ML20135F833
| ML20135F833 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 03/12/1997 |
| From: | Polich T NRC (Affiliation Not Assigned) |
| To: | Terry C TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| Shared Package | |
| ML20135B349 | List: |
| References | |
| TAC-M94165, TAC-M94166, NUDOCS 9703140160 | |
| Download: ML20135F833 (13) | |
Text
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UNITED STATES y
g NUCLEAR RE2ULATORY COMMISSION WASHINGTON, o.C. 30000-4001 March 12, 1997 Mr. C. Lance Terry l
TU Electric Group Vice President, Nuclear Attn: Regulatory Affairs Department i
P. O. Box 1002
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Glen Rose, TX 76043
SUBJECT:
SECOND ROUND REQUEST FOR ADDITIONAL INFORMATION (RAI) ON RISK-INFORMED INSERVICE TESTING (RI-IST) PILOT PLANT - COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2 (TAC NOS. M94165 AND M94166) i
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REFERENCES:
1.
TU Electric letter logged TXX-95260, from C. L. Terry i
to the NRC, dated November 27, 1995 t
2.
NRC letter from Timothy J. Polich to C. Lance Terry, dated March 15, 1996 3.
TU Electric letter logged TXX-96371, from C. L. Terry
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to the NRC, dated June 3, 1996 4.
TU Electric letter from Hossein G. Hamzehee to Mike Cheok, dated July 2, 1996 1
5.
TU Electric letter logged TXX-96458, from C. L. Terry to the NRC, dated September 12, 1996
Dear Mr. Terry:
On November 27, 1995, Texas Utilities Electric Company (TV Electric) submitted a request to the Nuclear Regulatory Comission (NRC) (Reference 1) to utilize a risk-informed inservice testing (RI-IST) program at the Comanche Peak Steam Electric Station (CPSES), Units 1 and 2, to determine inservice test frequencies for certain valves and pumps that were categorized as low safety significant. The request was part of a pilot plant effort with Arizona Public Service Company (APS). The NRC staff provided an initial request for additional information (RAI) to TU Electric related to the proposed RI-IST program via Reference 2.
The NRC staff met with TU Electric at the CPSES site on April 25, 1996, to discuss the RAI. TU Electric responded to the NRC staff's initial RAI and submitted a revised RI-IST program to the NRC via
~ Reference 3.
TU Electric submitted additional information to the NRC in support of their proposed RI-IST program via References 4 and 5.
The NRC staff used the information provided by both pilot plant licensees to help develop a draft RI-IST Regulatory Guide (DG-1062) and draft Standard Review Plan section (SRP Section 3.9.7).
Enclosed is an additional RAI aimed at determining the extent to which the RI-IST program proposed by TU Electric is consistent with the guidance being considered by the staff in the draft RI-IST Regulatory Guide and Standard Review Plan section (which will soon be made available for public comment). Several of the questions in the RAI were based on information provided to the NRC by Oak Ridge National Laboratory hh hh $@@ @@
9703140160 970312 PDR ADOCK 05000445 P
n Mr. C. Lance Terry ;
. 6 (ORNL) in letter reports. These letter reports (Attachments to the Enclosure) document ORNL's review of Nuclear Plant Reliability Data' System (NPRDS)
- failure records associated with CPSES..
Because of ongoing work on the draft rOk-informed Regulatory. Guides aiid-Standard Review Plan sections, the. staff may need to. ask the' pilot' plant licensees questions in addition to those contained:in the attachment to this letter. These additional questions may relate to the! policy 11ssues discussed in the January 22, 1997, Staff Requirements Memorandum.
It is;antici)ated that the final RAI will be sent to the RI-IST. pilot!plarit licensees stortly after the draft RI-IST regulatory guld.e (RG) andsstandard review plan'(SRP)J are sent out for public comment. While we regret,that,a comprehensive.. set pf
.RAIs cannot be provided to the pilot plant; licensees at this time,.we are,
confident that significant progress'wil1 continue to be made towards4 >
implementing RI-IST programs at CPSES. t -
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Sincerely, '
c.
