RC-13-0037, Permanent ILRT Interval Extension Risk Impact Assessment

Permanent ILRT Interval Extension Risk Impact Assessment

ML13095A110
Person / Time
Site:	Summer
Issue date:	03/01/2013
From:	Westinghouse
To:	Office of Nuclear Reactor Regulation
References
CR-13-00705, RC-13-0037
Download: ML13095A110 (56)

Text

Document Control Desk Attachment VI CR-1 3-00705 RC-1 3-0037 Page 1 of 55 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

ATTACHMENT VI SCE&G ILRT INTERVAL EXTENSION RISK ANALYSIS

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC V. C. Summer Nuclear Station Unit 1 Permanent ILRT Interval Extension Risk Impact Assessment 3/1/2013

© 2013 Westinghouse Electric Company LLC All Rights Reserved

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table of Contents Section Page 1

Purpose of Analysis................

.............................. 4 1.1 Purpose......................................................................................................

4 1.2 Background...................................................................................................

4 1.3 Criteria................................................................................................................

5 2

M ethodology......................................................................................................

6 3

Ground Rules....................................................................................................

7 4

In p u ts.........................................................................................................................

8 4.1 General Resources Available.........................................................................

8 4.2 Plant Specific Inputs....................................................................................

12 4.3 Impact of Extension on Detection of Component Failures That Lead to Leakage (Sm all and Large)........................................................................................

22 4.4 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage 2 3 5

Results.....................................................................................................................

28 5.1 Step 1 - Quantify the Base-Line Risk in Terms of Frequency Per Reactor Year30 5.2 Step 2 - Develop Plant Specific Person-Rem Dose (Population Dose) Per Reactor Y e a r..................................................................................................................

3 3 5.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval From 10 to 15 Years................................................................................................................

38 5.4 Step 4 - Determine the Change in Risk in Terms of Large Early Release Frequency (LERF)......................................................................................

43 5.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability (CCFP).......................................................................................................

43 5.6 Sum m ary of Results....................................................................................

44 6

Sensitivities.....................................................................

........................................ 46 6.1 Sensitivity to Corrosion Im pact Assum ptions................................................

46 6.2 Sensitivity to Class 3B Contribution to LERF...............................................

47 6.3 Potential Im pact From External Events Contribution....................................

47 7

Conclusions.............................................................................................................

51 8

References...............................................................................................................

52 Page 2

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC List of Tables Table Page Table 4-1: VCSNS Unit 1 Level 2 LERF Release Categories and Frequencies............ 14 Table 4-2: VCSNS Unit 1 Level 2 Release Categories and Frequencies...................... 15 Table 4-3: Summary Accident Progression Bin (APB)....................................................

15 Table 4-4: Calculation of Surry Population Dose Risk at 50 Miles..................................

17 Table 4-5: Calculation of VCSNS Unit 1 Population Dose Risk at 50 Miles................... 19 Table 4-6: VCSNS Unit 1 Level 2 Model Assumptions for Application to the NUREG/CR-4551 Accident Progression Bins and EPRI Accident Classes........................................ 20 Table 4-7: EPRI Containment Failure Classification (Reference 2)...............................

21 Table 4-8: Steel Liner Corrosion Base Case.................................................................

26 Table 5-1: A ccident C lasses..........................................................................................

29 Table 5-2: VCSNS Unit 1 Categorized Accident Classes and Frequencies.................. 31 Table 5-3: Radionuclide Release Frequencies as a Function of Accident Class (VCSNS U n it 1 B a se C a se )...............................................................................................................

3 3 Table 5-4: VCSNS Unit 1 Population Dose Estimates for Population...........................

35 Table 5-5: VCSNS Unit 1 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 3/10 Years.......................................................................

36 Table 5-6 VCSNS Unit 1 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/10 Years.......................................................................

39 Table 5-7: VCSNS Unit 1 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/15 Years.......................................................................

41 Table 5-8: VCSNS Unit 1 ILRT Cases: Base, 3 to 10, and 3 to 15 Yr Extensions (Including Age Adjusted Steel Liner Corrosion Likelihood)............................................................

45 Table 6-1: Steel Liner Corrosion Sensitivity Cases........................................................

47 Table 6-2: VCSNS External Events Summary..............................................................

50 Table 6-3: VCSNS Estimated Total LERF Including External Events Impact................. 51 Page 3

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC I

Purpose of Analysis 1.1 Purpose The purpose of this analysis is to provide a risk assessment of extending the currently allowed containment Type A Integrated Leak Rate Test (ILRT) to a permanent fifteen years. The extension would allow for substantial cost savings as the ILRT could be deferred for additional scheduled refueling outages for the V. C. Summer Nuclear Station Unit 1

(VCSNS Unit 1).

The risk assessment follows the guidelines from NEI 94-01 (Reference 1), the methodology used in EPRI TR-1 04285 (Reference 2), the NEI "Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals" from November 2001 (Reference 3), the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in Regulatory Guide 1.200 as applied to ILRT interval extensions, and risk insights in support of a request for a plant's licensing basis as outlined in Regulatory Guide (RG) 1.174 (Reference 4), the methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion induced leakage of steel liners going undetected during the extended test interval (Reference 5), and the methodology used in EPRI 1009325, Revision 2-A (Reference 21).

1.2 Background

Revisions to 10CFR50, Appendix J (Option B) allow individual plants to extend the Integrated Leak Rate Test (ILRT) Type A surveillance testing frequency requirement from three in ten years to at least once in ten years. The revised Type A frequency is based on an acceptable performance history defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage rate was less than the limiting containment leakage rate of 1La1.

The basis for the current fifteen year test interval is provided in Section 11.0 of NEI 94-01, Revision 3-A, and was established in 2008.

Section 11.0 of NEI 94-01 states that NUREG-1493, "Performance-Based Containment Leak Test Program," September 1995 (Reference 6), provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the NRC's rulemaking basis, NEI undertook a similar study. The results of that study are documented in Electric Power Research Institute (EPRI) Research Project Report TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals."

La (percent/24 hours) is the maximum allowable leakage rate at pressure Pa (calculated peak containment internal pressure related to the design basis accident) as specified in the technical specifications.

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC The NRC report on performance-based leak testing, NUREG-1493, analyzed the effects of containment leakage on the health and safety of the public and the benefits realized from the containment leak rate testing. In that analysis, it was determined that for a representative PWR plant (i.e., Surry) that containment isolation failures contribute less than 0.1 percent to the latent risks from reactor accidents. Consequently, it is desirable to show that extending the ILRT interval will not lead to a substantial increase in risk from containment isolation failures for VCSNS Unit 1.

The Guidance provided in Appendix H of EPRI Report No. 1009325, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," (Reference 21) for performing risk impact assessments in support of ILRT extensions builds on the EPRI Risk Assessment methodology, EPRI TR-104285. This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed ILRT changes.

It should be noted that containment leak-tight integrity is also verified through periodic in-service inspections conducted in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI.

More specifically, subsection IWE provides the rules and requirements for in-service inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants. Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct visual inspections of the accessible areas of the interior of the containment. The associated change to NEI 94-01 will require that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being conducted. These requirements will not be changed as a result of the extended ILRT interval. In addition, Appendix J, Type B local leak tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are also not affected by the change to the Type A test frequency.

1.3 Criteria The acceptance guidelines in RG 1.174 are used to assess the acceptability of this permanent extension of the Type A test interval beyond that established during the Option B rulemaking of Appendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in Core Damage Frequency (CDF) less than 10-6 per reactor year and increases in Large Early Release Frequency (LERF) less than 10-7 per reactor year. As VCSNS Unit 1 does not credit containment overpressure for the mitigation of design basis accidents, the Type A test does not impact CDF. Therefore, the relevant risk metric is the change in LERF. RG 1.174 also defines small changes in LERF as below 10-6 per reactor year. RG 1.174 discusses defense-in-depth and encourages the use of risk analysis techniques to help ensure and show that key principles, such as the defense-in-depth philosophy, are met. Therefore, the increase in the Conditional Containment Failure Probability (CCFP) that helps to ensure that the defense-in-depth philosophy is maintained is also calculated. The criteria described below are taken from the NRC Final Safety Evaluation for NEI 94-01 and EPRI Report No. 1009325 (Reference 29).

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Regarding CCFP, the NRC concluded that a small increase in CCFP should be defined as a value marginally greater than that accepted in previous one time fifteen year ILRT extension requests. To this end the NRC has endorsed a small increase in CCFP as an increase in CCFP be less than or equal to 1.5% (Reference 29).

In addition, the total annual risk (person rem/yr population dose) is examined to demonstrate the relative change in this parameter. The NRC concluded that for purposes of assessing the risk impacts of the Type A ILRT extension in accordance with the EPRI methodology, a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1 percent of the total population dose, whichever is less restrictive (Reference 29).

2 Methodology A simplified bounding analysis approach consistent with the EPRI approach is used for evaluating the change in risk associated with increasing the test interval to fifteen years.

The approach is consistent with that presented in Appendix H of EPRI Report No. 1009325, Revision 2-A, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals" (Reference 21), EPRI TR-104285 (Reference 2), NUREG-1493 (Reference 6) and the Calvert Cliffs liner corrosion analysis (Reference 5). The analysis uses results from the current VCSNS Unit 1 Level 2 PRA model to establish frequency of fission product releases.

Fission product release magnitudes are extrapolated from results of NUREG/CR-4551 to account for VCSNS Unit 1 specific characteristics.

This risk assessment is applicable to VCSNS Unit 1.

The six general steps of this assessment are as follows:

1. Quantify the baseline risk in terms of the frequency of events (per reactor year) for each of the eight containment release scenario types identified in the EPRI report No. 1009325, Revision 2-A (Reference 21).

2.

Develop plant specific person-rem (population dose) per reactor year for each of the eight containment release scenario types from plant specific consequence analyses.

3.

Evaluate the risk impact (i.e., the change in containment release scenario type frequency and population dose) of extending the ILRT interval to fifteen years.

4. Determine the change in risk in terms of Large Early Release Frequency (LERF) in accordance with RG 1.174 (Reference 4) and compare with the acceptance guidelines of RG 1.174.

5. Determine the impact of the ILRT interval extension on the Conditional Containment Failure Probability (CCFP) and the population dose and compare with the acceptance guidance of Reference 29.

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC

6. Evaluate the sensitivity of the results to assumptions in the liner corrosion analysis, external events and to the fractional contribution of increased large isolation failures (due to liner breach) to LERF.

This approach is based on the information and approaches contained in the previously mentioned studies. Furthermore:

Consistent with the other industry containment leak risk assessments, the VCSNS Unit 1 assessment uses LERF and delta LERF in accordance with the risk acceptance guidance of RG 1.174. Changes in population dose and conditional containment failure probability are also considered to show that defense-in-depth and the balance of prevention and mitigation is preserved.

This evaluation for VCSNS Unit 1 uses ground rules and methods to calculate changes in risk metrics that are similar to those used in Appendix H of EPRI Report No. 1009325, Revision 2-A, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals."

3 Ground Rules The following ground rules are used in the analysis:

The technical adequacy of the VCSNS Unit 1 PRA is consistent with the requirements of Regulatory Guide 1.200 as is relevant to this ILRT interval extension.

The current VCSNS Unit 1 Level 1 and Level 2 internal events PRA models are explicitly used in this analysis to assess fission product release frequencies.

It is appropriate to use the VCSNS Unit 1 internal events PRA model as a gauge to effectively describe the risk change attributable to the ILRT extension. It is reasonable to assume that the impact from the ILRT extension (with respect to percent increases in population dose) will not substantially differ if fire and seismic events were to be included in the calculations; this is evaluated in the sensitivity analysis which uses available information from the VCSNS Unit 1 IPEEE (Reference 27).

Dose results for the containment failures modeled in the PRA can be characterized by scaling information provided in NUREG/CR-4551 (Reference 7). Specifically, VCSNS population dose estimates are obtained by scaling the NUREG/CR-4551 reference plant results by differences in population, reactor power level (assumed proportional to fission product inventory), and nominal containment maximum leakage rate (La).

Accident classes describing radionuclide release end states are defined consistent with EPRI methodology (Reference 2) and are summarized in Section 4.2.

The representative containment leakage for Class 1 sequences is 1 La Class 3 accounts for increased leakage due to Type A inspection failures.

