The NRC issues regulatory guides to describe and make available to the public methods that the NRC staff considers acceptable 

for use in implementing specific parts of the agency=s regulations, techniques that the staff uses in evaluating specific problems or 

postulated accidents, and data that the staff needs in reviewing applications for permits and licenses. Regulatory guides are not 

substitutes for regulations, and compliance with them is not required. Methods and solutions that differ from those set forth in 

regulatory guides will be deemed acceptable if they provide a basis for the findings required for the issuance or continuance of a 

permit or license by the Commission. 

This guide was issued after consideration of comments received from the public. 

Regulatory guides are issued in 10 broad divisionsC1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials 

Facilities; 4, Environmental and Siting; 5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health; 

9, Antitrust and Financial Review; and 10, General. 

Electronic copies of this guide and other recently issued guides are available through the NRC=s public Web site under the 

Regulatory Guides document collection of the NRC=s Electronic Reading Room at 

http://www.nrc.gov/reading-rm/doc-collections/ and through the NRC=s Agencywide Documents Access and Management 

System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML100910006. 

U.S. NUCLEAR REGULATORY COMMISSION May 2011

Revision 2

REGULATORY GUIDE

OFFICE OF NUCLEAR REGULATORY RESEARCH

REGULATORY GUIDE 1.174 

(Draft was issued as DG-1226, dated August 2009)

AN APPROACH FOR USING PROBABILISTIC RISK 

ASSESSMENT IN RISK-INFORMED DECISIONS ON PLANTSPECIFIC CHANGES TO THE LICENSING BASIS 

A. INTRODUCTION 

The U.S. Nuclear Regulatory Commission’s (NRC’s) policy statement on probabilistic risk 

assessment (PRA) (Ref. 1) encourages greater use of this analysis technique to improve safety 

decisionmaking and improve regulatory efficiency. A description of current risk-informed initiatives may 

be found in (1) recent updates to the NRC staff’s Risk-Informed and Performance-Based Plan (RPP) 

formerly known as the Risk-Informed Regulation Implementation Plan (Ref. 2), and (2) the agency 

Internet site at http://www.nrc.gov/about-nrc/regulatory/risk-informed.html. 

One significant activity undertaken in response to the policy statement is the use of PRA to 

support decisions to modify an individual plant’s licensing basis (LB).1

 This regulatory guide provides 

guidance on the use of PRA findings and risk insights to support licensee requests for changes to a plant’s 

LB, as in requests for license amendments and technical specification changes under Title 10 of the Code 

 

1

 These are modifications to a plant’s design, operation, or other activities that require NRC approval. These 

modifications could include items such as exemption requests under 10 CFR 50.11 and license amendments under 10 

CFR 50.90. 
Rev. 2 of RG 1.174, Page 2 

of Federal Regulations (10 CFR) Sections 50.90, “Application for Amendment of License, Construction 

Permit, or Early Site Permit,” through 50.92, “Issuance of Amendment.” It does not address licenseeinitiated changes to the LB that do NOT require NRC review and approval (e.g., changes to the facility as 

described in the final safety analysis report (FSAR), the subject of 10 CFR 50.59, “Changes, Tests, and 

Experiments”). 

The staff usually evaluates licensee-initiated LB changes that are consistent with currently 

approved staff positions (e.g., regulatory guides, standard review plans, branch technical positions, or the 

Standard Technical Specifications) using traditional engineering analyses. Generally, the staff does not 

expect a licensee to submit risk information in support of the proposed change. 

The staff may evaluate licensee-initiated LB change requests that go beyond current staff 

positions using traditional engineering analyses as well as the risk-informed approach set forth in this 

regulatory guide. The staff may request a licensee to submit supplemental risk information if such 

information is not submitted by the licensee. If the licensee does not provide risk information on the 

proposed LB change, the staff will review the information provided by the licensee to determine whether 

the application can be approved. Based on the information provided, using traditional methods, the NRC 

staff will either approve or reject the application. 

However, licensees should be aware that special circumstances may arise in which new 

information reveals an unforeseen hazard or a substantially greater potential for a known hazard to occur, 

such as the identification of an issue related to the requested LB change that may substantially increase 

risk. In such circumstances, the NRC has the statutory authority to require licensee action above and 

beyond existing regulations and may request an analysis of the change in risk related to the requested LB 

change to demonstrate that the level of protection necessary to avoid undue risk to public health and 

safety (i.e., “adequate protection”) would be maintained upon approval of the requested LB change. 

This regulatory guide describes an acceptable method for the licensee and NRC staff to use in 

assessing the nature and impact of LB changes when the licensee chooses to support, or is requested by 

the staff to support, the changes with risk information. The NRC staff will review these LB changes by 

considering engineering issues and applying risk insights. Licensees that submit risk information 

(whether on their own initiative or at the request of the staff) should address each of the principles of riskinformed regulation discussed in this regulatory guide. Licensees should identify how their chosen 

approaches and methods (whether quantitative or qualitative, deterministic or probabilistic), data, and 

criteria for considering risk are appropriate for the decision to be made. 

Appendix D to Section 19.2 of the Standard Review Plan (SRP) (Ref. 3) provides the staff with 

additional guidance regarding the circumstances and process under which NRC staff reviewers would 

request and use risk information in the review of nonrisk-informed license amendment requests. 

The guidance provided in this regulatory guide does not preclude other approaches for requesting 

changes to the LB. Rather, the staff intends for this regulatory guide to improve consistency in regulatory 

decisions in areas in which the results of risk analyses are used to help justify regulatory action. As such, 

the principles, process, and approach discussed herein also provide useful guidance for the application of 

risk information to a broader set of activities than plant-specific changes to a plant’s LB (i.e., generic 

activities), and licensees are encouraged to use this guidance in that regard. 

The NRC issues regulatory guides to describe to the public methods that the staff considers 

acceptable for use in implementing specific parts of the agency=s regulations, to explain techniques that 

the staff uses in evaluating specific problems or postulated accidents, and to provide guidance to 
Rev. 2 of RG 1.174, Page 3 

applicants. Regulatory guides are not substitutes for regulations and compliance with them is not 

required. 

This regulatory guide contains information collection requirements covered by 10 CFR Part 50, 

“Domestic Licensing of Production and Utilization Facilities,” that 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. The NRC has determined that this 

regulatory guide is not a major rule as designated by the Congressional Review Act and has verified this 

determination with OMB. 

B. DISCUSSION 

Background 

Both the NRC and the nuclear industry recognize that PRA has evolved to the point that it can be 

used increasingly as a tool in regulatory decisionmaking. In August 1995, the NRC adopted the 

following policy statement (Ref. 1) regarding the expanded use of PRA: 

• The use of PRA technology should be increased in all regulatory matters to the extent 

supported by the state-of-the-art in PRA methods and data and in a manner that 

complements the NRC’s deterministic approach and supports the NRC’s traditional 

defense-in-depth philosophy. 

• PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and 

importance measures) should be used in regulatory matters, where practical within the 

bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current 

regulatory requirements, regulatory guides, license commitments, and staff practices. 

Where appropriate, PRA should be used to support the proposal for additional regulatory 

requirements in accordance with 10 CFR 50.109 (Backfit Rule). Appropriate procedures 

for including PRA in the process for changing regulatory requirements should be 

developed and followed. It is, of course, understood that the intent of this policy is that 

existing rules and regulations shall be complied with unless these rules and regulations 

are revised. 

• PRA evaluations in support of regulatory decisions should be as realistic as practicable 

and appropriate supporting data should be publicly available for review. 

• The Commission’s safety goals for nuclear power plants and subsidiary numerical 

objectives are to be used with appropriate consideration of uncertainties in making 

regulatory judgments on need for proposing and backfitting new generic requirements on 

nuclear power plant licensees. 

In its approval of the policy statement, the Commission articulated its expectation that 

implementation of the policy statement will improve the regulatory process in three areas: (1) foremost, 

through safety decisionmaking enhanced by the use of PRA insights, (2) through more efficient use of 

agency resources, and (3) through a reduction in unnecessary burdens on licensees. 

In parallel with the publication of the policy statement, the staff developed an implementation 

plan to define and organize the PRA-related activities being undertaken (Ref. 2). These activities cover a 

wide range of PRA applications and involve the use of a variety of PRA methods (with variety including 
Rev. 2 of RG 1.174, Page 4 

both types of models used and the detail of modeling needed). For example, one application involves the 

use of PRA in the assessment of operational events in reactors. The characteristics of these assessments 

permit relatively simple PRA models to be used. In contrast, other applications require the use of detailed 

models. 

The activities described in the RPP (Ref. 2) with its updates, and on the agency’s public Internet 

site (see Part A of this regulatory guide) relate to a number of agency interactions with the regulated 

industry. With respect to reactor regulation, activities include, for example, developing guidance for 

NRC inspectors on focusing inspection resources on risk-important equipment and reassessing plants with 

relatively high core damage frequencies (CDFs) for possible backfit. 

This regulatory guide focuses on the use of PRA in a subset of the applications described in the 

staff’s implementation plan. Its principal focus is the use of PRA findings and risk insights in decisions 

on proposed changes to a plant’s LB. 

This regulatory guide also makes use of the NRC’s Safety Goal Policy Statement (Ref. 4). As 

discussed below, one key principle in risk-informed regulation is that proposed increases in CDF and risk 

are small and are consistent with the intent of the Commission’s Safety Goal Policy Statement. The 

safety goals (and associated quantitative health objectives (QHOs)) define an acceptable level of risk that 

is a small fraction (0.1 percent) of other risks to which the public is exposed. The acceptance guidelines 

defined in this regulatory guide (in Section 2.4) are based on subsidiary objectives derived from the safety 

goals and their QHOs. 

Purpose of This Regulatory Guide 

Changes to many of the activities and design characteristics in a nuclear power plant’s LB require 

NRC review and approval. This regulatory guide provides the staff’s recommendations for using risk 

information in support of licensee-initiated LB changes to a nuclear power plant that require such review 

and approval. The guidance provided does not preclude other approaches for requesting LB changes. 

Rather, this regulatory guide is intended to improve consistency in regulatory decisions in areas in which 

the results of risk analyses are used to help justify regulatory action. As such, this regulatory guide, the 

use of which is voluntary, provides general guidance concerning one approach that the NRC has 

determined to be acceptable for analyzing issues associated with proposed changes to a plant’s LB and for 

assessing the impact of such proposed changes on the risk associated with plant design and operation. 

This guidance does not address the specific analyses needed for each nuclear power plant activity or 

design characteristic that may be amenable to risk-informed regulation. Additional or revised guidance 

might be provided for new reactors (e.g., advanced light-water reactors) licensed under 10 CFR Part 52, 

"Licenses, Certifications, and Approvals for Nuclear Power Plants. 

Scope of This Regulatory Guide 

This regulatory guide describes an acceptable approach for assessing the nature and impact of 

proposed LB changes by considering engineering issues and applying risk insights. 

Assessments should consider relevant safety margins and defense-in-depth attributes, including 

consideration of success criteria as well as equipment functionality, reliability, and availability. The 

analyses should reflect the actual design, construction, and operational practices of the plant. Acceptance 

guidelines for evaluating the results of such assessments are provided. This guide also addresses 

implementation strategies and performance monitoring plans associated with LB changes that will help to 

ensure that assumptions and analyses supporting the change are verified. 
Rev. 2 of RG 1.174, Page 5 

Consideration of the Commission’s Safety Goal Policy Statement (Ref. 4) is an important 

element in regulatory decisionmaking. Consequently, this regulatory guide provides acceptance 

guidelines consistent with this policy statement. 

In theory, one could construct a more generous regulatory framework for consideration of those 

risk-informed changes that may have the effect of increasing risk to the public. Such a framework would 

include, of course, assurance of continued adequate protection (that level of protection of the public 

health and safety that must be reasonably assured regardless of economic cost). But it could also include 

provision for possible elimination of all measures not needed for adequate protection, which either do not 

effect a substantial reduction in overall risk or result in continuing costs that are not justified by the safety 

benefits. Instead, in this regulatory guide, the NRC has chosen a more restrictive policy that would 

permit only small increases in risk and only when it is reasonably assured, among other things, that 

sufficient defense-in-depth and sufficient margins are maintained. This policy is adopted because of 

uncertainties and to account for the fact that safety issues continue to emerge regarding design, 

construction, and operational matters notwithstanding the maturity of the nuclear power industry. These 

factors suggest that nuclear power reactors should operate routinely only at a prudent margin above 

adequate protection. The safety goal subsidiary objectives are used as an example of such a prudent 

margin. 

