Text

CPSES Risk-Based IST Program Risk Ranking Determination Study Summary Rept
	ML20094P782
Person / Time
Site:	Comanche Peak
Issue date:	11/22/1995
From:	; TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	;
Shared Package
ML20094P773	List: ML20094P782; ML20094P797; TXX-9526, Requests Exemption from Requirements (Testing Frequency) of 10CFR50.55a(f)(4)(i) & (II) & Approval to Utilize risk-based IST Program to Determine IST Frequencies for Valves & Pumps, Identified as Less Safety Significant;
References
	NUDOCS 9511300129
	Download: ML20094P782 (54)
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Comanche Peak Steam Electric Station Risk-Based In-Service Testing Program j

'I Risk Ranking Determination Study Summary Report Engineering Analysis 1

November 22,1995 1

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9511300129 951127 PDR ADOCK 05000445

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. Table of Contents -

1.0 B AC KGROUND..................................................... 1 - 1 '

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2.0 -. PROJECT SCOPE AND OBJECTIVES...................... '.............. 2-1 l

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PROJECT APPROACH

.............................................. 3-1 3.1 Methodology................................................. 3 -2 3.2

' Risk Importance Determination.................................... 3-5 i

i 3.3 Completeness Issues............................................ 3-6 3.4 Cumulative Effects of Test Interval Changes........................ 3-7 3.5 Expert Panel................................................. 3 -8 3.6 Identification of Component Degradation and Feedback Process.......... 3-9 3.7 Quality and Technical Adequacy of CPSES IPE...................... 3-11 1

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SUMMARY

AND CONCLUSIONS..................................... 4 l 4.1 Summary of Expert Panel Process.................................. 4-1 j

4.2 Results and Conclusions......................................... 4-3 5.0 REFEREN CES...................................................... 5 - 1 3

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1.0 BACKGROUND

In-service Testing (IST) programs were developed to ensure the reliable operation of safety-related pumps and valves at nuclear power plants. The codes, standards and guides for these tests were developed by the American Society of Mechanical Engineers (ASME) Operations and Maintenance (O&M) Codunittee. The essential Nuclear Regulatory Commission (NRC) regulation goveming this process of testing has been 10CFR50.55 and has been implemented using ASME BaPV Code Section XI (R.ef.1), both for passive component examination (welding, studs, etc.) and for active component testing (pumps and valves).

For the past several years, both the nuclear industry and the NRC have devoted significant e.ltention and resources aimed at improving the performance of pumps and valves. In a letter (Ref. 2) dated September 9,1991 from James E. Richardson of the NRC to Forrest T. Rhodes of ASME, the NRC requested that the ASME O&M Committee consider revising existing requirements for in-service testing. The letter requested revisions to ensure the ability of certain pumps and valves to perform their intended hydraulic and mechanical safety functions. The revisions requested would:

Expand the scope to include specific components that are not constructed in accordance with ASME B&PV Code Section III rules for construction or tested in accordance with ASME B&PV Code Section XI; Require verification of each safety function for each included component; Require such verification be accomplished at design basis conditions, or, where such verification is not possible, a test at less than design basis conditions combined with an analysis may be substituted; and Data collected during component testing may be compared with data taken during previous tests to allow determination of the condition of the component.

This request was rnade in part due to NRC concems with the ability of some components to perform their safety functions under design basis conditions, such as motor-operated valves and check valves, and concem that the in-service tests required by ASME B&PV Code Section XI and incorporated by reference into 10CFR50.55a(f) do not: a) include each component that has a hydraulic or safety-related function; b) accomplish verification of each safety function of each safety-related component; or c) require that such verification be accomplished at the design basis conditions.

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The intent of current IST programs is to include all active safety-related pumps and valves that are credited in the plant design basis safety analysis. In general, the IST equipment lists are developed by review of plant drawings showing ASME Code Class 1,2 and 3 classification boundaries. All components within the boundaries are then reviewed to determine whether or not they were credited with an active safety function under the plant licensing basis. The FSAR analyses and other design basis documentation are reviewed to make these determinations.

Older plants not initially designed to ASME B&PV Code Section III have applied ANSI Safety Class 1,2 and 3 classification rules to piping and components for purposes of establishing ASME B&PV Code Section XI test requirements, even though the systems and components were not designed or constructed in accordance with ASME B&PV Code Section III.

As a result of the NRC request for IST program enhancement, there are industry concerns involving the restrictive nature and basis for these requirements and their impact on plant operation. Overly restrictive requirements can complicate plant operation, cause unwarranted operating costs, and most importantly, degrade plant safety through needless component testing and undue burden during plant outages.

Developments in the industry demonstrate an acceptance of the use of risk-based approaches using a plant's probabilistic safety analysis (PSA) to identify prescriptive regulations that have marginal safety benefits. The momentum in this direction is evidenced by recent NRC interest in graded QA and EPRI's applications of risk-based technologies, and most recently, in the issuance of the Nuclear Regulatory Commission's final policy statement on the use of PSA in nuclear regulatory activities (Ref. 3).

Similarly, improvements to IST programs using a risk-based approach can reduce operating l

costs while maintaining a high level of plant safety. Possible savings from improved IST programs include:

Reduced costs of engineering analyses to develop test criteria that adequately demonstrate functional capability at design basis conditions; Reduced costs of plant modifications where current configurations do not support testing at or near design basis conditions; Reduced costs for development of new test procedures implementing the new test criteria; and l-2 i

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. Reduction ofincremental costs associated with performing the new tests, including:

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' ' Additional time required to perfonn the tests and analyze results; Costs of specialized test equipment or vendor services;-

Possible effects on critical _ path outage duration; and Possible increases in radiation exposure.

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~ For these reasons it is advantageous for utilities to pursue IST program improvements. The

. impact of changes on plant safety is of primary interest and is the controlling factor in implementing such changes. _ However, changes that negligibly reduce plant sefety should not be ruled out, especially if such changes can lead to significant plant performance improvements -

in other areas.

