Text

License Amendment Request 256, One-Time Extension of Containment Integrated Leakage Test Interval, Revised Risk Assessment, Enclosure 3
	ML073520398
Person / Time
Site:	Point Beach
Issue date:	12/06/2007
From:	Kindred G, Krantz E; SCIENTECH
To:	; Office of Nuclear Reactor Regulation
References
	17670-0001, Rev. 3
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ENCLOSURE 3 FPL ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE AMENDMENT REQUEST 256 ONE-TIME EXTENSION OF CONTAINMENT NTEGRATED LEAKAGE RATE TEST INTERVAL REVISED RISK ASSESSMENT 161 pages follow

,SC1FENTFCH, CLIENT: Nucleair MIana 'ernen CoIn any BY: F..A. Krantz PACE: I OF 101 FILE NO. 17670-0001. Rev. 3 CH1ECK(ED HY: (,AV. Kindred IDate: 122'

,) 7 I FPL Energy Point Beach Nuclear Power Plant RISK IMPACT ASSESSMENT FOR EXTENDING CONTAINMENT TYPE A TEST INTERVAL Analysis File 17670-0001, Rev. 3 December 2007 Prepared By:

Reviewed By:.

Date:

/Z-zi 6

Date:

Accepted By:

Scientech, a Curtiss-Wright Flow Control Company Idaho Falls, Idaho

SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 2 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table of Contents Description 1.0 CLIENT 2.0 TITLE 3.0 AUTHOR 4.0 PURPOSE 5.0 INTENDED USE OF ANALYSIS RESULTS 6.0 TECHNICAL APPROACH Page No.

4 4

7.0 8.0 9.0 10.0 11.0 12.0 13.0 14.0 INPUT INFORMATION REFERENCES MAJOR ASSUMPTIONS IDENTIFICATION OF COMPUTER CODES DETAILED ANALYSIS COMPUTER INPUT AND OUTPUT 6

7 8

8 9

SUMMARY

OF RESULTS CONCLUSIONS 36 36 36 APPENDIX A APPENDIX B 48 81 17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE:30OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval List of Tables Table I -

Detailed Description for the Eight Accident Classes as defined by EPRI TR-104285 37 Table 2 -

Containment Frequency Measures for a Given Accident Class 38 Table 3 -

Conditional Person-Rem Measures for a Given Accident Class 39 Table 4a -

Unit 1 Baseline Mean Consequence Measures for a Given Accident Class 40 Table 4b -

Unit 2 Baseline Mean Consequence Measures for a Given Accident Class 41 Table 5a -

Mean Consequence Measures for 10 - Year Test Interval for a Given Accident Class Unit 1 42 Table 5b -

Mean Consequence Measures for 10 - Year Test Interval for a Given Accident Class Unit 2 43 Table 6a -

Mean Consequence Measures for 15-1/2 - Year Test Interval for a Given Accident Class Unit 1 44 Table 6b -

Mean Consequence Measures for 15-1/2 - Year Test Interval for a Given Accident Class Unit 2 45 Table 7a -

Effect of Internal Events Hazard Risk on PB ILRT Risk Assessment (Unit 1) 46 Table 7b -

Effect of Internal Events Hazard Risk on PB ILRT Risk Assessment (Unit 2) 47 17670-0001 PB ILRT Rev 3.doc

(

SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 4 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval ANALYSIS FILE: 17670-0001, Rev. 3 1.0 CLIENT FPL Energy - Point Beach Nuclear Power Plant 2.0 TITLE Risk Informed/Risk Impact Assessment for Extending Containment Type A Test Interval 3.0 AUTHOR Eddie A. Krantz 4.0 PURPOSE The purpose of this calculation is to assess the risk impact for extending the Integrated Leak Rate Test (ILRT) interval for the Point Beach Nuclear Plant (PBNP) from ten to fifteen and a half years. In October 26, 1995, the Nuclear Regulatory Commission (NRC) revised 10 CFR 50, Appendix J. The revision to Appendix J allowed individual plants to select containment leakage testing frequency under Option A "Prescriptive Requirements" or Option B "Performance-Based Requirements". PBNP selected the requirements under Option B as its testing program.

The surveillance testing requirements (for Option B of Appendix J) as proposed in NEI 94-01

[Reference 1] for Type A testing is at least once per 10 years 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 was less than 1.00La. PBNP will use this analysis to seek a one-time exemption from a 10 year test interval to a 15-1/2 year test interval.