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC The representative containment leakage for Class 3a sequences is 10La based on the previously approved methodology performed for Indian Point Unit 3 (References 8 and 9).

The representative containment leakage for Class 3b sequences is 100La based on the guidance provided in EPRI Report No. 1009325, Revision 2-A.

" The Class 3b is very conservatively categorized as LERF based on the previously approved methodology (References 8 and 9).

" The impact on population doses from containment bypass scenarios is not altered by the proposed ILRT extension, but is accounted for in the EPRI methodology as a separate entry for comparison purposes. Since the containment bypass contribution to population dose is fixed, no changes on the conclusions from this analysis will result from this separate categorization.

" The reduction in ILRT frequency does not impact the reliability of containment isolation valves to close in response to a containment isolation signal.

4 Inputs This section summarizes the general resources available as input (Section 4.1) and the plant specific resources required (Section 4.2).

4.1 General Resources Available Various industry studies on containment leakage risk assessment are briefly summarized here:

1. NUREG/CR-3539 (Reference 10)

2. NUREG/CR-4220 (Reference 11)

3. NUREG-1273 (Reference 12)

4. NUREG/CR-4330 (Reference 13)

5. EPRI TR-105189 (Reference 14)

6. NUREG-1493 (Reference 6)

7.

EPRI TR-1 04285 (Reference 2)

8. NUREG-1150 (Reference 15) and NUREG/CR-4551 (Reference 7)

9.

NEI Interim Guidance (Reference 3, Reference 18)

10. Calvert Cliffs Liner Corrosion Analysis (Reference 5)

11. EPRI Report No. 1009325, Revision 2-A, Appendix H (Reference 21)

The first study is applicable because it provides one basis for the threshold that could be used in the Level 2 PRA for the size of containment leakage that is considered significant and is to be included in the model. The second study is applicable because it provides a basis of the probability for significant pre-existing containment leakage at the time of a core damage accident. The third study is applicable because it is a subsequent study to Page 8

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC NUREG/CR-4220 that undertook a more extensive evaluation of the same database. The fourth study provides an assessment of the impact of different containment leakage rates on plant risk. The fifth study provides an assessment of the impact on shutdown risk from ILRT test interval extension. The sixth study is the NRC's cost-benefit analysis of various alternative approaches regarding extending the test intervals and increasing the allowable leakage rates for containment integrated and local leak rate tests. The seventh study is an EPRI study of the impact of extending ILRT and LLRT test intervals on at-power public risk.

The eighth study provides an ex-plant consequence analysis for a 50-mile radius surrounding a plant that is used as the bases for the consequence analysis of the ILRT interval extension for VCSNS Unit 1. The ninth study includes the NEI recommended methodology (promulgated in two letters) for evaluating the risk associated with obtaining a one-time extension of the ILRT interval. The tenth study addresses the impact of age-related degradation of the containment liners on ILRT evaluations. Finally, the eleventh study builds on the previous work and includes a recommended methodology and template for evaluating the risk associated with a permanent fifteen year extension of the ILRT interval.

4.1.1 NUREG/CR-3539 (Reference 10)

Oak Ridge National Laboratory (ORNL) documented a study of the impact of containment leak rates on public risk in NUREG/CR-3539. This study uses information from WASH-1400 (Reference 16) as the basis for its risk sensitivity calculations. ORNL concluded that the impact of leakage rates on LWR accident risks is relatively small.

4.1.2 NUREG/CR-4220 (Reference 11)

NUREG/CR-4220 is a study performed by Pacific Northwest Laboratories for the NRC in 1985. The study reviewed over two thousand LERs, ILRT reports and other related records to calculate the unavailability of containment due to leakage.

4.1.3 NUREG-1273 (Reference 12)

A subsequent NRC study, NUREG-1273, performed a more extensive evaluation of the NUREG!CR-4220 database. This assessment noted that about one-third of the reported events were leakages that were immediately detected and corrected. In addition, this study noted that local leak rate tests can detect "essentially all potential degradations" of the containment isolation system.

4.1.4 NUREG/CR-4330 (Reference 13)

NUREG/CR-4330 is a study that examined the risk impacts associated with increasing the allowable containment leakage rates. The details of this report have no direct impact on the modeling approach of the ILRT test interval extension, as NUREG/CR-4330 focuses on leakage rate and the ILRT test interval extension study focuses on the frequency of testing intervals. However, the general conclusions of NUREG/CR-4330 are consistent with NUREG/CR-3539 and other similar containment leakage risk studies:

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC

"...the effect of containment leakage on overall accident risk is small since risk is dominated by accident sequences that result in failure or bypass of containment."

4.1.5 EPRI TR-105189 (Reference 14)

The EPRI study TR-105189 is useful to the ILRT test interval extension risk assessment because it provides insight regarding the impact of containment testing on shutdown risk.

This study contains a quantitative evaluation (using the EPRI ORAM software) for two reference plants (a BWR-4 and a PWR) of the impact of extending ILRT and LLRT test intervals on shutdown risk. The conclusion from the study is that a small but measurable safety benefit is realized from extending the test intervals.

4.1.6 NUREG-1493 (Reference 6)

NUREG-1493 is the NRC's cost-benefit analysis for proposed alternatives to reduce containment leakage testing intervals and/or relax allowable leakage rates. The NRC conclusions are consistent with other similar containment leakage risk studies:

Reduction in ILRT frequency from three per ten years to one per twenty years results in an "imperceptible" increase in risk.

Given the insensitivity of risk to the containment leak rate and the small fraction of leak paths detected solely by Type A testing, increasing the interval between integrated leak rate tests is possible with minimal impact on public risk.

4.1.7 EPRI TR-104285 (Reference 2)

Extending the risk assessment impact beyond shutdown (the earlier EPRI TR-105189 study), the EPRI TR-104285 study is a quantitative evaluation of the impact of extending ILRT and LLRT test intervals on at-power public risk. This study combined IPE Level 2 models with NUREG-1150 Level 3 population dose models to perform the analysis. The study also used the approach of NUREG-1493 in calculating the increase in pre-existing leakage probability due to extending the ILRT and LLRT test intervals.

EPRI TR-104285 uses a simplified Containment Event Tree to subdivide representative core damage frequencies into eight classes of containment response to a core damage accident:

1. Containment intact and isolated

2. Containment isolation failures dependent upon the core damage accident

3. Type A (ILRT) related containment isolation failures

4. Type B (LLRT) related containment isolation failures

5. Type C (LLRT) related containment isolation failures

6. Other penetration related containment isolation failures

7. Containment failures due to core damage accident phenomena

8. Containment bypass Page 10

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Consistent with the other containment leakage risk assessment studies, this study concluded:

"... the proposed CLRT (containment leak rate tests) frequency changes would have a minimal safety impact. The change in risk determined by the analyses is small in both absolute and relative terms. For example, for the PWR analyzed, the change is about 0.04 person-rem per year... "

4.1.8 NUREG-1150 (Reference 15) and NUREG/CR 4551 (Reference 7)

NUREG-1 150 and the technical basis, NUREG/CR-4551, provide an ex-plant consequence analysis for a spectrum of accidents including a severe accident with the containment remaining intact (i.e., Tech Spec leakage). This ex-plant consequence analysis is calculated for the 50-mile radial area surrounding Surry. The ex-plant calculation can be delineated to total person-rem for each identified Accident Progression Bin (APB) from NUREG/CR-4551. With the VCSNS Unit 1 Level 2 model end-states assigned to one of the NUREG/CR-4551 APBs, it is considered adequate to represent VCSNS Unit 1. (The meteorology and site differences other than population are assumed not to play a significant role in this evaluation.)

4.1.9 NEI Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals (Reference 3, Reference 18)

The guidance provided in this document builds on the EPRI risk impact assessment methodology (Reference 2) and the NRC performance-based containment leakage test program (Reference 6), and considers approaches utilized in various submittals, including Indian Point 3 (and associated NRC SER) and Crystal River.

4.1.10 Calvert Cliffs Response to Request for Additional Information Concerning the License Amendment for a One-Time Integrated Leakage Rate Test Extension (Reference 5)

This submittal to the NRC describes a method for determining the change in likelihood, due to extending the ILRT, of detecting liner corrosion, and the corresponding change in risk.

The methodology was developed for Calvert Cliffs in response to a request for additional information regarding how the potential leakage due to age-related degradation mechanisms were factored into the risk assessment for the ILRT one-time extension. The Calvert Cliffs analysis was performed for a concrete cylinder, dome and a concrete basemat, each with a steel liner. Licensees may consider approved LARs for one-time extensions involving containment types similar to their facility. The VCSNS Unit 1 assessment has addressed the plant specific differences from the Calvert Cliffs design, and how the Calvert Cliffs methodology was adapted to address the specific design features. In the case where no similar analysis has been performed the licensee will use judgment based on the available analyses and plant specific features to perform the analysis.

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC 4.1.11 EPRI Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (Reference 21)

This report provides a generally applicable assessment of the risk involved in extension of ILRT test intervals to permanent 15-year intervals. Appendix H of this document provides guidance for performing plant specific supplemental risk impact assessments and builds on the previous EPRI risk impact assessment methodology (Reference 2) and the NRC performance-based containment leakage test program (Reference 6), and considers approaches utilized in various submittals, including Indian Point 3 (and associated NRC SER) and Crystal River.

The approach included in this guidance document is used in the VCSNS Unit 1 assessment to determine the estimated increase in risk associated with the ILRT extension. This document includes the bases for the values assigned in determining the probability of leakage for the EPRI Class 3a and 3b scenarios in this analysis as described in Section 5.

4.2 Plant Specific Inputs The plant specific information used to perform the VCSNS Unit 1 ILRT Extension Risk Assessment includes the following:

Level 1 Model results Level 2 Model results Release category definitions used in the Level 2 Model Population within a 50-mile radius for the year 2040 which is based on an extrapolation of the 2000 census (Reference 17)

VCSNS Unit 1 ILRT results 2 which demonstrate the adequacy and integrity of the administrative procedures and containment penetrations (Reference 23 and Reference 24)

Containment failure probability data 4.2.1 Level I Model The Level 1 PRA model that is used for VCSNS Unit 1 is characteristic of the as-built plant.

The current Level 1 model is a linked fault tree model, and was quantified with the total Core Damage Frequency (CDF) = 4.34E-06/yr.

2 The two most recent Type A tests at VCSNS Unit 1 have been successful, so the current Type A test interval requirement is one time in ten years. Operating Experience Evaluation CR-1 0-02513 states that VCSNS has not experienced any events which contributed significant degradation to the containment liner.

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC 4.2.2 Level 2 Model The Level 2 Model that is used for VCSNS Unit 1 was developed to calculate the LERF contribution as well as the other release categories evaluated in the model. Tables 4-1 and 4-2 summarize the pertinent VCSNS Unit 1 results in terms of release category. Note that the enumerated total internal events Level 2 release frequency is approximately 3%

larger than that of the internal events CDF. This difference arises as a result of the numerical truncation issues resulting from the full integration of core damage end-states into the Level 2 model and the impact of the CAFTA small number approximation as applied to the detailed containment isolation failure model.

The small number approximation is a standard modeling practice. While this difference is observable, it does not significantly impact the results of the simplified Level 2 PRA or the associated conclusions drawn with regard to the ILRT extension.

