ML15051A410

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Evaluation of Risk Significance of Permanent ILRT Extension
ML15051A410
Person / Time
Site: Calvert Cliffs  Constellation icon.png
Issue date: 02/17/2015
From: Mary Johnson
Exelon Generation Co
To:
Office of Nuclear Reactor Regulation
References
0054-0001 -000-CALC -001
Download: ML15051A410 (111)
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ATTACHMENT (3)

EVALUATION OF RISK SIGNIFICANCE OF PERMANENT ILRT EXTENSION Calvert Cliffs Nuclear Power Plant February 17, 2015

JENSEN HUGHES Calvert Cliffs Nuclear Power Plant:

Evaluation of Risk Significance of Permanent ILRT Extension 0054-0001 -000-CALC -001 Prepared for:

Calvert Cliffs Nuclear Power Plant Project Number: 0054-0001-000 Project

Title:

Permanent ILRT Extension Revision: 4 Name and Date 0 NOitally signed by Matt Preparer: Matthew Johnson pIoson

' .Bat: 2015.02.10 1'6

... 0:54-06'00' igi tally signed by Nicholas Reviewer: Nicholas Lovelace Dae 2015.02.10 18:11:42-06'00' Review Method Design Review 0 Alternate Calculation E

¶ Digitally signed by Richard Anoba Approved by: Richard Anoba Ri c h a rd A n o b DN:cn=Richard Anoba°o=Jensen Hughes, ou=Risk I'nfiormed

. Engineering, email=ranoba@haifire.com, c=US Date: 2015.02.1109:06:47 -05'00' Revision 4 Page 1 of 110

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension REVISION RECORD

SUMMARY

Revision Revision Summary 0 Initial Issue 1 Incorporated True North review comments.

2 Incorporated minor comments regarding NFPA 805 transition in Section 5.1.2.

3 Removed generic QA condition statement and generic containment overpressure discussion in section 2.0.

Incorporated RAI responses. Edited Table 1 of the attachment to indicate and discuss which Supporting Requirements were "not met." Added Tables 5 5-36 with CCFP and annual dose rates for the steel liner corrosion senstivities. Added phrases in Section 5.3.2 to introduce these tables. Updated Tables 5-25, 5-26, and 5-27 and other calculated values in Section 5.3.1 to reflect the Seismic CDF values given in Section 8 of the IPEEE report for both units. Updated Table 5-28, 5-29, and 5-30 and other calculated values in Section 5.3.1.1 to reflect the Unit 2 Fire CDF value given in Section 8 of the IPEEE. Added additional discussion to Section 5.2.4 to clarify the release timing. Added Attachment 2, which contains the release timing plots from NC-94-020 [Reference .18] for MAAP cases HRIF, GIOY, and MRIF. Revised Section 5.1.4 to include disposition of recent corrosion data. Revised section 5.3.2 to clarify that the dose increase estimate is based on the class 3b contribution with containment spray success and late sequences removed. Added Attachment 3 as placeholder for RAI responses.

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IRCA-54001-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension TABLE OF CONTENTS 1.0 PURPOSE ...................................................................................................................... 4 2.0 SCOPE ........................................................................................................................... 4

3.0 REFERENCES

....................................................................................................... 6 4.0 ASSUMPTIONS AND LIMITATIONS ........................................................................... 8 5.0 METHODOLOGY and analysis .................................................................................... 8 5.1 Inputs ........................................................................................................................... 8 5.1.1 General Resources Available ............................................................................. 8 5.1.2 Plant Specific Inputs ............................................................................................ 11 5.1.3 Impact of Extension on Detection of Component Failures that Lead to Leakage (Small and Large) .......................................................................................... 14 5.1.4 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage

.......................................................................................................................... 16 5.2 Analysis ...................................................................................................................... 18 5.2.1 Step 1 - Quantify the Baseline Risk in Terms of Frequency per Reactor Year ..... 19 5.2.2 Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose) ........... 23 5.2.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval from 10 to 15 Years ................................................................................................................. 24 5.2.4 Step 4 - Determine the Change in Risk in Terms of LERF ............................... 27 5.2.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability (CCFP) .......................................................................................................... 31 5.3 Sensitivities ................................................................................................................. 32 5.3.1 Potential Impact from External Events Contribution ........................................ 32 5.3.1.1 Potential Impact from External Events Contribution Using IPEEE Fire Analysis ............................................................................................... 34 5.3.2 Potential Impact from Steel Liner Corrosion Likelihood .................................... 36 5.3.3 Expert Elicitation Sensitivity ............................................................................ 40 5.3.4 Large Leak Probability Sensitivity Study ................................. :............................ 42 6.0 RESULTS ..................................................................................................................... 43

7.0 CONCLUSION

S AND RECOMMENDATIONS ........................................................ 45 A. Attachment 1 ................................................................................................................. 46 B. Attachment 2 ................................................................................................................. 97 C. Attachment 3 ................................................................................................................ 101 Page 3 of 110 Revision 4 Revision Page 3 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 1.0 PURPOSE The purpose of this analysis is to provide a risk assessment of extending the currently allowed containment Type A Integrated Leak Rate Test (ILRT) to permanent fifteen years. The extension would allow for substantial cost savings as the ILRT could be deferred for additional scheduled refueling outages for the Calvert Cliffs Nuclear Power Plant (CCNPP). The risk assessment follows the guidelines from NEI 94-01, Revision 3-A [Reference 1], the methodology used in EPRI TR-104285 [Reference 2], the NEI "Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals" from November 2001 [Reference 3], the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in Regulatory Guide 1.200 as applied to ILRT interval extensions, and risk insights in support of a request for a plant's licensing basis as outlined in Regulatory Guide (RG) 1.174 [Reference 4], the methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval [Reference 5],

the methodology used in EPRI 1009325, Revision 2-A [Reference 24], and the methodology improvements in EPRI 1018243 [Reference 24].

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

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 the 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], provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessment of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the NRC's rulemaking basis, NEI undertook a similar study. The results of that study are documented in Electric Power Research Institute (EPRI) Research Project TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals".

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

NEI 94-01 Revision 2-A contains a Safety Evaluation Report that supports using EPRI Report No. 1009325 Revision 2-A, Risk Impact Assessment of Extended IntegratedLeak Rate Testing Intervals, for performing risk impact assessments in support of ILRT 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. This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed ILRT changes.

Revision 4 Page 4 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension It should be noted that containment leak-tight integrity is also verified through periodic in-service inspections conducted in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI. More specifically, Subsection IWE provides the rules and requirements for in-service inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants. Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct visual inspections of the accessible areas of the interior of the containment. The associated change to NEI 94-01 will require that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being conducted. These requirements will not be changed as a result of the extended ILRT interval. In addition, Appendix J, Type B local leak tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are also not affected by the change to the Type A test frequency.

The acceptance guidelines in RG 1.174 are used to assess the acceptability of this permanent extension of the Type A test interval beyond that established during the Option B rulemaking of Appendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in Core Damage Frequency (CDF) less than 10-6 per reactor year and increases in Large Early Release Frequency (LERF) less than 10-7 per reactor year. 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-depth and encourages the use of risk analysis techniques to help ensure and show that key principles, such as the defense-in-depth philosophy, are met. Therefore, the increase in the Conditional Containment Failure Probability (CCFP), which helps ensure the defense-in-depth philosophy is maintained, is also calculated.

Regarding CCFP, changes of up to 1.1% have been accepted by the NRC for the one-time requests for extension of ILRT intervals. In context, 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 small.

In additional, the total annual risk (person rem/year population dose) is examined to demonstrate the relative change in this parameter. While no acceptance guidelines for these additional figures of merit are published, examinations of NUREG-1493 and Safety Evaluation Reports (SER) for one-time interval extension (summarized in Appendix G) indicate a range of incremental increases in population dose that have been accepted by the NRC. The range of incremental population dose Increases is from <0.01 to 0.2 person-rem/year and/or 0.002% to 0.46% of the total accident dose. 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 very small population dose is defined as an increase from the baseline interval (3 tests per 10 years) dose of <1.0 person-rem per year or 1% of the total baseline dose, whichever is less restrictive for the risk impact assessment of the proposed extended ILRT interval.

For those plants that credit containment overpressure for the mitigation of design basis accidents, a brief description of whether overpressure is required should be included in this section. In addition, if overpressure is included in the assessment, other risk metrics such as CDF should be described and reported.

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I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension

3.0 REFERENCES

The following references were used in this calculation:

1. Revision 3-A to Industry Guideline for Implementing Performance-BasedOption of 10 CFR Part50, Appendix J, 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 Assessments in Support of One-Time Extensions for ContainmentIntegrated Leakage Rate Test Surveillance Intervals, Revision 4, developed for NEI by EPRI and Data Systems and Solutions, November 2001.
4. An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, May 2011.
5. Response to Request for Additional Information Concerning the License Amendment Request for a One-Time IntegratedLeakage Rate Test Extension, Letter from Mr. C. H.

Cruse (Calvert Cliffs Nuclear Power Plant) to NRC Document Control Desk, Docket No.

50-317, March 27, 2002.

6. Performance-Based Containment Leak-Test Program, NUREG-1493, September 1995.
7. Evaluationof 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. Nuclear Regulatory Commission, IPN-01-007, January 18, 2001.
9. United States Nuclear Regulatory Commission, Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Frequency of Performance-Based Leakage Rate Testing (TAC No. MB0178), April 17, 2001.
10. Impact of Containment Building Leakage on LWR Accident Risk, Oak Ridge National Laboratory, NUREG/CR-3539, ORNI/TM-8964, April 1984.
11. ReliabilityAnalysis of ContainmentIsolation Systems, 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 Requirements, Pacific Northwest Laboratory, NUREG/CR-4330, PNL-5809, Volume 2, June 1986.
14. Shutdown Risk Impact Assessment for Extended Containment Leakage Testing Intervals Utilizing ORAMTM, EPRI, Palo Alto, CA, TR-105189, Final Report, May 1995.
15. Severe Accident Risks: An Assessment for Five U. S. Nuclear Power Plants, NUREG-1150, December 1990.
16. United States Nuclear Regulatory Commission, Reactor Safety Study, WASH-1400, October 1975.
17. Calculation No. CO-QU-001, Revision 1, Calvert Cliffs Nuclear Power Plant, Unit 1, "PRA Quantification (QU) Notebook," July 2012.
18. Calculation No. NC-94-020, "Severe Accident Analysis of Calvert Cliffs for IPE Level I1,"

December 1994.

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1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension

19. Massoud, M., Calculation No. CA07463, Revision 0, "2010 Update of Dose Analysis for Level 3 PRA Release Categories," August 2001.
20. Anthony R. Pietrangelo, One-time extensions of containment integrated leak rate test interval - additional information, NEI letter to Administrative Points of Contact, November 30, 2001.
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. FarleyNuclear Plant Regarding ILRT (Type A)

Extension Request, prepared for Southern Nuclear Operating Co. by ERIN 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 IntegratedLeak Rate Testing Intervals, Revision 2-A of 1009325, EPRI, Palo Alto, CA. 1018243, October 2008.
25. Risk Assessment for Vogtle Electric Generating Plant Regarding the ILRT (Type A)

Extension Request, prepared for Southern Nuclear Operating Co. by ERIN Engineering and Research, February 2003.

26. Perspectives Gained from the IPEEE Program, USNRC, NUREG-1742, April 2002.
27. Procedure STP M-662-1, Revision 6, Calvert Cliffs Nuclear Power Plant, Unit 1, "Integrated Leak Rate Test Unit 1 Containment."
28. Procedure STP M-662-2, Revision 7, Calvert Cliffs Nuclear Power Plant, Unit 2, "Integrated Leak Rate Test Unit 2 Containment."
29. Calculation No. CO-QU-002, Revision 1, Calvert Cliffs Nuclear Power Plant, Unit 2, "PRA Quantification (QU) Notebook," August 2010.
30. Calculation No. RSC 10-21, Revision 0, Calvert Cliffs Nuclear Power Plant, Unit 2, "Evaluation of Risk Significance of ILRT Extension," August 2010.
31. Armstrong, J., Simplified Level 2 Modeling Guidelines: WOG PROJECT: PA-RMSC-0088, Westinghouse, WCAP-16341-P, November 2005.
32. Calculation CO-LE-001, Revision 1, CENG, Units 1 and 2, "PRA Level 2 Notebook," May 2010.
33. Landale, J., PRAER No. CO-2010-012, CENG, CO-2010-012, August 2010.
34. Harrison, D., Generic Component Fragilities for the GE Advanced BWR Seismic Analysis, International Technology Corporation, September 1988.
35. Calculation No. RAN 97-031, IPEEE, Calvert Cliffs Nuclear Power Plant, "Individual Plant Examination of External Events," August 1997.
36. Report KLD-TR-21, Development of Evacuation Time Estimates, April 2008.
37. Letter L-14 -121, ML14111A291, FENOC Evaluation of the Proposed Amendment, Beaver Valley Power Station, Unit Nos. 1 and 2, April 2014.
38. Technical Letter Report ML112070867, Containment Liner Corrosion Operating Experience Summary, Revision 1, August 2011.

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I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 4.0 ASSUMPTIONS AND LIMITATIONS The following assumptions were used in the calculation:

" The technical adequacy of the CCNPP PRA is consistent with the requirements of Regulatory Guide 1.200 as is relevant to this ILRT interval extension, as detailed in Attachment 1.

" The CCNPP Level 1 and Level 2 internal events PRA models provide representative results.

" It is appropriate to use the CCNPP internal events PRA model as a gauge to effectively describe the risk change attributable to the ILRT extension. An extensive sensitivity study is done in Section 5.3.1 to show the effect of including external event models for the ILRT extension. The IPEEE simplified seismic PRA [Reference 35] and the detailed Fire PRA (model 6.1 M) are used for this sensitivity analysis. It is reasonable to assume that the impact from the ILRT extension (with respect to percent increases in population dose) will not substantially differ if detailed analysis of seismic events were to be included in the calculations.

" Accident classes describing radionuclide release end states are defined consistent with EPRI methodology [Reference 2].

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

" The representative containment leakage for Class 3a sequences is 10La based on the previously approved methodology performed for Indian Point Unit 3 [Reference 8, Reference 9].

" The representative containment leakage for Class 3b sequences is 1 00La based on the guidance provided in EPRI Report No. 1009325, Revision 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 bypass scenarios is not altered by the proposed ILRT extension, but is accounted for in the EPRI methodology as a separate entry for comparison purposes. Since the containment bypass contribution to population dose is fixed, no changes on the conclusions from this analysis will result from this separate categorization.

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

5.0 METHODOLOGY AND ANALYSIS 5.1 Inputs This section summarizes the general resources available 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 risk assessment are briefly summarized here:

1. NUREG/CR-3539 [Reference 10]
2. NUREG/CR-4220 [Reference 11]
3. NUREG-1273 [Reference 12]
4. NUREG/CR-4330 [Reference 13]

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5. EPRI TR-105189 [Reference 14]
6. NUREG-1493 [Reference 6]
7. EPRI TR-1 04285 [Reference 2]
8. NUREG-1150 [Reference 15] and NUREG/CR-4551 [Reference 71
9. NEI Interim Guidance [Reference 3, Reference 20]
10. Calvert Cliffs liner corrosion analysis [Reference 5]
11. EPRI Report No. 1009325, Revision 2-A (EPRI 1018243), Appendix H [Reference 24]

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 leakage that is considered significant and is to be included in the model. The second study is applicable because it provides a basis of the probability for significant pre-existing containment leakage at the time of a core damage accident. The third study is applicable because it is a subsequent study to NUREG/CR-4220 that undertook a more extensive evaluation of the same database. The fourth study provides an assessment of the impact of different containment leakage rates on plant risk. The fifth study provides an assessment of the impact on shutdown risk from ILRT test interval extension. The sixth study is the NRC's cost-benefit analysis of various alternative approaches regarding extending the test intervals and increasing the allowable leakage rates for containment integrated and local leak rate tests. The seventh study is an EPRI study of the impact of extending ILRT and LLRT test intervals on at-power public risk. The eighth study provides an ex-plant consequence analysis for a 50-mile radius surrounding a plant that is used as the basis for the consequence analysis of the ILRT interval extension for CCNPP. The ninth study includes the NEI recommended methodology (promulgated in two letters) for evaluating the risk associated with obtaining a one-time extension of the ILRT interval. The tenth study addresses the impact of age-related degradation of the containment liners on ILRT evaluations. Finally, the eleventh study builds on the previous work and includes a recommended methodology and template for evaluating the risk associated with a permanent 15-year extension of the ILRT interval.

NUREG/CR-3539 [Reference 101 Oak Ridge National Laboratory documented a study of the impact of containment leak rates on public risk in NUREG/CR-3539. This study uses information from WASH-1400 [Reference 16]

as the basis for its risk sensitivity calculations. ORNL concluded that the impact of leakage rates on LWR accident risks is relatively small.

NUREG/CR-4220 [Reference 111 NUREG/CR-4220 is a study performed by Pacific Northwest Laboratories for the NRC in 1985.

The study reviewed over two thousand LERs, ILRT reports and other related records to calculate the unavailability of containment due to leakage.

NUREG-1273 [Reference 121 A subsequent NRC study, NUREG-1273, performed a more extensive evaluation of the NUREG/CR-4220 database. This assessment noted that about one-third of the reported events were leakages that were immediately detected and corrected. In addition, this study noted that local leak rate tests can detect "essentially all potential degradations" of the containment isolation system.

NUREG/CR-4330 [Reference 131 NUREG/CR-4330 is a study that examined the risk impacts associated with increasing the allowable containment leakage rates. The details of this report have no direct impact on the Revision 4 Page 9 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension modeling approach of the ILRT test interval extension, as NUREG/CR-4330 focuses on leakage rate and the ILRT test interval extension study focuses on the frequency of testing intervals.

However, the general conclusions of NUREG/CR-4330 are consistent with NUREG/CR-3539 and other similar containment leakage risk studies:

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

EPRI TR-105189 [Reference 141 The EPRI study TR-1 05189 is useful to the ILRT test interval extension risk assessment because it provides insight regarding the impact of containment testing on shutdown risk. This study contains a quantitative evaluation (using the EPRI ORAM software) for two reference plants (a BWR-4 and a PWR) of the impact of extending ILRT and LLRT test intervals on shutdown risk. The conclusion from the study is that a small, but measurable, safety benefit is realized from extending the test intervals.

NUREG-1493 [Reference 61 NUREG-1493 is the NRC's cost-benefit analysis for proposed alternatives to reduce containment leakage testing intervals and/or relax allowable leakage rates. The NRC conclusions are consistent with other similar containment leakage risk studies:

Reduction in ILRT frequency from 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 interval between integrated leak rate tests is possible with minimal impact on public risk.

EPRI TR-104285 [Reference 21 Extending the risk assessment impact beyond shutdown (the earlier EPRI TR-1 05189 study),

the EPRI TR-104285 study is a quantitative evaluation of the impact of extending ILRT and LLRT test intervals on at-power public risk. This study combined IPE Level 2 models with NUREG-1 150 Level 3 population dose models to perform the analysis. The study also used the approach of NUREG-1493 in calculating the increase in pre-existing leakage probability due to extending the ILRT and LLRT test intervals.

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

1. Containment intact and isolated
2. Containment isolation failures dependent upon the core damage accident
3. Type A (ILRT) related containment isolation failures
4. Type B (LLRT) related containment isolation failures
5. Type C (LLRT) related containment isolation failures
6. Other penetration related containment isolation failures
7. Containment failures due to core damage accident phenomena
8. Containment bypass Consistent with the other containment leakage risk assessment studies, this study concluded:

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

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I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension NUREG-1150 [Reference 151 and NUREG/CR-4551 [Reference 71 NUREG-1 150 and the technical basis, NUREG/CR-4551, provide an ex-plant consequence analysis for a spectrum of accidents including a severe accident with the containment remaining intact (i.e., Tech Spec Leakage). This ex-plant consequence analysis is calculated for the 50-mile radial area surrounding Surry. The ex-plant calculation can be delineated to total person-rem for each identified Accident Progression Bin (APB) from NUREG/CR-4551. With the CCNPP Level 2 model end-states assigned to one of the NUREG/CR-4551 APBs, it is considered adequate to represent CCNPP. (The meteorology and site differences other than population are assumed not to play a significant role in this evaluation.)

NEI Interim Guidance for Performinq Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals [Reference 3.

Reference 201 The guidance provided in this document builds on the EPRI risk impact assessment methodology [Reference 2] and the NRC performance-based containment leakage test program

[Reference 6], and considers approaches utilized in various submittals, including Indian Point 3 (and associated NRC SER) and Crystal River.

Calvert Cliffs Response to Request for Additional Information Concerning the License Amendment for a One-Time Integrated Leakage Rate Test Extension [Reference 51 This submittal to the NRC describes a method for determining the change in likelihood, due to extending the ILRT, of detecting liner corrosion, and the corresponding change in risk. The methodology was developed for Calvert Cliffs in response to a request for additional information regarding how the potential leakage due to age-related degradation mechanisms was factored into the risk assessment for the ILRT one-time extension. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete base-mat, each with a steel liner.

EPRI Report No. 1009325, Revision 2-A. Risk Impact Assessment of Extended Inte-grated Leak Rate Testing Intervals [Reference 241 This report provides a generally applicable assessment of the risk involved in extension of ILRT test intervals to permanent 15-year intervals. Appendix H of this document provides guidance for performing plant-specific supplemental risk impact assessments and builds on the previous EPRI risk impact assessment methodology [Reference 2] and the NRC performance-based containment leakage test program [Reference 6], and considers approaches utilized in various submittals, including Indian Point 3 (and associated NRC SER) and Crystal River.

The approach included in this guidance document is used in the CCNPP assessment to determine the estimated increase in risk associated with the ILRT extension. This document includes the bases for the values assigned in determining the probability of leakage for the EPRI Class 3a and 3b scenarios in this analysis, as described in Section 5.2.

5.1.2 Plant Specific Inputs The plant-specific information used to perform the CCNPP ILRT Extension Risk Assessment includes the following:

" Level 1 Model results: Unit 1 [Reference 17] and Unit 2 [Reference 29]

" Level 2 Model results [Reference 17, Reference 18, Reference 19]

" Release category definitions used in the Level 2 Model [Reference 18, Reference 19]

" Dose within a 50-mile radius [Reference 19]

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" ILRT results to demonstrate adequacy of the administrative and hardware issues

[Reference 30]

" Containment failure probability data [Reference 18, References 32 and 33]

Level 1 Model The Level 1 Internal Events PRA Model that is used for CCNPP is characteristic of the as-built plant. The current Level 1 model (CCNPP PRA Model Version 6.2a) [Reference 17] is a linked fault tree model, and was quantified with the total Internal Events Core Damage Frequency (CDF) = 1.61 E-5/year for Unit 1 and CDF = 1.41 E-5/year for Unit 2. The total External Event CDF (excluding seismic) = 3.24E-5/year for Unit 1 and 3.71 E-5/year for Unit 2. Table 5-1 provides a summary of the Internal Events CDF results for CCNPP PRA Model Version 6.2a.

Table 5-2 provides a summary of the External Events CDF results. The High Winds are included in CCNPP PRA Model Version 6.2a. The Fire PRA results come from Model Version 6.1M. The Seismic PRA results come from the IPEEE Seismic Analysis [Reference 35].

Table 5 Internal Events CDF (CCNPP PRA Model Version 6.2a)

Internal Events Unit I Frequency (per year) Unit 2 Frequency (per year)

LOCAs 5.88E-6 7.70E-6 Internal Floods 6.18E-6 1.06E-6 Transients 3.40E-6 4.70E-6 ISLOCA 1.97E-7 1.97E-7 SGTR 4.71E-7 4.60E-7 Total Internal Events CDF 1.61 E-5 1.41 E-5 Total Internal Events CDF (Excluding ISLOCA & SGTR)

Table 5 External Events CDF External Events Unit I Frequency (per year) Unit 2 Frequency (per year)

Fire 3.15E-5 3.59E-5 High Winds 9.19E-7 1.23E-6 Seismic 1.29E-5 1.52E-5 Total External Events CDF 4.53E-5 5.23E-5 Note that the above Fire PRA values reflect the anticipated configuration of the plant upon full implementation of NFPA 805 and related plant modifications to resolve fire protection issues.

