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.
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-45

QUD1b(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.

QUD1c(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

-- -- --

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 QU31 (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

QUD5b(5)No Confirm that this

requirement is met.

No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-46

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.

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.
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-47

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 Partial QU-31, QU-32, (2)

QU-34

None No objection

QU-F2 Yes MU-7, QU-4, (2)

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

QU-F3 Partial QU-31 Use the ASME standard for (2)

requirements at the time of

doing an application.

No objection

QU-F4 No QU-27, QU-28, (2)

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 No Use the ASME standard for (2)

requirements at the time of

doing an application. NEI

00-02 does not address this

supporting requirement.

No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-48

QU-F6 No Use the ASME standard for (3)

requirements at the time of

doing an application. NEI

00-02 does not address this

supporting requirement.

No objection

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-49

LE-A2 Partial L2-7, L2-8, AS21Confirm 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

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-50

LE-A4 Partial L2-7,L2-8, L2-

9, 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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-51

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

LE-B3 No NEI 00-02 does not address (3)

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-52

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 LEC2a.

No objection

LE-C2b Partial L2-9, L2-12, (1)

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

LE-C3 Partial L2-8, L2-24,

L2-25

Confirm that the

justification for inclusion of

any of the features listed in

LE-C3 meet 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 Partial L2-11, L2-12 Confirm requirements of (1)

LE-C8b are implemented in

the PRA.

No objection
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-53

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 Partial AS-20, L2-11, (1)

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 selfassessment.

Containment bypass is

explicitly identified in the

failure modes addressed by

the LERF analysis.

No objection

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

LED1b(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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-54

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

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-55

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

LE-E3 No NEI 00-02 does not address (3)

this supporting

requirement. Use the

ASME PRA Standard for

Supporting Requirement

LE-E3.

 No objection

LE-E4 Partial QU sub- (7)

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 Yes L2-26 None No objection (1)

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
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-56

LE -F3 No NEI 00-02 does not address (3)

this supporting

requirement. Use the

ASME standard for

Supporting Requirement

LE-F3

No objection

LE-G1 Yes L2-26, L2-27, (2)

L2-28

None No objection

LE-G2 Partial L2-26, L2-27, (2)

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 Partial L2-26, L2-27, (2)

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 Partial QU-27, QU-28, (2)

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.
Table B-4. NRC Regulatory Position on Industry Self-Assessment Actions

NEI Assessment

Regulatory Position

ASM E

STD SR

Addressed by 

NEI 00-02?

Applicable

NEI 00-02

Elements

Industry Self-Assessment

Actions

Appendix B to Rev. 1 of RG 1.200, Page B-57

LE-G5 Partial L2-26, L2-27, (2)

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 No NEI 00-02 does not address (3)

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.

NRC regulatory position on NEI-05-04, “Process for Performing Follow-On PRA Peer Review Using the

ASME PRA Standard,” is provided below in Table B.5.
Appendix B to Rev. 1 of RG 1.200, Page B-58

Table B-5. NRC Regulatory Position on NEI 05-04

Report Section Regulatory

Position

Commentary/Resolution

Section 1.0. INTRODUCTION

1.1 Purpose No objection -----------------------------------

1.2 Background No objection -----------------------------------

1.3 Scope No objection

Section 2.0. GENERAL OVERVIEW OF PEER REVIEW PROCESS

1 paragraph Clarification A follow-on peer review of an at-power, internal events PRA (including st

internal flooding) that uses as criteria the supporting requirements of

Chapter 4 of the ASME PRA Standard needs to address the staff’s position

provided in Appendix A to this regulatory guide to be acceptable to the staff

for a regulatory application.

4 paragraph Clarification Per Section 6.3 of the ASME PRA Standard, the staff position is that, in th

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.