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' PROJECT SCOPE AND OBJECTIVES The scope of this project is to perform a review of the Comanche Peak Steam Electric Station j

IST program that optimizes the safety benefits in assuring pump and valve performance, it uses a methodology for a risk-based approach to IST program review and enhancement that is i

founded on a blend of probabilistic and detenninistic methods and that has as its principal results, recommendations for adjustments to test frequency intervals for these components.-

Thus, it is not aimed at reducing the number of components within the scope of an IST program, i

rather at optimizing what is tested and when. In this study, all components within the scope j

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of the IST program were examined. However, only those determined to be less safety significant will be considered for a code exemption. The ASME O&M Committee is reviewing the more safety significant components to ensure that the appropriate tests are identified and.

j performed on those components for their respective failure modes.

i The objectives of this project are to apply risk-based technologies to IST components to determine their risk significance; to apply risk-based technologies to risk-significant components identified in the IPE and outside of ASME Code Classes 1,2 and 3 to detennine whether additional compensatory measures are appropriate; and to apply a combination of detarministic and risk-based methods to determine appropriate testing frequencies and/or compensatory measures for IST components. The results of this project will be the basis for the CPSES code exemption submittal to the NRC and will be part of a pilot study for the industry.

Several safety enhancements to a plant IST program can be derived, both directly and indirectly, l

by using the probabilistic and deterministic approach presented in this report. These safety enhancements are very similar to those attendant with the optimized performance of motor-b

~ operated valves discussed in NUMARC 93-05 (Ref. 4), from which elements of the following discussion were taken, f

Direct Safety Enhancements e

i Greater attention and resources devoted to the high priority IST components' could translate into l

many direct safety enhancements. First, this group of components could be subjected to, where practical and meaningful, more frequent periodic tests than the lower priority groups. The timeliness of any problem identification and resolution would be improved.

Second, requirements associated with the high priority group ofIST components are expected to be more 2-1 1

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rigorous and demanding in nature than for the other groups. These requirements provide added assurance that any problems that may impact the functionality of the components will be identified and resolved. Third, the resulting risk-based IST program will consider whether some risk-significant components that are outside the scope of ASME Code Classes 1,2 and 3 should be added to the IST program to improve safety. Finally, because extensive testing can have

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adverse safety and operational consequences, reduction of testing may reduce component l

wearout and operator burden. These changes are expected to improve safety.

indirect Safety Enhancements There are other indirect safety benefits to this approach that are as important. Risk-based prioritization efforts identify the safety-significant IST components and the impact of their i

potential failures on plant safety. -In addition, these analyses identify important scenarios that provide information with respect to the operational demand that may be placed on a given component. Such information is valuable because it relates the performance of the IST component to the broader context of plant safety. This allows more rational decision making, more efficient use of resources, and is central to optimizing safety benefits.

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PROJECT APPROACH The TU Electric risk-based IST project was developed and implemented as part of a tailored collaboration (TC) effort with EPRI. The project was conducted under the direction of a Steering Committee that interfaced with the American Society of Mechanical Engineers (ASME) research program furided by the NRC, the Westinghouse Owners Group (WOG), the Nuclear Energy Institute (NEI) and other utilities, and coordinated its activities with other industry efforts such as the WOG check valve program and various NEI activities on risk-based regulation. The TC project was designed to provide plant-specific benefits to TU Electric and, as a pilot project, to provide generic insights and tools that will benefit similar industry projects.

In particular, the project developed generic methods for identifying opportunities to reduce those IST-related regulatory requirements and commitments that require significant resources to comply with and/or implement, but contribute insignificantly to safe and reliable operation.

This work is being provided to NEI's Risk-Based IST Task Force and ASME B&PV Code Section XI IST Research Task Force to assist them in their formulation of guidelines and in-service testing requirements.

The Steering Committee developed the overall project objectives and milestones and commissioned various work activities and studies in doing this work. The Steering Committee consisted of members with expertise in the areas oflicensing, probabilistic safety analysis, ASME B&PV Code Section XI and WOG analysis activities. In addition to providing overall coordination, the Steering Committee served as the central point of decision making for major technical issues and provided technology transfer and guidance to the expert panel in performing its work. These latter activities were accomplished through common membership of several members on the Steering Committee and the expert panel. It was concluded that the strength of this risk-based IST program and the integrity ofits results lie both in the rcbustness of the methodology and in the work of the Steering Committee and expert panel. Further, the robustness of the methodology provides consistency in the results.

The project was divided into two phases.

Phase 1 included the development of an implementation guidelines document and actual implementation of the methodology to prioritize components in the IST program. Phase 2 involved the development of tools for evaluating test intervals for the risk-significant IST components. The work activities in each of these phases were reviewed by the Steering Committee and presented to various other peer groups at strategic points in the project. In this way the methodology was refined, and a fairly 3-1

mature process was arrived at before involvement of the expert panel. The various tasks that

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support the project are described in more detail in the sections that follow.

3.1 Methodology The process described above lead to development of the methodology. The methodology was leveloped consistent with NUMARC Guides 93-01 (Ref. 5)(Maintenance Rule) and 93-05 (Motor Operated Valve (MOV) testing). The system level ranking approach from the Maintenance Rule process was merged with the component level ranking approach used for MOV testing. The merging of the two approaches was designed to ensure that the new IST program would benefit from and be consistent with the Maintenance Rule process and other industry risk-based programs.

The Risk Achievement Worth (RAW) and Risk Reduction Worth (RRW) risk measures of the Maintenance Rule were combined with the Fussell-Vesely (FV) risk measure of MOV testing.

Because this initiative was to reduce existing regulatory burden rather than focus on new regulatory initiatives, the methodology applies these risk measures in a manner intended to ensure a safety-neutral outcome.

Because RRW and FV provide similar insights, only the FV importance measure was utilized in this analysis. Fussell-Vesely provides a measure ofincremental change in total core damage frequency (CDF) that indicates the importance ofincremental changes in reliability that might result from changing in-service test intervals. Risk Achievement Worth provides an indicator of the importance ofdegradations in component reliability. These measures were combined into a decision criteria such as that shown in Figure 3-1.

As the figure indicates, components with a significant FV were considered "more risk significant". Components with an insignificant FV were considered "less risk significant".

However, it was important to ensure that a reduction in test intervals did not allow unintended consequences, i.e., a compromise in safety resulting from a degradation in reliability.