Finally, this regulatory guide indicates an acceptable level of documentation that will enable the 

staff to reach a finding that the licensee has performed a sufficiently complete and scrutable analysis and 

that the results of the engineering evaluations support the licensee’s request for a regulatory change. 

Relationship to Other Guidance Documents 

Directly relevant to this regulatory guide is the SRP (Ref. 3) designed to guide the NRC staff 

evaluations of licensee requests for changes to the LB that apply risk insights, as well as guidance 

developed in selected application-specific regulatory guides and the corresponding SRP chapters. The 

NRC has developed related regulatory guides on inservice testing, inservice inspection and technical 

specifications (Refs. 5, 7, and 8). The agency withdrew a regulatory guide covering graded quality 

assurance (Ref. 6), which was superseded by Regulatory Guide 1.201, Trial Use, “Guidelines for 

Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety 

Significance,” issued May 2006 (Ref. 9). An NRC contractor report (Ref. 10) is also available that 

provides a simple screening method for assessing one measure used in the regulatory guide—large early 

release frequency (LERF). The staff recognizes that the risk analyses necessary to support regulatory 

decisionmaking may vary with the relative weight that is given to the risk assessment element of the 

decisionmaking process. The burden is on the licensee who requests a change to the LB to justify that the 

chosen risk assessment approach, methods, and data are appropriate for the decision to be made. 

This regulatory guide also invokes Regulatory Guide 1.200, “An Approach for Determining the 

Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities.” (Ref. 11) 

Regulatory Guide 1.200 provides the necessary guidance for determining the technical adequacy of the 

base PRA. Figure 1 (taken from Regulatory Guide 1.200) shows the relationship of this regulatory guide 

to risk-informed activities, other application-specific guidance, Regulatory Guide 1.200, consensus PRA 

standards, and industry programs. 
Rev. 2 of RG 1.174, Page 6 

Figure 1 Relationship of Regulatory Guide 1.174 to other risk-informed guidance 

C. REGULATORY POSITION 

In its approval of the policy statement on the use of PRA methods in nuclear regulatory activities 

(Ref. 1), the Commission stated its expectation that “the use of PRA technology should be increased in all 

regulatory matters...in a manner that complements the NRC’s deterministic approach and supports the 

NRC’s traditional defense-in-depth philosophy.” The use of risk insights in licensee submittals 

requesting LB changes will assist the staff in the disposition of such licensee proposals. 

The staff has defined an acceptable approach to analyzing and evaluating proposed LB changes. 

This approach supports the NRC’s desire to base its decisions on the results of traditional engineering 

evaluations, supported by insights (derived from the use of PRA methods) about the risk significance of 

the proposed changes. Decisions concerning proposed changes 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. 

In implementing risk-informed decisionmaking, LB changes are expected to meet a set of key 

principles. Some of these principles are written in terms typically used in traditional engineering 

decisions (e.g., defense-in-depth). While written in these terms, it should be understood that risk analysis 

techniques can be, and are encouraged to be, used to help ensure and show that these principles are met. 

These principles include the following: 

1. The proposed change meets the current regulations unless it is explicitly related to a requested 

exemption (i.e., a specific exemption under 10 CFR 50.12, “Specific Exemptions”). 

2. The proposed change is consistent with a defense-in-depth philosophy. 
Rev. 2 of RG 1.174, Page 7 

3. The proposed change maintains sufficient safety margins. 

4. When proposed changes result in an increase in CDF or risk, the increases should be small and 

consistent with the intent of the Commission’s Safety Goal Policy Statement (Ref. 4).2

5. The impact of the proposed change should be monitored using performance measurement 

strategies. 

Each of these principles should be considered in the risk-informed, integrated decisionmaking 

process, as illustrated in Figure 2. 

Figure 2 Principles of risk-informed integrated decisionmaking 

The staff’s proposed evaluation approach and acceptance guidelines follow from these principles. 

In implementing these principles, the staff expects the following: 

• All safety impacts of the proposed change are evaluated in an integrated manner as part of an 

overall risk management approach in which the licensee is using risk analysis to improve 

operational and engineering decisions broadly by identifying and taking advantage of 

opportunities to reduce risk and not just to eliminate requirements the licensee sees as 

undesirable. For those cases in which risk increases are proposed, the benefits should be 

described and should be commensurate with the proposed risk increases. The approach used to 

identify changes in requirements should be used to identify areas in which requirements should be 

increased as well as those in which they can be reduced. 

• The scope, level of detail, and technical acceptability of the engineering analyses (including 

traditional and probabilistic analyses) conducted to justify the proposed LB change should (1) be 

appropriate for the nature and scope of the change, (2) be based on the as-built and as-operated 

and maintained plant, and (3) reflect operating experience at the plant. 

 

2

 For purposes of this guide, a proposed LB change that meets the acceptance guidelines discussed in Section 2.2.4 of 

this regulatory guide is considered to have met the intent of the policy statement. 
Rev. 2 of RG 1.174, Page 8 

• The plant-specific PRA supporting the licensee’s proposals has been demonstrated to be of 

sufficient technical adequacy. 

• Appropriate consideration of uncertainty is given in the analyses and interpretation of findings, 

including use of a program of monitoring, feedback, and corrective action to address key sources 

of uncertainty. NUREG-1855, “Guidance on the Treatment of Uncertainties Associated with 

PRAs in Risk-Informed Decision Making,” (Ref. 12) provides further guidance. 

• The use of CDF and LERF3

 as bases for PRA acceptance guidelines is an acceptable approach to 

addressing Principle 4. Use of the Commission’s Safety Goal QHOs in lieu of LERF is 

acceptable in principle, and licensees may propose their use. However, in practice, implementing 

such an approach would require an extension to a Level 3 PRA, in which case the methods and 

assumptions used in the Level 3 analysis, and associated uncertainties, would require additional 

attention. 

• Increases in estimated CDF and LERF resulting from proposed LB changes will be limited to 

small increments. The cumulative effect of such changes, risk increase and risk decrease (if 

available), should be tracked and considered in the decision process. 

• The acceptability of proposed changes should be evaluated by the licensee in an integrated 

fashion that ensures that all principles are met. 

• Data, methods, and assessment criteria used to support regulatory decisionmaking must be well 

documented and available for public review. 

Given the principles of risk-informed decisionmaking discussed above, the staff has identified a 

four-element approach to evaluating proposed LB changes. This approach, which Figure 3 presents 

graphically, supports the NRC’s decisionmaking process. This approach is not sequential in nature; rather 

it is iterative. 

Figure 3 Principal elements of risk-informed, plant-specific decisionmaking 

 

3

 In this context, LERF is being used as a surrogate for the early fatality QHO. It is defined as the sum of the frequencies 

of those accidents leading to rapid, unmitigated release of airborne fission products from the containment to the 

environment occurring before the effective implementation of offsite emergency response and protective actions such 

that there is the potential for early health effects. Such accidents generally include unscrubbed releases associated with 

early containment failure shortly after vessel breach, containment bypass events, and loss of containment isolation. 

This definition is consistent with accident analyses used in the safety goal screening criteria discussed in the 

Commission’s regulatory analysis guidelines. An NRC contractor’s report (Ref. 10) describes a simple screening 

approach for calculating LERF. 
Rev. 2 of RG 1.174, Page 9 

1. Element 1: Define the Proposed Change 

Element 1 involves three primary activities. First, the licensee should identify those aspects of 

the plant’s LB that may be affected by the proposed change, including but not limited to rules and 

regulations, FSAR, technical specifications, licensing conditions, and licensing commitments. Second, 

the licensee should identify all structures, systems, and components (SSCs), procedures, and activities 

that are covered by the LB change being evaluated and should consider the original reasons for including 

each program requirement. 

When considering LB changes, a licensee may identify regulatory requirements or commitments 

in its LB that it believes are overly restrictive or unnecessary to ensure safety at the plant. Note that the 

corollary is also true; that is, licensees are also expected to identify design and operational aspects of the 

plant that should be enhanced consistent with an improved understanding of their safety significance. 

Such enhancements should be embodied in appropriate LB changes that reflect these enhancements. 

Third, with this staff expectation in mind, the licensee should identify available engineering 

studies, methods, codes, applicable plant-specific and industry data and operational experience, PRA 

findings, and research and analysis results relevant to the proposed LB change. With particular regard to 

the plant-specific PRA, the licensee should assess the capability to use, refine, augment, and update 

system models as needed to support a risk assessment of the proposed LB change. 

The above information should be used collectively to describe the LB change and to outline the 

method of analysis. The licensee should describe the proposed change and how it meets the objectives of 

the NRC’s PRA Policy Statement (Ref. 1), including enhanced decisionmaking, more efficient use of 

resources, and reduction of unnecessary burden. In addition to improvements in reactor safety, this 

assessment may consider benefits from the LB change such as reduced fiscal and personnel resources and 

radiation exposure. The licensee should affirm that the proposed LB change meets the current regulations 

unless the proposed change is explicitly related to a proposed exemption (i.e., a specific exemption under 

10 CFR 50.12). 

1.1 Combined Change Requests

Licensee proposals may include several individual changes to the LB that have been evaluated 

and will be implemented in an integrated fashion. With respect to the overall net change in risk, 

combined change requests (CCRs) are to fall in one of the following two broad categories, each of which 

may be acceptable: 

1. CCRs in which any individual change increases risk; or 

2. CCRs in which each individual change decreases risk. 

In the first category, the contribution of each individual change in the CCR must be quantified in 

the risk assessment and the uncertainty of each individual change must be addressed. For CCRs in the 

second category, qualitative analysis may be sufficient for some or all individual changes. Guidelines for 

use in developing CCRs are discussed below. 

1.2 Guidelines for Developing Combined Change Requests

The changes that make up a CCR should be related to one another (e.g., they affect the same 

single system or activity, they affect the same safety function or accident sequence or group of sequences, 

or they are the same type, such as changes in outage time allowed by technical specifications). However, 

this does not preclude acceptance of unrelated changes. When CCRs are submitted to the NRC staff for 
Rev. 2 of RG 1.174, Page 10 

review, the relationships among the individual changes and how they have been modeled in the risk 

assessment should be addressed in detail, since this will control the characterization of the net result of the 

changes. Licensees should evaluate not only the individual changes, but also the changes taken together, 

against the safety principles and qualitative acceptance guidelines in Part C of this regulatory guide. In 

addition, the acceptability of the cumulative impact of the changes that make up the CCR with respect to 

the quantitative acceptance guidelines discussed in Section 2.4 of this guide should be assessed. 

In implementing CCRs in the first category, the risk from significant accident sequences will not 

be increased and the frequencies of the lower ranked contributors will not be increased so that they 

become significant contributors to risk. No significant new sequences or cutsets are to be created. In 

assessing the acceptability of CCRs, (1) risk increases related to the more likely initiating events (e.g., 

steam generator tube ruptures) should not be traded against improvements related to unlikely events (e.g., 

earthquakes) even if, for instance, they involve the same safety function, and (2) risk should be 

considered in addition to likelihood. The staff also expects that CCRs will lead to safety benefits, such as 

simplifying plant operations or focusing resources on the most important safety items. 

Proposed changes that modify one or more individual components of a previously approved CCR 

must also address the impact on the previously approved CCR. Specifically, the licensee should address 

whether the proposed modification would cause the previously approved CCR to become unacceptable. 

If this is the case, the submittal should address the actions the licensee is taking with respect to the 

previously approved CCR. 

2. Element 2: Perform Engineering Analysis 

The scope, level of detail, and technical adequacy of the engineering analyses conducted to justify 

any proposed LB change should be appropriate for the nature and scope of the proposed change. The 

licensee should appropriately consider uncertainty in the analysis and interpretation of findings. The 

licensee should use judgment on the complexity and difficulty of implementing the proposed LB change 

in deciding upon appropriate engineering analyses to support regulatory decisionmaking. Thus, the 

licensee should consider the appropriateness of qualitative and quantitative analyses, as well as analyses 

using traditional engineering approaches and those techniques associated with the use of PRA findings. 