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 4-1: VCSNS Unit I Level 2 LERF Release Categories and Frequencies Release Case Definition Frequency/yr Category LERF-01 Non-SBO with SG FW available and a Low 2.37E-08 Pressure CFE (CFE1)

LERF-02 Non-SBO with a High Pressure CFE (CFE 3)

O.OOE+00 Non-SBO with no SG FW available, no RCS LERF-03 depressurization early, and a Low Pressure CFE O.OOE+00 (CFE5)

LERF-04 Non-SBO with a TI-SGTR 5.15E-09 Non-SBO with no SG FW available, RCS LERF-05 depressurization early, and a Low Pressure CFE O.OOE+00 (CFE 5)

LERF-06 Non-SBO with a PI-SGTR 2.96E-08 LERF-07 Non-SBO with a Low Pressure CFE (CFE 1) 3.70E-09 LERF-08 Non-SBO with Containment Isolation Failure 4.97E-09 LERF-09 Non-SBO with a Large Bypass Event 1.75E-08 LERF-10 SBO with SG FW available and a Low Pressure 424E09 CFE (CFE 1)

LERF-1 1 SBO with a High Pressure CFE (CFE 3)

O.OOE+00 SBO with no SG FW available, no RCS LERF-12 depressurization early, and a Low Pressure CFE O.OOE+00 (CFE 5)

LERF-13 SBO with a TI-SGTR 6.14E-09 SBO with no SG FW available, RCS LERF-14 depressurization early, and a Low Pressure CFE O.OOE+00 (CFE 5)

LERF-15 SBO with a PI-SGTR 9.23E-09 LERF-16 SBO with a Low Pressure CFE (CFE 1) 1.08E-1 3 LERF-17 SBO with Containment Isolation Failure 6.94E-10 LERF-18 SBO with a Large Bypass Event 1.63E-1 1 Total LERF Release Category Frequency 1.05E-07 (LERF-01 through LERF-18)

Table 4-2 summarizes all of the Level 2 release categories and frequencies. The CDF including uncategorized releases is determined by adding together all Level 2 release categories.

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 4-2: VCSNS Unit I Level 2 Release Cateaories and Freauencies 4-2 1

Le el2 R e ea e......sa d

re u

nc e

Release Category Definition Frequency/yr INTACT Containment Intact 3.41 E-06 SERF Small Early Release 1.99E-08 LATE Late Release 9.62E-07 LERF Total Large Early Releases 1.05E-07 CDF (including uncategorized releases) 4.50E-06 4.2.3 Population Dose Calculations The population dose is calculated by using data provided in NUREG/CR-4551 and adjusting the results for VCSNS Unit 1. Each of the release categories from Table 4-1 was associated with an applicable Collapsed Accident Progression Bin (APB) from NUREG/CR-4551 (see below). The collapsed APBs are characterized by 5 attributes related to the accident progression. Unique combinations of the 5 attributes result in a set of 7 bins that are relevant to the analysis. The definitions of the 7 collapsed APBs are provided in NUREG/CR-4551 and are reproduced in Table 4-3 for reference purposes.

Table 4-4 summarizes the calculated population dose for Surry associated with each APB from NUREG/CR-4551.

Table 4-3: Summary Accident Progression Bin (APB)

Descriptions (Reference 7)

Summary APB Description Number 1

CD, VB, Early CF, Alpha Mode Core damage occurs followed by a very energetic molten fuel-coolant interaction in the vessel; the vessel fails and generates a missile that fails the containment as well. Includes accidents that have an Alpha mode failure of the vessel and the containment except those follow Event V or an SGTR. It includes Alpha mode failures that follow isolation failures because the Alpha mode containment failure is of rupture size.

2 CD, VB, Early CF, RCS Pressure > 200 psia Core Damage occurs followed by vessel breach. Implies Early CF with the RCS above 200 psia when the vessel fails. Early CF means at or before VB, so it includes isolation failures and seismic containment failures at the start of the accident as well as containment failure at VB. It does not include bins in which containment failure at VB follows Event V or an SGTR, or Alpha mode failures.

3 CD, VB, Early CF, RCS Pressure < 200 psia Core damage occurs followed by vessel breach. Implies Early CF with the RCS below psia when the containment fails. It does not include bins in which the containment failure at VB or an SGTR, or Alpha mode failures.

Page 15

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 4-3: Summary Accident Progression Bin (APB)

Descriptions (Reference 7)

Summary APB Description Number 4

CD, VB, Late CF Core Damage occurs followed by vessel breach. Includes accidents in which the containment was not failed or bypassed before the onset of core-concrete interaction (CCI) and in which the vessel failed. The failure mechanisms are hydrogen combustion during CCI, Basemat Melt-Through (BMT) in several days, or eventual overpressure due to the failure to provide containment heat removal in the days following the accident.

5 CD, Bypass Core Damage occurs followed by vessel breach. Includes Event V and SGTRs no matter what happens to the containment after the start of the accident. It also includes SGTRs that do not result in VB.

6 CD, VB, No CF Core Damage occurs followed by vessel breach. Includes accidents not evaluated in one of the previous bins. The vessel's lower head is penetrated by the core, but the containment does not fail and is not bypassed.

7 CD, No VB, No CF Core Damage occurs but is arrested in time to prevent vessel breach.

Includes accident progressions that avoid vessel failures except those that bypass the containment. Most of the bins placed in this reduce bin have no containment failure as well as no VB. It also includes bins in which the containment is not isolated at the start of the accident and the core is brought to a safe stable state before the vessel fails.

Page 16

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 4-4: Calculation of Surry Population Dose Risk at 50 Miles (Reference 7)

NUREG/CR-4551 NUREG/CR-4551 Fractional APB Pulation NUREG/CR-4551 Pulation Dose Colase Cntibtins Population Dose Collapsed Bin Population Dose Collapsed Contributions Risk at 50 miles ClasdBnat Bin #

to Risk (MFCR)

Rson-res Frequencies 50 miles (1)

(person-remlyr, (per year) (3) 5 ie mean) (2)

(person-rem) (4) 1 0.029 0.158 1.23E-07 1.28E+06 2

0.019 0.106 1.64E-07 6.46E+05 3

0.002 0.013 2.012E-08 6.46E+05 (5) 4 0.216 1.199 2.42E-06 4.95E+05 5

0.732 4.060 5.OOE-06 8.12E+05 6

0.001 0.006 1.42E-05 4.23E+02 7

0.002 0.011 1.91 E-05 5.76E+02 Totals 1.000 5.55 4.1 E-05 (1) Mean Fractional Contribution to Risk calculated from the average of two samples delineated in Table 5.1-3 of NUREG/CR-4551.

(2) The total population dose risk at 50 miles from internal events in person-rem is provided as the average of two samples in Table 5.1-1 of NUREG/CR-4551. The contribution for a given APB is the product of the total PDR50 and the fractional APB contribution.

(3) NUREG/CR-4551 provides the conditional probabilities of the collapsed APBs in Figure 2.5-3. These conditional probabilities are multiplied by the total internal CDF to calculate the collapsed APB frequency.

(4) Obtained from dividing the population dose risk shown in the third column of this table by the collapsed bin frequency shown in the fourth column of this table.

(5) Assumed population dose at 50 miles for Collapsed Bin #3 equal to that of Collapsed Bin

1. 2. Collapsed Bin Frequency #3 was then back calculated using that value. This does not influence the results of this evaluation since Bin #3 does not appear as part of the results for VCSNS Unit 1.

4.2.4 Population Dose Estimate Methodology The person-rem results in Table 4-4 can be used as an approximation of the dose for the VCSNS Unit 1 if it is corrected for allowable containment leak rate (La), reactor power level and the population density surrounding VCSNS Unit 1.

La adjustment:

FLeakage = La of VCSNS Unit 1 (%w/o/day) / La of reference plant (applicable only to those APBs affected by normal leakage)

La for VCSNS Unit 1 is 0.2%w/o/day (Reference 22, Page 6-12b). La for Surry is 0.1 %w/o/day.

FLeakage = 0.2 / 0.1 FLeakage = 2 Page 17

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Power level adjustment:

FPower = Rated power level of VCSNS Unit 1 (MWt) / Rated power level of reference plant The rated power level for VCSNS Unit 1 is 2900 MWt (Reference 22, Page 1-5). The rated power level for Surry is 2441 MWt.

FPower = 2900 MWt / 2441 MWt FPower = 1.188 Population density adjustment:

The total population within a 50-mile radius of VCSNS is 1,648,935 (Reference 17, Page 2.1-5).

This number is based on the 2000 census and includes the impact of population growth since it is the projected estimate for the year 2040. This population value is compared to the population value that is provided in NUREG/CR-4551 in order to get a "Population Dose Factor" that can be applied to the APBs to get dose estimates for VCSNS Unit 1.

Total VCSNS Population 50 miles = 1.65E+06 Surry Population within a 50 mile radius from the NUREG/CR-4551 reference plant =

1.23E+06 FPopulation = 1.65E+06 / 1.23E+06 = 1.341 The factors developed above are used to adjust the population dose for the surrogate plant (Surry) for VCSNS Unit 1. For intact containment endstates, the total population dose factor is as follows:

Flntact = FPopulation

• FPower

• FLeakage Flntact = 1.341

• 1.188* 2 Flntact = 3.185 For EPRI accident classes not dependent on containment leakage, the population dose factor is as follows:

FOthers = FPopulation

• FPower FOthers = 1.341

• 1.188 FOthers = 1.593 The difference in the doses at 50 miles is assumed to be in direct proportion to the difference in the population within 50 miles of each site. The above adjustments provide an Page 18

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC approximation for VCSNS Unit 1 of the population doses associated with each of the release categories from NUREG/CR-4551.

Table 4-5 shows the results of applying the population dose factor to the NUREG/CR-4551 population dose results at 50 miles to obtain the adjusted population dose at 50 miles for VCSNS Unit 1.

Table 4-5: Calculation of VCSNS Unit I Population Dose Risk at 50 Miles Accident NUREGICR-4551 Bin Multiplier used VCSNS Adjusted Progression Population Dose at to obtain VCSNS Population Dose at 50 miles Popobtin Dose 50 miles Bin (APB)

(person-rem)

Population Dose (person-rem) 1 1.28E+06 1.593 2.04E+06 2

6.46E+05 1.593 1.03E+06 3

6.46E+05 1.593 1.03E+06 4

4.95E+05 1.593 7.88E+05 5

8.12E+05 1.593 1.29E+06 6

4.23E+02 3.185 1.35E+03 7

5.76E+02 3.185 1.83E+03 4.2.5 Application of VCSNS Unit 1 PRA Model Results to NUREG/CR-4551 Level 3 Output A major factor related to the use of NUREG/CR-4551 in this evaluation is that the results of the VCSNS Unit 1 PRA Level 2 model are not defined in the same terms as reported in NUREG/CR-4551. The VCSNS Unit 1 PRA Level 2 model results are defined as four main release categories including INTACT, SERF, LATE, and LERF. In order to use the Level 3 model presented in that document, it was necessary to match the VCSNS PRA Level 2 release categories to the collapsed APBs. The VCSNS Level 2 release categories and frequencies are from the current simplified Level 2 model. The assignments are shown in Table 4-6, along with the corresponding EPRI classes (see below). The EPRI classes and descriptions are listed in Table 4-7 in addition to the VCSNS Level 2 release categories.

Page 19

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 4-6: VCSNS Unit I Level 2 Model Assumptions for Application to the NUREG/CR-4551 Accident Progression Bins and EPRI Accident Classes VCSNS Level 2 Release Frequency Definition NUREG/CR-EPRI Category (per yr) 4551 APB Class Frequency INTACT 3.41 E-06 Containment Intact 6

1 SERF 1.99E-08 Small Early Release 3

3 LATE 9.62E-07 Late Release 4

7 LERF-01 2.37E-08 Non-SBO with a Low 3

6 Pressure CFE LERF-02 O.OOE+00 Non-SBO with a High 2

7 Pressure CFE LERF-03 O.OOE+OO Non-SBO with a Low 3

6 Pressure CFE LERF-04 5.15E-09 Non-SBO with a TI-SGTR 5

8 LERF-05 O.OOE+OO Non-SBO with a Low 3

6 Pressure CFE LERF-06 2.96E-08 Non-SBO with a PI-SGTR 5

8 LERF-07 3.70E-09 Non-SBO with a Low 3

6 Pressure CFE Non-SBO with LERF-08 4.97E-09 Containment Isolation 1

2 Failure LERF-09 1.75E-08 Non-SBO with a Large 5

8 Bypass Event LERF-1O0 4.24E-09 SBO with a Low Pressure 3

6 CFE LERF-1 1 O.OOE+00 SBO with a High Pressure 2

7 CFE LERF-12 O.OOE+OO SBO with a Low Pressure 3

6 CFE LERF-15 9.23E-09 SBO with a PI-SGTR 5

8 LERF-16 1.08E-13 SBO with a Low Pressure 3

6 CFE LERF-17 6.94E-10 SBO with Containment 1

2 Isolation Failure LERF-18 1.63E-11 SBO with a Large Bypass 5

8 Event Page 20

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC 4.2.6 Release Category Definitions Table 4-7 defines the accident classes used in the ILRT extension evaluation, which is consistent with the EPRI methodology (Reference 2). These containment failure classifications are used in this analysis to determine the risk impact of extending the Containment Type A test interval as described in Section 5 of this report.

