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

November 15, 2015

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:STATUS OF IKATA PROBABILISTIC RISK ASSESSMENT MODELSAND RECOMMENDED RESEARCH

Dear Dr. Apostolakis:
During the fourth meeting of the Technical Advisory Committee of the Nuclear Risk
Research Center (NRRC), October 26-30, 2015, we met with representatives of
Shikoku Electric Power Company, Ltd. (Yonden) and their contractor, Mitsubishi
Heavy Industries, Ltd. (MHI) to receive a briefing on the status of the probabilistic
risk assessment (PRA) models for Ikata Nuclear Power Plant Unit 3.
Yonden has volunteered Ikata Unit 3 to be the lead pressurized water reactor (PWR)
to focus NRRC support and guidance for the development of a plant-specific Level 2
PRA that initially evaluates the risk from internal initiating events, seismic events,
and tsunamis during plant power operation. The eventual goal is to extend that PRA
to include a full-scope evaluation of the risk from all internal events, all external
hazards, and all major site radiological sources during all operating modes that is
consistent with the current international state-of-practice in PRA methods, models,
and technical quality.

CONCLUSION AND RECOMMENDATION
1. The scope, level of detail, and technical quality of the Ikata Unit 3 PRA event
sequence models, initiating events, system fault trees, and plant-specific data forthe analyses of internal initiating events have advanced substantially beyond themodels that we examined one year ago. The PRA team is making excellent
progress toward the development of a state-of-practice PRA for the evaluation ofinternal initiating events. Additionally, these models will also provide a solid basis
for PRA models for other hazards such as internal flooding, fires, and external
events.
2. The NRRC should work with the Japanese nuclear industry to develop research,
methods, and guidance for improved treatment of the following phenomena in all-1-

relevant PRA models. These topics are not associated uniquely with Ikata Unit 3
or its PRA.
 Consequential failures of steam generator tubes during certain severe
accident scenarios
 Initiating events caused by failures of multiple tubes in a steam generator
 Leakage rates through isolation valves and low pressure systems exposed toreactor coolant system pressure Treatment of common cause failures in models for the quantification of plantspecific initiating event frequencies
 Treatment of national, regional, and site-specific data for the causes,
frequency, and duration of offsite power failures

BACKGROUND
Our November 1, 2014, letter report on "Suitability of Models for Ikata Site
Probabilistic Risk Assessment" concluded that the event sequence models that weredeveloped to support the Ikata Unit 3 Periodic Safety Review process provided agood technical foundation for extension and eventual development of a full-scope
Level 2 PRA. We also noted that we will perform periodic reviews of technical tasks
in the PRA, its supporting analyses, and results, and report our conclusions andrecommendations to the NRRC project team.
We have had extensive and very detailed ongoing technical exchanges with the
Ikata PRA team over the last year, including full Technical Advisory Committee
briefings in January and May. The topics of those exchanges have covered theLevel 1 PRA event trees, the selection of generic and plant-specific initiating events,
the fault tree models for a sample of frontline and support systems, and thecompilation of plant-specific data. This letter report provides a brief summary of thecurrent status of the Ikata PRA, and it contains recommendations for focusedresearch activities that will support enhanced PRAs for other plants.

DISCUSSION
Status of Ikata Unit 3 PRA Models
The scope and level of detail in the Level 1 PRA event trees are consistent withcurrent state-of-practice event sequence models. Event trees have been developed
for a full range of initiating events caused by plant transients, losses of offsite power,
support system failures, loss of coolant accidents (LOCAs), steam generator tube
ruptures, and LOCAs through interfacing systems that bypass the containment. The
models account explicitly for conditions such as failures of the reactor trip (ATWS)
and turbine trip functions, consequential failures of electric power supplies,
overcooling transients, failures of the reactor coolant pump seals, LOCAs throughthe pressurizer relief valves, and mismatches between reactor coolant system
makeup and letdown flows. We have not yet examined the Level 2 PRA models for
accident sequence progression and containment performance. Nevertheless, wehave noted that the Level 1 event trees are structured to provide a direct interface tothose models. The Level 1 event sequence models currently include limited credit-2-

