AEP-NRC-2020-40, Calculation PRA-QNT-009, Rev. 0, Evaluation of Risk Significance of Short-Term ILRT Extension

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Calculation PRA-QNT-009, Rev. 0, Evaluation of Risk Significance of Short-Term ILRT Extension
ML20164A046
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Site: Cook American Electric Power icon.png
Issue date: 05/22/2020
From: Sattler J
Jensen Hughes
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American Electric Power, Indiana Michigan Power Co, Office of Nuclear Material Safety and Safeguards
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AEP-NRC-2020-40 PRA-QNT-009
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  • Enclosure 3 to AEP-NRC-2020-40 Evaluation of Risk Significance of Short-Term ILRT Extension

D. C. COOK NUCLEAR PLANT C!JAMERICAN" EUCTRIC POWER

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CALCULATION/REPORT COVER SHEET D Full Rev D Addendum Document No. PRA-QNT-009 Rev No. 0 D Status Change

Title:

Evaluation of Risk Significance of Short-Term ILRT Extension STATUS: ~ Approved D Superseded 0 Voided I D Information Only Document Type/Class: I~ Calculation D Report ID Class 1 D Class 2 ~ Class 3 QUALITY SYSTEM UNIT COMPUTER REVIEW METHOD:

CLASSIFICATION: CODE: NO.: MEDIA: ~ Detailed Review D Safety-Related 0Yes 0 Alternate Calculation 0 Non-Safety Related with ~No 0 Other NAPL 12 Special Requirements 0 N/A - Status/Class Change Only

~ Non-Safety Related Do any assumptions require later verification? 0Yes ~No / If yes, Action No.

Description:

The purpose of this analysis is to provide a risk assessment of temporarily extending the currently allowed containment Type A Integrated Leak Rate Test (ILRT) interval by 6 months from 15 years to 15.5 years for Unit 1 of the Donald C. Cook Nuclear Plant (CNP).

If the Reviewer is the Preparer's supervisor, the supervisor review is needed and is approved: ~NIA Supervisor's Manager's Name Title Signature Date Qualification Matrix Verification

  • The responsible Engineering Supervisor/Manager approval signature also serves to signify that the qualifications of the individual(s) assigned as Preparer(s) and Reviewer(s) and Independent Design Verifier(s) were verified in the Plant Qualification Matrix.

Preparation & Review PREPARED BY: REVIEWED BY: *APPROVED BY:

Name: James Heyeck Chris Peckat Greg Hill

Title:

PRA Contractor Nuclear Engineer Sr Supervisor Organization: Jensen Hughes NSA/PRA NSA/PRA Signature:

Date:

D Sign-offs for additional Preparer(s) and Reviewer(s) on next page

! This document includes the following pages: 1-32 (32 pages total)  ! Page 1

0 JENSEN HUGHES Advancing the Science of Safety Donald C. Cook Nuclear Plant:

Evaluation of Risk of Short-Term ILRT Extension PRA-QNT-009 Prepared for:

,,,,.. INDIANA

,ii,jf MICHIGAN POWER" A unit ofAmerican Electric Power Donald C. Cook Nuclear Plant Indiana Michigan Power Company Project Number: 1EAT1V046 Project

Title:

Short-Term ILRT Extension Revision: O Name and Date Digitally signed by Justin Sattler Preparer: Justin Sattler Date: 2020.05.22 10:22:04-05'00'

=- 7

~ig itally signed by Matt Johnson Reviewer: Matthew Johnson 1 IOJl'<~,,_____ _ _ Qate: 2020.05.22 12:29:36-05'00'

___ .J Review Method Design Review r;gJ Alternate Calculation Er Th Orn S bLI 0 -- igitally signed by Eric Thornsbury Approver: Eric Thornsbury *1 C ry~o/'l:,c;=us, E=~thornsbu,y@jens_enhughes.com, O=Jensan

-Hughes, OU=R1sk Informed Services, CN=Enc Thornsbury 1 -=.O*tej2020.os.22 13:4B:s2-04*00*

Revision 0 Page 2 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension REVISION RECORD

SUMMARY

Revision Revision Summary 0 Initial Issue Revision 0 Page 3 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension TABLE OF CONTENTS 1.0 PURPOSE ..............*........................................................................................................ 5 2.0 SCOPE ........................................................................................................................... 5

3.0 REFERENCES

............................................................................................................... *7 4.0 ASSUMPTIONS AND LIMITATIONS ............................................................................. 10 5.0 METHODOLOGY AND INPUTS .................................................................................... 11 5.1 General Resources Available ...................................................................................... 11 5.2 Plant Specific Inputs ................................................................................................... 12 5.3 Impact of Extension on Detection of Component Failures that Lead to Leakage (Small and Large) ............................................................................................................... 14 6.0 CALCULATIONS ...........................................................................................................15 6.1 Step 1 - Quantify the Risk in Terms of Frequency per Reactor Year .......................... 16 6.2 Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose) ....................... 19 6.3 Step 3- Evaluate Risk Impact of Extending Type A Test Interval from 15 to 15.5 Years

                                                                                                                                                                                                                                                                • 20 6.4 Step 4 - Determine the Change in Risk in Terms of LERF .......................................... 22 6.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability .... 23 6.6 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage ...... 24
6. 7 Impact from External Events Contribution ................................................................... 27 7.0 RESULTS ......................................................................................................................29

8.0 CONCLUSION

S AND RECOMMENDATIONS .............................................................. 30 A. PRA Acceptability ..........................................................................................................32 Revision 0 Page 4 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension 1.0 PURPOSE The purpose of this analysis is to provide a risk assessment of temporarily extending the currently allowed containment Type A Integrated Leak Rate Test (ILRT) interval by 6 months from 15 years to 15.5 years for Unit 1 of the Donald C. Cook Nuclear Plant (CNP). The risk assessment follows the guidelines from Nuclear Energy Institute (NEI) 94-01, Revision 3-A

[Reference 1], 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 Nuclear Regulatory Commission (NRC) regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in Regulatory Guide (RG) 1.200 as applied to ILRT interval extensions, risk insights in support of a request for a plant's licensing basis as outlined in RG 1.174 [Reference 4], the methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval [Reference 5], and the methodology used in Electric Power Research Institute (EPRI) 1018243, which is Revision 2-A of EPRI 1009325

[Reference 24].