A-ORIGINALr' SIGNED BY:
. Timothy J.'Polich, Project Manager
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Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear _ Reactor Regulation Docket Nos. 50-445 and 50-446
Enclosure:
Request for Additional Information with Attachments (3) cc w/ encl: See next page DISTRIBUTION:
Docket OGC PUBLIC JRoe GHill (4)
ACRS TPolich (2)
AHowell, RIV PDIV-1 r/f CHawes (2)
CGrimes EAdensam (EGA1) l!
Document Name:
CP94165.RAI 0FC TM/PD4-1 (A)LA/PD4-1 W CHawes 0/UH
}!
/
'NAME TPolich/vw DATC 3 /@/97 3/Ih97 COPY YES/NO YES/N0 0FFICIAL RECORD COPY 140035 2 ".
i Mr. C. Lance Terry '
(ORNL) in letter reports. These letter reports (Attachments to the Enclosure) document ORNL's review of Nuclear Plant Reliability Data System (NPRDS) failure records associated with CPSES.
l Because of ongoing work on the draft risk-informed Regulatory Guides and Standard Review Plan sections, the staff may need to ask the pilot plant licensees questions in addition to those contained in the attachment to this letter. These additional-questions may relate to the policy issues discussed in the January 22, 1997, Staff Requirements Memorandum.
It is anticizated that the final RAI will be sent to the RI-IST pilot plant licensees s tortly after the draft RI-IST regulatory guide (RG) and standard review plan (SRP) are sent out for public comment. While we regret that a comprehensive set of RAls cannot be provided to the pilot plant licensees at this time, we are confident that significant progress will continue to be made towards implementing RI-IST programs at CPSES.
Sincerely, A
Timothy J.
olich, Project Manager Project Directorate IV-I Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket Nos. 50-445 and 50-446 l
Enclosure:
Request for Additional Information with Attachments (3) cc w/ enc 1: See next page
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Mr. C. Lance Terry TU Electric Company Comanche Peak, Units 1 and 2 l
cc:
Senior Resident Inspector Honorable Dale McPherson i
U.S. Nuclear Regulatory Commission County Judge P. O. Box 1029 P. O. Box 851 Granbury, TX 76048 Glen Rose, TX 76043 Regional Administrator, Region IV Office of the Governor i
U.S. Nuclear Regulatory Cosnission ATTN: John Howard, Director 611 Ryan Plaza Drive, Suite 400 Environmental and Natural Arlingcon, TX 76011 Resources Policy P. O. Box 12428 Mrs. Juanita Ellis, President Austin, TX 78711 i
Citizens Association for Sound Energy 1426 South Polk Arthur C. Tate, Director Dallas, TX 75224 Division of Compliance & Inspection Bureau of Radiation Control Mr. Roger D. Walker Texas Department of Health i
TU Electric 1100 West 49th Street Regulatory Affairs Manager Austin, TX 78756-3189 P. O. Box 1002 4
j Glen Rose, TX 76043 i
Texas Utilities Electric Company c/o Bethesda Licensing i
3 Metro Center, Suite 610 Bethesda, MD 20814 j
George L. Edgar, Esq.
Morgan, Lewis & Bockius 1800 M Street, N.W.
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Washington, DC 20036-5869 l
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COMANCHE PEAK STEAM ELECTRIC STATION. UNITS 1 AND 2 i
f DDCKET NOS. 50-445. AND 50-446 3
REQUEST FOR ADDITIONAL INFORMATION REGARDING RISK-INFORMED INSERVICE TESTING (RI-IST) PILOT PLANT 1
The following are the supplemental questions and coments that have been developed by several Nuclear Regulatory Commission (NRC) staff reviewers who have bee. valuating the proposed risk-informed inservice testing (RI-IST) program for Comanche Peak Steam Electric Station (CPSES).
l These questions and cosmients are comprised of two parts:
(1) additional questions that remain from our review of the responses received for the first set of Request for Additional Information (RAls), and (2) relatively new l
issues that have been identified during the ongoing development of the NRC's guidance documents on risk-informed regulations. We recognize that for some i
of these questions, the licensee's response to the first RAls provided part of the answer that is being sought, and we.6ncourage the licensee to refer to the previous RAI responses where appropriate.
