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 Japan

February 16, 2021

Dr. George Apostolakis
Head, Nuclear Risk Research Center
Central Research Institute of Electric Power Industry
1-6-1 Otemachi, Chiyoda-ku
Tokyo, 100-8126 JapanSUBJECT:PROPOSED NRRC RESEARCH ON SELECTED SEISMIC ISSUESFOR FISCAL YEAR 2021

Dear Dr. Apostolakis:
This is a companion report to our February 16, 2021 letter report on "Proposed
NRRC Research Plan for Fiscal Year 2021." It provides more detailed discussions
and recommendations for two specific areas of the Nuclear Risk Research Center(NRRC) seismic research program:(1) Development of a guide for probabilistic seismic hazard analysis (PSHA),
including the specific plan for applying the Senior Seismic Hazard Analysis
Committee (SSHAC) methodology-based implementation to a region; and(2) Development of a simplified elasto-plastic assessment method for evaluating
piping system fragilities, based on experimental and analytical approaches.

CONCLUSIONS AND RECOMMENDATIONS
1. To establish technical adequacy, meet the applicable standards, and reflect theinternational state-of-practice, the PSHA guidance and a regional approach forevaluating seismic hazards should include activities to demonstrate how toincorporate local site effects and site response into a site-specific PSHA.
2. As part of its fragility analysis research, NRRC should consider developing
selection and screening criteria for piping systems. Those analyses should
examine various direct and indirect failure modes to determine which areimportant, considering plant-specific PRA success criteria, past studies, andexperience data. We agree with the NRRC that a simplified approach would bedesirable for practical applications.
3. We would like to be briefed on these hazard and fragility assessment research
activities as more progress is made and detailed plans are available.

-1-

BACKGROUND
Some of the most important issues related to the restart of Japanese nuclear power
plants are related to the seismic hazard and seismic response. These seismic
issues, particularly those related to the seismic hazard, have played a very importantrole for safety decisions at the national regulatory level, all levels of prefecture andlocal governments, and in the courts. The NRRC seismic research program has
produced very useful methods and analytical tools to address these issues. The
effort related to implementation of the SSHAC methodology for the Ikata site has
been a critical step in advancing the state-of-practice for seismic hazard analysis inJapan. The planned work in FY2021 for developing a PSHA guide and the specificplan for applying the SSHAC methodology-based implementation to a region is avery important next step.
Similarly, NRRC's research related to the development of realistic seismic fragilities
has addressed and advanced the state-of-practice in several important areas.
Examples include ground and slope failure fragilities and realistic seismic responses.
One activity in the seismic fragility research plan for FY2021 is an advanced
methodology for the evaluation of piping systems.
We have not had opportunities to discuss the planned work with the research teams
in any detail, partly because of the ongoing pandemic. Considering the importanceof these elements of a seismic PRA, our comments are intended to provide aperspective and some insights into these two subject areas that may help in
determining the scope of the planned research and defining specific activities.

