ML20249C110

Submits Response to NRC 971223 RAI Re risk-informed ISI Pilot Program Submitted on 971031.Supporting Info & Calculations,Encl

ML20249C110
Person / Time
Site:	Surry
Issue date:	06/18/1998
From:	Hartz L VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML18150A471	List: ML20249C110
References
98-001, 98-1, NUDOCS 9806260026
Download: ML20249C110 (17)

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YlRGINIA EIECTRIC ANI> PowEn Cous ANv Ricniuoss>, VIRGINI A 2J261 June 18, 1998 l

United States Nuclear Regulatory Commission Serial No.98-001 Attention: Document Control Desk NL&OS/GDM R2 Washington, D.C. 20005 Docket No. 50-280 License No. DPR-32 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT 1 REQUEST FOR ADDITIONAL INFORMATION RISK-INFORMED INSERVICE INSPECTION PILOT PROGRAM in a letter dated October 31,1997 (Serial No.97-640), Virginia Electric and Power Company submitted a Risk-Informed inservice Inspection (RI-ISI) Pilot Program for NRC review and approval. The proposed program is an alternative to current ASME Section XI inspection requirements for piping. In a letter dated December 23,1997, the NRC made a request for additional information based on a preliminary review of the program submittal. Our response to the request is provided in the attachment.

Use of ASME Code Cases Westinghouse Owners Group Topical Report, WCAP-14572, Revision 1,

" Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report," referenced ASME Code Cases, N-416-1, N-498-1 and N-532 and noted licensees should consider requesting NRC approval for these Code Cases if they plan to implement a RI-ISI program. (The Code Cases are discussed on pages 192 and 220-221 of the WOG Topical Report.) ASME Code Cases N-416-1 and N-498-1 eliminate the need to perform elevated system pressure tests. Code Case N-416-1 was previously approved for use at Surry in a NRC letter dated October 14, 1994. A request for NRC approval to use ASME Code Case N-498-1 for Surry Unit 1 will be provided in a separate submittal. (This Code Case has already been approved for Surry Unit 2.)

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ASME Code Case N-532 provides for repor*:ng of examination and pressure test results I on needatoperiodic basis use Code Case forat ASME N-532 Code this time and, Class I,2 consequently, will and not be 3 items.

requesting We do not antic NRC approval for its use at Surry. However, as noted in our response to Question 2 in the attachment, we will continue reporting Class 1 and 2 examinations as required by I the Code NIS-1 form, as well as the high safety significant examinations regardless of l pq 9006260026 98061e F / / D PDR AOOCK 05000230 b ( O 'y .

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ASME class. The remaining ASME Code Cases previously approved for use at Surry remain valid for implementation of the RI-ISI program.

Technical Specifications The Surry Technical Specifications (TS) were reviewed to determine whether a TS change is required to permit the implementation of the RI-ISI program. Currently, TS 4.0.5 states, in part, the following:

"4.0.5 Surveillance requirements for inservice inspection and testing of ASME Code Class 1,2 and 3 components shall be applicable as follows:

a. Inservice inspection of ASME Code Class 1,2, and 3 components and inservice testing of ASME Code Class pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i)."

The Risk-Informed ISI Pilot Program has been submitted as an alternative to current ASME Section XI inspection requirements for piping as permitted by 10 CFR 50.55a(a)(3). Since alternatives to the ASME Section XI program are permitted by 10 CFR 50.55a, and TS 4.05 states that the Surry Power Station inservice inspection program is applicable as required by 10 CFR 50.55a(g), we do not believe a Technical Specifications change is required to implement the RI-ISI program.

If you have any questions or require additional information, please contact us.

Very truly yours,

~

L. N. Hartz Vice President - Nuclear Engineering and Services Attachment

- - - _ - - _ - _ - _ _ _-__ - - _ - _ - _ _ = _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ - __

cc: U. S. Nuclear Regulatory Commission . ,

, Region ll .  !

- Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85  ;

Atlanta, Georgia 30303 )

Mr. R. A. Musser i NRC Senior Resident inspector Surry Power Station Mr. Donnie W. Whitehead Risk Assessment and Systems Modeling  ;

1 Department 6412 Sandia National Laboratories Albuquerque, New Mexico 87185-0747

' Dr. Fredric A. Simonen Theoretical & Applied Mech. Group Battelle Pacific Northwest National Laboratories P.O. Box 999

- Richland, WA 99352-0999 Commitment Summary

1. . A request for NRC approval to use ASME Code Case N-498-1 for Surry Unit 1

' will be provided in a separate submittal.

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. 2. The current code required NIS-1 (IWA-6000) requires reporting of Class 1 and 2 examinations and documentation at the plant of Class 3 examinations. One  !

modification to this requi;ement will be a commitment to report the high safety  !

significant examinations regardless of class in addition to the normal NIS-1 )

reporting requirements. The modification will require a minor modification to our l NIS-1 procedure upon program approval. i

3. The NIS-2 form in conjunction with the Owner's summary report required by ASME Section XI (IWA-6000) will be used to report repairs or replacements. 4 These requirements will be applicable to the high safety significant piping locations regardless of ASME Class. Procedural control of these activities will be in place prior to implementation of the new program.

O 8 ATTACHMENT ,

REQUEST FOR ADDITIONAL INFORMATION NRC Question No.1 Provide the results of probability risk assessment (PRA) calculations used to develop the pipe segment ranking and estimate the change in risk due to implementation of the '

program. Also, provide the input and output of the SRRA program which develops the pipe failure frequency or probability as appropriate to the consequence model. A discussion of specific elements to be submitted can be found on page 49 of the Draft Regulatory Guide DG-1063. Submittal of the quantitative PRA and pipe failure information on a computer diskette, in addition to the hard copy, will greatly expedite the review.

Virginia Power Response Enclosed (Enclosure 1) for yo review is Calculation Number SM-1124, " Segment Definitions for Surry 1 Ri-ISI Program, Rev. 0," and Engineering Transmittal, ET No.

MAT-97-0014, " Estimated Failure Probabilities for Risk-Based ISI Surry Unit 1, Rev 0." q Additionally, tables expanding the information provided in Tables 3.3-3 and 3.3-4 of our i original submittal are enclosed (Enclosure 2), as well as information utilized in our SRRA calculations (e.g., input sheets, known failure history, and system engineer I comments). Electronic copies of Tables 3.3-3, 3.3-4 and the calculation spreadsheet used for Core Damage Frequency (CDF) without operator action have been sent to the Surry NRC project manager, Mr. G. E. Edison.

l NRC Question No. 2 Provide a discussion of the proposed monitoring and corrective action programs as described in Section 5.3 of the draft regulatory guide. The submittal should be specific enough to allow a staff review to determine whether the monitoring and corrective action program is adapted to plant-specific ISI program changes. l Va. Power Response:

No specific plant implementatic.n nocedures have been developed for the proposed Risk-informed ISI program. Necessary procedures and changes to existing procedures, if required, will be made prior to implementation of the program. The proposed monitoring and corrective act;on program will contain the following elements.

(References to the Code pertain to ASME Section XI periodically updated as required

~ by 10 CFR 50.55a, currently 1989 edition)

A. Identify - The Rl-ISI program establishes, in part, an inspection for cause program. The examinations performed address the most likely failure l

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mechanisms postulated for a particular location. The Surry Power Station Unit 1 Rl-ISI ' Pilot Submittal, Attachment 4, contains a database listing of the inspections to be performed. This database is maintained and can be updated as necessary. The selected locations represent the areas and volumes determined most likely to require examination under the risk-informed inservice inspection process. The use of existing examination procedures will address most of the required inspections. Any remaining inspections requiring examination procedure _ development or revision (e.g., examination volume expansion, new calibration blocks, etc.) will be completed during the inspection interval as required. These procedures will be approved through the existing approval process, which includes review by the Station Nuclear Safety and Operating Committee (SNSOC) as applicable. Acceptance standards are listed in Table 4.1-1 of the WCAP-14572 Rev.1, which for the most part are from the ASME Section XI Code. Examinations not specifically addressed in the WCAP l table will use acceptance standards per IWA-3000 of ASME Section XI.