Refer to Section 5.3.1.

Level 2 Model The Level 2 Model that is used for CCNPP was developed with guidance from WCAP-1 6341-P to calculate the LERF contribution, as well as the other release end states evaluated in the model: INTACT, SERF (small early release frequency), and LATE [Reference 31]. The current LERF model (CCNPP PRA Model Version 6.2a) [Reference 17] is a linked fault tree model and was quantified with the total Unit 1 Internal Events LERF = 1.39E-6/year and Unit 2 Internal Revision 4 Page 12 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Events LERF = 1.56E-6/year. The total Unit 1 External Event LERF (excluding seismic) =

2.99E-6/year and Unit 2 External Event LERF (excluding seismic) = 4.21 E-6/year. Table 5-3 provides a summary of the Internal Events LERF results for CCNPP PRA Model Version 6.2a.

Table 5-4 provides a summary of the External Events CDF results. The High Winds are included in CCNPP PRA Model Version 6.2a. The Fire PRA results come from Model Version 6.1M. The Seismic PRA results come from the IPEEE Seismic Analysis [Reference 35].

Table 5 Internal Events LERF (CCNPP PRA Model Version 6.2a)

Internal Events Unit 1 Frequency (per year) Unit 2 Frequency (per year)

LOCAs 3.26E-7 4.01 E-7 Internal Floods 2.46E-7 2.17E-7 Transients 1.50E-7 2.84E-7 ISLOCA 1.97E-7 1.97E-7 SGTR 4.71E-7 4.60E-7 Total Internal Events LERF 1.39E-6 1.56E-6 Table 5 External Events LERF External Events Unit I Frequency (per year) Unit 2 Frequency (per year)

Fire 2.97E-6 4.17E-6 High Winds 2.21E-8 3.77E-8 Seismic 1.41 E-6 1.66E-6 Total External Events CDF 4.40E-6 5.87E-6 Note that the above Fire PRA values reflect the anticipated configuration of the plant upon full implementation of NFPA 805 and related plant modifications to resolve fire protection issues.

Refer to Section 5.3.1.

Population Dose Calculations The population dose calculation was performed for the CCNPP Severe Accident Mitigation Alternatives (SAMA) analyses [Reference 19] in 2010. Table 5-5 presents dose exposures calculated from methodology described in Reference 1 and data from Reference 19. Reference 19 provides the population dose (person-rem) for Classes 1, 2, 6, 7, and 8; Class 3a and 3b population dose values are calculated from the Class 1 population dose and represented as 1OL, and 1OOL a, respectively, as guidance in Reference 1 dictates.

Table 5 Population Dose Accident Class Description Release (person-rem) 1 Containment Remains Intact 3.20E+04 2 Containment Isolation Failures 2.OOE+07 3a Independent or Random Isolation Failures SMALL 3.20E+051 3b Independent or Random Isolation Failures LARGE 3.20E+06 2 3o1 Revisin4Pg Revision 4 Page 13 of 110

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Population Dose Accident Class Description Release (person-rem)

Isolation Failure in which pre-existing leakage is not n/a dependent on sequence progression. Type B test Failures 5 Isolation Failure in which pre-existing leakage is not n/a dependent on sequence progression. Type C test Failures 6 Isolation Failure that can be verified by IST/IS or 7.01 E+06 surveillance 7 Containment Failure induced by severe accident 5.61 E+07 8 Accidents in which containment is by-passed 2.25E+07

1. 10*La
2. 100 - La Release Category Definitions Table 5-6 defines the accident classes used in the ILRT extension evaluation, which is consistent with the EPRI methodology [Reference 2]. These containment failure classifications are used in this analysis to determine the risk impact of extending the Containment Type A test interval, as described in Section 5.2 of this report.

Table 5 EPRI Containment Failure Classification [Reference 2]

Class Description Containment remains intact including accident sequences that do not lead to containment failure in the 1 long term. The release of fission products (and attendant consequences) is determined by the maximum allowable leakage rate values La, under Appendix J for that plant.

Containment isolation failures (as reported in the Individual Plant Examinations) including those accidents in which there is a failure to isolate the containment.

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

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

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

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 (either as an initial condition or induced by phenomena) are included in Class 8. Changes in Appendix J testing 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 surfaces, which can lead to leakage. The proposed ILRT test interval extension may influence the conditional probability of detecting these types of failures. To ensure that this effect is properly addressed, the EPRI Class 3 Revision 4 Page 14 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension accident class, as defined in Table 5-6, is divided into two sub-classes, Class 3a and Class 3b, representing small and large leakage failures respectively.

The probability of the EPRI Class 3a and Class 3b failures is determined consistent with the EPRI Guidance [Reference 24]. For Class 3a, the probability is based on the maximum likelihood estimate of failure (arithmetic average) from the available data (i.e., 2 "small" failures in 217 tests leads to "large" failures in 217 tests (i.e., 2/217 = 0.0092). For Class 3b, the probability is based on the Jeffrey's Non-Uniform Prior (i.e., 0.5/ 218 = 0.0023).

In a follow-up letter [Reference 20] to their ILRT guidance document [Reference 3], NEI issued additional information concerning the potential that the calculated delta LERF values for several plants may fall above the "very small change" guidelines of the NRC Regulatory Guide 1.174

[Reference 4]. This additional NEI information includes a discussion of conservatisms in the quantitative guidance for ALERF. NEI describes ways to demonstrate that, using plant-specific calculations, the ALERF is smaller than that calculated by the simplified method.

The supplemental information states:

The methodology employed for determining LERF (Class 3b frequency) involves conservatively multiplying the CDFby the failure probabilityfor this class (3b) of accident. This was done for simplicity and to maintain conservatism. However, some plant-specific accident classes leading to core damage are likely to include individual sequences that eithermay already (independently) cause a LERF or could never cause a LERF, and are thus not associatedwith a postulatedlarge Type A containment leakage path (LERF). These contributorscan be removed from Class 3b in the evaluation of LERF by multiplying the Class 3b probabilityby only that portion of CDF that may be impacted by Type A leakage.

The application of this additional guidance to the analysis for CCNPP, as detailed in Section 5.2, involves the following:

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

" A review of Class 1 accident sequences shows that several of these cases involve successful operation of containment sprays. For calculation of the Class 3b and Class 3a frequencies, the fraction of the Class 1 CDF associated with successful operation of containment sprays could also be subtracted. Successful operation of containment sprays result in lower containment pressure with subsequent reduction in containment leakage. This conservatism was removed for the CCNPP ILRT analysis, as detailed in Section 5.2.4.

Consistent with the NEI Guidance [Reference 3], the change in the leak detection probability can be estimated by comparing the average time that a leak could exist without detection. For example, the average time that a leak could go undetected with a three-year test interval is 1.5 years (3 years / 2), and the average time that a leak could exist without detection for a ten-year interval is 5 years (10 years / 2). This change would lead to a non-detection probability that is a factor of 3.33 (5.0/1.5) higher for the probability of a leak that is detectable only by ILRT testing.

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-detection probability of a leak.

Page 15 of 110 Revision 44 Revision Page 15 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension It should be noted that using the methodology discussed above is very conservative compared to previous submittals (e.g., the IP3 request for a one-time ILRT extension that was approved by the NRC [Reference 9]) because it does not factor in the possibility that the failures could be detected by other tests (e.g., the Type B local leak rate tests that will still occur). Eliminating this possibility conservatively over-estimates the factor increases attributable to the ILRT extension.

5.1.4 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage An estimate of the likelihood and risk implications of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is evaluated using the methodology from the Calvert Cliffs liner corrosion analysis [Reference 5]. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete base-mat, each with a steel liner.

The following approach is used to determine the change in likelihood, due to extending the ILRT, of detecting corrosion of the containment steel liner. This likelihood is then used to determine the resulting change in risk. Consistent with the Calvert Cliffs analysis, the following issues are addressed:

Differences between the containment base-mat and the containment cylinder and dome:

" The historical steel liner flaw likelihood due to concealed corrosion

" The impact of aging

" The corrosion leakage dependency on containment pressure

" The likelihood that visual inspections will be effective at detecting a flaw Assumptions

" Consistent with the Calvert Cliffs analysis, a half failure is assumed for base-mat concealed liner corrosion due to the lack of identified failures. (See Table 5-7, Step 1)

" The two corrosion events used to estimate the liner flaw probability in the Calvert Cliffs previous analysis are assumed to still be applicable.

" Consistent with the Calvert Cliffs analysis, the estimated historical flaw probability is also limited to 5.5 years to reflect the years since September 1996 when 10 CFR 50.55a started requiring visual inspection. Additional success data was not used to limit the aging impact of this corrosion issue, even though inspections were being performed prior to this date (and have been performed since the time frame of the Calvert Cliffs analysis), and there is no evidence that additional corrosion issues were identified (See Table 5-7, Step 1).

" Consistent with the Calvert Cliffs analysis, the steel liner flaw likelihood is assumed to double every five years. This is based solely on judgment and is included in this analysis to address the increased likelihood of corrosion as the steel liner ages (See Table 5-7, Steps 2 and 3). Sensitivity studies are included that address doubling this rate every ten years and every two years.

" In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching the outside atmosphere, given that a liner flaw exists, was estimated as 1.1% for the cylinder and dome, and 0.11% (10% of the cylinder failure probability) for the base-mat. These values were determined from an assessment of the probability versus containment pressure. For CCNPP, the ILRT maximum pressure is psig 50 [References 27 and 28]

and ultimate pressure of 132 psig [References 32 and 33]. Probabilities of 1% for the cylinder and dome, and 0.1% for the base-mat are used in this analysis, and sensitivity studies are included in Section 5.3.2 (See Table 5-7, Step 4).

" Consistent with the Calvert Cliffs analysis, the likelihood of leakage escape (due to crack Revision 4 Page 16 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension formation) in the base-mat region is considered to be less likely than the containment cylinder and dome region (See Table 5-7, Step 4).

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

To date, all liner corrosion events have been detected through visual inspection (See Table 5-7, Step 5).

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

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 Success data: based on 70 steel-lined containments and 5.5 years since the 10CFR 50.55a requirements of periodic visual inspections of containment surfaces Year Failure rate Year Failure rate Aged adjusted liner flaw likelihood During the 15-year interval, assume 1 2.05E03 1 5.13E04 failure rate doubles every five years average 5-10 5.19E-03 average 5-10 1.30E-03 (14.9% increase per year). The 15 1.43E-02 15 3.57E-03 average for the 5th to 10th year set to the historical failure rate. 15 year average = 6.44E-03 15 year average = 1.61 E-03 Increase in flaw likelihood between 3 and 15 years Uses aged adjusted 0.73% (1 to 3 years) 0.18% (1 to 3 years) 3 liner flaw likelihood (Step 2), 4.18% (1 to 10 years) 1.04% (1 to 10 years) assuming failure rate doubles every 9.66% (1 to 15 years) 2.41% (I to 15 years) five years.

4 Likelihood of breach in containment 1% 0.1%

given liner flaw 10%

5% failure to identify visual flaws plus 5% likelihood that the flaw is Visual inspection detection failure not visible (not through-cylinder 100%

likelihood but culd be detected by ILRT). Cannot be visually inspected All events have been detected through visual inspection. 5%

visible failure detection is a conservative assumption.

0.00073% (3 years) 0.000180% (3 years) 0.73% x 1% x 10% 0.18% x 0.1% x 100%

Likelihood of non-detected 0.00418% (10 years) 0.00104% (10 years) 6 containment leakage (Steps 3 x 4 x 4.18% x1% x10% 1.04% x 0.1% x 100%

5) 0.00966% (15 years) 0.00241% (15 years) 9.66% x 1% x 10% 2.41% x 0.1% x 100%

Revision 4 Page 17 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension The total likelihood of the corrosion-induced, non-detected containment leakage is the sum of Step 6 for the containment cylinder and dome, and the containment base-mat, as summarized below for CCNPP.

Table 5 Total Likelihood on Non-Detected Containment Leakage Due to Corrosion for CCNPP Description At 3 years: 0.00073% + 0.000180% = 0.00091%

At 10 years: 0.00418% + 0.00104% = 0.00522%

At 15 years: 0.00966% + 0.00241% = 0.01207%

The above factors are applied to those core damage accidents that are not already independently LERF or that could never result in LERF.

The two corrosion events that were initiated from the non-visible (backside) portion of the containment liner used to estimate the liner flaw probability in the Calvert Cliffs analysis are assumed to be applicable to this containment analysis. These events, one at North Anna Unit 2 (September 1999) caused by a timber embedded in the concrete immediately behind the containment liner, and one at Brunswick Unit 2 (April 1999) caused by a cloth work glove embedded in the concrete next to the liner, were initiated from the nonvisible (backside) portion of the containment liner. A search of the NRG website LER database identified two additional events have occurred since the Calvert Cliffs analysis 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 containment liner. The other event occurred in April 2009, where a through-liner hole approximately 3/8-inch by 1-inch in size was identified in the Beaver Valley Power Station Unit 1 (BVPS-1) containment 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 liner. Two other containment liner through wall hole events occurred at Turkey Point 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 designed for periodic immersion service, and are not considered to be applicable to this analysis. More recently, in October 2013, some through-wall containment liner holes were identified at BVPS-1, with a combined total area of approximately 0.395 square inches. The cause of these 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 original construction and was contacting the steel liner. For risk evaluation purposes, these five 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 flaw 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.

5.2 Analysis The application of the approach based on the guidance contained in EPRI Report No. 1009325, Revision 2-A, Appendix H [Reference 24], EPRI TR-104285 [Reference 2] 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 5-6.

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

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I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension

" Core damage sequences in which the containment remains intact initially and in the long term (EPRI TR-104285, Class 1 sequences [Reference 2]).

" Core damage sequences in which containment integrity is impaired due to random isolation failures of plant components other than those associated with Type B or Type C test components. For example, liner breach or bellow leakage (EPRI TR-104285, Class 3 sequences [Reference 2]).

" Accident sequences involving containment bypassed (EPRI TR-1 04285, Class 8 sequences [Reference 2]), large containment isolation failures (EPRI TR-104285, Class 2 sequences [Reference 2]), and small containment isolation "failure-to-seal" events (EPRI TR-1 04285, Class 4 and 5 sequences [Reference 2]) are accounted for in this evaluation as part of the baseline risk profile. However, they are not affected by the ILRT frequency change.

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

Table 5 EPRI Accident Class Definitions Accident Classes Description (Containment Release Type) 1 No Containment Failure 2 Large Isolation Failures (Failure to Close) 3a Small Isolation Failures (Liner Breach) 3b Large Isolation Failures (Liner Breach) 4 Small Isolation Failures (Failure to Seal - Type B) 5 Small Isolation Failures (Failure to Seal - Type C) 6 Other Isolation Failures (e.g., Dependent Failures) 7 Failures Induced by Phenomena (Early and Late) 8 Bypass (Interfacing System LOCA)

CDF All CET End States (Including Very Low and No Release)

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

Step 1 - Quantify the baseline risk in terms of frequency per reactor year for each of the accident classes presented in Table 5-9.

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

Step 3 - Evaluate risk impact of extending Type A test interval from 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 Large Early Release Frequency (LERF) in accordance with RG 1.174 [Reference 4].

Step 5 - Determine the impact on the Conditional Containment 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 isolation failures, Type B or Type C testing, or Revision 4 Page 19 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension containment failure induced by severe accident phenomena.

For the assessment of ILRT impacts on the risk profile, the potential for pre-existing leaks is included in the model. (These events are represented by the Class 3 sequences in EPRI TR-104285 [Reference 2].) The question on containment integrity was modified to include the probability of a liner breach or bellows failure (due to excessive leakage) at the time of core damage. Two failure modes were considered for the Class 3 sequences. These are Class 3a (small breach) and Class 3b (large breach).

The frequencies for the severe accident classes defined in Table 5-9 were developed for CCNPP by first determining the frequencies for Classes 1, 2, 6, 7, and 8. Table 5-10 provides a correlation of the adjusted release category frequencies and the EPRI release classes in Table 5-9. Table 5-10 provides the CCNPP-specific frequencies for each Level 2 release category.

Table 5-11 presents the grouping of each endstate in EPRI Classes based on the associated description. Table 5-12 presents the LERF sequence description, frequency and EPRI category for each sequence and the totals of each EPRI classification. Table 5-13 provides a summary of the accident sequence frequencies that can lead to radionuclide release to the public and have been derived consistent with the definitions of accident classes defined in EPRI TR-1 04285

[Reference 2], the NEI Interim Guidance [Reference 3], 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 account for the impact of undetected corrosion of the steel liner per the methodology described in Section 5.1.4. Note: calculations were performed with more digits than shown in this section. Therefore, minor differences may occur if the calculations in this sections are followed explicitly.

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 probability of leakage detectable by a Type A ILRT is calculated to determine the impact of extending the testing interval. The Class 3 calculation is divided into two classes: Class 3a is defined as a small liner breach (La < leakage

< 1OLa), and Class 3b is defined as a large liner breach (1OLa < leakage < 100L.).

Data reported in EPRI 1009325, Revision 2-A [Reference 24] states that two events could have been detected only during the performance of an ILRT and thus impact risk due to change in ILRT frequency. There were 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:

2 Pclass3a = - = 0.0092 21.7 Multiplying the CDF by the probability of a Class 3a leak yields the Class 3a frequency contribution in accordance with guidance provided in Reference 24. As described in Section 5.1.3, additional consideration is made to not apply failure probabilities on those cases that are already LERF scenarios (i.e., the Class 2 and Class 8 contributions). Therefore, these LERF contributions from CDF are removed. Therefore, the frequency of a Class 3a failure is calculated by the following equation:

Frequiciass3a = Pclass3 a * (CDFul - Class2u, - Class8ul) = --L *(1.61E5 - 5.01E8 - 6.77E-7) 217 Frequilcass3a= 1.42E-7 Frequ2 class3a = Pclass3 a * (CDFU2 - Class2u2 - Class8u2 ) = 2 *(1.41E-5 - 4.29E-8 - 6.72E-7)

Frequ2ctass3a= 1.23E-7 Page 20 of 110 Revision 44 Page 20 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension In the database of 217 ILRTs, there are zero containment leakage events that could result in a large early release. Therefore, the Jeffreys Non-Informed Prior is used to estimate a failure rate and is illustrated in the following equations:

Number of Failures+ 1/2 Jeffreys Failure Probability Number of Test + 1/

Number of Tests + 1 0 + 1/2 Pciass3b = 217 + 1 = 0.0023 The frequency of a Class 3b failure is calculated by the following equation:

Frequjc1 ass3 b = Pctass3b * (CDFui - Class2ul - Class8ui) = 2- .(1.61E-5 - 5.01E 6.77E-7)

Frequlclass3b = 3.52E-8 FreqU 2 cass3b = Pclass3b * (CDFu2 - Class2u2 - Class8u2 ) = 'S *(1.41E 4.34E 6.72E-7) 218 Frequ2ctass3b = 3.07E-8 For this analysis, the associated containment leakage for Class 3a is 10L, and for Class 3b is 100La. These assignments are consistent with the guidance 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 Technical Specification Leakage). The frequency per year is initially determined from the EPRI Accident Class 1 frequency listed in Table 5-12 and then subtracting the EPRI Class 3a and 3b frequency (to preserve total CDF), calculated below:

Frequclcass1=Frequ1 class1 3 a - Frequiclass3b)

- (Frequlclass Frequ2class1 = Frequ 2 ctass1 - (Frequclass3 a - Frequ2class3b)

Class 2 Sequences. This group consists of core damage accident progression bins with large containment isolation failures. The frequency per year for these sequences is obtained from the EPRI Accident Class 2 frequency listed in Table 5-12.

Class 4 Sequences. This group consists of all core damage accident progression bins for which containment isolation failure-to-seal of Type B test components occurs. Because these failures are detected by Type B tests which are unaffected by the Type A ILRT, this group is not evaluated any further in the analysis, consistent with 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 test components. Because the failures are detected by Type C tests which are unaffected by the Type A ILRT, this group is not evaluated any further in this analysis, 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 containment isolation valves following a test/maintenance evolution. For CCNPP, this class is defined as the SERF category. The frequency per year for these sequences is obtained from the EPRI Accident Class 6 frequency listed in Table 5-12.

Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs (e.g., overpressure). For this analysis, the frequency is determined from the EPRI Accident Class 7 frequency listed in Table 5-12.

Revision 4 Page 21 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Class 8 Sequences. This group consists of all core damage accident progression bins in which containment bypass or SGTR occurs. For this analysis, the frequency is determined from the EPRI Accident Class 8 frequency listed in Table 5-12.

Table 5 Release Category Frequencies Release Category EPRI Category Unit I Frequency (lyr) Unit 2 Frequency (lyr)

INTACT Class 1 6.76E-06 5.12E-06 LERF Classes 2,7, 8 1.39E-06 1.56E-06 SERF Class 6 1.87E-06 1.25E-06 LATE Class I 6.06E-06 6.17E-06 1 Total (CDF) N/A 1.61 E-05 1.41 E-05

1. Unit 2 LATE was quantified at 5E-12 truncation. The other end states were quantified at 1E-12.

LERF quantification is distributed into EPRI categories based on end states. Table 5-11 shows this distribution.

Table 5 Release Category Frequencies CCNPP LERF Description of Outcome EPRI Unit 1 Unit 2 End State Category Frequency (/yr) Frequency (/yr)

LERF01 Containment failure following high- 7 3.66E-07 3.80E-07 pressure (HP) vessel breach (VB)

LERF02 Containment failure following HP VB 7 4.87E-10 5.45E-08 LERF03 Containment failure following low pressure 7 4.25E-1 1 1.89E-07 (LP) VB LERF04 Temperature induced (TI) SGTR 8 0.OOE+00 3.17E-09 LERF05 Containment failure following LP VB 7 5.04E-08 9.61 E-08 LERF06 Pressure induced (PI) SGTR 8 O.OOE+00 0.OOE+00 LERF07 Containment failure following LP VB 7 1.09E-08 9.68E-09 LERF08 Loss of isolation 2 3.34E-08 3.72E-08 LERF09 Containment bypass 8 6.68E-07 6.56E-07 LERF10 Containment failure following LP VB 7 1.37E-07 5.59E-08 LERF11 Containment failure following HP VB 7 2.01 E-08 1.40E-08 LERF12 Containment failure following LP VB 7 5.14E-08 2.85E-08 LERF13 TI-SGTR 8 8.75E-09 1.21 E-08 LERF14 Containment failure following LP VB 7 1.27E-08 7.63E-09 LERF15 PI-SGTR 8 O.OOE+00 0.OOE+00 LERF16 Containment failure following LP VB 7 0.OOE+00 0.OOE+00 LERF17 Loss of isolation 2 3.05E-08 1.71 E-08 LERF18 Containment bvoass 8 4.94E-10 5.23E-10

.. .. II

.. Containment bypass Contribution to EPRI Classification 2 6.39E-08 5.43E-08 Contribution to EPRI Classification 7 6.49E-07 8.35E-07 Contribution to EPRI Classification 8 6.77E-07 6.72E-07 Total LERF 1.39E-06 1.56E-06 Page 22 of 110 Revision 44 Revision Page 22 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Release Category Frequencies Release Category EPRI Category Unit I Frequency (Iyr) Unit 2 Frequency (Iyr)

INTACT + LATE 1 Class 1 1.28E-05 6 1.13E-05 6 LERF 2 Class 2 5.01E-08 6 4.34E-086 SERF 3 Class 6 1.87E-06 1.25E-06 4

LERF Class 7 6.49E-07 8.35E-07 5

LERF Class 8 6.77E-07 6.72E-07 Total (CDF) 1.61E-5 1.41E-5

1. The EPRI Class 1 category consists of INTACT and LATE failures. A LATE failure is classified as intact due to the long time until failure and is consistent with guidance in Reference 24.
2. The EPRI Class 2 category consists of CCNPP assigned LERF contribution associated with isolation failures as re-categorized in Table 5-11 with pre-event containment liner failure removed (see note 6).
3. The EPRI Class 6 category consists of CCNPP assigned scrubbed isolation failures in SERF.
4. The EPRI Class 7 category consists of the CCNPP assigned LERF contribution associated with phenomenological failures as re-categorized in Table 5-11.
5. The EPRI Class 8 category consists of the CCNPP assigned LERF contribution associated with bypass or SGTR failures as re-categorized in Table 5-11.
6. The level 2 model contains a bounding contribution associated with pre-event containment liner failure. To preclude influencing the current detailed assessment, the contribution associated with this failure is adjusted by removal of the bounding estimate from Class 2 and adding it to the intact containment case (Class 1).

