POLICY ISSUE 

NOTATION VOTE 

 

June 6, 2012 SECY-12-0081

FOR: The Commissioners 

FROM: R. W. Borchardt 

 Executive Director for Operations 

SUBJECT: RISK-INFORMED REGULATORY FRAMEWORK FOR NEW 

REACTORS 

PURPOSE: 

This paper responds to the staff requirements memorandum (SRM) on SECY-10-0121, 

“Modifying the Risk-Informed Regulatory Guidance for New Reactors,” dated March 2, 2011 

(Agencywide Documents Access and Management System (ADAMS) Accession No. 

ML110610166). 

SUMMARY: 

The SRM on SECY-10-0121 directed the staff to continue to use the existing risk-informed 

framework, including current regulatory guidance, for licensing and oversight activities for new 

plants, pending additional analysis. Specifically, the SRM directed the staff to engage with 

external stakeholders in a series of tabletop exercises to test various realistic performance 

deficiencies, events, modifications, and licensing bases changes against current U.S. Nuclear 

Regulatory Commission (NRC) policy, regulations, guidance and all other requirements 

(e.g., technical specifications (TS), license conditions, code requirements) that are or will be 

relevant to the licensing bases of new reactors. The purpose of the tabletop exercises was to 

either confirm the adequacy of those regulatory tools (and make the NRC aware of these 

potential scenarios such that commensurate regulatory oversight can be applied) or identify 

areas for improvement, such as potential adjustments to the Reactor Oversight Process (ROP). 

CONTACTS: Donald A. Dube, NRO/DSRA Ronald K. Frahm, Jr., NRR/DIRS/IPAB 

 (301) 415-2986 (301) 415-1483 

 
The Commissioners - 2 - 

In response to the SRM on SECY-10-0121, the staff conducted a series of public workshops 

and meetings with stakeholders from May 4, 2011, through October 26, 2011. Based on the 

results of the tabletop exercises for licensing applications, and considering the Commission’s 

decision reaffirming the existing safety goals, the staff did not identify any potentially significant 

decreases in the enhanced safety margins for new reactors. Therefore, the staff has no specific 

list of options with regard to changes to existing guidance in this category to propose to the 

Commission.

The staff identified a potential gap in the Tier 2 change process regarding severe accident 

features that are not related to ex-vessel severe accident prevention and mitigation. The staff 

recommends addressing this potential gap by a) ensuring that there are sufficient details on all 

key severe accident features in Tier 1, and b) including a Tier 2 change process in future design 

certification rulemakings in Section VIII for non-ex-vessel severe accident features similar to 

Section VIII.B.5.c for ex-vessel severe accident features. 

Furthermore, the staff proposes three options for Commission consideration to address the 

different risk metrics used during new reactor licensing and the risk-informed framework for 

currently operating reactors. The staff recommends transitioning at or before initial fuel load 

from large release frequency (LRF) and conditional containment failure probability (CCFP), 

which are used during design certification and combined license (COL) application reviews, to 

large early release frequency (LERF), which is used in the risk-informed framework for currently 

operating reactors. Thereafter, the regulatory use of LRF and CCFP would be discontinued. 

With regard to the application of the ROP to new reactors, the tabletops demonstrated that 

current risk thresholds are appropriate; however, some changes to the ROP may be warranted 

to implement the existing risk-informed concepts of Regulatory Guide (RG) 1.174, “An Approach 

for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes 

to the Licensing Basis,” for new reactors, and the staff presents three options for the 

Commission’s consideration. The staff recommends an option in which after working with 

internal and external stakeholders, the staff identifies appropriate changes to augment the 

existing risk-informed guidance with deterministic backstops to ensure an appropriate regulatory 

response for the new reactor designs. 

BACKGROUND: 

Risk estimates for new reactor designs are one or more orders of magnitude lower than those 

for operating reactor designs when internally and externally initiated events that have been 

quantified are included. The lower risk values raised questions on how to apply acceptance 

guidelines for changes to the licensing basis and regulatory response in the ROP. 

 

The staff developed a white paper that identifies the issues posed by the lower risk estimates for 

new reactor designs in risk-informed applications and potential options for implementation. The 

white paper was provided to the Commission (ADAMS Accession No. ML090160004) and 

discussed with stakeholders at a public meeting on February 18, 2009. The white paper 

includes consideration of the Commission-established goals for new reactor designs of a core 

damage frequency (CDF) of less than 10-4 per year (/yr); a CCFP of less than 0.1, and an LRF 

of less than 10-6/yr. 
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The Nuclear Energy Institute (NEI) developed an additional white paper to discuss these issues 

and recommended no change to the current risk metrics. NEI submitted their paper on 

March 27, 2009 (ADAMS Accession No. ML090900674). Staff and industry representatives 

briefed the Advisory Committee on Reactor Safeguards (ACRS) on April 3, 2009, which was 

followed by a more detailed presentation to the ACRS Subcommittee on Reliability and 

Probabilistic Risk Assessment (PRA) on June 2, 2009. The staff held a second public meeting 

on September 29, 2009, which focused on the potential issues associated with the ROP 

(ADAMS Accession No. ML092780211). 

Based on these interactions, the staff developed a draft Commission paper (ADAMS Accession 

No. ML101090355) describing the staff's plans to identify appropriate changes to the 

risk-informed guidance for new reactors. The staff held a public meeting on June 3, 2010, and 

an ACRS briefing on June 10, 2010, to discuss the path forward. By letter to the Commission 

dated July 27, 2010 (ADAMS Accession No. ML102000422), ACRS agreed with the staff's 

position on the proposed framework as described in Option 2 of that draft paper. The staff 

reviewed the ACRS letter and responded on August 25, 2010 (ADAMS Accession No. 

ML102210553). The final Commission paper was issued on September 14, 2010 

(SECY-10-0121, ADAMS Accession No. ML102430197). A Commission briefing on the 

topic was held on October 14, 2010. 

Subsequently, the Commission issued an SRM on March 2, 2011, directing the staff to continue 

using the existing risk-informed framework, including current regulatory guidance, for licensing 

and oversight activities for new plants, pending additional analysis. In the SRM, the 

Commission stated that it “reaffirms that the existing safety goals, safety performance 

expectations, subsidiary risk goals and associated risk guidance (such as the Commission’s 

2008 Advanced Reactor Policy Statement and Regulatory Guide 1.174), key principles and 

quantitative metrics for implementing risk-informed decision making, are sufficient for new 

plants.” The Commission further stated that “new reactors with these enhanced margins and 

safety features should have greater operational flexibility than current reactors. This flexibility 

will provide for a more efficient use of NRC resources and allow a fuller focus on issues of true 

safety significance.” The Commission also directed the staff to undertake the following: 

• The staff should articulate, in a single document, a coherent overview of the 

Commission’s policies and decisions regarding new reactor safety performance for the 

purposes of public communication and NRC staff knowledge management. 

• The staff should engage with external stakeholders in a series of tabletop exercises to 

test various realistic performance deficiencies, events, modifications, and licensing 

bases changes against current NRC policy, regulations, guidance and all other 

requirements (e.g., TS, license conditions, code requirements) that are or will be 

relevant to the licensing bases of new reactors. The tabletop exercises should either 

confirm the adequacy of those regulatory tools (and make the NRC aware of these 

potential scenarios such that commensurate regulatory oversight can be applied) or 
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identify areas for improvement, such as potential adjustments to the ROP. Specific 

programs and processes highlighted in the SRM include: 

o Title 10 of the Code of Federal Regulations (10 CFR), Section 50.59, “Changes, 

Tests and Experiments,” during the construction and operational phases of new 

nuclear power plants 

o The change control process addressing severe accident design features under 

10 CFR Part 52, “Licenses, Certifications, and Approvals for Nuclear Power 

Plants” 

o Risk-managed TS 

o 10 CFR 50.69, “Risk-Informed Categorization and Treatment of Structures, 

Systems and Components for Nuclear Power Reactors” 

o ROP 

• Progress reports are to be provided to the Commission every 6 months. 

• A notation vote paper with options and recommendations is to be provided to the 

Commission by June 4, 2012. 

DISCUSSION: 

In response to the SRM on SECY-10-0121, the staff conducted a series of public workshops 

and meetings with stakeholders from May 4, 2011, through October 26, 2011. The results of 

the tabletop exercises including key observations and conclusions regarding regulatory and 

programmatic controls that strengthen the various programs and tend to limit the decrease in 

the enhanced safety margin of the new reactor designs are summarized below and discussed in 

greater detail in Appendices A and B. These controls include such features as deterministic 

backstops, defense-in-depth measures, integrated decisionmaking panels, and performance 

monitoring. The staff also completed the summary-level public communication 

brochure, “New Reactors: Striving for Enhanced Safety,” in November 2011 (NUREG/BR-0356, 

ADAMS Accession No. ML11343A026). The staff provided informational briefings to the ACRS 

Subcommittee on Reliability and PRA on September 20, 2011, and March 7, 2012. A briefing of 

the full ACRS was held on April 12, 2012. 