Regardless of the analysis methods chosen, the licensee must show that it has met the principles set forth 

in Part C of this regulatory guide through the use of scrutable acceptance guidelines established for 

making that determination. 

Some proposed LB changes can be characterized as involving the categorization of SSCs 

according to safety significance. An example is grading the application of special treatment requirements 

commensurate with the safety significance of equipment under 10 CFR 50.69, “Risk-Informed 

Categorization and Treatment of Structures, Systems and Components for Nuclear Power Plants.” Like 

other applications, the staff’s review of LB change requests for applications involving safety 

categorization will be in accordance with the acceptance guidelines associated with each key principle 

presented in this regulatory guide, unless the licensee proposes alternative, equivalent guidelines. Since 

risk-importance measures are often used in such categorizations, Appendix A to this regulatory guide 

provides guidance on their use. Other application-specific guidance documents address guidelines 

associated with the adequacy of programs (in this example, special treatment requirements) implemented 

for different safety-significant categories (e.g., more safety significant and less safety significant). 

Licensees are encouraged to apply risk-informed findings and insights to decisions (and potential LB 

requests). 

As part of the second element, the licensee will evaluate the proposed LB change with regard to 

the principles of maintaining adequate defense-in-depth, maintaining sufficient safety margins, and 
Rev. 2 of RG 1.174, Page 11 

ensuring that proposed increases in CDF and risk are small and are consistent with the intent of the 

Commission’s Safety Goal Policy Statement (Ref. 4). 

2.1 Evaluation of Defense-in-Depth Attributes and Safety Margins

One aspect of the engineering evaluations is to show that the fundamental safety principles on 

which the plant design was based are not compromised by the proposed change. Design-basis accidents 

(DBAs) play a central role in nuclear power plant design. DBAs are a combination of postulated 

challenges and failure events against which plants are designed to ensure adequate and safe plant 

response. During the design process, plant response and associated safety margins are evaluated using 

assumptions that are intended to be conservative. National standards and other considerations such as 

defense-in-depth attributes and the single-failure criterion constitute additional engineering considerations 

that also influence plant design and operation. The licensee’s proposed LB change may affect margins 

and defenses associated with these considerations; therefore, the licensee should reevaluate them to 

support a requested LB change. As part of this evaluation, the impact of the proposed LB change on 

affected equipment functionality, reliability, and availability should be determined. 

2.1.1 Defense-in-Depth 

The engineering evaluation should evaluate whether the impact of the proposed LB change 

(individually and cumulatively) is consistent with the defense-in-depth philosophy. In this regard, the 

intent of this principle is to ensure that the philosophy of defense-in-depth is maintained, not to prevent 

changes in the way defense-in-depth is achieved. The defense-in-depth philosophy has traditionally been 

applied in reactor design and operation to provide multiple means to accomplish safety functions and 

prevent the release of radioactive material. It has been and continues to be an effective way to account for 

uncertainties in equipment and human performance and, in particular, to account for the potential for 

unknown and unforeseen failure mechanisms or phenomena, which (because they are unknown or 

unforeseen) are not reflected in either the PRA or traditional engineering analyses. If a comprehensive 

risk analysis is done, it can provide insights into whether the extent of defense-in-depth (e.g., balance 

among core damage prevention, containment failure, and consequence mitigation) is appropriate to ensure 

protection of public health and safety. However, to address the unknown and unforeseen failure 

mechanisms or phenomena, traditional defense-in-depth considerations should be used or maintained. 

The evaluation should consider the intent of the general design criteria in Appendix A, “General Design 

Criteria for Nuclear Power Plants,” to 10 CFR Part 50, “Domestic Licensing of Production and Utilization 

Facilities,”, national standards, and engineering principles, such as the single-failure criterion. Further, 

the evaluation should consider the impact of the proposed LB change on barriers (both preventive and 

mitigative) to core damage, containment failure or bypass, and the balance among defense-in-depth 

attributes. As stated earlier, the licensee should select the engineering analysis techniques, whether 

quantitative or qualitative, traditional or probabilistic, appropriate to the proposed LB change. 

The licensee should assess whether the proposed LB change meets the defense-in-depth principle. 

Defense-in-depth consists of a number of elements and consistency with the defense-in-depth philosophy 

is maintained if the following occurs: 

• A reasonable balance is preserved among prevention of core damage, prevention of containment 

failure, and consequence mitigation. 

• Over-reliance on programmatic activities as compensatory measures associated with the change 

in the LB is avoided. 
Rev. 2 of RG 1.174, Page 12 

• System redundancy, independence, and diversity are preserved commensurate with the expected 

frequency, consequences of challenges to the system, and uncertainties (e.g., no risk outliers). 

• Defenses against potential common-cause failures are preserved, and the potential for the 

introduction of new common-cause failure mechanisms is assessed. 

• Independence of barriers is not degraded. 

• Defenses against human errors are preserved. 

• The intent of the plant’s design criteria is maintained. 

The licensee can use these elements as guidelines for making their assessment. Other equivalent 

acceptance guidelines may also be used. 

2.1.2 Safety Margin 

The engineering evaluation should assess whether the impact of the proposed LB change is 

consistent with the principle that sufficient safety margins are maintained. Here also, the licensee is 

expected to choose the method of engineering analysis appropriate for evaluating whether sufficient 

safety margins would be maintained if the proposed LB change were to be implemented. An acceptable 

set of guidelines for making that assessment is summarized below. Other equivalent acceptance 

guidelines may also be used. With sufficient safety margins, the following are true: 

• Codes and standards or their alternatives approved for use by the NRC are met. 

• Safety analysis acceptance criteria in the LB (e.g., FSAR, supporting analyses) are met or 

proposed revisions provide sufficient margin to account for analysis and data uncertainty. 

The NRC has developed application-specific guidelines reflecting this general guidance which 

may be found in the application-specific regulatory guides (Refs. 5–9). 

2.2 Evaluation of Risk Impact, Including Treatment of Uncertainties

The licensee may use its risk assessment to address the principle that proposed increases in CDF 

and risk are small and are consistent with the intent of the NRC’s Safety Goal Policy Statement (Ref. 4). 

For purposes of implementation, the licensee should assess the expected change in CDF and LERF. The 

necessary sophistication of the evaluation, including the scope of the risk assessment (e.g., internal 

hazards only, at-power only), depends on the contribution the risk assessment makes to the integrated 

decisionmaking, which depends to some extent on the magnitude of the potential risk impact. For LB 

changes that may have a more substantial impact, an in-depth and comprehensive risk assessment, in the 

form of a PRA (i.e., one appropriate to derive a quantified estimate of the total impact of the proposed LB 

change) will be necessary to provide adequate justification.4

 

4

 Regulatory Guide 1.200 (Ref. 11) defines a PRA: “For a method or approach to be considered a PRA, the method or 

approach (1) provides a quantitative assessment of the identified risk in terms of scenarios that result in undesired 

consequences (e.g., core damage or a large early release) and their frequencies, and (2) is comprised of specific 

technical elements in performing the quantification.” Section 1.2 of Regulatory Guide 1.200 defines the technical 

elements. 
Rev. 2 of RG 1.174, Page 13 

In other applications, calculated risk-importance measures or bounding risk estimates will be 

adequate. In still others, a qualitative assessment of the impact of the LB change on the plant’s risk may 

be sufficient. 

The quantitative risk metrics adopted in this regulatory guide are CDF and LERF. There are, 

however, risk impacts that are not reflected (or are inadequately reflected) by changes to CDF and LERF. 

Therefore, the impacts of the proposed change on aspects of risk not captured (or inadequately captured) 

by these metrics should be addressed. For example, changes affecting long-term containment 

performance would impact radionuclide releases from containment occurring after evacuation and could 

result in substantial changes to offsite consequences such as latent cancer fatalities. Recognizing that the 

containment function is an important factor in maintaining the defense-in-depth philosophy (as already 

described in Section 2.1.1), the impact of the proposed change on those aspects of containment function 

not addressed in the evaluation of LERF should be addressed in the licensee submittal documentation as 

described in Section 6.3. 

The remainder of this section discusses the use of quantitative PRA results in decisionmaking. 

This discussion has three parts: 

1. A fundamental element of NRC’s risk-informed regulatory process is a PRA of sufficient scope, 

level of detail, and technical adequacy, and which sufficiently represents the as-built and asoperated plant, for the intended application. Section 2.3 of this guide discusses the staff’s 

expectations with respect to the needed PRA’s scope, level of detail, technical adequacy, and asbuilt and as-operated plant representation. 

2. PRA results are to be used in this decisionmaking process in two ways: (1) to assess the overall 

baseline CDF/LERF of the plant and (2) to assess the CDF/LERF impact of the proposed change. 

Section 2.4 of this guide discusses the acceptance guidelines the staff will use for each of these 

measures. 

3. One of the strengths of the PRA framework is its ability to characterize the impact of uncertainty 

in the analysis, and it is essential that these uncertainties be recognized when assessing whether 

the principles are being met. Section 2.5 of this guide provides guidelines on how the uncertainty 

is to be addressed in the decisionmaking process. 

The staff will base its decision on the proposed LB change on its independent judgment and 

review of the entire application. 

2.3 Technical Adequacy of Probabilistic Risk Assessment Analysis

The technical adequacy of a PRA analysis used to support an application is measured in terms of 

its appropriateness with respect to scope, level of detail, technical adequacy, and plant representation. 

The scope, level of detail, and technical adequacy of the PRA are to be commensurate with the 

application for which it is intended and the role the PRA results play in the integrated decision process. 

The more emphasis that is put on the risk insights and on PRA results in the decisionmaking process, the 

more requirements that have to be placed on the PRA in terms of both scope and how well the risk and 

the change in risk is assessed. 

Conversely, emphasis on the PRA scope, level of detail, and technical adequacy can be reduced if 

a proposed change to the LB results in a risk decrease or a change that is very small, or if the decision 

could be based mostly on traditional engineering arguments, or if compensating measures are proposed 

such that it can be convincingly argued that the change is very small. 
Rev. 2 of RG 1.174, Page 14 

Since this regulatory guide is intended for a variety of applications, the required scope, level of 

detail, and technical adequacy may vary. One overriding requirement is that the PRA should realistically 

reflect the actual design, construction, operational practices, and operational experience of the plant and 

its owner. This should include the licensee’s voluntary actions as well as regulatory requirements, and the 

PRA used to support risk-informed decisionmaking should also reflect the impact of previous changes 

made to the LB. 

2.3.1 Scope 

The scope of a PRA is defined in terms of the causes of initiating events and the plant operating 

modes it addresses. The causes of initiating events are classified into hazard groups. A hazard group is 

defined as a group of similar hazards that are assessed in a PRA using a common approach, methods, and 

likelihood data for characterizing the effect on the plant. Typical hazard groups considered in a nuclear 

power plant PRA include internal events, internal floods, seismic events, internal fires, high winds, 

external flooding, etc. Although the assessment of the risk implications in light of the acceptance 

guidelines discussed in Section 2.4 of this guide requires that all plant operating modes and hazard groups 

be addressed, it is not always necessary to have a PRA of such scope. A qualitative treatment of the 

missing modes and hazard groups may be sufficient when the licensee can demonstrate that those risk 

contributions would not affect the decision; that is, they do not alter the results of the comparison with the 

acceptance guidelines in Section 2.4 of this guide. However, when the risk associated with a particular 

hazard group or operating mode would affect the decision being made, it is the Commission’s policy that, 

if a staff-endorsed PRA standard exists for that hazard group or operating mode, then the risk will be 

assessed using a PRA that meets that standard (Ref. 13). Section 2.5 of this guide discusses this further. 

2.3.2 Level of Detail Required To Support an Application 

The level of detail required of the PRA is that which is sufficient to model the impact of the 

proposed change. The characterization of the problem should include establishing a cause-effect 

relationship to identify portions of the PRA affected by the issue being evaluated. Full-scale applications 

of the PRA should reflect this cause-effect relationship in a quantification of the impact on the PRA 

elements. For applications like component categorization, sensitivity studies on the effects of the change 

may be sufficient. For other applications, it may be adequate to define the qualitative relationship of the 

impact on the PRA elements or to only identify the impacted elements. 