Revision 2 of this document incorporates the results of an upgrade of the Point Beach Units 1 and 2 Level 2 PRA analysis.

5.0 INTENDED USE OF ANALYSIS RESULTS The results of this calculation will be usedto obtain NRC approval to extend the. Integrated Leak Rate Test interval from one in ten years to one in fifteen and a half years.

6.0 TECHNICAL APPROACH The methodology used for this analysis is similar to the assessments originally performed for Crystal River 3 (CR3) [Reference 2] and Indian Point 3 (IP3) [Reference 3] with enhancements outlined in the EPRI Interim Guidance [Reference 4] and incorporated in numerous subsequent submittals, including Kewaunee [Reference 5] and D. C. Cook [Reference 6]. The ILRT interval extensions requested by these submittals have been approved by the NRC. The impact of age-related degradation of the containment is also evaluated in a sensitivity study (see Appendix B) using methodology similar to that first employed in the Calvert Cliffs Nuclear Plant (CCNPP) response.to an NRC Request for Additional Information (RAI) [Reference 7] and subsequently used in numerous other submittals including those for Comanche Peak and D. C. Cook [References 8 and 6].

This calculation was performed in accordance with NEI 94-01 [Reference 1] guidelines, and the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) findings and risk insights in support of a licensee request for changes to a plant's licensing basis, Regulatory Guide RG 1.174

[Reference 9]. This methodology is similar to that presented in EPRI TR-104285 [Reference 10] and NUREG-1493 [Reference 111 and incorporates the revised guidance and additional information of References 4 and 12.

It uses a simplified bounding analysis approach to evaluate the risk impact of increasing the ILRT Type A interval from 10 to 15-1/2 years by using core damage and containment 17670-0001 PB ILRT Rev 3.doc

,SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 5 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval failure frequency information from the most recent update of the PBNP PRA [Reference 131.

Specifically, the following were considered:

Core damage sequences in which the containment remains intact initially and in the long term (EPRI TR-104285 Class 1 sequences).

Core damage sequences in which containment integrity is impaired due to pre-existing isolation failures of plant components other than those subjected to Type B or Type C tests. For example, this includes sequences with pre-existing liner breach or steam generator manway leakage (EPRI TR-104285 Class 3 sequences). Type B tests measure component leakage across pressure retaining boundaries (e.g., gaskets, expansion bellows and air locks). Type C tests measure component leakage rates across containment isolation valves.

Core damage sequences in which containment integrity is impaired due to containment isolation failures of pathways left 'open' following a plant post-maintenance test. For example, this includes situations in which a valve fails to close following a valve stroke test (EPRI TR-1 04285 Class 6 sequences).

Accident sequences involving containment failure induced by severe accident phenomena (EPRI TR-104285 Class 7 sequences), containment bypassed (EPRI TR-104285 Class 8 sequences),

large containment isolation failures (EPRI TR-1 04285 Class 2 sequences) and small containment isolation 'failure-to-seal' events (EPRI TR-104285 Class 4 and 5 sequences). The sequences of these classes are impacted by changes in Type B and C test intervals, not changes in the Type A test interval (Type A test measures the containment air mass and calculates the leakage from the change in mass over time).

Detailed descriptions of Classes 1 through 8 are excerpted from Reference 10 and provided in Table 1 of this analysis.

The nine steps.of the methodology are:

1) Quantify the baseline risk in terms of frequency per reactor year for each of the eight containment release scenario types identified in the EPRI report.

2) Determine the containment leakage rates for applicable cases, 3a and 3b.

3) Develop the baseline population dose (person-rem) for the applicable EPRI classes.

4)

Determine the population dose rate; also know as population dose risk (person-rem/ry) by multiplying the dose calculated in step (3) by the associated frequency calculated in step (1).

5) Determine the change in probability of leakage detectable only by ILRT, and associated frequency for the new surveillance intervals of interest (Classes 3a and 3b). Note that with increases in the ILRT surveillance interval, the size of the postulated leak path and the associated leakage rate are assumed not to change, however the probability of leakage detectable only by ILRT does increase.

6)

Determine the population dose rate for the new surveillance intervals of interest.

7)

Evaluate the risk impact (in terms of population dose rate and percentile change in population dose rate) for the interval extension cases.

8)

Evaluate the risk impact in terms of LERF.

17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 6 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval

9) Evaluate the change in conditional containment failure probability.