Licensing Tabletop Exercise Summary

The staff held the following public workshops on licensing applications: 

• December 2, 2010: 10 CFR 50.59-like change process for ex-vessel severe accident 

(EVSA) design features under section VIII.B.5.c of each design certification rule 

• May 4, 2011: Risk-informed inservice inspection of piping (RI-ISI) 
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• May 26, 2011, and June 1, 2011: Risk-informed technical specifications (RITS) initiative 

4b on completion times and Section (a)(4) of the “maintenance rule,” 10 CFR 50.65, 

“Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power 

Plants” 

• June 29, 2011: RITS 5b, surveillance frequency control program 

• August 9, 2011: 10 CFR 50.69 and guidance in Appendix C of NEI 96-07, “Guidelines for 

10 CFR 50.59 Implementation,” specific to the change process for EVSA design features 

under 10 CFR Part 52 “Licenses, Certification, and Approvals for Nuclear Power Plants” 

• October 5, 2011: RG 1.174 and transition options from LRF as a risk metric to LERF 

The staff did not identify any potentially significant decreases in the enhanced safety margins for 

new reactors for RI-ISI, RITS 4b and 5b, 10 CFR 50.65(a)(4), 10 CFR 50.69, and the 

acceptance guidelines and key principles in RG 1.174. For several of these programs, the staff 

identified some potential regulatory and implementation issues that would need to be 

addressed, such as the lack of plant-specific operating experience. A more detailed discussion 

can be found in Appendix A. 

During the 2010 workshop on EVSA design features, the staff discussed the requirements as 

codified in Section VIII.B.5.c of each appendix to 10 CFR Part 52 regarding the criteria for 

departures from Tier 2 affecting resolution of EVSA design features that would require a license 

amendment. There was consensus among the staff and stakeholders present that the current 

rules and guidance may not be complete. Certain severe accident features (e.g., features to 

prevent containment bypass and containment hydrogen mitigation equipment such as igniters) 

do not address “ex-vessel” conditions, as defined in the statement of considerations for the rule. 

In the current regulation, changes to severe accident design features that are not specifically 

intended to address EVSAs are not addressed using severe accident criteria as in 

Section VIII.B.5.c. Depending on the nature of the change, the licensee must follow one of the 

other change procedures in the rule, either Section VIII.B.5.a or Section VIII.B.5.b. If the 

change falls under the requirements of Section VIII.B.5.a, then the licensee could make the 

change without prior NRC approval. Section VIII.B.5.b, however, explicitly excludes changes to 

severe accident features. Moreover, the criteria of Section VIII.B.5.b generally ask if there is 

“more than a minimal increase” in the frequency of occurrence or consequences of an accident 

previously evaluated in the plant-specific design control document (DCD). While not stated 

explicitly in the rule, it is clear from the statement of considerations that design-basis accidents 

are intended here. The EVSA criteria in Section VIII.B.5.c use “substantial increase” in 

probability and public consequences as the standard for determining if prior NRC approval is 

required. The latter criteria are less stringent and more appropriate for non-ex-vessel severe 

accident design features given the large uncertainties associated with severe accident 

phenomena and the ability of analysts to determine precisely the impact of the change on 

probability and consequences. 
The Commissioners - 6 - 

Unless such non-ex-vessel severe accident design features also happen to have a dual function 

such as also addressing design basis accidents or aircraft impacts, risk-significant Tier 2 

changes (e.g., Chapter 19 information related to prevention and mitigation of severe accidents 

other than those considered “ex-vessel”) could be screened out altogether and not receive prior 

NRC approval. Staff has observed that Tier 1 descriptions usually have sufficient detail that 

necessitates prior staff review for major changes to severe accident design features. The staff 

believes, however, that there is a “gap” in the Tier 2 change process for non-ex-vessel design 

features, in that: a) such changes may be screened out or less appropriate criteria are applied 

when determining if prior NRC approval is needed, and b) whether or not prior NRC approval is 

obtained may be highly dependent on the degree of detail in Tier 1, if any. The staff has 

provided a recommendation to the Commission in this paper to address the potential “gap” in 

the 10 CFR Part 52 change process. 

The first topic discussed during the October 5, 2011, workshop was application of the general 

principles of risk-informed regulation for proposed changes to the licensing basis as described 

in RG 1.174. The general consensus of the staff and participating stakeholders is that no 

substantive changes to the key principles and other guidance to address new reactors are 

necessary. However, the staff plans to augment the existing discussion on long-term 

containment performance in Section 2.2 of RG 1.174 by referring to the containment 

performance objectives for new reactors as described in Commission-approved guidance such 

as SECY-90-016, “Evolutionary Light Water Reactor (LWR) Certification Issues and Their 

Relationship to Current Regulatory Requirements,” dated January 12, 1990, and SECY-93-087, 

“Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water 

Reactor (ALWR) Designs,” dated April 2, 1993.1

 

The second topic discussed during the October 5, 2011, workshop was potential options for 

transitioning from LRF and, to a lesser extent, CCFP (the risk metrics used in new reactor 

design certifications and COL applications) to the LERF metric used in the risk-informed 

regulatory framework for currently operating reactors. 

As discussed in Enclosure 1 to SECY-10-0121, the Commission earlier had requested that the 

staff provide a definition of LRF, but in SECY-93-138, “Recommendation on Large Release 

Definition,” dated May 19, 1993, the staff recommended termination of work on a definition, and 

the Commission later approved. As a result, the definitions of LRF in the DCDs referenced in 

COL applications all differ to varying extents. Because of the conservative definitions of LRF 

used in the DCDs, however, the staff has been able to review the PRA and severe accident 

evaluation portions of the DCDs to provide reasonable assurance that the Commission’s 

objectives as described in key policy papers such as SECY-90-016 and SECY-93-087 are 

addressed fully. 

 

1

 The containment should maintain its role as a reliable, leak-tight barrier (for example, by ensuring that 

containment stresses do not exceed ASME Service Level C limits for metal containments, or Factored 

Load Category for concrete containments) for approximately 24 hours following the onset of core 

damage under the more likely severe accident challenges and, following this period, the containment 

should continue to provide a barrier against the uncontrolled release of fission products. 
The Commissioners - 7 - 

The staff has identified three possible options for transitioning from LRF to LERF, and discussed 

the advantages and disadvantages of each. The staff has provided options and a 

recommendation to the Commission in this paper to address the issues. 

ROP Tabletop Exercise Summary

The staff held a half-day public workshop to discuss the ROP-related tabletops on 

October 5, 2011. A brief followup public meeting was held on October 26, 2011. The ROP 

tabletops tested various scenarios that are or will be relevant to the licensing basis for new 

reactors to confirm the adequacy of the current ROP risk-informed processes for regulatory 

decisionmaking or to identify areas for improvement. In preparation for the ROP tabletops, the 

staff developed a broad cross-section of well-vetted cases from actual greater-than-green 

significance determination process (SDP) findings, mitigating systems performance index 

(MSPI) data, and event response (in the risk-informed reactor safety cornerstones of initiating 

events, mitigating systems, and barrier integrity) from the current fleet of reactors. For each 

case study, the staff applied similar situations to the new reactor designs, filling in any gaps with 

realistic hypothetical situations and reasonable assumptions, and then compared the risk values 

and resulting regulatory responses from the new reactor scenarios to those derived from the 

current fleet. A complete summary of the ROP tabletop examples and results was made 

publicly available. 

Appendix B provides details on the ROP framework, existing risk-informed guidance, and 

results from the ROP tabletop exercises that the staff considered when developing conclusions 

and recommendations on potential changes to the ROP to address new reactor designs. 

Appendix B also provides further detail and examples that support the staff’s conclusions. 

In summary, the tabletops demonstrated that current risk thresholds are appropriate; however, 

some changes to the ROP may be warranted to implement the existing risk-informed concepts 

of RG 1.174 for new reactors. These changes will not infringe on the additional operational 

flexibility provided within the licensing basis of the new, more robust reactor designs. 

The risk-informed concept of RG 1.174 places a strong emphasis on the quantified increase in 

probability of CDF or LERF, but it also provides that basic deterministic principles such as 

diversity and defense in depth be maintained. Because of the history of design and 

modifications for the existing operating fleet, the extended loss of diversity or defense in depth, 

such as the loss of an entire safety system function in the initiating events or mitigating systems 

cornerstones, will nearly always result in crossing a threshold that results in an increased 

regulatory response for quantified increase in CDF or LERF. As a result of the ROP tabletop 

exercises, the staff identified specific areas in which, consistent with the integrated 

risk-informed decisionmaking framework in RG 1.174, additional deterministic backstops would 

complement the current set of risk-informed ROP tools, thereby ensuring regulatory responses 

to performance issues, events, and degraded conditions are reasonable and appropriate. 

The SDP tabletops indicated that the existing risk thresholds for determining significance of 

inspection findings are generally acceptable, and greater-than-green thresholds could be 

crossed that would produce an increased regulatory response. However, these 

greater-than-green inspection findings would likely involve common cause failures that affect 

multiple systems and/or long exposures of risk-significant components. Further, the staff found 
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that the existing process does not always ensure an appropriate regulatory response for 

degradation of passive components and barriers. Based on the SDP tabletop exercises, the 

staff concluded that the SDP analyses could be augmented with additional qualitative 

considerations (such as deterministic backstops) to appropriately address performance issues. 

Potential deterministic backstops could include an emphasis on barrier integrity, placing a limit 

on extensive equipment outage times and addressing repetitive or common cause equipment 

failures. These deterministic backstops should not infringe upon the operational flexibility 

afforded by the more robust new reactor designs, but should ensure that infrequent yet 

potentially significant performance issues that would not otherwise be considered based on risk 

calculations receive an appropriate regulatory response. 

The tabletop exercises for event response demonstrated that the existing risk thresholds for 

invoking reactive inspections are adequate for new reactors, and that thresholds could be 

crossed to invoke reactive inspections. The tabletops revealed that although deterministic 

criteria already play an important role in the process, the criteria are used initially for event 

screening, and are then considered again within a range of responses determined by the risk 

values. As a result of this current structure, risk values heavily influence whether a reactive 

inspection is warranted and, if so, at what level. The staff concluded that the contribution of the 

existing deterministic criteria could be modified, or new deterministic criteria could potentially be 

developed for initiating reactive inspections for new reactors, similar to those for the SDP 

discussed previously. 