If the impacts of a change to the plant cannot be associated with elements of the PRA, the PRA 

should be modified accordingly or the impact of the change should be evaluated qualitatively as part of 

the integrated decisionmaking process, as discussed in Section 2.6 of this guide. In any case, the licensee 

should properly account for the effects of the changes on the reliability and unavailability of SSCs or on 

operator actions. For additional guidance, see Regulatory Position C.1.3 in Regulatory Guide 1.200 

(Ref. 11). 

2.3.3 Probabilistic Risk Assessment Technical Adequacy 

In the current context, technical adequacy will be understood as being determined by the 

adequacy of the actual modeling and the reasonableness of the assumptions and approximations. A PRA 

used in risk-informed regulation should be performed correctly, in a manner that is consistent with 

accepted practices and commensurate with the scope and level of detail required as discussed above 

(Section 2.3.1 and 2.3.2), and appropriately represents the plant as discussed below (Section 2.3.4). 

Regulatory Guide 1.200 (Ref. 11) describes one acceptable approach for determining whether the 

technical adequacy of the PRA, in total or the parts that are used to support an application, is sufficient to 
Rev. 2 of RG 1.174, Page 15 

provide confidence in the results, such that the PRA can be used in regulatory decisionmaking for lightwater reactors. 

 Specifically, RG 1.200 provides guidance on the following: 

• technical elements and their associated attributes and characteristics of a technically acceptable 

PRA, 

• the NRC’s position on PRA consensus standards and industry PRA peer review program 

documents, 

• demonstration that the baseline PRA (in total or specific parts) used in regulatory applications is 

of sufficient technical adequacy, and 

• documentation of the technical adequacy of the PRA to support a regulatory submittal. 

Other approaches may also be acceptable, but may increase the scope of the staff review or result 

in a lower priority based on the availability of staff resources. 

Regulatory Guide 1.200 endorses the American Society of Mechanical Engineers/American 

Nuclear Society (ASME/ANS) PRA standard5

. 

The ASME/ANS PRA standard provides technical supporting requirements in terms of three 

Capability Categories. The intent of the delineation of the Capability Categories within the supporting 

requirements is generally that the degree of scope and level of detail, the degree of plant specificity, and 

the degree of realism increase from Capability Category I to Capability Category III. In general, the staff 

anticipates that current good practice, i.e., Capability Category II of the ASME/ANS PRA standard, is the 

level of detail that is adequate for the majority of applications6

. However, for some applications, 

Capability Category I may be sufficient for some requirements, whereas for other applications it may be 

necessary to achieve Capability Category III for specific requirements. 

2.3.4 Plant Representation 

The PRA results used to support an application are derived from a PRA model that represents the 

as-built and as-operated plant7

 to the extent needed to support the application. At the time of the 

application, the PRA should realistically reflect the risk associated with the plant. Regulatory Guide 

1.200 (Ref. 11), Section 1.4 provides guidance in this area. 

2.4 Acceptance Guidelines

The risk-acceptance guidelines presented in this regulatory guide are based on the principles and 

expectations for risk-informed regulation discussed in Part C of this regulatory guide and are structured as 

follows. Regions are established in the two planes generated by a measure of the baseline risk metric 

(CDF or LERF) along the x-axis, and the change in those metrics (ΔCDF or ΔLERF) along the y-axis 

(Figures 4 and 5). Acceptance guidelines are established for each region as discussed below. These 

guidelines are intended for comparison with a full-scope (including internal and external hazards, atpower, low power, and shutdown) assessment of the change in risk metric and, when necessary, as 

 

5

 These references for RG 1.200 and the ASME/ANS PRA standard are intended to refer to the revision of the 

ASME/ANS PRA standard that is endorsed in the current revision of RG 1.200. 

6

 Regulatory Guide 1.200 defines current good practice as those practices that are generally accepted throughout the 

industry and have been shown to be technically acceptable in documented analyses or engineering assessments. 

7

 As-built, as-operated is a conceptual term that reflects the degree to which the PRA matches the current plant design, 

plant procedures, and plant performance data, relative to a specific point in time. 
Rev. 2 of RG 1.174, Page 16 

discussed below, the baseline value of the risk metric (CDF or LERF). However, it is recognized that 

many PRAs are not full scope and PRA information of less than full scope may be acceptable, as 

discussed in Section 2.5 of this guide. 

Figure 4 Acceptance guidelines* for core damage frequency 

Figure 5 Acceptance guidelines* for large early release frequency 

* The analysis will be subject to increased technical review and management attention as indicated 

by the darkness of the shading of the figure. In the context of the integrated decisionmaking, the 

boundaries between regions are not definitive; the numerical values associated with defining the 

regions in the figure are to be interpreted as indicative values only. 

There are two sets of acceptance guidelines, one for CDF and one for LERF, and both sets should 

be used. 

• If the application clearly shows a decrease in CDF, the change will be considered to have 

satisfied the relevant principle of risk-informed regulation with respect to CDF. (Because 

Figure 4 is drawn on a log scale, this region is not explicitly indicated on the figure.) 
Rev. 2 of RG 1.174, Page 17 

• When the calculated increase in CDF is very small, which is taken as being less than 10-6 per 

reactor year, the change will be considered regardless of whether there is a calculation of the total 

CDF (Region III). While there is no requirement to calculate the total CDF, if there is an 

indication that the CDF may be considerably higher than 10-4 per reactor year, the focus should be 

on finding ways to decrease rather than increase it. Such an indication would result, for example, 

if (1) the contribution to CDF calculated from a limited scope analysis, such as the individual 

plant examination (IPE) or the individual plant examination of external events (IPEEE), 

significantly exceeds 10-4, (2) a potential vulnerability has been identified from a margins-type 

analysis, or (3) historical experience at the plant in question has indicated a potential safety 

concern. 

• When the calculated increase in CDF is in the range of 10-6 per reactor year to 10-5 per reactor 

year, applications will be considered only if it can be reasonably shown that the total CDF is less 

than 10-4 per reactor year (Region II). 

• Applications that result in increases to CDF above 10-5 per reactor year (Region I) would not 

normally be considered. 

AND 

• If the application clearly shows a decrease in LERF, the change will be considered to have 

satisfied the relevant principle of risk-informed regulation with respect to LERF. (Because 

Figure 5 is drawn with a log scale, this region is not explicitly indicated on the figure.) 

• When the calculated increase in LERF is very small, which is taken as being less than 10-7 per 

reactor year, the change will be considered regardless of whether there is a calculation of the total 

LERF (Region III). While there is no requirement to calculate the total LERF, if there is an 

indication that the LERF may be considerably higher than 10-5 per reactor year, the focus should 

be on finding ways to decrease rather than increase it. Such an indication would result, for 

example, if (1) the contribution to LERF calculated from a limited scope analysis, such as the IPE 

or the IPEEE, significantly exceeds 10-5, (2) a potential vulnerability has been identified from a 

margins-type analysis, or (3) historical experience at the plant in question has indicated a 

potential safety concern. 

• When the calculated increase in LERF is in the range of 10-7 per reactor year to 10-6 per reactor 

year, applications will be considered only if it can be reasonably shown that the total LERF is less 

than 10-5 per reactor year (Region II). 

• Applications that result in increases to LERF above 10-6 per reactor year (Region I) would not 

normally be considered. 

These guidelines are intended to provide assurance that proposed increases in CDF and LERF are 

small and are consistent with the intent of the Commission’s Safety Goal Policy Statement (Ref. 4). 

As indicated by the shading on the figures, the change request will be subject to an NRC technical 

and management review that will become more intensive as the calculated results move closer to the 

region boundaries. 

The guidelines discussed above are applicable for at-power, low-power, and shutdown operations. 

However, during certain shutdown operations when the containment function is not maintained, the LERF 

guideline as defined above is not practical. In those cases, licensees may use more stringent baseline 
Rev. 2 of RG 1.174, Page 18 

CDF guidelines (e.g., 10-5 per reactor year) to maintain an equivalent risk profile or may propose an 

alternative guideline to LERF that meets the intent of Principle 4 (see Figure 2). 

The technical review that relates to the risk evaluation will address the scope, level of detail, and 

technical adequacy of the analysis, including consideration of uncertainties as discussed in the next 

section. Section 2.6 of this guide discusses aspects covered by the management review, which include 

factors that are not amenable to PRA evaluation. 

2.5 Comparison of Probabilistic Risk Assessment Results with the Acceptance Guidelines

This section provides guidance on comparing the results of the PRA with the acceptance 

guidelines described in Section 2.4 of this guide. In the context of integrated decisionmaking, the 

acceptance guidelines should not be interpreted as being overly prescriptive. They are intended to 

provide an indication, in numerical terms, of what is considered acceptable. As such, the numerical 

values associated with defining the regions in Figures 4 and 5 of this regulatory guide are approximate 

values that provide an indication of the changes that are generally acceptable. Furthermore, the state-ofknowledge, or epistemic, uncertainties associated with PRA calculations preclude a definitive decision 

with respect to the region in which the application belongs based purely on the numerical results. 

The intent of comparing the PRA results with the acceptance guidelines is to demonstrate with 

reasonable assurance that Principle 4, discussed in Part C of this regulatory guide, is being met. As 

discussed in Section 2.3.1 the scope of the PRA needed to support a particular application may include 

several hazard groups or plant operating modes. When it is necessary to combine the assessment of the 

risk implications from different hazard groups, it is important to develop an understanding of the relative 

level of realism associated with the modeling of each of the hazard groups. For example, the analysis of 

specific scope items, such as internal fire, internal flooding, or seismic initiating events, typically involves 

a successive screening approach that allows the detailed analysis to focus on the more significant 

contributions. The analysis of the less significant contributions is generally of a more conservative 

nature. In addition, for each of the risk contributors, there are unique sources of model uncertainty. The 

assumptions made in response to these sources of model uncertainty and any conservatism introduced by 

the analysis approach discussed above can bias the results. This is of particular concern for the 

assessment of importance measures with respect to the combined risk assessment and the relative 

contributions of the hazard groups to the various risk metrics. 

Therefore, this comparison of the PRA results with the acceptance guidelines must be based on an 

understanding of the contributors to the PRA results and on the robustness of the assessment of those 

contributors and the impacts of the uncertainties, both those that are explicitly accounted for in the results 

and those that are not. This is a somewhat subjective process, and the reasoning behind the decisions 

must be well documented. Section 2.5.4 of this guide provides guidance on what should be addressed. 

However, the types of uncertainty that impact PRA results and methods typically used for their analysis 

are briefly discussed first. More information can be found in NUREG-1855 (Ref. 12) and in some of the 

publications in the bibliography. 

2.5.1 Types of Uncertainty and Methods of Analysis 

There are two facets to uncertainty that, because of their natures, must be treated differently when 

creating models of complex systems. They have recently been termed aleatory and epistemic uncertainty. 

The aleatory uncertainty is associated with events or phenomena being modeled that are characterized as 

occurring in a “random” or “stochastic” manner and probabilistic models are adopted to describe their 

occurrences. It is this aspect of uncertainty that gives PRA the probabilistic part of its name. The 

epistemic uncertainty is associated with the analyst’s confidence in the predictions of the PRA model 
Rev. 2 of RG 1.174, Page 19 

itself and reflects the analyst’s assessment of how well the PRA model represents the actual system being 

modeled. This has been referred to as state-of-knowledge uncertainty. This section discusses the 

epistemic uncertainty; the aleatory uncertainty is built into the structure of the PRA model itself. 

Because they are generally characterized and treated differently, it is useful to identify three 

classes of epistemic uncertainty that are addressed in and impact the results of PRAs: parameter 

uncertainty, model uncertainty, and completeness uncertainty. Completeness uncertainty can be regarded 

as one aspect of model uncertainty, but because of its importance, it is discussed separately. The 

following sections summarize the treatment of PRA uncertainty. Detailed guidance is given in NUREG1855 (Ref. 12). The bibliography may also be consulted for additional information. 

2.5.2 Parameter Uncertainty 

Each of the models that is used, either to develop the PRA logic structure or to represent the basic 

events of that structure, has one or more parameters. Typically, each of these models (e.g., the Poisson 

model for initiating events) is assumed to be appropriate. However, the parameter values for these 

models are often not known perfectly. Parameter uncertainties are those associated with the values of the 

fundamental parameters of the PRA model, such as equipment failure rates, initiating event frequencies, 

and human error probabilities that are used in the quantification of the accident sequence frequencies. 