DISCUSSION
SSHAC Methodology Applications
The completion of the Ikata SSHAC project is a very significant milestone. The next
step of developing PSHA guidance and implementation of a region-based approachis needed for application of the methodology to other sites. The NRRC is also
exploring implementation of several seismic hazard methodology topics through amodel plant analysis.
There are three major elements in a PSHA methodology: (1) characterization ofseismic sources, (2) propagation of seismic energy to the site, and (3) local site
response. The Ikata SSHAC project did not have to deal explicitly with the third
element, local site response, because the Ikata plant foundation is on hard rock. To
address this issue, in our November 27, 2016 letter report, we recommended:"Before further adaptation of the SSHAC process, the NRRC should consider a
SSHAC project for one more site which has a seismo-tectonic environment withsome features that are different from the Ikata site. This will provide more robust
insights for developing guidance and pertinent research to facilitate
implementation of the SSHAC process in Japan."-2-One of the reasons for this recommendation was:"For example, a soft site that requires consideration of local subsurface featuresand site response evaluations may provide additional insights into technicalissues, uncertainties, and development of data, methods, and models."
This is an important issue. The NRRC research team has noted that some BWR
plants are located on soft rock sites and have some unique site-specific issues,compared to PWR plants, most of which are located on hard-rock sites.
The state-of-the-art and practices for evaluating local site response are evolving. It
is instructive to look at the Japanese and U.S. seismic PRA standards and guidancewith respect to this issue. Examples of relevant elements of the standards andcontemporary guidance for evaluating local site response are summarized inAppendix A to provide further context for our recommendation.
The preceding discussion and the references in Appendix A provide a perspective on
the importance of local site effects. We understand that for a variety of reasons, theNRRC research team has not been able to implement a full-scope SSHAC
procedure at a softer site, as we recommended in November 2016. The activities
that are summarized in the FY2021 seismic hazard research plan do not explicitly
address the evaluation of local site effects in the planned SSHAC guidancedevelopment, implementation of a regional model, or in the hazard studies related tothe model plant PRA demonstration. The scope of the seismic hazard researchshould include a specific task for development and demonstration of a methodology
to include local site effects for successful application at all sites, to meet thestandards and codes requirements, and to reflect the international state-of-practice.
Piping Fragility Research
The seismic fragility research activities for FY2021 include development of asimplified elasto-plastic assessment method for piping systems, based on
experimental and analytical approaches. The research team noted that they intend
to adopt fatigue as a piping system failure criterion, instead of the guillotine break(rupture) criterion that has been used conventionally. They explained that the piping
fragility is traditionally evaluated using an elastic analysis, with a guillotine break set
as the failure mode. The research team indicated that this methodology isconservative, and it does not account for the realistic piping system behavior. They
also noted that consideration of the piping system supports remains a key issue forthe fragility evaluation. The team plans to investigate fatigue as a failure mode forone key piping system in FY2021. The NRRC is also conducting shaking table tests
to support this activity.
It is also instructive to look at the Japanese seismic PRA standards and U.S.
guidance with respect to this issue. Examples of relevant elements of the standards
and an engineering approach for evaluating piping failures are summarized inAppendix B to provide further context for our recommendation.

-3-

The piping system failure modes of concern have been generally related to anchor
motions or the existence of some sort of prior degradation. Many tests of piping
components and piping systems have been conducted over the years. For example,
NUPEC and JNES conducted piping tests with and without any degradation. In the
U.S., EPRI and NRC conducted similar tests in the 1990s. These testing programs
provide support for the general conclusion that a seismically-caused guillotine breakof un-degraded piping has very low probability for most nuclear power plant systems.
Incorporation of specific failure modes for any spatially distributed system, like piping
systems, in a seismic PRA tends to be a problematic issue for a variety of reasons.
Therefore, it is very important to identify what are the vulnerable systems. Thosesystems depend on details of the piping designs, configurations, and the plantspecific PRA models and success criteria. For example, in principle, the fragility
analyses should quantify the conditional probability that the seismic damage toindividual or multiple piping sections has the following effects:







Very small, small, medium, or large LOCA
Small or large steam line break
Functionally disable one or more trains of high pressure or low pressure injection
Functionally disable one or more trains of open- or closed-loop cooling water
Flood one or more plant compartments
Other relevant effects for the specific plant design and risk models

As part of its fragility analysis research, NRRC should consider developing selection
and screening criteria for piping systems. Those analyses should examine various
direct and indirect failure modes to determine which are important, considering plantspecific PRA success criteria, past studies, and experience data. We agree with the
NRRC that a simplified approach would be desirable for practical applications.

We would like to be briefed on these hazard and fragility assessment research
activities as more progress is made and detailed plans are available.

Sincerely,

John W. Stetkar
Chairman

REFERENCES
1.

"NRRC Overview: Research Program for FY2021, External Natural Events,"
Presentation to NRRC Technical Advisory Committee, November 4, 2020,
Proprietary

-4-

2.

Technical Advisory Committee individual members' comments and questions
on "NRRC Overview: Research Program for FY2021, External Natural
Events," December 4, 2020, Confidential.

3.

NRRC Research Team responses to comments and questions on "NRRC
Overview: Research Program for FY2021, External Natural Events," January
30, 2021, Confidential.

4.

Technical Advisory Committee of the Nuclear Risk Research Center,"Proposed NRRC Research Plan for Fiscal Year 2021," February 16, 2021.

5.

Technical Advisory Committee of the Nuclear Risk Research Center, "Interim
Report on Probabilistic Seismic Hazard Analysis Enhancements in Japan andFault Displacement Evaluation," November 27, 2016.

6.

Atomic Energy Society of Japan, "A Standard for Procedure of Seismic
Probabilistic Risk Assessment for Nuclear Power Plants: 2015," AESJ-SCP006E:2015, December 2015, English translation.

7.