The case studies developed for the MSPI tabletops showed that the existing Mitigating System 

Performance Index (MSPI) is not adequate and would be largely ineffective in determining an 

appropriate regulatory response for active new reactor designs. Furthermore, a meaningful 

MSPI may not even be possible for passive systems using the current formulation of the 

indicator. The staff noted that the existing performance limit approach, which incorporates a 

backstop, potentially could be modified and emphasized for active new reactor designs. The 

staff concluded that alternate performance indicators in the mitigating systems cornerstone 

could be developed and/or additional inspection could be used for the new reactors to 

supplement insights currently gained through MSPI for the current fleet. 

OPTIONS AND RECOMMENDATIONS: 

Based on the staff’s observations of the tabletop exercises for licensing applications, and 

considering the Commission’s decision reaffirming the existing safety goals, the staff did not 

identify any potentially significant decreases in the enhanced safety margins for new reactors. 

Therefore, the staff has no specific list of options with regard to changes to existing guidance in 

this category to propose to the Commission. 

1. Tier 2 Change Process

As noted above, the staff identified a potential gap in the Tier 2 change process regarding 

severe accident features that are not related to ex-vessel severe accident prevention and 

mitigation. The current change process does not address all of the severe accidents defined in 

10 CFR 52.47(a)(23) and 10 CFR 52.79(a)(38). The current regulation and its implementation 

through the guidance in draft Appendix C of NEI 96-07 could result in the licensee screening out 

changes in Chapter 19 (and other sections) of Tier 2 of the DCD and final safety analysis report 
The Commissioners - 9 - 

(FSAR) that do not affect ex-vessel severe accident design features. In a worst-case scenario, 

significant Tier 2 changes to non-ex-vessel severe accident features, up to and including 

permanent removal from service, could be made without prior NRC approval. 

Recommendation 1: Staff recommends addressing the potential gap in the Tier 2 change 

process, by a) ensuring that there are sufficient details on all key severe accident features in 

Tier 1, and b) including a change process in future design certification rulemakings in 

Section VIII for non-ex-vessel severe accident features similar to Section VIII.B.5.c for ex-vessel 

severe accident features. By letter dated April 26, 2012 (ADAMS Accession No. 

ML12107A199), the ACRS agreed with the staff’s recommendation to address the potential gap 

in the Tier 2 change process. 

Unless otherwise directed by the Commission, the staff will issue appropriate regulatory 

guidance to ensure that, for future design certifications, severe accident features that the staff 

previously reviewed and deemed important to prevent and/or mitigate the full range of severe 

accidents have sufficient detail in Tier 1 so as to preclude major plant changes that could 

remove (or significantly degrade) the performance of this equipment without prior NRC 

approval. Additionally, future rulemaking on design certifications would include a change 

process for non-ex-vessel severe accident design features that uses criteria similar to the 

current criteria in Section VIII.B.5.c. Implementation of this recommendation has the advantage 

of addressing the identified gap in the Tier 2 change control process and ensuring that 

significant changes to non-ex-vessel severe accident features receive the same level of 

screening and review as ex-vessel severe accident features. A possible disadvantage of 

implementing this recommendation is that it may add one or more steps to the overall change 

control process that licensees use. The new guidance would not apply to already-certified 

designs unless backfits were pursued. Based on reviews of the DCDs for the AP1000 and the 

Advanced Boiling Water Reactor (ABWR) the staff does not believe backfits are justified at this 

time since sufficient detail of severe accident features is already included in Tier 1. Similarly, 

the staff concludes there is sufficient detail of severe accident features included in Tier 1 of the 

DCD for the Economic Simplified Boiling Water Reactor design; this design is currently in the 

final stages of rulemaking. 

2. Transition from LRF to LERF

As noted above, the staff proposes several options for the Commission’s consideration to 

address the different risk metrics used during new reactor licensing and the risk-informed 

framework for currently operating reactors. 

Option 2A: Continue use of LRF (and CCFP) indefinitely (status quo) 

Under this option, COL holders would continue to apply the same risk metrics used during 

design certification and in the COL application (i.e., CDF, LRF, and CCFP) for all risk-informed 

applications throughout commercial operation. This option has the advantage of maintaining 

the same definition of LRF and using the same risk metrics as documented in the DCD, FSAR, 

and the staff review documented in the final safety evaluation report (FSER). A major 

disadvantage is that there is presently no common definition of LRF and CCFP, neither within 

the NRC nor in existing or proposed consensus standards on PRA. Moreover, this option may 

be inconsistent with the direction provided to the staff in the SRM on SECY-10-0121, in which 
The Commissioners - 10 - 

the Commission reaffirmed that the existing quantitative metrics for implementing risk-informed 

decision making are sufficient for new reactors. Implementing acceptance guidelines in 

RG 1.174 as well as the ROP, both of which explicitly use LERF, would be problematic for new 

reactors. 

Option 2B: Continue use of LRF (and CCFP) indefinitely and add LERF at or before initial fuel 

load 

Under this option, COL holders would continue to apply CDF, LRF, and CCFP throughout 

commercial operation. LERF would be added at or before initial fuel load, depending on 

whether the licensee proposes to implement one or more risk-informed initiatives. Similar to 

Option 2A, this option has the advantage of maintaining the same definition of LRF and using 

the same risk metrics as documented in the DCD, FSAR, and the staff review documented in 

the FSER. The use of CDF and LERF for risk-informed changes to the licensing basis is 

consistent with RG 1.174 for currently operating reactors. Another advantage is that continued 

use of LRF and CCFP would support the calculation of long-term containment performance as 

discussed in Section 2.2 of RG 1.174. A major disadvantage is that there is presently no 

common definition of LRF and CCFP, neither within the NRC nor in existing or proposed 

consensus standards on PRA. The continued use of both LRF and LERF in commercial 

operation adds confusion and may be viewed as an unnecessary burden on licensees. Similar 

to Option 2A, this option may be inconsistent with the direction provided to the staff in the SRM 

on SECY-10-0121 in which the Commission reaffirmed that the existing quantitative metrics for 

implementing risk-informed decisionmaking are sufficient for new reactors. 

Option 2C: Transition from LRF to LERF at or before initial fuel load; discontinue regulatory 

use of LRF (and CCFP) thereafter 

 

Under this option, COL holders would not need to calculate LRF and CCFP in regulatory 

applications following their transition to LERF at or before initial fuel load, consistent with the 

requirement in 10 CFR 50.71(h)(1) that states that no later than the scheduled date for initial 

loading of fuel, each holder of a COL shall develop a level 1 and a level 2 PRA. The PRA must 

cover those initiating events and modes for which NRC-endorsed consensus standards on PRA 

exist 1 year before the scheduled date for initial loading of fuel. The current level 1 PRA 

standard endorsed by the NRC uses LERF. A consensus level 2 PRA standard using 

radionuclide release categories as end states has been drafted but, as of the issuance of this 

paper, has not been issued for use or endorsed by the staff. It is expected that by the time of 

initial fuel load for the first set of COL holders, a level 2 PRA standard will be endorsed by the 

NRC. The staff expects that COL holders would use the level 2 PRA radionuclide release 

category frequencies and appropriately roll these up to LERF to meet 10 CFR 50.71(h)(1). No 

existing or proposed PRA standard provides a universal definition of LRF. 

Option 2C has the advantage of harmonizing the risk metrics for all reactors, both currently 

operating and new reactors, going forward after commercial operation. Moreover, the option is 

consistent with the direction provided to the staff in the SRM on SECY-10-0121 in which the 

Commission reaffirmed that the existing quantitative metrics for implementing risk-informed 

decisionmaking are sufficient for new reactors. A disadvantage of this option is that LRF and 

CCFP, which form part of the original design objective in the design certification, would no 

longer be tracked into commercial operation. LRF, which could be used as a measure of 
The Commissioners - 11 - 

determining the impact of risk-informed changes on late containment failure, no longer would be 

available. This disadvantage would be partly offset by the staff’s plan as discussed above to 

augment the existing discussion on long-term containment performance for new reactors in 

RG 1.174 by referring to the containment performance objectives in SECY-90-016 and 

SECY-93-087. 

Recommendation 2: The staff recommends Option 2C. This option harmonizes the 

risk-informed applications for the new reactors consistent with the risk metrics that the currently 

operating fleet uses. Moreover, the option does not introduce confusion and create burden on 

licensees as would be the case if both LRF and LERF were used going forward. The staff 

believes that adding a reference to containment performance for new reactors in RG 1.174 per 

SECY-90-016 and SECY-93-087 is appropriate and consistent with established Commission 

policy. In its letter of April, 26, 2012, the ACRS agreed with this recommendation. 

3. ROP

As a result of the tabletop exercises, the staff developed three options for the Commission’s 

consideration for applying the risk-informed regulatory framework of the ROP to new reactors. 

These options do not change the existing risk thresholds in the current ROP risk-informed 

processes and are consistent with the integrated risk-informed, decisionmaking framework in 

RG 1.174. In addition, these options do not infringe upon the greater operational flexibility 

afforded by the enhanced safety margins of the new reactor designs. These options apply to 

the risk-informed reactor safety cornerstones of initiating events, mitigating systems, and barrier 

integrity, while the other four more deterministic cornerstones would not be affected directly. 

These options provide a transparent and predictable way for the NRC to respond to events and 

performance issues at new reactors and to perform assessments of licensee performance. 