for the additional equipment and enhanced accident mitigation capabilities that are
being implemented at Ikata in response to the accidents at Fukushima Daiichi. Workis in progress to more fully examine a few detailed technical issues and to refine
elements of the timing and success criteria for some event sequences, based onplant-specific best-estimate thermal-hydraulic analyses. Our interactions with the
PRA team are continuing, and we do not anticipate any major technical impediments.
The PRA team is performing detailed failure modes and effects analyses (FMEAs) to
identify a full scope of plant-specific initiating events. This effort is especially
important for failures of support systems such as cooling water supplies, instrument
air supplies, ventilation and room cooling, high voltage and low voltage AC power
supplies, and DC power supplies. This work is currently in progress. The scope andlevel of detail in the FMEAs are consistent with current state-of-practice methods.
The team will use the FMEAs to identify initiating events that are uniquely associatedwith specific features of the Ikata Unit 3 plant design, system configurations, andoperating practices. We will continue to review the results of these analyses as the
FMEAs are completed, and the final set of initiating event groups is developed.
Our examinations of the system fault tree models have involved a deep "verticalslice" focused review of the fault trees for a representative frontline system. That
examination also includes the models for the associated support systems (e.g., AC
power, DC power, actuation signals, cooling water, room cooling and ventilation,etc.) to check how those systems and dependencies are integrated into the PRA.
This level of review is not intended to be comprehensive. However, it provides agood understanding of how the models are developed and their important supporting
assumptions. Our technical exchanges with the PRA team in this area are in
progress. We have not identified any significant concerns with the scope of the fault
tree models, their level of detail, failure modes, and treatment of issues such as theunavailability of equipment due to testing and maintenance, personnel errors during
those activities, and common cause failures. Our discussions have also addressed
possible enhancements to the models for normally-operating systems to facilitate
their future use for risk-informed applications. Those enhancements do not directly
affect the capability of the current models to quantify the average plant risk.
However, they are intended to improve use of the models as engineering tools for
integrated plant risk management. Fault tree models are also being developed toquantify the frequency of initiating events that are caused by plant-specific support
system failures. That approach is consistent with current state-of-practice guidance
and methods.
The PRA team is using the Ikata Unit 3 Equipment Maintenance Information
Database (EAM) to derive detailed plant-specific data for equipment failures andunavailability due to maintenance. The available data cover approximately sevenyears of operating experience for all plant systems and equipment. Other sources of
information such as plant operating and testing records are being used to derive datafor component demands, operating hours, and standby exposure periods, which are
not directly available in EAM. This effort is in progress, and it is very encouraging. It
is consistent with the best practices in contemporary PRAs. We will continue tofollow the team's progress as the databases are developed and more information
becomes available.

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Needs for Focused Research, Methods, Guidance, and Data
During our discussions and technical exchanges with the Ikata PRA team, thefollowing topics were identified as potential needs for focused research, methods,
guidance, and supporting data. These topics are not associated uniquely with Ikata
Unit 3 or its PRA. Therefore, the NRRC should work with the Japanese nuclear
industry to develop the necessary information and guidance to facilitate theirconsistent treatment in all relevant PRA models.
Consequential Failures of Steam Generator Tubes
Research performed by the U.S. Nuclear Regulatory Commission (NRC) Office of
Nuclear Regulatory Research has indicated that certain severe accident scenarios
may result in conditions that increase the potential for thermally-induced failures ofsteam generator tubes. If they occur, these failures can result in a containment
bypass event and a release of fission products to the environment. The main
accident scenarios of interest are those that lead to core damage with high reactor
pressure, a dry steam generator, and low steam generator pressure (so-called "highdry-low" conditions).
Because these conditions are a consequence of only specific severe accident
scenarios, they do not affect the frequency of core damage. However, they affectthe scope and the logic structure of the Level 1 PRA models, because those models
must be developed to consistently identify and quantify these scenarios for theirfurther evaluation in the Level 2 PRA analyses.
Research is needed to examine the applicability of the U.S. NRC's analyses andconclusions to Japanese nuclear power plant steam generator designs, fabrication,and materials. Guidance is also needed to ensure that the integrated Level 1 andLevel 2 PRA models are developed in a manner to ensure that this issue isevaluated consistently, as needed on a plant-specific basis.
Initiating Events Caused by Multiple Steam Generator Tube Failures
Contemporary international PRAs have evaluated the risk from initiating events thatinvolve failures of multiple tubes in a single steam generator. The models for those
events are logically and functionally very similar to those for single tube failures.
However, the detailed success criteria and the time windows that are available for
operator actions to mitigate the event are typically much more challenging.
There is substantial uncertainty about the frequency of these multiple tube rupture
events. Estimates are available from studies that have been conducted in different
countries over a period of several years. Research is needed to perform a survey ofthe international literature, the supporting technical analyses, and the available data
to determine their applicability to the Japanese steam generator designs, operating
practices, and inspection programs.