1.
What components are the major contributors to the change in core damage frequency (aCDF) associated with the proposed RI-IST program at CPSES?
What testing or other measures (duce risk) can the Vicensee take toincluding activities that would tend to re j
i reduce the negative impact on ACDF? Can the licens6e make a more realistic estimate of the aggregate effect on CDF of the proposed RI-IST program (i.e., as opposed to a " conservative" estimate where component failure rate (A) is linearly extrapolated)? This reassessment should include an identification of potential areas in which there was an overly conservative treatment in the quantification of both the baseline PRA risk levels and the change in risk associated with the proposed RI-IST program.
Qualitative information should be provided for those areas that cannot be quantified.
l 2.
Does the licensee's PRA assume that the current Code-required testing is 100% effective in assessing a components operational readiness? Does i
l the current Code-required testing provide adequate information relative to the failure modes modeled in the licensee's PRA (e.g., failure of a valve to remain open for a 24-hour period)? What consideration has been given to test effectiveness in establishing the proposed risk-informed IST program? Are any PRA model or test strategy adjustments warranted?
I 3.
On a component-s >ecific basis, the licensee should identify each instance where tie proposed IST program change will affect the current licensing basis of the plant (e.g., commitments made in response to NRC general letters (GLs) such as GL 89-10. TMI action plan items, components relied on by the staff in concluding that the system and i
plant designs were acceptable). These commitments, which may be incorporated into plant procedures, may not be modeled in the licensee's PRA. The licensee should identify the source and nature of the j
commitment (or requirement), and document the basis for the ENCLOSURE 4
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acceptability of the proposed change. The licensee should consider the original acceptance conditions, criteria, and limits as well as the risk significance of the component. Consideration should also be given to i
diversity, redundancy, defense in depth, and other aspects of the General Design Criteria.
If the licensing basis is not affected by the proposed IST program changes, the licensee should so indicate in its j
risk-informed IS" Program description.
i 4.
Provide any component-specific exemption requests, technical specification amendment requests, and relief requests necessary to implement the proposed RI-IST program. Has the licensee submitted evised relief requests for high safety significant components (HS5Cs) that were the subject of previously approved relief requests? These relief requests should be reevaluated in light of the components risk i
significance. Has the licensee submitted relief requests for high and i
low safety significant components (LSSCs) not tested in accordance with the Code test method requirements or methods described in an NRC
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endorsed Code Case? Has the licensee submitted relief requests for HSSCs that will not be tested in accordance with the Code test frequency
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requirements?
i 5.
Please provide a detailed risk-informed IST Program implementation plan l
(i.e., for both HSSCs and LSSCs). This implementatbn plan should 1
contain details on how each component, or group of components, catego-i rized as being LSSC, will have its test interval extended. For example, l
the staff needs to see a detailed description, or draft procedure, i
documenting how component test intervals will be extended in a step-wise j
manner (i.e., not just the " speed limit" test interval). The implemen-tation plan should describe how various component groupings were i
selected (e.g., using the guidance contained in NRC GL 89-04, Position 2 l
for check valves; Supplement 6 to NRC GL 89-10 and Sectfs 3.5 of ASME l
Code Case OMN-1 for motor-operated valves). The implew tb lon plan i
should document how the licensee proposes to use past p/.rfo sance, service condition, etc. in establishing the test strasg %r specific j
components (See question 10 below).
If the licensee wanM to take credit for other operations and maintenance activities to justify less l
frequent inservice testing, then the details of these other activities and how they relate to the IST strategy needs to be described explicitly.
6.