The Unit 1 pre-event containment liner failure value is 1.385E-8; the Unit 2 value is 1.094E-8. These values are the LERF contributions from events FAILLEAK and FAILLEAK_2 for Units 1 and 2, respectively.

Table 5 Baseline Risk Profile Class Description Unit I Frequency (/yr) Unit 2 Frequency (/yr) 1 No containment failure 1.27E-05 2 1.12E-05 2 2 Large containment isolation failures 5.01E-08 4.34E-08 3a Small isolation failures (liner breach) 1.42E-07 1.23E-07 3b Large isolation failures (liner breach) 3.52E-08 3.07E-08 4 Small isolation failures - failure to seal (type B) E1 E1 5 Small isolation failures - failure to seal (type C) El El 6 Containment isolation failures (dependent failure, 1.87E-06 1.25E-06 personnel errors) 7 Severe accident phenomena induced failure (early 6.49E-07 8.35E-07 and late) 8 Containment bypass 6.77E-07 6.72E-07 Total 1.61E-05 1.41E-05

1. E represents a probabilistically insignificant value.
2. The Class 3a and 3b frequencies are subtracted from Class 1 to preserve total CDF.

5.2.2 Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose)

Plant-specific release analyses were performed to estimate the person-rem doses to the population within a 50-mile radius from the plant. The releases are based on CCNPP-specific dose calculations summarized on Table 5-5. Table 5-14 provides a correlation of CCNPP population dose to EPRI Accident Class. Table 5-15 provides population dose for each EPRI accident class.

Revision 4 Page 23 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension The population dose for EPRI Accident Classes 3a and 3b were calculated based on the guidance provided in EPRI Report No. 1009325, Revision 2-A [Reference 24] as follows:

EPRI Class 3a PopulationDose = 10

  • 3.20E+4 = 3.20E+5 EPRI Class 3b PopulationDose = 100
  • 3.20E+4 = 3.20E+6 Table 5 Mapping of Population Dose to EPRI Accident Class Release EPRI Unit I Frequency Unit I Dose Unit 2 Frequency Unit 2 Dose Category Category (/yr) (person-rem) (/yr) (person-rem)

INTACT + Class 1 1.28E-05 3.20E+04 1.13E-05 3.20E+04 LATE LERF Class 2 5.01E-08 2.OOE+07 4.34E-08 2.OOE+07 SERF Class 6 1.87E-06 7.01E+06 1.25E-06 7.01 E+06 LERF Class 7 6.49E-07 5.61E+07 8.35E-07 5.61 E+07 LERF Class 8 6.77E-07 2.25E+07 6.72E-07 2.25E+07 Table 5 Baseline Population Doses Class Description Population Dose (person-rem) 1 No containment failure 3.20E+04 2 Large containment isolation failures 2.OOE+07 3a Small isolation failures (liner breach) 3.20E+05 1 2

3b Large isolation failures (liner breach) 3.20E+06 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) 7.01 E+06 7 Severe accident phenomena induced failure (early and late) 5.61 E+07 8 Containment bypass 2.25E+07 0*La

1. 10
2. 1i00*La 5.2.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval from 10 to 15 Years The next step is to evaluate the risk impact of extending the test interval from its current 10-year interval to a 15-year interval. To do this, an evaluation must first be made of the risk associated with the 10-year interval, since the base case applies to 3-year interval (i.e., a simplified representation of a 3-to-10 interval).

Risk Impact Due to 10-Year Test Interval As previously stated, Type A tests impact only Class 3 sequences. For Class 3 sequences, the release magnitude is not impacted by the change in test interval (a small or large breach remains the same, even though the probability of not detecting the breach increases). Thus, only the frequency of Class 3a and Class 3b sequences is impacted. The risk contribution is Revision 4 Page 24 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension changed based on the NEI guidance as described in Section 5.1.3 by a factor of 10/3 compared to the base case values. The Class 3a and 3b frequencies are calculated as follows:

10 21 Frequlclass3aloyr =13

  • 217 2 * (CDFul - Class2u, - Class8 1 u) = 3
  • 2172
  • 1.54E5 = 4.72E-7 10 21 Frequ2class3aloyr= * - * (CDFU2 - Class2u2 - Class8u2 ) = 0
  • 2
  • 1.34E-5 = 4.11E7 3 217 3 217 10 .5 10 5 ulclass3blOyr 10 **

3 210

  • (CDFul - Class2ul - Class8uj) = 10 3
  • 2 218
  • 1.54E-5 = 1.17E-7 Frequzclass3blOyr =1 * -- * (CDFu2 - Class2u2 - ClassAu 2 ) =-- * -
  • 1.34E-5 = 1.02E-7 3 218 3 218 The results of the calculation for a 10-year interval for Units 1 and 2 interval are presented in Tables 5-16 and 5-17, respectively.

Risk Impact Due to 15-Year Test Interval The risk contribution for a 15-year interval is calculated in a manner similar to the 10-year interval. The difference is in the increase in probability of leakage in Classes 3a and 3b. For this case, the value used in the analysis is a factor of 5 compared to the 3-year interval value, as described in Section 5.1.3. The Class 3a and 3b frequencies are calculated as follows:

Frequiclass 3 al 5 yr = 15

  • 2 3 217
  • (CDFul - Class2u, - Class8ul) = 5
  • 2 217 1.54E5 = 7.08E7 15 22 Frequzczass3al5yr =

3 217

- * (CDFu2 - Class2u2 - Class8u2) = 5 * -

217

  • 1.34E-5 = 6.17E-7 Frequlclass3blSyr=
  • 21 * (CDFU

( Classu l- assu ) =5* .S4E- = 1.76E7 FreqU-ass3bl5yr= -i 3

218 (CDFU2 - Class2U2 - ClassU 2 ) = 5

  • 218 5
  • 1.34E-5 = 1.53E-7 The results of the calculation for a 15-year interval for Units 1 and 2 are presented in Table 5-18 and 5-19.

Table 5 Unit I Risk Profile for Once in 10 Year ILRT Class Description Frequency Contribution (%) Population Population (lyr) Dose Dose Rate (person- (person-rem) rem/yr) 1 No containment failure1 1.23E-05 76.17% 3.20E+04 3.92E-01 2 Large containment isolation failures 5.01 E-08 0.31% 2.OOE+07 1.OOE+00 3a Small isolation failures (liner 4.72E-07 2.93% 3.20E+05 1.51 E-01 breach) 3b Large isolation failures (liner 1 .17E-07 0.73% 3.20E+06 3.76E-01 breach) 4 Small isolation failures - failure to El El El El seal (type B) 5 Small isolation failures - failure to F1 El El El seal (type C)

Containment isolation failures 6 (dependent failure, personnel 1.87E-06 11.61% 7.01E+06 1.31 E+01 errors)

Revision 4 Page 25 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Unit I Risk Profile for Once in 10 Year ILRT Class Description Frequency Contribution (%) Population Population

(/yr) Dose Dose Rate (person- (person-rem) rem/yr) 7 Severe accident phenomena 6.49E-07 4.04% 5.61 E+07 3.64E+01 induced failure (early and late) 8 Containment bypass 6.77E-07 4.21% 2.25E+07 1.52E+01 Total 1.61 E-05 6.67E+01

1. E represents a probabilistically insignificant value.
2. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.

Table 5 Unit 2 Risk Profile for Once in 10 Year ILRT Class Description Frequency Contribution (%) Population Population

(/yr) Dose Dose Rate (person- (person-rem) rem/yr) 1 No containment failure1 1.08E-05 76.51% 3.20E+04 3.45E-01 2 Large containment isolation failures 4.34E-08 0.31% 2.OOE+07 8.67E-01 3a Small isolation failures (liner breach) 4.11E-07 2.92% 3.20E+05 1.32E-01 3b Large isolation failures (liner breach) 1.02E-07 0.73% 3.20E+06 3.28E-01 Small isolation failures - failure to seal (type B)

Small isolation failures - failure to seal (type C)

Containment isolation failures 6 Cotimn slto alrs 1.25E-06 8.85% 7.01 E+06 8.75E+00 (dependent failure, personnel errors) 7 Severe accident phenomena 8.35E-07 5.92% 5.61E+07 4.69E+01 induced failure (early and late) 8 Containment bypass 6.72E-07 4.76% 2.25E+07 1.51E+01 Total 1.41 E-05 7.24E+01

1. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.

Table 5 Unit I Risk Profile for Once in 15 Year ILRT Class Description Frequency (/yr) Contribution (%) Population Population Dose Dose Rate (person- (person-rem) rem/yr) 1 No containment failure' 1.20E-05 74.34% 3.20E+04 3.83E-01 2 Large containment isolation 5.01 E-08 0.31% 2.OOE+07 1.OOE+00 failures 3a Small isolation failures (liner 7.08E-07 4.40% 3.20E+05 2.26E-01 breach) 3b Large isolation failures (liner 1.76E-07 1.09% 3.20E+06 5.64E-01 breach)

Revision 4 Page 26 of 110

I RCA-54001 -000-CALC-001 . Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Unit I Risk Profile for Once in 15 Year ILRT Class Description Frequency (/yr) Contribution (%) Population Population Dose Dose Rate (person- (person-rem) rem/yr) 4 Small isolation failures - failure El El El El to seal (type B) 5 Small isolation failures - failure El El E El to seal (type C)

Containment isolation failures 6 (dependent failure, personnel 1.87E-06 11.61% 7.01E+06 1.31 E+01 errors)

Severe accident phenomena 6.49E-07 4.04% 5.61E+07 3.64E+01 induced failure (early and late) 8 Containment bypass 6.77E-07 4.21% 2.25E+07 1.52E+01 Total 1.61 E-05 6.69E+01

1. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.

Table 5 Unit 2 Risk Profile for Once in 15 Year ILRT Class Description Frequency (/yr) Contribution (%) Population Population Dose Dose Rate (person- (person-rem) rem/yr) 1 No containment failure' 1.05E-05 74.69% 3.20E+04 3.37E-01 2 Large containment isolation 4.34E-08 0.31% 2.00E+07 8.67E-01 failures 3a Small isolation failures (liner 6.17E-07 4.37% 3.20E+05 1.97E-01 breach) 3b Large isolation failures (liner 1.54E-07 1.09% 3.20E+06 4.91 E-01 breach) 4 Small isolation failures - failure El El El El to seal (type B) 5 Small isolation failures - failure El El El El to seal (type C)

Containment isolation failures 6 (dependent failure, personnel 1.25E-06 8.85% 7.01 E+06 8.75E+00 errors) 7 Severe accident phenomena 8.35E-07 5.92% 5.61 E+07 4.69E+01 induced failure (early and late) 8 Containment bypass 6.72E-07 4.76% 2.25E+07 1.51 E+01 Total 1.41 E-05 7.26E+01

1. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.

5.2.4 Step 4 - Determine the Change in Risk in Terms of LERF The risk increase associated with extending the ILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from an intact containment could, in fact, result in a larger release due to the increase in probability of failure to Revision 4 Page 27 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension detect a pre-existing leak. With strict adherence to the EPRI guidance, 100% of the Class 3b contribution would be considered LERF.

Regulatory Guide 1.174 [Reference 4] provides guidance 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 less than 10-6/year and increases in LERF less than 10-7/year, and small changes in LERF as less than 10-6/year. Since containment overpressure is not required in support of ECCS performance to mitigate design basis accidents at CCNPP, the ILRT extension does not impact CDF. Therefore, the relevant risk-impact metric is LERF.

For CCNPP, 100% of the frequency of Class 3b sequences can be used as a very conservative first-order estimate to approximate the potential increase in LERF from the ILRT interval extension (consistent with the EPRI guidance methodology). Based on a 10-year test interval from Tables 5-16 and 5-17, the Class 3b frequency is 1.1 7E-7/year for Unit 1 and 1.02E-7 for Unit 2; based on a 15-year test interval from Tables 5-18 and 5-19, the Class 3b frequency is 1.76E-7 for Unit 1 and 1.54E-7 for Unit 2. 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 1.41 E-7/year for Unit 1 and 1.23E-7 for Unit 2. Similarly, the increase due to increasing the interval from 10 to 15 years is 5.87E-8/year for Unit 1 and 5.12E-8 for Unit 2. As can be seen, even with the conservatisms included in the evaluation (per the EPRI methodology), the estimated change in LERF is below the threshold criteria for a small change when comparing the 15-year results to the current 1 0-year requirement, and slightly greater than the criteria when compared to the original 3-year requirement. Table 5-20 summarizes these results.

Table 5 Impact on LERF due to Extended Type A Testing Intervals ILRT Inspection Unit 1:3 Unit 1:10 Unit 1:15 Unit 2:3 Unit 2:10 Unit 2:15 Interval Years Years Years Years Years Years (baseline) I (baseline)

Class 3b (Type A 3.52E-08 1.17E-07 1.76E-07 3.07E-08 1.02E-07 1.54E-07 LERF) basEF( e aine)8.22E-08 1.41 E-07 7.16E-08 1.23E-07 ALERF (10 year * ]5.87E-08 5.12E-08 baseline)

The increase in the overall probability of LERF due to Class 3b sequences being slightly greater than 1E-7 is not unexpected. Since the target is exceeded, some refinement is necessary. One method to remove some conservatism is to examine the source term expected to be available for release during the accident sequence. The source term is greatly reduced if the debris expelled from the reactor remains covered with water. Therefore, if the accident sequence contains containment spray success, the source term is not considered to lead to a large early release. The methodology developed in Reference 33 is used for this containment spray success sensitivity. Excluding INTACT scenarios where containment spray is credited and therefore scrubbing the source term release results in a frequency reduction.

Conservatisms are further reduced by analyzing the source term release times. Early release timing is defined by time short enough that ability to evacuate nearby population is impaired such that a fatality is possible. Reference 36 contains the development of Evacuation Time Estimates for the Calvert Cliffs Nuclear Power Plant. Review of this report shows that the average evacuation time is estimated to be 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Therefore, for this assessment, an early Page 28 of 110 Revision 44 Page 28 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension release is defined as occurring before 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. By reviewing CCNPP's MAAP runs in the Level 2 severe accident report [Reference 18], it was determined three cases had source terms released after the 6.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> mark. The first case is HRIF, which simulates a loss of main feedwater due to a station blackout (SBO). The last two cases, GIOY and MRIF, evaluate small LOCAs inside containment. These three MAAP cases are matched with a corresponding plant damage state (PDS) in CCNPP's Level 2 notebook [Reference 32]. Table 5-21 displays CCNPP's PDSs.

Table 5 Summary of CCNPP Plant Damage States PDS Containment RCS Pressure at Feedwater Pressurizer CHR? AC Power Bypass? Time of Core Availability? PORV/SRV Status? Available?

Damage?

ot NNot 1 No High available Not stuck open Available Available Not 4 No Low Available Not stuck open Available Available Not 5 No High Available Not stuck open Available Available 6 No Low Available Not stuck open Available Available 7 SGTR N/A N/A N/A N/A N/A 8 ISLOCA N/A N/A N/A N/A N/A Not Not 9 No High Available Not stuck open Available Available 10 NoHighNot 10 No High Available Not stuck open Available Available Not Not Not 14 No High Available Stuck open Available Available Not Not Not Available 15 No High Available Not stuck open Available 16 No High Available Stuck open Not Available Not Available Not Not 17 No Low Available Not Stuck Open Available Available The HRIF MAAP case models a SBO that leads to a loss of main feedwater. The analysis assumes a loss of containment heat removal and AC power. The reactor coolant system is isolated and the containment remains intact. Core damage occurs while the reactor coolant system is at high pressure. Based on the information in Table 5-21, this case can be used to represent PDS 15. Table 2-2 of Reference 32 contains the list of all the Level I core damage Revision 4 Page 29 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension accident sequences and how each is mapped to a PDS. Using the correlation of the HRIF case and the SBO cases that contain a loss of feedwater (PDS 15), it is determined that the frequency contribution can be removed from the LERF contribution because the release occurs at approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, which is greater than 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The impacted sequences are SBOO04, SBO005, SBOO10, SBOO13, SBO015, SBOO18, SBO019, and SB0039.

Another MAAP case evaluated is GIOY, which involves a Small LOCA inside containment with an equivalent break size of 0.005 ft 2 and the containment isolated. The reactor coolant system is at high pressure with auxiliary feedwater (AFW) and AC power available; containment air cooling (CAC) provides containment cooling and maintains containment pressure. The radionuclide release occurs at approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> [Reference 18], which is greater than 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Therefore, this case can be used to represent PDS 5. The following sequences are Small LOCA cases assigned to PDS 5 and are excluded based on their late release: SLOCA002, SLOCA003, and SLOCA012.

Another MAAP case evaluated is MRIF, which involves a Small LOCA inside containment with an equivalent break size of 0.02 ft2. Containment is isolated; the reactor coolant pressure is high; AFW and containment heat removal are not available; AC power is available. These characteristics map to PDS 1. The release occurs at approximately 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> [Reference 181 which is greater than 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The following sequences are Small LOCA cases assigned to PDS 5 and are removed from the PDS 1 frequency: TRAN003, TRAN004, TRAN005, TRAN007, TRAN008, TRAN009, SLOCA007, and SLOCA011 [Reference 18].

In addition to the late release timing associated with each of these MAAP cases, the release fractions of the iodine group are less than 1 E-03 in all cases. Typically, a release fraction of 10% of iodine is used to represent a large release and thus none of the releases associated with these cases can be considered large. For ease of review, the release timing and fractions from Reference 18 are contained in Attachment 2.

The exclusion of these frequencies yields new Level 2 results. Table 5-22 shows adjusted release category frequencies after some conservatisms from containment spray success and release timing are excluded.

Table 5 Adjusted Release Category Frequencies Release Category EPRI Category Unit 1 Frequency (/yr) Unit 2 Frequency (/yr)

INTACT Class 1 3.28E-06 1.18E-06 LERF Classes 2, 7, 8 9.51E-07 1.08E-06 SERF Class 6 1.54E-06 5.74E-07 LATE Class 1 2.37E-06 2.18E-06 1 Total (CDF) N/A 8.14E-06 5.01E-06

1. Unit 2 LATE was quantified at 5E-12 truncation. The other end states were quantified at 1E-12.

Substituting these values into the previously defined equations and calculation method yields the final results displayed in Table 5-23.

Page 30 of 110 Revision 44 Page 30 of 110

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension I RCA-54001 -O00-CALC-O01 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Impact on LERF due to Extended Type A Testing Intervals with Adjusted CDF ILRT Inspection Unit 1: 3 Unit 1:10 Unit 1:15 Unit 2: 3 Unit 2:10 Unit 2:15 Interval Years Years Years Years Years Years (baseline) (baseline)

Class 3b (Type A LERF) 1.14E-08 3.78E-08 5.68E-08 6.14E-09 2.05E-08 3.07E-08 ALERF (3 year 4.54E-08 2.46E-08 baseline)

ALERF (10 year 1.89E-08 1.02E-08 baseline)

The adjusted containment spray and PDS inputs allow the Unit 1 and 2 values to be much less than the 1E-7 LERF metric. The delta LERF between the 3 years and the 15 years is 4.54E-8/yr for Unit 1 and 2.46E-8/yr for Unit 2. These values show that the proposed extension meets the definition of a very small change in risk as defined in Regulatory Guide 1.174.

5.2.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability (CCFP)

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 the conditional containment failure probability (CCFP). The CCFP is defined as the probability of containment failure given the occurrence of an accident. This probability can be expressed using the following equation:

CCFP = 1 - f(ncf)

CDF where f(ncf) is the frequency of those sequences that do not result in containment failure; this frequency is determined by summing the Class 1 and Class 3a results.

Since CCFP is only concerned with a containment failure and not whether the release is small or large, the Class 1 results without refinement must be used to calculate the CCFP. Table 5-24 shows the steps and results of this calculation. The difference in CCFP between the 3-year test interval and 15-year test interval is 0.88% for Unit 1 and 0.87% for Unit 2.

Table 5 Impact on CCFP due to Extended Type A Testing Intervals ILRT Inspection Unit 1:3 Unit 1:10 Unit 1:15 Unit 2:3 Unit 2:10 Unit 2:15 Interval Years Years Years Years Years Years (baseline) (baseline) f(ncf) (/yr) 1.28E-05 1.27E-05 1.27E-05 1.13E-05 1.12E-05 1.12E-05 f(ncf)/CDF 0.796 0.791 0.787 0.799 0.794 0.791 CCFP 0.204 0.209 0.213 0.201 0.206 0.209 r.

ACCFP (3 year baseline) 0.876% 0.871%

ACCFP (10 year 0.365% 0.363%

baseline)

As stated in Section 2.0, a change in the CCFP of up to 1.5% is assumed to be small. The increase in the CCFP from the 3 in 10 year interval to I in 15 year interval is 0.876% for Unit I and 0.871% for Unit 2. Therefore, this increase is judged to be very small.

Revision 4 Page 31 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 5.3 Sensitivities 5.3.1 Potential Impact from External Events Contribution An assessment of the impact of external events is performed. The primary basis for this investigation is the determination of the total LERF following an increase in the ILRT testing interval from 3 in 10 years to 1 in 15 years.

Calvert Cliffs is transitioning to NFPA 805 licensing basis for fire protection and submitted a License Amendment Request (LAR) on September 24, 2013 (ADAMS Accession No. ML13301A673). This transition included performing a Fire PRA and committing to modifications to reduce the fire-induced core damage and large early release frequencies to those reported in the NFPA 805 LAR. Compensatory actions have been implemented to reduce the fire risk until the modifications are implemented. The Unit 1 ILRT is scheduled for 2016, which is prior to the scheduled implementation of all the modifications by 2018. It is anticipated that many, but not all, of the NFPA 805 modifications will be completed by the Unit 1 refueling outage. Risk mitigation strategies will be in place for any open modification. These strategies may be actions to reduce fire initiating event probabilities, actions to improve suppression probability, and/or actions to recover or protect systems that mitigate core damage and large early release accident sequences. The Unit 2 ILRT is scheduled for 2023, so the NFPA 805 modifications will be implemented prior to the extension. The section evaluates the fire risk using the Fire PRA.

Section 5.3.1.1 uses the IPEEE fire.risk values to evaluate fire risk.

The Fire PRA model 6.1M was used to obtain the fire CDF and LERF values. To reduce conservatism in the model, the plant damage state methodology described in Section 5.2.4 was also applied to the CDF portion of the Fire PRA model. The following shows the calculation for Class 3b for Units 1 and 2:

0.5 Frequctass3b = Pclass3b * (CDF1 - PDScDFl) = -218 * (3.18E-05 - 2.20E-5) = 2.25E-8 0.5 Frequ2 ciass3 b Pclass3b * (CDF2 - PDScDF2 ) = * (3.62E05 - 2.26E5) 3.12E8 10 10 *0.5 Frequlclass3blOyr=

  • Pciass3b * (CDF1 - PDScDFI) = 3:' * (3.18E05 - 2.20E5) = 7.49E8 10 10 05 *(3.62E 2.26E-5) = 1.04E-07 Frequ 2 cl as s3 blOyr = -
  • Pclass3b * (CDF2 - PDScDF2 ) = -*- *1-:"

i * &las3b * (CDF1 - PDSCDFl) =.= 5 * * (3.18E 2.20E-5) = 1.12E-07 s3 15 218 0

Frequ2class3blSyr = 7

  • Pciass3b * (CDF2 - PDScDF2 ) = 5
  • o.-

218

  • (3.62E 2.26E-5) = 1.56E-07 Seismic events were addressed through a simplified seismic PRA in Section 3 of the IPEEE for CCNPP [Reference 35]. The Seismic PRA method screened all the components that met a high confidence low probability of failure (HCLPF) for the review level seismic event occurring with a magnitude of 0.3g. The remaining components were grouped together as a proxy component. It was assumed that if this proxy component failed it would result in core damage. This method is considered conservative.

