Technical Advisory Committee of the Nuclear Risk Research Center
Central Research Institute of Electric Power Industry
1-6-1 Otemachi, Chiyoda-ku, Tokyo, 100-8126 JapanMay 24, 2023
Dr. George Apostolakis
Director, Nuclear Risk Research Center
Central Research Institute of Electric Power Industry
1-6-1 Otemachi, Chiyoda-ku
Tokyo, 100-8126 JapanSUBJECT:

RISK-INFORMED CHANGES TO CONTAINMENT VESSEL LEAK
RATE TESTING INTERVAL

Dear Dr. Apostolakis:
During the 18th meeting of the Technical Advisory Committee of the Nuclear Risk
Research Center (NRRC), May 15-19, 2023, we met with representatives of theNRRC staff to discuss a proposed methodology to evaluate the risk from changes tothe containment vessel integrated leak rate testing intervals at Japanese nuclear
power plants.
CONCLUSIONS AND RECOMMENDATIONS
1. The proposed methodology and simplified computation framework are very useful
because they can be applied by every Japanese plant now, without the need todevelop a detailed Level 2 probabilistic risk assessment (PRA).
2. For this risk-informed application, the change in containment failure frequency(ΔCFF) provides an appropriately conservative bound for the change in large
early release frequency (ΔLERF).
3. The change in risk can be appropriately approximated using the core damage
frequency (CDF) derived from the full-power PRA model for internal initiating
events, when such an estimate is supplemented by considerations of additional
contributors such as internal hazards, external events, and events which canoccur during low power and shutdown modes when containment integrity must bemaintained.
4. The total industry exposure time should not be used to derive estimates for thecontainment standby failure rate.
5. The proposed methodology should contain guidance for quantification ofuncertainties in the conditional containment failure probability (CCFP). Those-1-uncertainties should be combined with the CDF uncertainties and supplemented,
as needed, by additional quantitative or qualitative assessments to evaluate theoverall uncertainties in the analysis results.
BACKGROUND
The NRRC is developing methods and guidance that can be used by Japaneseutilities to justify extensions to their containment leak rate testing intervals, while
maintaining a high level of overall plant safety. Operating nuclear power plants inthe United States have used risk-informed, performance-based methodologies tomeet the requirements described in Option B of 10 CFR 50, Appendix J to extend
their Type A containment integrated leak rate testing intervals to 15 years.
U.S. Nuclear Regulatory Commission Regulatory Guide 1.174 describes thefundamental principles of a risk-informed integrated decision-making process. It also
provides guidance for the types of information and supporting analyses which areneeded to implement that process. Regulatory Guide 1.163 contains additional
guidance that is focused specifically on implementation of a risk-informed,performance-based containment leak rate testing program. The current draft
Revision 1 to Regulatory Guide 1.163 endorses the analytical methods that aredescribed in Nuclear Energy Institute report NEI 94-01, Revision 3-A and ElectricPower Research Institute report EPRI 1018243.
DISCUSSION
Risk-informed, performance-based changes to a plant's licensing basis should bejustified by integrated decisions which implement all five of the fundamental
principles that are described in Regulatory Guide 1.174. Our discussions with theNRRC staff about the proposed methodology were focused primarily on the analyses
which would be performed to support a conclusion that the change in overall plantrisk is acceptably small. We were not briefed on details of the additional guidance
and methods that will be used to support the other four elements of the integrated
decision process and the plant-specific licensing submittal.
Therefore, ourconclusions and recommendations in this letter report pertain only to the proposedrisk evaluation methodology.
Regulatory Guide 1.174 provides guidelines for acceptable changes to the coredamage frequency (ΔCDF) and the large early release frequency (ΔLERF) which
may result from a proposed change to a plant's licensing basis. Every Japanese
nuclear power plant has a Level 1 probabilistic risk assessment (PRA) that quantifies
the CDF. Those PRAs are currently developed to varying levels of completenessand technical quality. Some plants have Level 2 PRAs that can be used to estimate
the LERF. However, there is more variability in the scope and fidelity of those PRAs,
compared to the Level 1 PRAs.
The proposed methodology is tailored to the current status of the Japanese industry
PRAs. In particular, it does not require explicit quantification of the baseline LERF
that is associated with the existing containment leak rate testing interval or the LERF-2-change which results from an increase in that interval. It relies rather on an estimateof the change in the total conditional containment failure probability (ΔCCFP) due toall conditions which are identified by the Type A integrated leak rate tests. Thechange in the containment failure frequency (ΔCFF) is then used as a surrogate
metric for ΔLERF. It is quantified by:
ΔCFF = CDF * ΔCCFP
The baseline CCFP and its change are derived from a time-based incipientdegradation model which uses containment testing experience data. The methods
and guidance in NEI 94-01, Revision 3-A and EPRI 1018243 use a conceptuallysimilar standby failure rate model to estimate the change in LERF. However, incontrast to much of the detailed guidance for the U.S. plant submittals, the NRRC
simplified computation framework is very useful because it can be applied by every
Japanese plant now, without the need to develop a detailed Level 2 PRA.
The proposed methodology quantifies a conservative bound for the actual ΔLERF.
The analysis accounts for the total probability of all containment failure conditions
which result in a release of any size that exceeds the Type A testing acceptancecriteria. Use of those leakage criteria in these analyses ensures that the actual
LERF will, at most, be equal to the CFF. In practice, experience from full-scope
Level 2 PRAs and numerous integrated leak rate tests shows that the LERF is
typically much smaller than the CFF, because most containment flaws will not resultin a large release. Therefore, use of ΔCFF in the proposed computationalframework provides an appropriately conservative bound for the ΔLERF that is
associated with an increased leak rate testing interval.
Evaluation of Core Damage Frequency
The guidance in Regulatory Guide 1.174 emphasizes that risk-informed decisions
should account for how the overall plant risk is affected by the proposed change.
This requirement has two implications for the scope of the Level 1 PRA models andother supporting analyses that are used to estimate the CDF.
The first implication is that the analyses should account for the total CDF from all
initiating events (i.e., internal events, internal hazards, and external events) which
may occur during all reactor operating modes when containment integrity must bemaintained (i.e., full power, low power, and several shutdown plant operating states).
If the scope of the available plant-specific PRA models does not include all of thesecontributions, the licensing submittal must describe how the missing elements are
considered and evaluated to reach the risk-informed conclusion. In some cases, it
may be possible to use appropriately conservative quantitative estimates for somemissing contributors. In other cases, combinations of quantitative and qualitative
justifications may suffice. In practice, these issues of PRA completeness are often
addressed by an integrated decision-making panel of experts who are familiar withthe available plant-specific PRA, its results, risk assessments for similar plants, andanalytical techniques that are most effective to supplement the lacking information.
A second possible implication regarding the CDF evaluation may apply for specificplants. The licensing basis accident analyses for some U.S. plants include credit for-3-

