ML22307A313
| ML22307A313 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 11/03/2022 |
| From: | Sattler J Jensen Hughes |
| To: | Ameren Missouri, Office of Nuclear Reactor Regulation, Union Electric Co |
| Shared Package | |
| ML22307A310 | List: |
| References | |
| ULNRC-06765 54032-CALC-01, Rev 1 | |
| Download: ML22307A313 (1) | |
Text
Attachment 1 to Enclosure for ULNRC-06765 Page 1 of 45 Attachm ent 1
Evaluation of Risk Significance of Permanent ILRT Extension
(44 pages follow this cover sheet)
HUGHES
JENSEN 54032-CALC-01 Evaluation of Risk Significance of Pe rmanent ILRT Extension
REVISION RECORD
SUMMARY
Revision Revision Summary
0 Initial Issue
1 Minor revision to wording in Section 5.2.9 to add ress client comment
Revision 1 Page 2 of 44 54032-CALC-01 Evaluation of Risk Significance of Pe rmanent ILRT Extension
TABLE OF CONTENTS
1.0 PURPOSE......................................................................................................................... 4 2.0 SCOPE.............................................................................................................................. 4
3.0 REFERENCES
.................................................................................................................. 6 4.0 ASSUMPTIONS AND LIMITATIONS.............................................................................. 10 5.0 METHODOLOGY AND ANALYSIS................................................................................. 11 5.1 Inputs............................................................................................................................ 11 5.1.1 General Resources Available................................................................................ 11 5.1.2 Plant Specific Inputs.............................................................................................. 14 5.1.3 Impact of Extension on Detection of Component Failu res that Lead to Leakage (Small and Large)................................................................................................ 15 5.2 Analysis........................................................................................................................ 16 5.2.1 Step 1 - Quantify the Baseline Risk in Terms of Fre quency per Reactor Year..... 17 5.2.2 Step 2 - Develop Plant-Specific Person-Rem Dose (Po pulation Dose)................ 20 5.2.3 Step 3 - Evaluate Risk Impact of Extending Type A T est Interval from 10 to 15 Years................................................................................................................... 22 5.2.4 Step 4 - Determine the Change in Risk in Terms of I nternal Events LERF.......... 24 5.2.5 Step 5 - Determine the Impact on the Conditional Co ntainment Failure Probability
............................................................................................................................ 25 5.2.6 Impact of Extension on Detection of Steel Liner Cor rosion that Leads to Leakage
............................................................................................................................ 26 5.2.7 Impact from External Events Contribution............................................................. 29 5.2.7.1 Other External Hazards.............................................................................. 31 5.2.8 Defense-In-Depth Impact...................................................................................... 31 5.2.9 Containment Overpressure................................................................................... 32 5.3 Sensitivities................................................................................................................... 33 5.3.1 Potential Impact from Steel Liner Corrosion Likelih ood........................................ 33 5.3.2 Expert Elicitation Sensitivity.................................................................................. 35 6.0 RESULTS........................................................................................................................ 37
7.0 CONCLUSION
S AND RECOMMENDATIONS............................................................... 38 A. PRA Acceptability............................................................................................................ 40
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1.0 PURPOSE The purpose of this analysis is to provide a risk a ssessment of permanently extending the currently allowed containment Type A Integrated Lea k Rate Test (ILRT) from ten years to fifteen years. The extension would allow for substantial co st savings as the ILRT could be deferred for additional scheduled refueling outages for the Call away Plant. The risk assessment follows the guidelines from NEI 94-01, Revision 3-A and 2-A [Re ferences 1 and 36], the NEI Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillan ce Intervals from November 2001
[Reference 3], the NRC regulatory guidance on the u se of Probabilistic Risk Assessment (PRA) as stated in Regulatory Guide 1.200 Revision 3 as a pplied to ILRT interval extensions, risk insights in support of a request for a plants lice nsing basis as outlined in Regulatory Guide (RG) 1.174 [Reference 4], the methodology used for Calve rt 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 us ed in EPRI 1018243, Revision 2-A of EPRI 1009325 [Reference 24].
2.0 SCOPE Revisions to 10 CFR 50, Appendix J (Option B) allow individual plants to extend the Integrated Leak Rate Test (ILRT) Type A surveillance testing f requency requirement from three in ten years to at least once in ten years. The revised Ty pe A frequency is based on an acceptable performance history defined as two consecutive peri odic Type A tests at least 24 months apart in which the calculated performance leakage rate wa s less than the limiting containment leakage rate of 1L a.
The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, and established in 1995 during development of th e performance-based Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493, Performance-Based Containment Leak Test Program, September 1995 [Reference 6], provid es the technical basis to support rulemaking to revise leakage rate testing requireme nts contained in Option B to Appendix J. The basis consisted of qualitative and quantitative ass essments of the risk impact (in terms of increased public dose) associated with a range of e xtended leakage rate test intervals. To supplement the NRCs rulemaking basis, NEI undertoo k a similar study. The results of that study are documented in Electric Power Research Ins titute (EPRI) Research Project TR-104285, Risk Impact Assessment of Revised Containm ent Leak Rate Testing Intervals
[Reference 2].
The NRC report on performance-based leak testing, N UREG-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), containment isolation failures contribute less than 0.1% 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 contain ment isolation failures for Callaway.
NEI 94-01 Revision 3-A supports using EPRI Report N o. 1009325 Revision 2-A (EPRI 1018243), Risk Impact Assessment of Extended Integ rated Leak Rate Testing Intervals, for performing risk impact assessments in support of IL RT extensions [Reference 24]. The Guidance provided in Appendix H of EPRI Report No. 1009325 Revision 2-A builds on the EPRI Risk Assessment methodology, EPRI TR-104285 [Refere nce 2]. This methodology is followed to determine the appropriate risk information for u se in evaluating the impact of the proposed ILRT changes.
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It should be noted that containment leak-tight inte grity is also verified through periodic in-service inspections conducted in accordance with the requir ements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Ves sel 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 at tachments, and of metallic shell and penetration liners of Class CC pressure-retaining c omponents and their integral attachments in light-water cooled plants. Furthermore, NRC regulat ions 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct visual inspections of the acce ssible areas of the interior of the containment. The associated change to NEI 94-01 req uires that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being conducted. These requirements are not changed as a result of the extended ILRT interval. In addition, Appendix J, Type B local leak tests perfo rmed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, a nd gaskets are also not affected by the change to the Type A test frequency.
The acceptance guidelines in RG 1.174 are used to a ssess the acceptability of this permanent extension of the Type A test interval beyond that e stablished during the Option B rulemaking of Appendix J. RG 1.174 defines very small changes i n 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. Since the Type A test does not impact CDF, the relevant criterion 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-de pth 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. T herefore, the increase in the Conditional Containment Failure Probability (CCFP), which helps ensure the defense-in-depth philosophy is maintained, is also calculated.
Regarding CCFP, changes of up to 1.1% have been acc epted by the NRC for the one-time requests for extension of ILRT intervals. In contex t, it is noted that a CCFP of 1/10 (10%) has been approved for application to evolutionary light water designs. Given these perspectives, a change in the CCFP of up to 1.5% is assumed to be s mall [Reference 1].
In addition, the total annual risk (person-rem/yr p opulation dose) is examined to demonstrate the relative change in this parameter. While no acc eptance guidelines for these additional figures of merit are published, examinations of NUR EG-1493 [Reference 6] and Safety Evaluations (SEs) for one-time interval extension ( summarized in Appendix G of Reference 24) indicate a range of incremental increases in popula tion dose that have been accepted by the NRC. The range of incremental population dose incre ases is from 0.01 to 0.2 person-rem/yr and/or 0.002% to 0.46% of the total accident dose [ Reference 24]. The total doses for the spectrum of all accidents (NUREG-1493 [Reference 6], Figure 7-2) result in health effects that are at least two orders of magnitude less than the NRC Safety Goal Risk. Given these perspectives, a small population dose is defined as an increase from the baseline interval (3 tests per 10 years) dose of 1.0 person-rem per yea r or 1% of the total baseline dose, whichever is less restrictive for the risk impact a ssessment of the proposed extended ILRT interval [Reference 1].
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3.0 REFERENCES
The following references were used in this calculat ion:
- 1. Industry Guideline for Implementing Performance-Bas ed Option of 10 CFR Part 50, Appendix J, Revision 3-A, NEI 94-01, July 2012.
- 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 Assessm ents in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Revision 4, developed for NEI by EPRI and Data Systems and S olutions, November 2001.
- 4. An Approach for Using Probabilistic Risk Assessm ent in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 3, January 2018.
- 5. Response to Request for Additional Information Conc erning 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 D ocument 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, October 1990.
- 8. Letter from R. J. Barrett (Entergy) to U. S. Nuc lear Regulatory Commission, IPN-01-007, January 18, 2001.
- 9. United States Nuclear Regulatory Commission, Ind ian Point Nuclear Generating Unit No. 3
- Issuance of Amendment Re: Frequency of Performanc e-Based Leakage Rate Testing (TAC No. MB0178), April 17, 2001.
- 10. Impact of Containment Building Leakage on LWR Accid ent Risk, Oak Ridge National Laboratory, NUREG/CR-3539, ORNL/TM-8964, April 1984.
- 11. Reliability Analysis of Containment Isolation Syste ms, Pacific Northwest Laboratory, NUREG/CR-4220, PNL-5432, June 1985.
- 12. Technical Findings and Regulatory Analysis for Generic Safety Issue II.E.4.3 Containment Integrity Check, NUREG-1273, April 1988.
- 13. Review of Light Water Reactor Regulatory Requiremen ts, Pacific Northwest Laboratory, NUREG/CR-4330, PNL-5809, Volume 2, June 1986.
- 14. Shutdown Risk Impact Assessment for Extended Co ntainment Leakage Testing Intervals Utilizing ORAM', EPRI, Palo Alto, CA, TR-105189, Fi nal Report, May 1995.
- 15. Severe Accident Risks: An Assessment for Five U. S. Nuclear Power Plants, NUREG-1150, December 1990.
- 16. United States Nuclear Regulatory Commission, Re actor Safety Study, WASH-1400, October 1975.
- 17. PRA-IE-QUANT, At-Power Internal Events PRA, Qua ntification Analysis Notebook, Revision 3.
Revision 1 Page 6 of 44 54032-CALC-01 Evaluation of Risk Significance of Pe rmanent ILRT Extension
- 19. Callaway Plant Unit 1, Environmental Report for License Renewal, Appendix F - Severe Accident Mitigation Alternatives, 2010.
- 20. Anthony R. Pietrangelo, One-time extensions of containment integrated leak rate test interval - additional information, NEI letter to Ad ministrative Points of Contact, November 30, 2001.
- 21. 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.
- 22. Risk Assessment for Joseph M. Farley Nuclear Plant Regarding ILRT (Type A) Extension Request, prepared for Southern Nuclear Operating Co. by ER IN Engineering and Research, P0293010002-1929-030602, March 2002.
- 23. Letter from D. E. Young (Florida Power, Crystal River) to U. S. Nuclear Regulatory Commission, 3F0401-11, dated April 25, 2001.
- 24. Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325, EPRI, Palo Alto, CA, 1018243, October 2 008.
- 25. Risk Assessment for Vogtle Electric Generating Plan t Regarding the ILRT (Type A)
Extension Request, prepared for Southern Nuclear Operating Co. by ER IN Engineering and Research, February 2003.
- 26. AMN#PES00042-REPT-002, "Callaway Energy Center Fire Probabilistic Risk Assessment Peer Review F&Os Closure Review," February 2021.
- 27. Enclosure 4, License Amendment Request: Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Pr ovide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Informat ion Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models.
- 28. PRA-SEISMIC-QUANT, Seismic Probabilistic Risk A ssessment, Quantification Analysis Notebook, Revision 2.
- 29. AMN#PES00031-REPT-003, "Callaway Energy Center Probabilistic Risk Assessment Focused Scope Peer Review," November 2020.
- 30. CALILRT.14-R141231B, Integrated Leakage Rate T est Report, October 12-14, 2014.
- 31. Surveillance Task Sheet S487844, Containment I ntegrated Leak Rate Test, September 28, 1999.
- 32. PRA-OEH-ANALYSIS, Other External Hazards: Scre ening Assessment Notebook, Revision 0.
- 33. NEI Letter to NRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," (ADAMS Accession No. ML17086A431), dated February 21, 2017.
- 34. Technical Letter Report ML112070867, Containmen t Liner Corrosion Operating Experience Summary, Revision 1, August 2011.