Data from Table 3-6 of the IPEEE Seismic Analysis [Reference 35] is used to calculate a Class 3b frequency due to seismic. As noted in Table 3-6 of the IPEEE Seismic Analysis, the values given in Table 3-6 reflect quantification without the surrogate top event LA. Top event LA represents seismic failure of rugged plant systems at a conservative screening fragility.

Therefore, the total CDF is higher than the 1.07E-05/yr value given in Table 3-6. The CDF Revision 4 Page 32 of 110

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension values given in Section 8.1 of the IPEEE Seismic Analysis [Reference 35] are 1.29E-5/yr for Unit 1 and 1.52E-5/yr for Unit 2. The CDF contribution from surrogate top event LA was not included in the Unit 1 containment failure frequencies provided in the IPEEE (no containment failure frequencies are provided for Unit 2). To conservatively add in the difference between the CDF including surrogate top event LA and the total Unit 1 CDF of 1.07E-5/yr from Table 3-6 of the IPEEE for the Class 3b frequency calculation, it is added to Containment Category I Failure (Intact). Note that the Intact category CDF is slightly rounded so that the total seismic CDF is preserved. Then, the percent each category contributes to the total CDF is calculated forthe Unit 1 values and applied to the Unit 2 values because it is assumed that Unit 2 would have similar containment failure fractions to Unit 1.

Table 5 Seismic Contribution to Frequencies of Containment Failure Categories Containment Failure Unit I Seismic CDF Percent of CDF Unit 2 Seismic CDF (lyr)

Category (/yr)

I. Intact Containment 2.69E-06 20.85% 3.17E-06 I1.Late Containment Failure 8.63E-06 66.90% 1.02E-05 111.Early Small Containment 1.70E-07 to 1.27E-06 1.32% to 9.84% 2.00E-07 to 1.50E-06 Failure IV. Early Large Containment 3.13E-07 to 1.41E-06 2.43% to 10.93% 3.69E-07 to 1.66E-06 Failure V. Small Containment Bypass 0 0% 0 VI. Large Containment Bypass 0 0% 0 Total 1.29E-05 1.52E-05 Note: The Seismic contribution to Containment failure categories III and IV is shown as a range of values. A range is shown because the contribution of a certain PDS will be apportioned between the small and large early containment failures, but the ratio is unknown. Therefore, we show a range of values which reflect the contribution of this PDS from being attributed entirely to early-large containment failures (conservative) to early-small containment failures. See section 3.1.6.1 of the IPEEE Seismic Analysis for a more detailed explanation.

Using this seismic data, the Class 3b frequency can be calculated by the following formulas:

Frequliass3b l=

Plass3b * (CDF - CatIV - CatVI) = 5 218

  • (1.29E 3.13E 0) = 2.89E-8 Frequ2 class3 b = Pclass3b * (CDF - Cat!V - CatV!) = 0.5 * (1.52E 3.69E 0) = 3.40E-8 218 Frequlclass3 blOyr = 10 **Pctass3b * (CDF - Cat!V - CatVI) 10
  • 0.5

= -0 0 * (1.29E-5 - 3.13E-7-0) = 9.62E-8 Frequf2 cjass3 blOyrl = 1010

  • Pclass13b * (CDF - Cat! V - CatVI) = 10 10
  • 0.5
0. * (1.5 2E 3. 69E-7-0) =1.13E-7 33 218 15 0 Frq~ls~ly = 3* Pcasb* (CDF - Cat! V - CatVI) = 3Ž * .21. * (1.29E-5 - 3.13E-7-0) =1.44E-7 15* * (CDF - Cat!V - CatVI) = L- * -' * (1.52E 3. 69E-7-0) = 1.70E-7 Freq2cass3blSyr ="3
  • Pcass3b *D tV3 218 CNNPP topographical location presents the opportunity for high wind events. These events include tornadoes, thunderstorms, freezing precipitation, and hurricanes. Hurricanes pose approximately one threat per year and one significant threat per 10 years (Reference 24). These natural disasters are modeled in the internal events model. As shown in Table 5-2 and 5-4 show that high wind risk is approximately two orders of magnitude lower than fire risk. Since high wind Page 33 of 110 Revision 44 Page 33 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension risk is already included in the internal events PRA, no further analysis is necessary to include its contribution to Class 3b frequency.

The seismic and fire contributions to Class 3b frequencies 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 defined for the external events in Tables 5-26 and 5-27 for Units I and 2, respectively.

Table 5 CCNPP Unit I External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to I 3 per 10 year I per 10 year I per 15 years per 15 years)

External Events 5.13E-08 1.71E-07 2.57E-07 2.05E-07 Internal Events 1.14E-08 3.78E-08 5.68E-08 4.54E-08 Combined 6.27E-08 2.09E-07 3.13E-07 2.51E-07 Table 5 CCNPP Unit 2 External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 3 per 10 year I per 10 year 1 per 15 years per 15 years)

External Events 6.52E-08 2.17E-07 3.26E-07 2.61 E-07 Internal Events 6.14E-09 2.05E-08 3.07E-08 2.46E-08 Combined 7.13E-08 2.38E-07 3.57E-07 2.85E-07 The internal event results are also provided to allow a composite value to be defined. When both the internal and external event contributions are combined the total change in Unit 1 and 2 LERF meet the guidance for small change in risk, as it exceeds the 1.OE-7/yr and remains less than 1.OE-6 change in LERF for both units. For this change in LERF to be acceptable, total LERF must be less than 1.OE-5.

Conservatively using the highest seismic LERF value and not crediting containment spray success or plant damage state adjustments for the Internal Events or Fire PRA, the total LERF values are calculated below:

Unit 1: LERFul = LERFulintemal + LERFuoseismic + LERFulfre + LERFulclass3Bincrease

= 1.39E-6/yr + 1.41E-6/yr + 2.97E-6/yr + 2.51E-07/yr = 6.02E-6/yr Unit 2: LERFu2 = LERFU2internal + LERFu2seismic + LERFutfire + LERFuzclass3Bincrease

= 1.56E-6/yr + 1.66E-6/yr + 4.17E-6/yr + 2.85E-07/yr = 7.68E-6/yr Since the total LERF for both units is less than 1.OE-5, it is acceptable for the ALERF to be between 1.OE-7 and 1.OE-6.

5.3.1.1 Potential Impact from External Events Contribution Using IPEEE Fire Analysis An assessment of the impact of external events is also performed using fire risk analysis from the IPEEE [Reference 35] rather than the Fire PRA model 6.1 M. Table 4.7 from the simplified IPEEE fire PRA shows the frequencies of major containment failure categories for Unit 1 Revision 4 Page 34 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension

[Reference 35]. The same containment failure category percentages are assumed for Unit 2; as given in Section 8.2 of the IPEEE, the estimated Unit 2 fire CDF is 9.6E-5/yr. The Level 2 results are shown here in table 5-28.

Table 5 Fire Contribution to Frequencies of Containment Failure Categories Containment Failure Category Percentage Unit I Fire CDF (/yr) Unit 2 Fire CDF (/yr)

I. Intact Containment 36.4% 2.67E-05 3.50E-05 II. Late Containment Failure 55.5% 4.07E-05 5.33E-05 III. Early Small Containment Failure 1.7% 1.21E-06 1.58E-06 IV. Early Large Containment Failure 6.5% 4.67E-06 6.13E-06 V. Small Containment Bypass 0.0% 0.OOE+00 0.00E+00 VI. Large Containment Bypass 0.0% 0.OOE+00 0.OOE+00 Total 7.32E-05 9.60E-05 Note: the Unit 1 CDF obtained by summing the CDF values given in Table 4.7 of the IPEEE [Reference 35] is 7.32E-05, as cited in this table. This number differs from 7.29E-05 cited in Section 8 of the IPEEE [Reference 35]. The CDF value used in this table and subsequent calculations is conservatively chosen as 7.32E-05. Also, the percentages given in Table 4.7 of the IPEEE [Reference 35] sum to 101.8%. These percentages are adjusted based on the individual values of the Unit 1 Containment Failure Categories to sum to 100.0%.

Using the IPEEE fire data, the Class 3b frequency can be calculated by the following formulas:

Unit 1: Frequl1 1ass3 b = Pc1ass3b * (CDF - CatlV - CatVI) = 218 * (7.32E 4.67E 0)= 1.57E-7 10 Unit 1: Frequlciass3bloyr= 7

  • P10ss~b
  • 0 .5 (CDF - CatIV - CatVI) 10

-- * -- * (7.32E 4.67E 0) = 5.24E-7 3 15 218 Unit 1: Frequlicass3blSyr =-5

  • Pclass3b * (CDF - Cat!V - CatVJ)

= 5 * * (7.32E 4.67E 0)= 7.86E-7 218 Unit 2: Frequ2cass3b = Pdcass3b * (CDF - CatIV - CatVJ) = .- 5

  • 218 (9.60E 6.13E 0)= 2.06E-7 10 Unit 2: Frequ2cjass3bloyr= 10
  • Piass~ i * (CDF - Cat!V - CatV!)

10 0.5

--

  • 0- * (9.60E 6.13E 0)

= 6.87E-7 3 15 218 Unit 2: FreqU2ctass3bl~yr= -i

  • Pclass3b * (CDF - CatlV - CatV!)

0.5*

= 5

  • 0*5 * (9.60E 6.13E 0)= 1.03E-6 218 As done in Section 5.3.1, the IPEEE seismic and fire contributions to Class 3b frequencies 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 defined for the external events in Tables 5-29 and 5-30 for Units I and 2, respectively.

Page 35 of 110 Revision 44 Revision Page 35 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 CCNPP Unit I External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 3 per 10 year I per 10 year I per 15 years per 15 years)

External Events 1.86E-07 6.20E-07 9.30E-07 7.44E-07 Internal Events 1.14E-08 3.78E-08 5.68E-08 4.54E-08 Combined 1.97E-07 6.58E-07 9.87E-07 7.89E-07 Table 5 CCNPP Unit 2 External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 3 per 10 year I per 10 year I per 15 years per 15 years)

External Events 2.40E-07 8.OOE-07 1.20E-06 9.61 E-07 Internal Events 6.14E-09 2.05E-08 3.07E-08 2.46E-08 Combined 2.46E-07 8.21 E-07 1.23E-06 9.88E-07 The internal event results are also provided to allow a composite value to be defined. When both the internal and external event contributions are combined the total change in Unit 1 and 2 LERF meet the guidance for small change in risk, as it exceeds the 1.OE-7/yr and remains less than 1.OE-6 change in LERF for both units. For this change in LERF to be acceptable, total LERF must be less than 1.OE-5.

Conservatively using the highest seismic LERF value and not crediting containment spray success or plant damage state adjustments for the Internal Events PRA, the total LERF values are calculated below:

Unit 1: LERFul = LERFuiinternal + LERFuiseismic "+LERFuifire + LERFulclass3Bincrease

= 1.39E-6/yr + 1.41E-6/yr + 4.67E-6/yr + 7.89E-7/yr = 8.26E-6/yr Unit 2: LERFu 2 = LERFu2internal + LERFu2seismic + LERFu2fire+ LERFUzcass3Bincrease

= 1.56E-6/yr + 1.66E-6/yr + 6.13E-6/yr + 9.84E-7/yr = 1.03E-5/yr The Unit 2 LERF is barely greater than 1.OE-5. However, the Unit 2 Seismic LERF is between 3.69E-07 and 1.66E-06, and the highest Seismic LERF value was conservatively used to calculate 1.03E-5. If the 7 4 th percentile or smaller value of this range (< 1.32E-6) is used, the total Unit 2 LERF is less than 1.OE-5. Moreover, the IPEEE does not include recent significant plant modifications designed specifically to reduce fire risk. Therefore, it is reasonable to conclude that the total Unit 2 LERF is less than I.OE-5. Since the total LERF for both units is less than 1.OE-5, it is acceptable for the ALERF to be between 1.OE-7 and 1.OE-6.

5.3.2 Potential Impact from Steel Liner Corrosion Likelihood A quantitative assessment of the contribution of steel liner corrosion likelihood impact was performed for the risk impact assessment for extended 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 Revision 4 Page 36 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension corrosion likelihood using the relationships described in Section 5.1.4. 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 10 years to 1 in 10 years, or to 1 in 15 years are provided in Tables 5 5-36. Since CCFP is only concerned with a containment failure and not whether the release is small or large, the Class 1 results without containment spray and PDS refinement is used to calculate the CCFP. The Annual Dose Rate calculations are performed using the containment spray and PDS adjustments. The steel liner corrosion likelihood was increased by a factor of 1000, 10000, and 100000. Except for extreme factors of 10000 and 100000, the corrosion likelihood is relatively insensitive to the results.

Table 5 Unit 1 Steel Liner Corrosion Sensitivity Cases 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-M0 to year ILRT) year ILRT) year ILRT) I-per-10) 1-per-15) 1-per-15)

Internal Event 3B 1.14E-08 3.78E-08 5.68E-08 2.65E-08 4.54E-08 1.89E-08 Contribution Corrosion Likelihood 1.15E-08 3.98E-08 6.36E-08 2.84E-08 5.22E-08 2.38E-08 X 1000 Corrosion Likelihood 1.24E-08 5.76E-08 1.25E-07 4.52E-08 1.13E-07 6.77E-08 X 10000 Corrosion Likelihood 2.17E-08 2.35E-07 7.42E-07 2.14E-07 7.20E-07 5.06E-07 X 100000 Table 5 Unit 2 Steel Liner Corrosion Sensitivity Cases 3b 3b 3b LERF LERF LERF Frequency Frequency Frequency Increase Increase Increase (3-per-10 (I-per-10 (1-per-15 (3-per-10 to (3-per-10 to (I-per-10 to year ILRT) year ILRT) year ILRT) I-per-10) 1-per-15) 1-per-15)

Internal Event 3B 6.14E-09 2.05E-08 3.07E-08 1.43E-08 2.46E-08 1.02E-08 Contribution Corrosion Likelihood 6.19E-09 2.15E-08 3.44E-08 1.53E-08 2.82E-08 1.29E-08 X 1000 Corrosion Likelihood 6.70E-09 3.11E-08 6.77E-08 2.44E-08 6.1OE-08 3.66E-08 X 10000 Corrosion Likelihood 1.17E-08 1.27E-07 4.01E-07 1.16E-07 3.89E-07 2.74E-07 X 100000 Page 37 of 110 Revision 4 Page 37 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Unit I Steel Liner Corrosion Sensitivity Cases 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) I-per-10) 1-per-15) 1-per-15)

Baseline 2.04E-01 2.09E-01 2.13E-01 5.11 E-03 8.76E-03 3.65E-03 CCFP Corrosion Likelihood 2.04E-01 2.09E-01 2.13E-01 .5.16E-03 8.84E-03 3.68E-03 X 1000 Corrosion Likelihood 2.04E-01 2.10E-01 2.14E-01 5.57E-03 9.56E-03 3.98E-03 X 10000 Corrosion Likelihood 2.06E-01 2.16E-01 2.23E-01 9.76E-03 1.67E-02 1.29E-02 X 100000 Table 5 Unit 2 Steel Liner Corrosion Sensitivity Cases CCFP CCFP CCFP CCFP CCFP CCFP Increase Increase Increase (3-per-10 (I-per-10 (1-per-15 (3-per-10 to (3-per-10 to (1-per-M0 to year ILRT) year ILRT) year ILRT) 1-per-10) I-per-15) I-per-15)

Baseline CCFP 2.01E-01 2.06E-01 2.09E-01 5.08E-03 8.71E-03 3.63E-03 CCFP Corrosion Likelihood 2.01 E-01 2.06E-01 2.09E-01 5.13E-03 8.79E-03 3.66E-03 X 1000 Corrosion Likelihood 2.01 E-01 2.06E-01 2.10E-01 5.54E-03 9.50E-03 3.96E-03 X 10000 Corrosion Likelihood 2.03E-01 2.12E-01 2.19E-01 9.70E-03 1.66E-02 6.93E-03 X 100000 Table 5 Unit I Steel Liner Corrosion Sensitivity Cases 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)

Dose Rate 3.63E-02 1.21 E-01 1.82E-01 8.48E-02 1.45E-01 6.06E-02 Corrosion Likelihood 3.67E-02 1.27E-01 2.04E-01 9.08E-02 1.67E-01 7.61E-02 X 1000 Corrosion Likelihood 3.96E-02 1.84E-01 4.01E-01 1.45E-01 3.61E-01 2.17E-01 X 10000 Corrosion Likelihood 6.94E-02 7.53E-01 2.37E+00 6.84E-01 2.30E+00 1.62E+00 X 100000 Page 38 of 110 Revision 4 Page 38 of I 10

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Unit 2 Steel Liner Corrosion Sensitivity Cases 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-1 5) 1-per-1 5)

Dose Rate 1.96E-02 6.55E-02 9.82E-02 4.58E-02 7.86E-02 3.27E-02 Corrosion Likelihood 1.98E-02 6.89E-02 1.10E-01 4.91E-02 9.02E-02 4.12E-02 X 1000 Corrosion Likelihood 2.14E-02 9.96E-02 2.17E-01 7.82E-02 1.95E-01 1.17E-01 X 10000 Corrosion Likelihood 3.75E-02 4.07E-01 1.28E+00 3.70E-01 1.25E+00 8.76E-01 X 100000 Page 39 of 110 Revision 44 Page 39 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 5.3.3 Expert Elicitation Sensitivity Another sensitivity case on the impacts of assumptions regarding pre-existing containment defect or flaw probabilities of occurrence and magnitude, 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 extrapolated into a probability versus magnitude relationship for pre-existing containment defects. The failure mechanism analysis also used the historical ILRT data augmented with expert judgment to develop the results.

Details of the expert elicitation process and results are contained in Reference 24. The expert elicitation process has the 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 for large magnitude leakage events.

The expert elicitation results are used to develop sensitivity cases for the risk impact assessment. Employing the results requires the application of the ILRT interval methodology using the expert elicitation to change the probability of pre-existing leakage in the containment.

The baseline assessment uses the Jefferys non-informative prior and the expert elicitation sensitivity study uses the results of the expert elicitation. In addition, given the relationship between leakage magnitude and probability, larger leakage 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 La for small and 100 La for large) are used here. Table 5-37 presents the magnitudes and probabilities associated with the Jefferys non-informative prior and the expert elicitation use in the base methodology and this sensitivity case.

Table 5 CCNPP Summary of ILRT Extension Using Expert Elicitation Values (from Reference 24)

Leakage Size (La) Jefferys Non-Informative Expert Elicitation Percent Reduction Prior Mean Probability of Occurrence 10 2.70E-02 3.88E-03 86%

100 2.70E-03 9.86E-04 64%

Taking the baseline analysis and using the values provided in Tables 5 5-19 for the expert elicitation yields the results in Tables 5-38 and 5-39 for Units 1 and 2, respectively, are developed.

Table 5 CCNPP Unit I Summary of ILRT Extension Using Expert Elicitation Values Accident ILRT Interval Class 3 per 10 Years I per 10 Years I 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 1.28E-05 1.28E-05 3.40E+02 2.70E-04 1.26E-05 2.50E-04 1.25E-05 2.36E-04 2 5.01E-08 5.01 E-08 2.OOE+07 1.OOE+00 5.01E-08 1.00E+00 5.01E-08 1.00E+00 3a N/A 5.98E-08 3.40E+03 2.03E-04 1.99E-07 6.78E-04 2.99E-07 1.02E-03 3b N/A 1.52E-08 3.40E+04 5.17E-04 5.06E-08 1.72E-03 7.60E-08 2.58E-03 6 1.87E-06 1.87E-06 7.01E+06 1.31E+01 1.87E-06 1.31E+01 1.87E-06 1.31E+01 Revision 4 Page 40 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 CCNPP Unit I Summary of ILRT Extension Using Expert Elicitation Values Accident ILRT Interval Class 3 per 10 Years 1 per 10 Years I 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) 7 6.49E-07 6.49E-07 5.61E+07 3.64E+01 6.49E-07 3.64E+01 6.49E-07 3.64E+01 8 6.77E-07 6.77E-07 2.25E+07 1.52E+01 6.77E-07 1.52E+01 6.77E-07 1.52E+01 Totals 1.61E-05 1.61E-05 1.06E+08 6.58E+01 1.61E-05 6.58E+01 1.61E-05 6.58E+01 ALERF (3 per 10 N/A 3.55E-08 6.08E-08 yrs base)

ALERF (1 per 10 N/A N/A 2.53E-08 yrs base)

CCFP 20.26% 20.48% 20.64%

Table 5 CCNPP Unit 2 Summary of ILRT Extension Using Expert Elicitation Values Accident ILRT Interval Class 3 per 10 Years 1 per 10 Years I per 15 Years Base Adjusted Dose Dose Rate Frequency Dose Rate Frequency Dose Rate Frequency Base (person (person- (person- (person-Frequency -rem) rem/yr) rem/yr) rem/yr) 1 1.13E-05 1.12E-05 3.40E+02 3.81E-03 1.10E-05 3.73E-03 1.08E-05 3.67E-03 2 4.34E-08 4.34E-08 2.OOE+07 8.67E-01 4.34E-08 8.67E-01 4.34E-08 8.67E-01 3a N/A 5.21E-08 3.40E+03 1.77E-04 1.74E-07 5.90E-04 2.60E-07 8.86E-04 3b N/A 5.21E-08 3.40E+04 1.77E-03 1.74E-07 5.90E-03 2.60E-07 8.86E-03 6 1.25E-06 1.25E-06 7.01E+06 8.75E+00 1.25E-06 8.75E+00 1.25E-06 8.75E+00 7 8.35E-07 8.35E-07 5.61E+07 4.69E+01 8.35E-07 4.69E+01 8.35E-07 4.69E+01 8 6.72E-07 6.72E-07 2.25E+07 1.51E+01 6.72E-07 1.51E+01 6.72E-07 1.51E+01 Totals 1.41E-05 1.41E-05 1.06E+08 7.16E+01 1.41E-05 7.16E+01 1.41E-05 7.16E+01 ALERF (3 per 10 N/A 1.22E-07 2.09E-07 yrs base)

ALERF (1 per 10 N/A N/A 8.69E-08 yrs base)

CCFP 20.21% 21.08% 21.69%

The results illustrate how the expert elicitation reduces the overall change in LERF and the overall results are more favorable with regard to the change in risk.

Page 41 of 110 Revision 44 Revision Page 41 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 5.3.4 Large Leak Probability Sensitivity Study The large leak probability is a vital portion of determining the Class 3b frequency. CCNPP had previously calculated the large leak probability using the WCAP method. Table 5-40 presents the large leak probabilities for the baseline test, 10 year test interval, and 15 year test interval.

Table 5-41 was developed using the same process as to calculate Class 3b.

Table 5 CCNPP Large Leak Probabilities Using the WCAP Method Test Interval WCAP Large Leak EPRI Accident EPRI Accident Class 3b Probability Class 3b Frequency: Unit 2 Frequency: Unit 1 3 per 10 years 2.47E-4 1.38E-09 8.21E-10 10 years 7.41E-4 4.05E-09 2.41E-09 15 years 1.11E-3 5.96E-09 3.55E-09 Using the same EPRI approach, but with an updated Class 3b frequency calculated from the WCAP large leak probability data, Table 5-41 contains the final results for both units.