2.0 SCOPE 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 assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the NRC's rulemaking basis, NEI undertook a similar study. The results of that study are documented in EPRI Research Project TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals" [Reference 2].

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), containment isolation failures contribute less than 0.1 % to the latent risks from reactor accidents. Consequently, it is desirable to show that extending the ILRT interval will not lead to a substantial increase in risk from containment isolation failures for CNP.

NEI 94-01 Revision 3-A supports using EPRI Report No. 1018243, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," for performing risk impact assessments in support of ILRT extensions [Reference 24]. The Guidance provided in Appendix Hof EPRI Report No. 1018243 builds on the EPRI Risk Assessment methodology, EPRI TR-104285. This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed ILRT changes.

It should be noted that containment leak-tight integrity is also verified through periodic in-service inspections conducted in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI. More specifically, Subsection IWE provides the rules and requirements for in-service inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants. Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct visual inspections of the accessible areas of the interior of the containment. The associated change to NEI 94-01 will require that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being conducted. These requirements will not be changed as a result of the extended ILRT interval. In Revision 0 Page 5 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension 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 temporary 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 1o-s per reactor year and increases in Large Early Release Frequency (LERF) less than 10-7 per reactor year. Since containment accident pressure is not credited in support of ECCS performance, the Type A test does not impact CDF; therefore, the relevant criterion is the change in LERF. RG 1.174 also defines "small" changes in LERF as below 1o-s 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 addition, 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 Evaluations (SEs) for one-time interval extension (summarized in Appendix G of Reference 24) 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 small population dose is defined as an increase from the baseline interval dose of ::.1.0 person-rem per year or 1% of the total dose, whichever is less restrictive for the risk impact assessment of the proposed extended ILRT interval [Reference 1].

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PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension

3.0 REFERENCES

The following references were used in this calculation:

1. Revision 3-A to Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, 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 Containment Integrated 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, Revision 3, January 2018.
5. Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, Letter from Mr. C. H.

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

50-317, March 27, 2002.

6. Performance-Based Containment Leak-Test Program, NUREG-1493, September 1995.
7. Evaluation of Severe Accident Risks: Surry Unit 1, Main Report NUREG/CR-4551, SAND86-1309, Volume 3, Revision 1, Part 1, 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, ORNL/TM-8964, April 1984.
11. Reliability Analysis of Containment Isolation Systems, Pacific Northwest Laboratory, NUREG/CR-4220, PNL-5432, June 1985.
12. Technical Findings and Regulatory Analysis for Generic Safety Issue I1.E.4.3

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

13. Review of Light Water Reactor Regulatory Requirements, Pacific Northwest Laboratory, NUREG/CR-4330, PNL-5809, Volume2, June 1986.
14. Shutdown Risk Impact Assessment for Extended Containment Leakage Testing Intervals Utilizing ORAM', 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 PRA-NB-QU, "Internal Events Quantification Notebook," Revision 5, April 2018.

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PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension

18. Calculation PRA-NB-LER, "Large Early Release Frequency Notebook," Revision 2, October 2018.
19. Calculation PRA-ILRT, "Risk Impact Assessment For Permanently Extending Containment Type A Test Interval," Revision 2, March 2014.
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 Commi~sion, Docket No. 50-278, License No. DPR-56, LAR-01-00430, dated May 30, 2001.
22. Risk Assessment for Joseph M. Farley Nuclear Plant Regarding ILRT (Type A)

Extension Request, prepared for Southern Nuclear Operating Co. by 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 Integrated Leak Rate Testing lnteNals, 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. Report PRA-NB-SPRA-QU, "SPRA Model Quantification Notebook," Revision 2, October 2019.
28. Calculation PRA-NB-FIRE-FQ, "Fire PRA Model Quantification Notebook," Revision 3, October 2019.
29. NTS 2003 009 REP, Revision O," Cook Nuclear Plant Severe Accident Mitigation Alternatives Analysis," October 20, 2003.
30. 1-EHP-4030-134-202, "Integrated Leak Rate Test," Revision 3.
31. NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition.
32. "Individual Plant Examination of External Events Summary Report," April 1992.
33. NUREG/CR-6850 Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements," September 2010.
34. Technical Letter Report ML112070867, Containment Liner Corrosion Operating Experience Summary, Revision 1, August 2011.
35. NUREG-0800, Section 6.5.2, "Containment Spray as a Fission Product Cleanup System," Revision 3.
36. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009.
37. ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," 2009.

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PRA-QNT -009 Evaluation of Risk Significance of Short-Term ILRT Extension

38. NEI 05-04, "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard," Revision 2, November 2008.
39. ACUBE 2.0 Software Manual, EPRI Report 3002003169, December 2014.
40. NEI Letter to USNRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," February 21, 2017, Accession Number ML17086A431.
41. USN RC Letter to Mr. Greg Krueger (NEI), "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Observations (F&Os)," May 3, 2017, Accession Number ML17079A427.
42. NUREG-2169, "Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database: United States Fire Event Experience Through 2009," January 2015.
43. LTR-RAM-11-10-041, "Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the DC Cook Fire Probabilistic Risk Assessment," July 20, 2010.
44. Erin Engineering Report #D0403140002, "D.C. Cook Focused Scope Peer Review for Fire PRA," November 2015.
45. PWROG-17027-P, "Focused Scope Peer Review of the DC Cook Internal Fire Probabilistic Risk Assessment," July 2017.
46. PWROG-18062-P, "Peer Review of the D.C. Cook 1&2 Seismic Probabilistic Risk Assessment," Revision 0, January 2019.
47. Report AEPDCC-00058-REPT-001, "SPRA F&O Independent Assessment and Focused-Scope Peer Review, Donald C. Cook Nuclear Plant Units 1 and 2," Revision A.
48. AEP-NRC-2019-56, "Seismic Probabilistic Risk Assessment in Response to Near Term Task Force Recommendation 2.1: Seismic," November 4, 2019.

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PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension 4.0 ASSUMPTIONS AND LIMITATIONS The following assumptions were used in the calculation:

The acceptability (i.e., technical adequacy) of the CNP PRA [Reference 17] is either consistent with the requirements of Regulatory Guide 1.200, or where gaps exist, the gaps have been addressed, as detailed in Appendix A.