Table 4-7: EPRI Containment Failure Classification (Reference 2)

VCSNS Level 2 Release Class Description Category Frequency 1

Containment remains intact including accident INTACT sequences that do not lead to containment failure in the long term. The release of fission products (and attendant consequences) is determined by the maximum allowable leakage rate values La, under Appendix J for that plant 2

Containment isolation failures (as reported in LERF-08, LERF-17 the IPEs) include those accidents in which there is a failure to isolate the containment.

3 Independent (or random) isolation failures SERF include those accidents in which the pre-existing isolation failure to seal (i.e., provide a leak-tight containment) is not dependent on the sequence in progress.

4 Independent (or random) isolation failures N/A include those accidents in which the pre-existing isolation failure to seal is not dependent on the sequence in progress. This class is similar to Class 3 isolation failures, but is applicable to sequences involving Type B tests and their potential failures. These are the Type B-tested components that have isolated but exhibit excessive leakage.

5 Independent (or random) isolation failures N/A include those accidents in which the pre-existing isolation failure to seal is not dependent on the sequence in progress. This class is similar to Class 4 isolation failures, but is applicable to sequences involving Type C tests and their potential failures.

6 Containment isolation failures include those LERF-01, LERF-03, leak paths covered in the plant test and LERF-05, LERF-07, maintenance requirements or verified per in LERF-10, LERF-12, service inspection and testing (ISI/IST)

LERF-14, LERF-16 program.

Page 21

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 4-7: EPRI Containment Failure Classification (Reference 2)

VCSNS Level 2 Release Class Description Category Frequency 7

Accidents involving containment failure induced LATE, LERF-02, LERF-11 by severe accident phenomena. Changes in Appendix J testing requirements do not impact these accidents.

8 Accidents in which the containment is bypassed LERF-04, LERF-06, (either as an initial condition or induced by LERF-09, LERF-13, phenomena) are included in Class 8. Changes LERF-15, LERF-18 in Appendix J testing requirements do not impact these accidents.

4.3 Impact of Extension on Detection of Component Failures That Lead to Leakage (Small and Large)

The ILRT can detect a number of component failures such as liner breach, failure of certain bellow arrangements and failure of some sealing surfaces, which can lead to leakage. The proposed ILRT test interval extension may influence the conditional probability of detecting these types of failures. To ensure that this effect is properly accounted for, the EPRI Class 3 accident class as defined in Table 4-7, it is divided into two sub-classes, Class 3a and Class 3b, representing small and large leakage failures, respectively.

The probability of the EPRI Class 3a and 3b failures is determined consistent with the EPRI Guidance (Reference 21). For Class 3a, the probability is based on the maximum likelihood estimate of failure (arithmetic average) from the available data (i.e., 2 "small" failures in 217 tests leads to 2/217=0.0092). For Class 3b, Jefferys non-informative prior distribution is assumed for no "large" failures in 217 tests (i.e., 0.5 / (217+1) = 0.0023).

In a follow on letter (Reference 18) to their ILRT guidance document (Reference 3), NEI issued additional information concerning the potential that the calculated delta LERF values for several plants may fall above the "very small change" guidelines of the NRC Regulatory Guide 1.174. This additional NEI information includes a discussion of conservatisms in the quantitative guidance for delta LERF. NEI describes ways to demonstrate that, using plant specific calculations, the delta LERF is smaller than that calculated by the simplified method.

The supplemental information states:

The methodology employed for determining LERF (Class 3b frequency) involves conservatively multiplying the CDF by the failure probability for this class (3b) of accident. This was done for simplicity and to maintain conservatism. However, some plant specific accident classes leading to core damage are likely to include individual sequences that either may already (independently) cause a LERF or could never cause a LERF, and are thus not associated with a postulated large Type A containment leakage path (LERF). These contributors can be removed from Page 22

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Class 3b in the evaluation of LERF by multiplying the Class 3b probability by only that portion of CDF that may be impacted by type A leakage.

The application of this additional guidance to the analysis for VCSNS Unit 1, as detailed in Section 5, involves the following:

" The Class 2 and Class 8 sequences are subtracted from the CDF that is applied to Class 3b. To be consistent, the same change is made to the Class 3a CDF, even though these events are not considered LERF. Class 2 and Class 8 events refer to sequences with either large preexisting containment isolation failures or containment bypass events. These sequences are already considered to contribute to LERF in the VCSNS Unit 1 Level 2 PRA analysis.

Class 1 accident sequences may involve availability and or successful operation of containment sprays. It could be assumed that, for calculation of the Class 3b and 3a frequencies, the fraction of the Class 1 CDF associated with successful operation of containment sprays can also be subtracted.

However, in this assessment VCSNS Unit 1 does not credit containment spray as a means of reducing releases from Class 3 events.

Consistent with the NEI Guidance (Reference 3), the change in the leak detection probability can be estimated by comparing the average time that a leak could exist without detection. For example, the average time that a leak could go undetected with a three year test interval is 1.5 years (3 yr / 2), and the average time that a leak could exist without detection for a ten year interval is five years (10 yr/2). This change would lead to a non-detection probability that is a factor of 3.33 (5.0/1.5) higher for the probability of a leak that is detectable only by ILRT testing. An extension of the ILRT interval to fifteen years can be estimated to lead to about a factor of 5.0 (7.5/1.5) increase in the non-detection probability of a leak compared to a three year interval.

It should be noted that using the methodology discussed above is very conservative compared to previous submittals (e.g., the IP3 request for a one-time ILRT extension that was approved by the NRC (Reference 9)) because it does not factor in the possibility that the failures could be detected by other tests (e.g., the Type B local leak rate tests that will still occur.) Eliminating this possibility conservatively over-estimates the factor increases attributable to the ILRT extension.

4.4 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage An estimate of the likelihood and risk implications of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is evaluated using the methodology from the Calvert Cliffs liner corrosion analysis (Reference 5). The Calvert Cliffs analysis was performed for a concrete cylinder, dome and a concrete basemat, each with a steel liner. VCSNS Unit 1 has a similar type of containment (Reference 22, Page 5-1).

Page 23

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC The following approach is used to determine the change in likelihood, due to extending the ILRT, of detecting corrosion of the containment steel liner. This likelihood is then used to determine the resulting change in risk. Consistent with the Calvert Cliffs analysis, the following issues are addressed:

Differences between the containment basemat and the containment cylinder and dome The historical steel liner flaw likelihood due to concealed corrosion The impact of aging The corrosion leakage dependency on containment pressure The likelihood that visual inspections will be effective at detecting a flaw 4.4.1 Assumptions Consistent with the Calvert Cliffs analysis, a half failure is assumed for basemat concealed liner corrosion due to the lack of identified failures (See Table 4-8, Step 1)

The two corrosion events used to estimate the liner flaw probability in the Calvert Cliffs analysis are assumed to be applicable to this VCSNS Unit 1 containment analysis.

These events, one at North Anna Unit 2 and one at Brunswick Unit 2, were initiated from the nonvisible (backside) portion of the containment liner.

Consistent with the Calvert Cliffs analysis, the estimated historical flaw probability is based on 70 steel-lined containments.

The Calvert Cliffs analysis used the estimated historical liner flaw probability of 5.5 years to reflect the years since September 1996 when 10 CFR 50.55a started requiring visual inspection.

Additional success data was not used to limit the aging impact of this corrosion issue, even though inspections were being performed prior to this date. Two additional relevant liner corrosion events involving concealed corrosion (corrosion initiated on the inaccessible liner surface) are considered in this VCSNS Unit 1 report. These events occurred at Beaver Valley Unit 1 and D.C. Cook Unit 2 (Reference 25 and Reference 26, respectively).

The historical liner flaw probability using 13.25 years to account for the additional 7.75 year time period since the Calvert Cliffs analysis was submitted in 2002, results in a historical liner flaw likelihood of 4.3E-03/year ((2+2) / [70 * (5.5 + 7.75)] = 4.3E-03/year). This value is smaller than the value of 5.2E-03 which is used in the Calvert Cliffs analysis. The conservative value of 5.2E-03 will be used in this VCSNS Unit 1 report to remain consistent with the Calvert Cliffs analysis.

Consistent with the Calvert Cliffs analysis, the steel liner flaw likelihood is assumed to double every five years. This is based solely on judgment and is included in this analysis to address the increased likelihood of corrosion as the steel liner ages. (See Page 24

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 4-8, Steps 2 and 3). Sensitivity studies are included that address doubling this rate every ten years and every two years.

In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching the outside atmosphere given that a liner flaw exists was estimated as 1.1% for the cylinder and dome and 0.11% (10% of the cylinder failure probability) for the basemat. These values were determined from an assessment of the probability versus containment pressure, and the selected values are consistent with a pressure that corresponds to the ILRT target pressure of 37 psig. For VCSNS Unit 1, the containment failure probabilities are less than these values at 37 psig based on the containment fragility curve which is documented in the VCSNS Unit 1 IPE. A containment bypass model is utilized for LERF.

Conservative probabilities of 1% for the cylinder and dome and 0.1%

for the basemat are used in this analysis, and sensitivity studies are included that increase and decrease the probabilities by an order of magnitude (See Table 4-8, Step 4).

Consistent with the Calvert Cliffs analysis, the likelihood of leakage escape (due to crack formation) in the basemat region is considered to be less likely than the containment cylinder and dome region (See Table 4-8, Step 4).

Consistent with the Calvert Cliffs analysis, a 5% visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 10% is used.

To date, all liner corrosion events have been detected through visual inspection (See Table 4-8, Step 5). Sensitivity studies are included that evaluate total detection failure likelihood of 5% and 15%, respectively.

Consistent with the Calvert Cliffs analysis, all non-detectable containment failures are assumed to result in early releases. This approach avoids a detailed analysis of containment failure timing and operator recovery actions.

Page 25

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC 4.4.2 Analysis Table 4-8: Steel Liner Corrosion Base Case Step Description Containment Cylinder Containment Basemat and Dome 1

Historical Steel Liner Events: 2 Events: 0 Flaw Likelihood (assume half a failure)

(2)/(70

• 5.5) = 5.2E-03 Failure Data: Containment 0.5/(70

• 5.5) = 1.3E-03 location specific 2

Age Adjusted Steel Liner Year Failure Rate Year Failure Rate Flaw Likelihood 1

2.1E-03 1

5.0E-04 avg 5-10 5.2E-03 avg 5-10 1.3E-03 During 15-year interval, 15 1.4E-02 15 3.5E-03 assume failure rate doubles every five years (14.9% increase per year).

The average for 5th to 10th 15 year average =

15 year average =

year is set to the historical 6.27E-03 1.57E-03 failure rate (consistent with Calvert Cliffs analysis).

3 Flaw Likelihood at 3, 10, 0.71% (1 to 3 years) 0.18% (1 to 3 years) and 15 years 4.06% (1 to 10 years) 1.02% (1 to 10 years) 9.40% (1 to 15 years) 2.35% (1 to 15 years)

Uses age adjusted liner (Note that the Calvert (Note that the Calvert flaw likelihood (Step 2),

Cliffs analysis presents Cliffs analysis presents assuming failure rate the delta between 3 and the delta between 3 and doubles every five years 15 years of 8.7% to utilize 15 years of 2.2% to (consistent with Calvert in the estimation of the utilize in the estimation of Cliffs analysis - See delta-LERF value. For this the delta-LERF value.

Table 6 of Reference 5).

analysis, however, the For this analysis, values are calculated however, the values are based on the 3, 10, and calculated based on the 15 year intervals 3, 10, and 15 year consistent with the intervals consistent with intervals of concern in this the intervals of concern analysis.)

in this analysis.)

Page 26

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 4-8: Steel Liner Corrosion Base Case Step Description Containment Cylinder Containment Basemat and Dome 4

Likelihood of Breach in Containment Given Steel Liner Flaw The failure probability of 1%

0.1%

the cylinder and dome is assumed to be 1%

(compared to 1.1% in the Calvert Cliffs analysis).

The basemat failure probability is assumed to be a factor of ten less, 0.1%, (compared to 0.11%

in the Calvert Cliffs analysis).