- 35. Nuclear Regulatory Commission (NRC) Letter to M r. Greg Krueger (NEI), "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Obse rvations (F&Os)," May 3, 2017, Accession Number ML17079A427.
- 36. Industry Guideline for Implementing Performance-Bas ed Option of 10 CFR Part 50, Appendix J, Revision 2-A, NEI 94-01, November 2008.
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- 37. Regulatory Guide 1.200, An Approach for Determ ining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Info rmed Activities, Revision 2, March 2009.
- 38. USNRC Memorandum, "US Nuclear Regulatory Commis sion Staff Expectations for an Industry Facts and Observations Independent Assessm ent Process," May 1st, 2017 (ADAMS Access ML17121A271).
- 39. USNRC Memorandum, United States Nuclear Regula tory Commission Audit Report on Observation of Industry Independent Assessment Team Close-Out of Facts and Observations (F&Os), May 1st, 2017 (ADAMS Access M L17095A252).
- 40. ULNRC-06690, "Enclosure 5: LAR Supplement to Ad dress Audit Discussion Points Summarized in NRC Letter Dated September 14, 2021 ( ML21238A138)."
- 41. PRA-FLOOD-QUANT, At-Power Internal Flooding PRA, Modeling and Quantification Analysis Notebook, Revision 2.
- 42. PRA-HW-QUANT, Quantification and Results of Pla nt Response Model, Revision 2.
- 43. RG 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, December 2020.
- 44. PWROG-19020-NP Revision 1, "Newly Developed Met hod Peer Review Pilot - General Screening Criteria for Loss of Room Cooling in PRA Modeling Risk Management Committee," PA-RMSC-1647, Revision 1, April 2020.
- 45. Nuclear Energy Institute (NEI) Topical Report ( TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technica l Specifications (RMTS) Guidelines,"
Revision 0-A, October 12, 2012 (ADAMS Accession No. ML12286A322).
- 46. Letter from Jennifer M. Golder (NRC) to Biff Br adley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technica l Specifications (RMTS) Guidelines,"
May 17, 2007 (ML071200238).
- 47. NEI 17-07, Revision 2, "Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard," July 2019 (ADAMS Accession No. ML19228A2 42).
- 48. ASME/ANS RA-S-2009, Addenda to ASME/ANS RA-S-20 08, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessme nt for Nuclear Power Plant Applications," February 2009.
- 49. PWROG-18027-NP Revision 0, "Loss of Room Coolin g in PRA Modeling," April 2020.
- 50. PWROG-19012-P, "Peer Review of the Callaway Int ernal Events and Internal Flood Probabilistic Risk Assessment Model," April 2019.
- 51. PWROG-19034-P, "Independent Assessment of Facts & Observations Closure and Focused Scope Peer Review of the Callaway Probabili stic Risk Assessments," November 2019.
- 52. AMN#PES00031-REPT-001, "Callaway Energy Center Probabilistic Risk Assessment Focused Scope Peer Review," July 2020.
- 53. AMN#PES00031-REPT-002, "Callaway Energy Center Probabilistic Risk Assessment Peer Review F&Os Closure," July 2020.
- 54. PWROG-19022-P, "Peer Review of the Callaway Ext ernal Hazard Screening Assessment and High Winds Probabilistic Risk Assessment," Apri l 2019.
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- 55. ASME/ANS RA-S CASE 1, Case for ASME/ANS RA-Sb-2 013, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessme nt for Nuclear Power Plant Applications," ASME and ANS, November 2017.
- 56. PWROG-18044-P, "Peer Review of the Callaway Sei smic Probabilistic Risk Assessment,"
June 2018.
- 57. NRC Letter, U.S. Nuclear Regulatory Commission Acceptance of ASME/ANS RA-S Case 1, March 12, 2018 (ADAMS access ML18017A964 and ML1801 7A966).
- 58. PWROG-19011-P, "Independent Assessment of Facts & Observations Closure and Focused Scope Peer Review of the Callaway Seismic P robabilistic Risk Assessment,"
March 2019.
- 59. NUREG/CR-6850 (also EPRI 1011989), "Fire PRA Me thodology for Nuclear Power Facilities," September 2005, with Supplement 1 (EPR I 1019259), September 2010.
- 60. LTR-RAM-II-10-019, "Fire PRA Peer Review Agains t the Fire PRA Standard SRs From Section 4 of the ASME/ANS Standard for Level 1/Larg e Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Pl ant Applications for The Callaway Nuclear Plant Fire PRA," October 2009.
- 61. AMN#PES00021-REPT-001, "Callaway Energy Center Fire Probabilistic Risk Assessment Peer Review F&Os Closure," June 2019.
- 62. Final Safety Analysis Report, Section 6.3.2.2: Net Positive Suction Head, Revision OL-25.
- 63. Calculation EJ-29, Residual Heat Removal Pump NPSH Margin During Recirculation, Revision 2.
- 64. NUREG/CR-6762, GSI-191 Technical Assessment: Pa rametric Evaluations for Pressurized Water Reactor Recirculation Sump Performance, US NR C, August 2002.
- 65. Letter from T. Witt (Ameren Missouri) to NRC, R esponse to Request for Additional Information Regarding Request for License Amendment and Regulatory Exemptions for risk-Informed Approach to Address GSI-191 and Respo nd to Generic Letter 2004-02 (LDCN 19-0014) (EPID L-2021-LLA-0059 and EPID L-2021-LLE- 0021). (ULNRC-06735)
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4.0 ASSUMPTIONS AND LIMITATIONS The following assumptions were used in the calculat ion:
The acceptability (i.e., technical adequacy) of the Callaway PRA [Reference 17] is consistent with the requirements of Regulatory Guid e 1.200, as detailed in Appendix A.
The Callaway Level 1 and 2 internal events PRA mode ls [Reference 17] provide representative results.
It is appropriate to use the Callaway internal eve nts PRA model to effectively describe the risk change attributable to the ILRT extension. An analysis is performed in Section 5.2.7 to show the effect of including external even t models for the ILRT extension
[References 17, 18, 28, 40, and 41].
Accident classes describing radionuclide release en d states are defined consistent with EPRI methodology [Reference 24].
The representative containment leakage for Class 1 sequences is 1L a. Class 3 accounts for increased leakage due to Type A inspection fail ures [Reference 24].
The representative containment leakage for Class 3a sequences is 10L a based on the previously approved methodology performed for India n Point Unit 3 [Reference 8, Reference 9].
The representative containment leakage for Class 3b sequences is 100L a based on the guidance provided in EPRI Report No. 1009325, Revis ion 2-A (EPRI 1018243)
[Reference 24].
The Class 3b can be very conservatively categorized as LERF based on the previously approved methodology [Reference 8, Reference 9].
The impact on population doses from containment byp ass 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 containmen t bypass contribution to population dose is fixed, no changes in the conclusions from t his 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 isolat ion signal [Reference 24].
While precise numbers are maintained throughout the calculations, some values have been rounded when presented in this report. Therefo re, summing individual values within tables may yield a different result than the sum result shown in the tables.
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5.0 METHODOLOGY AND ANALYSIS 5.1 Inputs This section summarizes the general resources avail able as input (Section 5.1.1) and the plant specific resources required (Section 5.1.2).
5.1.1 General Resources Available Various industry studies on containment leakage ris k 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-104285 [Reference 2]
- 8. NUREG-1150 [Reference 15] and NUREG/CR-4551 [Ref erence 7]
- 9. NEI Interim Guidance [Reference 3, Reference 20]
- 10. Calvert Cliffs liner corrosion analysis [Refere nce 5]
This 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 leak age that is considered significant and is to be included in the model. The second study is applicab le because it provides a basis of the probability for significant pre-existing containmen t leakage at the time of a core damage accident. The third study is applicable because it is a subsequent study to NUREG/CR-4220 that undertook a more extensive evaluation of the s ame database. The fourth study provides an assessment of the impact of different containment l eakage rates on plant risk. The fifth study provides an assessment of the impact on shutdown ri sk from ILRT test interval extension. The sixth study is the NRCs cost-benefit analysis of v arious alternative approaches regarding extending the test intervals and increasing the all owable leakage rates for containment integrated and local leak rate tests. The seventh s tudy is an EPRI study of the impact of extending ILRT and local leak rate test (LLRT) inte rvals 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 basis for the consequence analysis of t he ILRT interval extension for Callaway. The ninth study includes the NEI recommended methodolog y (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 deg radation 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 15-year extension of the ILRT interval.
NUREG/CR-3539 [Reference 10]
Oak Ridge National Laboratory (ORNL) documented a s tudy of the impact of containment leak rates on public risk in NUREG/CR-3539. This study u ses information from WASH-1400
[Reference 16] as the basis for its risk sensitivit y calculations. ORNL concluded that the impact of leakage rates on LWR accident risks is relativel y small.
NUREG/CR-4220 [Reference 11]
NUREG/CR-4220 is a study performed by Pacific North west Laboratories for the NRC in 1985.
The study reviewed over two thousand license event reports (LERs), ILRT reports and other
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related records to calculate the unavailability of containment due to leakage.
NUREG-1273 [Reference 12]
A subsequent NRC study, NUREG-1273, performed a mor e 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 co rrected. In addition, this study noted that local leak rate tests can detect essentially all p otential degradations of the containment isolation system.
NUREG/CR-4330 [Reference 13]
NUREG/CR-4330 is a study that examined the risk imp acts 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 extensi on, as NUREG/CR-4330 focuses on leakage rate and the ILRT test interval extension study foc uses on the frequency of testing intervals.
However, the general conclusions of NUREG/CR-4330 a re consistent with NUREG/CR-3539 and other similar containment leakage risk studies:
the effect of containment leakage on overall acci dent risk is small since risk is dominated by accident sequences that result in failure or bypass of containment.
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 extendi ng ILRT and LLRT test intervals on shutdown risk. The conclusion from the study is tha t a small, but measurable, safety benefit is realized from extending the test intervals.
NUREG-1493 [Reference 6]
NUREG-1493 is the NRCs cost-benefit analysis for p roposed alternatives to reduce containment leakage testing intervals and/or relax allowable leakage rates. The NRC conclusions are consistent with other similar conta inment leakage risk studies:
Reduction in ILRT frequency from 3 per 10 years to 1 per 20 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 i nterval between integrated leak rate tests is possible with minimal impact on public risk.
EPRI TR-104285 [Reference 2]
Extending the risk assessment impact beyond shutdow n (the earlier EPRI TR-105189 study),
the EPRI TR-104285 study is a quantitative evaluati on of the impact of extending ILRT and LLRT test intervals on at-power public risk. This s tudy combined IPE Level 2 models with NUREG-1150 Level 3 population dose models to perfor m 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 containmen t response to a core damage accident:
- 1. Containment intact and isolated
- 2. Containment isolation failures dependent upon th e core damage accident
- 3. Type A (ILRT) related containment isolation fail ures
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- 4. Type B (LLRT) related containment isolation fail ures
- 5. Type C (LLRT) related containment isolation fail ures
- 6. Other penetration related containment isolation failures
- 7. Containment failures due to core damage accident phenomena
- 8. Containment bypass Consistent with the other containment leakage risk assessment studies, this study concluded:
the proposed CLRT (Containment Leak Rate Tests) f requency changes would have a minimal safety impact. The change in risk determine d by the analyses is small in both absolute and relative terms. For example, for the PWR analyz ed, the change is about 0.02 person-rem per year NUREG-1150 [Reference 15] and NUREG/CR-4551 [Refere nce 7]
NUREG-1150 and the technical basis, NUREG/CR-4551 [ Reference 7], provide an ex-plant consequence analysis for a spectrum of accidents in cluding a severe accident with the containment remaining intact (i.e., Tech Spec Leaka ge). 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.
NEI Interim Guidance for Performing Risk Impact Ass essments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals [Reference 3, Reference 20]
The guidance provided in this document builds on th e EPRI risk impact assessment methodology [Reference 2] and the NRC performance-b ased containment leakage test program
[Reference 6], and considers approaches utilized in various submittals, including Indian Point 3 (and associated NRC SER) and Crystal River.
Calvert Cliffs Response to Request for Additional I nformation Concerning the License Amendment for a One-Time Integrated Leakage Rate Te st Extension [Reference 5]
This submittal to the NRC describes a method for de termining the change in likelihood, due to extending the ILRT, of detecting liner corrosion, a nd the corresponding change in risk. The methodology was developed for Calvert Cliffs in res ponse to a request for additional information regarding how the potential leakage due to age-rela ted degradation mechanisms was factored into the risk assessment for the ILRT one-time exte nsion. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a co ncrete basemat, each with a steel liner.