Section 3.0. GRADING PROCESS

1 paragraph Clarification NEI 05-04 indicates that one of the outcomes of the follow-on peer review st

process is the assignment of grades for each SR that are used to indicate the

relative capability level of each PRA technical element. Since the use of a

PRA for risk-informed applications needs to be determined at the SR level,

the staff does not utilize an overall PRA technical element capability level in

the assessment of a PRA for specific applications.

2 paragraph Clarification NEI states that it is essential to focus the peer review on the specific

nd

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.

3.1 Grading

Process for Peer

Reviews Against

ASME PRA

Standard

2 paragraph nd

Clarification A follow-on peer review of an at-power, internal events PRA (including

internal flooding) that uses as criteria the supporting requirements of

Chapter 4, and the requirements of Chapter 5 of the ASME PRA Standard

needs to address the staff’s position provided in Appendix A to this

regulatory guide to be acceptable to the staff for a regulatory application.

5 paragraph Clarification NEI 05-04 indicates that although no grades are assigned to HLRs, a th

qualitative assessment of the HLRs will be made based on the associated SR

grades. The staff’s position is consistent with the ASME 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.
Table B-5. NRC Regulatory Position on NEI 05-04

Report Section Regulatory

Position

Commentary/Resolution

Appendix B to Rev. 1 of RG 1.200, Page B-59

3.2 Comparison

Against Grading

Process for 

NEI 00-02

Clarification 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 PRA

Standard, the staff’s 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 The NEI 00-02 peer review process grades each PRA element from 1 to 4,

while the ASME 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)

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

05-04. (Staff comment on Section 3.3 on NEI 00-02)
Table B-5. NRC Regulatory Position on NEI 05-04

Report Section Regulatory

Position

Commentary/Resolution

Appendix B to Rev. 1 of RG 1.200, Page B-60

Section 4.0. FOLLOW-ON PEER REVIEW: ASME PRA STANDARD SCOPE

4.1 Scope Clarification The staff accepts that in addition to performing a follow-on peer review of a

PRA update, the process in NEI 05-04 can be used to validate the selfassessment performed under NEI 00-02 Appendix D guidance (referred to in

NEI 05-04 as a gap-analysis), as endorsed in this appendix. The use of the

results of the NEI 00-02 self-assessment can be used to focus such a review. 

However, for a follow-on peer review of a PRA upgrade, the staff’s position

is that all pertinent SRs must be reviewed.

4.2 Host Utility

Requirements

No objection ---------------------------------

4.3 SelfAssessmentClarification The staff interpretation of NEI 00-02 Grades 2, 3, and 4 is that, they

correspond broadly to the ASME PRA Standard 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. Thus, although it is

reasonable to assign an SR that received a Grade 3 or 4 in the NEI 00-02

review as a Capability Category II, it is not reasonable to assume a Grade 2

corresponds to Capability Category I. (Staff comment on Section 3.3 on

NEI 00-02)

4.5 Peer Review

Schedule

No objection ------------------------------------------

4.6 Peer Review

Process

4 paragraph th

Qualification NEI 05-04 states that a reviewer’s assessment whether each SR meets the

ASME PRA Standard should be derived from what is in the standard and not

based on the staffs clarifications and qualifications of the SRs provided in

Appendix A to this regulatory guide. The staff’s position is that, when used

to support a regulatory application, the assigned SR grades accepted by the

NRC for a specific application will include consideration of the

clarifications and qualifications to the ASME PRA Standard provided in

Appendix A.

9 and 10 th th

paragraphs

Clarification Section 6.1 of the ASME 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 staff’s 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 staff’s 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.
Table B-5. NRC Regulatory Position on NEI 05-04

Report Section Regulatory

Position

Commentary/Resolution

Appendix B to Rev. 1 of RG 1.200, Page B-61

APPENDICES

Appendix A

Sample Fact and

Observation

Form

No objection ------------------------------------

Appendix B

Sample

Summary Tables

No objection ------------------------------------

Appendix C

Maintenance and

Update Process

Review

Checklist

No objection ---------------------------------