Option 3A: Use as is (status quo)

Under this option, the staff would use the existing risk-informed ROP tools for new reactor 

applications without making any changes. The staff would: 

(1) Continue to use the current SDP to process inspection findings in the reactor safety 

cornerstones as stipulated in Appendix A to Inspection Manual Chapter (IMC) 0609, 

“Significance Determination Process” 

(2) Continue to use the current event response guidance as stipulated in Management 

Directive (MD) 8.3, “NRC Incident Investigation Program,” and IMC 0309, “Reactive 

Inspection Decision Basis for Reactors” 

(3) Continue to use the existing MSPI guidance as stipulated in NEI 99-02, “Regulatory 

Assessment Performance Indicator Guideline,” and IMC 0608, “Performance Indicator 

Program,” recognizing that the current MSPI guidance does not apply to passive 

systems 
The Commissioners - 12 - 

An advantage to Option 3a is that no additional action or resources would be necessary. A 

disadvantage to this option is that the existing tools may not always provide for an appropriate 

regulatory response to address performance issues and plant events for new reactor designs. 

Option 3B: Augment existing processes 

Under this option, the staff would use the existing risk-informed ROP tools, but augment the 

qualitative tools with deterministic backstops to ensure an appropriate regulatory response for 

the new reactor designs. These deterministic backstops would be consistent with the licensing 

basis and the existing defense in depth, safety margins, and other traditional engineering 

principles described in RG 1.174. The staff would: 

(1) Develop deterministic backstops or other qualitative considerations for characterizing the 

significance of inspection findings in the reactor safety cornerstones to compensate for 

shortfalls noted during the tabletop exercises and allow for a transparent and predictable 

process for determining the appropriate regulatory response to address performance 

issues 

(2) Modify the contribution of existing deterministic criteria or develop new deterministic 

criteria for initiating a reactive inspection for events or degraded conditions at new 

reactor facilities, to provide a transparent and predictable process for determining the 

appropriate regulatory response to plant events 

(3) For active new reactor designs develop a risk-informed alternative to MSPI (new PIs or 

risk-informed inspection), or augment the existing MSPI guidance to place more 

emphasis on the performance limit (backstop) or revise the performance limit (backstop); 

also, for passive new reactor designs increase inspection of passive mitigating systems 

as necessary to supplement insights that will not be afforded with MSPI 

Under this option, the deterministic backstops would be developed and inserted into the existing 

risk-informed framework to address the noted shortfalls identified during the tabletop exercises. 

The staff envisions that augmenting existing tools with deterministic backstops would provide for 

a more appropriate response to performance issues and plant events for the new reactor 

designs. Furthermore, the impact on the current fleet would be minimal because the existing 

risk thresholds would likely be surpassed, in most cases, before the potentially new backstops 

were applied. The staff also anticipates that deterministic backstops would only affect the inputs 

to the ROP Action Matrix (i.e., risk-informed SDP and PIs), and the ROP Action Matrix would 

function the same for new reactors as it does for the current fleet in determining regulatory 

response. The staff expects that the proposed process enhancements could be researched and 

developed over the next few years, using existing resources, well in advance of their potential 

implementation in the oversight of new reactor operations. These process enhancements, if 

approved and implemented, would be refined based on experience and lessons learned 

consistent with existing provisions for ROP continuous improvement. More specific potential 

guidance changes to address the noted gaps in the risk-informed ROP processes are presented 

in Appendix B. 

 
The Commissioners - 13 - 

Under this option, the staff would obtain Commission approval for the proposed changes to the 

ROP before implementation. 

Option 3C: Develop deterministic tools 

Under this option, the staff would not use the existing risk-informed ROP tools for new reactor 

applications, but capture risk insights in a more simplified manner using deterministic guidance 

for regulatory decisionmaking consistent with new reactor design certification and licensing 

basis. The staff would: 

(1) Develop a deterministic SDP to characterize the significance of inspection findings in the 

reactor safety cornerstones for new reactor designs 

(2) Develop deterministic event response criteria for MD 8.3 that would apply to new 

reactors and guide reactive inspection outcomes 

 

(3) Develop deterministic PIs or additional inspection activities in the mitigating systems 

cornerstone to replace the MSPI for the new reactor designs 

New deterministic guidance would be developed to replace the existing risk-informed ROP tools 

under this option. Risk insights would be captured, but to a much lesser extent than they are for 

the current fleet. Additional resources may be necessary to research and develop the new 

guidance documents. 

Under this option, the staff would obtain Commission approval for the proposed changes to the 

ROP before implementation. 

Recommendation 3: The staff recommends Option 3B. The staff would work with internal and 

external stakeholders to formulate the process changes as necessary to provide for an 

appropriate regulatory response for new reactor applications. The staff would provide a paper 

to the Commission with its proposed ROP guidance changes at least 1 year before their 

scheduled implementation. 

By their letter dated April 26, 2012, the ACRS recommended that the staff consider a relative 

risk Option 3D for the ROP and RG 1.174 acceptance guidelines. While the staff understands 

the ACRS recommendation, the staff believes that an approach involving relative risk was 

previously considered but was not pursued for reasons discussed below. In addition, the staff’s 

proposed approach to use deterministic backstops to supplement the risk insights is a simpler 

approach to achieving the desired outcome while remaining consistent with the existing ROP 

framework and program goals of being objective, risk informed, understandable, and 

predictable. In the February 12, 2009, white paper (Enclosure 1 to SECY-10-0121), the staff 

considered the merits of a relative risk metric. Impediments to this approach were identified by 

both internal and external stakeholders. Therefore, the staff did not consider this option further 

or include it in SECY-10-0121. In its SRM to SECY-10-0121, the Commission did not approve 

the development of lower numeric thresholds for new reactors in which the ACRS 

recommendation would effectively result. The Commission directed the staff to conduct tabletop 

exercises to either confirm the adequacy of current regulatory tools or identify areas for 

improvement. The ROP tabletop exercises confirmed the adequacy of the current ROP 
The Commissioners - 14 - 

framework and risk thresholds. However, the staff acknowledged that the current SDP and 

reactive inspection guidance may not be sufficient by itself to support an appropriate regulatory 

response to licensee performance deficiencies and events because of the lower risk profiles 

associated with new reactor designs. Therefore, the staff proposes deterministic backstops to 

ensure regulatory responses are appropriate. The existing ROP framework provides for 

deterministic considerations in regulatory decisionmaking in accordance with RG 1.174. 

Deterministic backstops for new reactors would provide a clear, efficient and reliable way of 

ensuring appropriate and predictable regulatory responses within the existing ROP framework, 

consistent with the principles of good regulation. 

RESOURCES:

Implementation of Recommendation 1 would require minor changes to existing guidance 

documents as part of the routine revision to such documents, as well as changes to rule 

language on future design certifications under 10 CFR Part 52. The additional resources would 

be minimal. 

Implementation of Option 2A (status quo) would require no additional resources. Options 2B or 

2C would necessitate some changes to existing guidance documents (e.g., RG 1.174), but the 

resources would be minimal. 

For the ROP, implementation of Option 3A (use as is) would require no additional resources. 

Option 3B (augment existing processes) would require staff resources to engage stakeholders, 

evaluate proposed changes, and draft updates to guidance documents. Based on recent 

experience with the development of risk-informed regulatory guidance, this effort is estimated to 

require no more than 1.0 full-time equivalent during fiscal year (FY) 2013 and FY 2014. 

Although this activity is not specifically included in the FY 2013 and FY 2014 budgets, resources 

for oversight support are available within the operating reactor business line to complete this 

work. Option 3C (develop deterministic tools) likely would require additional staff resources 

beyond those currently budgeted, depending on the extent of effort needed to evaluate potential 

deterministic tools and the level of stakeholder involvement (e. g., number of public meetings 

needed). The extent of the resultant changes to existing guidance documents or the creation of 

new ones is not yet known; therefore, resource estimates would be developed following 

stakeholder interaction and appropriate adjustments would be made through the planning, 

budgeting, and performance management process. 
The Commissioners - 15 - 

COORDINATION:

This paper has been coordinated with the Office of the General Counsel, which has no legal 

objection. The Office of the Chief Financial Officer has reviewed this paper for resource 

implications and has no objections. A copy of this paper has been provided to the Advisory 

Committee on Reactor Safeguards. The ACRS issued a letter dated April 26, 2012 (ADAMS 

Accession No. ML12107A199), about its conclusions and recommendations on the paper. The 

staff provided a response to each of the recommendations in the ACRS letter (ADAMS 

Accession No. ML12123A695). 

/RA by Michael Johnson for/ 

R. W. Borchardt 

Executive Director 

 for Operations 

Enclosures: 

1. Appendix A: Licensing Tabletop 

 Exercise Results 

2. Appendix B: ROP Background and 

 Tabletop Exercise Results 
The Commissioners - 15 - 

COORDINATION:

This paper has been coordinated with the Office of the General Counsel, which has no legal 

objection. The Office of the Chief Financial Officer has reviewed this paper for resource 

implications and has no objections. A copy of this paper has been provided to the Advisory 

Committee on Reactor Safeguards. The ACRS issued a letter dated April 26, 2012 (ADAMS 

Accession No. ML12107A199), about its conclusions and recommendations on the paper. The 

staff provided a response to each of the recommendations in the ACRS letter (ADAMS 

Accession No. ML12123A695). 