Please identify (if any) human actions that were used to compensate for t
a basic event probability increasing as a result of a test interval extension, specify the human failure probability used and describe how the licensee will ensure performance at this functional level.
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l 3-7.
How specifically will each of the following factors be considered by the licensee's integrated decision making process to establish an appropriate test strategy (i.e., test frequency and test method) for components:
i past performance history, j
service condition, l
design, and safety significance?
j a
j In response to a similar RAI (i.e., El-5), TU electric indicated that "For the IST study, the expert panel had available to it infomation j
pertaining to plant components and this was considered by the panel in the course of its evaluation. Thus, performance history was considered in the determination of a connonents safety sianificant classification
[ emphasis added)". The response also discussed how future performance might be used to adjust the test interval (and classification), but it l
does not describe how these factors were explicitly considered in l
establishing each component's test interval.
i Either describe in detail the process that was used by TU Electric to factors these variables into the test strategy determination or propose a process. The staff recognizes that, to some extent, these factors are i
embedded in the models and data supporting the licensee's PRA. However, the staff expects licensee's to augment its PRA with a component-specific evaluation of perfomance, conditions, and design to arrive at an appropriate test strategy (including test interval).
8.
A November 22,1996, (Attachment 1) letter report from A. B. Poole j
(ORNL) to J. E. Jackson (NRC) states (in part) that:
1 The failure rate for [all) pumps and related equipment is i
significantly higher at PWRs than BWRs with the exception of l
electric motors.
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Electric-driven pumps at PWRs have a 40% larger failure rate than at BWRs and turbine-driven pumps at PWRs have a 86% larger failure rate j
than at BWRs.
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Required Regulatory / Code inspections only found approximately 1/3 of the significant failures of pumps and related equipment.
..., the PWR pump trend shows a failure rate increase beyond 15 i
years to a new linear aging rate (a) of approximately 25% increase j
per year.
I These increased aging rates for older PWR pumps certainly indicate
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the need for consideration of special IST intervals for this portion
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of the pump population. Unless specific plant maintenance records can support a lower aging trend, the unavailability of PWR pumps 215 years in age should be evaluated using a value of e=0.25.
The report also showed higher than average failure rates for pumps in certain systems (e.g., essential service water pumps at both PWRs and BWRs have pump failures about 70% above the average). Failure rate breakdown for safety-related pumps and related equipment (1994-1995) was provided as follows:
PWR All pump hardware:
1 PWR Signif fail.
Years Hours Rate f/hr Pumps 56 2812 24633120 2.3E-06 l
Motors 14 2403 21050280 6.7E-07 Turbine 25 146 1278960 2.0E-05 1
drives 5
Circuit 62 2403 21050280 2.9E-06 breakers Overall for pump systems:
2.5E-05 How do the failure rates used in the CPSES PRA compare to those in the 4
table above? Please connent on the reasons for any significant differences. Do the failure rates used in the licensee's PRA take aging and unique system service conditions into consideration? If not, how does the licensee justify not considering these factors? How would the licensee's integrated decision making process use this information in adjusting the test strategy for pumps?
9.
A January 9, 1997, (Attachment ?) letter report from A. B. Poole (ORNL) to J. E. Jackson (NRC) indicates that ORNL did a review of the available NPRDS failure recor'is and performance data (i.e., data from 1990 to 1995) for check vab es at CPSES. Of the 33 failure records in the NPRDS database for CPSES check valves, all but 2 of the failed valves are included in the list of IST deferral candidates (i.e., LSSC).
Nine of the thirty-three failures involved repeat failures (considering bothunits). Two individual components (2FW-0013 and 200-0258) had repeat significant failures. Diesel Lube Oil valve DO-0258 had a total of three significant failures, considering both units. While the data did not provide statistically meaningful results and did not point to 4
any abnormal failure patterns or causes, the data could not be used to validate the proposed IST interval or to evaluate the effect of interval extension on check valve performance.
(More general studies by ORNL seem to imply that many check valves should receive IST at intervals of 4
2-4 years.)