The first seven steps of the methodology calculate the change in dose. The eighth step in the interim methodology calculates the change in LERF and compares it to the guidelines in Regulatory Guide 1.174 (Reference 9). Because the change in ILRT test interval does not impact the CDF, the relevant criterion is LERF. The final step of NEI's interim methodology calculates the change in containment failure probability given the change of ILRT test interval from once-per-10 years to once-per-i 5-1/2 years.

The technical approach for the sensitivity study evaluating the potential impact of age-related corrosion of the steel containment is provided in Appendix B along with the detailed calculations and results.

7.0 INPUT INFORMATION

1.

Updated PRA total Core Damage Frequency (CDF) based upon the calculations done for Reference 15. Release category results based upon 2007 Level 2 model update. The revised Level 2 model incorporated updated model of the induced steam generator tube rupture (ISGTR) modeling. The approach used was a plant specific application of the modeling approach provided in NUREG-1570 (Reference 19).

2.

Population Doses for containment failure modes. Provided from "Applicant's Environmental Report.Operating License Renewal Stage, Point Beach Nuclear Plant, Units 1 and 2, Docket Nos. 50-266 and 50-301, License Nos. DPR-24 and DPR-27, February 2004" [Reference 15].

3.

Probability of Containment Isolation Failure (2.3E-04) from Input #2.

4.

To calculate the probability that a liner leak will be small (Class 3a), use was made of the data presented in NUREG-1493 [Reference 11] and the EPRI Interim Guidance [Reference 4].

NUREG-1493 states that 144 ILRTs have been conducted. The data reported that 23 of 144 tests had allowable leak rates in excess of 1 La. However, of these 23 'failures,' only 4 were found by an ILRT. The others were found by Type B and C testing or were errors in test alignments. Therefore, the number of failures considered for 'small releases' are 4 of 144.

The EPRI Interim Guidance stated that one failure found by an ILRT was found in 38 ILRTs performed after NUREG-1493. Thus, the best estimate of the probability of a small leak, Prob(Class 3a), is calculated as 5/182 = 0.027 [Reference 4].

5.

To calculate the probability that a liner leak will be large (Class 3b), use was made of the data presented in NUREG-1493 [Reference 11] and new data presented by the EPRI Interim Guidance [Reference 4]. One data set found in NUREG-1493 reviewed 144 ILRTs and the EPRI Interim Guidance reviewed additional 38 ILRTs. The largest reported leak rate from those 144 tests was 21 times the allowable leakage rate (La). Since 21 La does not constitute a large release, no large releases have occurred based on the 144 ILRTs reported in NUREG-1493. One failure was found in the 38 ILRTs discussed in the EPRI Interim Guidance and this failure was not considered large.

Because no Class 3b failures have occurred in 182 ILRT tests, the EPRI Interim Guidance suggested that the Jeffery's non-informative prior distribution would be appropriate for the Class 3b distribution. (The rationale for using the Jeffery's non-informative prior distribution was discussed in Reference 4.)

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Prob(Class 3b) = Failure probability = (# of failures (0)+ 11/2)/(Number of tests (182) + 1)

The number of large failures is zero and the probability is Prob(Class 3b) = 0.5/183 = 0.0027

8.0 REFERENCES

1.

NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10CFR Part 50, Appendix J, July 26, 1995, Revision 0.

2.

"Crystal River-Unit 3 - License Amendment Request #267, Revision 2, Supplemental Risk-Informed Information in Support of License Amendment Request #267," Florida Power, 3F0601-06, June 20, 2001.

3.

"Supplemental Information Regarding Proposed Change to Section 6.14 of the Administrative Section of the Technical Specification", Entergy, IPN-01-007, Indian Point 3 Nuclear Power Plant, January 18, 2001.

4.

J. Haugh, J. M. Gisclon, W. Parkinson, K. Canavan, "Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals", Rev. 4, EPRI, November, 2001.

5.

"License Amendment Request 198 to the Kewaunee Nuclear Power Plant Technical Specifications for one-time extension of containment integrated leak rate test interval," Nuclear Management Company, June 20, 2003.

6.

"Donald C. Cook Nuclear Plant Units 1 and 2, Response to Nuclear Regulatory Commission Request for Additional Information Regarding the License Amendment Request for a One-time Extension of Integrated Leakage Rate Test Interval," Indiana Michigan Power Company, November 11, 2002.

7.

"Calvert Cliffs Nuclear Power Plant Unit No. 1; Docket No. 50-317," Constellation Nuclear letter to USNRC, March 27, 2002.

8.