5 Visual Inspection 10%

100%

Detection Failure Likelihood 5% failure to identify Cannot be visually visual flaws plus 5%

inspected.

Utilize assumptions likelihood that the flaw is consistent with Calvert not visible (not through-Cliffs analysis.

cylinder but could be detected by ILRT)

All events have been detected through visual inspection. 5% visible failure detection is a conservative assumption.

6 Likelihood of Non-0.00071% (at 3 years) 0.00018% (at 3 years)

Detected Containment 0.71%

• 1%

• 10%

0.18%

• 0.1%

• 100%

Leakage 0.0041% (at 10 years) 0.0010% (at 10 years) 4.1%* 1%* 10%

1.0%* 0.1%* 100%

(Steps 3

• 4* 5) 0.0094% (at 15 years) 0.0024% (at 15 years) 9.4%

• 1%

• 10%

2.4%

• 0.1%

• 100%

The total likelihood of the corrosion-induced, non-detected containment leakage is the sum of Step 6 for the containment cylinder and dome and the containment basemat as summarized below for VCSNS Unit 1.

Total Likelihood of Non-Detected Containment Leakage Due To Corrosion for VCSNS Unit 1:

At 3 years:

0.00071% + 0.00018% = 0.00089%

Page 27

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC At 10 years: 0.0041% + 0.0010% = 0.0051%

At 15 years: 0.0094% + 0.0024% = 0.012%

The above factors are applied to those core damage accidents that are not already independently LERF or that could never result in LERF. For example, the three in ten year case is calculated as follows:

Per Table 4-6, the VCSNS Unit 1 CDF associated with accidents that are not independently LERF or could never result in LERF are VCSNS Level 2 Release Categories INTACT, SERF and LATE. Therefore the VCSNS Unit 1 CDF associated with accidents that are not independently LERF or could never result in LERF is equal to 3.41 E-06/yr + 1.99E-08/yr + 9.62E-07/yr = 4.39E-06/yr.

Per Table 5-3, the EPRI Class 3b frequency is 1.01 7E-08/yr.

" The increase in the base case Class 3b frequency due to the corrosion-induced concealed flaw issue is calculated as 4.39-06/yr

• 0.00089% = 3.91E-11/yr, where 0.00089% was previously shown above to be the cumulative likelihood of non-detected containment leakage due to corrosion at three years.

" The three in ten year Class 3b frequency including the corrosion-induced concealed flaw issue is then calculated as 1.017E-08/yr + 3.91 E-1 1/yr = 1.021 E-08/yr.

5 Results The application of the approach based on the guidance contained in EPRI Report No. 1009325, Revision 2-A, Appendix H, EPRI-TR-104285 (Reference 2) and previous risk assessment submittals on this subject (References 5, 8, 19, 20) have led to the following results. The results are displayed according to the eight accident classes defined in the EPRI report. Table 5-1 lists these accident classes.

The analysis performed examined VCSNS Unit 1 specific accident sequences in which the containment remains intact or the containment is impaired. Specifically, the breakdown of the severe accidents contributing to risk were considered in the following manner:

Core damage sequences in which the containment remains intact initially and in the long term (EPRI TR-1 04285 Class 1 sequences).

Core damage sequences in which containment integrity is impaired due to random isolation failures of plant components other than those associated with Type B or Type C test components. For example, liner breach or bellows leakage. (EPRI TR-104285 Class 3 sequences).

Core damage sequences in which containment integrity is impaired due to containment isolation failures of pathways left "opened" following a plant post-maintenance test. (For Page 28

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC example, a valve failing to close following a valve stroke test. (EPRI TR-1 04285 Class 6 sequences). Consistent with the NEI Guidance, this class is not specifically examined since it will not significantly influence the results of this analysis.

Accident sequences involving containment bypassed (EPRI TR-104285 Class 8 sequences), large containment isolation failures (EPRI TR-1 04285 Class 2 sequences),

and small containment isolation "failure-to-seal" events (EPRI TR-1 04285 Class 4 and 5 sequences) are accounted for in this evaluation as part of the baseline risk profile.

However, they are not affected by the ILRT frequency change.

Class 4 and 5 sequences are impacted by changes in Type B and C test intervals; therefore, changes in the Type A test interval do not impact these sequences.

Table 5-1: Accident Classes Accident Classes (Containment Description Release Type) 1 No Containment Failure 2

Large Isolation Failures (Failure to Close) 3a Small Isolation Failures (Liner Breach) 3b Large Isolation Failures (Liner Breach) 4 Small Isolation Failures (Failure to Seal-Type B) 5 Small Isolation Failures (Failure to Seal-Type C) 6 Other Isolation Failures (e.g., Dependent Failures) 7 Failures Induced by Phenomena (Early and Late) 8 Bypass (Interfacing System LOCA)

CDF All CET End states (including Very Low and No Release)

The steps taken to perform this risk assessment evaluation are as follows:

Step 1 - Quantify the base-line risk in terms of frequency per reactor year for each of the eight accident classes presented in Table 5-1.

Step 2 - Develop plant specific person-rem dose (population dose) per reactor year for each of the eight accident classes.

Step 3 - Evaluate risk impact of extending Type A test interval from three to fifteen and ten to fifteen years.

Step 4 - Determine the change in risk in terms of Large Early Release Frequency (LERF) in accordance with RG 1.174.

Step 5 - Determine the impact on the Conditional Containment Failure Probability (CCFP).

Page 29

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC 5.1 Step 1 - Quantify the Base-Line Risk in Terms of Frequency Per Reactor Year As previously described, the extension of the Type A interval does not influence those accident progressions that involve large containment isolation failures, Type B or Type C testing, or containment failure induced by severe accident phenomena.

For the assessment of ILRT impacts on the risk profile, the potential for pre-existing leaks is included in the model. (These events are represented by the Class 3 sequences in EPRI TR-104285). The question on containment integrity was modified to include the probability of a liner breach or bellows failure (due to excessive leakage) at the time of core damage.

Two failure modes were considered for the Class 3 sequences. These are Class 3a (small breach) and Class 3b (large breach).

The frequencies for the severe accident classes defined in Table 5-1 were developed for VCSNS Unit 1 by first determining the frequencies for Classes 1, 2, 7 and 8 using the categorized sequences and the identified correlations shown in Table 4-6, scaling these frequencies to account for the uncategorized sequences, determining the frequencies for Classes 3a and 3b, and then determining the remaining frequency for Class 1.

Furthermore, adjustments were made to the Class 3b and hence Class 1 frequencies to account for the impact of undetected corrosion of the steel liner per the methodology described in Section 4.4.

The total frequency of the categorized sequences is 4.45E-06/yr, the total CDF is 4.50E-06/yr, and the scale factor is 1.011. The scaling factor is determined by dividing the total core damage frequency (including the uncategorized frequency) by the total categorized release category frequency (4.50E-06/4.45E-06 = 1.011). This process ensures that the CDF of 4.50E-06/yr is maintained for the determination of Class 3 states (see below) and effectively distributes the dose impact of the non-represented classes (Classes 4, 5, and 6) proportionately (per frequencies identified in the last column of Table 5-2) over the evaluated classes 1, 2, 7, and 8. The results are summarized below in Table 5-2 and in Table 5-3. Table 5-2 contains the frequencies from the categorized sequences, and the resulting frequencies due to the scale factor.

Page 30

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 5-2: VCSNS Unit I Categorized Accident Classes and Frequencies Frequency Based Adjusted Frequency EPRI VCSNS on Using Scale Class Release Categorized Factor of 1.011 Category Results (per yr)

(per yr) 1 Intact Containment 3.41 E-06 3.45E-06 (INTACT)

Containment Isolation 2

Failures (LERF-08 &

5.66E-09 5.73E-09 LERF-17)

Late Containment Failure (LATE) and High 7

Pressure Containment 9.62E-07 9.73E-07 Failure (LERF-02 &

LERF-1 1)

Containment Bypass (LERF-09 & LERF-18) 8 and SGTR (LERF-04, 6.76E-08 6.84E-08 LERF-06, LERF-13 &

LERF-1 5)

Total Frequency 4.45E-06 4.50E-06 Class 1 Sequences. This group consists of all core damage accident progression bins for which the containment remains intact (modeled as Technical Specification Leakage). The frequency per year is initially determined from the Containment Intact Level 2 Release Category listed in Table 4-6, minus the EPRI Class 3a and 3b frequency, which are calculated below.

Class 2 Sequences. This group consists of all core damage accident progression bins for which a failure to isolate the containment occurs. The frequency per year for these sequences is obtained from the Large Containment Isolation Failures Level 2 Release Category listed in Table 4-6.

Class 3 Sequences. This group consists of all core damage accident progression bins for which a pre-existing leakage in the containment structure (e.g., containment liner) exists.

The containment leakage for these sequences can be either small (in excess of design allowable but <1OLa) or large (>1OOLa).

The respective frequencies per year are determined as follows:

PROBciass_3a

= probability of small pre-existing containment liner leakage

= 0.0092 [see Section 4.3]

PROBc~ass_3b

= probability of large pre-existing containment liner leakage Page 31

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC

= 0.0023 [see Section 4.3]

As described in Section 4.3, additional consideration is made to not apply these failure probabilities on those cases that are already LERF scenarios (i.e., the Class 2 and Class 8 contributions).

Class 3a Frequency = 0.0092 * (CDF - (Class 2 + Class 8))

= 0.0092 * (4.50E-06/yr - (5.73E-09/yr + 6.84E-08/yr)) = 4.067E-08/yr Class 3b Frequency = 0.0023 * (CDF- (Class 2 + Class 8))

=0.0023 * (4.50E-06/yr - (5.73E-09/yr + 6.84E-08/yr)) = 1.01 7E-08/yr For this analysis, the associated containment leakage for Class 3a is 1OLa and for Class 3b is 10OLa. These assignments are consistent with the guidance provided in EPRI Report No. 1009325, Revision 2-A.

Note, in the above equations for the Class 3a and 3b release frequencies, the total adjusted release frequency from the last column of Table 5-2 has been substituted for CDF.

As discussed previously this process marginally over-estimates the Class 3 releases.

Class 4 Sequences. This group consists of all core damage accident progression bins for which containment isolation failure-to-seal of Type B test components occurs. Because these failures are detected by Type B tests which are unaffected by the Type A ILRT, this group is not evaluated any further in the analysis.

Class 5 Sequences. This group consists of all core damage accident progression bins for which containment isolation failure-to-seal of Type C test components occurs. Because the failures are detected by Type C tests which are unaffected by the Type A ILRT, this group is not evaluated any further in this analysis.

Class 6 Sequences. This group is similar to Class 2. These are sequences that involve core damage accident progression bins for which a failure-to-seal containment leakage due to failure to isolate the containment occurs. These sequences are dominated by misalignment of containment isolation valves following a test/maintenance evolution, typically resulting in a failure to close smaller containment isolation valves. All other failure modes are bounded by the Class 2 assumptions. Consistent with guidance provided in EPRI Report No. 1009325, Revision 2-A, this accident class is not explicitly considered since it has a negligible impact on the results.

Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs (e.g.,

overpressure). For this analysis, the frequency is determined from the Severe Accident Phenomena-Induced Failures Release Category from the VCSNS Unit 1 Level 2 results shown in Table 4-6.

Page 32

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Class 8 Sequences. This group consists of all core damage accident progression bins in which containment bypass occurs. For this analysis, the frequency is determined from the Containment Bypass Release Category from the VCSNS Unit 1 Level 2 results shown in Table 4-6.

5.1.1 Summary of Accident Class Frequencies In summary, the accident sequence frequencies that can lead to radionuclide release to the public have been derived consistent with the definitions of accident classes defined in EPRI-TR-104285 the NEI Interim Guidance, and guidance provided in EPRI Report No. 1009325, Revision 2-A. Table 5-3 summarizes these accident frequencies by accident class for VCSNS Unit 1.