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n Figure 3-1 r

Decision Criteria Rink Imnortance Meannres Fussell-Vesely importance > 0.001 Risk Achievment Worth > 2.0 10 I

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2 Risk Achievment Worth.

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0.001 0.1 Fussell-Vesely importance i

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More Safety Significant Component (MSSC) 11:

Less Safety Significant Component (LSSC) With Compensatory Measures 111: Less Safety Significant Component (LSSC)

IV; More Safety Significant Component (MSSC)

Not Modeled - Components Reviewed By Expert Panel For Determination Of Ranking i

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' Therefore, if FV was insignificant,' it was also required usat RAW be insignificant for a component to be classified as "less' risk significant". If RAW was significant, the component.

was considered by the expert panel for placement in the high category. If the panel' decided the.

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. component could be ranked low, an additional requirement was imposed before a component -

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could be classified as "less risk significant". A compensatory measure was requi ed to be selected by the expert panel th limit degradations in reliability.

f During the development of this methodology, EPRI and NEI began working with NRC on the development of the EPRI PSA Applications Guide (Ref. 6). In general, this methodology is consistent with the guide. The guide did provide a specific acceptance criteria for permanent risk increases that was used in this evaluation. A few minor differences between this methodology and the EPRI PSA Applications Guide exit, most of which are more conservative j

in this study.

The general approach taken included four steps. First, risk importance was determined. This i

determination was based on the results of the IPE and the IPEEE and other plant operatmg modes, such as outage modes. In addition to this complete spectrum of core damage accidents, severe accidents leading to large and early fission product releases were also given special attention. Finally, the importance of components not in the IPE and IPEEE models or not in the

. IST program were evaluated.

The next step addressed the completeness and adequacy of these models through a number of sensitivity analyses to compensate for the limitations of the quantitative models. The third step evaluated the cumulative impact of low risk significant components on plant risk if their in-service test intervals were extended. This step provided technical justification for proposed test intervals for less risk significant components in the existing IST. Tiie fourth and last step was

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to review the process and results with an expert panel that was knowledgeable of plant risk, j

plant design, plant operations practices, and plant performance. This process blended deterministic safety insights with quantitative risk insights to ensure that risk significance was appropriately identified.

The following sections further describe the methodology and provide some additional L

a background to this work.

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3.2 Risk Imnortance Determination i

i In this study, risk importance rankings of the IST components were determined based on the j

results of the CPSES IPE. These risk rankings were then complemented with rankings based on consideration of other accident initiators and plant operating modes. These other accident initiators are extemal events sdch as fires, tomados, and earthquakes. The other plant operating mode is the outage mode. Each of these evaluations considered importance with respect to core damage prevention. Core damage prevention has been found to be a good measure of the spectrum of releases that can result from severe accidents. However, unicue risk contributions can occur if severe accident releases are large and early. Hence, risk rankings were also complemented by considering components important to preventing large, early releases. This approach is consistent with the intent of the safety goal and the severe accident policy statement and is a requirement of the EPRI PSA Applications Guide.

In applying the above method, it was found that a significant action ofIST components are not in the IPE While it is likely that such components are not risk significant, this study specifically evaluated each component and the design basis functions addressed by the IST program. Most components that are not in the IPE were found to be implicitly modeled by the i

study. That is, the IPE found that the components either were not required for the system to prevent severe accidents, were in systems that provided a highly redundant function, or performed functions that were extremely unlikely to be required. The systematic review of these components used quantitative and qualitative insights to determine whether components should be considered more or less risk significant and whether risk insights implied that compensatory actions should be considered.

1 The risk ranking process also identified some IPE components that were more risk significant but which were not in the IST program. These components typically were found to be outside the code class boundary and therefore not subject to IST requirements. These components were considered for compensatory action equivalent to those defined for components in the IST program.

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3.3 Completeness Issues Quantitative risk models have limitations associated with the structure of the models and the assumptions and the input data used. The limitations were compensated for by evaluating truncation limits, identifying IST components masked by the IPE, applying a conservative treatment of common cause failures, requiring an expert panel to identify components with operational concerns, and performing selected sensitivity studies.

The risk ranking process described above used the FV and RAW importance measures. The values for these importance measures are calculated based on cutsets. The cumulative effects analysis described below also is based on cutsets. Cutsets are obtained by solving the model with a truncation limit. Experience has shown that setting the truncation limit arbitrarily low creates inefficiencies such that analysis costs quickly exceed the value of risk insights gained.

This project evaluated the truncation limit used in the CPSES IPE and found it to be sufficient for both risk ranking and estimating cumulative effects.

The IPE model may " mask" certain components because they are associated with supercomponents, human events or initiating events but not explicitly identified. The components masked by the IPE model are typically small contributors to the overall probability of the event. However, it was considered appropriate to verify this consideration for this effort.

The project evaluated those IST components that were: 1) contained in supercomponents (e.g.,

some components on the diesel generator skid),2) required to function as part of a human action, and 3) might cause a significant plant initiator.

Risk ranking results can be strongly affected by the contribution of common cause failure. The approach taken in the project was to conservatively assume that a common cause event in the cutsets should have its entire risk significance assigned to all components represented by the event. This approach lead to the inclusion of a significant number of components in the more risk significant category which otherwise would have been considered less risk significant. The expert panel confirmed that the approach identified potentially important components.

Both risk ranking measures used are influenced by the reliability data assigned to the component. The CPSES IPE uses generic data since an insufficient amount of plant-specific data was available. Generic data (and indeed, most interpretations of plant specific data) 1 i

considers components in groups. But ranking was done on a component basis. Consequently, 3-6 1

the expert panel considered whether or not plant specific operational insights indicated component reliability problems that might affect the ranking of an individual component or small group of components. Components with operational concerns were considered more risk 1

significant by the expert panel.

Finally, the completeness of tlie models, assumptions and input data were tested by sensitivity studies. In one sensitivity study designed to consider the impact of human event modeling, risk ranking results were compared assuming operator events in the IPE always failed to occur.

Another sensitivity study was designed to consider whether changes to in-service testing offered the potential for common-cause-like degradations in components in different systems. Less risk significant components were assumed to be influenced two at a time. Four such components were identified which, together with other components, otTered the potential of becoming more risk significant. Appropriate compensatory actions designed to limit reliability degradations were imposed on these components. A similar sensitivity study was performed where less risk significant comporints were assumed to be influenced three at a time.