EPRI Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals [Reference 24]
This report provides a generally applicable assessm ent of the risk involved in extension of ILRT test intervals to permanent 15-year intervals. Appe ndix H of this document provides guidance for performing plant-specific supplemental risk imp act 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 associate d NRC SER) and Crystal River.
The approach included in this guidance document is used in the Callaway assessment to determine the estimated increase in risk associated with the ILRT extension. This document includes the bases for the values assigned in deter mining the probability of leakage for the EPRI Class 3a and 3b scenarios in this analysis, as desc ribed in Section 5.2.
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5.1.2 Plant Specific Inputs The plant-specific information used to perform the Callaway ILRT Extension Risk Assessment includes the following:
CDF and LERF Model results [References 17, 18, 28, 41, and 42]
Dose within a 50-mile radius [Reference 19]
ILRT results to demonstrate adequacy of the adminis trative and hardware issues
[Reference 30 and 31]
Callaway Model The Internal Events (IE) and Internal Flood (IF) PR A Models that are used for Callaway are characteristic of the as-built plant. The current C DF and LERF model is a linked fault tree model
[Reference 17]. The IE+IF CDF is 7.08E-6/yr; the LE RF is 3.20E-8/yr [References 17 and 41].
Table 5-1 and Table 5-2 provide a summary of the IE +IF CDF and LERF results for the Callaway PRA Model. Note: for the rest of this repo rt, internal events risk includes internal floods.
Refer to Section 5.2.7 for further details on exter nal events as they pertain to this analysis.
Table 5 Internal Events CDF Internal Events Frequency (per year)
Internal Floods 5.11E-6 Transients 8.81E-7 FLB/SLB 7.45E-7 LOCAs 1.33E-7 LOOP 1.80E-7 SGTR 1.93E-8 ISLOCA 2.19E-9
Total Internal Events CDF 7.08E-6
Table 5 Internal Events LERF Internal Events Frequency (per year)
Internal Floods 8.09E-9 Transients 1.40E-9 FLB/SLB 9.14E-10 LOCAs 1.54E-10 LOOP 2.24E-10 SGTR* 1.90E-8 ISLOCA 2.21E-9 Total Internal Events LERF 3.20E-8
- Note: induced SGTR (thermal or pressure induced) f requency is removed from the original initiator frequency and c ollectively reported as SGTR in this table (with nominal SGTR f requency).
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Release Category Definitions Table 5-3 defines the accident classes used in the ILRT extension evaluation, which is consistent with the EPRI methodology [Reference 24]. These containment failure classifications are used in this analysis to determine the risk imp act of extending the Containment Type A test interval, as described in Section 5.2 of this repor t.
Table 5 EPRI Containment Failure Classification [Reference 24]
Class Description Containment remains intact including accident seque nces that do not lead to containment failure in the 1 long term. The release of fission products (and att endant consequences) is determined by the maximum allowable leakage rate values L a, under Appendix J for that plant.
2 Containment isolation failures (as reported in the Individual Plant Examinations) including those acci dents in which there is a failure to isolate the containm ent.
3 Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal (i.e., provide a leak-tight contain ment) is not dependent on the sequence in progress.
Independent (or random) isolation failures include those accidents in which the pre-existing isolation 4 failure to seal is not dependent on the sequence in progress. This class is similar to Class 3 isolati on failures, but is applicable to sequences involving Type B tests and their potential failures. These ar e the Type B-tested components that have isolated, but ex hibit excessive leakage.
Independent (or random) isolation failures includin g those accidents in which the pre-existing isolati on 5 failure to seal is not dependent on the sequence in progress. This class is similar to Class 4 isolati on failures, but is applicable to sequences involving Type C test and their potential failures.
6 Containment isolation failures including those leak paths covered in the plant test and maintenance requirements or verified per in-service inspection and testing (ISI/IST) program.
7 Accidents involving containment failure induced by severe accident phenomena. Changes in Appendix J testing requirements do not impact these accidents.
8 Accidents in which the containment is bypassed (eit her as an initial condition or induced by phenomena ) are included in Class 8. Changes in Appendix J test ing requirements do not impact these accidents.
5.1.3 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 bellows arrangements, and failure of some sealing s urfaces, 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 addressed, the EPRI Class 3 accident class, as defined in Table 5-3, is divided into two sub-classes, Class 3a and Class 3b, representing small and large leakage failures respe ctively.
The probability of the EPRI Class 3a and Class 3b f ailures is determined consistent with the EPRI Guidance [Reference 24]. For Class 3a, the pro bability 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, the probability is based on the Jeffreys non-informative prior for no large failures in 21 7 tests (i.e., 0.5 / (217+1) = 0.0023).
In a follow-up letter [Reference 20] to their ILRT guidance document [Reference 3], NEI issued additional information concerning the potential tha t the calculated delta LERF values for several plants may fall above the very small change guide lines of the NRC Regulatory Guide 1.174
[Reference 4]. This additional NEI information incl udes a discussion of conservatisms in the quantitative guidance for LERF. NEI describes ways to demonstrate that, using plant-specific calculations, the LERF is smaller than that calcul ated by the simplified method.
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The supplemental information states:
The methodology employed for determining LERF (Clas s 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, s ome plant-specific accident classes leading to core damage are likely to include individual seq uences 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 (L ERF). These contributors can be removed from Class 3b in the evaluation of LERF by multiply ing the Class 3b probability by only that portion of CDF that may be impacted by Type A leaka ge.
The application of this additional guidance to the analysis for Callaway, as detailed in Section 5.2, involves subtracting LERF risk from the CDF th at is applied to Class 3b because this portion of LERF is unaffected by containment integr ity. To be consistent, the same change is made to the Class 3a CDF, even though these events are not considered LERF.
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 unde tected with a three-year test interval is 1.5 years (3 years / 2), and the average time that a le ak could exist without detection for a ten-year interval is 5 years (10 years / 2). This change wou ld 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.
Correspondingly, an extension of the ILRT interval to 15 years can be estimated to lead to a factor of 5 ((15/2)/1.5) increase in the non-detect ion probability of a leak.
It should be noted that using the methodology discu ssed above is very conservative compared to previous submittals (e.g., the Indian Point 3 re quest 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 and C local leak rate tests that will still occur).
Eliminating this possibility conservatively over-es timates the factor increases attributable to the ILRT extension.
5.2 Analysis The application of the approach based on the guidan ce contained in EPRI 1009325 [Reference 24] and previous risk assessment submittals on this subject [References 5, 8, 21, 22, and 23]
have led to the following results. The results are displayed according to the eight accident classes defined in the EPRI report, as described in Table 5-4.
The analysis performed examined Callaway-specific a ccident sequences in which the containment remains intact or the containment is im paired. Specifically, the breakdown of the severe accidents, contributing to risk, was conside red in the following manner:
Core damage sequences in which the containment rema ins intact initially and in the long term (EPRI 1009325, Class 1 sequences [Reference 24 ]).
Core damage sequences in which containment integrit y is impaired due to random isolation failures of plant components other than t hose associated with Type B or Type C test components. For example, liner breach or bello w leakage (EPRI 1009325, Class 3 sequences [Reference 24]).
Accident sequences involving containment bypassed ( EPRI 1009325, Class 8 sequences [Reference 24]), large containment isolat ion failures (EPRI 1009325, Class 2 sequences [Reference 24]), and small containment is olation failure-to-seal events (EPRI 1009325, Class 4 and 5 sequences [Reference 2 4]) are accounted for in this
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evaluation as part of the baseline risk profile. Ho wever, 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 n ot impact these sequences.
Table 5 EPRI Accident Class Definitions Accident Classes (Containment Release Type) Descrip tion 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 Failure s) 7 Failures Induced by Phenomena (Early and Late) 8 Bypass (SGTR and Interfacing System LOCA)
CDF All CET End States (Including Very Low and No R elease)
The steps taken to perform this risk assessment eva luation are as follows:
Step 1 - Quantify the baseline risk in terms of fre quency per reactor year for each of the accident classes presented in Table 5-4.
Step 2 - Develop plant-specific person-rem dose (po pulation dose) per reactor year for each of the eight accident classes.
Step 3 - Evaluate risk impact of extending Type A t est interval from 3 in 10 years to 1 in 15 years and 1 in 10 years to 1 in 15 years.
Step 4 - Determine the change in risk in terms of L arge Early Release Frequency (LERF) in accordance with RG 1.174 [Reference 4].
Step 5 - Determine the impact on the Conditional Co ntainment Failure Probability (CCFP).
5.2.1 Step 1 - Quantify the Baseline 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 isolati on failures, Type B or Type C testing, or containment failure induced by severe accident phen omena.
For the assessment of ILRT impacts on the risk prof ile, the potential for pre-existing leaks is included in the model (these events are represented by the Class 3 sequences in EPRI 1009325 [Reference 24]). The question on containmen t integrity was modified to include the probability of a liner breach or bellows failure (d ue to excessive leakage) at the time of core damage. Two failure modes were considered for the C lass 3 sequences. These are Class 3a (small breach) and Class 3b (large breach).
The frequencies for the severe accident classes def ined in Table 5-4 were developed for Callaway by first determining the frequencies for C lasses 1, 2, 6, 7, and 8.
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Table 5-5 presents the grouping of each release cat egory in EPRI Classes based on the associated description. Table 5-6 provides a summar y of the accident sequence frequencies that can lead to radionuclide release to the public and have been derived consistent with the NEI Interim Guidance [Reference 3] and the definiti ons of accident classes and guidance provided in EPRI Report No. 1009325, Revision 2-A [ Reference 24]. Adjustments were made to the Class 3b and hence Class 1 frequencies to accou nt for the impact of undetected corrosion of the steel liner per the methodology described in Section 5.2.6.
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 that can only be detected by performing a Type A ILRT. The probab ility of leakage detectable by a Type A ILRT is calculated to determine the impact of exten ding the testing interval. The Class 3 calculation is divided into two classes: Class 3a i s defined as a small liner breach (L a < leakage
< 10L a), and Class 3b is defined as a large liner breach (10L a < leakage < 100L a).
Data reported in EPRI 1009325, Revision 2-A [Refere nce 24] states that two events could have been detected only during the performance of an ILR T and thus impact risk due to change in ILRT frequency. There was a total of 217 successful ILRTs during this data collection period.
Therefore, the probability of leakage is determined for Class 3a as shown in the following equation:
= 2217= 0.0092
Multiplying the CDF by the probability of a Class 3 a leak yields the Class 3a frequency contribution in accordance with guidance provided i n Reference 24. As described in Section 5.1.3, additional consideration is made to not appl y failure probabilities on those cases that are already LERF scenarios. Therefore, these LERF contr ibutions from CDF are removed. The frequency of a Class 3a failure is calculated by th e following equation:
= = 7.08E 3.20E-8 = 6.49E-8
In the database of 217 ILRTs, there are zero contai nment leakage events that could result in a large early release. Therefore, the Jeffreys non-in formative prior is used to estimate a failure rate and is illustrated in the following equations:
Jeffreys Failure Probability = 5678 9: ;<=6 > + 1/25678 9: A >B> + 1
C = 0 + 1/2217 + 1= 0.0023
The frequency of a Class 3b failure is calculated b y the following equation:
C = C =.D E 7.08E 3.20E-8 = 1.62E-8
For this analysis, the associated containment leaka ge for Class 3a is 10L a and for Class 3b is 100L a. These assignments are consistent with the guidanc e provided in Reference 24.
Class 1 Sequences. This group consists of all core damage accident progression bins for which the containment remains intact (modeled as Technica l Specification Leakage). The SAMA
[Reference 19] provides the most recent plant-speci fic risk profile; since the model [Reference 17] does not calculate Intact frequency, the SAMA I ntact frequency is scaled using the CDF, calculated below:
FGHH = IJB;KB LMNM / LMNM FO = 8.08E-6 / 1.66E-5
- 7.08E-6 = 3.45E-6
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The Intact frequency for internal events is 3.45E-6. The EPRI Accident Class 1 frequency is then adjusted by subtracting the EPRI Class 3a and 3b frequency (to preserve total CDF),
calculated below:
= FGHH C = 3.45E 6.49E 1.62E-8 = 3.37E-6 Class 2 Sequences. This group consists of accident progression bins with large containment isolation failures. The large isolation failure is in internal events cutsets that contribute 0.242 of LERF. Multiplying by the total LERF, the EPRI Accid ent Class 2 frequency is 7.72E-9, as shown in Table 5-5.
Class 4 Sequences. This group consists of all core damage accident progression bins for which containment isolation failure-to-seal of Type B tes t components occurs. Because these failures are detected by Type B tests which are unaffected b y the Type A ILRT, this group is not evaluated any further in the analysis, consistent w ith approved methodology.