Table 5-3: Radionuclide Release Frequencies as a Function of Accident Class (VCSNS Unit I Base Case)

Accident Frequency Classes (per x-yr)

(Containment Description (pereRCase (CntimetBase Case Base Case Release Type)

Plus Corrosion1 1

No Containment Failure 3.397E-06 3.397E-06 2

Large Isolation Failures 5.728E-09 5.728E-09 (Failure to Close)

Small Isolation Failures 3a 4.067E-08 4.067E-08 (liner breach) 3b Large Isolation Failures 1.017E-08 1.021 E-08 (liner breach)

Small Isolation Failures N/A N/A (Failure to seal-Type B)

Small Isolation Failures N/A N/A (Failure to seal-Type C) 6 Other Isolation Failures N/A N/A (e.g., dependent failures)

Failures Induced by 7

Phenomena (Early and 9.728E-07 9.728E-07 Late) 8 Bypass (Interfacing System 6.840E-08 6.840E-08 LOCA)

CDF All CET end states 4.495E-06 4.495E-06

1. Note that this the corrosion.

is based on data developed in Section 4.4. Only Class 3b is impacted by 5.2 Step 2

- Develop Plant Specific Person-Rem Dose (Population Dose) Per Reactor Year Plant specific release analyses were performed to estimate the person-rem doses to the population within a 50-mile radius from the plant. The releases are based on information Page 33

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC provided by NUREG/CR-4551 with adjustments made for the site demographic differences compared to the reference plant as described in Section 4.2, and summarized in Table 4-5.

The results of applying these releases to the EPRI containment failure classification are as follows:

Class 1 = (1.35E+03 person-rem (at 1.OLa) + 1.83E+03 person-rem (at 1.OLa)) / 2

=

1.59E+03 person-rem (1)

Class 2 = 1.03E+06 person-rem (2)

Class 3a = 1.59E+03 person-rem x 1 OLa = 1.59E+04 person-rem (3)

Class 3b = 1.59E+03 person-rem x 1 00La = 1.59E+05 person-rem (3)

Class 4 = Not analyzed Class 5 = Not analyzed Class 6 = Not analyzed Class 7 = 7.88E+05 person-rem (4)

Class 8 = 1.29E+06 person-rem (5)

(1) The derivation is described in Section 4.2 for VCSNS Unit 1. Class 1 is assigned the dose from the "no containment failure" APBs from NUREG/CR-4551 (i.e., APB #6 and APB #7). The dose is calculated as an arithmetic average of the dose for these bins and is bounding.

(2) The Class 2, containment isolation failures, dose is assigned from APB #2 (Early CF).

(3) The Class 3a and 3b dose are related to the Class 1 leakage rate as shown. While no pre-existing leakage in excess of 21 La has been identified for any historical ILRT event, Class 3b releases are conservatively assessed at 100La. Class 3a releases are conservatively assessed at 10La. This is consistent with the guidance provided in EPRI Report No. 1009325, Revision 2-A.

(4) The Class 7 dose is assigned from APB #4 (Late CF).

(5) Class 8 sequences involve containment bypass failures; as a result, the person-rem dose is not based on normal containment leakage. The releases for this class are assigned from APB #5 (Bypass).

In summary, the population dose estimates derived for use in the risk evaluation per the EPRI methodology (Reference 2) containment failure classifications, and consistent with the NEI guidance (Reference 3) as modified by EPRI Report No. 1009325, Revision 2-A are provided in Table 5-4.

Page 34

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 5-4: VCSNS Unit I Population Dose Estimates for Population Within 50 Miles Accident Classes Person-Rem (Containment Description (50 miles)

Release Type) 1 No Containment Failure 1.59E+03 2

Large Isolation Failures (Failure to Close) 1.03E+06 3a Small Isolation Failures (liner breach) 1.59E+04 3b Large Isolation Failures (liner breach) 1.59E+05 4

Small Isolation Failures (Failure to seal-Type B)

N/A 5

Small Isolation Failures (Failure to seal-Type C)

N/A 6

Other Isolation Failures (e.g., dependent failures)

N/A 7

Failures Induced by Phenomena (Early and Late) 7.88E+05 8

Bypass (Interfacing System LOCA) 1.29E+06 The above dose estimates, when combined with the results presented in Table 5-3, yield the VCSNS Unit 1 baseline mean consequence measures for each accident class. These results are presented in Table 5-5.

Page 35

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 5-5: VCSNS Unit I Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 3/10 Years Accident EPRI EPRI Methodology Change Due Classes Person-Methodology Plus Corrosion to Corrosion (Cnmt Description Rem (50 quency PPerson-Release miles)

Frequency Rem/yr Frequency Person-Rem/yr Remlyr(1)

Type)

(per Rx-yr)

(0mils (per Rx-yr)

(50 miles)

Type)(50 miles) 1 No Containment 1.59E+03 3.40E-06 5.41 E-03 3.40E-06 5.41 E-03

-6.22E-08 Failure (2)

Large Isolation 2

Failures (Failure to 1.03E+06 5.73E-09 5.89E-03 5.73E-09 5.89E-03 O.OOE+00 Close)

Small Isolation 3a Failures (liner 1.59E+04 4.07E-08 6.47E-04 4.07E-08 6.47E-04 O.OOE+00 breach)

Large Isolation 3b Failures (liner 1.59E+05 1.02E-08 1.62E-03 1.02E-08 1.62E-03 6.22E-06 breach)

Small Isolation 4

Failures (Failure to N/A N/A N/A N/A N/A N/A seal -Type B)

Small Isolation 5

Failures (Failure to N/A N/A N/A N/A N/A N/A seal-Type C)

Other Isolation 6

Failures (e.g.,

N/A N/A N/A N/A N/A N/A dependent failures)

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 5-5: VCSNS Unit I Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 3/10 Years Accident EPRI EPRI Methodology Change Due Classes Person-Methodology Plus Corrosion to Corrosion (Cnmt Description Rem (50 Person-Frequency Person-Rem/yr Person-Release miles)

Frequency Persnmrquny Prsnyelr Rem/yr(1)

Type)

(per Rx-yr)

Remlyr (per Rx-yr)

(50 miles)

Type)__

(50 miles)

Failures Induced 7

by Phenomena 7.88E+05 9.73E-07 7.67E-01 9.73E-07 7.67E-01 O.OOE+00 (Early and Late)

Bypass 8

(Interfacing 1.29E+06 6.84E-08 8.85E-02 6.84E-08 8.85E-02 O.OOE+00 System LOCA)

CDF All CET end states N/A 4.50E-06 8.69E-01 4.50E-06 8.69E-01 6.16E-06

1) Only release Classes 1 and 3b are affected by the corrosion analysis.

2) Characterized as 1 La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs.

Release classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

Page 37

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC 5.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval From 10 to 15 Years The next step is to evaluate the risk impact of extending the test interval from its current ten year value to fifteen years. To do this, an evaluation must first be made of the risk associated with the ten year interval since the base case applies to a three year interval (i.e., a simplified representation of a three in ten interval).

5.3.1 Risk Impact Due to 10-year Test Interval As previously stated, Type A tests impact only Class 3 sequences. For Class 3 sequences, the release magnitude is not impacted by the change in test interval (a small or large breach remains the same, even though the probability of not detecting the breach increases). Thus, only the frequency of Class 3a and 3b sequences is impacted.

The risk contribution is changed based on the NEI guidance as described in Section 4.3 by a factor of 3.33 compared to the base case values. The results of the calculation for a ten year interval are presented in Table 5-6.

5.3.2 Risk Impact Due to 15-Year Test Interval The risk contribution for a fifteen year interval is calculated in a manner similar to the ten year interval. The difference is in the increase in probability of leakage in Classes 3a and 3b. For this case, the value used in the analysis is a factor of 5.0 compared to the three year interval value, as described in Section 4.3.The results for this calculation are presented in Table 5-7.

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 5-6 VCSNS Unit I Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/10 Years Accident EPRI Methodology EPRI Methodology Change Due Classes Person-Plus Corrosion to Corrosion (Cnmt Description Rem (50 Person-Person-Person-Release miles)

Frequency Rem/yr Frequency Rem/yr Rem/yr(1)

Type)

(per Rxyr)

(50 miles)

(per Rxyr)

(50 miles) 1 No Containment 1.59E+03 3.28E-06 5.22E-03 3.28E-06 5.22E-03

-2.07E-07 Failure (2)

Large Isolation 2

Failures (Failure to 1.03E+06 5.73E-09 5.89E-03 5.73E-09 5.89E-03 O.OOE+00 Close)

Small Isolation 3a Failures (liner 1.59E+04 1.35E-07 2.16E-03 1.35E-07 2.16E-03 O.OOE+00 breach)

Large Isolation 3b Failures (liner 1.59E+05 3.39E-08 5.39E-03 3.40E-08 5.41 E-03 2.07E-05 breach)

Small Isolation 4

Failures(Failure to N/A N/A N/A N/A N/A N/A seal-Type B)

Small Isolation 5

Failures (Failure to N/A N/A N/A N/A N/A N/A seal-Type C)

Other Isolation 6

Failures (e.g.,

N/A N/A N/A N/A N/A N/A dependent failures)

Failures Induced by 7

Phenomena (Early 7.88E+05 9.73E-07 7.67E-01 9.73E-07 7.67E-01 O.OOE+00 and Late)

I

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 5-6 VCSNS Unit 1 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/10 Years Accident EPRI Methodology EPRI Methodology Change Due Classes Person-Plus Corrosion to Corrosion (Cnmt Description Rem (50 Person-Frequency Person-Person-Release miles)

Frequency Rem/yr Freqeny Rem/yr Remlyr(1)

Type)

(per Rx-yr)

(50 miles)

(per Rx-yr)

(50 miles) 8 Bypass (Interfacing 1.29E+06 6.84E-08 8.85E-02 6.84E-08 8.85E-02 0.OOE+00

________ System LOCA)_____

CDF All CET end states N/A 4.50E-06 8.74E-01 4.50E-06 8.74E-01 2.05E-05

1) Only release Classes 1 and 3b are affected by the corrosion analysis.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs.

Release classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

Page 40

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 5-7: VCSNS Unit I Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/15 Years Accident EPRI Methodology EPRI Methodology Plus Change Due Classes Person-Rem Corrosion to Corrosion (Cnmt Description (50 Person-Person-Person-Release miles)

Frequency Rem/yr Frequency Rem/yr Rem/yr(l)

Type)

(50 miles)

(per Rx-yr)

(50 miles) 1 No Containment 1.59E+03 3.19E-06 5.08E-03 3.19E-06 5.08E-03

-3.11E-07 Failure (2)

Large Isolation 2

Failures (Failure to 1.03E+06 5.73E-09 5.89E-03 5.73E-09 5.89E-03 O.OOE+00 Close)

Small Isolation 3a Failures (liner 1.59E+04 2.03E-07 3.24E-03 2.03E-07 3.24E-03 O.OOE+00 breach)

Large Isolation 3b Failures (liner 1.59E+05 5.08E-08 8.09E-03 5.1OE-08 8.12E-03 3.11E-05 breach)

Small Isolation 4

Failures (Failure to N/A N/A N/A N/A N/A N/A seal-Type B)

Small Isolation 5

Failures (Failure to N/A N/A N/A N/A N/A N/A seal-Type C)

Other Isolation 6

Failures (e.g.,

N/A N/A N/A N/A N/A N/A dependent failures)

Failures Induced by 7

Phenomena (Early 7.88E+05 9.73E-07 7.67E-01 9.73E-07 7.67E-01 O.OOE+00 and Late)

Page 41

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 5-7: VCSNS Unit I Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/15 Years Accident EPRI Methodology EPRI Methodology Plus Change Due Classes Person-Rem Corrosion to Corrosion (Cnmt Description (50 Person-F Person-Person-Release miles)

Frequency Rem/yr Frequency Rem/yr Rem/yr(1)

Type)(per Rx-yr) miles)

(per Rx-yr)

(50 miles) 8 Bypass (Interfacing 1.29E+06 6.84E-08 8.85E-02 6.84E-08 8.85E-02 O.OOE+00 System LOCA)

CDF All CET end states N/A 4.50E-06 8.78E-01 4.50E-06 8.78E-01 3.08E-05

1) Only release Classes 1 and 3b are affected by the corrosion analysis.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs.

Release classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

Page 42

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC 5.4 Step 4 - Determine the Change in Risk in Terms of Large Early Release Frequency (LERF)

The risk increase associated with extending the ILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from an intact containment could in fact result in a larger release due to the increase in probability of failure to detect a pre-existing leak. With strict adherence to the EPRI guidance, 100% of the Class 3b contribution would be considered LERF.