3.4 Cumdative Effects of Test Interval Changes A risk ranking approach based on importance measures such as was used in this project does not necessarily guarantee that acceptable levels of risk will result. Risk importance measures 1

are based on changes to components one at a time.

Changes to many components simultaneously may cause unintended increases in risk despite meeting the selected conservative risk ranking measures.

An analysis was performed to determine the potential risk impact ofincreasing in-service testing intervals simultaneously on all less risk significant components. Consideration was given to available information on how changes in test intervals will change component unavailability.

Uncertainty in this information, together with the complexity required to model such an approach, dictated the use of a very conservative approach. That is, risk impact was measured assuming that component unavailability (including both on demand and time dependent failure rates) increased by the same factor that the test interval increased. Despite the use of this conservative assumption, calculations indicate that test intervals could be increased from quarterly to six years or more with acceptable increases in risk. If consideration were given to improvements in performance that are possible to occur from a risk-based IST program, it is plausible that core damage risk may not increase at all.

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3.5 Exnert Panel For the CPSES Risk-Based In-Service Testing (RBIST) Program, an expert panel (EP) was established to make the final determination of risk ranking for the pumps and valves in the CPSES Unit I and 2 IST program. The panel was constituted in part ofindividuals who were members of the Steering Coinmittee and of others who were members of the expert panel established for the implementation of the Maintenance Rule.

The members of the panel were selected based on their nuclear power plant experience which included expertise in the areas of ASME codes and standards, plant operations, maintenance engineering, systems engineering, design engineering and probabilistic safety assessment (PSA). The minimal education and experience requirements for panel members were a BS in an engineering discipline and eight years in nuclear power. The operations representative currently holds a USNRC Senior Reactor Operator License and has held it for at least two years.

The chairman has significant technical expertise in PSA applications and project management.

The expert panel also utilized the expertise of other consultants and engineers in doing its evaluations.

The minimum quorum necessary for the EP to conduct business was four (4) members consisting of the representatives from operations, probabilistic risk assessment, system engineering /in-service test engineering, and codes and standards. It was decided that the panel would be living and it would participate in periodic updates to the ranking whenever the IPE study is updated.

l The scope of the expert panel activities included both risk ranking and application ofit. The panel's principal responsibility was to ensure the risk ranking information was consistent with plant design, operating procedures, and with plant-specific operating experience. The panel made a qualitative assessment of the risk importance categories that were developed for the components using the IPE results and insights discussed in the preceding sections of this report.

This assessment was based on deterministic insights, plant-specific history, engineering judgements, regulatory requirements, and probabilistic safety analysis insights. The panel reviewed the IPE component risk rankings, compared the IPE and IST functions to ensure consistency with plant design, analyzed applicable deterministic information and determined the final safety significance categorizations for all the IST components. At the end of the expert panel evaluation process, every component in the CPSES IST program was reviewed and 3-8

A evaluated by the expert panel members. A summary of the expert panel process is provided in section 4 of this report.

3.6 Identification of Component Degradation and Feedback Process At CPSES, various station l#ccedures are used to govern the activities related to the IST program and other areas such as corrective action and root cause programs. These procedures form a consistent means of controlling and integrating site-wide activities. The ASME B&P Code Section XI in-service testing of pumps and valves is implemented by procedure STA-711, "ASME Section XI In-service Testing Program for Pumps and Valves.' This procedure provides guidance to ensure effective, consistent and coordinated implementation of the code requirements. It provides guidance on how the in-service testing program interfaces with other station procedures to perform surveillances, to maintain test records, to assure deficiencies are identified, tracked and resolved, and to assure that corrective actions are performed and documented. These procedures provide the means by which feedback of failures ofIST components to the IST program is accomplished. They provide assurance that failures ofIST components will be promptly identified and addressed and modifications to the in-service testing program (e.g., change to surveillance intervals) are made in a timely manner.

A failure of an IST component may be identified in the course of doing ordinary maintenance and tests or as part of a surveillance activity. These activities are controlled primarily by STA-606, " Work Requests and Work Orders," and STA-704, " Surveillance Program". When a failure is identified as part of a surveillance test or maintenance activity, a ONE Form is prepared per STA-421," Operations Notification and Evaluation (ONE FORM)", depending on the nature of the failure. This form is used at CPSES to report potential adverse conditions and resolve issues and to assure that corrective actions are performed and documented. Resolution of a ONE Form is accomplished in accordance with the requirements of STA-422, " Processing of Operations Notification and Evaluation (ONE) Forms". Resolution of a ONE Form includes:

i Assigning a unique identification number and logging in appropriate plant information systems, and initial distribution for trending purposes.

Reviewing the reported condition to determine the category of correction action required.

Considering the generic implications of the item, i.e., the potential for the condition to exist elsewhere and initiating works order as required to investigate.

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I Determining the probable cause of failure.

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.. Identifying and performing corrective action.

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- Depending upon'the nature of the adverse condition, the corrective actions may includeL

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reporting to outside agencies, performing an engineering evaluation or performing a root cause evaluation.~ Root cause evaluations are preformed in accordance with STA-515," Root Cause

. Analysis." These evaluations include a structured analysis ofissues in order to identify causes -

of and contributing factors to component failure. As appropriate, root cause evaluations' consider human performance issues and require failure analysis of components.

In addition to these activities, the implementation ~ of the Maintenance Rule at CPSES requires that failures of components in systems within the scope of the rule be reviewed to determine whether these failures are maintenance preventable functional failures. The IST systems are

.within the scope of the maintenance rule and thus will come under these provisions.

Maintenance preventable failures that result in system functional failures receive root cause analysis and corrective action evaluations, if the Maintenance Rule has been implemented on the system.

For deficiencies arising from surveillance work orders, records of corrective action are documented on work orders per the requirements of STA-606, " Work Requests and Work t

Orders". Work orders contain details of all corrective actions performed. Records ofin-service testing to confirm operational adequacy following corrective actions are documented on post-i work test reports per the requirements of STA-623, " Post-Work Test Program." The IST engineer reviews all closed IST-related surveillance work orders and post-work tests. The IST engineer also reviews in-service valve test results during the work order post-work review process and extracts and records any trendable data for early identification of equipment problems that may require modification to the IST program.