/RA by Michael Johnson for/ 

R. W. Borchardt 

Executive Director 

 for Operations 

Enclosures: 

1. Appendix A: Licensing Tabletop 

 Exercise Results 

2. Appendix B: ROP Background and 

 Tabletop Exercise Results 

WITS 201100112/EDATS: SECY-2011-0132 

WITS 201100113/EDATS: SECY-2011-0133 

ADAMS Accession Number: ML12117A012 *via email SECY-012

OFFICE NRO/DSRA NRR/DIRS/IPAB D: NRO/DSRA* D: NRR/DIRS* D:NRR* D:NSIR* 

NAME DDube RFrahm CAder HNieh ELeeds (BBoger for) JWiggins 

DATE 4/25 /12 4/25/12 4/25/12 4/26/12 5/9/12 5/9/12 

OFFICE RA:RI* RA:RII* RA:RIII* RA:RIV * OCFO * NRO/DARR* 

NAME WDean (DLew for ) VMcCree CPederson ECollins 

(AHowell for) 

JDyer (DPelton for) MMayfield 

(SCoffin for) 

DATE 5/11/12 4/26/12 5/7/2012 5/14/12 5/10/12 5/10/12 

 OFFICE Tech Editor* OGC* D:NRO EDO 

NAME JDougherty JMartin GHolahan RWBorchardt 

(MJohnson for) 

 

DATE 5/15/12 5/8/12 5/23/12 6/ 6 /12 

 OFFICIAL RECORD COPY
 

 Enclosure 1 

Appendix A: Licensing Tabletop Exercise Results

1. Risk-Informed Inservice Inspection of Piping

 

The NRC staff held a 1-day public workshop on May 4, 2011. The Electric Power Research 

Institute (EPRI) first provided an overview of its methodologies on risk-informed in-service 

inspection of piping (RI-ISI). The staff then provided sample results from actual licensee 

submittals for currently operating reactors supporting the point that the theoretically calculated 

changes in core damage frequency (CDF) and large early release frequency (LERF) are 

sometimes positive, sometimes negative, but virtually always low in absolute magnitude. EPRI 

also provided scoping calculations of the potential impact of an RI-ISI program for a new active 

plant and a new passive plant that showed the effects continued to be risk-neutral, even when 

sensitivity studies using more restrictive acceptance criteria were assumed. The staff observed 

that RI-ISI simply changes the locations for inservice inspection in a risk-informed manner. 

Neither the design nor plant operational configurations are affected by RI-ISI. 

The staff identified the features as well as regulatory and programmatic controls that strengthen 

the RI-ISI program and tend to limit the decrease in enhanced safety margin of the new reactor 

designs. These controls are discussed in greater detail in the meeting summary package 

(ADAMS Accession No. ML111330381). Several key features include, for example: 

• The guidelines on potential increases to the baseline CDF and LERF are imposed at a 

system level as well as the overall totals. This ensures that no one system absorbs 

most of the change in risk and it helps to address uncertainty in system-level risk results. 

• Inspection of a minimum set of weld locations is required regardless of what the risk 

levels are calculated to be. This deterministic feature provides additional safety margins. 

• A number of programs remain in place to address degradation mechanisms such as flow 

accelerated corrosion and microbiologically induced corrosion. These programs provide 

added levels of assurance to the risk-informed elements of RI-ISI. 

The general consensus of the staff and participating stakeholders from the tabletop exercise is 

that use of the current guidance for RI-ISI for a new reactor design with sufficient operating 

experience would not result in any significant decrease in enhanced safety. 

 

2. Risk-Informed TS Initiative 4b (Completion Times) and Maintenance Rule Section (a)(4)

The staff held a 2-day public workshop on May 26, 2011, and June 1, 2011. The staff 

addressed RITS 4b and (a)(4) of Title 10 of the Code of Federal Regulations (10 CFR) 50.65, 

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

together because of the complementary nature of risk-managed technical specifications (TS) 

and the management of the risk increase that may result from the proposed maintenance 

activities. The staff led the workshop by providing a brief overview of the RITS 4b methodology. 

A representative of the South Texas Project (STP) provided a comprehensive discussion of the 

implementation of RITS 4b at STP Units 1 and 2. The representative also highlighted the use of 

the on-line risk monitoring tools, case studies, best practices, and important lessons learned. 

The staff then presented calculation results using standardized plant analysis Risk (SPAR) 
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models for internal events for the Advanced Boiling Water Reactor (ABWR) and AP1000 reactor 

designs. The configurations tested spanned a wide range from single equipment outages to 

multiple equipment outages across safety divisions. The staff identified some plant 

configurations that could create a large incremental core damage probability (ICDP), such as 

having multiple trains of equipment out of service simultaneously. Repeated entry into such 

conditions over time could increase the baseline CDF by one or more orders of magnitude. The 

staff, however, believes these configurations are unlikely or unrealistic, would not occur 

repeatedly, and there are numerous programmatic and regulatory controls that would limit the 

aggregated risk increase. 

Representatives from Mitsubishi Heavy Industries (MHI), General Electric-Hitachi, and AREVA 

presented the results of case study calculations for the US-APWR, Economic Simplified 

Boiling-Water Reactor (ESBWR), and U.S. EPR designs, respectively. In addition, the 

Westinghouse representative provided comments on calculations he performed that generally 

confirmed staff calculations using the AP1000 SPAR model. Taken as a whole, the calculations 

indicated that two major features of RITS 4b tended to limit the potential risk increase from 

various maintenance configurations, specifically: 

• The risk-informed completion time is limited to a deterministic maximum of 30 days 

(referred to as the backstop completion time) from the time the TS action was first 

entered. 

• Voluntary use of the risk-managed TS for a configuration that represents a loss of 

TS-specified safety function, or inoperability of all required safety trains, is not permitted 

by the guidance. 

When comparing RITS 4b to existing standard TS the provide fewer controls on the frequency 

of entering certain limiting conditions for operation, the staff notes that it is conceivable that 

implementation of RITS 4b may be shown to be no worse than “risk neutral.” Key to ensuring 

that risk does not increase over time is to limit the frequency of entering higher risk maintenance 

configurations that could otherwise have the effect of increasing the baseline CDF. 

Performance monitoring under NEI 06-09, “Risk-Managed Technical Specifications (RMTS) 

Guidelines,” RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed 

Decisions on Plant-Specific Changes to the Licensing Basis,” and RG 1.177, “An Approach for 

Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications,” are therefore key 

programmatic controls. Additional discussion on regulatory and programmatic controls can be 

found in the meeting summary packages (ADAMS Accession Nos. ML111650176 and 

ML111650341). 

The general consensus of the staff and participating stakeholders from the tabletop exercise is 

that no substantive changes to methodology are necessary to implement RITS 4b for new 

reactor designs. However, certain implementation and process issues may need to be 

addressed before implementing RITS 4b. 

A related topic to RITS 4b is implementation of 10 CFR 50.65(a)(4) at new reactor designs. 

Industry representatives began the discussion by providing a detailed overview of the 

10 CFR 50.65(a)(4) experience, as well as a demonstration of a widely used risk monitor tool. 

The blended approach whereby the probabilistic risk assessment (PRA) is combined with inputs 
A-3 

on the degree of defense in depth and plant transient assessment was highlighted. The staff 

observed that factors other than PRA often were more limiting in terms of the risk management 

action level. Finally, the staff presented the results of 10 CFR 50.65(a)(4) inspection experience 

over a 10-year timeframe. Of 116 violations that occurred during this time, all were categorized 

as very low-risk significance (green) findings. 

The general consensus of the staff and participating stakeholders from the tabletop exercise on 

Section (a)(4) of the maintenance rule, as it pertains to new reactor designs, is that Section 11 

of NUMARC 93-01, “Industry Guideline for Monitoring the Effectiveness of Maintenance at 

Nuclear Power Plants” (endorsed in RG 1.182, “Assessing and Managing Risk Before 

Maintenance Activities at Nuclear Power Plants”), does not appear to need substantive change 

to address qualitative and quantitative considerations in the assessment and management of 

risk for new reactor designs. Of note is Section 11.3.7.2 of NUMARC 93-01 which states: 

Due to differences in plant type and design, there is acknowledged variability in 

baseline core damage frequency and large early release frequency. Further, 

there is variability in containment performance that may impact the relationship 

between baseline core damage frequency and baseline large early release 

frequency for a given plant or class of plants. Therefore, determination of the 

appropriate method or combination of methods as discussed above, and the 

corresponding quantitative risk management action thresholds, are plant-unique 

activities. 

It is noted that some changes to NUMARC 93-01 may be necessary to address changes of 

scope because of new and different structures, systems and components (SSCs) in the new 

reactor designs. 

3. Risk-Informed TS Initiative 5b (Surveillance Frequency Control Program)

The staff held a 1-day public workshop on June 29, 2011. The staff began by providing an 

overview of the surveillance frequency control program (SFCP). The implementation process 

and administrative controls were also discussed. Representatives from industry discussed their 

experience of implementing RITS 5b per NEI 04-10, “Risk-Informed Method for Control of 

Surveillance Frequencies,” at several operating nuclear power plants. Issues such as scope, 

commitment review, defense-in-depth review, operating experience review, and performance 

monitoring were highlighted. The results of several sensitivity studies that Westinghouse 

performed for the AP1000 reactor design also were summarized. Potential risk increases were 

calculated to be very low, with changes in CDF in the range of 10-9/yr to 10-7/yr. 

The staff identified the regulatory and programmatic controls in the existing guidance. These 

controls and other features of the RITS 5b program provide reasonable assurance that the 

enhanced safety margins of the new reactor designs are maintained. These controls and 

features include, for example: 

• Surveillance frequencies that are controlled by other programs are excluded from the 

SFCP. Equipment covered by in-service testing (e.g., major pumps and valves) tend to 
A-4 

have some of the highest risk importance values but are excluded. What remains to be 

implemented under RITS 5b are generally lower risk importance components. 

• The integrated decisionmaking panel’s (IDP) review of proposed changes is seen as 

strengthening the process. A broad range of expertise is brought to this subject matter. 

• RITS 5b is implemented using a phased approach whereby surveillance test intervals 

are gradually increased from, for example, monthly to quarterly to annually. This 

provides reasonable assurance that failure rate changes are identified and addressed 

before becoming unacceptably high. 