American Nuclear Society, "American National Standard, Probabilistic
Seismic Hazard Analysis," ANSI/ANS-2.29-2020, April 2020.

8.

U.S. Nuclear Regulatory Commission, "Updated Implementation Guidelines
for SSHAC Hazard Studies," NUREG-2213, October 2018 (ML18282A082).

9.

U.S. Nuclear Regulatory Commission, "Central and Eastern United States
Seismic Source Characterization for Nuclear Facilities," NUREG-2115,
Vols. 1-6, January 2012 (ML12048A776).

10.

U.S. Nuclear Regulatory Commission, "Seismic Considerations for theTransition Break Size," NUREG-1903, February 2008 (ML080880140).

11.

U.S. Nuclear Regulatory Commission, "Probability of Pipe Failure in theReactor Coolant Loops of Westinghouse PWR Plants," NUREG/CR-3660,
UCID-19988, Vols. 1-4, 1985.

12.

U.S. Nuclear Regulatory Commission, "Probability of Pipe Failure in theReactor Coolant Loops of Combustion Engineering PWR Plants,"
NUREG/CR-3663, UCRL-53500, Vols. 1-3, 1985.

13.

U.S. Nuclear Regulatory Commission, "Probability of Pipe Failure in theReactor Coolant Loops of Babcock and Wilcox PWR Plants,"
NUREG/CR-4290, UCRL-53644, Vols. 1-3, 1985.

14.

U.S. Nuclear Regulatory Commission, "Probability of Failure in BWR Reactor
Coolant Piping," NUREG/CR-4792, UCID-20914, Vols. 1-4, 1988.

-5-

APPENDIX AREFERENCE STANDARDS AND TECHNICAL GUIDANCE FOREVALUATION OF LOCAL SITE RESPONSE

Japanese Seismic PRA Standard AESJ-SC-P006E:2015
Section 6.4.2 of the Standard addresses the seismic ground motion propagation
characteristics in the site vicinity and around the unit of interest. It states, in part:"In characterizing the seismic ground-motion propagation, first analyze theseismic-motion propagation characteristics in the site vicinity and around the unitof interest at the site in order to understand whether local uniqueness of those
propagation characteristics exists".
The Standard contains guidance for various aspects of this issue. For example,
Annex AP, "Difference in Ground-Motion Propagation Characteristics within a Same
Site," apparently expands on this discussion. However, we did not have an English
translation of that annex available for this review.
U.S. Probabilistic Seismic Hazard Analysis Standard ANSI/ANS-2.29-2020
Section 4.4 of the Standard addresses site effects. It states:"This section identifies acceptable procedures to incorporate local site effects into
the calculation of ground motions at different elevations in a site profile overlyinga reference horizon, typically the shallowest rock stratum that satisfies theassumptions associated with an elastic half-space. The purpose of a site
response analysis is to quantify the influence of the geologic profile above areference horizon on the amplitude and frequency of seismic waves propagating
to the profile surface. In cases where major above- or below-grade topographic
effects are minimal and where the geologic layers are generally flat, onedimensional ground response analysis is adequate to represent the wave
propagation conditions in which the response is assumed to be dominated byvertically propagating and horizontally polarized shear waves.""The guidance provided in this section covers the development of site-spectral
amplification factors that can either be (1) used to scale 5% damped uniform
hazard response spectra (UHRS) at the reference horizon or (2) used with the
PSHA results for the reference conditions in order to develop surface or control
point elevation hazard curves. Consistent with this, the site response analysis
shall properly account for the uncertainties as well as the variability in the site
effects on the ground motions computed at the surface of the geologic profile ofinterest."
Subsequent subsections of the ANSI/ANS Standard expand on available
approaches for the site effects analyses.