Because the IST engineer is a member of the systems engineering group, his activities are closely integrated with those of the system engineers. The pump and valve performance records maintained by the IST engineer are used extensively by systems engineers to determine corrective actions and to monitor system performance. The IST engineer is also a member of j

the expert panel for implementation of the Maintenance Rule and the risk-based IST program.

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.He participates in periodic reviews of the performance of systems within the scope of these 3-10

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programs, and through these means, he can provide timely feedback of performance of e

components in the systems.

- Thus, the various procedures and programs in place at CPSES provide assurance that failures -

ofIST components will be promptly identified and addressed and modifications to the in-service l'

testing program will be considered and made in a timely manner.

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3.7 Onnlity and Technical Adeanacy of CPSES IPE k

In general, the IPE study for CPSES fully satisfies the requirements of a full-scope Level-I and 1

Level-II PRA. One of the main objectives of the IPE development was to be able to utilize its results and insights toward the enhancement of plant safety through risk-based applications.

With this objective in mind, the IPE elements were developed in detail and integrated in a I

j manner sufficient to satisfy both the NRC Generic Letter 88-20 requirements and support future

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plant applications.

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. The CPSES IPE study was performed by developing large fault trees and small event trees. The large fault trees were then linked together according to the event tree logics for quantifying 1

l accident sequences. The major elements of the IPE study were developed and reviewed in a manner consistent with and in excess of the good practices of the time. In general, it is believed that the CPSES IPE meets or exceeds the quality standards subsequently suggested by the EPRI PSA Applications Guide. These major elements are briefly described below.

Initiatino Event Analysis A detailed review of plant equipment and operating procedures was performed to identify all the potential plant-specific initiating events as well as those initiating events that were identified in the industry. The loss of support system initiators such as service water, component cooling 4

water, safety chilled water, HVAC, Instrument Air, Electrical Power subsystems were also identified and evaluated in the IPE study. In addition, other special initiators including l

interfacing systems LOCA, SGTR, ATWS, internal flooding and station blackout were analyzed i

in detail and documented in the IPE.

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' Accident Seanence Annivsis 7

.A detailed accident sequence analysis was performed and resulted in the development of j

functional event trees for all the initiating events identified in'the IPE study. This also included induced LOCA initiating events such as stuck open primary side safety valves, stuck opcn 4

PORVs, and most importantly, reactor coolant pump seal LOCA.

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The accident sequences were quantified using the fault tree linking methodology. The common

. concern in the industry is the truncation limit which could potentially impact the importance evaluation. The total core damage frequency for CPSES was estimated to be 5.72 E-05. The truncation limit chosen for the CPSES accident sequence quantification was set at L0E-09 which is approximately 2.0E-05 below the total core damage frequency. The recommended

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truncation limit in the EPRI PSA Application Guide document is 10" below the baseline IPE l

core damage frequency. The analysis of truncation limits for this application is described in h

section 4 of the main report. Most assumptions related to IST components were in effect validated by the treatment of not-modeled IST components. In addition, ATWS mitigating IST components have been ranked appropriately.

1 Systems Analysis One of the major elements of the CPSES IPE study was the system analysis task. A total of 15 systems including support systems and front-line systems required for accident mitigation were analyzed. For all 15 systems, detailed system notebooks were developed which are' found to

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be excellent documents for plant support activities. The impact of the loss of room cooling on I

equipment operability was carefully evaluated by the plant-specific room heat up calculations and other available information in the industry. As part of this effort, the impact ofloss of room cooling on the control room and switchgear room were also evaluated.

Common Cause Failure Analysis t

l Common Cause Failures (CCF) impacting two or more components in a system were carefully examined and appropriately placed in the system fault tree models. The Multiple Greek Letter (MGL) method described in NUREG/CR-4780, " Procedures for Treating Common Cause

. Failures in Safety and Reliability Studies," was used to quantify the effect of common cause

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-1 jo failure events.- The evaluation process is consistent with the NRC and EPRI guidelines. The

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- typical IST-related component types are included in the CCF analysis. These are:

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Motor operated valves I

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Air operated valves i

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L Check valves

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Electro-hydraulic valves a

4 Solenoid valves i

a' Operating pumps Standby pumps a

Turbine-driven pumps

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Positive displacement pumps

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Human Reliability Analysis TU Electric spent extensive amount of time to review, analyze and document human interactions that were modeled in the IPE study. This analysis is consistent with the guidelines of SHARP methodology developed by EPRI. This analysis included an evaluation of operator -

timing and emergency operating procedures that might create more demands on the operator, i

In general, three groups of human interactions were considered, namely, latent human errors, j

human errors associated with initiating events, and dynamic human errors. In addition, a detailed recovery analysis was performed to properly account for the possible recovery actions.

The approach adopted for the CPSES IPE follows the general guidelines in the EPRI recovery

)

' analysis (EPRI RP 3206-03,"Modeling of Pecovery Actions in PRAs"). The recovery analysis included the interview of operations staff with extensive plant experience, development of decision trees, review of related procedures and drawings, and consideration of the available time for each critical recovery action. The human reliability analysis process and results were all documented in a separate notebook.

3-13 r

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i

',.m IPE Review Process 1

i To ensure a high-quality IPE and to provide quality control to the IPE Process, two types of independent reviews were conducted. One was done internally by TU Electric staff, and the other was done extemally by outside PSA exoerts. In general, both reviews were applied to the j

t entire examination process except when it was not possible due to the availability of resources '

or required skills. In those few cases, as a minimum, each task was reviewed thoroughly by either an internal or external independent reviewer. Furthermore, a final independent review 4

l was performed after the IPE study was completed. A team of PRA experts was selected from the industry to independently review the entire IPE study and its supporting analyses. The j

review team spent one week at the TU Electric offices where documents, procedures and supporting calculations and analyses were available for use. The results of all independent review activities performed by intemal and external reviewers were well documented as part of the IPE documentation requirements.

5 5

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3-14 l

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^s 4.0

SUMMARY

AND CONCLUSIONS.