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

It is appropriate to use the CNP internal events PRA model to effectively describe the risk change attributable to the ILRT extension. An analysis is performed in Section 6.7 to show the effect of including external event models for the ILRT extension. The Seismic PRA [Reference 28] and Fire PRA [Reference 18] are used for this sensitivity analysis.

Accident classes describing radionuclide release end states are defined consistent with EPRI methodology [Reference 24].

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

The representative containment leakage for Class 3a sequences is 1Ola 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 1OOLa based on the guidance provided in EPRI Report No. 1018243 [Reference 24].

The sequences of 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 in 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 [Reference 24].

While precise numbers are maintained throughout the calculations, some values have been rounded when presented in this report. Therefore, rounding differences may result in differences in table summations.

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PRA-QNT -009 Evaluation of Risk Significance of Short-Term ILRT Extension 5.0 METHODOLOGY AND INPUTS This section summarizes the general resources available as input (Section 5.1) and the plant specific resources required (Section 5.2).

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

1. NUREG/CR-3539 [Reference 10]
2. NUREG/CR-4220 [Reference 11]
3. NUREG-1273 [Reference 12]
4. NUREG/CR-4330 [Reference 13]
5. EPRI TR-105189 [Reference 14]
6. NUREG-1493 [Reference 6]
7. EPRI TR-104285 [Reference 2]
8. NUREG-1150 [Reference 15] and NUREG/CR-4551 [Reference 7]
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 I LRT 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 local leak rate test (LLRT) 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 CNP. 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.

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PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension 5.2 Plant Specific Inputs The plant-specific information used to perform the CNP ILRT Extension Risk Assessment includes the following:

CDF and LERF Model results [Reference 17, Reference 18]

Dose within a 50-mile radius [Reference 19, Reference 29]

CNP Model The Internal Events PRA Model that is used for CNP is characteristic of the as-built plant. The current Level 1, LERF, and Level 2 model is a linked fault tree model [Reference 17]. The CDF is 2.40E-5/year for Unit 1; the LERF is 2.25E-6/year for Unit 1 [Reference 17]. Table 5-1 and Table 5-2 provide a summary of the Internal Events CDF and LERF results for the CNP PRA Model.

The total Fire CDF is 6.47E-5/year for Unit 1; the total Fire LERF is 4.0SE-6/year for Unit 1

[Reference 18]. The total Seismic CDF is 2.44E-5/year for Unit 1 and the total Seismic LERF is 5.65E-6/year [Reference 28]. Refer to Section 6. 7 for further details on external events as they pertain to this analysis.

Table 5 Internal Events CDF Internal Events Unit 1 Frequency (per year)

Internal Floods 4.0?E-06 Transients 1.46E-05 Main Steam Break Inside Containment 2.15E-07 LOCAs 3.31E-06 ISLOCA 5.13E-07 SGTR 6.30E-07 RPV Rupture 2.90E-08 Loss of Offsite Power (LOOP) 6.39E-07 Total Internal Events CDF 2.40E-05 Table 5 Internal Events LERF Internal Events Unit 1 Frequency (per year)

Internal Floods 3.44E-07 Transients 4.06E-07 Main Steam Break Inside Containment 3.91E-09 LOCAs 6.93E-07 ISLOCA 2.30E-07 SGTR 4.57E-07 RPV Rupture 2.84E-08 LOOP 8.50E-08 Total Internal Events LERF 2.25E-06 Revision 0 Page 12 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension Population Dose Calculations The population dose calculation was reported in the permanent 15-year ILRT extension

[Reference 19], which used data from the SAMA [Reference 29]. Table 5-3 presents these dose exposures.

Table 5 Population Dose EPRI Category Dose (person-rem)

Class 1 1.01E+03 Class 2 3.84E+06 Class 7 3.84E+06 Class 8 9.68E+06 Release Category Definitions Table 5-4 defines the accident classes used in the ILRT extension evaluation, which is consistent with the EPRI methodology [Reference 24]. These containment failure classifications are used in this analysis to determine the risk impact of extending the Containment Type A test interval, as described in Section 6.0 of this report.

Table 5 EPRI Containment Failure Classification [Reference 24]

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

Containment isolation failures (as reported in the Individual Plant Examinations) including those accidents 2

in which there is a failure to isolate the containment.

Independent (or random) isolation failures include those accidents in which the pre-existing isolation 3

failure to seal (i.e., provide a leak-tight containment) 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 4

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.

Containment isolation failures including those leak paths covered in the plant test and maintenance 6

requirements or verified per in-service inspection and testing (IS1/IST) program.

Accidents involving containment failure induced by severe accident phenomena. Changes in Appendix J 7

testing requirements do not impact these accidents.

Accidents in which the containment is bypassed (either as an initial condition or induced by phenomena) 8 are included in Class 8. Changes in Appendix J testing requirements do not impact these accidents.

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PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension 5.3 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 accident class, as defined in Table 5-4, 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 2 / 217 = 0.0092). For Class 3b, the probability is based on the Jeffreys non-informative prior for no "large" failures in 217 tests (i.e., 0.5 / (217+1) = 0.0023).

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

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

The supplemental information states:

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

The application of this additional guidance to the analysis for CNP, as detailed in Section 6.0, involves subtracting LERF risk (except the risk associated with a pre-existing leak) from the GDF that is applied to Class 3b because this portion of LERF is unaffected by containment integrity. To be consistent, the same change is made to the Class 3a GDF, even though these events are not considered LERF.

Consistent with the NEI Guidance [Reference 3], the change in the leak detection probability can be estimated by comparing the average time that a leak could exist without detection. For example, the average time that a leak could go undetected with a three-year test interval is 1.5 years (3 years/ 2), the average time that a leak could exist without detection for a 15-year test interval is 7.5 years (15 years/ 2), and the average time that a leak could exist without detection for a 15.5-year interval is 7.75 years (15.5 years/ 2). The change from a three-year test interval to a 15-year test interval would lead to a non-detection probability that is a factor of 5.0 (7.5/1.5) higher for the probability of a leak that is detectable only by ILRT testing. Correspondingly, an extension of the ILRT interval to 15. 5 years can be estimated to lead to a factor of 5.17

((15.5/2)/1.5) increase in the non-detection probability of a leak.