B. Characterize - The requirement to characterize a condition remains essentially unchanged from the existing ISI program requirements established for the plant at this time. Relevant conditions and any unacceptable flaws identified will be addressed promptly for system / component operability considerations in accordance with Technical Specifications requirements. Reporting will routinely be addressed through the NIS-1 process of the current code as currently required. Expedited reporting will be done, as it is now, through the Licensee Event Report process, if required. The current code required NIS-1 (IWA-6000) requires reporting of Class 1 and 2 examinations and documentation at the plant of Class 3 examinations. One modification to this requirement will be a commitment to report the high safety significant examinations regardless of class in addition to the normal NIS-1 reporting requirements. This modification will require a minor revision to our NIS-1 procedure upon program approval.

C. Evaluato (1) Determine the cause and extent of the condition identified - Existing procedures and programs involving failure analysis, including engineering evaluations, will be used to evaluate the cause of an unacceptable condition.

These procedures address current ASME Section XI Code requirements and consider the cause of failure in determining the suitability of a repair or replacement. Upon determination, additional examinations will be conducted if required as described in Section 3.8 of our submittal. Procedural control of additional examinations will be in place prior to implementation of the new program.

l (2) Develop a corrective action plan or plans - An engineering evaluation will be performed of the service conditions and degradation mechanism to establish whether the affected piping will still perform its intended safety function during 2

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subsequent operation. The evaluation may include analytical evaluation as

. described in the Code, if applicable. If the piping does not meet the continued service criteria of the evaluation, then the piping will be repaired or replaced.

As previously stated, the cause of failure is included in the planning of the l repair or replacement. Information regarding the failure is also fed back into the RI-ISI process as new information. Ti.is is discussed in more detail in our response to question 4 below. The NIS-2 form in conjunction with the Owner's '

summary report required by ASME Section XI (IWA-6000) will be used to report repairs or replacements. These requirements will be applicable to the high safety significant piping locations regardless of ASME Class. Procedural control of these activities will be in place prior to implementation of the new program. The commitment here does not change any other aspect of our current ASME Section XI repair / replacement program (e.g., repair of non-class, low safety significant piping would not be reported on a NIS-2 form).

D. Decide - The approval of the appropriate corrective action plan will involve the appropriate levels of management as is currently practiced through procedural controls. Procedural modifications may be requ! red to include non-class high safety significant piping. These controls will be in place prior to the new program's implementation. l l

1. Implement - Procedural controls will be in place to assure implementation t' l

an approved corrective action plan. Successive examinations required for high l safety significant piping locations will follow the requirements of ASME Section XI, IWB-2420, regardless of ASME Class when analytical evaluations accepting flaws or conditions are used.

F. Monitor - Appropriate procedural controls will be put in place by management to ensure completion of corrective action plans. Corrective action plans may result in the total elimination of the problem with a repair or replacement, the temporary elimination of a problem using components with a finite service life (e.g., erosion / corrosion replacements), or return of the problem or condition l unexpectedly. To prevent the return of the problem or condition unexpectedly the affected location will be monitored. Monitoring the effectiveness of the corrective action may be in the form of another examination, monitoring the existence of conditions associated with the failure previously (e.g., thermal stratification temperature monitoring), plant walkdowns, or combinations of these activities. Procedures controlling these activities will be in place prior to implementing the new program.

G. Trend - The use of plant specific history (e.g., inspection results, corrective action plans, system engineer insights, etc.), owner's group and NRC bulletins, NRC generic letters, etc. will be used to periodically review and update the piping failure probabilities. The periodic review, as a minimum, will be in conjunction

with each ASME period (3,4,3 years). The new failure probabilities will be used 3

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along with other new risk information to update the risk ranking of the piping segments. Appropriate inspection changes as a result of the new rankings and expert panel interaction will be made accordingly.

l N_PC Question No. 3 Address the following items as described in " areas of review" specified in the draft Safety Review Plan (SRP) 3.9.8, Subsection 1.2.1.

. Traditional engineering analysis is needed to assess whether the impact of the proposed ISI changes are consistent with the principles of defense-in-depth.

. Adequate safety margins are maintained.