• Monitoring and feedback, and periodic reassessment (e.g., every 6 months) are fed back 

to the IDP. Actual changes in the reliability of equipment modeled in the PRA are 

included in the periodic updates. If a change is found to result in unacceptable 

equipment performance, the IDP may take corrective action by returning the surveillance 

frequency to the previous setting. 

• The impact of changes under the SFCP on defense in depth, maintenance rule 

(e.g., Section (a)(1)), the mitigating systems performance index (MSPI), and other 

programs are generally assessed. Often, these programs limit the scope of RITS 5b 

changes because of the potential to reduce operational and safety margins. 

The general consensus of the staff and participating stakeholders from the tabletop exercise on 

RITS 5b is that the program is much more deterministically oriented than RITS 4b (completion 

times) where the quantitative risk assessment is key to the application. In RITS 5b, risk impact 

is only a secondary consideration in the criteria for changing surveillance test intervals. 

Participants in the workshop recognized the need to obtain sufficient baseline operating 

experience on affected equipment during the initial cycle or cycles of reactor operation before 

beginning full implementation of RITS 5b in the new plants. A detailed discussion of the 

workshop can be found in the meeting summary package (ADAMS Accession No. 

ML11182A976). 

4. Risk-Informed Categorization and Treatment of Structures, Systems and Components, 

10 CFR 50.69

The staff held a 1-day public workshop on August 9, 2011. EPRI first provided an overview of 

the methodologies on risk-informed categorization and treatment of active and passive 

components under 10 CFR 50.69 “Risk-informed Categorization and Treatment of Structures, 

Systems and Components for Nuclear Power Reactors,” per guidance in NEI 00-04, “10 CFR 

50.69 SSC Categorization Guideline.” ASME Code Cases N-660, “Risk-Informed Safety 

Classification for Use in Risk-Informed Repair/Replacement Activities,” and draft N-752, 

“Risk-Informed Safety Classification and Treatment for Repair/Replacement Activities in Class 2 

and 3 Moderate Energy Systems” were also briefly discussed. For passive components (e.g., 

piping), there is consistency with the RI-ISI program. The important role of the IDP also was 

highlighted. Representatives from industry discussed some of the considerations for new 

reactor designs. Participants noted that typically only 2,000 components may be modeled in the 

PRA, leaving thousands of other components to be classified on the basis of criteria other than 
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quantitative risk measures. Finally, EPRI discussed sample categorization for a new 

pressurized-water reactor design with active safety features. The RISC-1, 2, 3, and 4 

categorization distribution mirrored results from STP Units 1 and 2, recognizing that the 

classification for the new design was based strictly on risk importance measures and had not 

been through the IDP. A sample categorization for active and passive components in a new 

reactor safety injection system also was provided. 

Meeting participants identified the features as well as regulatory and programmatic controls that 

strengthen the 10 CFR 50.69 program and tend to limit the decrease in enhanced safety 

margins of the new reactor designs. These include, for example: 

• That 10 CFR 50.69 does not change the nuclear plant system designs, but simply 

changes the treatment of SSCs. 

• NRC staff review and approval of a license amendment request is required. Typically, 

methods and sample results are provided to the staff as part of the review cycle. 

• There is similarity between the regulatory treatment of nonsafety systems (RTNSS), 

reviewed by the staff for passive reactor designs, and the treatment of RISC-2 SSC 

components. 

• For RISC-3 components, there are regulatory requirements for performance monitoring 

and timely corrective action. Periodic review of the program is also required. 

The staff has concluded that passive system categorization (e.g., piping) is similar to RI-ISI, 

which an earlier tabletop demonstrated no decrease in enhanced level of safety. Categorization 

of active components (e.g., pumps and valves) are based on relative risk importances, 

independent of baseline risk, so adverse impact on enhanced safety for new designs should not 

be expected. A detailed discussion of the workshop can be found in the meeting summary 

package (ADAMS Accession No. ML112290891). 

5. Guidance in NEI 96-07 Appendix C on the 10 CFR Part 52 Change Process 

In conjunction with the NRC Working Group on Changes during Construction, the staff held a 

1-day public workshop on December 2, 2010, before the issuance of the Commission’s SRM on 

SECY-10-0121, to address the 10 CFR Part 52 “Licenses, Certification, and Approvals for 

Nuclear Power Plants,” change process specific to ex-vessel severe accident (EVSA) design 

features (ADAMS Accession No. ML110130408). Followup discussions were held on August 9, 

2011 (ADAMS Accession No. ML112290891), and again on November 15, 2011 (ADAMS 

Accession No. ML113320197), in which the staff provided further comments on some of the 

broader 10 CFR Part 52 change processes in the draft Appendix C, “Guideline for 

Implementation of Change Control Processes for New Nuclear Plants Licensed Under 10 CFR 

52,” of NEI 96-07. 

During the 2010 workshop, the staff discussed the requirements as codified in Section VIII.B.5.c 

of each appendix to 10 CFR Part 52 regarding the criteria for departures from Tier 2 affecting 

resolution of EVSA design features that would require a license amendment. There was 
A-6 

consensus on the part of staff and stakeholders present that the current rules and guidance may 

not be complete. Certain severe accident features (e.g., features to prevent containment 

bypass and containment hydrogen mitigation equipment such as igniters) do not address 

“ex-vessel” conditions, as defined in the statement of considerations for the rule. 

In the current regulation, changes to severe accident design features that are not specifically 

intended to address EVSAs are not addressed using severe accident criteria as in 

Section VIII.B.5.c. Depending on the nature of the change, the licensee must follow one of the 

other change procedures in the rule, either Section VIII.B.5.a or Section VIII.B.5.b, using the 

guidance in Section 4.4.3.2 of the draft Appendix C to NEI 96-07. If the change falls under the 

requirements of Section VIII.B.5.a, then the licensee could make the change without prior NRC 

approval. Section VIII.B.5.b, however, explicitly excludes changes to severe accident features. 

Moreover, the criteria of Section VIII.B.5.b generally ask if there is “more than a minimal 

increase” in frequency of occurrence or consequences of an accident previously evaluated in 

the plant-specific design control document (DCD). While not stated explicitly in the rule, it is 

clear from the statement of considerations that design-basis accidents are intended here. The 

EVSA criteria in Section VIII.B.5.c use “substantial increase” in probability and public 

consequences as the standard for determining if prior NRC approval is required. The latter 

criteria are less stringent and are more appropriate for non-ex-vessel severe accident design 

features, given the large uncertainties associated with severe accident phenomena and the 

ability of analysts to determine precisely the impact of the change on probability and 

consequences. 

Unless such non-ex-vessel severe accident design features also happen to have a dual 

function, such as also addressing design-basis accidents or aircraft impacts, risk-significant 

Tier 2 changes (e.g., Chapter 19 information related to prevention and mitigation of severe 

accidents other than those considered “ex-vessel”) could be screened out altogether and not 

receive prior NRC approval. The staff has observed that Tier 1 descriptions usually have 

sufficient detail as to necessitate prior staff review for major changes to severe accident design 

features. The staff believes, however, that there is a “gap” in the Tier 2 change process for 

non-ex-vessel design features, in that: a) such changes may be screened out or less 

appropriate criteria are applied when determining whether prior NRC approval is needed, and b) 

whether or not prior NRC approval is obtained may be highly dependent on the degree of detail 

in Tier 1, if any. 

6. Guidance on Risk-Informed Changes to the Licensing Basis, RG 1.174

The staff dedicated a significant portion of the October 5, 2011, public workshop to addressing 

the use of RG 1.174 for new reactors. The staff first discussed the five principles of 

risk-informed decisionmaking, including the defense-in-depth philosophy and the acceptance 

guidelines for change in risk. To test the application of RG 1.174, eight cases representing 

actual or hypothetical changes were presented and the assessment process for determining the 

acceptability of the change was discussed. The staff noted that since it was highly unlikely that 

a combined license (COL) holder would propose a license amendment request to completely 

remove a Tier 1 system, the most likely changes would be about how the existing system is to 

be categorized, operated, and maintained. 
A-7 

The following observations from the tabletop exercise are noted: 

• While a proposed change might have acceptably low change in CDF (ΔCDF) and low 

change in LERF (ΔLERF), if the change adversely impacts equipment that provides 

defense-in-depth capability through redundancy and/or diversity, this could be a cause of 

staff rejection of the proposed change. Thus, a low ΔCDF and low ΔLERF are 

necessary but not sufficient conditions for a change to be considered acceptable by the 

staff. 

• A significant change in risk profile, or if operator action is substituted for an automatic 

function, also would be areas of close review by the staff. 

• Proposed changes in or near the boundary of Region II in the acceptance guidelines 

would undergo close scrutiny by the staff, and there would have to be a compelling 

reason on the part of the license holder for the proposed change. Serious consideration 

of alternatives with lower risk impact would need to be assessed by the licensee. 

The general consensus of the staff and participating stakeholders from the tabletop exercise on 

RG 1.174 is that no substantive changes to the key principles and other guidance to address 

new reactors are necessary. However, the staff plans to augment the existing discussion on 

long-term containment performance in Section 2.2 of RG 1.174 by referring to the containment 

performance objectives for new reactors as described in Commission-approved guidance such 

as SECY-90-016 and SECY-93-087. A detailed discussion of the workshop can be found in the 

meeting summary package (ADAMS Accession No. ML11291A076). 

7. Transition from LRF to LERF

The second topic discussed during the October 5, 2011, workshop was about potential options 

for transitioning from LRF and to a lesser extent on conditional containment failure probability as 

risk metrics used in new reactor design certifications and COL applications, to LERF used in the 

risk-informed regulatory framework for currently operating reactors. 