-6-

NUREG-2213,
Studies""Updated

Implementation

Guidelines

for

SSHAC

HazardA more succinct discussion of regional and site-specific studies and relationships to
local site effects is provided in NUREG-2213. Section 3.2.2, Regional and SiteSpecific Studies, states:"Another decision factor at the outset of a SSHAC study is determining whether aregional or site-specific study is needed. A site-specific study is one that is done
for a single site or facility at a particular location, while a regional study is one thatis conducted over a geographically extended region that includes multiple sites.
An example of a regional SSHAC Level 3 study was the central and eastern
United States (CEUS) SSC study (NRC, 2012b), which included the entire CEUS
east of the Rocky Mountains. This regional study was intended to provide an
SSC model that could be used at multiple nuclear facility sites across the eastern
half of the U.S.""Experience has shown that there are essentially two alternative methods forwhich the concept of a regional study has been implemented (with subtle
variations within these two general approaches). In the first method, referred toherein as a phased regional study, a general SSC and/or GMC model isdeveloped that is applicable for the entire study region. At a future date, sitespecific refinements are made to this model to make it appropriate for sitespecific use. An illustration of this process is given in Figure 3-6(a). The second
method, herein referred to as an integrated regional study, incorporates a seriesof site-specific assessments that are conducted simultaneously within theregional study. In this integrated regional method, the SSC and GMC models
share some common elements, such as the seismic source geometries andrecurrence, but the details of the SSC and GMC models are constructed toinclude site-specific aspects [e.g., behavior of nearby faults, shear wave velocity
and kappa corrections to ground-motion prediction equations (GMPEs)]. An
illustration of this process is given in Figure 3-6(b). The key difference between
these two types of regional assessments is that the phased regional study uses
general SSC and/or GMC models assuming that site-specific applications will be
conducted at a later time. Presumably, these could be conducted under a lower
level SSHAC process such as Level 2. The integrated regional model isdeveloped for multiple sites in the region incorporating both regional and sitespecific data, such that the model is immediately adaptable to all of the sites
considered in the study region. An important additional feature of the integrated
regional model is the incorporation of site-specific information that has
traditionally been used for site response. In the integrated regional model
approach, many aspects of the site response can therefore be included in the
ground motion model. It may be possible for other sites to utilize much of theinformation developed by the integrated regional study at a future time, but that is
not the primary purpose. Although both approaches provide an adequate way ofincorporating available information and satisfying SSHAC goals, the choice of an
integrated regional study or a phased regional study will depend on the needs of
the Sponsor and status of existing SSHAC studies."
Figure 3-6 from NUREG-2213 is reproduced below to facilitate this discussion.

-7-

Figure 3-6. Illustration of two types of regional SSHAC Level 3 or 4 studies. A
phased regional study is illustrated in the first panel (A), and an integrated
regional study is shown in the second panel (B). (from NUREG-2213)-8-

APPENDIX BREFERENCE STANDARDS AND EXAMPLE OF AN ENGINEERING APPROACH
FOR EVALUATION OF PIPING FRAGILITIESJapanese Seismic PRA Standard AESJ-SC-P006E:2015
Annex CB of the Standard includes evaluation methods for realistic response ofcomponents and piping systems. Annex DE, "Example of a Method of Calculating
the Probability of the Occurrence of an Initiating Event as a Result of an
Earthquake," includes examples of analyses for seismic-induced small, medium, andlarge LOCAs. Those analyses should require careful consideration of the pipe
rupture failure criteria. However, we did not have an English translation of Annex DE
available for this review.
NUREG-1903, "Seismic Considerations for the Transition Break Size"
One study, which was conducted by the U.S. NRC in connection with a proposedrisk-informed revision of Emergency Core Cooling System (ECCS) requirements in
10 CFR 50.46, provides an example of evaluating piping fragilities consideringvarious failure mechanisms. It is documented in NUREG-1903, and it provides someinsights into this issue. The question was what break size can be postulated
considering seismic effects. Although the focus of this study was for a primary loop,
the findings can be extended to other piping systems. The study used the seismic
hazard estimates for the central and eastern U.S. sites, and it looked at PWR
primary loops in detail. The Japanese seismo-tectonic environment would needfurther examination.
The study evaluated the following four failure mechanisms and conditions:
1. Fatigue
2. Unflawed piping
3. Flawed piping
4. Indirect failure because of support movement
The analyses in NUREG-1903, in part, utilized an extensive study conducted by theLawrence Livermore National Laboratory (LLNL) in 1980s. The LLNL results were
updated considering newer data. The unflawed piping failure probabilities were
evaluated using the capacity information based on evaluation of the EPRI/NRC
piping system and component data. However, treatment of the flawed or degradedpiping issue is quite difficult. Instead of directly estimating probabilities of throughwall failures, the NRC staff examined what would be the size of a critical flaw (depth
and circumferential) that would lead to a break at the earthquake levels associatedwith annual exceedance frequencies of 1x10-5 and 1x10-6 event per year. Evaluationof the indirect failures included an update of key support fragilities and use of theavailable seismic hazard information.

-9-