pressure in the containment (so-called "containment overpressure" or "containment
accident pressure") to maintain adequate net positive suction head (NPSH) forpumps that can be aligned to the containment sump. For those plants, the analyses
should evaluate how the extended testing interval affects the probability ofcontainment leakage conditions which can functionally disable those pumps when
they are needed for core recirculation cooling. The risk-informed decision must thenaccount for how the proposed change affects ΔCDF, as well as ΔLERF. This issue
is addressed in the guidance in draft Revision 1 of Regulatory Guide 1.163 and EPRI
1018243.
Estimation of Containment Standby Failure Rate
The proposed methodology relies on the use of a standby failure rate model toestimate the ΔCCFP due to the increased testing interval. This type of model isendorsed by the methods and guidance in NEI 94-01, Revision 3-A and EPRI
1018243, and it has been used in the U.S. plant analyses.
The proposed NRRC formulation of the quantification method inappropriately uses
the total industry exposure time to derive an estimate for the containment standby
failure rate. This process results in a numerical estimate which is smaller than thefailure rate which can be justified by the available operating experience. When that
failure rate is applied over the extended testing interval, the calculation will
underestimate the ΔCCFP.
These analyses use a time-based standby failure rate model for the accumulation of
incipient failures and testing restoration. Therefore, it is essential that the supporting
data must correctly account for the testing intervals during which the relevant
operating experience was compiled. This is a subtle, but very important issue. It isbest illustrated conceptually by a simple example.
Suppose that the available operating experience includes 10 units. Each unit hasoperated for 10 years. During that 10-year period, each unit has performed four
Type A leak rate tests. Suppose also that no failures have occurred during any of
those tests. Thus, the available industry experience is that no failures have occurredin a total of 40 tests, which were performed with an average interval of 2.5 years
between tests. Application of the proposed methodology would incorrectly combine
the total industry operating time to infer that no failures have occurred in 100
consecutive years. The standby failure rate that is derived from that 100-year
exposure period is smaller than the failure rate which would be derived from theactual 2.5-year testing interval data.
The total industry exposure time should not be used to derive estimates for thecontainment standby failure rate. Other methods should be used to derive thebaseline CCFP and the appropriate containment standby failure rate. The guidancein EPRI 1018243 summarizes a calculation method that has been used for the U.S.
plant analyses.

-4-Evaluation of Uncertainties
The guidance in Regulatory Guide 1.174 emphasizes that risk-informed decisions
should account for quantitative and qualitative uncertainties in the supporting
analyses and results. It specifically refers to the extensive guidance on identificationand treatment of uncertainties that is provided in NUREG-1855.
In practice, the available Level 1 PRA models that are used to quantify the plantspecific CDF should quantify the corresponding uncertainty in that parameter.
There are also uncertainties in the CCFP. The proposed methodology should
contain guidance for quantification of those uncertainties. The CCFP uncertainties
should be combined with the CDF uncertainties to quantify the composite uncertainty
in the resulting estimate for ΔCFF.
In practice, it may also be necessary to supplement these explicitly quantified
uncertainties by additional quantitative or qualitative assessments of other sources ofuncertainty. For example, the integrated decision-making expert panel may evaluate
the effects from uncertainties associated with their assessments of the completenessand technical quality of the applied PRA models.
We look forward to continuing our review of this important NRRC project and its
benchmark application for practical and effective risk-informed decision-making.
Sincerely,John W. Stetkar
Chairman
REFERENCES
1.

Nuclear Risk Research Center, "Draft Guidelines for Containment Vessel
Leak Rate Test," Presentation to NRRC Technical Advisory Committee, May16, 2023, Proprietary.

2.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, "An Approachfor Using Probabilistic Risk Assessment in Risk-Informed Decisions on PlantSpecific Changes to the Licensing Basis," Revision 3, January 2018.

3.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.163,"Performance-Based Containment Leak-Test Program," Proposed Revision 1,Draft Regulatory Guide DG-1391, November 2022.

-5-4.

Nuclear Energy Institute, NEI 94-01, "Industry Guideline for Implementing
Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3-A,July 2012.

5.

Electric Power Research Institute, EPRI 1018243, "Risk Impact Assessmentof Extended Integrated Leak Rate Testing Intervals," (EPRI 1009325,Revision 2-A), October 2008.

6.

U.S. Nuclear Regulatory Commission, NUREG-1855, Revision 1, "Guidanceon the Treatment of Uncertainties Associated with PRAs in Risk-Informed
Decisionmaking", Revision 1, March 2017.

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