Table 5 Impact on LERF due to Extended Type A Testing Intervals with WCAP CDF ILRT Inspection Unit 1: 3 Unit 1:10 Unit 1:15 Unit 2: 3 Unit 2:10 Unit 2:15 Interval Years Years Years Years Years Years (baseline) (baseline)

Class 3b (Type A LERF) 1.38E-09 4.05E-09 5.96E-09 8.21E-10 2.41 E-09 3.55E-09 ALERF (3 year 2.73E-09 baseline) 4.57E-09 ALERF (10 year 1.91 E-09 1.14E-09 baseline)

These results demonstrate that the EPRI methodology is conservative when used to calculate a large leak probability as compared to the WCAP method.

Page 42 of 110 Revision 44 Revision Page 42 of 110

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 6.0 RESULTS The results from this ILRT extension risk assessment for CCNPP are summarized in Table 6-1 for Unit 1 and Table 6-2 for Unit 2.

Table 6 Unit I ILRT Extension Summary Class Dose Base Case Extend to Extend to (person- 3 in 10 Years 1 in 10 Years 1 in 15 Years rem)

CDF/Year Person- CDF/Year Person- CDF/Year Person-Rem/Year Rem/Year Rem/Year 3.20E+04 5.59E-06 1.79E-01 5.46E-06 1.75E-01 5.37E-06 1.72E-01 2 2.OOE+07 3.33E-08 6.66E-01 3.33E-08 6.66E-01 3.33E-08 6.66E-01 3a 3.20E+05 4.56E-08 1.46E-02 1.52E-07 4.87E-02 2.28E-07 7.30E-02 3b 3.20E+06 1.14E-08 3.63E-02 3.78E-08 1.21E-01 5.68E-08 1.82E-01 6 7.01E+06 1.54E-06 1.08E+01 1.54E-06 1.08E+01 1.54E-06 1.08E+01 7 5.61 E+07 2.49E-07 1.40E+01 2.49E-07 1.40E+01 2.49E-07 1.40E+01 8 2.25E+07 6.68E-07 1.50E+01 6.68E-07 1.50E+01 6.68E-07 1.50E+01 Total 8.14E-06 4.07E+01 8.14E-06 4.08E+01 8.14E-06 4.09E+01 ILRT Dose Rate from 3a and 3b From 3 N/A 1.15E-01 1.96E-01 ATotal Years Dose Rate From 10 N/A Years YearsN/A 8.18E-02 From 3 N/A 0.282% 0.483%

%ADose Years Rate From 10 N/A Years YearsN/A 0.201%

I 3b Frequency (LERF)

From 3 N/A 2.65E-08 4.54E-08 Years ALERF From 10 N/A N/A 1.89E-08 Years CCFP %

From 3 N/A 0.326% 0.558%

Years ACCFP%

From 10 N/A N/A 0.233%

Years Revision 4 Page 43 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 6 Unit 2 ILRT Extension Summary Class Dose Base Case Extend to Extend to (person- 3 in 10 Years 1 in 10 Years I in 15 Years rem)

CDF/Year Person- CDF/Year Person- CDF/Year Person-Rem/Year Rem/Year Rem/Year 1 3.20E+04 3.32E-06 1.06E-01 3.25E-06 1.04E-01 3.20E-06 1.02E-01 2 2.OOE+07 1.84E-08 3.69E-01 1.84E-08 3.69E-01 1.84E-08 3.69E-01 3a 3.20E+05 2.47E-08 7.89E-03 8.22E-08 2.63E-02 1.23E-07 3.95E-02 3b 3.20E+06 6.14E-09 1.96E-02 2.05E-08 6.55E-02 3.07E-08 9.82E-02 6 7.01 E+06 5.74E-07 4.02E+00 5.74E-07 4.02E+00 5.74E-07 4.02E+00 7 5.61E+07 4.01E-07 2.25E+01 4.01E-07 2.25E+01 4.01E-07 2.25E+01 8 2.25E+07 6.60E-07 1.49E+01 6.60E-07 1.49E+01 6.60E-07 1.49E+01 Total 5.01E-06 4.19E+01 5.01E-06 4.19E+01 5.01E-06 4.20E+01 ILRT Dose Rate from 3a and 3b From 3 N/A 6.19E-02 1.06E-01 ATotal Years Dose Rate From 10 N/A YearsN/A 4.42E-02 Years From 3 N/A 0.148% 0.254%

%ADose Years Rate From Years10I N/A N/A Years 0.106%

3b Frequency (LERF)

From 3 N/A 1.43E-08 2.46E-08 Years ALERF From 10 N/A N/A 1.02E-08 Years CCFP %

From 3 Years N/A 1 0.286% 0.490%

ACCFP%

From 10 N/A N/A 0.204%

Years Page 44 of 110 Revision 44 Page 44 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension

7.0 CONCLUSION

S AND RECOMMENDATIONS Based on the results from Section 5.2 and the sensitivity 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 years:

" Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of CDF less than 1.OE-06/year and increases in LERF less than 1.OE-07/year. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 4.54E-8/year for Unit 1 and 2.46E-8/year for Unit 2 using the EPRI guidance. As such, the estimated change in LERF is determined to be "very small" for both units 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 plant risk for those accident sequences influenced by Type A testing, is 0.20 person-rem/year for Unit 1 and 0.11 person-rem/year for Unit 2. EPRI Report No. 1009325, Revision 2-A [Reference 24] states that a very small population dose is defined as an increase of < 1.0 person-rem per year, or <

1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. This results of this calculation meet these criteria for both units. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible.

" The increase in the conditional containment failure from the 3 in 10 year interval to 1 in 15 year interval is 0.558% for Unit 1 and 0.490% for Unit 2. EPRI Report No. 1009325, Revision 2-A [Reference 24] states that increases in CCFP of < 1.5% is very small.

Therefore, this increase is judged to be very small.

Therefore, increasing the ILRT interval to 15 years is considered to be insignificant since it represents a very small change to the CCNPP risk profile.

Previous Assessments The NRC in NUREG-1493 [Reference 6] has previously concluded that:

" Reducing the frequency of Type A tests (ILRTs) from 3 per 10 years to 1 per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B or Type C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements.

" Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage-rate tests is possible with minimal impact on public risk. 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 findings for CCNPP confirm these general findings on a plant-specific basis considering the severe accidents evaluated for CCNPP, the CCNPP containment failure modes, and the local population surrounding CCNPP.

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I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension A. ATTACHMENT 1 PRA Quality Statement for Permanent 15-Year ILRT Extension The Calvert Cliffs Internal Events and Wind Model, Calvert-CAFTA-TREE-6.2a, was used for this analysis.

An independent PRA peer review was conducted under the auspices of the Pressurized Water Reactor Owners Group in June of 2010, and was performed against the guidance of Regulatory Guide 1.200, Revision 2, and requirements of American Society of Mechanical Engineers (ASME)/American National Standards (ANS) RA-Sa-2009. The scope of the review was a full-scope review of the Calvert Cliffs Nuclear Plant (Calvert Cliffs) at-power, internal initiator PRA.

Findings (generally, documentation issues or model concerns that have been evaluated as not significant using a sensitivity study) have been captured in the PRA Configuration Risk Management Program (CRMP) database. On an on-going basis, other potential PRA model and documentation changes are captured and prioritized in the CRMP database.

To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plant, the following configuration control activities are routinely performed:

" Design changes and procedure changes are reviewed for their impact on the PRA model. PRA screening is required for all design and procedure changes.

" New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.

" Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated based upon reviews of plant program data, particularly data supporting the Maintenance Rule.

The Calvert Cliffs Internal Events model is also updated to support the Calvert Cliffs Fire PRA.

The Calvert Cliffs Internal Events PRA is based on a detailed model of the plant developed from the Individual Plant Examination for Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities." The model is maintained and updated in accordance with Calvert Cliffs procedures, and has been updated to meet the ASME PRA Standard and Revision 2 of Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities."

The Calvert Cliffs internal events PRA model was peer reviewed in June 2010. All findings which had significant impact on this analysis have been addressed. This assessment is provided as Table 1. The ILRT application was determined to be an application requiring a Capability Category II PRA model per the Regulatory Guide 1.200 criteria, Revision 2. This is based on the requirement for numerical results for CDF and LERF to determine the risk impact of the requested change and the fact that this change is risk-informed, not risk-based. Table 1 includes discussion of all findings from the industry peer review along with the assessment and evaluation of the finding that shows that they have either been addressed or have no material impact on the ILRT interval extension request.

The peer review found that 97% of the SR's evaluated Met Capability Category IIor better.

There were 3 SRs that were noted as "not met" and eight that were noted as Category I. As noted in the peer review report, the majority of the findings were documentation related. Of the 11 SRs which did not meet Category IIor better, seven were related to conservatisms or documentation in LERF and two were related to internal floods. The 3 SRs that were noted as "not met" are LE-F2, LE-G5, and IFQU-A10. LE-F2 relates to LERF results. The dominant LERF contributors were reviewed and model changes implemented prior to the ILRT analysis. LE-G5 Revision 4 Page 46 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension relates to the documentation of limitations of applications of the PRA. IFQU-A1O relates to documentation of the treatment of the internal flood analysis in the event trees. None of the "not met" SRs impact the ILRT extension analysis. There were 39 findings. All findings which could be relevant to the ILRT extension evaluation were updated in the internal events model used to quantify the Level 2 release states. Thus, with the exception of minor documentation concerns, the internal events model meets Capability Category II or causes conservative results for all SRs relevant to the ILRT extension evaluation results. No significant changes have been implemented in the internal events PRA. As there are no new methods applied, no follow on or focused peer reviews were required.

The Calvert Cliffs Fire PRA peer review was performed January 16-20, 2012 using the NEI 07-12 Fire PRA peer review process, the ASME PRA Standard (ASME/ANS RA-Sa-2009) and Regulatory Guide 1.200, Rev. 2. The purpose of this review was to establish the technical adequacy of the Fire PRA for the spectrum of potential risk-informed plant licensing applications for which the Fire PRA may be used. The 2012 Calvert Fire PRA peer review was a full-scope review of all of the technical elements of the Calvert Cliffs at-power FPRA (2012 model of record) against all technical elements in Section 4 of the ASME/ANS Combined PRA Standard, including the referenced internal events SRs. The peer review noted a number of facts and observations (F&Os). The findings and their dispositions are provided in Table 2. All findings are being provided and have been dispositioned. All F&Os that were defined as suggestions have been dispositioned and will be available for NRC review. The Fire PRA is adequate to support the ILRT extension.

The Calvert Cliffs seismic PRA model is relatively conservative and, other than the high magnitude acceleration event, is not a dominant contributor. The Calvert Cliffs high winds PRA model is very conservative in the tornado area in that all tornados are grouped into the most conservative event. PRA risk for tornadoes and high winds are based upon IPEEE values. Calvert Cliffs has maintained and updated a high wind PRA model in order to perform risk assessment of tornado missile impacts and hurricane force winds. Although this model has not been peer reviewed in compliance with the ASME/ANS RA-Sa-2009 standard, the model is based upon accepted methodology and utilizes the ASME/ANS RA-Sa-2009 compliant internal events model. High winds updates are not expected to cause a significant increase in CDF or LERF. A more detailed assessment would be expected to cause a decrease in CDF.

Page 47 of 110 Revision 44 Revision Page 47 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to ILRT Extension 1-16 AS-B3 Systems Based on Sections 2.4 and 2.10 of Complete The PRA Internal Events No impact on SY-B6 Analysis the System Analysis Introduction Accident Sequence ILRT analysis.

Notebook (CO-SY-00, Rev. 0) this Notebook, CO-AS-O0i, Subsequent SR appears to be met. However, Section 3.3, has been analysis has there is a potential issue related to updated with an engineering found this issue this SR. Did not find reference to analysis of this issue. The to be non-any engineering analysis needed analysis identifies that significant: 1) the to support Containment Air Cooler during the Loss of Offsite temperature rise operation when this system is Power sequences, the is not likely to assumed to be available during Containment Air Coolers are challenge the LOSP when the containment heats credited for SBO conditions containment air up prior to electrical recovery. where the containment coolers, and 2) heats up, and then, after the importance of (This F&O originated from SR SY- power recovery, the air the air coolers is B6) coolers are credited for significantly containment pressure and reduced by the temperature control. For redundant these accident sequences, function provided offsite power is restored in by containment one hour, and the spray containment pressure and corresponding saturation temperature remain well below containment design parameters that would challenge the CACs.

Furthermore, failure of CACs is not risk significant, due to the potential availability of containment spray.

REFERENCE CO-AS-001 Page 48 of 110 Revision 44 Page 48 of 110

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension 1-17 IFSO-A1 Internal Examined Internal Flooding Complete An engineering analysis has Due to the QU-E3 Flooding Notebook (CO-IF-001, Rev. 1) subsequently been relatively low Sections 3.0 and 3.1. Part of the performed for AFW contribution to Internal Flood analysis may not be discharge piping flooding. CDF, this flood complete for assessing the Aux The fraction of at-power time has no impact on Feedwater Discharge Piping as a during which the AFW ILRT analysis.

Flood Source. system is in operation 0.6%

and the AFW Discharge (This F&O originated from SR Piping flood may be IFSO-A1) screened due to their low impact on CDF (<1 E-9).

REFERENCE CO-IF-001 1-18 IFSO-A4 Internal Examined Internal Flooding Open Human-induced impacts on No impact on IFEV-A7 Flooding Notebook (CO-IF-001, Rev. 1) the flood initiating event ILRT analysis.

Section 3.3 and 5.3. Consideration frequencies are not well This is a of human-induced mechanisms as documented. The issue has documentation potential flood sources not clear, been captured in the PRA issue.

Regarding human-induced impacts configuration control on the flood frequency, Section 5.3 database (CRMP), but not of the IF report states that they yet addressed.

were included, but their inclusion should be better documented or referenced from IF (e.g., a sample calculation showing human contribution would be helpful)

(This F&O originated from SR IFSO-A4) 1-19 IFEV-B3 Internal While some items are included in Open In the Internal Flood No impact on IFPP-B3 Flooding Section 7.0 of the IF report, many notebook, the discussions ILRT analysis.

IFQU-B3 other instances of uncertainties on uncertainties and This is a Revision 4 Page 49 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension IFSN-B3 and assumptions are cited assumptions should be documentation IFSO-B3 throughout the report, but not expanded. This issue has issue.

included in the discussion of been captured in the PRA Section 7.0 nor are the configuration control implications of these other database (CRMP), but not uncertainties and assumptions are yet addressed.

discussed.

(This F&O originated from SR IFPP-B3) 1-25 DA-C7 Data For the most part actual plant- Complete The ESFAS logic train The low risk specific data is used as a basis for testing has a very low risk significance of the number of demands significance and generally ESFAS logic train associated actual plant does not take the logic OOS. testing is experiences (See basis for DA- The train does go to 2-out- considered to C6), which includes both actual. of-3 logic. Occurrences have no impact planned and unplanned activities, where the train is in 2-out-of- on ILRT analysis.

However, there are a few ESFAS 3 logic is incorporated into testing and/or other logic channel the PRA Data Analysis testing that are not tracked via the Notebook, CO-DA-plant computer. 001, Section 2.6 and 3.5.

For the logic relays there is Created this F&O on non- a RAW of <1.04 and documentation of ESFAS/logic Bimbaum on the order of train testing, which needs to 4E-07. Any logic relay include actual practice. unavailability that does not cause the ESFAS channel to (This F&O originated from SR DA- be OOS and bypassed, is C7) therefore of low significance.

REFERENCE CO-DA-001 Page 50 of 110 Revision 44 Page 50 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension 2-7 IFPP-A5 Internal Flood Section 2.3 provides a discussion Complete A walkdown was performed No impact on that walkdowns used to confirm to assess the susceptibility ILRT analysis.

plant arrangement. The following to jet impingement or spray This is an note is contained in section 2.3: in rooms 105A and 203. All Internal Flood equipment is considered documentation Unfortunately, the walk-down failed by spray or issue.

documentation from the original impingement for flood flooding analysis no longer exists. sources originating in the A plant walk-down was performed room. Notebook CO-IF-001 as a part of this analysis to provide was updated with this familiarity with the plant design as additional documentation.

well as confirm findings from the original walk-down. This walk- REFERENCE down is documented in a set of CO-IF-001 notes and photographs included in Appendix B.

Walkdown photos for room 105A and 203 show equipment and potential flood propagation paths.

However, there is not enough spatial information to develop specific targets for flood impingement or spray.

(This F&O originated from SR IFPP-A5) 2-9 DA-D4 Data Evidence of meeting this SR at Complete Table 2-6 of the Data No impact on CC-I/Ill is found in the PRA Data Notebook CO-DA-001 listed ILRT analysis.

Notebook (CO-DA-001, Rev. 1) in incorrect data and Bayesian This was a Sections 2.1 and 2.7. Found update results for the documentation inconsistencies in the value of total SACMs. However, the issue. The number components of different correct values were used in Internal Events types (for both units) in Table 2-5 the models for peer review.

Revision 4 Page 51 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension of the PRA Data Notebook with the model includes actual total number for Calvert For the SACM EDGs in the correct data.

Cliffs. Also, found an inconsistency Table 2-6, the correct plant-between the prior distribution and specific data are in Table 2-posterior distribution for SACM 5. Table 2-6 lists incorrect EDG fail to start in Table 2-6 of the data and Bayesian Update Data Notebook. results for the SACMs.

However, the correct values (This F&O originated from SR DA- are used in the models.

D4)

The above errors have been corrected in CO-DA-001.

Other minor typographical errors were identified and corrected in the notebook.

REFERENCE CO-DA-001 3-3 SY-C2 Systems Section 2.3 of each system Complete Marked-up system boundary No impact on Analysis notebook states that marked up drawings were generated for ILRT analysis.

plant system drawings are each system notebook. This is an provided as supplements to the Where Unit 1 and Unit 2 are Internal Events system notebook, which depicts similar, just the Unit 1 documentation the boundary of the system in boundary is depicted. In issue that has terms of PRA modeling. The addition, the system been addressed.

drawings are not in the notebooks. notebooks include drawing snippets, sketches, and (This F&O originated from SR SY- descriptive text that also C2) depict the system boundary.

REFERENCES CO-SY-[AII]

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I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension 3-5 SY-Al 1 Systems The fault tree does not include Complete A bounding sensitivity case This finding does SY-A6 Analysis potential failures of the AFW was run to include failure of not impact the accumulator system. the AFW accumulators ILRT extension.

failing short-term AFW The random (This F&O originated from SR SY- operation. This issue has an failure probability Al1) insignificant contribution to of the CDF. Short-term failure of accumulators is the AFW operation is two orders of dominated by failure of magnitude lower electrical support systems than active and failure of active hardware failures hardware (i.e. valves and that support the instrumentation). The same system applicable system function.

notebooks were updated.

REFERENCES C0-SY-036 CO-SY-019 CO-SY-000 3-8 SY-Cl Systems Several system notebooks were Complete Some new flow diversions No impact on SY-A13 Analysis reviewed (AFW, EDG, SI, 120 were identified as part of the ILRT analysis.

VAC electrical, etc.). In general, Fire PRA Multiple Spurious This is an the documentation is complete and Operation review, and these Internal Events thorough. In most cases it clearly were added to the system documentation follows the RG 1.200 SRs. models and system issue that has In some places, assumptions were notebooks. Furthermore, a been addressed.

imbedded in the documentation comprehensive review of without sufficient reference or PRA mechanical systems justification. Examples include: notebooks and drawings was performed to identify Sl notebook page 11, last bullet and document potential flow

'Only one of the three HPSI pumps diversions. Flow diversion functions - For a cold leg break, it discussions were added to Revision 4 Page 53 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to ILRT Extension is assumed that only one-fourth Sections 3.4.d of the pump discharge is spilled via the applicable system break. For a hot leg break, the notebooks.

entire pump discharge reaches the core.'

SI notebook page 12, 2nd bullet T'he maximum time assumed for operation for the safety injection pumps is 30 seconds following SIAS initiation.'

CO-SY-000 states that each system notebook addresses flow diversions (where applicable) in section 3.4.d. Although flow diversions appear to be addressed (for example, the SW notebook talks about flow diversion), there is no consistent discussion in each system notebook.

(This F&O originated from SR SY-Cl) 3-9 DA-B1 Data DA notebook table 2-5 contains Complete The model has been No impact on the grouping of components for updated to add additional ILRT analysis.

plant specific failure data. Many of component types and failure The model used the groupings appear to take into modes to better reflect for the ILRT account differences in such things service conditions. Service analysis includes as size, type, mission type (e.g., Water and Salt Water the updated data FW TDP run vs. AFW TDP pumps were broken out. and failure standby). However, in some cases, AFW pumps and Safety modes.

it is not clear what the basis for the Injection pumps were broken grouping is. For example, SW out. This resulted in changes MDP RUN and SRW MDP RUN to the associated failure are grouped together even though rates. The change has been Revision 4 Page 54 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to ILRT Extension they are of different service reflected in the Data conditions (salt water vs. clean Notebook, CO-DA-001.

water), voltages (480 VAC vs.

4160 VAC), size, etc. Similarly, REFERENCE AFW MDP is included with HPSI CO-DA-001 MDP and LPSI MDP, even though the two SI pumps are pumping borated water, while the AFW pump is pumping condensate grade water. No documentation of the appropriateness of these groupings is provided.

(This F&O originated from SR DA-B1) 3-11 QU-B7 Quantification The mutually exclusive cutsets for Complete A comprehensive review of No impact on each system are described in the mutual exclusive modeling ILRT analysis.

system notebook section 3.4.e. was performed. Each The PRA model Several SY notebooks were system notebook and each that was updated reviewed to determine system model was reviewed as part of this appropriateness of the mutually to validate the review was used exclusive cutsets. All appeared appropriateness of the as the model for reasonable. A review was modeling and reconcile any the ILRT performed of the MUTEX gate differences, and to verify analysis.

within the fault tree model and the that a documented basis appropriate combinations identified exists for each mutually in the SY notebooks appear to exclusive event. The PRA have been included in the model. model was updated to reflect There are two gates under the new, deleted, or re-MUTEX gate which contain organized mutually exclusive mutually exclusive cutsets which modeling identified as part of are not documented in the system this review.

notebooks. While the majority of these are intuitively obvious (e.g.,

Revision 4 Page 55 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension 11 Steam Generator Tube Rupture REFERENCE occurs as an IE AND 12 Steam SY-CO-[ALL]

Generator Tube Rupture occurs as an IE), these should be included in an appropriate system notebook.

(This F&O originated from SR QU-B7) 3-12 QU-D3 Quantification A review of the top cutsets from Complete Documentation of the cutset No impact on each event tree was performed. reviews was presented to ILRT analysis.

The utility stated that during this the peer review team; The original review, cutsets were reviewed to although, the documentation internal events determine ifany mutually exclusive was separate from the cutset review events were contained within formal QU notebook notes have now cutsets, ifany flag settings were package. A note was added been archived.

inappropriate or if any recoveries to the QU notebook directing were overlooked or added the reader to the location of inappropriately. A review of a the cutest review notes and sampling of cutsets did not indicate spreadsheets. The PRA any inappropriate results. configuration control However, the QU notebook does procedure, CNG-CM-1.01-not include a discussion of this 3003, requires a review of review. cutsets for PRA changes. In (This F&O originated from SR QU- practice, the top CDF and D3) LERF cutsets are examined for even the most innocuous model changes.

REFERENCE CNG-CM-1.01-3003 CO-QU-001 CO-FRQ-001 4-5 IE-A10 Complete To address this finding, the No impact on SY-A10 Initiating Events The onlysystems shared mentionbetween in C0-SC-001 of the units Diesel Generator modeling ILRT analysis.

Revision 4 Page 56 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension IE-C3 is the SBO EDG, noted in Section was updated as described in The finding has SC-A4 4.1.2. It states that the SBO diesel Appendix H of CO-SY-023- been addressed can power any one bus on either 024, PRA DG System in the Internal unit. However, in the CAFTA Notebook. EOP-7 directs to Events model, model, there is an assumed bus align the OC DG to the unit which, in turn, is preference of 11, then 24, then 12, with redundant safety used in the ILRT then 23.* This is noted in the EDG equipment out-of-service, analysis.

system notebook but no basis is with a goal to restore at least provided. The procedures do not one 4KV bus. Since 4KV actually have a preference, which Buses 11 and 24 support yields a potentially non- AFW, those busses would conservative analysis. For have a preference over example, ifthere is a LOOP, the Busses 14 and 21, all else U2 diesels fail to start and the Ul being equal. No unit diesels fail to run after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The preference is modeled. If SBO diesel would then be aligned there is a conflict in the to U2, and it is non-conservative to order-of-preference, for give the U1 bus 11 full credit. If example, both 4KV Bus 11 such non-conservatism is and 4KV Bus 24 are not negligible, some analysis should powered, then a 50-50 be performed to demonstrate this. probability is assumed as to the preferred bus.