Class 5 Sequences. This group consists of all core damage accident progression bins for which a containment isolation failure-to-seal of Type C t est components occurs. Because the failures are detected by Type C tests which are unaffected b y the Type A ILRT, this group is not evaluated any further in this analysis, consistent with approved methodology.
Class 6 Sequences. 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 co ntainment isolation valves following a test/maintenance evolution. All other failure modes are bounded by the Class 2 assumptions.
This accident class is also not evaluated further.
Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure is induced by severe accident p henomena (e.g., overpressure). This frequency is calculated by subtracting the Class 1, 2, and 8 frequencies from the total CDF. For this analysis, the frequency is determined from the EPRI Accident Class 7 frequency listed in Table 5-5.
Class 8 Sequences. This group consists of all core damage accident progression bins in which containment is bypassed via SGTR or ISLOCA. The SGT R internal events (including induced SGTR) cutsets contribute 0.593 of LERF. The ISLOCA initiators are in internal events cutsets that contribute 0.069 of LERF. Thus, the total EPRI Accident Class 8 frequency is the summation of the SGTR and ISLOCA frequencies, 2.12E -8, as shown in Table 5-5 and Table 5-6.
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Table 5 Accident Class Frequencies (Core Damage )
EPRI Category Unit 1 Frequency (/yr)
Class 1 3.45E-6 Class 2 7.72E-9 Class 7 3.59E-6 Class 8 (SGTR) 1.90E-8 Class 8 (ISLOCA) 2.21E-9 Total (CDF) 7.08E-6
Table 5 Baseline Risk Profile Class Description Frequency (/yr) 1 No containment failure 3.37E-6 2 2 Large containment isolation failures 7.72E-09 3a Small isolation failures (liner breach) 6.49E-08 3b Large isolation failures (liner breach) 1.62E-08 4 Small isolation failures - failure to seal (type B) 1 5 Small isolation failures - failure to seal (type C) 1 6 Containment isolation failures (dependent failure, personnel errors) 1 7 Severe accident phenomena induced failure (early and late) 3.59E-06 8 Containment bypass 2.12E-08 Total 7.08E-06
- 1. represents a probabilistically insignificant v alue or a Class that is unaffected by the Type A IL RT.
- 2. The Class 3a and 3b frequencies are subtracted f rom Class 1 to preserve total CDF.
5.2.2 Step 2 - Develop Plant-Specific Person-Rem Do se (Population Dose)
Plant-specific release analyses were performed to e stimate the person-rem doses to the population within a 50-mile radius from the plant. The population dose was calculated using data in Table 3-14 in Attachment F (Severe Accident Mitigation Alternatives (SAMA) analysis) of the Environmental Report [Reference 19]; its dose a nd frequency data is presented in Table 5-7. Reference 19 INTACT Release Category correspon ds to EPRI Accident Class 1. LERF-CI (Containment Isolation Failure) Release Category co rresponds to EPRI Accident Class 2. Since they are not associated with other classes, three c ontainment end-states correspond to EPRI Accident Class 7 (LATE-COP, LATE-BMT, and LERF-CF R elease Categories); the EPRI Accident Class 7 dose is calculated via a weighted average using the frequencies provided in Reference 19. The LERF-SG (Steam Generator Tube Rup ture) and LERF-ITR (Induced Steam Generator Tube Rupture) Release Categories and LERF -IS (ISLOCA) Release Category correspond to EPRI Accident Class 8; dose used in t his analysis is weighted via the ISLOCA and SGTR frequencies in this calculation. Class 3a and 3b population dose values are calculated from the Class 1 population dose and are represented as 10L a and 100L a, respectively, as guidance in Reference 1 dictates. Since population dose (person-rem) is not presented directly in SAMA Table F.3-3, population dose rate (person-rem/yr) is divided by frequency (/yr) to calculate population dose.
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Table 5 Baseline Population Doses Release INTACT LATE-LATE-LERF-LERF-LERF-IS LERF-CI LERF-CF Total Category COP BMT SG ITR Frequency (/yr) 8.08E-06 3.19E-06 2.55E-06 2.33E-06 2.17E-07 1.73E-07 1.66E-10 1.13E-08 1.66E-05 Population Dose Rate (person-2.31E-02 1.72E+00 9.92E-02 2.13E+00 2.67E-01 3.46E-01 1.27E-04 9.27E-03 4.60E+00 rem/yr)
Population Dose 3.89E+04 9.13E+05 1.23E+06 2.00E+06 7.66E+05 8.24E+05 -
(person-rem) 2.86E+03 5.41E+05 EPRI Class 1 7 7 8 8 8 2 7
Table 5-8 presents dose exposures calculated from t he methodology described in Reference
- 24. Table 5-9 presents the baseline risk profile fo r Callaway.
The population dose for EPRI Accident Classes 3a an d 3b were calculated based on the guidance provided in EPRI Report No. 1009325, Revis ion 2-A [Reference 24] and the Class 1 dose as follows:
I =;>> 3; 9R6=;B<9J 9> = 10 2.86 +3 = 2.86 +4
I =;>> 38 9R6=;B<9J 9> = 100 2.86 +3 = 2.86 +5
Table 5 Baseline Population Doses Class Description Population Dose (person-rem) 1 No containment failure 2.86E+03 2 Large containment isolation failures 7.66E+05 3a Small isolation failures (liner breach) 2.86E+04 1 3b Large isolation failures (liner breach) 2.86E+05 2 4 Small isolation failures - failure to seal (type B) N/A 5 Small isolation failures - failure to seal (type C) N/A 6 Containment isolation failures (dependent failure, personnel errors) N/A 7 Severe accident phenomena induced failure (early and late) 3.19E+05 8 Containment bypass 1.09E+06
- 1. 10*L a
- 2. 100*L a
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Table 5 Baseline Risk Profile for ILRT Class Description Frequency Contribution Population Population
(/yr) (%) Dose (person-Dose Rate rem) (person-rem/yr) 1 No containment failure 2 3.37E-06 47.67% 2.86E+03 9.65E-03 2 Large containment isolation failures 7.72E-09 0.11% 7.66E+05 5.91E-03
3a Small isolation failures (liner breach) 6.49E-08 0.92% 2.86E+04 1.86E-03
3b Large isolation failures (liner breach) 1.62E-08 0.23% 2.86E+05 4.62E-03
4 Small isolation failures - failure to seal (type B) 1 1 1 1
5 Small isolation failures - failure to seal (type C) 1 1 1 1
Containment isolation failures 6 (dependent failure, personnel 1 1 1 1 errors)
7 Severe accident phenomena induced failure (early and late) 3.59E-06 50.77% 3.19E+05 1.15E+00
8 Containment bypass 2.12E-08 0.30% 1.09E+06 2.31E-02 Total 7.08E-06 100.00% 1.19E+00
- 1. represents a probabilistically insignificant v alue or a Class that is unaffected by the Type A IL RT.
- 2. The Class 1 frequency is reduced by the frequenc y of Class 3a and Class 3b to preserve total CDF.
5.2.3 Step 3 - Evaluate Risk Impact of Extending Ty pe A Test Interval from 10 to 15 Years The next step is to evaluate the risk impact of ext ending the test interval from its current 10-year interval to a 15-year interval. To do this, an eval uation must first be made of the risk associated with the 10-year interval, since the base case appl ies to a 3-year interval (i.e., a simplified representation of a 3-in-10 interval).
Risk Impact Due to 10-Year Test Interval As previously stated, Type A tests impact only Clas s 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 no t detecting the breach increases). Thus, only the frequency of Class 3a and Class 3b sequenc es is impacted. The risk contribution is changed based on the NEI guidance as described in S ection 5.1.3 by a factor of 10/3 compared to the base case values. The Class 3a and 3b freque ncies are calculated as follows:
S TUV = T = T 7.04E-6 = 2.16E-7
SC TUV = T.D E = T.D E 7.04E-6 = 5.38E-8
The results of the calculation for a 10-year interv al are presented in Table 5-10.
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Table 5 Risk Profile for Once in 10 Year ILRT Class Description Frequency Contribution Population Population
(/yr) (%) Dose (person-Dose Rate rem) (person-rem/yr) 1 No containment failure 2 3.18E-06 45.00% 2.86E+03 9.11E-03 2 Large containment isolation failures 7.72E-09 0.11% 7.66E+05 5.91E-03
3a Small isolation failures (liner breach) 2.16E-07 3.06% 2.86E+04 6.19E-03
3b Large isolation failures (liner breach) 5.38E-08 0.76% 2.86E+05 1.54E-02
4 Small isolation failures - failure to seal (type B) 1 1 1 1
5 Small isolation failures - failure to seal (type C) 1 1 1 1
Containment isolation failures 6 (dependent failure, personnel 1 1 1 1 errors)
7 Severe accident phenomena induced failure (early and late) 3.59E-06 50.77% 3.19E+05 1.15E+00
8 Containment bypass 2.12E-08 0.30% 1.09E+06 2.31E-02 Total 7.08E-06 1.21E+00
- 1. represents a probabilistically insignificant v alue or a Class that is unaffected by the Type A IL RT.
- 2. The Class 1 frequency is reduced by the frequenc y of Class 3a and Class 3b to preserve total CDF.
Risk Impact Due to 15-Year Test Interval The risk contribution for a 15-year interval is cal culated in a manner similar to the 10-year interval. The difference is in the increase in prob ability of leakage in Classes 3a and 3b. For this case, the value used in the analysis is a factor of 5 compared to the 3-year interval value, as described in Section 5.1.3. The Class 3a and 3b fre quencies are calculated as follows:
S DUV = D = 5 7.04E-6 = 3.25E-7
SC DUV = D.D E = 5.D E 7.04E-6 = 8.08E-8
The results of the calculation for a 15-year interv al are presented in Table 5-11.
Table 5 Risk Profile for Once in 15 Year ILRT Class Description Frequency Contribution Population Dose Population
(/yr) (%) (person-rem) Dose Rate (person-rem/yr) 1 No containment failure 2 3.05E-06 43.09% 2.86E+03 8.72E-03
2 Large containment isolation failures 7.72E-09 0.11% 7.66E+05 5.91E-03
3a Small isolation failures (liner breach) 3.25E-07 4.59% 2.86E+04 9.28E-03
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Table 5 Risk Profile for Once in 15 Year ILRT Class Description Frequency Contribution Population Dose Population
(/yr) (%) (person-rem) Dose Rate (person-rem/yr)
3b Large isolation failures (liner breach) 8.08E-08 1.14% 2.86E+05 2.31E-02
4 Small isolation failures - failure to seal (type B) 1 1 1 1
5 Small isolation failures - failure to seal (type C) 1 1 1 1 Containment isolation failures 6 (dependent failure, personnel 1 1 1 1 errors)
7 Severe accident phenomena induced failure (early and late) 3.59E-06 50.77% 3.19E+05 1.15E+00
8 Containment bypass 2.12E-08 0.30% 1.09E+06 2.31E-02 Total 7.08E-06 1.22E+00
- 1. represents a probabilistically insignificant v alue or a Class that is unaffected by the Type A IL RT.
- 2. The Class 1 frequency is reduced by the frequenc y of Class 3a and Class 3b to preserve total CDF.
5.2.4 Step 4 - Determine the Change in Risk in Term s of Internal Events LERF The risk increase associated with extending the ILR T interval involves the potential that a core damage event that normally would result in only a s mall radioactive release from an intact containment could, in fact, result in a larger rele ase due to the increase in probability of failure t o detect a pre-existing leak. With strict adherence t o the EPRI guidance, 100% of the Class 3b contribution would be considered LERF.
Regulatory Guide 1.174 [Reference 4] provides guida nce for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 [Reference 4] defines very small changes in risk as resulting in increases of CDF le ss than 10 -6/yr and increases in LERF less than 10 -7/yr, and small changes in LERF as less than 10 -6/yr. Since Callaway does not rely on containment overpressure for net positive suction h ead (NPSH) for ECCS injection [References 62 and 63], the ILRT extension does not impact CDF (see Section 5.2.9 for details). Therefore, the relevant risk-impact metric is LERF.
For Callaway, 100% of the frequency of Class 3b seq uences can be used as a very conservative first-order estimate to approximate th e potential increase in LERF from the ILRT interval extension (consistent with the EPRI guidan ce methodology). Based on a 10-year test interval from Table 5-10, the Class 3b frequency is 5.38E-8/yr; based on a 15-year test interval from Table 5-11, the Class 3b frequency is 8.08E-8/ 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 3 to 15 years is 6.46E-8/yr. Similarly, the in crease due to increasing the interval from 10 to 15 years is 2.69E-8/yr. Table 5-12 summarizes these results.