. The IS' Program changes meet or exceed the intent of ASME BPVC Code Section XI.

. There is no adverse impact to the augmented inspection programs, such as those for intergranular stress corrosion cracking (IGSCC) and erosion / corrosion.

Provide a description of such an engineering analysis method and the procedures j used, and a summary of the analysis results. Explain how the SRP goals stated above l are met. I Virginia Power Response: I a Traditional engineering analysis is needed to assess whether the impact of the proposed ISI changes are consistent with the principles of defense-in-depth.

Traditional engineering analyses are used in the RI-ISI program. Inputs to the program are the identification of structural mechanics parameters, possible degradation mechanisms, design limit considerations, operating practices and environments, and design and operational stress / strain limits. A failure effects evaluaboc is also conducted to identify the potential consequences of piping failure. Th:s includes an assessment of internal and external events, including resulting primary and secondary effects of pipe degradation. Secondary effects are evaluated through the assessment of pipe whip, flooding, and jet impingement (using information in the plant's Updated Final Safety Analysis Report, Chapter 14, Appendix B) and a plant walkdown. Traditional engineering analyses also include the evaluation of an indication, if found, through a flaw evaluation as discussed in the ASME Section XI Code (unchanged by the RI-ISI program).

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Defense-in-depth has traditionally been applied in reactor design and operation to provide a multiple means to accomplish. safety functions and prevent the release of radioactive material. As defined in the draft regulatory guide DG-1061, defense-in-depth is maintained by assuring that:

. a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation is preserved Characterizing the safety significance of piping on the basis of PRA insights reflects the balance between preventing core damage and consequence mitigation by directly addressing concerns regarding both CDF and large early release frequency (LErg),

Additionally, all piping segments are deterministically evaluated by the plant expert panel. These deterministic evaluations consider each piping segment's ability to cause initiating events, its potential use in mitigating design-base accidents, and its use in supporting other design basis analysis. The RI-ISI program should improve this balance by including piping that is currently exempt from examination or outside the scope of ASME Section XI. This can prove particularly useful in preventing or mitigating -

transients outside the traditional design-basis since it should result in a higher degree of confidence in the maintenance of structuralintegrity of the piping.

. over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided The_ RI-ISI program will not reduce design margins or defense-in-depth based on compensating programmatic activities.

. system redundancy, independence, and diversity are preserved commensurate with the expected frequency and consequences to the system l

No physical design changes are made that impact the system i redundancy, independence or diversity. The determination of the safety significance of piping segments is based on the expected frequency and consequences of challenges to the systems. The process includes piping currently outside the scope of Section XI that have been determined to provide a  ;

useful function for preventing or mitigating potential accidents.

In addition, deterministic questions asked as part of the plant expert panel process address the frequency and consequences of challenges of events outside the PRA model.

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I e defenses against potential common cause failures are preserved and the potential for introduction of new common cause failure '

mechanisms is assessed In the RI-ISI program, defenses against potential common cause failures are actually enhanced through the evaluation of the potential failure mechanisms and the examination of locations in which a likely failure mechanism, such as erosion-corrosion, is postulated. In addition, the monitoring and corrective action programs evaluate the impact. of potential common cause failures on like piping segments.

. independence of barriers is not degraded (the barriers are identified as the fuel cladding, reactor coolant pressure boundary, and containment structure)

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The implementation of RI-ISI will neither remove nor alter existing l physical barriers. In addition, the current levels of system redundancy and design will not be changed as a result of implementation of RI-ISI. With regard to the reactor coolant pressure boundary, the evaluation of the challenges to the reactor coolant system piping and the potential failure mechanisms of the piping allow for an explicit evaluation of this barrier. The RI-ISI program, including failure probability estimation, consequence evaluation, safety significance categorization, Perdue statistical model evaluation, selection of i NDE locations and methods, and monitoring and corrective  !

action programs, provides a measure of assurance equivalent to the current ASME Section XI program that the reactor coolant system boundary is not degraded. l

. defenses against human errors are preserved )

l Defenses against human errors are preserved with a risk-informed ISI program. No changes to the design, abnormal operating procedures'or emergency operating procedures are associated with the RI-ISI program. The RI-ISI program is being implemented according to the Nuclear Design Control l Program _ (NDCP). The NDCP incorporates engineering responsibilities and provides design control and document i control.