"Comanche Peak Steam Electric Station (CPSES), Docket Nos. 50-445 and 50-446, Response to Request for Additional Information Regarding License Amendment Request (LAR) 01-14 Revision to Technical Specification (TS) 5.5.16 Containment Leakage Rate Testing Program," TXU Energy letter to USNRC, June 12, 2002.

9.

Regulatory Guide 1.174, "An Approach for using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis", Revision 1, November 2002.

10. EPRI TR-1 04285, "Risk Assessment of Revised Containment Leak Rate Testing Intervals" August 1994.

11. NUREG-1493, "Performance-Based Containment Leak-Test Program, July 1995".

12. NEI Memo, "One-Time Extension of Containment Integrated Leak Rate.Test Interval - Additional Information", Nuclear Energy Institute, November 30, 2001.

13. Point Beach PRA Model, Revision 3.17, January 18, 2006.

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SUBJECT:

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14. United States Nuclear Regulatory Commission, "Individual Plant Examination: Submittal Guidance," NUREG-1335, August 1989.

15. Applicant's Environmental Report Operating License Renewal Stage, Point Beach Nuclear Plant, Units 1 and 2, Docket Nos. 50-266 and 50-301, License Nos. DPR-24 and DPR-27, February 2004

16. U.S. Nuclear Regulatory Commission, "Severe Accident Risks: An Assessment for Five U.S.

Nuclear Power Plants, NUREG-1150", December 1990.

17. 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, March 27, 2002.

18. "Impact on the Point Beach Nuclear Plant Large Early Release Frequency (LERF) Due to Level 2 Modeling Enhancements", a white paper, Scientech, December 6,,2007..

19. U.S. Nuclear Regulatory Commission, "Risk Assessment of Severe Accident-Induced Steam Generator Tube Rupture", NUREG-1 570, March 1998.

9.0 MAJOR ASSUMPTIONS:

1. The containment leakage for Class I sequences is assumed to be 1 La.

[Reference 4]

2. The containment leakage for Class 3a sequences is assumed to be 10 La.

[Reference 4]

3.

The containment leakage for Class 3b sequences is assumed to be 35 La.

[Reference 4]

4.

Because Class 8 sequences are containment bypass sequences (e.g., Steam Generator Tube Rupture - SGTR, Interfacing Systems Loss of Coolant Accidents - ISLOCA), potential releases are primarily directly to the environment. Therefore, the integrity of the containment structure will not significantly impact the release magnitude.

10.0 IDENTIFICATION OF COMPUTER CODES None used.

17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 9 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07 SLUBJECT: Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 11.0 DETAILED ANALYSIS:

11.1 Internal Events Analysis 11.1.1 Step I - Quantify the baseline freauencv per reactor year for each of the eiaht accident classes presented in Table 1.

As mentioned in the methods section above, step 1 quantifies the annual frequencies for the eight accident classes defined in Reference 10. Class 3 was evaluated based on Interim Guidance and Additional Information from EPRI and NEI [References 4 and 12].

Reference 18 provides the following results of the latest PBNP PRA update. Also included are the accident classes corresponding to the PBNP source term categories.

PBNP Source Term Description EPRI Unit 1 Unit 2 Category Accident Frequencies Frequencies Class 1

Intact 1

3.61 E-05 3.12E-05 2

Late Basemat 7b 2.95E-07 2.71 E-07 3

Late SGTR 8c 2.62E-06 2.66E-06 4

Late Rupture 7a 9.06E-06 1.55E-05 5

Early Liner 7c 2.94E-08 2.70E-08 6

Early Unisolated 2

1.50E-08 1.55E-08 7

Early ISLOCA 8a 2.37E-07 2.37E-07 8

Early SGTR 8b 1.83E-06 1.91 E-06 Total Internal Events CDF 5.02E-05 5.18E-05 Total Internal Events LERF 2.11E-06 2.19E-06 The following plant specific features and experience have been considered in the development of the frequencies presented in the previous table:

Any of three positive displacement charging pumps (not dependent upon component cooling water or service water) can adequately cool the RCP seals.

The Point Beach steam generators have been replaced and none have experienced any significant degradation and have very low percentages of plugged tubes.

The Point Beach plant has two steam generators per unit.

The annual frequencies for each accident class are assessed as follows:

Class 1 Sequences, This group consists of all core damage accident progression bins for which the containment remains intact. For this analysis the associated maximum containment leakage for this group is I La. The frequency for these sequences is determined as follows:

Class1 -Frequency

= NCF - Class_3aFrequency - Class_3bFrequency Where:

NCF

= Frequency in which containment leakage is at or below maximum allowable Technical Specification leakage.