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 NRG [Reference 91) because it does not factor in the possibility that the failures could be Revision 0 Page 14 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension 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.

6.0 CALCULATIONS The application of the approach based on the guidance contained in EPRI 1018243 [Reference 24] and previous risk assessment submittals on this subject [References 5, 8, 21, 22, and 23]

have led to the following results. The results are displayed according to the eight accident classes defined in the EPRI report, as described in Table 6-1.

The analysis performed examined CNP-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:

Core damage sequences in which the containment remains intact initially and in the long term (EPRI 1018243, Class 1 sequences [Reference 24]).

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 1018243, Class 3 sequences [Reference 24]).

Accident sequences involving containment bypassed (EPRI 1018243, Class 8 sequences [Reference 24]), large containment isolation failures (EPRI 1018243, Class 2 sequences [Reference 24]), and small containment isolation "failure-to-seal" events (EPRI 1018243, Class 4 and 5 sequences [Reference 24]) 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 6 EPRI Accident Class Definitions Accident Classes (Containment Release Type) Description 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 (SGTR and 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 6-1.

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

Step 3 - Evaluate risk impact of a one-time extension of the Type A test interval from 1 in 15 Revision 0 Page 15 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension years to 1 in 15.5 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).

6.1 Step 1 - Quantify the Risk in Terms of Frequency per Reactor Year As previously described, the extension of the Type A interval does not influence those accident progressions that involve large containment isolation failures, Type B or Type C testing, or containment failure induced by severe accident phenomena.

For the assessment of ILRT impacts on the risk profile, the potential for pre-existing leaks is included in the model (these events are represented by the Class 3 sequences in EPRI 1018243 [Reference 24]). 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 6-1 were developed for CNP by first determining the frequencies for Classes 1, 2, 6, 7, and 8.

Table 6-2 presents the grouping of each release category in EPRI Classes based on the associated description. Table 6-3 provides a summary of the accident sequence frequencies that can lead to radionuclide release to the public and have been derived consistent with the NEI Interim Guidance [Reference 3] and the definitions of accident classes and guidance provided in EPRI Report No. 1018243 [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 6.6.

Class 3 Sequences. This group consists of all core damage accident progression bins for which a pre-existing leakage in the containment structure (e.g., containment liner) exists that can only be detected by performing a Type A ILRT. The probability of leakage detectable by a Type A ILRT is calculated to determine the impact of temporarily extending the testing interval. The Class 3 calculation is divided into two classes: Class 3a is defined as a small liner breach (La <

leakage< 10La), and Class 3b is defined as a large liner breach (10La <leakage< 100La).

Data reported in EPRI 1018243 [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 Pctass3a = 217 = 0.0092 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 5.3, additional consideration is made to not apply failure probabilities on those cases that are already LERF scenarios. Therefore, these LERF contributions from CDF are removed (except the LERF associated with a pre-existing leak). The frequency of a Class 3a failure for a 3 in 10 year ILRT interval is calculated by the following equation (ratios of intervals are subsequently applied to calculate frequencies for longer ILRT intervals):

Frequ1class3a = Pctass3a * ( CDF - (LERF - LERF1eak))

2

= 217 * (2.40E (2.25E-6-1.09E-7)) = 2.0lE-7 Revision 0 Page 16 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension The risk contribution is changed based by a factor of 15/3 and 15.5/3 (for the 1 in 15 year and 1 in 15.5 year cases, respectively) compared to the 3 in 10 year ILRT interval case values. The Class 3a frequencies are calculated as follows:

15 2 Frequ1class3a15yr = 3

  • 217 * ( CDF - )

(LERF - LERFieak) = 3 15

  • 2172
  • 2.18E-5 = 1.0lE-6 15.5 2 FreqU1Class3a15 *5yr = - 3 * -217 * ( CDF - )

(LERF - LERFieak) = - 3 15.5 2

  • -217
  • 2.18E-5 =
  • 1.04E-6 In the database of 217 ILRTs, there are zero containment leakage events that could result in a large early release. Therefore, the Jeffreys non-informative prior is used to estimate a failure rate and is illustrated in the following equations:

Number of Failures+ 1/2 Jeffreys Failure Probability= N b f T um er o ests + 1 0 + 1/2 Pc1ass3b = 217 + l = 0.0023 The frequency of a Class 3b failure for a 3 in 10 year ILRT interval is calculated by the following equation (ratios of intervals are subsequently applied to calculate frequencies for longer ILRT intervals):

FreqU1class3b = Pclass3b * (CDF- (LERF-LERFieak))

= ~ * (2.40E (2.25E-6-1.09E-7)) = 5.0lE-8 218 The risk contribution is changed based by a factor of 15/3 and 15.5/3 (for the 1 in 15 year and 1 in 15.5 year cases, respectively) compared to the 3 in 1O year ILRT interval case values. The Class 3b frequencies are calculated as follows:

  • ~ *~

15 15 Frequ1c1ass3b15yr = 3 218

  • 2.18E-5 = 2.SlE-7
  • ~ *~

15 5 15 5 FreqU1Class3b15 *syr = 3 218

  • 2.18E-5 = 2.59E-7 For this analysis, the associated containment leakage for Class 3a is 1Ola and for Class 3b is 1OOLa. 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 Intact frequency for internal events is 5.04E-6 for Unit 1 [Reference 17]. The EPRI Accident Class 1 frequency is then adjusted by subtracting the EPRI Class 3a and 3b frequency (to preserve total CDF), calculated below:

Frequ1class1 = Frequ11ntact - (Frequ1class3a - Frequ1class3b)

Class 2 Sequences. This group consists of accident progression bins with large containment isolation failures. The large isolation failure is in internal events cutsets that contribute 12.9% of LERF for Unit 1. Multiplying by LERF, the EPRI Accident Class 2 frequency is 2.89E-7 for Unit 1, as shown in Table 6-2.

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 Revision 0 Page 17 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension a containment isolation failure-to-seal of Type C test components occurs. Because the failures are detected by Type C tests which are unaffected by the Type A ILRT, this group is not evaluated any further in this analysis, 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. All other failure modes are bounded by the Class 2 assumptions.

This accident class is also not evaluated further.

Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure is induced by severe accident phenomena (e.g., overpressure). This frequency is calculated by subtracting the Class 1, 2, and 8 frequencies from the total CDF. For this analysis, the frequency is determined from the EPRI Accident Class 7 frequency listed in Table 6-2.

Class 8 Sequences. This group consists of all core damage accident progression bins in which containment is bypassed via interfacing-systems loss of coolant accident (ISLOCA) or steam generator tube rupture (SGTR). The ISLOCA initiators are in internal events cutsets that contribute 2.14% of LERF for Unit 1. The SGTR initiator is in internal events cutsets that contribute 2.63% of CDF for Unit 1. Thus, the total EPRI Accident Class 8 frequency is the summation of the ISLOCA and SGTR frequencies, 8.60E-7 for Unit 1, as shown in Table 6-2 and Table 6-3.

Table 6 Accident Class Frequencies EPRI Category Unit 1 Frequency (/yr)

Class 1 2.17E-05 Class 2 2.89E-07 Class 7 1.10E-06 Class 8 (SGTR) 6.30E-07 Class 8 (ISLOCA) 2.30E-07 Total (CDF) 2.40E-05 Table 6 Risk Profile for 3 in 10 Year ILRT Class Description Unit 1 Frequency (/yr)

No containment failure 2.1 SE-05 2 2 Large containment isolation failures 2.89E-07 3a Small isolation failures (liner breach) 2.01 E-07 3b Large isolation failures (liner breach) 5.01 E-08 4 Small isolation failures - failure to seal (type B) E1 5 Small isolation failures - failure to seal (type C) E1 6 Containment isolation failures (dependent failure, personnel errors) E1 7 Severe accident phenomena induced failure (early and late) 1.1 0E-06 8 Containment bypass 8.60E-07 Total 2.40E-05

1. E represents a probabilistically insignificant value or a Class that is unaffected by the Type A ILRT.
2. The Class 3a and 3b frequencies are subtracted from Class 1 to preserve total CDF.

Revision 0 Page 18 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension 6.2 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. Table 5-3 provides population dose for each release category. Table 6-4 provides a correlation of CNP population dose to EPRI Accident Class. The population dose for EPRI Accident Classes 3a and 3b were calculated based on the guidance provided in EPRI Report No. 1018243 [Reference 24] as follows:

EPRI Class 3a Population Dose= 10

  • 1.01£+3 = 1.01£+4 EPRI Class 3b Population Dose= 100
  • 1.01£+3 = 1.01£+5 Table 6 Population Doses Unit 1 Population Class Description Dose (person-rem) 1 No containment failure 1.01E+03 2 Large containment isolation failures 3.84E+06 3a Small isolation failures (liner breach) 1.01E+041 3b Large isolation failures (liner breach) 1.01E+052 4 Small isolation failures - failure to seal (type B) N/AI 5 Small isolation failures - failure to seal (type C) N/A 6 Containment isolation failures (dependent failure, personnel errors) N/A 7 Severe accident phenomena induced failure (early and late) 3.84E+06 8 Containment bypass 9.68E+06
1. 10*La
2. 100*La Revision 0 Page 19 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension Table 6 Unit 1 Risk Profile for 3 in 10 Year ILRT Class Description Frequency Contribution Population Population

(/yr) (%) Dose (person- Dose Rate rem) (person-rem/yr) 1 No containment failure 2 2.15E-05 89.58% 1.01E+03 2.17E-02 2 Large containment isolation failures 2.89E-07 1.21% 3.84E+06 1.11E+00 Small isolation failures (liner 3a 2.01E-07 0.84% 1.01E+04 2.03E-03 breach)

Large isolation failures (liner 3b 5.01E-08 0.21% 1.01E+05 5.06E-03 breach)

Small isolation failures - failure to E1 E1 E1 E1 4

seal (type B)

Small isolation failures - failure to E1 E1 E1 E1 5

seal (type C)

Containment isolation failures 6 (dependent failure, personnel E1 E1 E1 E1 errors)

Severe accident phenomena 7 1.1 0E-06 4.58% 3.84E+06 4.22E+00 induced failure (early and late) 8 Containment bypass 8.60E-07 3.58% 9.68E+06 8.32E+00 Total 2.40E-05 1.37E+01

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

6.3 Step 3 - Evaluate Risk Impact of Temporarily Extending Type A Test Interval from 15 to 15.5 Years The next step is to evaluate the risk impact of a one-time extension of the test interval from its current 15-year interval to a 15.5-year interval. To do this, an evaluation must first be made of the risk associated with the 15-year interval to allow calculation of a change in risk for the one-time extension to a 15.5-year interval.

Risk Impact Due to 15-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 changed based on the NEI guidance as described in Section 5.3 by a factor of 15/3 compared to the 3 in 1O year ILRT interval case values. The Class 3a and 3b frequencies are calculated as follows:

Frequ 1class 3 al5yr = 315

  • 217 2

)

= 315

  • 217 2
  • 2.18E-S = l.0lE-6
  • 2218 * (CDF - = 153
  • 2218
  • 2.18E-5 = 2.SlE-7 15 Frequ1c1ass3b15yr = 3 (LERF - LERF1eak))

The results of the calculation for a 15-year interval are presented in Table 6-6.

Revision 0 Page 20 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension Table 6 Unit 1 Risk Profile for Once in 15 Year ILRT Class Description Frequency Contribution Population Population

(/yr) (%) Dose (person- Dose Rate rem) (person-rem/yr) 1 No containment failure 2 2.0SE-05 85.38% 1.01E+03 2.07E-02 2 Large containment isolation failures 2.89E-07 1.21% 3.84E+06 1.11 E+00 Small isolation failures (liner 3a 1.01 E-06 4.20% 1.01E+04 1.02E-02 breach)

Large isolation failures (liner 3b 2.51E-07 1.04% 1.01E+05 2.53E-02 breach)

Small isolation failures - failure to £1 £1 £1 £1 4

seal (type B)

Small isolation failures - failure to £1 £1 £1 £1 5

seal (type C)

Containment isolation failures 6 (dependent failure, personnel £1 £1 £1 £1 errors)

Severe accident phenomena 7 1.10E-06 4.58% 3.84E+06 4.22E+00 induced failure (early and late) 8 Containment bypass 8.60E-07 3.58% 9.68E+06 8.32E+00 Total 2.40E-05 1.37E+01

1. £ represents a probabilistically insignificant value or a Class that is unaffected by the Type A ILRT.
2. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total GDF.