Section 3.0 provides an overview of the process used to develop the risk based IST plan for.

CPSES. This section provides a discussion of the results and conclusions of the process.

i 4.1 Summary of Exnert Panel Process As described earlier, the expert panel process was integrated with a Steering Committee which in turn coordinated with other industry activities such as the ASME research program and the WOG check valve program. The expert panel for risk-based IST was essentially the same as

. the Maintenance Rule expert panel with the addition of the Steer ng Comm ttee c a rman and i

i hi i

the IST program coordinator, both of whom were knowledgeable ofIST requirements and commitments, IST plan implementation, and CPSES plant performance.

To prepare for the expert panel review, the risk ranking team developed a set of simplified P& ids for all the systems modeled in the IPE. The IPE risk category results, component tag numbers, and the location of the components in the systems were all shown on the simplified diagrams. Using this information and the design basis functions addressed by IST as documented in the IST plan, the panel reviewed and validated or adjusted the ranking results.

The panel's principal responsibility was to ensure the risk ranking information was consistent with plant design, operating procedures, and with plant-specific operating experience. The IPE and IST functions were compared to ensure consistency with plant design. In particular, reverse flow in check valves was evaluated to see ifit might be risk significant since the IPE assumed this to be low probability. If redundant trains could be affected by that failure mode, the risk ranking was adjusted accordingly. Also, information was fed back to the ranking process to reflect unmodeled operator actions that altered some ranking information, usually a RAW value rather than a FV value. Finally, the panel identified operational concerns about a specific

. component that might affect the risk ranking or might make in-service testing desirable for other reasons. In more than one case, a component's ranking was increased to high because in-service testing helped prevent entry into a limiting condition for operation (LCO).

The panel also reviewed the sensitivity of the component rankings to common cause failure.

Many of these components were valves in the lower half of the FV medium category (i.e., from 0.005 to 0.001). The panel felt that these were important components and that they should be retained as is in the IST program.

4-1

r i

The panel also reviewed the rankings and their associated technical bases for other sources of risk and other risk measures, namely, the IPEEE and outage risk sources and the large, early release risk measure. Based on this information, the ranking ofsome components was increased to high.

In the event that the panel found a component to be potentially high (low FV, but high RAW),

l the panel selected a compensatory measure to ensure that component functionality would still l

be evaluated on a regular basis by other plant programs. Because pumps were often ranked high

~

and potentially high components were often in the flow path for the IST pump test, the quarterly j

pump test was often found to be an effective compensatory measure for suction and discharge j

check valves. Potentially high MOVs were often " tested" by other technical specification-requirements, namely slave relay tests.

The panel also spent a significant portion ofits time reviewing IST components not modeled by the IPE, and IST components that were modeled by the IPE but for which not all IST

.l functions were modeled. Risk ranking of these functions was based on insights gained from the earlier work, e.g., components whose failures might affect redundant trains were ranked high.

l l

The final ranking step performed by the panel was to consider IPE components not in the IST j

which met the criteria for high risk. All the high risk components not in the IST program were confirmed by the expert panel. In general, the importance ofinstrument air and the decay heat removal related portions of main steam were the principal focus of the panel's considerations.

Evaluations were designated to determine how to best use in-service testing techniques to address the more safety significant failure modes modeled in the IPE.

i i

For high ranked components in the IST, the panel decided to maintain all in-service testing as t

l is, regardless of whether some failure modes (and therefore some tests) were not risk significant.

This conservative approach was adopted for ease of implementation and administrative consistency. For low ranked components in the IST, the panel discussed the technical basis for

. increasing test intervals and yet maintaining plant safety. In addition, the panel considered l

implementation issues associated with particular test intervals. The panel concluded that generally a staggered test implementation over 6 years would be the best implementation j

strategy.

l I

i t

4-2 l

4.2 L Results and Conclusions In this study, all components within the scope of the IST program were examined. In all, a total of 687 components were examined and ranked as either High-more safety significant or Low-less safety significant. Of this total,654 valves were evaluated,117 (17.9%) of which were -

ranked high and 537 (82.I'4) of which were ranked low. Thirty-three (33) pumps were evaluated,21 (63.6%) of which were ranked high and 12 (36.4%) of which were ranked low.

Of the total components, 375 (54.6%) were modeled in the IPE and 312 (45.4%) were in IST only, most (285) of the latter being low ranked valves. Only those determined to be less safety significant (low) will be considered for a code exemption.

Table 4-1 lists all the components by tag number that were examined in this evaluation. This table shows the entire spectrum of the review and the results of the expert panel evaluations.

The risk ranking process was concluded to be robust. It generated results that were consistent with deterministic insights from the expert panel and found to be safety neutral. The following spectrum of risk and deterministic insights demonstrates this conclusion:

a spectrum of risk sources were considered, i.e., IPE, extemal and outage, multiple risk measures were considered, i.e., CDF and LERF, a

diverse importance measures were used, i.e., FV and RAW, sensitivity studies consistently demonstrated that the risk significant components had e

been identified, both IPE and IST functions were compared and evaluated and considered in an integrated manner, and both PRA anddeterministic insights from the expert panel were incorporated into both the ranking results and the resulting IST plan.

The scope and level of detail of the results review by the expert panel, the emphasis placed on understanding why components were ranked high or low, the careful comparison of the IPE and

..the IST functions, and the sensitivity studies performed all demonstrated the technical adequacy of the IPE to serve as the basis for this and other risk based applications. The resulting risk based IST program is considered by the expert panel to have the appropriate changes (both increases as well as decreases in scope) and the appropriate checks and balances to ensure burden reduction can be achieved while maintaining or even improving plant safety.

4-3

l i

4 J

l The results of this analysis indicate that the risk increase associated with the proposed interval

-.3 changes is acceptable even with the very conservative assumptions used in the study. The total risk may in fact decrease if the overall IST program becomes more efficient by focusing on the more important components'. Each of the important components are represented more than once in nearly all of the cutsets containing pumps and valves. A small improvement in the unavailabilities ofimportant components would likely translate into a corresponding reduction j

in risk. This reduction in risk is probably larger than the increase that might result from I

increased test intervals since it is expected that the risk increase would be even less than the l

l amounts calculated here.