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval

= 3.61 E-05/yr (Unit 1)

= 3.12E-05/yr (Unit 2)

[From table above for STCI]

[From table above for STC1]

Class_3aFrequency Class_3bFrequency Therefore.

Class 1 Frequency Class 1 -Frequency

= Frequency of small pre-existing containment liner leakage

= 9.75E-07/yr (Unit 1)

[See below]

= 8.42E-07/yr (Unit 2)

[See below]

= Frequency of large pre-existing containment liner leakage

= 9.75E-08/yr (Unit 1)

[See below]

= 8.42E-08/yr (Unit 2)

[See below]

= 3.61 E -05/yr - 9.75E -07/yr - 9.75E -08/yr = 3.50E-05/yr (Unit 1)

= 3.12E -05/yr - 8.42E -07/yr - 8.42E -08/yr = 3.03E-05/yr (Unit 2)

Class 2 Sequences. This group consists of all core damage accident progression bins in which the containment isolation system fails to function during the accident progression. These sequences are dominated by failures to close of greater than 2-inch diameter but less than 5-inch diameter containment isolation valves. Failure to close of very large isolation valves (greater than 5 inches) that could lead to a large early release (LER) have a much lower frequency.

The frequency for these sequences is determined as follows:

Class 2 Frequency = Frequency of STC 6 Where Class 2 Frequency

= Frequency of EPRI Class 2 given a 3-in-10 years ILRT interval Class 2 Frequency

= 1.50E-08 (Unit 1) (Table above) and

= 1.55E-08 (Unit 2) (Table above)

Class 3a Sequences. This group consists of all core damage accident progression bins for which a small pre-existing leakage in the containment structure (i.e., containment liner) exists. This type of failure is identifiable only from an ILRT and, therefore, is affected by a change in ILRT testing frequency.. Evaluation of this class is based on EPRI TR-104285 [Reference 10], the EPRI Interim Guidance [Reference 4] and the NEI Additional Information [Reference 12].

Class_3aFrequency

= Probclass3a * (CDFTota l-CDFIndep)

Where, Class_3aFrequency

= Frequency of EPRI Class 3a given a 3-in-10 years ILRT interval ProbClass3a CODFT~tal

= Probability of small pre-existing containment liner leakage

= 0.027

[Section 7]

= PB Core Damage Frequency

= 5.02E-05/yr (Unit 1)

[Table above]

and 17670-0001 PB ILRT Rev 3.doc

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161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval

= 5.18E-05/yr (Unit 2)

[Table above]

CDFlndep

= CDF for those individual sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences:

EPRI Class 2 = 1.50E-08/yr EPRI Class 7a = 9.06E-06/yr EPRI Class 7b = 2.95E-07/yr EPRI Class 7c = 2.94E-08/yr EPRI Class 8a = 2.37E-07/yr EPRI Class 8b = 1.83E-06/yr EPRI Class 8c = 2.62E-06/yr

= 1.41 E-05/yr (Unit 1) and EPRI Class 2 = 1.55E-08/yr a

EPRI Class 7a = 1.55E-05/yr EPRI Class 7b = 2.71 E-07/yr a

EPRI Class 7c = 2.70E-08/yr EPRI Class 8a = 2.37E-07/yr EPRI Class 8b = 1.91 E-06/yr EPRI Class 8c = 2.66E-06/yr

= 2.06E-05/yr (Unit 2)

Therefore, Class_3aFrequency

= 0.027 * (5.02E-05/yr - 1.41 E-05/yr) = 9.75E-07 (Unit 1) and

= 0.027 * (5.18E-05/yr - 2.06E-05/yr) = 8.42E-07 (Unit 2)

Class 3b Sequences. This group consists of all core damage accident progression bins for which a large pre-existing leakage in the containment structure (i.e., containment liner) exists. This type of failure is identifiable only from an ILRT and, therefore, is affected by a change in ILRT testing frequency. Evaluation of this class is based on EPRI TR-104285 [Reference 10], the EPRI Interim Guidance [Reference 4] and the NEI Additional Information [Reference 12].