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Of the 27 internals-related failures,16 involved some type of age-i related failure mechanism such as debris accumulation or wear. Of the i
16 sinnificant failures, 8 were age-related. - This would tend to quest'on the technical validity of extending the test interval for certain check valves without assurance that corrective actions have been taken and/or inclusion of condition monitoring on at least some of the check valves.
Please describe the nature of and corrective actions associated with these failures (e.g., the five component cooling water (CCW) stop check a
valves that were sticking closed due to..." corrosion product accumulation between plug and bore during long periods of inactivity.
l Periodic stroking of valve [s] was less than adequate. Quarterly
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stroking of [the) valve [s] had to be initiated to prevent recurrence.").
l How was this information used as input to the risk-based calculations?
Were the unavailabilities of all check valves in a>plications susceptible to aging increased simultaneously by tie appropriate factor i
to cover the simultaneous effects of aging? If not, how will the j
licensee ensure that the impact on risk of unmitigated aging remains acceptably low?
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10.
The CCW stop check valves mentioned in 12 above (i.e., CC-0646, -0657,
-0687, -0694; 100-1075, -1076, -1077, -1078; 20C-0371, -0372, -0373,
-0374) are 2-inch, Category C, RCP Thermal Barrier Rupture Isolation
. Check Valves. According to the CPSES IST Program Plan, these valves are full stroke exercised during cold shutdowns. The licensee's IST Program states "These valves cannot be full-stroke exercised during plant operation because isolating the thermal barrier coolers to perform reverse flow testing would cause thermal overpressurization of the coolers and piping due to pump heat and challenge the relief valves on i
the system.
Part-stroke close exercising is not applicable".
It would appear that because " Quarterly stroking of [the) valve [s] had to be f
initiated to prevent recurrence." (As noted in the NPRDS entry) it is practicable to at least part-stroke open these valves at power (in accordance with'OM-10 5 4.2.1.2(b)). Are there other valves at Comanche Peak where testing at power (i.e., quarterly, full-stroke, or part-stroke) is practical and the licensee has decided to defer testing? The NRC staff was under the impression that components with poor performance histories at CPSES would not be categorized as LSSC.
According to the licensee's RI-IST program submittal, these ' valves have a risk achievement worth (RAW) of approximately 6.
Since these valves appear to be subject to a common mode failure mechanism, how can the
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licensee justify the proposed 6-year test interval?
l 11.
A January 8,1997, (Attachment 3) letter report from A. B. Poole (ORNL) to J. E. Jackson (NRC) indicates that ORNL reviewed NPRDS failure records for low risk significant motor operated valves (MOVs) (i.e.,
1 data from 1990 to 1995) at CPSES. This letter report states that " data j
examined are insufficient to support the extension of the in-service i
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- I test interval to 6 years without the inclusion of other information such as frequency of operation, preventive maintenance practices and l
scheoules, and operating environment". While a failure rate for MOVs derived from NPRDS data since 1990 may be overly conse'evt.tive (i.e.,
because it does not adequately reflect improvements to 'dOVs made as a result of GL 89-10), it may also be non-conservative (i.e., because MOV i
testing has not typically evaluated MOV performance under dynamic i
conditions). Describe the basis for the selection of failure rates used i
in the licensee's PRA. How were these failure rates adjusted based on l
plant-specific experience and operating environment?
12.
The licensee should describe in detail its performance monitoring plan and explain how sufficient data will be developed to facilitate PRA and risk-informed IST Program updates. Will there be sufficient monitoring of both HSSC and LSSC to support the periodic updates? As noted in RAI 4
- 1, have the components that contribute most to risk increase been identified and a monitoring program specifically planned that could be used to modify assumed failure rate data that is currently either under or overly conservative?
Does the proposed performance monitoring process ensure:
1 enough tests are included, over gradually extending time peric'ds, to provide meaningful data; incipient degradation is likely to be detected and corrective action i
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taken; and appropriate parameters, as required by the ASME Code or ASME Code case, are trended as necessary to provide validation of the PRA?