They are typically characterized by establishing probability distributions on the parameter values. These 

distributions can be interpreted as expressing the analyst’s degree of belief in the values these parameters 

could take, based on his or her state of knowledge and conditional on the underlying model being correct. 

It is straightforward and within the capability of most PRA codes to propagate the distribution 

representing uncertainty on the basic parameter values to generate a probability distribution on the results 

(e.g., CDF, accident sequence frequencies, LERF) of the PRA. However, the analysis must be done to 

correlate the sample values for different PRA elements from a group to which the same parameter value 

applies (the so-called state-of-knowledge dependency; see Reference 14). 

2.5.3 Model Uncertainty 

The development of the PRA model is supported by the use of models for specific events or 

phenomena. In many cases, the industry’s state of knowledge is incomplete, and there may be different 

opinions on how the models should be formulated. Examples include approaches to modeling human 

performance, common-cause failures, and reactor coolant pump seal behavior upon loss of seal cooling. 

This gives rise to model uncertainty. 

In many cases, the appropriateness of the models adopted is not questioned and these models 

have become, de facto, the consensus models8

 to use. Examples include the use of Poisson and binomial 

models to characterize the probability of occurrence of component failures. For some issues with wellformulated alternative models, PRAs have addressed model uncertainty by using discrete distributions 

over the alternative models, with the probability associated with a specific model representing the 

analyst’s degree of belief that the model is the most appropriate. A good example is the characterization 

of the seismic hazard as different hypotheses lead to different hazard curves, which can be used to 

develop a discrete probability distribution of the initiating event frequency for earthquakes. Other 

examples can be found in the Level 2 analysis. 

 

8

 NUREG-1855 (Ref. 12) defines a consensus model as one that has a publicly available published basis and has been 

peer reviewed and widely adopted by an appropriate stakeholder group. In addition, widely accepted PRA practices 

may be regarded as consensus models. Examples of the latter include the use of the constant probability of failure on 

demand model for standby components and the Poisson model for initiating events. For risk-informed regulatory 

decisions, the consensus model approach is one that NRC has utilized or accepted for the specific risk-informed 

application for which it is proposed. 
Rev. 2 of RG 1.174, Page 20 

Another approach to addressing model uncertainty has been to adjust the results of a single model 

through the use of an adjustment factor. However it is formulated, an explicit representation of model 

uncertainty can be propagated through the analysis in the same way as parameter uncertainty. More 

typically, however, particularly in the Level 1 analysis, the use of different models would result in the 

need for a different structure (e.g., with different thermal hydraulic models used to determine success 

criteria). In such cases, uncertainties in the choice of an appropriate model are typically addressed by 

making assumptions and, as in the case of the component failure models discussed above, adopting a 

specific model. 

PRAs model the continuum of possible plant states in a discrete way, and are, by their very 

nature, approximate models of the world. This results in some random (aleatory) aspects of the “world” 

not being addressed except in a bounding way (e.g., different realizations of an accident sequence 

corresponding to different loss-of-coolant accident (LOCA) sizes, within a category, are treated by 

assuming a bounding LOCA), with the time of failure of an operating component assumed to occur at the 

moment of demand. These approximations introduce biases (uncertainties) into the results. 

In interpreting the results of a PRA, it is important to develop an understanding of the impact of a 

specific assumption or choice of model on the predictions of the PRA. This is true even when the model 

uncertainty is treated probabilistically, since the probabilities, or weights, given to different models would 

be subjective. The impact of using alternative assumptions or models may be addressed by performing 

appropriate sensitivity studies or by using qualitative arguments, based on an understanding of the 

contributors to the results and how they are impacted by the change in assumptions or models. The 

impact of making specific modeling approximations may be explored in a similar manner. 

2.5.4 Completeness Uncertainty 

Completeness is not in itself an uncertainty, but a reflection of scope limitations. The result is, 

however, an uncertainty about where the true risk lies. The problem with completeness uncertainty is 

that, because it reflects an unanalyzed contribution, it is difficult (if not impossible) to estimate its 

magnitude. Some contributions are unanalyzed not because methods are unavailable, but because they 

have not been refined to the level of the analysis of internal hazards. 

Examples are the analysis of some external hazards and the low-power and shutdown modes of 

operation. There are issues, however, for which methods of analysis have not been developed, and they 

have to be accepted as potential limitations of the technology. Thus, for example, the impact on actual 

plant risk from unanalyzed issues such as the influences of organizational performance cannot now be 

explicitly assessed. 

The issue of completeness of scope of a PRA can be addressed for those scope items for which 

methods are in principle available, and therefore some understanding of the contribution to risk exists, by 

supplementing the analysis with additional analysis to enlarge the scope, using more restrictive 

acceptance guidelines, or by providing arguments that, for the application of concern, the out-of-scope 

contributors are not significant. The next section includes approaches acceptable to the NRC staff for 

dealing with incompleteness. 

2.5.5 Comparisons with Acceptance Guidelines 

The different regions of the acceptance guidelines require different depths of analysis. Changes 

resulting in a net decrease in the CDF and LERF estimates do not require an assessment of the calculated 

baseline CDF and LERF. Generally, it should be possible to argue on the basis of an understanding of the 
Rev. 2 of RG 1.174, Page 21 

contributors and the changes that are being made that the overall impact is indeed a decrease, without the 

need for a detailed quantitative analysis. 

If the calculated values of ΔCDF and ΔLERF are very small, as defined by Region III in 

Figures 4 and 5, a detailed quantitative assessment of the baseline value of CDF and LERF will not be 

necessary. However, if there is an indication that the CDF or LERF could considerably exceed 10-4 and 

10-5, respectively, in order for the change to be considered the licensee may be required to present 

arguments as to why steps should not be taken to reduce CDF or LERF. Such an indication would result, 

for example, if (1) the contribution to CDF or LERF calculated from a limited scope analysis, such as the 

IPE or the IPEEE, significantly exceeds 10-4 and 10-5, respectively, (2) there has been an identification of 

a potential vulnerability from a margins-type analysis, or (3) historical experience at the plant in question 

has indicated a potential safety concern. 

For larger values of ΔCDF and ΔLERF, which lie in the range used to define Region II, an 

assessment of the baseline CDF and LERF is required. 

To demonstrate compliance with the numerical guidelines, the level of detail required in the 

assessment of the values and the analysis of uncertainty related to model and incompleteness issues will 

depend on both (1) the LB change being considered and (2) the importance of the demonstration that 

Principle 4 has been met. In Region III of Figures 4 and 5, the closer the estimates of ΔCDF or ΔLERF 

are to their corresponding acceptance guidelines, the more detail will be required. Similarly, in Region II 

of Figures 4 and 5, the closer the estimates of ΔCDF or ΔLERF and CDF and LERF are to their 

corresponding acceptance guidelines, the more detail will be required. In a contrasting example, if the 

estimated value of a particular metric is very small compared to the acceptance goal, a simple bounding 

analysis may suffice with no need for a detailed uncertainty analysis. 

Because of the way the acceptance guidelines (Section 2.4) were developed, the appropriate 

numerical measures to use in the initial comparison of the PRA results to the acceptance guidelines are 

mean values. The mean values referred to are the means of the probability distributions that result from 

the propagation of the uncertainties on the input parameters and those model uncertainties explicitly 

represented in the model. While a formal propagation of the uncertainty is the best way to correctly 

account for state-of-knowledge uncertainties that arise from the use of the same parameter values for 

several basic event probability models, under certain circumstances, a formal propagation of uncertainty 

may not be required if it can be demonstrated that the state-of-knowledge correlation is unimportant. This 

will involve, for example, a demonstration that the bulk of the contributing scenarios (cutsets or accident 

sequences) do not involve multiple events that rely on the same parameter for their quantification. For 

more detail, see Section 4 of NUREG-1855 (Ref. 12). 

Consistent with the viewpoint that the guidelines are not to be used prescriptively, even if the 

calculated ΔCDF and ΔLERF values are such that they place the change in Region I or II, it may be 

possible to make a case that the application should be treated as if it were in Region II or III if, for 

example, it is shown that there are unquantified benefits that are not reflected in the quantitative risk 

results. However, care should be taken that there are no unquantified detrimental impacts of the change, 

such as an increase in operator burden. In addition, if compensatory measures are proposed to counter the 

impact of the major risk contributors, even though the impact of these measures may not be estimated 

numerically, such arguments will be considered in the decision process. 

While the analysis of parametric uncertainty is fairly mature and is addressed adequately through 

the use of mean values, the analysis of the model and completeness uncertainties cannot be handled in 

such a formal manner. Whether the PRA is full scope or only partial scope, and whether it is only the 

change in metrics or both the change and baseline values that need to be estimated, it will be incumbent 
Rev. 2 of RG 1.174, Page 22 

on the licensee to demonstrate that the choice of reasonable9

 alternative hypotheses, adjustment factors, or 

modeling approximations or methods to those adopted in the PRA model would not significantly change 

the assessment. This demonstration can take the form of well-formulated sensitivity studies or qualitative 

arguments. It is not the intent that the search for alternatives should be exhaustive or arbitrary. For the 

decisions that involve only assessing the change in metrics, the number of model uncertainty issues to be 

addressed will be smaller than for the case of the baseline values, when only a portion of the model is 

affected. The alternatives that would drive the result toward unacceptability (i.e., those associated with 

key sources of model uncertainty10) should be identified and sensitivity studies performed or reasons 

given as to why they are not appropriate for the current application or for the particular plant. In general, 

the results of the sensitivity studies should confirm that the guidelines are still met even under the 

alternative assumptions (i.e., change generally remains in the appropriate region). Alternatively, this 

analysis can be used to identify candidates for compensatory actions or increased monitoring. (Section 7 

of NUREG-1855 (Ref. 12) provides additional guidance on treating PRA uncertainty in the 

decisionmaking process.) The licensee should pay particular attention to those assumptions that impact 

the parts of the model being exercised by the change. 

When the PRA is not full scope, it is necessary for the licensee to address the significance of the 

out-of-scope items. The importance of assessing the contribution of the out-of-scope portions of the PRA 

to the base case estimates of CDF and LERF is related to the margin between the as-calculated values and 

the acceptance guidelines. When the contributions from the modeled contributors are close to the 

guidelines, the argument that the contribution from the missing items is not significant must be 

convincing and in some cases may require additional PRA analyses. When the margin is significant, a 

qualitative argument may be sufficient. The contribution of the out-of-scope portions of the model to the 

change in metric may be addressed by bounding analyses, detailed analyses, or by a demonstration that 

the change has no impact on the unmodeled contributors to risk. In addition, it should be demonstrated 

that changes based on a partial PRA do not disproportionately change the risk associated with those 

accident sequences that arise from the modes of operation not included in the PRA. 

One alternative to an analysis of uncertainty is to design the proposed LB change such that the 

major sources of uncertainty will not have an impact on the decisionmaking process. For example, in the 

region of the acceptance guidelines where small increases are allowed regardless of the value of the 

baseline CDF or LERF, the proposed change to the LB could be designed such that the modes of 

operation or the initiating events that are missing from the analysis would not be affected by the change. 

In these cases, incompleteness would not be an issue. Similarly, in such cases, it would not be necessary 

to address all the model uncertainties, but only those that impact the evaluation of the change. 

If only a Level 1 PRA is available, in general, only the CDF is calculated and not the LERF. 

Reference 10 presents an approach that allows a subset of the core damage accidents identified in the 

Level 1 analysis to be allocated to a release category that is equivalent to a LERF. The approach uses 

simplified event trees that can be quantified by the licensee on the basis of the plant configuration 

applicable to each accident sequence in the Level 1 analysis. The frequency derived from these event 

trees can be compared to the LERF acceptance guidelines. The approach described in Reference 10 may 

be used to estimate LERF only in those cases when the plant is not close to the CDF and LERF 

benchmark values. 

 

9

 In the ASME/ANS PRA standard 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 made. 

10 In the ASME/ANS PRA standard a source of model uncertainty is labeled “key” when it could impact the PRA results 

that are being used in a decision, and consequently, may influence the decision being made. 
Rev. 2 of RG 1.174, Page 23 

2.6 Integrated Decisionmaking

In making a regulatory decision, risk insights are integrated with considerations of defense-indepth and safety margins. The degree to which the risk insights play a role, and therefore the need for 

detailed staff review, is application dependent. 