Regulatory Guide 1.174 provides guidance for determining the risk impact of plant specific changes to the licensing basis. RG 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 10-6/yr and increases in LERF below 10-7/yr, and small changes in LERF as below 10-6/yr. Because the ILRT does not impact CDF, the relevant metric is LERF.

For VCSNS Unit 1, 100% of the frequency of Class 3b sequences can be used as a very conservative first-order estimate to approximate the potential increase in LERF from the ILRT interval extension (consistent with the EPRI guidance methodology). The baseline LERF based on a test frequency of three times in ten years is 1.02E-08/yr. Based on a ten year test interval from Table 5-6, the Class 3b frequency (conservatively including corrosion) is 3.40E-08/yr; and, based on a fifteen-year test interval from Table 5-7, it is 5.10E-08/yr. Thus, the increase in the overall probability of LERF due to Class 3b sequences that is due to increasing the ILRT test interval from three to fifteen years is 4.08E-08/yr as shown in Table 5-8. Similarly, the increase due to increasing the interval from ten to fifteen years is 1.70E-08/yr as shown in Table 5-8. As can be seen, even with the conservatisms included in the evaluation (per the EPRI methodology), the estimated change in LERF for VCSNS Unit 1 is below the threshold criteria for a very small change when comparing both the fifteen year results to the current ten year requirement, and the fifteen year results compared to the original three year requirement. See Table 5-8 for more information.

5.5 Step 5 -

Determine the Impact on the Conditional Containment Failure Probability (CCFP)

Another parameter that the NRC guidance in RG 1.174 states can provide input into the decision-making process is the change in the conditional containment failure probability (CCFP). The change in CCFP is indicative of the effect of the ILRT on all radionuclide releases, not just LERF. The CCFP can be calculated from the results of this analysis.

One of the difficult aspects of this calculation is providing a definition of the "failed containment." In this assessment, the CCFP is defined such that containment failure includes all radionuclide release end states other than the intact state. The conditional part of the definition is conditional given a severe accident (i.e., core damage).

The change in CCFP can be calculated by using the method specified in the EPRI Report No. 1009325, Revision 2-A. The NRC has previously accepted similar calculations (Reference 9) as the basis for showing that the proposed change is consistent with the defense-in-depth philosophy. The list below shows the CCFP values

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC that result from the assessment for the various testing intervals including corrosion effects.

CCFP = [1 - (Class 1 frequency + Class 3a frequency) / CDF]

• 100%

CCFP 3 = [1 - (3.40E-06/yr + 4.07E-08/yr) / 4.50E-06/yr]

• 100% = 23.54%

CCFP 3 = 23.54%

CCFP10 = [1 - (3.28E-06/yr + 1.35E-07/yr) /4.50E-06/yr]

• 100% = 24.11%

CCFP10 = 24.11%

CCFP 15 = [1 - (3.19E-06/yr + 2.03E-07/yr) / 4.50E-06/yr]

• 100% = 24.60%

CCFP 15 = 24.60%

ACCFP = CCFP 15-CCFP 3 = 1.06%

ACCFP = CCFP 15 - CCFP10= 0.49%

ACCFP = CCFP1O - CCFP 3= 0.57%

The change in CCFP of approximately 1.0% by extending the test interval to fifteen years from the original three in ten year requirement is judged to be very small.

5.6 Summary of Results The results from this ILRT extension risk assessment for VCSNS Unit 1 are summarized in Table 5-8.

Page 44

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 5-8: VCSNS Unit I ILRT Cases: Base, 3 to 10, and 3 to 15 Yr Extensions (Including Age Adjusted Steel Liner Corrosion Likelihood)

Base Case Extend to Extend to 3 in 10 Years I in 10 Years I in 15 Years EPRI DOSE Class Per-Rem Per-Per-CDFIYr Per-Rem/Yr Rem/Yr Rem/Yr 1

1.59E+03 3.40E-06 5.41E-03 3.28E-06 5.22E-03 3.19E-06 5.08E-03 2

1.03E+06 5.73E-09 5.89E-03 5.73E-09 5.89E-03 5.73E-09 5.89E-03 3a 1.59E+04 4.07E-08 6.47E-04 1.35E-07 2.16E-03 2.03E-07 3.24E-03 3b 1.59E+05 1.02E-08 1.62E-03 3.40E-08 5.41E-03 5.1OE-08 8.12E-03 7

7.88E+05 9.73E-07 7.67E-01 9.73E-07 7.67E-01 9.73E-07 7.67E-01 8

1.29E+06 6.84E-08 8.85E-02 6.84E-08 8.85E-02 6.84E-08 8.85E-02 Total N/A 4.50E-06 8.69E-01 4.50E-06 8.74E-01 4.50E-06 8.78E-01 ILRT Dose Rate from 3a D 3b P r-m 2.27E-03 7.56E-03 1.14E-02 3a and 3b Per-Rem/Yr Delta Total From 3 yr N/A 5.1OE-03 8.76E-03 Dose Rate1 From 10 yr N/A N/A 3.66E-03 change From 3 yr N/A 0.59%

1.01%

in dose rate from From 10 yr N/A N/A 0.42%

base 3b Frequency (LERF) 1.02E-08 3.40E-08 5.10E-08 Per-Rem/Yr Delta From 3 yr N/A 2.38E-08 4.08E-08 LERF From 10 yr N/A N/A 1.70E-08 CCFP %

23.54%

24.11%

24.60%

Delta From 3 yr N/A 0.57%

1.06%

CCFP %

From 10 yr N/A N/A 0.49%

1 The overall difference in total dose rate is less than the difference of only the 3a and 3b categories between two testing intervals. This is because the overall total dose rate includes contributions from other categories that do not change as a function of time, e.g., the EPRI Class 2 and 8 categories, and also due to the fact that the Class 1 person-rem/yr decreases when extending the IRLT frequency.

Page 45

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC 6

Sensitivities 6.1 Sensitivity to Corrosion Impact Assumptions The results in Tables 5-5, 5-6 and 5-7 show that including corrosion effects calculated using the assumptions described in Section 4.4 does not significantly affect the results of the ILRT extension risk assessment.

Sensitivity cases were developed to gain an understanding of the sensitivity of the results to the key parameters in the corrosion risk analysis. The time for the flaw likelihood to double was adjusted from every five years to every two and every ten years.

The failure probabilities for the cylinder and dome and the basemat were increased and decreased by an order of magnitude. The total detection failure likelihood was adjusted from 10% to 15% and 5%. The results are presented in Table 6-1. In every case the impact from including the corrosion effects is very minimal. Even the upper bound estimates with very conservative assumptions for all of the key parameters yield increases in LERF due to corrosion of only 3.90E-12/yr. The results indicate that even with very conservative assumptions, the conclusions from the base analysis would not change.

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 6-1: Steel Liner Corrosion Sensitivity Cases Age Containment Visual Increase in Class 3b (Step 3 in Breach Inspection &

Frequency the (Step 4 in the Non-Visual (LERF) for ILRT corrosion corrosion Flaws Extension 3 to 15 analysis) analysis)

(Step 5 in the Years (per Rx-yr) corrosion Total Increase Due analysis)

Increase to Corrosion Base Case Base Case Base Case Doubles (1% Cylinder, 10% Cylinder, 4.08E-08 1.56E-10 every 5 yrs 0.1% Basemat) 100% Basemat Doubles Base Base 4.10E-08 2.78E-10 every 2 yrs Doubles Base Base 4.07E-08 4.51E-11 every 10 yrs Base Base 15%

4.09E-08 2.19E-10 Base Base 5%

4.08E-08 9.40E-1 1 Base 10% Cylinder, 1%

Base 4.22E-08 1.56E-09 Basemat Base 0.1% Cylinder, Base 4.07E-08 1.56E-11 0.01% Basemat Lower Bound Doubles 0.1% Cylinder, 5% Cylinder, 4.07E-08 2.71 E-15 every 10 yrs 0.01% Basemat 100% Basemat Upper Bound Doubles 10% Cylinder, 1%

15% Cylinder, 407E-08 3.90E-12 every 2 yrs Basemat 100% Basemat 6.2 Sensitivity to Class 3B Contribution to LERF The Class 3b frequency for the base case of a three in ten year ILRT interval is 1.02E-08/yr (Table 5-5). Extending the interval to one in ten years results in a frequency of 3.40E-08/yr (Table 5-6). Extending it to one in fifteen years results in a frequency of 5.10E-08/yr (Table 5-7), which is an increase of 4.08E-08/yr from three in ten years to once in fifteen years. If 100% of the Class 3b sequences are assumed to have potential releases large enough for LERF, then the increase in LERF due to extending the interval from three in ten to one in fifteen is below the RG 1.174 threshold for very small changes in LERF of 1.OOE-07/yr.

6.3 Potential Impact From External Events Contribution The latest information related to external events for VCSNS is from the Individual Plant Examination for External Events (IPEEE) submittal (Reference 27). The external events considered included seismic, fire, high winds, external flooding, and nearby facility and transportation accidents. Programs are underway to update the seismic and fire Page 47

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC assessments, but they are not available for use in this evaluation of the ILRT extension.

The following is based on the results of the IPEEE evaluations.

Seismic Assessment The seismic assessment implemented a focused scope Seismic Margins Assessment (SMA) which consisted of walkdowns that concentrated on potential seismic vulnerabilities for equipment, large tanks, distribution systems, and structures.

The focus of the walkdowns was to assess the ability of this equipment to withstand the defined 0.3 g Review Level Earthquake (RLE) and still provide its safe shutdown function. The walkdowns did not identify any outliers that were operability issues at the plant and there was one case where the design practices were enhanced.

High Confidence of Low Probability of Failure (HCLPF) values were determined for those components that potentially had a capacity below the 0.3 g RLE. These included three large flat bottom tanks, the service water pond dams, and neutral grounding resistors.

The HCLPF values for the flat bottom tanks and resistors were determined to be greater than 0.3 g and the most conservative HCLPF for the service water pond dams was estimated to be 0.22 g. These values are all well above the Safe Shutdown Earthquake level of 0.15 g. Based on this, the seismic risk at VCSNS is not expected to be a large contributor to CDF or LERF and have little impact on the results of the ILRT extension assessment. However, for the purpose of this analysis, a seismic CDF value of 1.OOE-05/yr will be used. This is based on VCSNS being a more recently built plant, relative to other plants, and that no significant seismic related plant safety issues were identified during the SMA. Furthermore, based on the IPEEE results, this is consistent with other more recently built east coast plants.

Hi-gh Winds Assessment An evaluation of high winds was performed following the progressive screening methodology in NUREG-1407.

This included a review of the external high winds hazards at the plant and the licensing basis, an assessment of whether there have been any significant changes since the Operating License was issued, a site walkdown, and screening to determine if the plant meets the 1975 Standard Review Plan (SRP) criteria.

The results indicated there are no new high wind hazards since the Operating Licensing was issued, that the plant does comply with the 1975 SRP criteria, and that the frequency of event (design basis wind) is acceptable and low compared to the screening criteria (less than 1.OOE-06/yr).

Based on this, the high wind risk at VCSNS is not expected to be a significant contributor to CDF or LERF, and will have little impact on the results of the ILRT extension assessment. For the purpose of this analysis, a conservative high wind contribution to CDF of 1.00E-06/yr will be used.

External Flood Assessment An evaluation of external flooding was performed following the progressive screening methodology in NUREG-1407. This included a review of the external flood hazards in the vicinity of the plant, a review of the licensing basis, an assessment of whether there have been any significant changes since the Operating License was issued, a site Page 48

Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC walkdown, and a screening to determine if the plant meets the 1975 SRP criteria. As noted in the FSAR, VCSNS is considered a "dry site" which means the site is not subject to stream or river flooding due to topographic conditions. The results concluded that the plant was designed to provide adequate protection for safety-related structures, components, and systems from external flood hazards. Since the Operating License was issued, there have been no changes to indicate that the external flood hazards are significantly greater than originally considered or that the changes at the plant have created new vulnerabilities.

Based on this, the external flood risk at VCSNS is not expected to be a significant contributor to CDF or LERF, and will have little impact on the results of the ILRT extension assessment. Since this is considered a "dry site" and no issues were identified during the IPEEE, the contribution of external floods to CDF is expected to be very small. For the purpose of this analysis, the contribution to CDF from external flooding will be zero.