Does the proposed performance monitoring process ensure that degradation is not significant for components that are placed on an extended test interval, and that failure rate assumptions for these components are not compromised by test data?
13.
Does the licensee's corrective action program:
a.
Comply with 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action?
b.
Evaluate IST components that fail to meet the test acceptance criteria as well as IST components that are otherwise detemined to 4
i be in a nonconfoming condition (e.g., a failure or degraded condition discovered during normal plant operation).
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c.
For each cosponent failure:
i (1) couply with 10 CFR Part 50, Appendix B, Criterion XVI, i
Corrective Action, (ii) determine the impact of the failure or nonconfoming j
condition on system / train operability since the previous j
- test, (iii) determine and correct the root cause of the failure or nonconforming condition (e.g., improve testing practices, repair or replace the component),
(iv) assess the applicability of the failure or nonconforming condition to other components in the IST program (including any test sample expansion that may be required for grouped components such as relief valves),
(v) correct other susceptible similar IST components as necessary, 1
(vi) assess the validity of the PRA failure rate and unavailability assumptions in light of the failure (s), and (vii) consider the effectiveness of the component's test strategy in detecting the failure or no'iconforming condition. Adjust the test frequency and/or iisthods, as appropriate, where the I
component (or group of components) experiences repeated failures or nonconforming conditions.
d.
Provide the licensee's PRA group with the corrective action evaluations so that any necessary model changes and re-grouping are l
done as might be appropriate.
Is any credit taken for the corrective action program in the PRA? If not, do you thirik that it i
is feasible and justified to do so?
l 14.
Are there any RI-IST program changes that the licensee proposes to make I
without prior NRC ap roval other than changes explicitly described by the licensee in RI-IST program submittals and approved by the staff i
(e.g., component categorization /re-cate]orization in accordance with an NRC approved methodology, gradual extension of a components test i
interval in a step-wise fashion as approved by the staff in its safety evaluation)? Does the licensee have an adequate process or procedures in place to ensure that RI-IST program changes of the following two types get reviewed and approved by the NRC prior to implementation:
Test method changes that involve deviation from the NRC-endorsed Code requirements.
Changes to the risk-informed IST program that involve process changes (e.g., changes to the plant probabilistic model assumptions, i
changes to the grouping criteria or figures of merit used to group 4
components, changes in the acceptance guidelines used by the t
j licensee's integrated decision-making process (e.g., expert panel)).
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l 15.
Does the licensee's RI-IST program test components in the HSSC category 4
that are not in the licensee's current IST program commensurate with their safety significance? These components should be tested in accordance with the ASME Code where practical, including compliance with all administrative requirements. Where ASME Section XI or 0&M testing is not practical, has the licensee proposed alternative test methods to i
ensure operational readiness and to detect component degradation (i.e.,
degradation associated with failure modes identified as being important l
in the licensee's PRA)?
16.
Are IST components in the RI-IST program (with the exception of check i
valves) exercised or operated at least once every refueling cycle? Are i
components in the following categories exercised more frequently than once per operating cycle, if practical:
i a) Components with high risk significance; b) Components in adverse or harsh environmental conditions; or l
c) Components with any abnormal characteristics (operational, design, or maintenance conditions)?
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17.
How does the licensee plan to address, or deal with, the synergistic effects of implementing its risk-informed IST progra.m and other risk-informed initiatives? How does the licensee plan to maintain the level i
of commitment of plant resources (e.g., Quality Assurance or
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maintenance) that was assumed in justifying extended IST intervals?
18.
By letter dated January 31, 1997, TU Electric submitted Revision 8 to the CPSES Units 1 and 2 IST Plan for Pumps and Valves. This revision added several relief valves to the scope of the CPSES IST program. Will the relief valves added to the CPSES IST program by Revision 8 be added to the scope of the licensee's proposed RI-IST program? Will the RI-IST program scope have any other additions or deletions?
e i
19.