Quantitative risk results from PRA calculations are typically the most useful and complete 

characterization of risk, but they should be supplemented by qualitative risk insights and traditional 

engineering analysis where appropriate. Qualitative risk insights include generic results that have been 

learned from previous PRAs and from operational experience. For example, if one is deciding which 

motor-operated valves in a plant can be subject to less frequent testing, the plant-specific PRA results can 

be compared with results from similar plants. This type of comparison can support the licensee’s analysis 

and reduce the reliance of the staff review on the technical adequacy of the licensee PRA. However, as a 

general rule, applications that impact large numbers of SSCs will benefit from a PRA of sound technical 

adequacy. 

Traditional engineering analysis provides insight into available margins and defense-in-depth. 

With few exceptions, these assessments are performed without any quantification of risk. 

The results of the different elements of the engineering analyses discussed in Sections 2.1 and 2.2 

of this guide must be considered in an integrated manner. None of the individual analyses is sufficient in 

and of itself. In this way, it can be seen that the decision will not be driven solely by the numerical results 

of the PRA. These results are one decisionmaking input and help in building an overall picture of the 

implications of the proposed change on risk. The PRA has an important role in putting the change into its 

proper context as it impacts the plant as a whole. The PRA analysis is used to demonstrate that 

Principle 4 has been satisfied. As the discussion in the previous section indicates, both quantitative and 

qualitative arguments may be brought to bear. Even though the different pieces of evidence used to argue 

that the principle is satisfied may not be combined in a formal way, they need to be clearly documented. 

The acceptability of the change supported by the risk-informed decision will be a function of the 

confidence the NRC staff has in the results of the analysis. As indicated, one important factor that can be 

considered when determining the degree of implementation of the change is the ability to monitor the 

performance to limit the potential risk. In many applications, the potential risk can be limited by defining 

specific measures and criteria that must be monitored subsequent to approval. When relying on 

performance monitoring, the staff must have assurance that the measures truly represent the potential for 

risk increase and that the criteria are set at reasonable limits. Moreover, one must be sure that degrading 

performance can be detected in a timely fashion, long before a significant public health issue results. The 

impact of the monitoring can be fed back into the analysis to demonstrate how it supports the decision. 

The NRC review of an application will consider all these factors. The review of PRA technical 

adequacy in particular will focus on those aspects that impact the results used in the decision and on the 

degree of confidence required in those results. A limited application would lead the staff to conduct a 

more limited review of the risk estimates, therefore placing less emphasis on the technical adequacy of 

the PRA than would be the case for a broad-scope application. 

Finally, when implementing a decision, the licensee may choose to compensate for a lack of 

confidence in the analysis by restricting the degree of implementation. This has been the technique used 

in several applications involving SSC categorization into low or high safety significance. In general, 

unless there is compelling evidence that the SSC is of low safety significance, it is maintained as high 

safety significance. This requires a reasonable understanding of the limitations of the PRA. Another 

example of risk limitation is the placing of restrictions on the application. For example, risk-informed 
Rev. 2 of RG 1.174, Page 24 

technical-specification-completion time changes are accompanied by implementation of a configuration 

risk management program, which requires licensees to examine their plant configuration before 

voluntarily entering the approved condition. 

Section 2.4 of this guide indicates that the application would be given increased NRC 

management attention when the calculated values of the changes in the risk metrics, and their baseline 

values, when appropriate, approach the guidelines. Therefore, if the risk metrics approach the guidelines, 

the licensee’s submittal should address the following issues: 

• the cumulative impact of previous changes and the trend in CDF (the licensee’s risk management 

approach), 

• the cumulative impact of previous changes and the trend in LERF (the licensee’s risk 

management approach), 

• the impact of the proposed change on operational complexity, burden on the operating staff, and 

overall safety practices, 

• plant-specific performance and other factors (for example, siting factors, inspection findings, 

performance indicators, and operational events), and Level 3 PRA information, if available, 

• the benefit of the change in relation to its CDF/LERF increase, 

• the practicality of accomplishing the change with a smaller CDF/LERF impact, and 

• the practicality of reducing CDF/LERF when there is reason to believe that the baseline 

CDF/LERF are above the guideline values (i.e., 10-4 and 10-5 per reactor year, respectively). 

3. Element 3: Define Implementation and Monitoring Program 

Careful consideration should be given to implementation of the proposed change and the 

associated performance-monitoring strategies. The primary goal of Element 3 is to ensure that no 

unexpected adverse safety degradation occurs due to the change(s) to the LB. The staff’s principal 

concern is the possibility that the aggregate impact of changes that affect a large class of SSCs could lead 

to an unacceptable increase in the number of failures from unanticipated degradation, including possible 

increases in common cause mechanisms. Therefore, an implementation and monitoring plan should be 

developed to ensure that the engineering evaluation conducted to examine the impact of the proposed 

changes continues to reflect the actual reliability and availability of SSCs that have been evaluated. This 

will ensure that the conclusions that have been drawn from the evaluation remain valid. Applicationspecific regulatory guides (Refs. 5, 7-9) discuss additional details of acceptable processes for 

implementation in specific applications. 

Decisions concerning the implementation of LB changes should be made after considering the 

uncertainty associated with the results of the traditional and probabilistic engineering evaluations. Broad 

implementation within a limited time period may be justified when uncertainty is shown to be low (e.g., 

data and models are adequate, engineering evaluations are verified and validated). A slower, phased 

approach to implementation (or other modes of partial implementation) would be expected when 

uncertainty in evaluation findings is higher and when programmatic changes are being made that could 

impact SSCs across a wide spectrum of the plant, such as in inservice testing, inservice inspection, and 

graded quality assurance (IST, ISI, and graded special treatment). In such situations, the potential 

introduction of common cause effects must be fully considered and included in the submittal. 

The licensee should propose monitoring program(s) that include a means to adequately track the 

performance of equipment that, when degraded, can affect the conclusions of the licensee’s engineering 

evaluation and integrated decisionmaking that support the change to the LB. The program should be 

capable of trending equipment performance after a change has been implemented to demonstrate that 

performance is consistent with the assumptions in the traditional engineering and probabilistic analyses 
Rev. 2 of RG 1.174, Page 25 

conducted to justify the change. This may include monitoring associated with nonsafety-related SSCs if 

the analysis determines that those SSCs are risk significant. The program should be structured such that 

(1) SSCs are monitored commensurate with their safety importance (i.e., monitoring for SSCs categorized 

as having low safety significance may be less rigorous than that for SSCs of high safety significance), (2) 

feedback of information and corrective actions is accomplished in a timely manner, and (3) degradation in 

SSC performance is detected and corrected before plant safety can be compromised. The potential impact 

of observed SSC degradation on similar components in different systems throughout the plant should be 

considered. 

Licensees will integrate, or at least coordinate, their monitoring for risk-informed changes with 

existing programs that monitor equipment performance and other operating experience on their site and 

industry-wide. In particular, monitoring that is performed in conformance with 10 CFR 50.65, 

“Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants” (the 

Maintenance Rule) can be used when the monitoring performed under the Maintenance Rule is sufficient 

for the SSCs affected by the risk-informed application. If an application requires monitoring of SSCs that 

the Maintenance Rule does not include, or has a greater resolution of monitoring than the Maintenance 

Rule (component versus train- or plant-level monitoring), it may be advantageous for a licensee to adjust 

the Maintenance Rule monitoring program rather than to develop additional monitoring programs for 

risk-informed purposes. In these cases, the performance criteria chosen should be shown to be 

appropriate for the application. It should be noted that plant or licensee performance under actual design 

conditions may not be readily measurable. When actual conditions cannot be monitored or measured, 

whatever information most closely approximates actual performance data should be used. For example, 

establishing a monitoring program with a performance-based feedback approach may combine some of 

the following activities: 

• monitoring performance characteristics under actual design-basis conditions (e.g., reviewing 

actual demands on emergency diesel generators, reviewing operating experience), 

• monitoring performance characteristics under test conditions that are similar to those expected 

during a design-basis event, 

• monitoring and trending performance characteristics to verify aspects of the underlying analyses, 

research, or bases for a requirement (e.g., measuring battery voltage and specific gravity, 

inservice inspection of piping), 

• evaluating licensee performance during training scenarios (e.g., emergency planning exercises, 

operator licensing examinations), and 

• component quality controls, including developing pre- and post-component installation 

evaluations (e.g., environmental qualification inspections, reactor protection system channel 

checks, continuity testing of boiling water reactor squib valves). 

As part of the monitoring program, it is important that provisions for specific cause 

determination, trending of degradation and failures, and corrective actions be included. Such provisions 

should be applied to SSCs commensurate with their importance to safety as determined by the 

engineering evaluation used to support the LB change. A determination of cause is needed when 

performance expectations are not being met or when there is a functional failure of an application-specific 

SSC that poses a significant condition adverse to performance. The cause determination should identify 

the cause of the failure or degraded performance to the extent that corrective action can be identified that 

would preclude the problem or ensure that it is anticipated before becoming a safety concern. It should 

address failure significance, the circumstances surrounding the failure or degraded performance, the 

characteristics of the failure, and whether the failure is isolated or has generic or common-cause 

implications (as defined in Reference 15). 
Rev. 2 of RG 1.174, Page 26 

Finally, in accordance with Criterion XVI of Appendix B, “Quality Assurance Criteria for 

Nuclear Power Plants and Fuel Reprocessing Plants,” to 10 CFR Part 50, the monitoring program should 

identify any corrective actions to preclude the recurrence of unacceptable failures and/or degraded 

performance. The circumstances surrounding the failure may indicate that the SSC failed because of 

adverse or harsh operating conditions (e.g., operating a valve dry, over-pressurization of a system) or 

failure of another component that caused the SSC failure. Therefore, corrective actions should also 

consider SSCs with similar characteristics with regard to operating, design, or maintenance conditions. 

The results of the monitoring need not be reported to the NRC, but should be retained onsite for 

inspection. 

4. Element 4: Submit Proposed Change 

Requests for proposed changes to the plant’s LB typically take the form of requests for license 

amendments (including changes to or removal of license conditions), technical specification changes, 

changes to or withdrawals of orders, and changes to programs under 10 CFR 50.54, “Conditions of 

Licenses” (e.g., quality assurance program changes under 10 CFR 50.54(a)). Licensees should 

(1) carefully review the proposed LB change to determine the appropriate form of the change request, 

(2) ensure that information required by the relevant regulations in support of the request is developed, and 

(3) prepare and submit the request in accordance with relevant procedural requirements. For example, 

license amendments should meet the requirements of 10 CFR 50.90, 10 CFR 50.91, “Notice for Public 

Comment; State Consultation,” and 10 CFR 50.92, as well as the procedural requirements in 

10 CFR 50.4, “Written Communications.” Risk information that the licensee submits in support of the 

LB change request should meet the guidance in Section 6 of this regulatory guide. 

Licensees may submit risk information in support of their LB change request. If the licensee’s 

proposed change to the LB is consistent with currently approved staff positions, the staff’s determination 

generally will be based solely on traditional engineering analyses without recourse to risk information 

(although the staff may consider any risk information submitted by the licensee). If the licensee’s 

proposed change goes beyond currently approved staff positions, the staff normally will consider 

information based on both traditional engineering analyses and risk insights. If the licensee does not 

submit risk information in support of an LB change that goes beyond currently approved staff positions, 

the staff may request the licensee to submit such information. If the licensee chooses not to provide the 

risk information, the staff will review the proposed application using traditional engineering analyses and 

determine whether sufficient information has been provided to support the requested change. However, if 

new information reveals an unforeseen hazard or a substantially greater potential for a known hazard to 

occur, such as the identification of an issue related to the requested LB change that may substantially 

increase risk (see Reference 3), the NRC staff will request that the licensee submit risk-related 

information. The NRC staff will not approve the requested LB change until it has reasonable assurance 

that the public health and safety will be adequately protected if the requested LB change is approved. 

In developing the risk information set forth in this regulatory guide, licensees are likely to 

identify SSCs with high risk significance that are not currently subject to regulatory requirements or are 

subject to a level of regulation that is not commensurate with their risk significance. It is expected that 

licensees will propose LB changes that will subject these SSCs to an appropriate level of regulatory 

oversight, consistent with the risk significance of each SSC. Application-specific regulatory guides (Refs. 