In conclusion, modifying the test frequencies of the IST components in the low safety significance category to every 6 years is reasonable and at worst would result in an insignificant increase in total plant risk. By every indication from both engineering judgment and risk insights, the selected test interval increase for less safety significant components is prudent and l

the overall change to the IST program is believed to be safety neutral.

2 i

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'4-4

l 5.0

.EEEERENCES 1.

ASME Bioler and Pressure Vessel Code,Section XI 2.

Letter dated September 9,1991 from James E. Richardson of the NRC to Forrest T.

I Rhodes of ASME.

i

. 3.-

Nuclear Regulatory Commission,"Use of Probabilistic Risk Assessment Methods in

- Nuclear Regulatory Activities; Final Policy Statement," Federal Register, Vol. 60, No.

158, August 16,1995.

4.

NUMARC 93-05, " Guideline for Optimizing Safety Benefits in Assuring the

. Performance of Motor-Operated Valves," Nuclear Management and Resources Council, Inc., December 1993.

5.

NUMARC 93-01," Industry Guideline for Monitoring the Effectiveness of Maintenance 2

at Nuclear Power Plants," Nuclear Management and Resources Council, Inc., May 1993.

Electric Power Research Institute, "PSA Applications Guide," TR-105396, Project l

6.

3200-12, Final Report, August 1995.

.3 h

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5-1

s Table 4-1

$==

y of Ruk Rankas Results for IST Castponese Sorted By IST Plan C7 Plan Risk initialIPE Outage Risk Seismic Risk CDF Ranking Final Ranking Table Component Description Fussell-Vesely

Achievement Ranking Based Ranking Ranking Changes Based On IST

'9 NumW Worth

on FV "

Cnanges Changes wfout CCF Study Number MOTOR DR!VEN AUJut4ARY FEEDWATER Table O CP1-AFAPMD41 (1)

PUMP 141 0 0282 2 8296 High No change No change No @Juige No change High No Change High MOTOR DRIVEN AUXittARY FEEDWATER Tatso O CP1,AFAPMD42 PUMP 142 0 0394 3 3020 High No change No change No &ange Nochange Hgh No Change High TURBINE DRNEN AUAILLARY Tabio O CP1-AFAPTD-01 FEEDWATER PUMP 141 0 2351 12 9035 Hgh No change No change Nochange No change Hegh No Change Hgh COMPONENT COOLING WATER PUMP 1 Tatde O CP1-CCAPCC-01 01 0 0366 4 8323 High No change No change No change No change High No Change High COMPONENT COOLING WATER PUMP 1-Tatdo O CP1-CCAPCC-02 02 0 0303 38 5384 High No change H3 change No change No change High No Change Hgh SAFETY CHILLED WATER REClRC PUMP Table O CP14HAPCP45 145 0 0080 1T2TS Medium No change No change No change No change

' Medium No Change Hgh SAFETY CHILLED WATER RECIRC PUMP TatWe O CP1-CHAPCP-06 (2)(4 1-06 0 0003 1.3459 Medium No dange No @ange No mange No change Medium No Change High Tatde O CP1-CTAPCS41 CONTAINMENT SPRAY PUMP 141 Wa wa None No change Nochange No dev No change n/a No C-.p Hsgh Table O CP1-CTAPCS-02 CONTAINMENT SPRAY PUMP 142 wa Na None No change No J-p No change No mange n/a No Cnange High Iatde O CP1-CTAPCS43 CONTAINMENT SPRAY PUMP 1-03 n/a Na None No change No change No Gep No d=p Wa No Change Mgh Tatae O CP14TAPCS-04 CONTAINVENT SPRAY PUMP 144 n/a n/a None No change No change No change No C.-p nra No Change Hgh Table O CP100APRM41 REACTOR MAKEUP WATER PUMP 141 n/a n/a n/a n/a n/a n/a No change rva Low Lew DtESEL GENERATOR 141 FUEL Oil TatWe O CP1-DOAPFT41 TRANSFER PUMP 141 0 0478 140 0001)

Hgh No change No change No mange No change None Deceased Low DtESEL GENERATOR 1-01 FUEL OIL Table O CP1-DOAPFT-02 TRANSFER PUMP 142 0 0478 140 0000 High No change No change No change No change None Decreased Low OtESEL GENERATOR 1-02 FUEL OIL Tatdo O CP1-DOAPFT-03 TRANSFER DUMP 143 0 0478 140 0000

*Wh No change No change No change No change None Decreased Low OtESEL GENERATOR 142 FUEL Ott TatWe O CP100APFT-04 TRANSFER PUMP 144 0 0478 140 0000 High No change No change No change No change None Decreased Low TatHe O CP1-SWAPSW41 STATION SERVICE WATER PUMP 141 0 0969 T7 6709 High No change No change No mange No change Hgh No Change Hegh Tatdo O CP1-SWAPSW42 (1)

STATION SERVICE WATER PUMP 142 0 0386 107 0000 H6gh No dange No cbange No change No change High No Change Hgh SAFEGUARD BUILDING SUMP 141 Pt IMP TatWe O CP1-WPAPSS-01 141 We n/a Na da We nfa No d eige n/a low Low SAFEGUARD ButLDING SUMP 1-01 PUMP Table O CP1-WPAPSS-02 142 n/a n/a We We n/a n/a No change n/a Low Low SAFEGUARD BUILDING SUMP 142 PUMP Table O CP1-WPAPSS-03 143 rwa nts n/a n/a n/a n/a No change n/a Low Low SAFEGUARD BUILDtNG SUMP 142 PUMP Yable O CP1-WPAPSS-04 144 n/a n/a n/a n/a We n/a No change nfa low Low Tatdo O CPX-DDAPRM41 REACTOR MAKEUP WATER PUMP K41 n/a n/a da Wa rVa Na No change n/a low Low SPENT FUEL POOL COOLING WATER TatWe O CPX-SFAPSF-01 PUMP X-01 n/a n/a n/a n/a n/a n/a No change n/a Low Low SPENT FUEL POOL COOLING WATER TatWe O CPX-SFAPSF-02 PUMP X42 n/a n/a n/a n/a n/a n/a No chwige n/a low Low Table 0 TBX4SAPBA-01 BORIC ACID TRANSFER PUMP 141 n/a Na None No change No change No change No change n/a No Changa Hgh Tatde 0 TBX-CSAPBA-02 BORIC ACID TRANSFER PUMP 1-02 n/a n/a None No change No change No change No change n/a No Change High Tat 9e 0 TBX4SAPCH-01 (1)