Class_3bFrequency

= Probclass3b * (CDFTotal - CDFIndep)

Where, Class_3b_Frequency

= Frequency of IPRI Class 3b given a 3-in-10 years ILRT interval ProbClass3b

= Probability of large pre-existing containment liner leakage

= 0.0027

[Section 7]

CDFmotal

= PB Core Damage Frequency

= 5.02E-05/yr (Unit 1)

[Table above]

and

= 5.18E-05/yr (Unit 2)

[Table above]

CDFIndep

= 1.41 E-05/yr (Unit 1) (from above) and

= 2.06E-05/yr (Unit 2) (from above)

Therefore, Class_3bFrequency

= 0.0027 * (5.02E-05/yr - 1.41 E-05/yr) = 9.75E-08 (Unit 1) 17670-0001 PB ILRT Rev 3.doc

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval and

= 0.0027 * (5.18E-05/yr - 2.06E-05/yr) = 8.42E-08 (Unit 2)

Class 4 Sequences. This group consists of all core damage accident progression sequences in which the containment isolation system function fails due to a pre-existing failure-to-seal of Type B test component(s). Consistent with EPRI Interim Guidance (Reference 4), because these failures are detected by Type B tests and not by the Type A ILRT, this group is not evaluated further in this analysis.

Class 5 Sequences. This group consists of all core damage accident progression sequences in which the containment isolation system function fails due to a pre-existing failure-to-seal of Type C test component(s). Consistent with EPRI Interim Guidance (Reference 4), because these failures are detected by Type C tests, this group is not evaluated any further.

Class 6 Sequences. This group consists of all core damage accident sequences in which the containment isolation function is failed due to 'other' pre-existing failure modes (e.g., pathways left open or misalignment of containment isolation vales following a test/maintenance evolution).

Consistent with EPRI Interim Guidance (Reference 4), because these failures are detected by Type B or C tests, this group is not evaluated any further.

Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs (i.e., H2 combustion).

The EPRI Class 7 is subdivided in this report to reflect the subdivision into those sequences that are due to interfacing systems LOCAs, steam generator tube ruptures that occur early and steam generator tube ruptures that occur late.

0 Class 7a - Late containment rupture.

Class 7b - Late Basemat failure.

Class 7c - Early liner failure.

Class_7aFrequency

= Frequency of late containment rupture

= 9.06E-06 (Unit 1)

[From above table]

and

= 1.55E-05 (Unit 2)

[From above table]

Class_7bFrequency Class_7cFrequency

= Frequency of late Basemat failure

= 2.95E-07 (Unit 1) and

= 2.71 E-07 (Unit 2)

= Frequency of early liner failure

= 2.94E-08 (Unit 1) and

= 2.70E-08 (Unit 2)

[From above table]

Therefore, the Class 7 frequencies are:

Class_7_Frequency

= 9.38E-06 (Unit 1) and 17670-0001 PB ILRT Rev 3.doc

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval

= 1.58E-05 (Unit 2)

Class 8 Sequences. This group consists of all core damage accident progression bins in which containment bypass occurs.

The EPRI Class 8 is subdivided in this report to reflect the subdivision into those sequences that are due to interfacing systems LOCAs, steam generator tube ruptures that occur early and steam generator tube ruptures that occur late.

Class 8a - Containment bypass due to interfacing systems LOCAs. In these sequences it is assumed that containment bypass and core melt result in a high release. No credit for Reactor Building retention is taken for these sequences, Class 8b - Containment bypass due to steam generator tube rupture events leading to early core damage.

Class 8c - Containment bypass due to steam generator tube rupture events leading to late core damage.

Class_8aFrequency

= Frequency of ISLOCA

= 2.37E-07 (Unit 1)

[From above table]

and

= 2.37E-07 (Unit 2)

[From above table]

Class_8bFrequency

= Frequency of SGTRE

= 1.83E-06 (Unit 1)

[From above table]

and

= 1.91 E-06 (Unit 2)

[From above table]

Class_8c_Frequency

= Frequency of SGTRL

= 2.62E-06 (Unit 1)

[From above table]

and

= 2.66E-06 (Unit 2)

[From above table]

Note for this class the maximum releases are not based on normal containment leakage, because most of the releases are directly to the environment. Therefore, the integrity of the containment structure will not significantly impact the release magnitude.

The annual frequencies for the eight classes are summarized in Table 2.