5, 7-9) present specific information on the staff’s expectations in this regard. 

5. Quality Assurance 

As stated in Section 2 of this guide, the quality of the engineering analyses conducted should 

justify proposed LB changes will be appropriate for the nature of the change. In this regard, it is expected 
Rev. 2 of RG 1.174, Page 27 

that for traditional engineering analyses (e.g., deterministic engineering calculations), existing provisions 

for quality assurance (e.g., Appendix B to 10 CFR Part 50, for safety-related SSCs) will apply and 

provide the appropriate quality needed. Likewise, when a risk assessment of the plant is used to provide 

insights into the decisionmaking process, the PRA is to have been subject to quality control. 

To the extent that a licensee elects to use PRA information to enhance or modify activities 

affecting the safety-related functions of SSCs, the following (in conjunction with the other guidance 

contained in this guide), describes methods acceptable to the NRC staff to ensure that the pertinent quality 

assurance requirements of Appendix B to 10 CFR Part 50 are met and that the PRA is sufficient to be 

used for regulatory decisions: 

• Use personnel qualified for the analysis. 

• Use procedures that ensure control of documentation, including revisions, and provide for 

independent review, verification, or checking of calculations and information used in the 

analyses. (An independent peer review or certification program can be used as an important 

element in this process.) 

• Provide documentation and maintain records in accordance with the guidelines in Section 6 of 

this guide. 

• Use procedures that ensure that appropriate attention and corrective actions are taken if 

assumptions, analyses, or information used in previous decisionmaking are changed (e.g., 

licensee voluntary action) or determined to be in error. 

When performance monitoring programs are used in the implementation of proposed changes to 

the LB, it is expected that those programs will be implemented by using quality assurance provisions 

commensurate with the safety significance of affected SSCs. An existing PRA or analysis can be utilized 

to support a proposed LB change, provided it can be shown that the appropriate quality provisions are 

met. 

6. Documentation 

6.1 Introduction

To facilitate the NRC staff’s review to ensure that the analyses conducted were sufficient to 

conclude that the key principles of risk-informed regulation have been met, documentation of the 

evaluation process and findings are to be maintained. Additionally, the information submitted should 

include a description of the process used by the licensee to ensure its adequacy and some specific 

information to support the staff’s conclusion regarding the acceptability of the requested LB change. 

6.2 Archival Documentation

Archival documentation should include a detailed description of engineering analyses conducted 

and the results obtained, irrespective of whether they were quantitative or qualitative, or whether the 

analyses made use of traditional engineering methods or probabilistic approaches. This documentation 

should be maintained by the licensee, as part of its quality assurance program, so that it is available for 

examination. Documentation of the analyses conducted to support changes to a plant’s LB should be 

maintained as lifetime quality records in accordance with Regulatory Guide 1.33, Revision 2, “Quality 

Assurance Program Requirements (Operation),” issued February 1978 (Ref. 16). 
Rev. 2 of RG 1.174, Page 28 

6.3 Licensee Submittal Documentation

To support the NRC staff’s conclusion that the proposed LB change is consistent with the key 

principles of risk-informed regulation and NRC staff expectations, the licensee should submit the 

following information: 

• A description of how the proposed change will impact the LB (relevant principle: LB changes 

meet regulations). 

• A description of the components and systems affected by the change, the types of changes 

proposed, the reason for the changes, and results and insights from an analysis of available data 

on equipment performance. (Relevant staff expectation: All safety impacts of the proposed LB 

change must be evaluated). 

• A reevaluation of the LB accident analysis and the provisions of 10 CFR Part 20, “Standards for 

Protection against Radiation,” and 10 CFR Part 100, “Reactor Site Criteria,” if appropriate. 

(Relevant principles: LB changes meet the regulations, sufficient safety margins are maintained, 

and defense-in-depth philosophy is used). 

• An evaluation of the impact of the LB change on the breadth or depth of defense-in-depth 

attributes of the plant. (Relevant principle: Defense-in-depth philosophy is used). 

• Identification of how and where the proposed change will be documented as part of the plant’s 

LB (e.g., FSAR, technical specifications, licensing conditions). This should include proposed 

changes or enhancements to the regulatory controls for high-risk-significant SSCs that are not 

subject to any requirements or the requirements are not commensurate with the SSC’s risk 

significance. 

The licensee should also identify: 

• key assumptions11 in the PRA that impact the application (e.g., voluntary licensee actions), 

elements of the monitoring program, and commitments made to support the application, 

• SSCs for which requirements should be increased, 

• information to be provided as part of the plant’s LB (e.g., FSAR, technical specifications, 

licensing condition), and 

• whether provisions of Appendix B to 10 CFR Part 50 apply to the PRA. 

The latter item comes into play if the PRA forms part of the basis used to enhance or modify 

safety-related functions of SSCs subject to those provisions. Thus, the licensee would be expected to 

control PRA activity in a manner commensurate with its impact on the facility’s design and licensing 

basis and in accordance with all applicable regulations and its quality assurance program description. 

An independent peer review is an important element of ensuring technical adequacy. The 

licensee’s submittal should discuss measures used to ensure technical adequacy, such as a report of a peer 

review augmented by a discussion of the appropriateness of the PRA model for supporting a risk 

assessment of the LB change under consideration. The submittal should address any analysis limitations 

that are expected to impact the conclusion regarding acceptability of the proposed change. 

 

11 In the ASME/ANS PRA standard an assumption is labeled “key” when it may influence (i.e., have the potential to 

change) the decision being made. 
Rev. 2 of RG 1.174, Page 29 

The licensee’s resolution of the findings of the peer review should also be submitted. For 

example, this response could indicate whether the PRA was modified following the peer review or could 

justify why no change was necessary to support decisionmaking for the LB change under consideration. 

As discussed in Section 2.2 of this guide, the staff’s decision on the proposed license amendment will be 

based on its independent judgment and review. 

6.3.1 Risk Assessment Methods 

To have confidence that the risk assessment is adequate to support the proposed change, a 

summary of the risk assessment methods used should be submitted. Consistent with current practice, 

information submitted to the NRC for its consideration in making risk-informed regulatory decisions will 

be made publicly available, unless such information is properly identified as proprietary in accordance 

with the regulations. The following information should be submitted and is intended to illustrate that the 

scope, level of detail, and technical acceptability of the engineering analyses conducted to justify the 

proposed LB change are appropriate to the nature and scope of the change: 

• A description of risk assessment methods used. 

• Documentation of the adequacy of the scope of the PRA. 

• A description of the licensee’s process to ensure PRA technical adequacy and a discussion as to 

why the PRA is of sufficient technical adequacy to support the current application. 

• The key modeling assumptions12 that are necessary to support the analysis or that impact the 

application. 

• The event trees and fault trees that require modification to support analyses of the proposed 

change with a description of their modification. 

• A list of operator actions modeled in the PRA that impact the application and their error 

probabilities. 

The submitted information that summarizes the results of the risk assessment should include the 

following: 

• The effects of the proposed change on the more significant sequences (e.g., sequences that 

contribute more than 5 percent to the risk) to show that the LB change does not create risk 

outliers and does not exacerbate existing risk outliers. 

• An assessment of the change to CDF and LERF, including a description of the significant 

contributors to the change. 

• Information related to the assessment of the full-scope, baseline CDF; the extent of the 

information required will depend on whether the analysis of the change in CDF is in Region II or 

Region III of Figure 4. 

• Information related to the assessment of the full-scope, baseline LERF; the extent of the 

information required will depend on whether the analysis of the change in LERF is in Region II 

or Region III of Figure 5. 

 

12 In the ASME/ANS PRA standard, a modeling assumption is one that is related to a model uncertainty and is made with 

the knowledge that a different reasonable alternative assumption exists. 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 made. An assumption is labeled “key” when it may influence (i.e., have the 

potential to change) the decision being made. 
Rev. 2 of RG 1.174, Page 30 

• Results of sensitivity analyses that show that the conclusions regarding the impact of the LB 

change on plant risk will not vary significantly under a different set of plausible assumptions. 

6.3.2 Cumulative Risks 

As part of evaluation of risk, licensees should understand the effects of the current application in 

light of past applications. Optimally, the PRA used for the current application should already model the 

effects of past applications. However, qualitative effects and synergistic effects are sometimes difficult to 

model. Tracking changes in risk (both quantifiable and nonquantifiable) that are due to plant changes 

would provide a mechanism to account for the cumulative and synergistic effects of these plant changes 

and would help to demonstrate that the proposing licensee has a risk management philosophy in which 

PRA is not just used to systematically increase risk, but is also used to help reduce risk where appropriate 

and where it is shown to be cost effective. The tracking of cumulative risk will also help the NRC staff in 

monitoring trends. 

As part of the submittal, the licensee should track and submit the impact of all plant changes that 

have been submitted for NRC review and approval which have not yet been incorporated into the baseline 

PRA model, and are therefore not reflected in the baseline risk. Documentation should include the 

following: 

• the calculated change in risk for each application (CDF and LERF) and the plant elements (e.g., 

SSCs, procedures) affected by each change, 

• qualitative arguments used to justify the change (if any) and the plant elements affected by these 

arguments, 

• compensatory measures or other commitments used to help justify the change (if any) and the 

plant elements affected, and 

• summarized results from the monitoring programs (where applicable) and a discussion of how 

these results have been factored into the PRA or into the current application. 

As an option, the submittal could also list (but not submit to the NRC) past changes to the plant 

that reduced the plant risk, especially those changes that are related to the current application. A 

discussion of whether these changes are already included in the base PRA model should also be included. 

D. IMPLEMENTATION 

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

NRC’s plans for using this regulatory guide. The NRC does not intend or approve any imposition or 

backfit in connection with its issuance. 

In some cases, applicants or licensees may propose or use a previously established acceptable 

alternative method for complying with specified portions of the NRC’s regulations. Otherwise, 

the methods described in this guide will be used in evaluating compliance with the applicable regulations 

for license applications, license amendment applications, and amendment requests. 
Rev. 2 of RG 1.174, Page 31 

REFERENCES13

1. 60 FR 42622, “Use of Probabilistic Risk Assessment Methods in Nuclear Activities: Final Policy 

Statement,” Federal Register, Volume 60, Number 158, p. 42622, Washington, DC, August 16, 

1995. 

2. SECY-10-0143, “Semiannual Update of the Risk-Informed and Performance-Based Plan,” U.S. 

Nuclear Regulatory Commission, October 28, 2010 (ML102800313). 

3. NUREG-0800, Section 19.2, “Review of Risk Information Used to Support Permanent PlantSpecific Changes to the Licensing Basis: General Guidance,” June 2007. 

4. 51 FR 30028, “Safety Goals for the Operations of Nuclear Power Plants; Policy Statement,” 

Federal Register, Volume 51, p. 30028, Washington, DC, August 4, 1986. 

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

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

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

Graded Quality Assurance,” U.S. Nuclear Regulatory Commission, Washington DC (withdrawn). 

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

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

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

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

9. Regulatory Guide 1.201 (Trial Use), Revision 1, “Guidelines for Categorizing Structures, 

Systems, and Components in Nuclear Power Plants According to Their Safety Significance,” U.S. 

Nuclear Regulatory Commission, Washington, DC. 

10. W.T. Pratt et al., “An Approach for Estimating the Frequencies of Various Containment Failure 

Modes and Bypass Events,” NUREG/CR-6595, Revision 1, October 2004. 

11. Regulatory Guide 1.200, “An Approach for Determining the Technical Adequacy of Probabilistic 

Risk Assessment Results for Risk-Informed Activities,” U.S. Nuclear Regulatory Commission, 

Washington, DC. 

12. NUREG-1855, “Guidance on the Treatment of Uncertainties Associated with PRAs in RiskInformed Decision Making,” Volume 1, March 2009. 

13. Staff Requirements Memorandum to SECY-04-0118, “Plan for the Implementation of the 

Commission’s Phased Approach to Probabilistic Risk Assessment Quality,” U.S. Nuclear 

Regulatory Commission, Washington, DC, October 6, 2004. 