As discussed in Enclosure 1 to SECY-10-0121, “Modifying the Risk-Informed Regulatory 

Guidance for New Reactors,” the Commission had requested the staff earlier to provide a 

definition of LRF, but in SECY-93-138, “Recommendation on Large Release Definition,” the staff 

recommended, and the Commission approved, termination of work on a definition. As a result, 

the definitions of LRF in the DCDs referenced in COL applications all differ to varying extents. 

Because of the conservative definitions of LRF used in the DCDs, however, the staff has been 

able to review the PRA and severe accident evaluation portions of the DCDs to provide 

reasonable assurance that the Commission’s objectives as described in key policy papers such 

as SECY-90-016 and SECY-93-087 are fully addressed. Additional discussion on the evolution 

of LERF from LRF can be found in the attachment of Enclosure 2 to SECY-10-0121. 

The staff identified three possible options for the transition from LRF to LERF, and discussed 

the advantages and disadvantages of each. While the staff reserved final judgment during the 

workshop on which of three possible options it preferred, the industry representatives clearly 

preferred the option in which the use of LRF would no longer be required for regulatory 

applications once they transitioned to LERF at or before initial fuel load; LERF would be used 
A-8 

going forward during plant commercial operation. The staff has provided options and a 

recommendation to the Commission in this paper to address the issues. 

8. Other Risk-Informed Applications Not in Tabletop Exercises

Other risk-informed initiatives that were not discussed in any depth during the series of tabletop 

exercises were briefly reviewed during the October 5, 2011, workshop. These include: 

• RG 1.175, “An Approach for Plant-Specific, Risk-Informed Decisionmaking: Inservice 

Testing” 

• NEI 94-01, “Industry Guideline for Implementing Performance-Based Option of 

10 CFR 50, Appendix J” 

• RG 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis 

Accidents at Nuclear Power Reactors” 

• Draft Final Rule 10 CFR 50.46a, “Risk-Informed Changes to Loss-of-Coolant Accident 

Technical Requirements” 

• NFPA 806, “Performance-Based Standard for Fire Protection for Advanced Nuclear 

Reactor Electric Generating Plants Change Process” 

Participants in the workshop noted that for most of these activities there appeared to be very 

limited short-term interest by future COL holders, and the staff’s decision not to perform 

tabletops was reasonable. The staff noted that alternative radiological source terms already 

have been applied at most of the new reactor designs. Additionally, all risk-informed changes to 

the licensing basis would fall under the general guidance in RG 1.174. 
 

 Enclosure 2 

Appendix B: ROP Background and Tabletop Exercise Results 

ROP Framework and Risk-Informed Guidance

Some of the key tenets of the Reactor Oversight Process (ROP) and the drivers in its 

development were (1) to improve the objectivity of the oversight processes so that subjective 

decisionmaking is minimized, (2) to improve the scrutability and predictability of U.S. Nuclear 

Regulatory Commission (NRC) actions so that regulatory response has a clear tie to licensee 

performance, and (3) to risk-inform the processes so that NRC and licensee resources are 

focused on performance issues with the greatest impact on safe plant operation. Consistent 

with RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed 

Decisions on Plant-Specific Changes to the Licensing Basis,” the ROP’s risk-informed 

processes integrate risk insights with more traditional deterministic factors (such as defense in 

depth and maintaining safety margins) to guide regulatory decisionmaking. 

The regulatory framework for reactor oversight consists of three key strategic performance 

areas: reactor safety, radiation safety, and safeguards. Within each strategic performance area 

are cornerstones that reflect the essential safety aspects of facility operation. These seven 

cornerstones include: initiating events, mitigating systems, barrier integrity, emergency 

preparedness, public radiation safety, occupational radiation safety, and security. Satisfactory 

licensee performance in the cornerstones provides reasonable assurance of safe facility 

operation and that the NRC’s safety mission is being accomplished. Each cornerstone contains 

inspection procedures and performance indicators (PIs) to ensure that their objectives are being 

met. Both inspection findings and PIs are evaluated and given a color designation based on 

their safety significance, and this designation feeds the ROP Action Matrix to determine a 

predictable regulatory response. 

Significance determination process (SDP) implementation guidance is contained in Inspection 

Manual Chapter (IMC) 0609. IMC 0609, Appendix A, “Determining the Significance of Reactor 

Inspection Findings for At-Power Situations,” is used to determine risk-significance of 

performance deficiencies in the initiating events, mitigating systems, and barrier integrity 

cornerstones. Risk thresholds are a function of changes in core damage frequency (CDF) and 

large early release frequency (LERF) against a plant’s baseline risk. For those relatively 

infrequent cases when sufficient probabilistic risk assessment (PRA) methods and tools are not 

available or appropriate to provide reasonable and timely estimates of safety significance, the 

staff uses IMC 0609 Appendix M, “Significance Determination Process Using Qualitative 

Criteria,” and considers the best available information and factors such as defense in depth, 

safety margins, and the potential for plant-wide impacts attributable to the performance 

deficiency to determine the safety significance in those cases. Several additional SDPs are 

more subjective to determine an equivalent regulatory response (e.g., emergency 

preparedness, radiation safety, security). 

Implementation guidance for the PI program, including but not limited to migrating systems 

performance index (MSPI), is contained in IMC 0608 “Performance Indication Program.” More 

detailed guidance on the data collection and PI calculations are contained in NEI 99-02, 

“Regulatory Assessment Performance Indicator Guideline,” which is jointly produced and 

maintained by Nuclear Energy Institute (NEI) and the NRC. The MSPI covers five systems 

important to safety, and tracks the unavailability of monitored trains and the unreliability of 
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monitored components. The MSPI calculation reflects the deviation of a specific unit’s 

performance from an industry baseline, converted to a simplified ΔCDF. A performance limit, or 

deterministic backstop, is also used for determining degraded performance. 

The implementation guidance for NRC’s response to events is contained in Management 

Directive (MD) 8.3, “NRC Incident Investigation Program,” and supplemented by IMC 0309, 

“Reactive Inspection Decision Basis for Reactors.” Deterministic criteria are used for initial 

event screening, and a range of risk thresholds are subsequently applied to determine if a 

reactive inspection will be launched. The risk-informed reactive inspection thresholds are a 

function of conditional core damage probability (CCDP) and conditional large early release 

probability (CLERP). An overlap of options provides flexibility based on uncertainty and 

deterministic insights, and additional deterministic criteria are reviewed and documented as the 

basis for staff decisions on the appropriate regulatory response. While these ranges offer some 

flexibility for determining the level of event response, they still involve thresholds based on risk 

values. 

ROP Tabletop Discussions and Observations

The staff held a half-day public workshop to discuss the ROP-related tabletops on the afternoon 

of October 5, 2011 (ADAMS Accession No. ML11291A076). A brief followup public meeting 

was held on October 26, 2011 (ADAMS Accession No. ML11308A542). The ROP tabletops 

tested various scenarios that are or will be relevant to the licensing basis for new reactors to 

confirm the adequacy of the current ROP risk-informed processes for regulatory decisionmaking 

or to identify areas for improvement. In preparation for the ROP tabletops, the staff developed a 

broad cross-section of well-vetted cases from actual greater-than-green SDP findings, MSPI 

data, and event response (in the risk-informed reactor safety cornerstones of initiating events, 

mitigating systems, and barrier integrity) from the current fleet of reactors. For each case study, 

the staff applied similar situations to the new reactor designs, filling in any gaps with realistic 

hypothetical situations and reasonable assumptions, and then compared the risk values and 

resulting regulatory responses from the new reactor scenarios to those derived from the current 

fleet. A complete summary of the ROP tabletop examples and results was made publicly 

available (ADAMS Accession No. ML11308A354). 

As a result of the ROP tabletops, the staff noted the following observations and potential 

process improvements. 

• SDP: Although less likely (and less frequently), the case studies demonstrated that an 

increased regulatory response can be triggered based on surpassing greater-than-green 

risk-thresholds for inspection findings, but it would likely take common-cause failures 

that affect multiple systems and/or long-term exposures of risk-significant components. 

In addition, the analyses for the current fleet often are influenced by uncertainties and/or 

sensitivities of critical parameters that influence the numerical results, and this is 

expected to continue to be true for future reactors. The SDP analyses for new reactor 

designs may need to be augmented with additional qualitative considerations to provide 

a deterministic backstop to appropriately respond to performance issues. 
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• Event Response: Although less likely (and less frequently), the case studies 

demonstrated that the numeric thresholds for invoking reactive inspections, including 

augmented inspection teams (AITs), can be met. In the current guidance and practice, 

deterministic criteria are used for initial event screening, and ranges of risk values are 

subsequently applied to determine the level of reactive inspection. 

While these ranges offer some flexibility in determining the regulatory response, they 

involve thresholds based on risk values that significantly influence regulatory outcomes 

governing whether or not a reactive inspection is warranted and, if so, at what level. The 

tabletop exercises also revealed that an insufficient consideration of deterministic factors 

(such as defense in depth and barrier integrity) could result in an inadequate regulatory 

response. 

Therefore, the contribution of the existing deterministic criteria may need to be modified, 

or new deterministic criteria or backstops may need to be developed, to appropriately 

respond to plant events. 

• MSPI: The cases indicated that it would be rare and unlikely to cross greater-than-green 

MSPI thresholds for active new reactor designs. The performance limit (deterministic 

backstop) would play a more significant role and could be emphasized for the new 

reactor MSPI. Passive designs are too different to evaluate at this time and an MSPI 

may not be possible for passive systems without significantly altering the fundamental 

methodology in NEI 99-02 used for active safety systems. Given the anticipated low 

utility of this indicator for new reactor designs, it may be of limited value for licensees to 

create an MSPI basis document and track and report MSPI data. Alternate mitigating 

system performance indicators could be developed, and/or additional inspection could 

be used to supplement or complement insights currently gained through MSPI. 