Table 5 Impact on LERF due to Extended Type A Testing Intervals ILRT Inspection Interval 3 Years (baseline) 10 Year s 15 Years Class 3b (Type A LERF) 1.62E-08 5.38E-08 8.08E-08 LERF (3 year baseline) 3.77E-08 6.46E-08 LERF (10 year baseline) 2.69E-08
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As can be seen, even with the conservatisms include d in the evaluation (per the EPRI methodology), the estimated change in LERF meets th e criteria for a very small change when comparing the 15-year results to the current 10-yea r requirement and when comparing the 15-year results to the original 3-year requirement, as it remains less than 1.0E-7/yr in both cases.
NEI 94-01 Revision 2-A [Reference 36] states that a small population dose is defined as an increase of 1.0 person-rem/yr, or 1% of the tot al population dose, whichever is less restrictive for the risk impact assessment of the e xtended ILRT intervals. As shown in Table 5-13, the results of this calculation meet the dose rate criteria.
Table 5 Impact on Dose Rate due to Extended Ty pe A Testing Intervals ILRT Inspection Interval 10 Years 15 Years Dose Rate (3 year baseline) 1.457E-02 2.498E-02 Dose Rate (10 year baseline) 1.041E-02
%Dose Rate (3 year baseline) 1.224% 2.098%
%Dose Rate (10 year baseline) 0.863%
- 1. Dose Rate is the difference in the total dose r ate between cases. For instance, Dose Rate (3 yea r baseline) for the 1 in 15 case is the total dose r ate of the 1 in 15 case minus the total dose rate o f the 3 in 10 year case.
- 2. %Dose Rate is the Dose Rate divided by the tot al baseline dose rate. For instance, %Dose Rate ( 3 year baseline) for the 1 in 15 case is the Dose Rate (3 year baseline) of the 1 in 15 year case di vided by the total dose rate of the 3 in 10 year case.
5.2.5 Step 5 - Determine the Impact on the Conditio nal Containment Failure Probability Another parameter that the NRC guidance in RG 1.174 [Reference 4] states can provide input into the decision-making process is the change in t he conditional containment failure probability (CCFP). The CCFP is defined as the probability of c ontainment failure given the occurrence of an accident. This probability can be expressed usin g the following equation:
= 1 : JK:
where f(ncf) is the frequency of those sequences that do n ot result in containment failure; this frequency is determined by summing the Class 1 and Class 3a results. Table 5-14 shows the steps and results of this calculation.
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Table 5 Impact on CCFP due to Extended Type A Testing Intervals ILRT Inspection Interval 3 Years (baseline) 10 Year s 15 Years f(ncf) (/yr) 3.44E-06 3.40E-06 3.37E-06 f(ncf)/CDF 0.486 0.481 0.477 CCFP 0.514 0.519 0.523 CCFP (3 year baseline) 0.533% 0.913%
CCFP (10 year baseline) 0.381%
As stated in Section 2.0, a change in the CCFP of u p to 1.5% is assumed to be small. The increase in the CCFP from the 3 in 10 year interval to 1 in 15 year interval is 0.913%. Therefore, this increase is judged to be small.
5.2.6 Impact of Extension on Detection of Steel Lin er 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 ex tended test interval is evaluated using a methodology similar to the Calvert Cliffs liner cor rosion analysis [Reference 5]. The Calvert Cliffs analysis was performed for a concrete cylind er and dome and a concrete basemat, each with a steel liner.
The following approach is used to determine the cha nge in likelihood, due to extending the ILRT, of detecting corrosion of the containment ste el 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 th e containment cylinder and dome The historical steel liner flaw likelihood due to concealed corrosion The impact of aging The corrosion leakage dependency on containment pr essure The likelihood that visual inspections will be eff ective at detecting a flaw Assumptions Consistent with the Calvert Cliffs analysis, a hal f failure is assumed for basemat concealed liner corrosion due to the lack of identi fied failures (See Table 5-15, Step 1).
In the 5.5 years following September 1996 when 10 CFR 50.55a started requiring visual inspection, there were three events where a through wall hole in the containment liner was identified. These are Brunswick 2 on 4/27/99, N orth Anna 2 on 9/23/99, and D. C.
Cook 2 in November 1999. The corrosion associated w ith the Brunswick event is believed to have started from the coated side of th e containment liner. Although Callaway has a different containment type, this eve nt could potentially occur at Callaway (i.e., corrosion starting on the coated side of con tainment). Construction material embedded in the concrete may have contributed to th e corrosion. The corrosion at North Anna is believed to have started on the uninspectab le side of containment due to wood imbedded in the concrete during construction. The D.C. Cook event is associated with an inadequate repair of a hole drilled through the liner during construction. Since the hole was created during construction and not caused by corrosion, this event does not apply to this analysis. Based on the above data, th ere are two corrosion events from the 5.5 years that apply to Callaway.
Consistent with the Calvert Cliffs analysis, the e stimated historical flaw probability is also limited to 5.5 years to reflect the years since Sep tember 1996 when 10 CFR 50.55a
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started requiring visual inspection. Additional suc cess data was not used to limit the aging impact of this corrosion issue, even though i nspections were being performed prior to this date (and have been performed since the tim e frame of the Calvert Cliffs analysis)
(See Table 5-4, Step 1).
Consistent with the Calvert Cliffs analysis, the s teel liner flaw likelihood is assumed to double every five years. This is based solely on ju dgment and is included in this analysis to address the increased likelihood of corrosion as the steel liner ages (See Table 5-15, Steps 2 and 3). Sensitivity studies are included th at 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 pr obability) for the basemat. These values were determined from an assessment of the pr obability versus containment pressure. For Callaway, the ILRT maximum pressure i s 48.8 psig [Reference 30].
Probabilities of 1% for the cylinder and dome, and 0.1% for the basemat are used in this analysis, and sensitivity studies are included in S ection 5.3.1 (See Table 5-15, Step 4).
Consistent with the Calvert Cliffs analysis, the l ikelihood of leakage escape (due to crack formation) in the basemat region is considered to b e less likely than the containment cylinder and dome region (See Table 5-15, Step 4).
In the Calvert Cliffs analysis, it is noted that a pproximately 85% of the interior wall surface is accessible for visual inspections. Consi stent with the Calvert Cliffs analysis, a 5% visual inspection detection failure likelihood g iven the flaw is visible and a total detection failure likelihood of 10% is used. To dat e, all liner corrosion events have been detected through visual inspection (See Table 5-15, Step 5).
Consistent with the Calvert Cliffs analysis, all n on-detectable containment failures are assumed to result in early releases. This approach avoids a detailed analysis of containment failure timing and operator recovery ac tions.
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Table 5 Steel Liner Corrosion Base Case Step Description Containment Cylinder and Containment Basemat Dome (85%) (15%)
Historical liner flaw likelihood Events: 2 Events: 0 Failure data: containment location (Brunswick 2 and North Anna 2) Assume a half failure specific 2 / (70 x 5.5) = 5.19E-03 0.5 / (70 x 5.5) = 1.30E-03 1 Success data: based on 70 steel-lined containments and 5.5 years since the 10 CFR 50.55a requirements of periodic visual inspections of containment surfaces Year Failure rate Year Failure rate Aged adjusted liner flaw likelihood 1 2.05E-03 1 5.13E-04 During the 15-year interval, assume average 5-10 5.19E-03 average 5-10 1.30E-03 2 failure rate doubles every five years 15 1.43E-02 15 3.57E-03 (14.9% increase per year). The average for the 5th to 10th year set 15 year average = 6.44E-03 15 year average = 1.61E-03 to the historical failure rate.
Increase in flaw likelihood between 0.71% (1 to 3 years) 0.18% (1 to 3 years) 3 and 15 years Uses aged adjusted 3 liner flaw likelihood (Step 2), 4.14% (1 to 10 years) 1.04% (1 to 10 years) assuming failure rate doubles every 9.66% (1 to 15 years) 2.42% (1 to 15 years) five years.
4 Likelihood of breach in containment given liner flaw 1% 0.1%
10%
5% failure to identify visual flaws plus 5% likelihood that the flaw is not visible (not through-cylinder 100%
5 Visual inspection detection failure likelihood but could be detected by ILRT). Cannot be visually inspected All events have been detected through visual inspection. 5%
visible failure detection is a conservative assumption.
0.00071% (3 years) 0.00018% (3 years) 0.71% x 1% x 10% 0.18% x 0.1% x 100%
Likelihood of non-detected 0.00414% (10 years) 0.00104% (10 years) 6 containment leakage (Steps 3 x 4 x 4.14% x 1% x 10% 1.04% x 0.1% x 100%
5) 0.00966% (15 years) 0.00242% (15 years) 9.66% x 1% x 10% 2.42% x 0.1% x 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 t he containment basemat, as summarized below for Callaway.
Table 5 Total Likelihood on Non-Detected Conta inment Leakage Due to Corrosion for Callaway Description At 3 years: 0.00071% + 0.00018% = 0.00089%
At 10 years: 0.00414% + 0.00104% = 0.00517%
At 15 years: 0.00966% + 0.00242% = 0.01207%
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The above factors are applied to those core damage accidents that are not already independently LERF or that could never result in LE RF.
The two corrosion events that were initiated from t he non-visible (backside) portion of the containment liner used to estimate the liner flaw p robability in the Calvert Cliffs analysis are assumed to be applicable to this containment analys is. These events, one at North Anna Unit 2 (September 1999) caused by timber embedded in the c oncrete immediately behind the containment liner, and one at Brunswick Unit 2 (Apr il 1999) caused by a cloth work glove embedded in the concrete next to the liner, were in itiated from the nonvisible (backside) portion of the containment liner. A search of the NRC websi te LER database identified two additional events have occurred since the Calvert Cliffs analy sis was performed. In January 2000, a 3/16-inch circular through-liner hole was found at Cook Nuclear Plant Unit 2 caused by a wooden brush handle embedded immediately behind the contai nment liner. The other event occurred in April 2009, where a through-liner hole approximatel y 3/8-inch by 1-inch in size was identified in the Beaver Valley Power Station Unit 1 (BVPS-1) con tainment liner caused by pitting originating from the concrete side due to a piece of wood that was left behind during the original construction that came in contact with the steel li ner [Reference 34]. Two other containment liner through-wall hole events occurred at Turkey P oint Units 3 and 4 in October 2010 and November 2006, respectively. However, these events originated from the visible side caused by the failure of the coating system, which was not de signed for periodic immersion service, and are not considered to be applicable to this analysi s. More recently, in October 2013, some through-wall containment liner holes were identifie d at BVPS-1, with a combined total area of approximately 0.395 square inches. The cause of the se through-wall liner holes was attributed to corrosion originating from the outside concrete surface due to the presence of rayon fiber foreign material that was left behind during the or iginal construction and was contacting the steel liner. For risk evaluation purposes, these fi ve total corrosion events occurring in 66 operating plants with steel containment liners over a 17.1 year period from September 1996 to October 4, 2013 (i.e., 5/(66*17.1) = 4.43E-03) are bounded by the estimated historical f law probability based on the two events in the 5.5 year period of the Calvert Cliffs analysis (i.e.,
2/(70*5.5) = 5.19E-03) incorporated in the EPRI guidance [Refere nce 34].
5.2.7 Impact from External Events Contribution An assessment of the impact of external events is p erformed. The primary purpose for this investigation is the determination of the total LER F following an increase in the ILRT testing interval from 3 in 10 years to 1 in 15 years.