 

13 Publicly available NRC published documents such as Regulations, Regulatory Guides, NUREGs, and Generic Letters 

listed herein are available electronically through the Electronic Reading room on the NRC’s public Web site at: 

http://www.nrc.gov/reading-rm/doc-collections/. Copies are also available for inspection or copying for a fee from the 

NRC’s Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD; the mailing address is USNRC PDR, 

Washington, DC 20555; telephone 301-415-4737 or (800) 397-4209; fax (301) 415-3548; and e-mail 

PDR.Resource@nrc.gov. 
Rev. 2 of RG 1.174, Page 32 

14. G. Apostolakis and S. Kaplan, “Pitfalls in Risk Calculations,” Reliability Engineering, Vol. 2, 

pp. 135–145, 1981.14 

15. A. Mosleh et al., “Guidance on Modeling Common-Cause Failures in Probabilistic Risk 

Assessment,” NUREG/CR-5485, November 1998. 

16. Regulatory Guide 1.33, “Quality Assurance Program Requirements (Operation),” U.S. Nuclear 

Regulatory Commission, Washington, DC. 

 

14 Copies of the non-NRC documents included in these references may be obtained directly from the publishing 

organization. 
Rev. 2 of RG 1.174, Page 33 

BIBLIOGRAPHY 

Apostolakis, G.A., “Probability and Risk Assessment: The Subjectivist Viewpoint and Some 

Suggestions,” Nuclear Safety, 19(3), pp. 305-315, 1978. 

Bohn, M.P., T.A. Wheeler, G.W. Parry, “Approaches to Uncertainty Analysis in Probabilistic Risk 

Assessment,” NUREG/CR-4836, US Nuclear Regulatory Commission, January 1988. 

Hickman, J.W., “PRA Procedures Guide,” NUREG/CR-2300, US Nuclear Regulatory Commission, 

January 1983. 

Kaplan, S., and B.J. Garrick, “On the Quantitative Definition of Risk,” Risk Analysis, Vol. 1, pp. 11-28, 

March 1981. 

Mosleh, A., et al., “Proceedings of Workshop I in Advanced Topics in Risk and Reliability Analysis, 

Model Uncertainty: Its Characterization and Quantification” (held in Annapolis, MD, October 20-22, 

1993), US Nuclear Regulatory Commission, NUREG/CP-0138, October 1994. 

Parry, G.W., and P.W. Winter, “Characterization and Evaluation of Uncertainty in Probabilistic Risk 

Analysis,” Nuclear Safety, 22(1), pp. 28-42, 1981. 

Reliability Engineering and System Safety (Special Issue on the Meaning of Probability in Probabilistic 

Safety Assessment), Vol. 23, 1988. 

Reliability Engineering and System Safety (Special Issue on Treatment of Aleatory and Epistemic 

Uncertainty), Vol. 54, Nos. 2 and 3, November/December 1996.
Appendix A to Rev. 2 of RG 1.174, Page A-1 

APPENDIX A 

USE OF RISK-IMPORTANCE MEASURES TO CATEGORIZE 

STRUCTURES, SYSTEMS, AND COMPONENTS WITH RESPECT TO 

SAFETY SIGNIFICANCE 

A-1. Introduction 

For several of the proposed applications of the risk-informed regulation process, one of the 

principal activities is the categorization of structures, systems, and components (SSCs) and human actions 

according to safety significance. The purpose of this appendix is to discuss one way that this 

categorization may be performed to be consistent with Principle 4 (see Figure 2 of Regulatory Guide 

1.174) and the expectations discussed in Section 2.1 of Regulatory Guide 1.174. 

Safety significance of an SSC can be thought of as being related to the role the SSC plays in 

preventing the occurrence of the undesired end state. Thus the position adopted in this regulatory guide is 

that all the SSCs and human actions considered when constructing the PRA model (including those that 

do not necessarily appear in the final quantified model because they have been screened initially, assumed 

to be inherently reliable, or have been truncated from the solution of the model) have the potential to be 

safety significant since they play a role in preventing core damage. 

In establishing the categorization, it is important to recognize the purpose behind the 

categorization, which is, generally, to sort the SSCs and human actions into groups (e.g., those for which 

some relaxation of requirements is proposed and those for which no such change is proposed). It is the 

proposed application that is the motivation for the categorization, and it is the potential impact of the 

application on the particular SSCs and human actions and on the measures of risk that ultimately 

determines which of the SSCs and human actions must be regarded as safety significant within the 

context of the application. This impact on overall risk should be evaluated in light of the principles and 

decision criteria identified in this guide. Thus, the most appropriate way to address the categorization is 

through a requantification of the risk measures. 

However, the feasibility of performing such risk quantification has been questioned when a 

method for evaluating the impact of the change on SSC unavailability is not available for those 

applications. An acceptable alternative to requantification of risk is for the licensee to perform the 

categorization of the SSCs and human actions in an integrated manner, making use of an analytical 

technique, based on the use of PRA importance measures as input. This appendix discusses the technical 

issues associated with the use of PRA importance measures. 

A-2. Technical Issues Associated With the Use of Importance Measures 

In the implementation of the Maintenance Rule (Title 10 of the Code of Federal Regulations (10 

CFR) Section 50.65, “Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power 

Plants”) and in industry guides for risk-informed applications (e.g., the “PSA Applications Guide”1

), the 

Fussell-Vesely Importance, Risk Reduction Worth, and Risk Achievement Worth are the most commonly 

identified measures in the relative risk ranking of SSCs. However, in using these importance measures 

for risk-informed applications, there are several issues that should be addressed. Most of the issues are 

related to technical problems that can be resolved by the use of sensitivity studies or by appropriate 

 

1

 D. True et al., “PSA Applications Guide,” Electric Power Research Institute, TR-105396, August 1995. 
Appendix A to Rev. 2 of RG 1.174, Page A-2 

quantification techniques. These issues are discussed in detail below. In addition, the licensee should be 

aware of and adequately address two other issues: (1) that risk rankings apply only to individual 

contributions and not to combinations or sets of contributors and (2) that risk rankings are not necessarily 

related to the risk changes that result from those contributor changes. When performed and interpreted 

correctly, component-level importance measures can provide valuable input to the licensee. 

Risk-ranking results from a PRA can be affected by many factors, the most important being 

model assumptions and techniques (e.g., for modeling of human reliability or common-cause failures 

(CCFs)), the data used, or the success criteria chosen. The licensee should therefore make sure that the 

PRA is technically adequate. 

In addition to the use of a technically adequate PRA, the robustness of categorization results 

should also be demonstrated for conditions and parameters that might not be addressed in the base PRA. 

Therefore, when importance measures are used to group components or human actions as low-safetysignificant contributors, the information to be provided to the analysts performing qualitative 

categorization should include sensitivity studies or other evaluations to demonstrate the sensitivity of the 

importance results to the important PRA modeling techniques, assumptions, and data. Issues that should 

be considered and addressed are listed below. 

Truncation Limit: The licensee should determine that the truncation limit has been set low 

enough so that the truncated set of minimal cutsets contains all the significant contributors and their 

logical combinations for the application in question and is low enough to capture at least 95 percent of the 

core damage frequency (CDF). Depending on the PRA level of detail (module level, component level, or 

piece-part level), this may translate into a truncation limit ranging from 10-12 to 10-8 per reactor year. In 

addition, the truncated set of minimal cutsets should be determined to contain the important applicationspecific contributors and their logical combinations. 

Risk Metrics: The licensee should ensure that risk in terms of both CDF and large early-release 

frequency (LERF) is considered in the ranking process. 

Completeness of Risk Model: The licensee should ensure that the PRA model is sufficiently 

complete to address all important modes of operation for the SSCs being analyzed. Safety-significant 

contributions from internal hazards, external hazards, and shutdown and low-power initiators should be 

considered by using PRA or other engineering analyses. 

Sensitivity Analysis for Component Data Uncertainties: The sensitivity of component 

categorizations to uncertainties in the parameter values should be addressed by the licensee. Licensees 

should be satisfied that SSC categorization is not affected by data uncertainties. 

Sensitivity Analysis for Common-Cause Failures: CCFs are modeled in PRAs to account for 

dependent failures of redundant components within a system. The licensee should determine that the 

safety-significant categorization takes into account the combined effect of associated basic PRA events, 

such as failure to start and failure to run, including indirect contributions through associated CCF event 

probabilities. CCF probabilities can affect PRA results by enhancing or obscuring the importance of 

components. A component may be ranked as a high risk contributor mainly because of its contribution to 

CCFs, or a component may be ranked as a low risk contributor mainly because it has negligible or no 

contribution to CCFs. 

Sensitivity Analysis for Recovery Actions: PRAs typically model recovery actions, especially 

for significant accident sequences. Quantification of recovery actions typically depends on the time 

available for diagnosis and for performing the action, as well as the training, procedures, and knowledge 
Appendix A to Rev. 2 of RG 1.174, Page A-3 

of operators. A certain degree of subjectivity is involved in estimating the success probability for the 

recovery actions. The concerns in this case stem from situations in which very high success probabilities 

are assigned to a sequence, resulting in related components being ranked as low risk contributors. 

Furthermore, it is not desirable for the categorization of SSCs to be affected by recovery actions that 

sometimes are only modeled for the significant scenarios. Sensitivity analyses can be used to show how 

the SSC categorization would change if all recovery actions were removed. The licensee should ensure 

that the categorization has not been unduly affected by the modeling of recovery actions. 

Multiple Component Considerations: As discussed previously, importance measures are 

typically evaluated on an individual SSC or human action basis. One potential concern raised by this is 

that single-event importance measures have the potential to dismiss all the elements of a system or group 

despite the fact that the system or group has a high importance when taken as a whole. (Conversely, there 

may be grounds for screening out groups of SSCs, owing to the unimportance of the systems of which 

they are elements.) There are two potential approaches to addressing the multiple component issue. The 

first is to define suitable measures of system or group importance. The second is to choose appropriate 

criteria for categorization based on component-level importance measures. In both cases, it will be 

necessary for the licensee to demonstrate that the cumulative impact of the change has been adequately 

addressed. 

While there are no widely accepted definitions of system or group importance measures, if any 

are proposed the licensee should ensure that the measures capture the impact of changes to the group in a 

logical way. The remainder of this paragraph provides an example of the issues that can arise. For frontline systems, one could define a Fussell-Vesely-type measure of system importance as the sum of the 

frequencies of sequences involving failure of that system divided by the sum of all sequence frequencies. 

Such a measure would need to be interpreted carefully if the numerator includes contributions from 

failures of that system caused by support systems. Similarly, a Birnbaum-like measure could be defined 

by quantifying sequences involving the system, conditional on its failure, and summing up those 

quantities. This would provide a measure of how often the system is critical. However, again the support 

systems make the situation more complex. For examples, in a two-division plant, front-line failures can 

occur as a result of failure of support division A in conjunction with failure of front-line division B. 

Working with a figure of merit based on “total failure of support system” would miss contributions of this 

type. 

In the absence of appropriately defined group-level importance measures, the appropriate 

determination must rely on a qualitative categorization by the licensee, as part of the integrated 

decisionmaking process. 

Relationship of Importance Measures to Risk Changes: Importance measures do not directly 

relate to changes in risk. Instead, the risk impact is indirectly reflected in the choice of the value of the 

measure used to determine whether an SSC should be classified as being of high or low safety 

significance. This is a concern whether importances are evaluated at the component or at the group level. 

The PSA Applications Guide1

 suggested values of Fussell-Vesely importance of 0.05 at the system level 

and 0.005 at the component level, for example. However, the criteria for categorization into low and high 

significance should relate to the acceptance criteria for changes in CDF and LERF. This implies that the 

criteria should be a function of the base case CDF and LERF rather than being fixed for all plants. Thus 

the licensee should demonstrate how the chosen criteria are related to, and conform with, the acceptance 

guidelines described in this document. If component-level criteria are used, they should account for the 

risk increase resulting from simultaneous changes to all members of the category. 

SSCs Not Included in the Final Quantified Cutset Solution: Importance measures based on 

the quantified cutsets will not factor in those SSCs that have either been truncated or were not included in 
Appendix A to Rev. 2 of RG 1.174, Page A-4 

the fault tree models because they were screened on the basis of high reliability. SSCs that have been 

screened because their credible failure modes would not fail the system function can be argued to be 

unimportant. The licensee must ensure that these SSCs are considered.