11.1.2 Step 2 - Containment Leakage Rates This step defines the containment leakage rates for EPRI accident Classes 3a and 3b. As defined in Step 1, accident Class 3a and 3b are plant accidents with pre-existing containment leakage pathways (designated as 'small' and 'large') that are identifiable only when performing a Type A ILRT. The EPRI Interim Guidance (Reference 4) recommends containment leakage rates of 1 OLa and 35La for accident Classes 3a and 3b, respectively. These values are consistent with previous ILRT frequency extension submittal applications (e.g., Reference 5). La is the plant Technical Specification maximum allowable containment leak rate. By definition and per the EPRI Interim Guidance (Reference 4) and previously approved methodology (Reference 5) the containment leakage rate for Class 1 (i.e.,

accidents with containment leakage at or below maximum allowable Technical Specification leakage) is 1 La.

17670-0001 PB ILRT Rev 3.doc

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PAGE: 14 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 161 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 11.1.3 Step 3 - Develop plant specific person-rem dose (population dose) per reactor year for each of the eight accident classes In accordance with guidance given by Reference 10, this step estimates the baseline population dose for each of the eight EPRI accident classes. The EPRI Interim Guidance (Reference 4) recommends two options for calculating population dose:

a Use of NUREG-1150 dose calculations.(Reference 16)

Use of plant-specific dose calculations Because Point Beach has a Level 3 PSA (Reference 15) and associated plant-specific dose, this risk assessment uses plant specific dose results. The Point Beach population doses were calculated using the MACCS2 code and are provided below from Table F.1.4 of Reference 15.

PB Release Category Person-SV Dose (REM)

Late SGTR 1.39E+03 139000 Early SGTR 1.88E+03 188000 Isolation Failure 1.13E+03 113000 ISLOCA 1.13E+04 1130000' Internal Other CM 3.86E+01 3860 Sequences Reference 15 documents an assessment of the PBNP site population dose consequences due to the accidental release of radiological materials resulting from several severe accident scenarios.

1 The dose from ISLOCA was assumed to be 10 times the dose of Isolation Failure in the analysis of Reference 15.

17670-0001 PB ILRT Rev 3.doc

CaSCIENTECH.,

PAGE: 15 CLIENT: Nuclear Management Company BY: E. A. Krantz 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/0

SUBJECT:

Risk-Informed /Risk impact Assessment for Extending Containment Type A Test Interval OF 6/07 Pooulation Dose for EPRI Class 1.

The dose for the 'no containment failure" EPRI class 1 sequences is based on PB Release Category "Internal Other CM Sequences" (see table above). Therefore, Class_1_Dose

= 3.86E+01 person-sv *100 person-rem / person-sv

= 3.86E+03 person-rem

[Table 3]

Population Dose for EPRI Class 2.

The 50-mile population dose for the EPRI accident Class 2 (Large Containment Isolation Failures, failure-to-close) is based on the Point Beach release category "Isolation Failure" (see table above).

Therefore, Class_2_Dose

= 1.13E+03 person-sv *100 person-rem / person-sv

= 1.13E+05 person-rem

[Table 3]

Population Dose for EPRI Class 3.

The 50-mile population dose for the EPRI accident Class 3a (Small Isolation Failures-Liner Breach) and accident Class 3b (Large Isolation Failures-Liner Breach), per Reference 4), are taken as factors of 1 OLa and 35La (Reference 4), respectively, times the population dose of EPRI accident Class 1.

Therefore, Class 3a Dose Class_3b-Dose Class 3a Dose Class_3bDose Class_ 3a Dose Class_3bDose

= 10

Class_1 Dose

=35

Class_1 -Dose

= 10

3.86E+03 person-rem

= 35

3.86E+03 person-rem

= 3.86E+04 person-rem

= 1.35E+05 person-rem Population Dose for EPRI Class 4, 5, and 6.

Per the EPRI Interim Guidance (Reference 4), EPRI accident Classes 4 (Small Isolation Failure -

failure-to-seal, Type B test), 5 (Small Isolation Failure - failure-to-seal, Type C test), and 6 (Containment Isolation Failures, dependent failures, personnel errors) are not affected by ILRT frequency and are not analyzed as part of this risk assessment. Therefore no selections of population dose estimates are made for these accident classes.

17670-0001 PB ILRT Rev 3.doc

(OS CIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 16 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval PoDulation Dose for EPRI Class 7.

The 50-mile population dose for the EPRI accident Class 7 (Containment failure due to phenomenology) is assumed to be similar to and is based on the Point Beach release category "Late SGTR" (see table above). Therefore, Class_7_Dose

= 1.39E+03 person-sv*100 person-rem / person-sv

= 1.39E+05 person-rem

[Table 3]

Population Dose for EPRI Class 8a.

The 50-mile population dose for the EPRI accident Class 8a (Containment bypass due to interfacing systems LOCA) is based on the Point Beach release category "ISLOCA" (see table above).