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m - Adequate safety margins are maintained.

As proposed,. the implementation of the RI-ISI program does not involve changing any acceptance criteria in the current licensing basis. The acceptance criteria currently included in the ASME Section XI Code for flaws is used, which includes margin between a flaw and a leak. Furthermore, the monitoring and evaluation of flaws that can lead to leaks provide safety margin, and the use of acceptable target leak frequency goals (used with the Perdue statistical model as part of the RI-ISI process) provides additional confidence that the RI-ISI program will not lead to more leaks than would be expected under the current ASME Section XI program.

m The ISI Program changes meet or exceed the intent of ASME OPV Code Section XI.

As stated in the draft SRP chapter 3.9.8, the intent of 10 CFR 50.55a is to maintain the structural integrity of piping in a nuclear power plant.

10 CFR 50.55a references the ASME BPVC Section XI for the detailed requirements. The ASME Section XI requirements for piping involve inspections performed on a sample basis with additional inspections, in terms of locations as 1

well as frequency, mandated in response to detection of flaws, The risk-informed ISI program meets the intent of 10 CFR 50.55a (i.e., maintain the structural integrity of piping) and also meets the intent of ASME Section XI.

The RI-ISI program is essentially a re-allocation of the inspections, specifically NDE examinations, performed on a sample basis. It reallocates inspections to areas of potentially. high risk. Risk is defined in terms of 1) the postulated consequences measured by piping CDF and LERF and 2) the likelihood of the consequences as measured by the postulated piping failure mechanism and estimated piping failure probability; a There is no adverse impact to the augmented inspection programs, such as those for intergranular stress corrosion cracking (IGSCC) and erosion / corrosion.

No changes to the augmented inspection programs are being made with the 1 proposed change to the ASME Section XI ISI program.

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NRC Question No. 4  !

Provide the implementation and monitoring program, which is an integrated part of the RI-ISI Program. As described in the draft SRP, the program should include l performance-monitoring strategies, inspection methods and acceptance guidelines, inspection result evaluation strategies, circumstances for corrective actions,

~ documentation requirements, administrative control, and quality control requirements to

- ensure meeting Appendix B to 10 CFR Part 50.

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s Virginia Power Response Implementation and. monitoring procedures will be in place prior to implementation of

.the new program. As described in'our submittal, Surry Power Station Unit 1 Risk-Informed Inservice Inspection Pilot Program Submittal, Section 5, the new program will be integrated into the existing ASME Section XI interval. Currently, Surry Unit 1 is in interval 3 period 2 The - ASME Section XI Code requires that a schedule of examinations be established that define in which period of the 10 year interval an

, examination will be performed. The. schedule provided in our submittal establishes a o similar schedule divided in thirds by period. If approval of the program occurs in the L 'second_ period and one refueling outage remains to allow examinations, then 66% of the required risk-informed program examinations would be performed in the remaining

- part of the interval, if approved in the third interval then 33% would be_ performed.- The E

. submittal contains the first period schedule also for documentation purposes. These examinations would not take place until the fourth interval since the 1st period of the

' third interval has already been completed. Examinations required by the current ASME LSection XI program for Categories B-F, B-J, C-F-1, and C-F-2 would be eliminated.

The applicable aspects of the Code not affected by this change would be retained, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements. Existing ASME Section XI program implementing procedures would be retained and would be revised to-i - address the RI-ISI process. Additionally the procedures will be revised to include the

.high safety significant locations in'the program requirements regardless of their current ASME class.