The staff also noted that several current regulatory and programmatic controls exist and can be 

leveraged as necessary, including: 

• The ROP self-assessment process could be used to evaluate and potentially adjust the 

ROP for new reactors in the future as a result of additional experience and lessons 

learned. 

• All performance deficiencies (including green) are entered into the licensee’s corrective 

action program, and receive attention by licensees and the NRC. They would also be 

considered for cross-cutting aspects in accordance with the current process. 

• Deviations from the ROP Action Matrix could also be used to adjust the staff’s actions to 

provide for an appropriate regulatory response, if deemed necessary, and then each 

deviation would be evaluated for potential program improvements. 
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The staff also identified the following limitations in the scope of its tabletop analysis: 

• The best available data were used to perform the analyses for the tabletop exercises. 

The standardized plant analysis risk (SPAR) models and new reactor vendors’ risk 

models used for these case studies are still being refined and reviewed for quality and 

accuracy, and any future changes could potentially affect the risk values. 

• Only limited consideration was given to passive systems and components for these case 

studies. Potential passive design issues may need to be taken into account for the new 

reactor designs in the future. 

• The ROP tabletops were limited to the risk-informed reactor safety cornerstones of the 

ROP: initiating events, mitigating systems, and barrier integrity. The more deterministic 

cornerstones of emergency preparedness, public radiation safety, occupational radiation 

safety, and security were not directly addressed. 

Staff Conclusions from ROP Tabletop Exercises

Within the ROP, the SDP is used to characterize inspection findings. All inspection findings 

require a performance deficiency, the vast majority of which are based on violations. Where the 

licensing basis, including the TS completion times and minimum equipment configuration 

requirements, allow greater flexibility in operation, that flexibility is not encroached upon by the 

ROP’s inspection findings. The staff concluded that the existing risk-informed SDP is generally 

acceptable, and could generate an increased regulatory response based on greater-than-green 

results. The case studies, however, demonstrated that the performance deficiencies would 

likely have to involve common cause failures that affect multiple systems and/or involve longterm exposures of risk-significant components. In addition, the case study on reactor coolant 

system integrity demonstrates that the existing quantitative process does not provide the 

appropriate response for degradation of passive components and barriers. The intent of the 

ROP is to generally capture declining licensee performance by progression through the action 

matrix’s columns. The staff believes that this would not be the case for passive plants, and 

there would be little NRC intervention above the baseline inspection program before an event 

with actual consequences. 

More specifically, a case study involving reactor vessel head degradation resulted in a 

marginally white finding for the AP1000 and a green finding for the U.S. Advanced Pressurized 

Water Reactor (US-APWR) for medium break and large break loss-of-coolant accidents based 

on the risk numbers. The calculations included increasing the initiating event frequencies by 

two orders of magnitude per the SDP guidelines to address the significant unanalyzed 

parameters, as was done for a similar case at Davis-Besse in 2002. Based on this calculation, 

the resultant regulatory response would be to move the AP1000 facility to the Regulatory 

Response Column (column 2) of the ROP Action Matrix and perform a single supplemental 

inspection of about 40 hours to ensure the causes of the performance issues are identified and 

that appropriate corrective actions are planned or taken to prevent recurrence. In this case, the 

staff believes that a more robust and diagnostic supplemental inspection would be warranted to 

ensure that the NRC and licensees identify and arrest root causes that led to the degradation in 
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the integrity of barriers designed to protect the public from radionuclide releases caused by 

potential accidents. Unless the deterministic criteria and how they are applied are enhanced, 

this increased regulatory response would not be attainable using the current risk-informed SDP 

and a deviation from the ROP Action Matrix would need to be pursued to ensure an appropriate 

regulatory response. 

To address the identified shortfall, the staff determined that the SDP analyses for new reactor 

designs should be augmented with additional qualitative considerations, consistent with the 

integrated risk-informed decision-making framework in RG 1.174, to provide a deterministic 

backstop as necessary to appropriately address performance issues. For example, 

deterministic backstops could be developed to reinforce the importance of maintaining barrier 

integrity for fuel cladding, reactor coolant system pressure boundary, and containment. For 

active safety system plants, the staff could also explore the feasibility of implementing a 

deterministic backstop for equipment outages resulting from degraded conditions (similar to the 

RITS 4b backstop completion time) and/or for the number of repetitive equipment failures that 

could degrade the reliability or availability of structures, systems and components (SSCs) from 

performing their intended safety functions. These deterministic backstops should not infringe 

upon the operational flexibility afforded by the more robust new reactor designs, but should 

instead be designed to capture the infrequent yet potentially significant performance issues that 

would not otherwise be captured by the risk calculations to ensure an appropriate regulatory 

response. 

MD 8.3 and IMC 0309 establish the criteria for event response. Deterministic criteria are used 

for initial event screening, and risk thresholds are subsequently applied to determine if a 

reactive inspection will be launched. The threshold for launching the lowest level of reactive 

inspection, a special inspection team (SIT) is clearly defined as an estimated CCDP of 10-6. 

Similarly, ranges of risk values are specified for launching AIT and incident investigation team 

inspections. While these ranges offer some flexibility for determining the level of event 

response, they still involve thresholds based on risk values. As such, risk values significantly 

influence regulatory outcomes governing whether or not a reactive inspection is warranted and, 

if so, at what level. 

The specific tabletop scenarios yielded mixed results: while most would have resulted in an 

appropriate response to events, the staff noted that one outcome would have been different 

depending on the risk model used for determining the level of reactive inspection. 

Nevertheless, the scenarios indicated that risk thresholds are adequate for new reactors. The 

staff, however, recognized the potential for different responses as a function of variations in or 

minor revisions to the risk models used in the calculations. In addition, since MD 8.3 

determinations are largely influenced by risk information, the staff acknowledged the potential 

for inadequate response based on the results of the tabletop scenarios used in the SDP and 

MSPI applications. For these reasons, the staff determined the need to better balance the risk 

information with other deterministic considerations such that an appropriate level of regulatory 

response is achieved for new reactors. The staff concluded that the contribution of the existing 

deterministic criteria may need to be modified, or new deterministic criteria could be developed, 

for initiating a reactive inspection to assess events or degraded conditions at new reactor 

facilities. Similar backstops to those discussed above for the SDP could be explored to ensure 
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regulatory response to plant events and conditions is appropriate and commensurate with the 

significance of the event (or condition). 

For example, consider the case of a steam generator tube rupture at an AP1000 facility. Before 

the case study, the staff used the submitted Westinghouse PRA to calculate that only an SIT 

would be performed in accordance with the existing MD and IMC guidance. 

During the case study, the staff used the slightly more conservative SPAR model to calculate 

results that were just over the threshold where either a SIT or an AIT would be considered. The 

staff believes that performance of an AIT is appropriate for any steam generator tube rupture 

(an event that involves the breach of the reactor coolant system pressure boundary and 

potentially bypasses containment), and an insufficient consideration of deterministic factors 

(such as defense in depth and barrier integrity) could result in an inadequate regulatory 

response. 

Finally, the staff concluded that the existing MSPI would not be adequate and would be largely 

ineffective in providing meaningful input to the risk-informed regulatory decision-making 

process. There were numerous case studies that demonstrated this shortfall. The case studies 

demonstrated that it would be extremely rare to cross greater-than-green MSPI thresholds that 

would result in an increased regulatory response for active new reactor designs, and a 

meaningful MSPI may not even be possible for passive systems using the current formulation of 

the indicator. The existing performance limit approach, which incorporates a backstop that 

indicates when the performance of a monitored component in an MSPI system is significantly 

lower than expected industry performance, would play a more significant role and could 

potentially be emphasized and modified for the active new reactor designs. 

For example, one case study identified that it would take greater than 25 emergency diesel 

generator (EDG) start failures or greater than 25 EDG run failures for the U.S. Evolutionary 

Power Reactor (U. S. EPR) to exceed the green-white risk threshold, and 12 failures to reach 

the performance limit. In another case, it would take 14 or more turbine-driven emergency 

feedwater pump failures or greater than 25 motor-driven pump failures for the US-APWR to 

exceed the green-white threshold using the licensee’s PRA model, and the performance limit 

would not be exceeded until 6 or more pump failures in a 3-year timeframe occurred. Taken as 

a whole, it would be highly improbable to have a greater-than-green indicator for any MSPI 

system at any of the new reactor facilities, even taking into account the performance limit 

(backstop) as it is currently formulated. Therefore, the existing MSPI would provide little if any 

insight into plant performance for new reactors and would not trigger an appropriate regulatory 

response to address performance issues. The staff determined that alternate performance 

indicators (PIs) in the mitigating systems cornerstone could potentially be developed and/or 

additional inspection could be used to supplement insights currently gained through MSPI. 

Potential Changes to ROP Guidance Documents to Address Shortfalls

The staff could revise existing guidance to address the noted gaps in the risk-informed SDP, 

event response, and MSPI applications for new reactors identified by participants during the 

ROP tabletop exercises. 
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For example, the staff could revise relevant portions of IMC 0609 and its appendices to reflect 

the additional deterministic criteria for use in the SDP, which could involve revising Appendix A, 

Appendix M, and/or creating an additional appendix as necessary. The staff could also revise 

IMC 0309, and potentially MD 8.3, to add specific deterministic criteria and/or clarify the 

guidance to promote a more holistic and integrated approach to event response based on both 

risk-informed and deterministic factors. The staff could also revise the PI guidance in IMC 0608 

and/or inspection guidance in IMC 2515, “Light-Water Reactor Inspection Program – Operations 

Phase,” and work with industry to revise NEI 99-02 to reflect the program revisions necessary to 

compensate for the loss of MSPI insights.