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Enhanced Models for Interfacing System LOCA Events
Interfacing system LOCA events are typically not an important contribution to thefrequency of core damage. However, experience from numerous PRAs has shownthat they can account for a very important contribution to offsite releases. The
equipment and operator actions that can be used to effectively mitigate theconsequences from these events depend on both the size of the LOCA and its
location.
Limited research has been performed to examine the likelihood of leakage or ruptureof various types of isolation valves when they are exposed to large differential
pressures. Limited research has also been performed to examine the vulnerabilities
of low pressure piping, welds, equipment, seals, etc. to leakage or failure when they
are exposed to pressures that are beyond their rated designs. This research has
been used in contemporary state-of-practice PRAs to develop numerical correlations
for the frequency and size of interfacing system LOCA events, the locations of low
pressure system failures, and the consequential reactor coolant leakage rates.
These correlations are then used to determine the functional success criteria andtime windows that are available for operator actions to prevent core damage or to
mitigate offsite releases.
Research is needed to perform a survey of the international literature, the supporting
technical analyses, and the available data to determine their applicability to
Japanese system designs. This research may also require re-examination of
selected components, system material properties, or the development of focused
probabilistic models to refine previous estimates or to better characterize theassociated uncertainties.
Treatment of Common Cause Failures for Plant-Specific Initiating Events
As noted previously, fault tree models are being developed to quantify thefrequencies of initiating events that are caused by plant-specific support system
failures. Those fault trees are then integrated with the corresponding event
sequence models for plant response to each initiating event. This approach correctly
accounts for component failures that cause the initiating event and the unavailabilityof that equipment in the models for event mitigation. It also accounts for potential
dependencies between human errors that may contribute to the initiating event andoperator performance after the reactor trip occurs. This integrated approach isconsistent with current state-of-practice guidance and methods.
Experience has shown that the treatment and quantification of common cause
failures in the integrated models can have a very significant effect on the initiating
event frequency and the conditional probability for failures of similar equipment thatis needed for post-trip event mitigation. The issue is especially important fornormally-operating equipment such as pumps, ventilation chillers, air compressors,fans, etc. This is a common challenge for all PRAs.
Contemporary PRAs have applied different methods, data, and assumptions for thequantification of these common cause failures. Guidance is needed to ensure thatthese failures are modeled and quantified consistently in all Japanese PRAs.

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Data for the Frequency and Duration of Loss of Offsite Power Events
Loss of offsite power events are typically important contributors to overall plant risk.
International operating experience has shown that the causes, frequencies, anddurations of power failures are often affected significantly by national, regional, andlocal factors. For example, typical categories for the causes of power failures
include grid-related events, weather-related events, and switchyard-related events1.
Both the frequencies and the durations of offsite power failures are often correlated
to these causes, and they may vary significantly from site-to-site. Therefore, it isessential that each PRA should consistently account for all available operating
experience, design features, and data that affect the evaluation of loss and recovery
of offsite power supplies at the particular site.
Research and coordinated data collection are needed to develop a comprehensive
database for the causes, frequencies, and durations of offsite power failures that canbe used in a consistent manner for site-specific PRAs throughout Japan. The firststep in that process requires a clear definition of the scope of causes and analyticalboundaries for each category of loss of offsite power initiating event.
The data for these analyses may not be readily available to each nuclear power plantor its operating company. For example, relevant data may be available from sources
such as national grid reliability studies, regional power system operating data, andexperience from other large municipal or industrial consumers near each nuclear
power plant site. Therefore, such data should be compiled as a unified Japaneseindustry activity.

We look forward to continuing our review of this milestone NRRC project and its key
technical tasks as it evolves toward a state-of-practice full-scope PRA.

Sincerely,

John W. Stetkar
Chairman1

Some databases also cite frequencies for "plant-centered" events. However, experiencehas shown that the frequencies and consequences from those events are influenced verystrongly by the plant-specific electric power system design (e.g., transformers, operating andstandby bus alignments, automatic bus transfers, etc.). Therefore, state-of-practice PRAs
often develop plant-specific models to quantify the frequencies of those plant-centered
events, in a manner similar to initiating events caused by other support system failures.

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"Suitability of Models for Ikata Site Probabilistic Risk Assessment," TechnicalAdvisory Committee letter report to Dr. George Apostolakis, November 1,
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14.

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