L' The RI-ISI program is a living program requiring feedback of new relevant information to l' ensure the appropriate' identification of high safety significant piping locations. A procedure will be developed to address changes to the PSA model and to failure l l probabilities. _ As a minimum,. risk ranking of piping segments will be reviewed and L adjusted on an ASME period basis. Significant changes may require more frequent l> adjustment'as directed by NRC bulletin or Generic Letter requirements, or by plant specific feedback. Procedural screening criteria for both PSA changes and new failure information will be developed to determine the significance of changes. PSA changes

_may -include PSA modeling changes, PSA assumption changes, plant procedure I

changes, equipment performance changes and plant design changes. New failure information may include new structural reliability computer code modeling, new failure input data for the computer code, new industry information on failures or new failure L mechanisms to consider. Changes due to feedback may be applied at the system, L piping: segment,_ or piping structural element portion of the RI-ISI process as b appropriate. Additions to the inspection program may be done without expert panel involvement (based on new numerical results). Deletions will require expert panel concurrence. 1

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NRC Question No. 5 According to Subsection 11.2.3.1 of SRP 3.9.8, sensitivity studies should be performed which identify the impact of highly uncertain PRA modeling assumptions and techniques, and the rankings adjusted to minimize the influence of these assumptions on the classification. Verify that sensitivity studies were done, and provide a description of the studies with a summary of results.

Virginia Power Response When the original PRA for Surry was conducted, numerous sensitivity studies were performed as described in the Surry IPE submittal. These sensitivity studies addressed the typical PRA modeling uncertainty issues. In addition, several specific sensitivity studies were defined for the risk-informed ISI evaluation.

As discussed in the Surry submittal in Sections 3.5 and 3.7, each piping segment within the scope of the program was evaluated to determine its core damage frequency and large early release frequency (LERF) without operator action. Additionally, each pipe segment was reviewed and where possible consideration was given to cases "with operator action" to evaluate the sensitivity of the results to crediting an operator action to isolate the piping failure to reduce the overall consequences from the piping failure.

Each of these four cases was used to identify the high safety significant piping segments and those in the range which were deemed worthy of additional consideration by the plant expert panel.

In Table 3.7-3 the safety significant piping segments for each case are shown. In addition, another sensitivity study was performed which took "no credit for augmented programs." These results are also shown in Table 3.7-3.

An uncertainty analysis was conducted in which the conditional core damage probabilities / frequencies (CCDP/CCDF) and failure probabilities / frequencies were assigned distributions around each of these " point estimates" such that the median of the log-normal distribution is equal to the point estimate. This assumption causes the calculated mean to be higher and thus causes the risk reduction worth (RRW) values to be higher for each segment. Higher RRW values allow for more piping segments to be considered in the high safety-significant category. A similar analysis was conducted for LERF and LERF rankings.

In addition, the risk associated with three cases is compared to identify additional safety significant piping segments. The three cases that are compared are: 1) the no ISI case,

2) the current Section XI program, and 3) the proposed Rl-ISI program. The comparison of these cases identifies additionai piping segments that are categorized as low safety significant that require additional evaluation in order to show a net safety benefit or risk neutral position when moving from the current Section XI program to the 9

proposed RI-ISI program. This comparison includes the benefits gained based on l assumed detection during an ISI examination in the SRRA software, j l

l Draft SRP 3.9.8 and Draft DG-1063 Regarding Sensitivity Studies For sensitivity studies, Subsection 11.2.3.1 of SRP 3.9.8 refers to draft DG-1063. The draft DG-1063 section 4.2.6.3 refers to Reference 21 (draft NUREG-1602) and section A2.5.5 of Appendix 2 of draft DG-1063 with regard to uncertainty and sensitivity studies.

Section A2.5.5 of Appendix 2 describes sensitivity studies that, based on the changes

- that occurred in WOG risk-informed ISI process during the Surry RI-ISI program, are not considered valid sensitivity studies for the reasons described below:

i l . ' The 1E-08 cutoff value for a pipe rupture' failure probability was not used in the l l Surry study because leaks _ and disabling leak probabilities were estimated which were typically higher than the 1E-08 cutoff assumed for a pipe rupture.

The actual failure probability from the' SRRA software was used in the calculations.

. The issue of estimated probabilities for certain failure mechanisms being ,

systematically high or low was addressed by crediting the impact of augmented L programs already in place on the failure probability. In addition, the uncertainty evaluation addressed the estimated failure probabilities.