The Fire PRA calculated a CDF of 1.25E-5 and a LERF of 4.07E-8 [Reference 18]. As described in Section 5.1.3, additional consideratio n is made to not apply failure probabilities on those cases that are already LERF scenarios. Theref ore, the LERF contributions are removed from CDF; to reduce conservatism in the ILRT extens ion analysis, the methodology of subtracting existing LERF from CDF is also applied to the Fire PRA model. The following shows the calculation for Class 3b:
C = C = 0.5218 1.25 -5 4.07 -8 = 2.85E-8
C TUV = T C = T T.D E 1.25 -5 4.07 -8 = 9.51E-8
C DUV = D C = 5 T.D E 1.25 -5 4.07 -8 = 1.43E-7
The Seismic PRA calculated a CDF of 4.56E-5 and a L ERF of 3.41E-6 using ACUBE
[Reference 28]. As described in Section 5.1.3, addi tional consideration is made to not apply
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failure probabilities on those cases that are alrea dy LERF scenarios. Subtracting seismic LERF from CDF, the Class 3b frequency can be calculated by the following formulas:
C = C = T.D E 4.56 -5 3.41 -6 = 9.68E-8
C TUV = T C = T T.D E 4.56 -5 3.41 -6 = 3.23E-7
C DUV = D C = 5 T.D E 4.56 -5 3.41 -6 = 4.84E-7
The High Winds (HW) PRA calculated a CDF of 2.79E-6 and a LERF of 7.98E-9 using ACUBE
[Reference 42]. As described in Section 5.1.3, addi tional consideration is made to not apply failure probabilities on those cases that are alrea dy LERF scenarios. Subtracting seismic LERF from CDF, the Class 3b frequency can be calculated by the following formulas:
C = C = T.D E 2.79 -6 7.98 -9 = 6.39E-9
C TUV = T C = T T.D E 2.79 -6 7.98 -9 = 2.13E-8
C DUV = D C = 5 T.D E 2.79 -6 7.98 -9 = 3.19E-8
The external event contributions to Class 3b freque ncies are then combined to obtain the total external event contribution to Class 3b frequencies. The change in LERF is calculated for the 1 in 10 year and 1 in 15 year cases and the change de fined for the external events in Table 5-17.
Table 5 Callaway External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from LERF Increase (from 3 per 10 years to 1 1 per 10 years to 1 3 per 10 years 1 per 10 years 1 per 15 years per 15 years) per 15 years)
External Events 1.32E-07 4.39E-07 6.59E-07 5.27E-07 2.20E-07 Internal Events 1.62E-08 5.38E-08 8.08E-08 6.46E-08 2.69E-08 Combined 1.48E-07 4.93E-07 7.39E-07 5.92E-07 2.46E-07
The internal event results are also provided to all ow a composite value to be defined. When both the internal and external event contributions are combined, the increase due to increasing the interval from 10 to 15 years is 2.46E-7; the to tal change in LERF due to increasing the ILRT interval from 3 to 15 years is 5.92E-7, which meets the guidance for small change in risk, as it exceeds 1.0E-7/yr and remains less than a 1.0E-6 ch ange in LERF. For this change in LERF to be acceptable, total LERF must be less than 1.0E-5. The total LERF due to increasing the ILRT interval from 3 to 15 years is calculated below by adding external and internal events LERF and change in LERF:
LERF = LERF IE + LERF fire + LERF seismic + LERF HW + LERF class3Bincrease LERF = 3.20E-8/yr + 4.07E-8/yr + 3.41E-6/yr + 7.98E-9/yr + 5.92E-7/yr = 4.08E-6/yr
The total LERF due to increasing the ILRT interval from 10 to 15 years is calculated below by adding external and internal events LERF and change in LERF:
LERF = LERF IE + LERF fire + LERF seismic + LERF HW + LERF class3Bincrease LERF = 3.20E-8/yr + 4.07E-8/yr + 3.41E-6/yr + 7.98E-9/yr + 2.46E-7/yr = 3.73E-6/yr
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Several conservative assumptions were made in this ILRT analysis, as discussed in Sections 4.0, 5.1.3, 5.2.1, and 5.2.4; therefore, the total change in LERF is considered conservative for this application. As specified in Regulatory Guide 1.174 [Reference 4], since the total LERF is less than 1.0E-5, it is acceptable for the LERF to be between 1.0E-7 and 1.0E-6.
5.2.7.1 Other External Hazards Callaways Other External Hazards (OEHs) Screening Assessment Notebook [Reference 32]
provides an evaluation of OEHs performed in accorda nce with the process specified in Part 6, Section 6-2 of Reference 48. All the external hazar ds that could potentially affect the CEC site were screened out based on the criteria specified i n SRs EXT-B1 and EXT-C1 of the PRA Standard [Reference 32].
Similarly, Reference 27 analyzed and screened all t he hazards listed in Table D-1 of Regulatory Guide 1.200 Revision 3 [Reference 43]. The evaluati on concluded that all other hazards either do not present a design-basis challenge to Callaway, the challenge is adequately addressed in the PRA, or the hazard has a negligible impact [Ref erence 27].
5.2.8 Defense-In-Depth Impact Regulatory Guide 1.174, Revision 3 [Reference 4] de scribes an approach that is acceptable for developing risk-informed applications for a licensi ng basis change that considers engineering issues and applies risk insights. One of the consid erations included in RG 1.174 is Defense in Depth. Defense in Depth is a safety philosophy that employs successive compensatory measures to prevent accidents or mitigate damage if a malfunction, accident, or naturally caused event occurs at a nuclear facility. The foll owing seven considerations as presented in RG 1.174, Revision 3, Section C.2.1.1.2 will serve to evaluate the proposed licensing basis change for overall impact on Defense in Depth.
- 1. Preserve a reasonable balance among the layers o f defense.
The use of the risk metrics of LERF, population dos e, and conditional containment failure probability collectively ensures the balance betwee n prevention of core damage, prevention of containment failure, and consequence mitigation is preserved. The change in LERF is very small with respect to internal events and small when including external events per RG 1.174, and the change in population dose and CCFP are sma ll as defined in this analysis and consistent with NEI 94-01 Revision 3-A.
- 2. Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.
The adequacy of the design feature (the containment boundary subject to Type A testing) is preserved as evidenced by the overall small chang e in risk associated with the Type A test frequency change.
- 3. Preserve system redundancy, independence, and di versity commensurate with the expected frequency and consequences of challenges t o the system, including consideration of uncertainty.
The redundancy, independence, and diversity of the containment subject to the Type A test is preserved, commensurate with the expected frequency and consequences of challenges to the system, as evidenced by the overall small change in risk associated with the Type A test frequency change.
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- 4. Preserve adequate defense against potential CCFs.
Adequate defense against CCFs is preserved. The Typ e A test detects problems in the containment which may or may not be the result of a CCF; such a CCF may affect failure of another portion of containment (i.e., local penetra tions) due to the same phenomena. Adequate defense against CCFs is preserved via the continued performance of the Type B and C tests and the performance of inspections. The change to t he Type A test, which bounds the risk associated with containment failure modes including those involving CCFs, does not degrade adequate defense as evidenced by the overall small change in risk associated with the Type A test frequency change.
- 5. Maintain multiple fission product barriers.
Multiple Fission Product barriers are maintained. T he portion of the containment affected by the Type A test extension is still maintained as an ind ependent fission product barrier, albeit with an overall small change in the reliability of the ba rrier.
- 6. Preserve sufficient defense against human errors.
Sufficient defense against human errors is preserve d. The probability of a human error to operate the plant, or to respond to off-normal cond itions and accidents is not significantly affected by the change to the Type A testing freque ncy. Errors committed during test and maintenance may be reduced by the less frequent per formance of the Type A test (less opportunity for errors to occur).
- 7. Continue to meet the intent of the plants desig n criteria.
The intent of the plants design criteria continues to be met. The extension of the Type A test does not change the configuration of the plant or t he way the plant is operated.
5.2.9 Containment Overpressure The guidance in Reference 24 states that in general, CDF is not significantly impacted by an extension of the ILRT interval. Plants that rely on containment overpressure for net positive suction head (NPSH) for emergency core coolant syst em (ECCS) injection for certain accident sequences may experience an increase in CDF.
In the case of accident sequences that are the resu lt of the long-term loss of containment heat removal, containment pressurization and eventual fa ilure are assumed to result in a loss of core coolant injection systems.
In the case where containment overpressure may be a consideration, plants should examine their ECCS NPSH requirements to determine if contai nment overpressure is required (and assumed to be available) in various accident scenar ios. Examples include the following:
LOCA scenarios where the initial containment press urization helps to satisfy the NPSH requirements for early injection in BWRs or PWR sum p recirculation.
Total loss of containment heat removal scenarios w here gradual containment pressurization helps to satisfy the NPSH requiremen ts for long-term use of an injection system from a source inside of containment (for exa mple, BWR suppression pool).
Either of these scenarios could be impacted by a la rge containment failure that eliminates the overpressure contribution to the available NPSH cal culation. If either of these cases is susceptible to whether or not containment overpress ure is available (or other cases are identified), then the PRA model should be adjusted to account for this requirement. As a first-order estimate of the impact, it can be assumed tha t the EPRI Class 3b contribution would lead
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to loss of containment overpressure, and the system s that require this contribution to NPSH should be made unavailable when such an isolation f ailure exists. The impact on CDF can then be accounted for in a similar fashion to the LERF c ontribution as the EPRI Class 3b contribution changes for various ILRT test intervals. The combin ed impacts on CDF and LERF should then be considered in the ILRT evaluation and compared w ith the Regulatory Guide 1.174 acceptance guidelines.
Per the ECCS analysis of record, Callaway does not rely on containment overpressure for net positive suction head (NPSH) for ECCS injection [Re ferences 62 and 63].
A separate analysis for Generic Safety Issue (GSI) 191 pertains to debris accumulation in containment following a LOCA [Reference 64]. The Ca llaway Small LOCA analysis does not credit recirculation to achieve successful core coo ling, so overpressure credit for NPSH is not a concern [References 40 and 65]. Reference 40 provid es analysis for boiling (flashing) at the top of the sump strainers, deaeration, and NPSH (NPSH i s the relevant analysis for this ILRT application). Although spurious or significantly mi scalibrated sump level indication is unlikely, these calculations assume a 3-inch penalty to the n ormal sump level indicator. If the sump level indicator performs nominally, the pool will be 3 in ches deeper than the flashing calculation assumes, and no overpressure credit is required. Pe r Reference 40, if nominal performance of the normal sump level indicator is assumed (no assu med penalty for a spurious low-level reading), overpressure credit is not required.
Therefore, the impact of credit for containment ove rpressure is not significant to the ILRT extension PRA risk metrics, and the conclusion of t he application is based on the change in LERF, change in population dose, and change in cond itional containment failure probability metrics.
5.3 Sensitivities 5.3.1 Potential Impact from Steel Liner Corrosion L ikelihood A quantitative assessment of the contribution of st eel liner corrosion likelihood impact was performed for the risk impact assessment for extend ed ILRT intervals. As a sensitivity run, the internal event CDF was used to calculate the Class 3b frequency. The impact on the Class 3b frequency due to increases in the ILRT surveillance interval was calculated for steel liner corrosion likelihood using the relationships descri bed in Section 5.2.6. The EPRI Category 3b frequencies for the 3 per 10-year, 10-year, and 15-year ILRT intervals were quantified using the internal events CDF. The change in the LERF, change in CCFP, and change in Annual Dose Rate due to extending the ILRT interval from 3 in 1 0 years to 1 in 10 years, or to 1 in 15 years are provided in Table 5 Table 5-20. The steel liner corrosion likelihood was increased by a factor of 1000, 10000, and 100000. Except for extre me factors of 10000 and 100000, the corrosion likelihood is relatively insensitive to t he results.
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Table 5 Steel Liner Corrosion Sensitivity Case : 3B Contribution 3b 3b 3b LERF LERF LERF Frequency Frequency Frequency Increase Increase Increase (3-per-10 (1-per-10 (1-per-15 (3-per-10 to (3-per-10 to (1-per-10 to year ILRT) year ILRT) year ILRT) 1-per-10) 1-per-15) 1-per-15)
Corrosion Likelihood X 1 1.62E-08 5.38E-08 8.08E-08 3.77E-08 6.46E-08 2.69E-08 Corrosion Likelihood X 1000 1.63E-08 5.66E-08 9.05E-08 4.03E-08 7.42E-08 3.39E-08 Corrosion Likelihood 1.76E-08 8.17E-08 1.78E-07 6.41E-08 1.61E-07 9.66E-08 X 10000 Corrosion Likelihood 3.05E-08 3.32E-07 1.06E-06 3.02E-07 1.03E-06 7.24E-07 X 100000
Table 5 Steel Liner Corrosion Sensitivity: CCF P CCFP CCFP CCFP CCFP CCFP CCFP Increase Increase Increase (3-per-10 (1-per-10 (1-per-15 (3-per-10 to (3-per-10 to (1-per-10 to year ILRT) year ILRT) year ILRT) 1-per-10) 1-per-15) 1-per-15)
Corrosion Likelihood X 1 5.14E-01 5.19E-01 5.23E-01 5.33E-03 9.13E-03 3.81E-03 Corrosion Likelihood X 1000 5.14E-01 5.20E-01 5.23E-01 5.37E-03 9.21E-03 3.84E-03 Corrosion Likelihood 5.14E-01 5.20E-01 5.24E-01 5.80E-03 9.95E-03 4.14E-03 X 10000 Corrosion Likelihood 5.16E-01 5.26E-01 5.33E-01 1.01E-02 1.73E-02 7.19E-03 X 100000
Table 5 Steel Liner Corrosion Sensitivity: Dos e Rate Dose Rate Dose Rate Dose Rate Dose Rate Dose Rate Dose Rate Increase Increase Increase (3-per-10 (1-per-10 (1-per-15 (3-per-10 to (3-per-10 to (1-per-10 to year ILRT) year ILRT) year ILRT) 1-per-10) 1-per-15) 1-per-15)
Corrosion 4.62E-03 1.54E-02 2.31E-02 1.08E-02 1.85E-02 7.70E-03 Likelihood X 1 Corrosion Likelihood X 1000 4.66E-03 1.62E-02 2.59E-02 1.15E-02 2.12E-02 9.69E-03 Corrosion Likelihood X 10000 5.03E-03 2.34E-02 5.10E-02 1.83E-02 4.60E-02 2.76E-02 Corrosion Likelihood X 100000 8.73E-03 9.51E-02 3.02E-01 8.64E-02 2.93E-01 2.07E-01
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5.3.2 Expert Elicitation Sensitivity Another sensitivity case on the impacts of assumpti ons regarding pre-existing containment defect or flaw probabilities of occurrence and magn itude, or size of the flaw, is performed as described in Reference 24. In this sensitivity case, an expert elicitation was conducted to develop probabilities for pre-existing containment defects that would be detected by the ILRT only based on the historical testing data.