CENTRlFUGAL CHARGtNG PUMP 141 0 0125 1 $301 High No change No change Na change No change High No Change High Table 4-1 Page I

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6 t D M m H AS AS AS AS P P A A A5 A5 A5 A5 P H H H H MO MR T L 5 T T C C C S S o T T6 T8 T6 Ts T T T T T 5 5 T 5 T N N T C M M6 M6 M6 Me M M MV MV MV MV MV MVM M TV T T V SL SL C U V NW N NW S S R C d T T T L T N V N V TW Ai T T TW 3 2 Tn T L T L W T T T T 3 T E S CM A N NK NK NK NK N N N V N E E A E N N NR N E VL P R R P A F P C C2 C2 C2 C2 C V C CL CL CL CL CL C L CP C VL V V O C1 C1 W 0 0 0 0 O O t O 1 O 1 O O M O 1 L 1 t t 1 1 1 t t L t 1 L 1 t U u5 U5 u5 1U5 UV U uS uS UIS UIS UIS FIS UD U UIS UV uS R0 R0 L UR 5 1 5 t 5 5 I I I 3 3 4 5 6 7 9 9 6 7 8 9 0 1 2 3 2 3 4 1 2 0 0 4 4 4 4 4 4 4 5 5 5 5 6 6 6 6 8 8 6 1 1 0 0 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 0 0 0 3 3 1 5 l 5 5 5 5 5 5 5 5 5 5 5 5 5 5 4 4 6 7 7 4 n 5 V P W V-N N V-W- V N N V V V V N N V V V V C. S a / 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 4 4 M 4 4 H 4 4 4 4 4 4 4 H 4 P T 4 4 4 4 4 4 4 4 4 4 4 4 4 4 4 1 1 1 1 1 1 SIy n Pb_ te o o o e o o o e e o e e o o o o o o o e o lae 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 B l 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 de a d d d d d d d d d d d d d d S a t t ta ta t t t n d d ta t t ta t ta t d ta t d T b t ta a a a a a t d n tr T t t a a a a a a a a a o I T T T T T T T T T T T T T T T T T T T T T T S

1! TaNe 4-l Summary of Rash Ranksg Resuhs for IST Cesaposents Sorted ey tst pian CT Plan Risk InfuelIPE Outage Risk Selsenic Risk CDF Ranking Final Ranking '8 Talk Component Description Fusset-Vesety

ACC _ : 2 Ranking Based Ranking Ranking Changes Based On IST Number Worth
onFV" Changes Changes utout CCF Study C

PRZR 1-01 LeQ SPACE SMPL LN ORC RLF Table 17 1 # S 4501 VLv n/a n/a We n/a Wa n/a No change n/a low Low PRZR 141 STM SPACE SMPL LN ORC Table 17 1PS-0502 RLF VLV n/a n/a n/a Wa n/a n/a No chwige n/a low Low RC LOOP 141/144 HL SMPL LN ORC RLF Tabao 17 1PS4503 VLV We Wa n/a Wa Wa n/a No mange n/a low Low CNTMT PERS ARLOCK 1-01 EAT DOOR Table 17 IBS4015 MAN EQUAL VLV 0015 n/a Wa wa We n/a n/a No change n/a Low Low CN1MT PERS art OCK 141 EXT DOOR Tabao17 1BS4025 AUTO EQUAL VLV We n/a Wa We nra n/a No change Wa low Low CNIMT PERS A4RLOCK 1-01 EXT DOOR ' n/a low Low Table 17 1BS4029 MAN EQUAL VLV 0029 We Wa We n/a n/a n/a No change CNTMT PERS ARLOCK 141 INT DOOR Tabee 17 1BS4030 AUTO EOUAL VLV n/a n/a n/a n/a We n/a No change We Low Low CNTMT PERS AtRLOCK 1416NT DOOR Table 17 1BS4044 MAN EOUAL VLV 0044 n/a We n/a nte n's We No change Wa Low Low CNTMT PERS AIRLOCK 1414NT DOOR Tabte 17 1BS4056 MAN EQUAL VLV 0058 n/a Wa n/a n/a n/a n/a No change n/a low Low U1 CNTMT PERS EMER ARLOCK INT Tatsa 17 1BS4202 DOOR MAN EQUAL VLV n/a Wa n/a Wa n/a n/a No change n/a low Low U1 CNTMT PERS EMER A6RLOCK EXT Table 17 1BS4203 DOOR MAN EOUAL VLV Wa n/a n/a n/a Wa n/a No change n/a low Low Table 17 1CA4016 U1 CNTMT SERV AIR HOR CHK VLV n/a Na Wa n/a n/a n/a No change n/a low Low Table 17 1CH4024 U1 VENT CH WTR SPLY RC CHK VLV n/a n/a n/a We nfa We No mange We Low Low U1 CNTMT VENT CH WTR SPLY HDR Table 17 1CH4271 ORC PRESS RLF VLV Wa de We n/a Wa n/a No change n/a low Low U1 CNTMT VENT CH WTR RET HDR ORC Table 17 1CH-0277 PRESS RLF VLV Wa We We n/a Wa We No dange n/a Low Low Table 17 1Cl-0030 U1 INST AR HDR TO U1 CNTMT CHK VLV n/a n/a None No change No change No mange No dange n/a No Change Low Table 17 IWP-7178 LWPS RCDT 141 DRN HOR RLF VLV We Wa n/a n/a n/a Wa No change Na Low Low RC PASS SMPL RET TO RCDT 141 RLF Table 17 1WP-7177 VLV nfa n/a We n/a n/a n/a No @ange n/a low Low TnWe 4-1 PnSe 27

m se J l 4 1 1 e 1 1 L ENCLOSURE 4 TO TXX-95260-}}

ML20094P782

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SUMMARY

1.0 BACKGROUND

SUMMARY

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