Risk lm~act Due to 15.5-Year Test Interval The risk contribution for a 15.5-year interval is calculated in a manner similar to the 15-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 15.5/3 compared to the 15-year interval value, as described in Section 5.3. The Class 3a and 3b frequencies are calculated as follows:

15.5 2 Frequ1c1ass3a15

  • 5yr = - 3 * -217 * ( CDF - (LERF - LERF1eak) ) 15.5

= -3 2

  • -217
  • 2.18E-5 = 1.04E-6
  • ~ *~

15 5 15 5 Frequ1c1ass3b15 *5yr = 3 218

  • 2.18E-5 = 2.59E-7 The results of the calculation for a 15.5-year interval are presented in Table 6-7.

Revision 0 Page 21 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension Table 6 Unit 1 Risk Profile for Once in 15.5 Year ILRT Class Description Frequency Contribution Population Dose Population

(/yr) (%) (person-rem) Dose Rate (person-rem/yr)

No containment failure 2 2.04E-05 85.21% 1.01E+03 2.06E-02 Large containment isolation 2 2.89E-07 1.21% 3.84E+06 1.11 E+00 failures Small isolation failures (liner 3a 1.04E-06 4.34% 1.01 E+04 1.0SE-02 breach)

Large isolation failures (liner 3b 2.59E-07 1.08% 1.01 E+05 2.61E-02 breach)

Small isolation failures - failure 4 E1 E1 E1 E1 to seal (type B)

Small isolation failures - failure 5 E1 E1 E1 E1 to seal (type C)

Containment isolation failures 6 (dependent failure, personnel E1 E1 E1 E1 errors)

Severe accident phenomena 7 1.10E-06 4.58% 3.84E+06 4.22E+00 induced failure (early and late) 8 Containment bypass 8.60E-07 3.58% 9.68E+06 8.32E+00 Total 2A0E-05 1.37E+01

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

6.4 Step 4 - Determine the Change in Risk in Terms of LERF The risk increase associated with one-time extension of the I LRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from an intact containment could, in fact, result in a larger release due to the increase in probability of failure to detect a pre-existing leak. With strict adherence to the EPRI guidance, 100% of the Class 3b contribution would be considered LERF.

Regulatory Guide 1.174 [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 10-5/year and increases in LERF less than 1o-7/year, and "small" changes in LERF as less than 1o-6/year. Since containment overpressure is not required in support of ECCS performance to mitigate design basis accidents and no equipment in the shield building is credited in the CDF model at CNP, the ILRT extension does not impact CDF. Therefore, the relevant risk-impact metric is LERF.

For CNP, 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 15-year test interval from Table 6-6, the Class 3b frequency is 2.51 E-7/year for Unit 1; based on a 15.5-year test interval from Table 6-7, the Class 3b frequency is 2.59E-7/year for Unit 1. Thus, the increase in the overall probability of LERF due to Class 3b sequences that is due to the one-time ILRT test interval extension from 15 to 15.5 years is 8.35E-9/year for Unit 1. As can be seen, even with the conservatisms included in the evaluation (per the EPRI methodology), the estimated one-Revision 0 Page 22 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension time change in LERF meets the criteria for a "very small" change when comparing the 15.5-year results to the current 15-year requirement. Table 6-8 summarizes these results.

Table 6 Impact on LERF due to Extended Type A Testing Intervals ILRT Inspection Interval Unit 1: 15 Years Unit 1: 15.5 Years Class 3b (Type A LERF) 2.51E-07 2.59E-07 llLERF (15 year baseline) 8.35E-09 NEI 94-01 [Reference 1] states that a "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. As shown in Table 6-9, the results of this calculation meet the dose rate criteria.

Table 6 Impact on Dose Rate due to Extended Type A Testing Intervals ILRT Inspection Interval Unit 1: 15.5 Years llDose Rate (15 year baseline) 1.14E-03

%l1Dose Rate (15 year baseline) 0.0083%

1. llDose Rate is the difference in the total dose rate between cases. For instance, 'llDose Rate (15 year baseline)' for the 1 in 15.5 case is the total dose rate of the 1 in 15.5 case minus the total dose rate of the 1 in 15 year case.
2. %l1Dose Rate is the llDose Rate divided by the total baseline dose rate. For instance, 'o/ollDose Rate (15 year baseline)' for the 1 in 15.5 case is the 'llDose Rate (15 year baseline)' of the 1 in 15.5 year case divided by the total dose rate of the 1 in 15 year case.

6.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability Another parameter that the NRC guidance in RG 1.174 [Reference 4] states can provide input into the decision-making process is the change in 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 =l _ 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. Table 6-10 shows the steps and results of this calculation.

Table 6 Impact on CCFP due to Extended Type A Testing Intervals ILRT Inspection Interval Unit 1: 15 Years Unit 1: 15.5 Years f(ncf) (/yr) 2.149E-05 2.148E-05 f(ncf)/CDF 0.896 0.895 CCFP 0.104 0.105 llCCFP (15 year baseline) 0.035%

As stated in Section 2.0, a change in the CCFP of up to 1.5% is assumed to be "small." The one-time increase in the CCFP from the 1 in 15 year interval to 1 in 15.5 year interval is 0.035%

for Unit 1. Therefore, this increase is judged to be "small."

Revision 0 Page 23 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension 6.6 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 a methodology similar to the Calvert Cliffs liner corrosion analysis [Reference 5]. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete basemat, 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 basemat and the containment cylinder and dome The historical steel liner flaw likelihood due to concealed corrosion The impact of aging The corrosion leakage dependency on containment pressure The likelihood that visual inspections will be effective at detecting a flaw Assumptions Consistent with the Calvert Cliffs analysis, a half failure is assumed for basemat concealed liner corrosion due to the lack of identified failures (See Table 6-11, Step 1).

In the 5.5 years following September 1996 when 10 CFR 50.55a started requiring visual inspection, there were three events where a through wall hole in the containment liner was identified. These are Brunswick 2 on 4/27/99, North Anna 2 on 9/23/99, and D. C.