Using the expert knowledge, this information was ex trapolated into a probability-versus-magnitude relationship for pre-existing containment defects [Reference 24]. The failure mechanism analysis also used the historical ILRT da ta augmented with expert judgment to develop the results. Details of the expert elicitat ion process and results are contained in Reference 24. The expert elicitation process has th e advantage of considering the available data for small leakage events, which have occurred in the data, and extrapolate those events and probabilities of occurrence to the potential fo r large magnitude leakage events.
The expert elicitation results are used to develop sensitivity cases for the risk impact assessment. Employing the results requires the appl ication of the ILRT interval methodology using the expert elicitation to change the probabil ity of pre-existing leakage in the containment.
The baseline assessment uses the Jeffreys non-infor mative prior and the expert elicitation sensitivity study uses the results of the expert el icitation. In addition, given the relationship between leakage magnitude and probability, larger l eakage that is more representative of large early release frequency, can be reflected. For the purposes of this sensitivity, the same leakage magnitudes that are used in the basic methodology ( i.e., 10 L a for small and 100 L a for large) are used here. Table 5-21 presents the magnitudes a nd probabilities associated with the Jeffreys non-informative prior and the expert elici tation used in the base methodology and this sensitivity case.
Table 5 Callaway Summary of ILRT Extension Usi ng Expert Elicitation Values (from Reference 24)
Leakage Size (L a) Expert Elicitation Mean Probability of Occurrence Percent Reduction 10 3.88E-03 86%
100 2.47E-04 91%
Taking the baseline analysis and using the values p rovided in Table 5-10 and Table 5-11 for the expert elicitation sensitivity yields the results i n Table 5-22.
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Table 5 Callaway Summary of ILRT Extension Usi ng Expert Elicitation Values Accident ILRT Interval Class 3 per 10 Years 1 per 10 Years 1 per 15 Years
Base Adjusted Dose Dose Frequency Dose Frequency Dose Frequency Base (person-Rate Rate Rate Frequency rem) (person- (person- (person-rem/yr) rem/yr) rem/yr) 1 3.45E-06 3.42E-06 2.86E+03 9.80E-03 3.36E-06 9.60E- 03 3.31E-06 9.46E-03 2 7.72E-09 7.72E-09 7.66E+05 5.91E-03 7.72E-09 5.91E- 03 7.72E-09 5.91E-03 3a N/A 2.73E-08 2.86E+04 7.82E-04 9.11E-08 2.61E-03 1. 37E-07 3.91E-03 3b N/A 1.74E-09 2.86E+05 4.98E-04 5.80E-09 1.66E-03 8. 70E-09 2.49E-03 7 3.59E-06 3.59E-06 3.19E+05 1.15E+00 3.59E-06 1.15E+ 00 3.59E-06 1.15E+00 8 2.12E-08 2.12E-08 1.09E+06 2.31E-02 2.12E-08 2.31E- 02 2.12E-08 2.31E-02 Totals 7.08E-06 7.08E-06 2.49E+06 1.19E+00 7.08E-06 1.19E+ 00 7.08E-06 1.19E+00 LERF N/A (3 per 10 4.06E-09 6.96E-09 yrs base)
LERF N/A (1 per 10 N/A 2.90E-09 yrs base)
CCFP 51.21% 51.26% 51.31%
The results illustrate how the expert elicitation r educes the overall change in LERF and the overall results are more favorable with regard to t he change in risk.
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6.0 RESULTS The Internal Events results from this ILRT extensio n risk assessment for Callaway are summarized in Table 6-1.
Table 6 ILRT Extension Summary (Internal Events )
Class Dose Base Case Extend to Extend to (person-3 in 10 Years 1 in 10 Years 1 in 15 Years rem)
CDF/yr Person-CDF/yr Person-CDF/yr Person-Rem/yr Rem/yr Rem/yr 1 2.86E+03 3.37E-06 9.65E-03 3.18E-06 9.11E-03 3.05 E-06 8.72E-03 2 7.66E+05 7.72E-09 5.91E-03 7.72E-09 5.91E-03 7.72 E-09 5.91E-03 3a 2.86E+04 6.49E-08 1.86E-03 2.16E-07 6.19E-03 3.2 5E-07 9.28E-03 3b 2.86E+05 1.62E-08 4.62E-03 5.38E-08 1.54E-02 8.0 8E-08 2.31E-02 7 3.19E+05 3.59E-06 1.15E+00 3.59E-06 1.15E+00 3.59 E-06 1.15E+00 8 1.09E+06 2.12E-08 2.31E-02 2.12E-08 2.31E-02 2.12 E-08 2.31E-02 Total 7.08E-06 1.19E+00 7.08E-06 1.21E+00 7.08E-06 1.22E+00
ILRT Dose Rate from 3a and 3b Total From 3 Dose Rate Years N/A 1.46E-02 2.50E-02 (Person-From 10 Rem/yr) Years N/A N/A 1.04E-02
From 3
%Dose Years N/A 1.22% 2.10%
Rate From 10 Years N/A N/A 0.86%
3b Frequency (LERF/yr)
From 3 LERF Years N/A 3.77E-08 6.46E-08 From 10 Years N/A N/A 2.69E-08
CCFP %
From 3 CCFP% Years N/A 0.533% 0.913%
From 10 Years N/A N/A 0.381%
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7.0 CONCLUSION
S AND RECOMMENDATIONS Based on the results from Section 5.2 and the sensi tivity calculations presented in Section 5.3, the following conclusions regarding the assessment of the plant risk are associated with extending the Type A ILRT test frequency to 15 year s:
Regulatory Guide 1.174 [Reference 4] provides guid ance for determining the risk impact of plant-specific changes to the licensing basis. R egulatory Guide 1.174 defines very small changes in risk as resulting in increases of CDF less than 1.0E-06/year and increases in LERF less than 1.0E-07/year. Regulator y Guide 1.174 defines small changes in risk as resulting in increases of CDF gr eater than 1.0E-6/yr and less than 1.0E-5/yr and increases in LERF greater than 1.0E-7 /yr and less than 1.0E-6/yr. Since the ILRT does not impact CDF, the relevant criterio n is LERF. The increase in LERF resulting from a change in the Type A ILRT test int erval from 3 in 10 years to 1 in 15 years is estimated as 6.46E-8/yr using the EPRI gui dance; this value increases negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included. Therefore, the estimated change in LERF is determined to be very small usi ng the acceptance guidelines of Regulatory Guide 1.174 [Reference 4]. The risk chan ge resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 i n 15 years bounds the 1 in 10 years to 1 in 15 years risk change. Considering the increase in LERF resulting from a change in the Type A ILRT test interval from 1 in 10 years to 1 in 15 years is estimated as 2.69E-8/yr, the risk increase is very small using the a cceptance guidelines of Regulatory Guide 1.174 [Reference 4].
When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 5.92E-7/yr using the EPRI guidance, and total LERF is 4.0 8E-6/yr. As such, the estimated change in LERF is determined to be small using th e acceptance guidelines of Regulatory Guide 1.174 [Reference 4]. The risk chan ge resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 i n 15 years bounds the 1 in 10 years to 1 in 15 years risk change. When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test int erval from 1 in 10 years to 1 in 15 years is estimated as 2.46E-7/yr, and the total LER F is 3.73E-6/yr. Therefore, the risk increase is small using the acceptance guidelines of Regulatory Guide 1.174
[Reference 4].
The effect resulting from changing the Type A test frequency to 1-per-15 years, measured as an increase to the total integrated pla nt risk for those accident sequences influenced by Type A testing is 0.025 person-rem/yr. NEI 94-01 [Reference 36] states that a small population dose is defined as an inc rease of 1.0 person-rem per year, or 1% of the total population dose, whichever is les s restrictive for the risk impact assessment of the extended ILRT intervals. The resu lts of this calculation meet these criteria. Moreover, the risk impact for the ILRT ex tension when compared to other severe accident risks is negligible.
The increase in the conditional containment failur e probability from the 3 in 10 year interval to 1 in 15 year interval is 0.913%. NEI 94 -01 [Reference 1] states that increases in CCFP of 1.5% is small. Therefore, this incre ase is judged to be small.
Therefore, increasing the ILRT interval to 15 years is considered to be small since it represents a small change to the Callaway risk prof ile.
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Previous Assessments
The NRC in NUREG-1493 [Reference 6] has previously concluded that:
Reducing the frequency of Type A tests (ILRTs) fro m 3 per 10 years to 1 per 20 years was found to lead to an imperceptible increase in r isk. The estimated increase in risk is very small because ILRTs identify only a few potent ial containment leakage paths that cannot be identified by Type B or Type C testing, a nd the leaks that have been found by Type A tests have been only marginally above existi ng requirements.
Given the insensitivity of risk to containment lea kage rate and the small fraction of leakage paths detected solely by Type A testing, in creasing the interval between integrated leakage-rate tests is possible with mini mal impact on public risk. The impact of relaxing the ILRT frequency beyond 1 in 20 years has not been evaluated. Beyond testing the performance of containment penetrations, ILRTs also test integrity of the containment structure.
The conclusions for Callaway confirm these general conclusions on a plant-specific basis considering the severe accidents evaluated for Call away, the Callaway containment failure modes, and the local population surrounding Callawa y.
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A. PRA ACCEPTABILITY
A.1. Introduction This appendix provides information on the technical adequacy of the Callaway Plant, Unit No. 1 (Callaway) Probabilistic Risk Assessment (PRA) Inte rnal Events, Internal Flooding, High Winds, Fire, and Seismic PRA models.
Nuclear Energy Institute (NEI) Topical Report NEI 0 6-09-A, Revision 0 [Reference 45], as clarified by the NRC final safety evaluation of thi s report [Reference 46], defines the technical attributes of a PRA model and its associated Config uration Risk Management Program (CRMP) tool required to implement this risk-informed appli cation. Meeting these requirements satisfies Regulatory Guide (RG) 1.174, An Approach for Deter mining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Info rmed Activities, Revision 3 [Reference 4],
requirements for risk-informed plant-specific chang es to a plant's licensing basis.
Ameren Missouri employs a multi-faceted approach to establishing and maintaining the technical adequacy and fidelity of PRA models for C allaway. This approach includes both a PRA maintenance and update process procedure and th e use of self-assessments and independent peer reviews.
The Callaway PRA models are at-power models consist ing of four hazard models - Internal Flooding, Fire, Seismic, and High Wind. Each hazard model has the Internal Events model as the base with hazard specific initiators added and fault tree modifications and additions made, as necessary. Each model directly addresses plant c onfigurations during plant Modes 1, 2 and 3 of reactor operation. The models provide both core damage frequency (CDF) and large early release frequency (LERF). All five of these PRA mod els were developed to comply with RG 1.200 Revision 2 [Reference 37].
Section A.2 describes the peer review findings clo sure process.
Section A.3 describes the requirements related to the scope of the Callaway PRA models.
Section A.4 addresses the technical adequacy of th e Callaway PRA Internal Events and Internal Flooding model for this application.
Section A.5 addresses the technical adequacy of th e Callaway PRA High Winds model for this application.
Section A.6 addresses the technical adequacy of th e Callaway PRA Seismic model for this application.
Section A.7 addresses the technical adequacy of th e Callaway PRA Fire model for this application.