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-. The use of leak probabilities versus break probabilities was not an issue in the Surry study because leaks and disabling leaks were estimated and used in the l calculations (the rupture failure probabilities were not used except for pipe whip concerns).~

r . The evaluation of operator actions in mitigating the effects of a pipe break was

addressed by considering cases with and without operator action as base cases for which high safety significant piping segments were identified.

. The evaluation of no credit for ISI in the initial rankings is evaluated through the change in risk process in which the current Section XI ISI program is compared to the proposed RI-ISI program.

I Uncertainty Analysis Details l An uncertainty analysis was conducted in which the conditional core damage I l

probabilities / frequencies (CCDP/CCDF) and failure probabilities / frequencies were i assigned distributions around each of these " point estimates" such that the median of the log-normal distribution is equal to the point estimate. The given point estimates were taken as the median of the log-normal distribution, i.e., there is a 50% chance that the actual failure rate will be higher and a 50% chance that it will be lower. The only 10 l

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other reasonable interpretation would be to take the point estimates as the mean or '

i expected value. Interpreting them as medians is a conservative assumption from the i perspective of risk since, in a rig'it skewed probability distribution such as the log-normal, the mean is greater than the median, i.e., we are assuming that, on average, the actual value will be greater than the point estimate.

The " spread" of the distribution about the median is determined by the standard j deviation. To estimate the spread or variability of the log-normal distribution, factors of 5,10 and 20 were used for this analysis. These factors were used to estimate the 95*-% tile of the distribution, i.e.. the 95*-% tile was set equal to the selected factor times the initial point estimate (medir1). The factor used was determined by the value of the point estimate. If the pr, int estimate was less than 1E-04, a factor of 20 was used. If the point estimate was greater than or equal to 1E-02, a factor of 5 was used.

Otherwise, a factor of 10 was used.

Given these three points, i.e., the lower bound (= 0), the median, and the 95*-% tile, it is straightforward to fit a log-normal probability distribution by first estimating the mean and standard deviation of the underlying normal distribution:

px =1n(mediants)

In(rangefactor,x)-p, inverse normal (95%)

and then computing the parameters of the associated log-normal distribution:

ut, = e(n, + M) oty = 'e(2 y + oj.)e(oj. -1)'

This probability model assigns a 50% chance that the value is between 0 and the original point estimate, a 45% chance that the value is between the original point estimate and 95*-% tile (= the original estimate multiplied by its assigned factorj, and a 5% chance that the value is greater than the 95*-% tile.

The @ RISK software, a Monte Carlo simulation " add-in" to Microsoft Excel, was used to analyze the uncertainty and generate the range of outcomes for each piping segment. A simulation of 5000 iterations was performed, and simulation statistics were collected for Pressure boundary CDF/LERF, Total segment Piping CDF/LERF, RAW, RRW and the Total CDF/LERF. The mean CDF/LERF value was then used for each piping segment. The results of this uncertainty analysis showed that there was no significant change in the RRW ranking. The uncertainty analysis results are summarized in Tables 5-1 and 5-2 (these tables were reported in WOG WCAP-14572, Revision 1 as Tables 3.6-7 and 3.6-8 respectively).

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, , . . 1 l The uncertainty-analysis identified a total of 86 segments whose RRW value was greater than 1.005 (see Table 5-2). As discussed in previous tables in the submittal, SW-18 through SW-25, were also identified during the original case analyses and were considered not to be safety significant. Approximately 75% of the piping segments identified using point estimates were also identified through the uncertainty analysis. l The remaining 25% were in the previous RRW range of 1.001 to 1.004 which were targeted for expert panel special consideration. Several of the piping segments identified through the uncertainty evaluation were added to the high safety-significant category by the plant expert panel. Additional piping segments were also added to the  !

evaluation as part of the defense-in-depth or change in risk calculations.