Cook 2 in November 1999. The corrosion associated with the Brunswick event is believed to have started from the coated side of the containment liner. Although CNP has a different containment type than Brunswick 2 and North Anna 2, these events could also potentially occur at CNP (i.e., corrosion starting on the coated side of containment).

Construction material embedded in the concrete may have contributed to the corrosion.

The corrosion at North Anna is believed to have started on the uninspectable side of containment due to wood embedded in the concrete during construction. The D.C. Cook event is associated with an inadequate repair of a hole drilled through the liner during construction. Since the hole was created during construction and not caused by corrosion, this event does not apply to this analysis. Based on the above data, there are two corrosion events from the 5.5 years that apply to CNP.

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)

(See Table 6-11, 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 6-11, Steps 2 and 3). Sensitivity studies are included that address doubling this rate every ten years and every two years.

In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching the outside atmosphere, given that a liner flaw exists, was estimated as 1.1 % for the cylinder and dome, and 0.11 % (10% of the cylinder failure probability) for the basemat. These Revision 0 Page 24 of 32

PRA-QNT -009 Evaluation of Risk Significance of Short-Term ILRT Extension values were determined from an assessment of the probability versus containment pressure. For CNP, the ILRT maximum pressure is 12.0 psig [Reference 30].

Probabilities of 1% for the cylinder and dome, and 0 .1 % for the bas em at are used in this analysis.

Consistent with the Calvert Cliffs analysis, the likelihood of leakage escape (due to crack formation) in the basemat region is considered to be less likely than the containment cylinder and dome region (See Table 6-11, Step 4).

In the Calvert Cliffs analysis, it is noted that approximately 85% of the interior wall surface is accessible for visual inspections. 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 6-11, 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.

Revision 0 Page 25 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension Table 6 Steel Liner Corrosion Base Case Step Description Containment Cylinder and Containment Basemat (15%)

Dome (85%)

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 10 CFR 50.55a requirements of periodic visual inspections of containment surfaces Aged adjusted liner flaw likelihood Year Failure rate Year Failure rate During the 15.5-year interval, 1 2.0SE-03 1 5.13E-04 assume failure rate doubles every 15.5 1.54E-02 15.5 3.84E-03 2 five years (14.9% increase per year). The average for the 5th to 10th year set to the historical failure 15.5 year average= 7.00E-03 15.5 year average= 1.75E-03 rate.

Increase in flaw likelihood between 3 and 15 years Uses aged adjusted 0.71% (1 to 3 years) 0.18% (1 to 3 years) 3 liner flaw likelihood (Step 2), 9.66% (1 to 15 years) 2.42% (1 to 15 years) assuming failure rate doubles every 11.19% ( 1 to 15.5 years) 2.80% (1 to 15.5 years) five years.

Likelihood of breach in containment 4 1% 0.1%

given liner flaw 10%

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

Visual inspection detection failure 5 but could be detected by ILRT).

likelihood Cannot be visually inspected All events have been detected through visual inspection. 5%

visible failure detection is a conservative assumption.

0.00071 % (3 years) 0.00018% (3 years) 0.71%x1%x10% 0.18% X 0.1% X 100%

Likelihood of non-detected 0.00966% (15 years) 0.00242% (15 years) 6 containment leakage (Steps 3 x 4 x

5) 9.66% X 1% X 10% 2.42% X 0.1 % X 100%

0.01119% (15.5 years) 0.00280% (15.5 years) 11.19% X 1% X 10% 2.80% X 0.1% X 100%

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

Table 6 Total Likelihood on Non-Detected Containment Leakage Due to Corrosion for CNP Description At 3 years: 0.00071% + 0.00018% = 0.00089%

At 15 years: 0.00966% + 0.00242% = 0.01207%

At 15.5 years: 0.01119% + 0.00280% = 0.01399%

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PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension 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 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 NRC 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 [Reference 34]. Two other containment liner through-wall hole events occurred at Turkey Point Units 3 and 4 in October 201 O 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 [Reference 34].

6.7 Impact from External Events Contribution An assessment of the impact of external events is performed. The purpose for this investigation is to determine the change in LERF associated with a one-time 6 month extension in ILRT testing.

The CNP Fire PRA model was created to satisfy the ASME/ANS PRA standard [Reference 37]

and support risk-informed NFPA 805 applications [Reference 31]. The Fire PRA model was used to obtain the fire CDF and LERF values [Reference 28]. To reduce conservatism in the ILRT extension analysis, the methodology of subtracting existing LERF from CDF (except the risk associated with a pre-existing leak) is also applied to the Fire PRA model. The following shows the calculation for Class 3b:

15 FreqUlclass3b15yr = 3

  • Pclass3b * (CDF - (LERF - LERF1eak))

= 315

  • 218 0.5
  • (6.47£ (4.05£ 2.83£-7)) = 6.99E-7 Frequlclass3b15.5yr = -15.5 3

(

-

  • Pclass3b

)

=

15 5 3

  • *~218
  • (6.47£ (4.0SE 2.83£-7)) = 7.ZZE-7 Revision 0 Page 27 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension Reference 27 was developed for risk-informed applications and provides an assessment of the seismic hazard. The Unit 1 Seismic CDF and LERF are estimated to be 2.44E-5/yr and 5.65E-6/yr, respectively.

Subtracting seismic LERF from CDF (except the risk associated with a pre-existing leak), the Class 3b frequency can be calculated by the following formulas:

15 Freqclass3b15yr = 3

  • Pc1ass3b * (CDF - (LERF - LERF1eak))

= -153 * -218 0.5

  • (2.44£ (5.65£ 5.07£-8)) = 2.16E-7 Freqclass3b15.5yr = -15.5 3

(

-

  • Pclass3b

)

= 153*5 * ~ 218

  • (2.44£ (5.65£ 5.07 E-8)) = 2.23E-7 Reference 32 provides a high winds and tornado assessment, along with other external hazards. Due to the low frequency of strong wind, tornado, and tornado-induced missile events and the high level of protection afforded the CNP to these events, it is concluded that the contribution to plant risk from severe wind events is insignificant [Reference 32]. Similarly, it is concluded that the contribution to plant risk from external flooding, transportation accidents, other external events is insignificant [Reference 32].