(This F&O originated from SR IE-Al 0) REFERENCE C0-SY-023-024

  • Note: Peer review finding was not precise. Itshould have stated bus preference for Unit 1 is 11, then 24, and for Unit 2, is 24 then 21.

4-12 HR-Cl Human One basic event calculated in the Complete The basic event has been No impact on Reliability appendix (ESFOHFCISZEFG) was added to the model. A ILRT analysis.

not included in the fault tree sensitivity run with the basic The missing models. CCNPP staff noted that it event included in the current basic event has Revision 4 Page 57 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to ILRT Extension had previously been modeled, but model showed no increase been added to inadvertently deleted in an update. in risk. The system notebook the internal CO-SY-048 was updated. events model (This F&O originated from SR HR- used in the ILRT Cl) REFERENCE analysis.

CO-SY-048 4-15 IFEV-A6 Internal The internal flooding analysis did Open This finding has been No impact on Flooding not have a formal process to identified in the PRA ILRT analysis.

gather plant specific design configuration control The review of information, operating practices, database (CRMP), but has condition reports etc. that could potentially affect the not yet been addressed. did not identify generic flooding frequencies. In any design response to an NRC RAI on the issues or CCNPP ISI program plan, CCNPP operating mentioned a review of Condition practices that Reports that did not find any items would affect the that would increase the flooding generic flooding frequency. frequencies.

The CR review meets part of the requirement, but the SR also calls for reviews of plant design, operating practices, etc. that should be considered. The evaluation should be documented in the PRA.

(This F&O originated from SR IFEV-A6)

Revision 4 Page 58 of 110

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to ILRT Extension 4-19 LE-C13 Large Early The sources of uncertainty are well Complete Dominant LERF cutsets No significant LE-F3 Release identified in Table 5-1 of the LE were reviewed to identify impact on ILRT LE-G4 notebook and quantified in Table uncertainties that could be analysis. The 5-2 of the QU notebook. However, addressed. Two changes dominant LERF no discussion of the uncertainties have been implemented to contributors were or insights from them is provided, address significant reviewed and For example, Sensitivity 1 shows a uncertainties and reduced model changes 74% reduction in LERF, but this LERF. First, a reverse-flow implemented.

large reduction is not investigated, check valve in the CVCS The Calvert Cliffs Letdown line was credited LERF Also, conservatisms in the ISLOCA as a potential ISLOCA contribution is analyses were discussed in the AS recovery. Second, a new now similar to review. SGTR was treated in an human action was added other PWRs.

overly conservative manner by with realistic timing for categorizing all SGTR as LERF. Steam Generator isolation and RCS depressurization (This F&O originated from SR LE- on a SGTR. These and less F3) significant model updates resulted in a LERF-to-CDF ratio change from approximately 17% to approximately 10%. This newer ratio is in the typical range for other PWRs.

REFERENCE CO-LE-001 4-20 LE-F1 Large Early The relative contribution to LERF Complete The contributions to LERF No impact on LE-G3 Release is presented in the QU notebook are documented in the ILRT analysis.

by PDS and by initiating event, but Quantification Notebook and This is an internal not by accident progression are noted as such in the events sequence, phenomena, Level 2 Notebook. Accident documentation containment challenges or progression sequences are issue.

containment failure mode. located in Section 4.2.3 and Revision 4 Page 59 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to ILRT Extension Appendix C. The Level 2 (This F&O originated from SR LE- notebook has been updated G3) to point to additional phenomena and containment challenges and failure mode Tables/Figures in the QU Notebook REFERENCE CO-QU-001 CO-LE-001 4-21 LE-G5 Large Early The LE notebook states that Complete Section 5.5.2.7 of CO-LE- This internal Release limitations in the LE analysis that 001, Revision 2 - added events finding could impact applications are discussion of results of does not impact documented in the QU notebook, impact on application of the the ILRT but it is not. Therefore, SR LE-G5 Unit 2 ILRT extension analysis.

was determined to be not met. request.

Given the conservative modeling of SGTR and ISLOCA, the impact REFERENCE on applications should include CO-LE-001 assessment of how this conservatism can skew the LERF results.

(This F&O originated from SR LE-G5) 4-22 LE-C10 Large Early The LERF contributors have not Complete The LERF results were No significant LE-C12 Release been reviewed for reasonableness reviewed for conservatisms impact on ILRT LE-F2 (per SR LE-F2). Therefore, SR LE- as described in the SRs. analysis. The LE-C3 F2 was determined to be not met. After conservatisms were dominant LERF The QU notebook discusses the addressed (see discussion contributors were top 20 LERF cutsets (which total for F&O 4-19 above), no reviewed and Revision 4 Page 60 of 110

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension 73% of the total LERF). It notes significant issues were model changes conservatism in the cutsets and identified. implemented.

says it will be evaluated in Section The Calvert Cliffs 5.2, but is not. Section 4.3.6 of the REFERENCE LERF QU notebook compares the total CO-LE-001 contribution is LERF of CCNPP to St. Lucie, but now similar to does not even break the results other PWRs.

down by contributor (e.g., SGTR, ISLOCA, etc.).

Also, the ASME PRA Standard SRs C-3, C-10 and C-13 require a review of the LERF results for conservatism in the following areas:

1. Engineering analyses to support continued equipment operation or operator actions during severe accident progression that could reduce the LERF
2. Engineering analyses to support continued equipment operation or operations after containment failure.
3. Potential credit for repair of equipment.

No such review has been performed, despite the large conservatism noted in the containment bypasses.

(This F&O originated from SR LE-F2) 5-10 LE-D7 Large Early Following the failure of one or Complete The merits have been No impact to Release more containment penetrations to considered of adding an ILRT analysis.

isolate on CIAS, a feasible operator action in order Modeling of an Revision 4 Page 61 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension operator action is to manually close containment operator action to close the failed valves from the penetration from the Main manually close Main Control Room. Control Room to recover failed valves from from a containment isolation the main control (This F&O originated from SR LE- failure. A review of cutsets room would not D7) shows that a recovery is not significantly feasible for top LERF reduce LERF, as sequences, because the such an action is sequence includes either 1) not feasible for a loss of CR indication, 2) the significant includes a station black-out sequences where condition, or 3) includes containment non-recoverable pipe isolation has breaks. failed.

REFERENCE CO-LE-001 Attachment S 5-17 IE-C1 Initiating Bayesian updates of non-time- Complete CENG understands the No impact on IE-C13 Events based LOCA data were improper. general concern on ILRT analysis.

IE-C4 The small and medium LOCA Bayesian updating of rare The approach frequencies were obtained from events. However, the used for LOCA draft NUREG 1829 then Bayesian method used was based on frequencies has updated (in App E) with CCNPP a white paper developed by been validated by experience from 2004 to 2008. The industry experts regarding industry experts Very Small LOCA prior having LOCA frequencies. These and is the same alpha = 0.4, Mean = 1.57E-03; was experts included INL, NRC approach as was Bayesian updated to a Posterior and Industry experts. In used for the having a mean value of 7.02E-04. addition, the approach used NRC's SPAR This represents an excessive drop for the Calvert PRA was the model.

associated with CCNPP same as used for the NRC experience of 4 to 5 years. SPAR model. This issue is Similarly, the Small and Medium captured in the PRA LOCAs were Bayesian updated configuration control Revision 4 Page 62 of 110

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension with the whole industry experience database (CRMP).

rcy data. The draft NUREG 1829 LOCA frequencies were obtained REFERENCE from expert elicitations (not time- CO-lE-001 based) that included crack propagation analysis. The Bayesian update for VSLOCA used the Alpha parameter and the mean value to justify that the prior mean was based on 255 rcy. This may not have been the basis for the expert elicitations in NUREG 1829.

Also, the Medium LOCA frequency may be classified as extremely rare event. Itwould require no Bayesian updating. The current CCNPP SLOCA and MLOCA frequencies are very close even though the source data in NUREG 1829 indicates a negative exponential drop in these frequencies.

(This F&O originated from SR IE-Cl)

(Note: rcy - reactor year) 5-18 IE-C2 Initiating Justify the exclusion of LOOP Complete The event is not counted No impact on IE-C7 Events event at CCNPP in 1987. No time following guidance provided ILRT analysis.

trend analysis was provided to in NUREG/CR-6928, based The data analysis justify the exclusion, upon trend analysis. A full is acceptable.

discussion is included in the (This F&O originated from SR IE- Initiating Event notebook, C2) CO-IE-001.

Revision 4 Page 63 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension REFERENCE CO-IE-001 5-23 HR-A2 Human The Pre-Initiator HRAs did not Complete It is agreed that the No impact on Reliability include the miscalibration of SIT miscalibration of SIT ILRT analysis.

pressure. For example, in the pressure could have a Given the event where SIT pressure is negative impact on various pressure of the miscalibrated high, various accident scenarios involving CCNPP SITs accident scenarios requiring SI are LLOCA and VLLOCA they are only negatively impacted. Add SIT initiators. However, this required and pressure miscalibrated high or, instrumentation is not provide justify no impact on CDF / LERF. modeled explicitly and is significant benefit therefore deemed included on Large LOCAs.

(This F&O originated from SR HR- within the component The frequency of A2) boundary for the SIT. As a Large LOCA such the miscalibration times the pre-probability would be initiator included in the SIT frequency is unavailability, negligible.

REFERENCE CO-HR-001 5-25 SC-Cl Success Simplify the traceability of Tsw. In Complete Where applicable, the Tsw No impact on HR-12 Criteria the post initiator HRA details, the of each HFE that could be ILRT analysis.

SC-C2 HRA success criteria are often traced to the Success This is an internal provided as a positive re-statement Criteria notebook (CS-SC- events of the HRA title. And, the 001) was updated and documentation consequence of failure is often referenced in the HRA finding.

stated as core damage. Consider Calculator. C0-HR-001 was adding Tsw to the success criteria also updated.

and linking that to the PCTran case where Tsw was developed.

Revision 4 Page 64 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension Also, in the SC report (Table B-3), REFERENCE consider adding the actual time to CO-HR-001 core uncovery (or core damage) instead of providing a "Yes" entry in the column of "core damage?"

(This F&O originated from SR HR-12) 5-30 LE-D1 Large Early Section 3.2.11 discussed the Complete CCNPP's Level 2 PRA No impact on LE-B2 Release containment challenge from follows the analysis in ILRT analysis.

Hydrogen Combustion. It WCAP-16341-P, Simplified The methodology concluded that the challenge may Level 2 Modeling in WCAP-16341-be significant for some accident Guidelines. In the industry- P is appropriate scenarios. The CCNP entry in supported analysis, the for Calvert Cliffs Table 6.11-2 of the Level 2 WCAP percentage of cladding level 2 analysis showed a potentially significant oxidation is the main factor for internal impact from Hydrogen burn. used to develop a maximum events initiators.

Provide an estimate of the impact H2 concentration in the of Hydrogen burn on containment containment, and, in turn, a pressure. Use an accident containment pressure is scenario that is likely to produce calculated ifthe H2 larger amounts of H2 with failed completely bums. These are containment spray. The optimal then mapped to site-specific time to estimate the impact of containment failure Hydrogen burn is approximately at probabilities.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> which is the time when the EOF and TSC personnel have A simplifying assumption is convened and are ready to guide made that "no pre-burning of the Main Control Room into hydrogen generated in the periodic Hydrogen burns before core melt progression is the formation of explosive considered." Calvert Cliffs' mixtures. severe action management procedures do include Revision 4 Page 65 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension (This F&O originated from SR LE- actions to reduce H2 D1) concentration in the containment, but these actions are not credited in the PRA model. Also, Containment Spray is not questioned for the LERF accident sequences.

Containment Spray is a factor in LATE containment failure accident sequences.

REFERENCES CO-LE-001 5-31 DA-D4 Data The summary table for Bayesian Complete The aforementioned No impact on the updated parameters (on Page 53 footnote was incorporated ILRT analysis for of the PRA Data Notebook, CO- into Table 2-6 of CO-DA-001. this minor DA-001, Rev. 1) shows the CS- internal events MDP was Bayesian updated with REFERENCE documentation plant experience containing 1 CO-DA-001 issue and no failure and Zero run-hours. The changes were CCNPP PRA staff responded to required for the this issue as an isolated case. CS-MDP failure There is an actual FTR > 1 hr rate.

(This F&O originated from SR DA-D4) 6-3 SC-B2 Success Expert judgment was not used as Complete The approach for SLOCA The existing Criteria the sole basis for any success break size analysis is analysis meets criteria. However, upon inspection discussed in the Success the intent of the of the PCTran run tables in the SC Criteria notebook. SR and therefore report appendices, many instances Furthermore, a review was there is no Revision 4 Page 66 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension of surrogate or inferred results conducted of this issue; in impact on the were found. Instead of running addition, TH analyses were ILRT analysis.

specific PCTran calculations to completed to verify the cover the whole SLOCA break size break-size ranges. It was spectrum, intermediate break sizes found that the computer have been calculated simulations adequately supplemented with expert represented the various judgment to derive limiting time break-size ranges.

delay for operators to actuate SI (30 min) or limiting time delay for REFERENCE OTCC (SGL<350'+10min). C0-SC-001 (This F&O originated from SR SC-B2) 6-5 SY-A20 Systems When appropriate, the Complete AFW basic event No impact on Analysis simultaneous unavailability within a AFWOTMMAINT6-F7 was ILRT analysis.

system is documented in the determined to not be needed The offending system notebooks and included in in the plant model. The basic basic event was the PRA model. However, a further event was removed. All removed from the review of these items is required remaining AFW equipment model. A review for completeness. unavailability events in the did not discover model and notebooks were other missing or (This F&O originated from SR SY- reviewed for consistency. incorrect A20) AFWOTMMAINT-TF was simultaneous determined to be modeled unavailability correctly, its description was events.

found to be in error in the system notebook. Notebook CO-SY-036 was updated. A review for concurrent maintenance was previously performed and documented in the Data Notebook.

Revision 4 Page 67 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to ILRT Extension REFERENCE CO-SY-036 6-8 HR-H2 Human Some recovery actions included in Complete For each screening HRA, No impact on Reliability the model (thus credited) are set to the internal events analysis ILRT analysis.

screening values. In the HEP was updated to include a The evaluation (appendices of the HR specific reference to the documentation report) there are no indications that earlier HRA analysis. for internal procedures, training, or other Included are the applicable events HRAs shaping factors are available on a success criteria for each was updated to plant-specific basis. recovery. Refer to CO-HR- address this 001, Internal Events Human finding.

(This F&O originated from SR HR- Reliability Analysis, and the H2) associated HRA Calculator file.

For Fire PRA development, the internal events HRAs with screening values were analyzed to assure that they were sufficiently conservative for fire scenarios. Refer to Section 4.1 of CO-HRA-001, Fire PRA Human Reliability Analysis. Documentation for fire HRA actions are similar Page 68 of 110 Revision 44 Revision Page 68 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension to that done for the updated internal events HRA actions.

REFERENCE CO-HR-001 CO-HRA-001 6-9 HR-I1 Human The HR report is well documented Complete Updated the notebooks in No impact on Reliability in general and will facilitate the reference section so Fire PRA. This is upgrades, however, some basic HRA designator names and a documentation event names are not consistent designar name an finding. HRA HR report and the between the system the HR Calteboorks.Rmodel Calculator, HR in the names and system notebooks. notebook, CAFTA Model notebook are (This F&O originated from SR HR- 6.0. Changes included notebonsiste I1) adding the "-B" extension now consistent.

and removing the "(-2)"

event where applicable.

REFERENCE CO-HR-001 CO-SY-[Many]

6-10 IFPP-A2 Internal Plant design features such as Complete The Internal Flood notebook No impact on IFSN-A2 Flooding open rooms or as built divisions has been updated to ILRT analysis.

are used to define the flood areas incorporate an analysis This is a and was well documented. More describing the screening of documentation detail is needed as to why the the containment building finding for the containment buildings were from flooding analysis. Internal Flood screened from the analysis. Essentially, the containment notebook.

is designed for LOCA condition, which screens Revision 4 Page 69 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension (This F&O originated from SR reactor coolant system and IFPP-A2) related piping system. Other piping systems have limited inventory, are normally isolated, or have a low flow rate. Reference CO-IF-001.

REFERENCE CO-IF-001 6-14 IFSO-B1 Internal While the flooding calculations Open This is a documentation No impact on IFSN-A9 Flooding have been performed and are finding for the internal floods ILRT analysis.

thought to be correct and well notebook. The issue has This is a done, additional documentation of been captured in the PRA documentation data would enhance the IF report. configuration control issue.

It appears that the input reports database (CRMP), but not and references are based on yet closed-out.

poorly documented or non-officially revisioned reports and information sources.

(This F&O originated from SR IFSN-A9) 6-16 IFQU- Internal Walkdowns have been conducted Open This is an internal floods No impact on All Flooding and are documented in Appendix documentation finding. The ILRT analysis.

IFPP-B2 B of the IF report. It is stated in the finding has been captured in This is a IF report that prior information is the PRA configuration documentation no longer available; this fact control database (CRMP), issue.

should be corrected as required for but not yet addressed.

analysis updates and information verifications.

Revision 4 Page 70 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension (This F&O originated from SR IFQU-A1 1) 6-17 IFQU- Internal By including the flooding events Open The level of modeling detail No impact on A10 Flooding under the transient fault tree, the in the CCPRA is sufficiently ILRT analysis.

LERF impacts are automatically robust such that the model This is a accounted for in the same manner logic for flood impacts documentation as the general transient events in propagate appropriately issue.

the LERF analysis. Very little through the system fault documentation is found related to trees so that the equivalent the IF analysis in the LE report, general transient initiator although the IF report states that (e.g loss of CCW) is the LERF impacts due to flooding appropriately defined in the are documented and analyzed in transient fault tree. In the LE report. Therefore, SR addition, cutset reviews IFQU-A10 was determined to be have not revealed the not met. current modeling to be deficient in this regard.

(This F&O originated from SR This documentation of the IFQU-A10) above basis is captured in the PRA configuration control database (CRMP),

but not yet addressed.

6-18 HR-H2 Human The system time window Tsw for Complete Itwas determined that the No impact on Reliability post initiator HRAs was frequently text in Section 3.1.5.7 was ILRT analysis. As associated with 'core damage'. incorrect and does not described in this Post initiator HRAs that appear in capture how stress is F&O for internal the top cutsets may require actually applied in the EPRI events, the stress success criteria linked to beginning HRA Calculator. CO-HR-001, levels in the of core uncovery (about 20 Internal Events PRA Human model are minutes before 'core damage'). Or, Reliability Analysis, has appropriate, but the operator actions that may fall been updated to show the updates to the into that final 20-minute time stress level applied to each documentation Revision 4 Page 71 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension period should be overridden to HFE and the justification for are required. The assume a high stress level. While stress selection. Also internal events section 3.1.5.7 described this included is a correlation documentation approach, there is no evidence of between stress level and was updated.

its proper application in the HRA failure of execution quantifications. probability. New text has been provided for inclusion (This F&O originated from SR HR- in a future update of the H2) HRA notebook.

For the Fire PRA development, the internal events HRA stress levels were carried forward. As described in CO-HRA-001, Fire PRA Human Reliability Analysis, additional stresses were evaluated and incorporated due to the fire initiator.

REFERENCE CO-HR-001 CO-HRA-001 6-22 HR-El Human Upon RAS, LPSI stops and EOP- Complete As documented in CR-2009- No impact on Reliability 5, Step S.1(d) requires the 005881, shutting the RWT ILRT analysis.

Operators to 'Shut RWT OUT outlet valves upon a RAS The system is Valves SI-4142, 4143'. This does not impact station operable without manual action was not modeled in operability. The Safety the manual the PRA. The CCNPP PRA staff Injection Pumps and action to shut the provided reasonable response to Containment Spray Pumps RWT outlet this issue. Based on CR-2009- will not fail ifthe RWT valves. There is 005581, there is no impact on isolation valves do not close no impact on Revision 4 Page 72 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension pump operability. Also, the staff with a RAS signal. A design internal events will continue to track the CR. If margin issue has been CDF. The issue there are any changes to the identified. This issue has was added to the disposition of pump operability, been added to the plant's plant's margin then a new HRA may be added to margin management management the PRA model (if warranted). program. No model changes program.

have been made, but the (This F&O originated from SR HR- PRA configuration El ) management program, CNG-CM-1.01-3003, would capture any design changes concerning this issue.

REFERENCE C0-SY-052 CR-2009-005881

  • CNG-CM-l.01-3003 6-23 HR-G7 Human When the Calculator reads in the Open New HRA events, No impact on Reliability combinations, it assumes that CVC0HFBHEOTA-B-8HRS ILRT actions occur in the order of the and AFWOHF-CC-SGDEC- analysis. The time delay (Td). However, the time 8HR were added to model new HRA events delay is not the same for all Td variances where CST are not sequences, and care must be depletion occurs early and significant.

taken to make the combinations when it occurs later. This appropriate for the sequences in accounts for appropriate which they occur. Page 88 of the sequencing of events.

HRA notebook indicates this was considered, since the Td was This specific issue with time modified for events occurring prior delay and CST depletion to reactor trip, and also for OTCC has been addressed and after SG overfill. However, not all incorporated into the PRA occurrences have been model. An updated addressed. The combination dependency analysis has Revision 4 Page 73 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to ILRT Extension examined by the review team is been performed, which Combination 770 (OTCC after CST includes these new HRA depletion). In this event the CST events. The dependency depletion should come first. analysis shows that these new HRA actions are not (This F&O originated from SR HR- significant for CDF or G7) LERF. A PRA configuration control database (CRMP) item has been initiated to formally incorporate the updated dependency analysis into the model.

REFERENCE CO-HR-001 CO-HRA-001 7-13 QU-A2 Quantification Discrepancy between Complete The top flood cutset was No impact on documentation and result files. incorrectly flagged as being ILRT analysis for SB0037 and SB0038 sequences SBO sequence 37 (offsite this internal appear to be inverted in Tables D- power recovered < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) events 1, 4.2.2, 4.2.4, 4.2.5, B-3). instead of sequence 38 documentation (offsite power not issue.

(This F&O originated from SR QU- recovered). Updated tables A2) B-2, C-1, and D 1 in CO-QU-001. Spot-check was performed to identify other errors. In CO-QU-002, fixed sequence 12 table 4.2-5, which incorrectly showed sequence 37 instead of 38.

Page 74 of 110 Revision 44 Page 74 of 110

IRCA-54001-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to ILRT Extension REFERENCE CO-QU-001 CO-QU-002 Page 75 of 110 Revision 44 Page 75 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis PP- PP-B3 Plant Complete The containment is partitioned CO-PP-001, Calvert Cliffs Fire PRA Plant No impact to B3- PP-B6 Partitioning into 2 PAUs. There are Partitioning Notebook, was updated to ILRT analysis, 01 PP-C3 intervening combustibles and include an analysis that justifies the as this affects this was accounted for in the partitioning of the containment into two the FPRA plant PRA by treating the 20 feet as plant partitioning units with a 20-foot spatial partitioning an overlap region and failing separation (known as the buffer zone). The analysis.

components affected in both only potential intervening combustibles in PAUs. There is no justification this buffer zone were identified as qualified given for the 20 foot cables that were verified to be encased assumption. The turbine deck within marinate covered raceways. The is continuous from unit I to unit covers prevent the cables from becoming

2. This area is divided into 2 potential combustibles and therefore are PAUs, TURB1 and TURB2, but not considered intervening combustibles.

there is no discussion for the basis of the partitioning. The unit 1 and unit 2 Turbine Deck was Finding level of significance is walked down to assess for the acceptability baseparaon wredithino r site of the Appendix R partitioning into distinct separation with no requisite PAUs. The boundary was assessed to justification. have at least a 20-foot separation between Maintain the containment as 1 potential ignition sources and potential PAU and discern the targets, assessed for intervening separation of east from west in combustibles, and the Turbine deck the fire modeling. Document volume assessed for damaging hot gas Sthe spatial separation and no layer development. The partitioning was intervening combustibles for found acceptable and consistent with the turbine deck. NUREG/CR-6850, Section 1.5.2, where main turbine decks are typical applications where spatial separation has been credited.