A.2. Peer Review Findings Closure Process All of the PRA models discussed in this appendix ha ve been peer reviewed and assessed against RG 1.200 Revision 2 [Reference 37].
The review and closure of finding-level F&Os was pe rformed by an independent assessment team using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, "Close-out of Facts and Observations" (F&Os) [Refer ence 33] as accepted by NRC in the letter dated May 3, 2017 (ML17079A427) [Reference 35]. All of the reviews also met the requirements of NEI 17-07 Revision 2 [Reference 47].
The assessment team assessed whether each F&O was c losed through application of a PRA maintenance or upgrade activity, as defined by the ASME/ANS PRA Standard, or through application of a new method. Note that, per APC 17-13,
Subject:
NRC Acceptance of Industry
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Guidance on Closure of PRA Peer Review Findings, d ated May 8, 2017 with attachment Appendix X, a new method represents a fundamentally new approach in addressing a technical aspect of PRA. The results of the peer reviews and independent assessments have been documented and are available for NRC audit.
The PRA scope and technical adequacy is met for thi s application as the Standard requirements for all models are met at Capability Category II (C CII) or higher. There are no open Finding F&Os against any of the models discussed in this ap plication, and all Finding F&Os have been independently assessed and closed using the process es discussed above. The resolved findings and the basis for resolution are documente d in the Callaway PRA documentation and the F&O Closure Review reports.
A.3. Scope of the Callaway PRA Models The Internal Events, Internal Flooding, Fire, High Winds, and Seismic PRA models are at-power models (i.e., they directly address plant configura tions during plant Modes 1, 2 and 3 of reactor operation). The models provide both core damage fre quency (CDF) and large early release frequency (LERF).
Note that the Callaway PRA models do not incorporat e the risk impacts of external events except for High Winds and Seismic. The treatment of non-modeled external risk hazards are discussed in the OEH Notebook [Reference 32] and En closure 4 [Reference 27], which show that all non-modeled external risk hazards screen.
A.4. Technical Adequacy of the Callaway Internal Ev ents and Internal Flooding PRA Model Topical Report NEI 06-09-A requires that the PRA be reviewed to the guidance of RG 1.200
[Reference 37] for a PRA that meets Capability Cate gory II (CCII) for the supporting requirements of the Internal Events at power ASME/A NS PRA Standard [Reference 48].
The information provided in this section demonstrat es that the Callaway Internal Events PRA model (including Internal Flooding) meets the expec tations for PRA scope and technical adequacy as presented in ASME/ANS RA-Sa-2009 [Refer ence 48] and RG 1.200 to fully support this ILRT extension application. The Ameren Missouri risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for Callaway.
Related to the technical adequacy of the Internal E vents model, the Internal Events discussion below describes implementation of the methodology p rovided in PWROG-18027-NP [Reference 49] for assessing the loss of room cooling in PRA m odeling. Following, but unrelated to, implementation of the method provided in PWROG-1802 7-NP into the Callaway PRA, this method was chosen by the PWROG and NEI to pilot the Newly Developed Methods (NDM) peer review process established in NEI 17-07 [Reference 47]. The NEI 17-07 process was successfully completed with all applicable NDM attr ibutes met at Capability Category I/II (CC I/II) and no open peer review Findings against the method in PWROG-18027-NP.
In addition, an implementation peer review and asso ciated F&O closure review have been completed using NRC-approved processes, with no ope n Findings identified against implementation of the method. While the NEI 17-07 p rocess was completed successfully, it is recognized that this process was not an endorsed pr ocess until RG 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Info rmed Activities," Revision 3 was issued in December 2020. As a result, the NRC staff may decid e to independently review the method in PWROG-18027-NP for technical adequacy. The PWROG-18 027-NP report contains the technical basis for the acceptability of the method and is available for NRC audit.
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Peer Review Summary The Internal Events/Internal Flooding PRA was peer reviewed in April 2019. This peer review was a full-scope review of the technical elements o f the Internal Events and Internal Flooding at-power PRA as documented in PWROG-19012-P [Reference 50]. As a full scope review, it included those supporting requirements (SRs) specif ied in PWROG-19020-NP [Reference 44]
for implementation of the methodology for loss of r oom cooling modeling provided in PWROG-18027-NP [Reference 49].
An Independent Assessment of F&Os was conducted in November 2019 and documented in PWROG-19034-P [Reference 51]. The scope of the asse ssment included all Facts and Observations (F&Os) generated in the April 2019 pee r review. All F&Os except for one were closed. The remaining F&O was related to implementa tion of the methodology provided in PWROG-18027-NP [Reference 49] for assessing the los s of room cooling in PRA modeling.
Following, but unrelated to, incorporation of the m ethod provided in PWROG-18027-NP into the Callaway PRA, this method was chosen by the PWROG a nd NEI to pilot the Newly Developed Methods (NDM) peer review process established in NE I 17-07 [Reference 47]. Despite the Callaway assessment, and acknowledgement by the PWR OG, that the method provided in PWROG-18027-NP did not necessarily meet the definit ion of a NDM, Callaway decided to suspend resolution of the associated F&O until the NDM peer review and closure of any F&Os were completed using the process established in NEI 17-07. Also, during the November 2019 independent assessment, two F&O resolutions were de termined to be upgrades to the Internal Events/Internal Flooding PRA. Thus, a focused-scope peer review was required. Based on this focused scope peer review, one new Internal Events F&O was generated.
During February and March 2020, a new peer review, following the guidance in NEI 17-07 Revision 2, was conducted on the method provided in PWROG-18027-NP and documented in PWROG-19020-NP. Based on the results of this review all applicable NDM attributes are met at CC I/II and there are no open peer review Findings against the method in PWROG-18027-NP.
In June 2020, an independent assessment of F&O reso lution and a focused scope peer review, completing the review of PWROG-18027-NP implementat ion, were conducted on the Callaway Internal Events and Fire PRA models. The focused sc ope peer review determined that all of the SRs that were examined, including the SR associated with the F&O related to implementation of the method in PWROG-18027-NP, satisfy CCII or highe r requirements as documented in AMN#PES00031-REPT-001 [Reference 52]. The independe nt assessment of F&Os included an assessment of all remaining open F&O Findings. The results of this review are documented in AMN#PES00031-REPT-002 [Reference 53].
There are no open peer review Findings for the Inte rnal Events/Internal Flooding PRA model.
A.5. Technical Adequacy of Callaway High Winds PRA Model The information provided in this section demonstrat es that the Callaway High Winds PRA model meets the expectations for PRA scope and technical adequacy as presented in ASME/ANS RA-Sa-2009 [Reference 48] and RG 1.200 to fully suppor t this ILRT extension application. The Ameren Missouri risk management process ensures tha t the PRA model used in this application reflects the as-built and as-operated plant for Cal laway.
Peer Review Summary The High Winds PRA was peer reviewed in April 2019 and documented in PWROG-19022-P
[Reference 54]. The scope of this work was to revie w the Callaway External Hazards Screening Assessment and High Winds PRA against the technical elements in Sections 6 and 7 of the ASME/ANS RA-Sa-2009 Standard, and in RG 1.200.
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An Independent Assessment of F&O resolution was con ducted in November 2019 and documented in PWROG-19034-P [Reference 51]. The sco pe of the assessment included all F&Os generated in the April 2019 peer review. All F &Os were closed.
There are no open peer review Findings for the Othe r External Hazards Screening or the High Winds PRA model.
A.6. Technical Adequacy of Callaway Seismic PRA Mod el The information provided in this section demonstrat es that the Callaway Seismic PRA model meets the expectations for PRA scope and technical adequacy as presented in ASME/ANS RA-S CASE 1, Case for ASME/ANS RA-Sb-2013 [Reference 5 5] and RG 1.200 to fully support this ILRT extension application. The Ameren Missouri ris k management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for Callaway.
Peer Review Summary The Seismic PRA was peer reviewed in June 2018 and documented in PWROG-18044-P
[Reference 56]. This peer review was conducted agai nst the requirements of the Code Case for ASME/ANS RA-Sb-2013 [Reference 55], as amended by t he Nuclear Regulatory Commission (NRC) on March 12, 2018 [Reference 57]. The Code Ca se is an approved alternative to Part 5 of ASME/ANS RA-Sb-2013 Addendum B, the American Soc iety of Mechanical Engineers (ASME) / American Nuclear Society (ANS) Probabilist ic Risk Assessment (PRA) Standard.
An Independent Assessment of F&Os was conducted in March 2019. The scope of the assessment included all but two of the F&Os generat ed in the June 2018 peer review. All in-scope F&Os were closed as documented in PWROG-19011 -P [Reference 58]. Also, in the March 2019 review documented in PWROG-19011-P, thre e SRs were the subject of a focused-scope peer review based on the closures of associat ed F&Os being assessed as upgrades. As a result of that peer review, the three SRs were de termined to be met at CCII.
Subsequently, another Independent Assessment of F&O s was conducted in June 2020 and documented in AMN#PES00031-REPT-002 [Reference 53]. The scope of the assessment included all remaining F&Os generated in the June 2 018 peer review. All F&Os were closed.
There are no open peer review Findings for the Seis mic PRA model.
A.7. Technical Adequacy of Callaway Fire PRA Model The information provided in this section demonstrat es that the Callaway Fire PRA model meets the expectations for PRA scope and technical adequa cy as presented in ASME/ANS RA-Sa-2009 [Reference 48] and RG 1.200 to fully support t his ILRT extension application. The Ameren Missouri risk management process ensures that the P RA model used in this application reflects the as-built and as-operated plant for Callaway.
The Internal Fire PRA model was developed consisten t with NUREG/CR-6850 [Reference 59]
and only utilizes methods previously accepted by th e NRC. Callaway was approved to implement NFPA-805 in January 2014, and since that time, there have been numerous updates to the approved methods through the issuance of Fir e PRA frequently asked questions and new or revised guidance documents. New or revised guida nce is specifically addressed through the Callaway PRA maintenance and update process. The Am eren Missouri risk management process ensures that the PRA model used in this app lication reflects the as-built and as-operated plant for Callaway.
It should also be noted that, as part of transition to NFPA 805, there were several committed modifications and implementation items as documente d in NFPA 805 LAR Attachment S, "Plant
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Modifications and Items to be Completed during Impl ementation," which described the Callaway plant modifications necessary to implement the NFPA 805 licensing basis. All NFPA 805 LAR Attachment S items have been implemented; therefore, there are no NFPA 805 open items impacting this application.
Peer Review Summary The Fire PRA was prepared using the methodology def ined in NUREG/CR-6850, "Fire PRA Methodology for Nuclear Power Facilities," to suppo rt a transition to National Fire Protection Association (NFPA) Standard 805, "Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants." The Fir e PRA was peer reviewed to ASME/ANS RA-Sa-2009 and RG 1.200 Revision 2 in October 2009. The review is documented in LTR-RAM-11-10-019 [Reference 60].
An Independent Assessment of F&Os was conducted in June 2019 and documented in AMN#PES00021-REPT-001 [Reference 61].
In June 2020, an independent assessment of F&Os and a focused scope peer review were conducted for the Callaway Internal Events and Fire PRA models. The focused scope peer review generated additional Fire PRA related F&Os a s documented in AMN#PES00031-REPT-001 [Reference 52]. The independent assessment of F &Os included an assessment of all remaining open F&O Findings. As documented in AMN#P ES00031-REPT-002 [Reference 53],
all Finding F&Os were closed, including the Fire PR A Findings identified in the Focused Scope peer review.
In fulfillment of Commitment 50437 in Enclosure 4 t o ULNRC-06550 (ML20304A456) and associated with closure of NFPA 805 LAR Table S-3 I mplementation Item 13-805-001, a focused scope peer review was conducted in November 2020, as documented in AMN#PES00031-REPT-003 [Reference 29], for the resol ution of Fire PRA Suggestion F&O FSS-B1-03, which a July 2019 F&O closure review had determined to be an upgrade, as documented in AMN#PES00021-REPT-001 [Reference 61].
As documented in AMN#PES00042-REPT-002 [Reference 2 6], the F&Os from this focused scope peer review were closed during an F&O closure review in February 2021. The results of this review formally closed Commitment 50437.
There are no open peer review Findings for the Fire PRA model.
A.8. Summary The PRA scope and technical adequacy is met for thi s application as the Standard requirements for all models are met at CCII or higher. There are no open Finding F&Os against any of the models discussed in this Enclosure, and all Finding F&Os have been independently assessed and closed using the processes discussed in Section 2 of this Enclosure. In addition, all of the reviews also met the requirements of NEI 17-07 Revi sion 2 [Reference 47].
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