The external event contributions to Class 3b frequencies are then combined to obtain the total external event contribution to Class 3b frequencies. The LERF increase is conservatively assumed to be the change in Class 3b frequency.

Table 6 Unit 1 CNP External Event Impact on ILRT LERF Calculation Haz.ard EPRI Accident Class 3b Frequency LERF Increase (from 1 per 15 years to 1 per 15.5 years) 1 per 15 year 1 per 15.5 years External Events 9.14E-07 9.45E-07 3.0SE-08 Internal Events 2.51E-07 2.59E-07 8.35E-09 Combined 1.16E-06 1.20E-06 3.BBE-08 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 increase due to temporarily increasing the interval from 15 to 15.5 years is 3.88E-8 for Unit 1, which meets the guidance for "very small" change in risk, as it is less than 1.0E-7/yr [Reference 4].

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PRA-QNT -009 Evaluation of Risk Significance of Short-Term ILRT Extension 7.0 RESULTS The results from this ILRT extension risk assessment for CNP are summarized in Table 7-1 for Unit 1.

Table 7 Unit 1 ILRT Extension Summary Class Dose 1 in 15 Years Extend to (person- 1 in 15.5 Years rem)

CDF/Year Person- CDF/Year Person-Rem/Year Rem/Year 1 1.01 E+03 2.05E-05 2.07E-02 2.04E-05 2.06E-02 2 3.84E+06 2.89E-07 1.11E+00 2.89E-07 1.11E+00 3a 1.01 E+04 1.01E-06 1.02E-02 1.04E-06 1.05E-02 3b 1.01E+05 2.51E-07 2.53E-02 2.59E-07 2.61E-02 7 3.84E+06 1.10E-06 4.22E+00 1.10E-06 4.22E+00 8 9.68E+06 8.60E-07 8.32E+00 8.60E-07 8.32E+00 Total 2.40E-05 1.37E+01 2.40E-05 1.37E+01 ILRT Dose Rate from 3a and 3b t.Total From 15 N/A 1.14E-03 Dose Rate Years

°lot.Dose From 15 N/A 0.0083%

Rate Years From 15 t.LERF N/A 8.35E-09 Years From 15 t.CCFP% N/A 0.035%

Years Revision 0 Page 29 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension

8.0 CONCLUSION

S AND RECOMMENDATIONS Based on the results from Section 6.0, the following conclusions regarding the assessment of the plant risk are associated with a one-time extension of the Type A ILRT test frequency to 15.5 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 1.0E-06/year and increases in LERF less than 1.0E-07/year. Since the ILRT does not impact CDF, the relevant criterion is LERF. The one-time increase in LERF resulting from a change in the Type A ILRT test interval from 1 in 15 years to 1 in 15.5 years is estimated as 8.35E-9/year for Unit 1 using the EPRI guidance; this value increases negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included. Therefore, the estimated change in LERF is determined to be "very small" using the acceptance guidelines of Regulatory Guide 1.174 [Reference 4].

When external event risk is included, the one-time increase in LERF resulting from a change in the Type A ILRT test interval from 1 in 15 years to 1 in 15.5 years is estimated as 3.88E-8/year for Unit 1 using the EPRI guidance. As such, the estimated change in LERF is determined to be "very small" using the acceptance guidelines of Regulatory Guide 1.174 [Reference 4].

" The effect resulting from temporarily changing the Type A test frequency to 1-per-15.5 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing is 0.0014 person-rem/year for Unit 1. NEI 94-01

[Reference 1] states that a "small" population dose is defined as an increase of :s; 1.0 person-rem per year, or :s; 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The results of this calculation meet these criteria. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible.

The one-time increase in the conditional containment failure probability from the 1 in 15 year interval to 1 in 15.5 year interval is 0.035% for Unit 1. NEI 94-01 [Reference 1]

states that increases in CCFP of :s; 1.5% is "small." Therefore, this increase is judged to be "small."

Therefore, the one-time ILRT interval extension to 15.5 years is considered to be "small" since it represents a small change to the CNP risk profile.

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

Reducing the frequency of Type A tests (I LRTs) 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 Revision 0 Page 30 of 32

PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension of relaxing the ILRT frequency beyond 1 in 20 years has not been evaluated. Beyond testing the performance of containment penetrations, ILRTs also test integrity of the containment structure.

The conclusions for CNP confirm these general conclusions on a plant-specific basis considering the severe accidents evaluated for CNP, the CNP containment failure modes, and the local population surrounding CNP.

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PRA-QNT-009 Evaluation of Risk Significance of Short-Term ILRT Extension A. PRA TECHNICAL ADEQUACY FOR ILRT The CNP Internal Events PRA (including internal flooding) received a full scope peer review in 2015, followed by several focused-scope reviews on various portions of the model, such as pre-initiator HRA and containment hydrogen analysis [Reference 17]. For the purposes of this ILRT extension evaluation, which only requires an assessment of CC-I, only those SRs that are currently "Not Met" are evaluated. Of the remaining Internal Events SRs, no impacts were identified on the ILRT extension evaluation.

The CNP Fire PRA [Reference 28] was subject to a full scope peer review during initial model development in 2010 [Reference 43], with follow-on focused-scope reviews occurring in 2015

[Reference 44] and 2017 [Reference 45]. For the purposes of ILRT extension evaluation, which only requires an assessment of CC-I, only those SRs that are currently "Not Met" are evaluated.

Of the remaining Fire PRA SRs, open items related to IGN-A 1 are identified as a potential impact on the ILRT extension evaluation. This SR is evaluated as Not Met due to the use of fire ignition frequencies based on NUREG/CR-6850 Supplement 1 [Reference 33] instead of NUREG-2169 [Reference 42]. For the purposes of the ILRT extension evaluation, the Fire PRA model is quantified with each set of fire ignition frequencies, and the more limiting results are used.

The CNP Seismic PRA peer review was conducted in 2018 [Reference 46], with a formal F&O closure review in 2019 [Reference 47]. For the purposes of ILRT, which only requires an assessment of CC-I, only those SRs that are "Not Met" are listed and evaluated. However, no SPRA-related SRs are currently "Not Met" [Reference 48], so no evaluation is provided for the purposes of the ILRT extension evaluation.

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