PP- PP-B5 Plant Complete The water curtain in the CCW The Component Cooling Water room water No impact to B5- PP-C3 Partitioning room was credited as an active curtain is an approved Appendix R ILRT analysis, 01 fire barrier. The justification exemption, as identified in the exemption as this affects Revision 4 Page 76 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis was that the water curtain was issued by the NRC in response to Calvert the FPRA plant part of the original regulatory Cliffs exemption request ER820816. The partitioning fire protection program. This validity of crediting CCW Room Water analysis.

meets CAT 1, but needs Curtains is discussed in Southwest enhancement for CAT Il/111. Research Institute Report No. 01-0763-Finding level was used 201. A reference to the Southwest because the requirements for Research Institute report was added to CO-CAT Il/111 were not met. PP-001, Plant Partitioning Notebook.

Calvert Cliffs should provide a direct reference to their Appendix R program as the basis for the acceptability for this or provide a design basis justification for the water curtain and document that in the PP notebook ifthe Appendix R program reference cannot be found.

PP- PP-B7 Plant Complete 1. The walk down A table was created to correlate the No impact to B7- PP-C3 Partitioning nomenclature does not match building or area nomenclature that was ILRT analysis, 01 PP-C4 Qualitative the PP notebook. Example used for the plant walkdown as this affects QLS- Screening page 561 of the walkdown documentation, to the plant analysis unit the FPRA plant documentation uses identifiers used in the Fire PRA analysis. partitioning Al nomenclature in the This table was added to CO-PP-001, documentation.

containment that does not Calvert Cliffs Fire PRA Plant Partitioning match the PP notebook. Notebook as Table 17.

2. There are many areas inaccessible such as: #23 The facilities and rooms that were not Charging Pump Room, U1 originally walked-down were reviewed.

Service Water Pump Room, Supplemental walkdowns were performed UI East Battery Room, E/W and supplemental walkdown datasheets Revision 4 Page 77 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis Corridor. These areas appear were generated. For areas that were not to be accessible with a little accessible at the time of the supplemental effort. In some of the areas walkdowns (for radiological safety reasons, screened out in QLS, the areas personnel safety concerns, or access were inaccessible and did not otherwise denied), The reason for have a confirmatory walkdown. inaccessibility was added to Table 17.

Finding level assessed due to the incompleteness of the walkdown documentation.

1. Prepare a table that correlates the PAUs from the PP notebook with the area nomenclature used in the walkdown documentation.
2. Complete the walkdowns, particularly for areas screened in the QLS task.

CS- CS-B1 Fire PRA Complete Current Breaker coordination The breaker coordination study has been No impact to B1- CS-C4 Cable study still in progress. This completed. As described in ECP ILRT analysis, 01 Selection study needs to be completed in 000321, Form 12, Engineering Evaluation, as this affects and order to receive a category II all PRA common power supplies are the FPRA plant Location met for CS-B1. assumed to meet - or will meet - the Cable coordination requirements of NFPA 805, Selection Complete the breaker except as noted in C0-CS-001, Fire PRA analysis and coordination study. Cable Selection Notebook. As described in the item has the cable selection notebook, two 120VAC been lighting panels are not validated as completed.

coordinated, and these panels are assumed to fail for all Fire PRA scenarios.

Also, as described in the PRA notebook a breaker for 480V motor control center Revision 4 Page 78 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis MCC101BT has not been validated as coordinated. This breaker, 52-10150, is modeled so that a fire-induced electrical fault on the breaker's power cabling will fail MCC101BT. Finally, the notebook identifies that selected 120V power panels have coordination issues, but that these will be addressed by design changes and referenced in Attachment S - Modifications and Implementation Items.

PRM- PRM- Fire Complete The FPRA model did not Loss of Control Room HVAC can affect the No impact to B3- B3 PRA/Plant address events involving loss operability and availability of equipment in the ILRT 01 PRM- Response of both HVAC trains to the the control room and cable spreading analysis, as the B4 Model MCR, long term heatup of room. As described in Calvert PRA System loss of MCR MCR and need for operator Analysis Notebooks CO-SY-002, CO-SY- HVAC PRM-actions outside the MCR to 017, and CO-SY-030, loss of HVAC is modeling has B5 compensate for the loss of modeled to have the effect of increasing been electronic controls in the MCR, the failure rate of 120VAC and 125VDC implemented in which was assumed as a instruments and controls in the cable the models CCDP of 1.0 for the plant. The spreading room. For the control room, used in the basis for excluding this degradation of the 125VDC system is used ILRT analysis potential Core Damage as a conservative surrogate for control sequence was addressed in room I&C degradation.

questions to the Calvert Cliffs PRA team. This sequence is a Loss of Control Room HVAC and new sequence outside the current FPRA model logic subsequent temperature increases may adversely affect operator responses. The trees.

model reflects degradation of human actions by the degradation of the 125VDC Consider using a combination system used for instruments and controls.

of MCR heatup calculations to Loss of Control Room HVAC is not define the time when operators would leave MCR and consider expected to cause abandonment by Page 79 of 110 Revision 4 4 Page 79 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis a recovery action for restoring operations staff of the control room due to cooling the MCR. high temperatures. On complete loss of HVAC with no mitigation, such as no use of emergency fans, calculation CA02725 shows a CR temperature of 123 deg F at 24-hours. While this is a challenging environment, this temperature is assessed as insufficient to solely drive a complete CR abandonment scenario. NUREG/CR-6738 describes operational experience where operators will continue to occupy the control room even under severe environments.

Operations staff says that in consideration of high temperatures in the control room, that Operations would do what was needed to keep the cores safe and covered. The site safety director says that for a temperature of 123 deg F, the site would implement a mitigation strategy which would include stay-times, assessment of individuals for heat-related conditions, use of ice vests, and call-in of additional qualified operations staff to rotate into the control room.

The above discussion was included in CO-SY-030, Control Room HVAC PRA System Notebook.

Page 80 of 110 Revision 44 Page 80 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis FSS- FSS- Fire Complete A range of ignition source / FDS modeling was used for fire scenario No impact to A5- A5 Scenario target set combinations has evaluations in the Cable Spreading Rooms the ILRT 01 Selection been represented for and Switchgear Rooms. In both cases, analysis, as and unscreened PAUs. These thermocouple location was adjusted as this affects the Analysis combinations are identified in identified in F&O FSS-D3-02. For the CSR, FPRA model relevant calculation sheets for consequences were divided into scenarios and the item is unscreened PAUs. In some based on mitigation potential. First, if the complete.

PAUs, sub-PAUs are defined scenario was suppressed by the Halon and damage from a potential system then the limit of damage was based fire within the sub-PAU is on what was predicted by FDS in terms of addressed. However, it is not temperature and energy. If it was clear how or why damage unsuppressed it went to total room burn, would be limited to the which assumes failure of all targets in the specified sub-PAU because room, regardless of the initial scenario there are no physical barriers boundary. For the Switchgear Room FDS between specified sub-PAUs. analysis, the analysis was updated to add The documentation is such clarity to the analysis. A discussion of the that it cannot be determined if application of sub-PAUs has been added to the selected fire scenarios Addendum 1 to CO-FSS-004, Fire PRA provide reasonable assurance Detailed Fire Modeling Notebook. Damage that the risk contribution of was not limited to specified sub-PAUs.

each unscreened PAU can be Specific examples of the treatment of fire characterized. Another issue growth and the application of sub-PAUs that influences the potential for have been provided.

fire propagation across sub-PAU boundaries is that the temperature measurement As described in CO-FSS-004, the sub-PAU analysis included spatial information from locations specified in the walkdown, along with engineering detailed FDS fire modeling judgment, to determine if fire sources could evaluations do not generally fail additional components, cables, or other coincide with locations where combustibles, potentially leading to more maximum temperature are damage to surrounding equipment or cables. For scenarios that leveraged FDT Page 81 of 110 Revision 44 Page 81 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis expected (e.g., within the fire modeling, the issue related to whether the plume). analysis had correctly addressed the impact of transients along the edge of a consequence, for some boundary interface for a sub-PAU. A fire a csecenarios dame tom As comparable consideration was also related fire scenarios damage to to secondary combustion and oil fires.

targets is not predicted when it Resolution involved selection of several shoufiedb mased oither. representative PAUs for a sensitivity study specified damage criteria, that expanded the existing sub-PAUs and Somthe bascios artepercreed examined secondary ignition potential.

on the basis of temperature measurements that do not represent conditions at targets within the fire plume. (See F&O FSS-D3-02) This could have a significant impact on the potential for fire propagation across sub-PAU boundaries and needs to be discussed more thoroughly.

FSS- FSS- Fire Complete There were indications that The PAUs were considered representative No impact to A5- A5 Scenario Calvert Cliffs had the tools and of the work performed based on several the ILRT 01 Selection information in place to properly criteria. The analysis indicated that the analysis, as and evaluate the propagation of methods mentioned were indeed this affects the Analysis fires across the sub-PAU appropriate. Sub-PAU impacts did not FPRA model boundaries given no physical change from the expanded assessment and the item is barriers but there were no and that secondary ignition was bounded complete.

examples showing that this by the existing analysis and was evaluation was performed or appropriately addressed. The analysis was any explicit descriptions of how incorporated into the documentation for they were performed in CO-FSS-004.

general. The concern here is that without an explicit Revision 4 Page 82 of 110

1RCA-54001-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis description of the process for evaluating the spread of fires across sub-PAU boundaries with no physical barriers and detailed examples, there is the potential that in the future, new people updating the PRA may not know that they have to evaluate this.

Calvert Cliffs needs to describe their process for evaluating fire growth and propagation between sub-PAUs and as applicable, between PAUs.

Specific examples of the sub-PAU fire growth need to be provided. Iffire propagation from sub-PAU to sub-PAU was not treated, Calvert Cliffs needs to evaluate all sub-PAUs to determine ifthere is any potential for fire spread and then model the potential for spreading fires and for damage occurring across sub-PAU boundaries.

FSS- FSS- Fire Complete Where used, the FDS model FDS modeling was used for fire scenario No impact to D2- D2 Scenario was generally used with a level evaluations in the Cable Spreading Rooms the ILRT 01 Selection of grid resolution that was and Switchgear Rooms. analysis, as and below the level of grid this affects the Analysis resolution documented in the FPRA model Revision 4 Page 83 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis NUREG-1824 Verification and For the Cable Spreading Room FDS fire and the item is Validation study for the FDS scenarios, a grid study was performed on complete.

model. A validation study was the updated FDS model. The study not conducted to support the recommended a grid size that was within use of this lower level of grid the range in NUREG/CR-1824. That grid resolution. Grid resolution has size was used for CSR FDS scenario a bearing on the results of FDS evaluations. The study and results were calculations. Grid resolutions incorporated into C0-FSS-004, Fire PRA outside the validation range in Detailed Fire Modeling Notebook.

NUREG-1824 should be justified and validated. The Unit 1 27' and 45' Switchgear Rooms Increase the level of grid were updated to increase the level of grid Inreaselutin thell o g Uid e resolution to a value that is within the resolution in the FDS PAU Fire vldto ag ouetdi Evaluations (C0-FSS-004 R1) validation range documented in EvthaluationsidCreFolutioRi) NUREG/CR-1824. Results calculated in within the validation range the Unit 1 FDS models were applied to Unit documented in NUREG-1824. 2. Results of the updated model are incorporated into C0-FSS-004 as Addendum 1.

FSS- FSS- Fire Complete This SR is not met because FDS modeling was used for fire scenario No impact to D3- D3 Scenario detailed FDS fire modeling evaluations in the Cable Spreading Rooms the ILRT 01 FSS- Selection evaluations of PAUs 302, 306, and Switchgear Rooms. analysis, as B2 and 311, 317,407 and 430 assume this affects the FSS- Analysis that material surfaces are For the Cable Spreading Room FDS fire FPRA model 04 "inert." As noted on p. 44 of scenarios, the Unit 1 CSR was modified to and the item is CO-FSS-004 R1, this include actual material properties and complete.

assumption was made "... so sensitivity analysis. Actual material the PAU structure (walls, floor, properties were used in the updated or ceiling) itself would absorb U1CSR FDS model rather than the prior use of "inert" material conditions. Adiabatic any heat from the various fire conditions were used for any items with scenarios, producing a more material properties that are unknown or of Revision 4 Page 84 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis conservative or worst case a high uncertainty to bound the analysis result for all fire scenarios' and prevent heat transfer into those impacts to the components objects. The CSR FDS model was and cables within the PAU executed and the results compared to the model. As such, no detailed baseline results. This study was then material properties were documented in FSS-004. The results were required to be defined in FDS applied to Unit 2 CSR. This study was then for the scenarios to function documented in FSS-004, Fire PRA correctly." However, Detailed Fire Modeling Notebook.

specification of material surfaces as "inert" in FDS does The Unit 1 27' and 45' Switchgear Rooms not prevent heat absorption into material surfaces. On the were updated to specify representative material properties as referenced by contrary, this specification NUREG 1805. This adjustment enabled the maintains material surfaces at analysis to obtain more realistic estimates ambient temperature in FDS, which tends to maximize heat of environmental conditions for these fire absorption into these surfaces. scenarios. Results calculated in the Unit 1 FDS models were applied to Unit 2.

To prevent heat absorption into Results of the updated model are material surfaces, they should incorporated into CO-FSS-004 as have been specified as Addendum 1.

"adiabatic" rather than as "inert." The "inert" parameter in FDS maximizes heat transfer to surfaces rather than minimize it. This can result in lower calculated gas temperatures.

Specify materials surfaces as "adiabatic" rather than as "inert" in FDS to prevent them from absorbing heat in order to achieve the stated goal of Page 85 of 110 Revision 44 Revision Page 85 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis producing a more conservative or worst case result. This may prove to be overly conservative, in which case specification of realistic material properties could be used to achieve more realistic estimates of environmental conditions for these fire scenarios.

FSS- FSS- Fire Complete Temperature measurement FDS modeling was used for fire scenario No impact to D3- D3 Scenario locations specified in the evaluations in the Cable Spreading Rooms the ILRT 02 FSS- Selection detailed FDS fire modeling and Switchgear Rooms. analysis, as A5 and evaluations do not generally this affects the Analysis coincide with locations where FPRA model maximum temperature are For the Cable Spreading Room FDS fire and the item is scenarios, new measurement devices were expected (e.g., within the fire complete.

included in the updated U 1CSR FDS plume). As a consequence, for model. The thermocouples were placed some fire scenarios damage to targets is not predicted when it directly above the fire source in the updated FDS model and the scenarios re-should be based on the specified damage criteria. evaluated. The results were applied to Unit 2 CSR. This study and the results were Some scenarios are screened on the basis of temperature then documented in FSS-004, Fire PRA measurements that do not Detailed Fire Modeling Notebook.

represent conditions at targets within the fire plume. The Unit 1 27' and 45' SWGR rooms were updated to alter the location of the Re-run FDS simulations with thermocouples such that the centerline temperature measurement plume temperature was recorded and used probes located within the fire to determine target impacts. Results Page 86of110 Revision 44 Page 86 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis plume or use other fire calculated in the Unit 1 FDS models were modeling tools such as FDTs applied to Unit 2. Results of the updated to calculate fire plume model are incorporated into CO-FSS-004 temperatures for these as Addendum 1.

scenarios.

FSS- FSS- Fire Complete Fire detection timing is FDS modeling was used for fire scenario No impact to D8- D8 Scenario evaluated for detailed fire evaluations in the Cable Spreading Rooms the ILRT 01 Selection modeling cases that use FDS. and Switchgear Rooms. analysis, as and This fire detection timing is For the updated Cable Spreading Room this affects the Analysis then used to estimate FDS fire scenarios, cable tray obstructions FPRA model automatic fire suppression were placed in the ceiling area of the and the item is timing and fire brigade updated U1CSR FDS model. Additional complete.

response timing for these thermocouple and heat flux data recording scenarios. However, the fire devices were added to the U1CSR model detection timing is based on under the new cable tray obstructions in modeling that does not include the vicinity of the fire source. The scenarios obstructions located beneath were re-evaluated. The results were the ceiling that could have an applied to Unit 2. A sensitivity study was impact on fire detector also performed. The study and new response. The fire detection scenario results were incorporated into CO-timing is also based on an FSS-004, Fire PRA Detailed Fire Modeling unjustified assumption Notebook.

regarding the type of smoke detectors installed in the affected PAUs. Obstructions to The Unit 1 27' and 45' SWGR rooms were the flow of fire gases can have also updated to include significant an impact on smoke obstructions such as cable trays and beam concentrations and velocities, pockets within the switchgear rooms.

which in turn influence smoke Results calculated in the Unit 1 FDS detector response. Without models were applied to Unit 2. Results and including such obstructions in details of this analysis are documented in fire modeling simulations, their CO-FSS-004 as Addendum 1.

Revision 4 Page 87 of 110

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis impact on fire detection times is not evaluated.

Include obstructions located beneath the ceiling for the affected fire scenarios in order to evaluate their impact on fire detection timing. Provide justification for the selection of the type of smoke detector specified in the FDS simulations for these fire scenarios.

FSS- FSS- Fire Complete To achieve CC Il/111 for this SR, The Turbine Building was reviewed for No impact to F3-01 F3 Scenario a quantitative assessment of potential fire scenarios where structural the ILRT Selection the risk of the selected fire steel can be adversely affected. From the analysis, as and scenarios involving a) exposed scenarios examined, those that can this affects the Analysis structural steel and b) the damage structural steel were selected for FPRA model presence of a high-hazard fire further analysis. The frequency, severity and the item is sources must be completed factor and non-suppression probability of complete.

consistent with the FQ each scenario were developed and requirements including the included in the Structural Failure Analysis collapse of the exposed Notebook. These impacts were then added structural steel and any to FRANX database and quantified as part attendant damage. Such an of the final Fire PRA risk quantification in assessment has not been Fire Quantification Notebooks CO-FRQ-001 done or was not documented and CO-FRQ-002.

in a readily discernible manner.

This has a potential impact on fire risk quantification.

Revision 4 Page 88 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis Complete a quantitative assessment of the risk of the selected exposed structural steel fire scenarios consistent with the FQ requirements.

FSS- FSS- Fire Complete An assessment of the Generic probabilities were used for No impact to G4- G4 Scenario effectiveness, reliability and credited passive fire barrier features in the the ILRT 01 Selection availability of credited passive multi-compartment analysis. At Calvert analysis, as and fire barrier features has not Cliffs, the fire barriers are verified to be this affects the Analysis been documented in the multi- effective through test procedures. An FPRA model compartment analysis. To unreliability value was applied to all and the item is achieve a CC II capability normally closed doors that represents the complete.

assessment, the effectiveness, probability of the door being propped open reliability and availability of given a fire in the exposing compartment.

credited passive fire barrier The probability of finding a failed sealed features must be assessed. wall penetration is assumed to be very small to warrant propagation scenarios. A discussion of the effectiveness, reliability, Assess the effectiveness, and availability of fire barriers was added to reliability and availability of CO-FSS-008, Calvert Fire PRA Multi-credited passive fire barrier Compartment Analysis.

features and document this assessment.

FSS- FSS- Fire Complete The effectiveness, reliability Active fire barriers were evaluated as No impact to G5- G5 Scenario and availability of credited effective in studies used to support the ILRT 01 Selection active fire barrier features have Appendix R analysis. An unreliability value analysis, as and not been quantified in the has been applied to all normally open, self this affects the Analysis multi-compartment analysis. closing dampers and doors; A discussion FPRA model To achieve a CC II capability of the effectiveness of credited active fire and the item is assessment, the effectiveness, barriers was added to CO-FSS-008, Calvert complete.

reliability and availability of Fire PRA Multi-Compartment Analysis.

credited active fire barrier Revision 4 Page 89 of 110

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance Of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis features must be quantified.

Quantify the effectiveness, reliability and availability of credited active fire barrier features and document this assessment.

HRA- HRA- Human Complete Improve documentation of the CO-HRA-001, Fire Human Reliability No impact to B2- B2 Reliability adverse operator actions notebook, was updated to detail the the ILRT 01 Analysis needed to address the impact adverse operator actions added to the analysis, as of grounded or shorted model following the fire AOP review this affects the electrical buses that might process. Table 3 was added to Section 2.2 FPRA model have an impact on other plant detailing each basic event, set to true (1.0) documentation buses if not isolated and re used in the model to annotate the adverse and the item is energized in the areas operator actions in the model. These complete.

identified. Very difficult to find include actions to de-energize electrical the information within the HRA busses to isolate them from potential notebook alone, because the shorts and grounds. Table 2 shows the actions are modeled as inputs HFEs added to the model as part of the to FRANX. AOP review, including actions to restore AC power to busses lost due to fire failure Provide new tables listing the sequences.

actions considered or references to specific locations.

HRA- HRA- Human Complete Documentation for what was CO-HRA-001, Fire Human Reliability No impact to El- El Reliability done was very good, however, Notebook, was updated detailing the Alarm the ILRT 01 Analysis the details for not selecting any Response Procedure review process. analysis, as spurious alarms is not clear. Table 12 was expanded to show the ARP this affects the The documentation of the review of alarm impact and operator FPRA model adverse actions put into the interview notes for CR annunciators that and Revision 4 Page 90 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis model as "true" are not in the could result in a manual reactor trip. No documentation HRA report, actions identified annunciators were identified that would and the item is in the cutset reviews are not cause the operator to terminate a systems complete.

clearly identified, rational for or components operation based solely on not using specific HFEs in the the alarm itself, but several were identified RCP trip actions, for identifying that could potentially result in the operator actions from procedures and tripping the Unit unnecessarily.

the process for assigning uncertainty range for the CO-HRA-001 was also updated to detail the combos. Doesn't permit adverse operator actions added to the verification of the rational for model following the fire AOP review judgments made in deciding process. Table 3 was added to Section 2.2 what is in and out of the Fire detailing each basic event, set to true (1.0)

HRA. Also, from the used in the model to annotate the adverse calculation viewpoint the need operator actions in the model. These to know the use of all include actions to de-energize electrical manpower requirements during busses to isolate them from potential early time after fire initiator for shorts and grounds. Table 2 shows the dependency analysis.

HFEs added to the model as part of the AOP review, including actions to restore Enhance documentation of the AC power to busses lost due to fire failure specific issues needed to sequences.

reproduce the assumptions and calculations used in the HRA. New HFEs added as part of the cutset review process are identified in Table 1 of C0-HRA-001, Fire Human Action Reliability notebook. These are annotated with "identified during the development of the PRM Notebook." The cutset reviews are described in CO-QNS-001, Fire PRA Quantitative Screening Notebook. A new dependency analysis was performed after the new HFEs were added to the model, Page 91 of 110 Revision 44 Page 91 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis ensuring new dependency combinations are considered.

Additional information was added to Table 1 of the Human Reliability Analysis Notebook, CO-HRA-O001, detailing why each HFE was either retained or removed.

For example, event FGAFWOSGTRISOL, Operator Feeds Affected SG with SGTR to Assure Heat Removal, was "Not retained for fire scenarios, because these actions are SGTR specific. Modeling was not necessary to ensure these actions did not appear in the cutsets, because the SGTR initiator is not being used for fire scenarios."

Combination event multipliers are used in cutsets of multiple HEP actions to account for dependencies between HEP actions. To account for the uncertainty in HEP actions, an uncertainty parameter is added to the HEP action. When performing uncertainty analysis, the uncertainty parameters for combination events is increased proportionally when they are multiplied by the combination event multipliers.