Does the licensee have procedures for conducting the periodic risk-
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informed IST program review to ensure that it:
Prompts the licensee to conduct overall program assessments
. periodically (i.e., at least once every two refueling outages) to i
reflect changes in plant configuration, component performance, test results, industry experience, and to reevaluate the effectiveness of j
the IST program, prompts the licensee to compare actual component performance to predicted levels to determine if component performance and conditions are acceptable (i.e., as compared to predicted levels).
If performance or conditions are not acceptable then the cause(s) j should be determined and corrective action implemented,'
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prompts the licensee to review and revise as necessary the assumptions, reliability data, and failure rates used to group j
components to determine if component groupings have changed, and rompts the licensee to reevaluate equipment performance (based on xth plant-specific and generic information) and test effectiveness j
to determine if the inservice test program should be adjusted i
(Plant-specific data should be incorporated into the generic data using appropriate updating techniques)?
1 Does the licensee have procedures to ensure that the results of its corrective action program for IST program components get fed back into j
its periodic IST program reassessment?
Does the licensee have procedures in place to identify the need for more emergent RI-IST program updates (e.g., following a major plant i
modification, or significant equipment performance problem)?
l 20.
To avoid being overly prescriptive in its guidance, yet still ensure l
that certain topics having major safety importance for all risk-informed programs are addressed in licensee's proposals, the staff has identified a set of five key safety principles in the draft risk-informed guidance i
l documents.
It is currently intended that the five key principles given below must be explicitly addressed in all licensee applications for i
risk-informed programs. The regulatory guides that are under j
development are to provide an example of acceptable means for satisfying these key principles. Because that guidance has not been finalized, it would be useful to have the pilot plant licensees describe how their proposed RI-IST program satisfies each of the following key safety principles:
i a.
The proposed change meets the current regulations.
[This principle i
applies unless the proposed change is explicitly related to a j
requested exemption or rule change.]
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b.
The defense in depth philosophy is maintained.
l c.
Sufficient safety margins are maintained.
I.
d.
Proposed increases in risk, and their cumulative effect, are small and do not cause the NRC Safety Goals to be exceeded.
I' e.
Performance-based implementation and monitoring strategies are proposed that address uncertainties in analysis models and data and l'
provide for timely feedback and corrective action.
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In addressing these principles, the licensee should describe how:
l All safety impacts of the proposed changes were evaluated on a j
j component-specific basis as well as in an integrated manner as part i
a b
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of an overall risk e nagement approach in which the licensee uses i
risk analysis to iiprove operational and engineering decisions i
broadly and not just to eliminate requirements that the licensee sees as undesirable. The approach used to identify changes in l-requirements should be used to identify areas where requirements l
should be increased as well as where they could be reduced.
l The acceptability of proposed changes should be evaluated by the licensee 1n an integrated fashion that ensures that all principles are met.3
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Core damage frequency (CDF) and large early release frequency (LERF) can be used as suitable metrics for making risk-informed regulatory decisions.
Increases in estimated CDF and LERF resulting from proposed current licensing bases (CLB) changes will be limited to small increments.
i The scope and quality of the engineering analyses (including traditional and probabilistic analyses) conducted to justify the proposed CLB change should be appropriate for the nature and scope
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of the changes proposed and should be based on the as-built and as-i operated and maintained plant.
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Appropriate consideration of uncertainty is given in analyses and interpretation of findings.
The plant-specific PRA supporting decisions has been subjected to j
quality controls such as an independent peer review.
Data, methods, and assessment criteria used to support the proposed IST program changes (e.g., those used by the licensee's expert i
panel) must be available for public review.
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1 Attachments: As stated (3) i i-i
' One important element of integrated decision making can be the use of an " expert panel." Such a panel is not a necessary component of risk-informed i
decision making; but when it is used, the key principles and associated j
decision criteria still apply and must be shown to have been met or to be irrelevant to the issue at hand.
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