Transportation and Nearby Facility Accidents Assessment An evaluation of transportation and nearby facility hazards was performed following the progressive screening methodology in NUREG-1407.

This included a review of the external hazards related to transportation and nearby facility accidents in the vicinity of the plant, a review of the licensing basis, an assessment of whether there have been any significant changes since the Operating License was issued, and a screening to determine if the plant meets the 1975 SRP criteria. The results concluded that the plant complies with the applicable acceptance criteria of the 1975 SRP and since the Operating License was issued there have been no significant changes with regard to industrial, military, and transportation facilities near the site. Therefore, it was concluded that there have been no changes that indicate that these hazards are significantly greater than originally considered in the plant design or that changes at the plant have created new vulnerabilities. Based on this, the transportation and nearby facility risk at VCSNS is not expected to be a significant contributor to CDF or LERF, and will have little impact on the results of the ILRT extension assessment. Since the contribution of transportation and nearby facility accidents to CDF is expected to be very small, for the purpose of this analysis, the contribution to CDF from transportation and nearby facility accidents will be zero.

Fire Assessment The fire assessment (Reference 28) followed the Fire-Induced Vulnerability Evaluation (FIVE).

Following the FIVE screening analysis method, a PRA assessment of unscreened fire scenarios was completed and fire induced CDF values were determined.

For VCSNS the fire CDF was 8.52E-05/yr.

Five fire areas/zones were identified with CDF contributions greater than 1.OOE-06/yr. These were two of the ESF Switchgear Rooms, Relay Room, Control Room, and Turbine Building.

Based on insights from this evaluation, VCSNS enhanced Station fire response by revising operator training, fire brigade training and the Fire Emergency Procedures. However, for the purpose of this sensitivity evaluation, the original IPEEE fire CDF value (8.52E-05/yr) will be utilized.

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC External Events Summary Table 6-2 below lists the VCSNS CDF values for each external event type that are used to determine the potential impact from the External Events contribution.

Table 6-2: VCSNS External Events Summary External Event Type CDF Seismic 1.OOE-05/yr High Winds 1.OOE-06/yr External Flood O.OOE+00/yr Transportation and Nearby 0.00E+00/yr Facility Accidents Fire 8.52E-05/yr Combining the External Events CDF values and the Internal Events CDF of 4.34E-06/yr results in a total CDF value of 1.01E-04/yr. Note that the External Event CDF values used in this analysis are conservative.

The change in LERF from extending the Type A test interval can be conservatively estimated using the total CDF value of 1.01E-04/yr to determine the external event contribution. This CDF value was specifically used to determine the Class 3b frequency based on the external events contribution. The factors for determining the increase in the non-detection probability of a leak described in Section 4.3 were applied to the Class 3b base value frequency to determine the 3b frequencies for the once per ten year test and once per fifteen year test.

Class 3b Frequency (three per ten year test) = 0.0023 * (CDF- (Class 2 + Class 8)

=0.0023 * (1.01 E-04/yr - (5.73E-09/yr + 6.84E-08/yr)) = 2.32E-07/yr Class 3b Frequency (once per ten year test) = 3.33

• 2.32E-07/yr = 7.73E-07/yr Class 3b Frequency (once per fifteen year test) = 5.00

• 2.32E-07/yr = 1.16E-06/yr Table 6-3 shows the results of these calculations.

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC Table 6-3: VCSNS Estimated Total LERF Including External Events Impact 3b 3b 3b 3b 3b 3b LERF Increase Frequency Frequency Frequency (3 per 10 to Case (3 per 10 (1 per 10 (1 per 15

(

per 10t year test) year test) year test)

I per 15)

Internal Events Contribution 1.02E-08 3.40E-08 5.1OE-08 4.08E-08 (From Table 5-8)

External Events Contribution 2.32E-07 7.73E-07 1.16E-06 9.28E-07 Combined (ined 2.42E-07 8.07E-07 1.21 E-06 9.69E-07 (Internal and External Events)

Combining the Internal and External Events LERF values results in a total LERF value of 1.21E-06/yr, following the ILRT and remains well below the Regulatory Guide 1.174 criteria of 1.OOE-05/yr following the ILRT extension. Furthermore, the increase in total LERF from the three per ten year test to the once per fifteen year test is 9.69E-07/yr, which is within the range of the Regulatory Guide 1.174 criteria of 1.OOE-07/yr to 1.OOE-06/yr for a small change in risk.

7 Conclusions Based on the results from Section 5 and the sensitivity calculations presented in Section 6, the following conclusions regarding the assessment of the plant risk are associated with permanently extending the Type A ILRT test frequency to once in fifteen years:

Reg. Guide 1.174 (Reference 4) provides guidance for determining the risk impact of plant specific changes to the licensing basis. Reg. Guide 1.174 defines very small changes in risk as resulting in increases of CDF below 1006/yr and increases in LERF below 10-7/yr. Since the ILRT does not impact CDF, the relevant criterion is LERF.

The increase in LERF resulting from a change in the Type A ILRT test interval from three in ten years to one in fifteen years is conservatively estimated as 4.08E-08/yr using the EPRI guidance as written. As such, the estimated change in LERF is determined to be "very small" using the acceptance guidelines of Reg. Guide 1.174.

Regulatory Guide 1.174 (Reference 4) also states that when the calculated increase in LERF is in the range of 1.OOE-07 per reactor year to 1.OOE-06 per reactor year, applications will be considered only if it can be reasonably shown that the total LERF is less than 1.OOE-05 per reactor year. An additional assessment of the impact from External Events was also made. In this case, the total LERF including External Events was conservatively estimated as 1.21 E-06/yr for VCSNS. This is below the RG 1.174 acceptance criteria for total LERF of 1.OOE-05/yr and therefore this change satisfied both the incremental and absolute expectations with regard to the RG 1.174 LERF metric.

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC The change in Type A test frequency to once per fifteen years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 8.76E-03 person-rem/yr. Note that this value is based on internal events only and does not consider external events. EPRI Report No. 1009325, Revision 2-A states that a very small population dose is defined as an increase of

<1.0 person-rem per year or <1 % of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. This is consistent with the NRC Final Safety Evaluation for NEI 94-01 and EPRI Report No. 1009325 (Reference 29).

Moreover, the risk impact when compared to other severe accident risks is negligible.

" The increase in the conditional containment failure probability from the three in ten year interval to a permanent one time in fifteen year interval is 1.06%. EPRI Report No. 1009325, Revision 2-A states that increases in CCFP of 51.5 percentage points are very small. This is consistent with the NRC Final Safety Evaluation for NEI 94-01 and EPRI Report No. 1009325 (Reference 29). Therefore this increase is judged to be very small.

Therefore, permanently increasing the ILRT interval to fifteen years is considered to be a very small change to the V.C. Summer Nuclear Station risk profile.

7.1.1 Previous Assessments The NRC in NUREG-1493 (Reference 6) has previously concluded that:

Reducing the frequency of Type A tests (ILRTs) from three per ten years to one per twenty years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B and C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements.

Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage rate tests is possible with minimal impact on public risk. Beyond testing the performance of containment penetrations, ILRTs also test the integrity of the containment structure.

The findings for VCSNS Unit 1 confirm these general findings on a plant specific basis considering the severe accidents evaluated for VCSNS Unit 1, the VCSNS Unit 1 containment failure modes, and the local population surrounding VCSNS Unit 1.

8 References

1. Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, NEI 94-01, Revision 3-A, July 2012.

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC

2. Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, EPRI, Palo Alto, CA EPRI TR-104285, August 1994.

3.

Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Rev. 4, Developed for NEI by EPRI and Data Systems and Solutions, November 2001.

4. An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 2, May 2011.

5. Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, Letter from Mr. C.

H. Cruse (Calvert Cliffs Nuclear Power Plant) to NRC Document Control Desk, Docket No. 50-317, March 27, 2002.

6. Performance-Based Containment Leak-Test Program, NUREG-1493, September 1995.

7. Evaluation of Severe Accident Risks: Surry Unit 1, Main Report NUREG/CR-4551, SAND86-1309, Volume 3, Revision 1, Part 1, December 1990.

8. Letter from R. J. Barrett (Entergy) to U.S. Nuclear Regulatory Commission, IPN-01-007, January 18, 2001.

9. United States Nuclear Regulatory Commission, Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Frequency of Performance-Based Leakage Rate Testing (TAC No. MB01 78), April 17, 2001.

10. Impact of Containment Building Leakage on LWR Accident Risk, Oak Ridge National Laboratory, NUREG/CR-3539, ORNL/TM-8964, April 1984.

11. Reliability Analysis of Containment Isolation Systems, Pacific Northwest Laboratory, NUREG/CR-4220, PNL-5432, June 1985.

12. Technical Findings and Regulatory Analysis for Generic Safety Issue I1.E.4.3

'Containment Integrity Check', NUREG-1273, April 1988.

13. Review of Light Water Reactor Regulatory Requirements, Pacific Northwest Laboratory, NUREG/CR-4330, PNL-5809, Vol. 2, June 1986.

14. Shutdown Risk Impact Assessment for Extended Containment Leakage Testing Intervals Utilizing ORAMTM, EPRI, Palo Alto, CA TR-105189, Final Report, May 1995.

15. Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, NUREG-1 150, December 1990.

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Westinghouse Non-Proprietary Class 3 WESTINGHOUSE ELECTRIC COMPANY LLC

16. United States Nuclear Regulatory Commission, Reactor Safety Study, WASH-1400, October 1975.

17. V.C. Summer Nuclear Station Units 2 and 3 Updated Final Safety Analysis Report, Revision 0.

18. Anthony R. Pietrangelo, One-time extensions of containment integrated leak rate test interval -

additional information, NEI letter to Administrative Points of Contact, November 30, 2001.

19. Letter from J.A. Hutton (Exelon, Peach Bottom) to U.S. Nuclear Regulatory Commission, Docket No. 50-278, License No. DPR-56, LAR-01-00430, dated May 30, 2001.

20. Letter from D.E. Young (Florida Power, Crystal River) to U.S. Nuclear Regulatory Commission, 3F0401-11, dated April 25, 2001.

21. Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325, EPRI, Palo Alto, CA: 2008. 1018243.

22. V.C. Summer Nuclear Station Technical Specification, October 2012.

23. RC-12-0058, Virgil C. Summer Nuclear Station (VCSNS) Unit 1, April 2012.

24. RC-1 1-0158, Virgil C. Summer Nuclear Station (VCSNS) Unit 1, October 2011.

25. Letter from P.P. Sena III (FENOC) to Document Control Desk (NRC), dated June 18, 2009, Beaver Valley Power Station, Unit No. 1, Docket No. 50-334, License No.

DPR-66, LER 2009-003-00, "Containment Liner Through Wall Defect Due to Corrosion."

26. Letter from J.E. Pollock (AEP Indiana Michigan Power) to Document Control Desk (NRC), dated March 16, 2001, submitting LER 316/2000-001-01, "Through-Liner Hole Discovered in Containment Liner."

27. South Carolina Electric and Gas Company, RC-95-0180,

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Transmittal of the IPEEE Report Generic Letter 88-20, Supplement 4 (LTR 880020-4).

28. South Carolina Electric and Gas Company, RC-99-0017,

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Request for Additional Information Regarding Generic Letter 88-20 (TAC NO. M83680).

29. Final Safety Evaluation For NEI Topical Report 94-01 Revision 2, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J" and EPRI Report No. 1009325 Revision 2, "Risk Impact Assessment of Extended Integrated Leak Rate Test Intervals".

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Document Control Desk Attachment VII CR-13-00705 RC-13-0037 Page 1 of 1 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

ATTACHMENT VII LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by SCE&G, Virgil C. Summer Nuclear Station in this document. Any other statements in this submittal are provided for information purposes and are not considered to be commitments. Please direct questions regarding these commitments to Mr. Bruce L. Thompson, Manager, Nuclear Licensing, (803) 931-5042.

COMMITMENT Due DatelEvent ILRT Type A test shall be performed no later than October 15, 2018.

October 15, 2018 Update Station Procedures with NEI 94-01 Rev 3-A April 4, 2014 Update Station Procedures with ANSI/ANS-56.8-2002 April 4, 2014