Based on interviews, there are sufficient non-control room personnel for fire recovery actions. Appendix D of CO-HRA-001 notes that there are no control room Revision 4 Page 92 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis operators assigned to the fire brigade.

There were no identified staffing issues or interferences between operators performing fire recovery actions and members of the fire brigade.

FQ- FQ-A1 Fire Risk Complete Treatment of 0 CCDPs The fire risk quantification process has No impact to Al- Quantificati scenarios is not clear and been updated in notebooks CO-FRQ-001 the ILRT 01 on appears to result in an and CO-FRQ-002 to address the issue with analysis, as underestimate of total risk (the FRANX fire scenarios having a zero this affects the underestimate appears to be conditional probability for CDF and LERF. FPRA model small based on the sensitivity and the item is evaluations performed): 1. When documented analysis shows that complete.

1 - with respect to opposite unit selected fire scenarios for one unit are quantification, use CCDP for screened from impact for the opposite unit reactor trip initiator unless (typically, no trip would be initiated), then confirmation of no trip is that scenario may be excluded from the documented; opposite unit's fire risk quantification.

2 - address use of 0 CCDP for Otherwise, a nominal conditional control room HVAC loss probability, as described in item 3 below, scenarios, apply CCDP would apply.

consistent with control room abandonment 2. F&O PRM-B3-01 identifies the concern 3 - for scenarios with limited with loss of Control Room HVAC with impact with a 0 CCDP, due to control room abandonment. As discussed cutsets below truncation limit, in more detail with the resolution to PRM-apply a baseline CCDP based 83-01, subsequent investigation revealed on reactor trip initiator that loss of CR HVAC is not expected to cause abandonment by the operations staff More than 50% of the of the control room due to high scenarios have a 0 CCDP but temperatures. Loss of CR HVAC and no clear discussion of the subsequent temperature increases may adversely affect operator responses, and Revision 4 Page 93 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis basis for the 0 CCDP is the model reflects degradation of human provided, actions with loss of CR HVAC. CO-SY-030, Control Room HVAC PRA System Treatment of 0 CCDPs Notebook, was updated to include this Trenatmtof 0discussion.

scenarios:

1 - with respect to opposite unit quantification, use CCDP for 3. The new quantification process reactor trip initiator unless described in the FRQ notebooks is to confirmation of no trip is assure a nominal conditional value is documented; calculated for these low significant 2- address use of 0 CCDP for scenarios by 1) recalculating the zero-scntrol, room HVACy losvalue conditional scenarios to assure resolution in the truncation at a lower scenario scenarios, apply CCDP cutset file and conditional probabilities, consistent with control room and/or to 2) use a baseline conditional abandonment probability for CDF and LERF for the 3 - for scenarios with limited internal events reactor trip initiating vent -

impact with a 0 CCDP, due to IEOPT for Unit 1 or IEOPT-2 for Unit 2 cutsets below truncation limit, apply a baseline CCDP based on reactor trip initiator FQ- FQ-B1 Fire Risk Complete We observed zero CCDPs for The fire risk quantification process has No impact to B1- Quantificati some PAU CDF and LERF been updated in notebooks C0-FRQ-001 the ILRT 01 on values in the FRANX tables and CO-FRQ-002 to address the issue with analysis, as (e.g., PAU 512) which FRANX fire scenarios having a zero this affects the eliminated loss of HVAC to the conditional probability for CDF and LERF. FPRA model MCR as a potential MCR and the item is abandonment sequence. 1. When documented analysis shows that complete.

Treatment of 0 CCDPs selected fire scenarios for one unit are scenarios: screened from impact for the opposite unit 1 - with respect to opposite unit (typically, no trip would be initiated), then quantification, use CCDP for that scenario may be excluded from the Revision 4 Page 94 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis reactor trip initiator unless opposite unit's fire risk quantification.

confirmation of no trip is Otherwise, a nominal conditional documented; probability, as described in item 3 below, 2 - address use of 0 CCDP for would apply.

control room HVAC loss scenarios, apply CCDP 2. F&O PRM-B3-01 identifies the concern consistent with control room with loss of Control Room HVAC with abandonment (F&O FQ-A1-01 control room abandonment. As discussed (F)) in more detail with the resolution to PRM-3 - for scenarios with limited B3-01, subsequent investigation revealed impact with a 0 CCDP, due to that loss of CR HVAC is not expected to cutsets below truncation limit, cause abandonment by the operations staff apply a baseline CCDP based of the control room due to high on reactor trip initiator temperatures. Loss of CR HVAC and Allowing zero CCDPs allows subsequent temperature increases may scenarios in the fire model to adversely affect operator responses, and quantify with no contribution to the model reflects degradation of human the CDF or LERF value and actions with loss of CR HVAC. CO-SY-030, this under represents those Control Room HVAC PRA System frequencies especially when Notebook, was updated to include this considering delta risk discussion.

evaluations.

3. The new quantification process Replace the zero entries with described in the FRQ notebooks is to the lowest CCPD for a plant assure a nominal conditional value is trip with only random failures of calculated for these low significant the safety equipment as in the scenarios by 1) recalculating the zero-internal events model. We conditional scenarios at a lower truncation discussed this with the Calvert value to assure resolution in the scenario Cliffs PRA team and some of cutset file and conditional probabilities, the zeros are due to fire areas and/or to 2) use a baseline conditional in one unit potentially probability for CDF and LERF for the contributing to the CCDP of the Revision 4 Page 95 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis opposite unit. With the internal events reactor trip initiating vent -

exception of these cases a IEOPT for Unit 1 or IEOPT-2 for Unit 2 method for handling the zeros needed to be developed and applied in the frequency quantifications.

Page 96 of 110 Revision 44 Page 96 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension B. ATTACHMENT 2 This attachment contains the release timing plots from NC-94-020 [Reference 18] for MAAP cases HRIF, GIOY, and MRIF. The MAAP release groups FRELI through FREL_2 correlate to the release categories shown in the figure below, and the MAAP outputs for each are shown in the figures on the following pages.

N(C 944"~

Dr. Mahmoud Massoud August 17, 1993 Revised Release Category Definitions Cataor Dscrnton :MAAP,',S 1 Noble Gases Noble Gases (Group 1) 2 Iodine CsI (Group 2) 3 cesium + Rubidium CsI + CsOH (Mole Fraction of Group 2 + Group 6) 4 Tellurium + Antimony TeO2 + Tej + Sb (Mole Fraction of Group 3, Group 11, + Group 10) 5 Strontium SrO2 (Group 4) 6 Ruthenium, Rhodium, MoO2 (Group 5)

Molybdenum,

+ Technetium 7 Yttrium, Lanthanum, La2O3 + Pr 2 O3 + Nd2O3 Zirconium, Niobium, + Sm2 O3 + Y2 03 Praseodymium, (Group 8)

Neodymium, Americium,

+ Curium 8 Cerium, Neptunium, CeOz + U02 + NpO2

+ Plutonium + PuO2 (Mole Fraction of Group 9 + Group 12) 9 Barium BaO (Group 7)

Figure 1 - Release Category Definitions from NC-94-020 [Reference 18].

Revision 4 rage 9/ oTyIu

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension W, qq -nip .;0 plot Oct. 7.1993 10:58:40AM (fission product; tape 35; page 1 of 5)

L2-HRIF, Loss of Main Feedwoter

.Ii I MW own 1 Nto I*"-

Figure 2 - MAAP Case HRIF Release Fractions Revision 4 Page 98 of 110

I Ri~A..~AAflI .Aflfl-(~AI C-fbi Evaluation of Risk Sianificance of Permanent ILRT Extension 4 PrA-94001.nOO-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension No q5-);,o plot Dec. 5. 1 993 5:39:04 PM (fission product: tape 35; page 1 of 5)

L2-GIOY, Smoll LOCiA,

/

I.0, CAI 4-s 0.4 0.6 I0.,1

_j,I 20-08

  • J 20-0
l. -0.2 30 a IQ=
  • 000 4-f-46 X.

I-Gm I

.w 30 0 I000 0000 300 40000

'WE-S Io0o 2o000 0 40 010 3E-A X0-6 6[-"

X0-1 01-1 IMI IE-6 20-11 IC-8 It-fl NOM0 "00o a00 ICB to 2 30000 0 am0 At-. NAILS TwL5 0,A 0.C A4.

.' 02*

CCO

-0.4

_a4 -0.4 30010 'Cm 2,0000 30000 4W0O0 " 0 0000 S300 400000 F 3C 1LS fo0-5 "TMrS Figure 3 - MAAP Case GIOY Relea* ,e Fractions Page 990? 110 Revision 4 Page 99 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Nq-o~o-3 i

plot OcI. 10.1993 5:31:38PM (fission product; tope 35; page 1 of 5)

L2-MRIF, Small LOCA IJ-f

-OA Ta" 9

I..

-I Figure 4 - MAAP Case MRIF Release Fractions Revision 4 rage iuu OT iiu Page I UU OTr- -1lU

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension C. ATTACHMENT 3 Probabilistic Risk Assessment Licensing Branch (APLA) RAI 1 In the safety evaluation report for Electric Power Research Institute (EPRI) Technical Report (TR) 1009325, Revision 2, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," the Nuclear Regulatory Commission (NRC) staff, in part, stated that for licensee requests for a permanent extension of the integrated leak rate testing (ILRT) surveillance interval to 15 years "[c]apability category I of ASME RA-Sa-2003 shall be applied as the standard, since approximate values of CDF and LERF and their distribution among release categories are sufficient for use in the EPRI methodology." of the license amendment request (LAR) states that the 2010 full scope peer review of the internal events PRA model identified three supporting requirements (SRs) that were 'Not Met'. Table 1 of Attachment 3 to the LAR lists the findings from the 2010 peer review, but does not identify the three "Not Met' SRs. Identify which SRs were considered not met. For each SR, summarize why not meeting Capability Category I requirements will have no impact on the ILRT extension application.

Response

The 3 SRs that were noted as "not met" are LE-F2, LE-G5, and IFQU-A10. LE-F2 relates to LERF results. The dominant LERF contributors were reviewed and model changes implemented prior to the ILRT analysis. LE-G5 relates to the documentation of limitations of applications of the PRA. IFQU-A1 0 relates to documentation of the treatment of the internal flood analysis in the event trees. Therefore, none of the "not met" SRs impact the ILRT extension analysis.

APLA RAI 2 Section 5.3.2 of Attachment 3 to the LAR uses the Calvert Cliffs Nuclear Power Plant methodology from 2002 in evaluating the impact of steel liner corrosion on the extension of ILRT testing intervals. This assessment was based on two observed corrosion events at North Anna Power Station, Unit 2 and Brunswick Steam Electric Plant, Unit 2.

a. Ifthere have been additional instances of liner corrosion that could be relevant to this assessment, provide an updated list of observed corrosion events relevant to Calvert Cliffs containment, and an evaluation of the impact on risk results when all relevant corrosion events are included in the risk assessment.

Response

A search of the NRG website LER database identified two additional events have occurred since the Calvert Cliffs analysis 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 containment liner. The other event occurred in April 2009, where a through-liner hole approximately 3/8-inch by 1-inch in size was identified in the Beaver Valley Power Station Unit 1 (BVPS-1) containment 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 liner. Two other containment liner through wall hole events occurred at Turkey Point 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 Revision 4 Page 101 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension coating system, which was not designed for periodic immersion service, and are not considered to be applicable to this analysis. More recently, in October 2013, some through-wall containment liner holes were identified at BVPS-1, with a combined total area of approximately 0.395 square inches. The cause of these 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 original construction and was contacting the steel liner. For risk evaluation purposes, these five 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 flaw 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.

b. Per EPRI TR-1009325, Revision 2, the risk metrics associated with the ILRT extension application include changes in large early release frequency (LERF),

population dose, and conditional containment failure probability (CCFP). The steel liner corrosion assessment in Section 5.3.2 of Attachment 3 to the LAR calculates only the change in LERF. Include an estimate of change in population dose and CCFP due to increase in steel liner corrosion likelihood and demonstrate acceptability of the risk results.

Response

An estimate of change in population dose and CCFP due to increase in steel liner corrosion likelihood are provided in the following tables.

Page 102 of 110 Revision 4 4 Page 102 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Unit I Steel Liner Corrosion Sensitivity CCFP CCFP CCFP CCFP CCFP CCFP CCFP Increase Increase Increase (3-per-10 (I-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)

Baseline 2.04E-01 2.09E-01 2.13E-01 5.11E-03 8.76E-03 3.65E-03 CCFP Corrosion Likelihood 2.04E-01 2.09E-01 2.13E-01 5.16E-03 8.84E-03 3.68E-03 X 1000 Corrosion Likelihood 2.04E-01 2.10E-01 2.14E-01 5.57E-03 9.56E-03 3.98E-03 X 10000 Corrosion Likelihood 2.06E-01 2.16E-01 2.23E-01 9.76E-03 1.67E-02 1.29E-02 X 100000 Unit 2 Steel Liner Corrosion Sensitivity CCFP 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-C0 to year ILRT) year ILRT) year ILRT) 1-per-10) 1-per-15) 1-per-15)

Baseline 2.01E-01 2.06E-01 2.09E-01 5.08E-03 8.71E-03 3.63E-03 CCFP Corrosion Likelihood 2.01E-01 2.06E-01 2.09E-01 5.13E-03 8.79E-03 3.66E-03 X 1000 Corrosion Likelihood 2.01 E-01 2.06E-01 2.10E-01 5.54E-03 9.50E-03 3.96E-03 X 10000 Corrosion Likelihood 2.03E-01 2.12E-01 2.19E-01 9.70E-03 1.66E-02 6.93E-03 X 100000 Unit I Steel Liner Corrosion Sensitivity Dose 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-M0 to year ILRT) year ILRT) year ILRT) 1-per-10) 1-per-15) I-per-15)

Dose Rate 3.63E-02 1.21E-01 1.82E-01 8.48E-02 1.45E-01 6.06E-02 Corrosion Likelihood 3.67E-02 1.27E-01 2.04E-01 9.08E-02 1.67E-01 7.61 E-02 X 1000 Corrosion Likelihood 3.96E-02 1.84E-01 4.01E-01 1.45E-01 3.61E-01 2.17E-01 X 10000 Corrosion Likelihood 6.94E-02 7.53E-01 2.37E+00 6.84E-01 2.30E+00 1.62E+00 X 100000 Page 103 of 110 Revision 4 Page 103 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Unit 2 Steel Liner Corrosion Sensitivity Dose Rate 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) I-per-15) I-per-15)

Dose Rate 1.96E-02 6.55E-02 9.82E-02 4.58E-02 7.86E-02 3.27E-02 Corrosion Likelihood 1.98E-02 6.89E-02 1.10E-01 4.91 E-02 9.02E-02 4.12E-02 X 1000 Corrosion Likelihood 2.14E-02 9.96E-02 2.17E-01 7.82E-02 1.95E-01 1.17E-01 X 10000 Corrosion Likelihood 3.75E-02 4.07E-01 1.28E+00 3.70E-01 1.25E+00 8.76E-01 X 100000 Page 104 of 110 Revision 44 Page 104 of 110

d IRCA-54001-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension APLA RAI 3 Section 4.2.7 of EPRI TR-1009325, Revision 2-A states that "[w]here possible, the analysis should include a quantitative assessment of the contribution of external events (for example, fire and seismic) in the risk impact assessment for extended ILRT intervals." EPRI TR-1 009325, Revision 2-A further states that the "assessment can be taken from existing, previously submitted and approved analyses or another alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval."

Section 5.3.1 in Attachment 3 to the LAR assesses the potential impact from external events contribution.

a. The results of the seismic PRA performed for the Individual Plant Examinations for External Events (IPEEE) were used to assess the seismic risk with a reported core damage frequency (CDF) of 1.07E-5/year for both Unit 1 and Unit 2. Section 8, "Summary and Conclusions," of the IPEEE report (Calculation No. RAN 97-031, IPEEE, Calvert Cliffs Nuclear Power Plant, "Individual Plant Examination of External Events," August 1997) reports seismic CDF values of 1.29E-5/year for Unit 1 and 1.52E-5/year for Unit 2. Justify the use of the 1.07E-5/year seismic CDF value in the external events sensitivity study.

Response

Data from Table 3-6 of the IPEEE Seismic Analysis is used to calculate a Class 3b frequency due to seismic. As noted in Table 3-6 of the IPEEE Seismic Analysis, the values given in Table 3-6 reflect quantification without the surrogate top event LA.

Top event LA represents seismic failure of rugged plant systems at a conservative screening fragility. Therefore, the total CDF is higher than the 1.07E-05/yr value given in Table 3-6. The CDF values given in Section 8.1 of the IPEEE Seismic Analysis are 1.29E-5/yr for Unit 1 and 1.52E-5/yr for Unit 2. The CDF contribution from surrogate top event LA was not included in the Unit 1 containment failure frequencies provided in the IPEEE (no containment failure frequencies are provided for Unit 2). In lieu of justification for the value of 1.07E-5/year seismic CDF, the contribution is conservatively added to the Unit 1 CDF of 1.07E-5/yr from Table 3-6 of the IPEEE to Containment Category I Failure (Intact). Note that the Intact category CDF is slightly rounded so that the total seismic CDF is preserved. Then, the percent each category contributes to the total CDF is calculated for the Unit 1 values and applied to the Unit 2 values because it is assumed that Unit 2 would have similar containment failure fractions to Unit 1. The resulting containment failure frequencies and total external events contribution using the revised IPEEE seismic CDF for each Unit are shown in the following tables:

Page 105 of 110 Revision 44 Revision Page 105 of 110

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Seismic Contribution to Frequencies of Containment Failure Categories Containment Failure Unit 1 Seismic CDF Percent of CDF Unit 2 Seismic CDF (/yr)

Category (/yr)

I. Intact Containment 2.69E-06 20.85% 3.17E-06 II. Late Containment Failure 8.63E-06 66.90% 1.02E-05 Ill. Early Small Containment 1.70E-07 to 1.27E-06 1.32% to 9.84% 2.00E-07 to 1.50E-06 Failure IV. Early Large Containment 3.13E-07 to 1.41E-06 2.43% to 10.93% 3.69E-07 to 1.66E-06 Failure V. Small Containment Bypass 0 0% 0 VI. Large Containment Bypass 0 0% 0 Total 1.29E-05 1.52E-05 CCNPP Unit I External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 3 per 10 year I per 10 year I per 15 years per15years)

External Events 5.13E-08 1.71E-07 2.57E-07 2.05E-07 Internal Events 1.14E-08 3.78E-08 5.68E-08 4.54E-08 Combined 6.27E-08 2.09E-07 3.13E-07 2.51E-07 CCNPP Unit 2 External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 3 per 10 year I per 10 year I per 15 years per 15 years)

External Events 6.52E-08 2.17E-07 3.26E-07 2.61E-07 Internal Events 6.14E-09 2.05E-08 3.07E-08 2.46E-08 Combined 7.13E-08 2.38E-07 3.57E-07 2.85E-07 Page 106 of 110 Revision 44 Page 106 of 110

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension

b. In Section 5.3.1.1 of Attachment 3 to the LAR the results from the IPEEE fire analysis were used to assess fire risk (CDF of 1.10E-5/year and LERF of 7.15E-7/year for Unit 2). Section 8, "Summary and Conclusions," of the IPEEE report (Calculation No. RAN 97-031, IPEEE, Calvert Cliffs Nuclear Power Plant, "Individual Plant Examination of External Events," August 1997) reports a Unit 2 fire CDF of 9.6E-5/year. Justify the use of selected IPEEE Unit 2 fire CDF/LERF values and discuss acceptability of Unit 2 risk results when using the IPEEE fire CDF/LERF.

Response

In lieu of justification of selected fire IPEEE CDF and LERF, the risk results have been revised using a Unit 2 fire CDF of 9.6E-5/year as given in Section 8 of the IPEEE. The Unit 2 containment failure frequencies and frequency 3b change are shown in the following tables:

Page 107 of 110 Revision 44 Revision Page 107 of I 10

.4 4.

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Unit 2 Fire Contribution to Frequencies of Containment Failure Categories Containment Failure Category Percentage Unit 2 Fire CDF (/yr)

I. Intact Containment 36.4% 3.50E-05 II. Late Containment Failure 55.5% 5.33E-05 Ill. Early Small Containment Failure 1.7% 1.58E-06 IV. Early Large Containment Failure 6.5% 6.13E-06 V. Small Containment Bypass 0.0% 0.OOE+00 VI. Large Containment Bypass 0.0% 0.OOE+00 Total 9.60E-05 CCNPP Unit 2 External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 3 per 10 year I per 10 year 1 per 15 years per 15 years)

External Events 2.40E-07 8.OOE-07 1.20E-06 9.61 E-07 Internal Events 6.14E-09 2.05E-08 3.07E-08 2.46E-08 Combined 2.46E-07 8.21 E-07 1.23E-06 9.88E-07 Revision 4 Page 108 of 110

I RCA-54001 -000-CALC-001 ' Evaluation of Risk Significance of Permanent ILRT Extension The internal event results are also provided to allow a composite value to be defined.

When both the internal and external event contributions are combined the total change in Unit 1 and 2 LERF meet the guidance for small change in risk, as it exceeds the 1.OE-7/yr and remains less than 1.OE-6 change in LERF for both units. For this change in LERF to be acceptable, total LERF must be less than 1.OE-5.

Conservatively using the highest seismic LERF value and not crediting containment spray success or plant damage state adjustments for the Internal Events PRA, the total LERF values are calculated below:

Unit 2: LERFU2 = LERFu2internal + LERFU2seismic + LERFU2fire+ LERFU2class3Bincrease

= 1.56E-6/yr + 1.66E-6/yr + 6.13E-6/yr + 9.84E-7/yr = 1.03E-5/yr The Unit 2 LERF is barely greater than 1.OE-5. However, the Unit 2 Seismic LERF is between 3.69E-07 and 1.66E-06, and the highest Seismic LERF value was conservatively used to calculate 1.03E-5. If the 74th percentile or smaller value of this range (<- 1.32E-6) is used, the total Unit 2 LERF is less than I.OE-5. Moreover, the IPEEE does not include recent significant plant modifications designed specifically to reduce fire risk. Therefore, it is reasonable to conclude that the total Unit 2 LERF is less than 1.OE-5. Since the total LERF for both units is less than 1.OE-5, it is acceptable for the ALERF to be between I.OE-7 and 1.OE-6.

APLA RAI 4 Section 5.2.4 of Attachment 3 to the LAR refines the calculation of the Class 3b frequencies for internal events by examining the source term. The conservatism in Class 3b frequency is reduced by analyzing the source term release time and defines an early release as occurring before 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, which allows the removal of three accident scenarios from the Class 3b frequency. Elimination of these three scenarios appears to reduce the Class 3b frequency for internal events by a factor of 3 to 5. Section 5.3.1 of Attachment 3 to the LAR indicates that the same approach is used in the calculation of the fire Class 3b frequency in the external events sensitivity study when using the NFPA-805 fire PRA.

a. Provide the calculated timing of the expected release for each of these three accident scenarios.

Response

The calculated timing of the expected release for the scenarios is as follows:

HRIF: 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> GIOY: 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> MRIF: 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br />

b. Provide the basis for the 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> delineation between early and late release.

Response

The 6.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> delineation between early and late release is based on the calculation of evacuation time estimates for Calvert Cliffs Nuclear Power Plant. Since early release timing is defined by time short enough that ability to evacuate nearby population is impaired such that a fatality is possible, and the calculation shows that Page 109 of 110 Revision 44 Revision Page 109 of 110

C.

IRCA-54001-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is sufficient for evacuation, this is chosen as the delineation between early and late release.

c. Explain whether releases from these scenarios were included in the analysis to calculate the increase in the total integrated dose risk for all accident sequences.

Response

The releases from these scenarios were excluded in the analysis to calculate the total integrated dose risk, since the scenarios are excluded from the class 3b contribution based on either containment spray success or the late timing of the release.

Page 110 of 110 Revision 44 Page 110 of 110

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