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TABLE 5-1 (WCAP-14572-R1, Table 3.6-7)

Surry Unit 1 Uncertainty Analysis MEAN PIPING RISK CONTRIBUTION BY SYSTEM 3 System # of CDF CDF LERF LERF {

Segments NO WITH NO WITH l OPERATOR OPERATOR OPERATOR OPERATOR ACTION ACTION ACTION ACTION ACC 15 3.07E-9 3.10E-9 7.40E-10 8.74E-10 AFW 32 2.82E-5 1.41 E-6 1.87E-6 1.74E-7 1 AS 2 3.24E-8 3.29E-8 3.31 E-8 3.41 E-8 I BD 12 1.20E-6 1.18E-6 6.98E-7 6.95E-7 l CC 66 1.38E-7 1.16E-7 1.03E-7 8.70E-8 CH 44 1.94E-6 1.97E-6 1.51 E-8 1.32E-8 CN 9 8.39E-6 5.23E-7 4.61 E-7 1.07E-7 CS 16 2.54E-6 4.98E-7 3.45E-7 4.61 E-7 CW 16 5.20E-7 5.23E-7 2.30E-8

)

2.41 E-8 1 ECC 8 7.50E-10 8.28E-10 1.09E-10 9.83E-11 l EE 7 4.56E-9 3.80E-9 3.26E-10 2.79E-10 i FC 9 N/A N/A N/A N/A )

FW 20 2.27E-6 2.22E-6 1.56E-7 1.56E-7 HHI 27 4.01 E-6 1.27E-6

)

6.05E-7 2.45E-7 i LHI 18 4.49E-7 2.00E-8 6.04E-8 2.58E-9 MS 38 2.13E-6 2.09E-6 8.94E-8 8.65E-8 RC 96 1.26E-5 1.29E-5 7.00E-8 7.26E-8 RH 11 4.60E-7 4.70E-7 5.42E-7 5.08E-7 RS 13 1.16E-6 1.25E-6 2.75E-10 0 SW 54 1.17E-4 6.98E-7 1.76E-5 7.03E-8 VS 2 2.82E-5 0 1.43E-6 0 l

TOTAL 515- 2.11 E-4 2.71 E-5 2.41 E-5 2.74E-6 l

l l

13 L _ _

o ,

l Table 5-2 (WCAP-14572-R1 Table 3.6-8) ,

SURRY UNIT 1 l UNCERTAINTY ANALYSIS RESULTS l HIGH SAFETY SIGNIFICANT PIPING SEGMENTS System CDF Without CDF With Operator LERF LERF With Operator Action - Without Operator Action - Uncertainty Operator Action -

Uncertainty Analysis Action - Uncertaldv Analysis Uncertainty Analysis Analysis ACC None None None None AFW AFW-15, 16, AFW-4, 5, 6, 28 AFW-15, 16. AFW-4, 5, 6 17,18,19 17,18,19 13,14,28 AS None None None AS-1, 2 BD None BD-2B, 3, 58, 6, 88, BD-2B, 3, 58, BD-2B, 3 9 6, 8B, 9 SB,6,88,9 I CC None None None CC-25, 30 l 33 i CH None CH-8, 9,10 None None l CN CN-8 CN-8 CN-8 CN-8 CS None CS-9 CS-5, 6 CS-7, 8, 9 CW None CW-5, 6, 7, 8 None None ,

ECC None None None None EE None None None None FC N/A N/A N/A N/A FW None FW-12,13,14 None FW-1, 2, 5 12,13,14 HHI HHi-5A, 6A HHI-10,12A,13,15, HHI-4A, 5A, HHI-10, 12A 17 6A 13,15,17 LHI None None None None MS None MS-33, 34 None MS-33, 34 RC RC-18, 41, RC-10, 11, 12, 13, None RC-41, 42 42,43 14, 15, 16, 17, 18, 43 19, 37, 38, 39, 41, 42,43 RH None RH-3B RH-3, 3B RH-2, 3, 3B RS RS-10 RS-9,10 None None SW SW-18, 19,~ SW-4, 5, 6, 9,10 SW-18, 19, SW-4, 6 20, 21, 22, 20, 21, 22, 23,24,25 23,24,25 VS VS-2 None VS-2 None Notes for Table 5-2:

For risk calculations, the segments identified as high safety significant had calculated values for RRW of greater than 1.005.

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w Enclosure 1 Calculation No. SM-1124 Segment Definitions for Surry 1 RI-ISI Program, Rev. O and Engineering Transmittal ET No. MAT-97-0014 Estimated Failure Probabilities for Risk-Based ISI Surry Unit 1, Rev. O