Therefore, Class_8aDose

= 1.13E+06 person-rem

[Table 3]

Population Dose for EPRI Class 8b.

The 50-mile population dose for the EPRI accident Class 8b (Containment bypass due to early steam generator tube ruptures) is based on the Point Beach release category "Early SGTR" (see table above). Therefore, Class_8bDose

= 1.88E+03 person-sv*100 person-rem / person-sv

= 1.88E+05 person-rem

[Table 3]

Population Dose for EPRI Class 8c.

The 50-mile population dose for the EPRI accident Class 8c (Containment bypass due to late steam generator tube ruptu res) is based on the Point Beach release category "Late SGTR" (see table above). Therefore, Class_8cDose

= 1.39E+03 person-sv*100 person-rem / person-sv

= 1.39E+05 person-rem

[Table 3]

The 50-mile population dose (person-rem) for each release category is summarized in Table 3. The doses provided in this table are applicable to each unit in accordance with Reference 15 (Note: The use of dose results for the 50-mile radius around the plant as a "figure of merit" in the risk evaluation is consistent with past ILRT frequency extension submittals, and the EPRI Interim Guidance (Reference 4)).

11.1.4 Step 4 - Estimate Baseline Population Dose Rate per reactor year for each of the eight accident classes This step calculates the baseline dose rates for each of the eight EPRI's accident classes. The calculation is performed by multiplying the dose calculated in Step 3 (Table 3) by the associated frequency calculated in Step 1 (Table 2). Since the conditional containment pre-existing leakage probabilities for EPRI accident classes 3a and 3b are based on a 3-per-1 0 year ILRT frequency, the calculated baseline results reflect a 3-per-10 year ILRT surveillance frequency.

17670-0001 PB ILRT Rev 3.doc

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PAGE: 17 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 161 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class 1 DoseRate Class 2 DoseRate Class 3a_DoseRate Class 3b_DoseRate Class 7 DoseRate Class 8a_DoseRate Class 8b_DoseRate Class_8c_DoseRate Class_1_Dose Class 2 Dose Class 3a Dose Class 3b Dose Class 7 Dose Class 8a Dose Class 8b Dose Class_8cDose Class 1 Frequency Class 2 Frequency Class 3a Frequency Class 3b Frequency Class_7_Frequency Class 8a Frequency Class_8bFrequency Class 8c Frequency Where ClassIDoseRate Class_2 DoseRate Class 3a_DoseRate Class 3b_DoseRate Class 7_DoseRate Class 8a_DoseRate Class 8bDoseRate ClassScDoseRate Class 1 Dose

=

Class 2 Dose

=

Class 3a Dose

=

Class 3b Dose

=

Class 7 Dose

=

Class 8a Dose

=

Class 8b Dose

=

Class_8cDose

=

For Unit.1, Class 1 Frequency Class_2_Frequency Class_3aFrequency Class_3bFrequency Class 7 Frequency Class_8aFrequency Class_8bFrequency ClassScFrequency EPRI accident Class_1_dose rate given a 3-in-10 years ILRT interval EPRI accident Class_2_dose rate given a 3-in-10 years ILRT interval EPRI accident Class_3a dose rate given a 3-in-10 years ILRT interval EPRI accident Class 3b dose rate given a 3-in-10 years ILRT interval EPRI accident Class 7 dose rate given a 3-in-10 years ILRT interval EPRI accident Class 8a dose rate.given a 3-in-1 0 years ILRT interval EPRI accident Class_8b dose rate given a 3-in-10 years ILRT interval EPRI accident Class_8c dose rate given a 3-in-1 0 years ILRT interval EPRI accident Class 1 dose EPRI accident Class 2 dose EPRI accident Class 3a dose EPRI accident Class_3b._dose EPRI accident Class 7 dose EPRI accident Class 8a dose, EPRI accident Class_8b dose EPRI accident Class_8c-dose 3.86E+03 person-rem (Table 3) 1.13E+05 person-rem (Table 3) 3.86E+04 person-rem (Table 3) 1.35E+05 person-rem (Table 3) 1.39E+05 person-rem (Table 3) 1.13E+06 person-rem (Table 3) 1.88E+05 person-rem (Table 3) 1.39E+05 person-rem (Table 3)

=

Frequency of EPRI accident Class I given a 3-in-10 years ILRT interval

=

3.50E-05/ry (Table 2)

=

Frequency of EPRI accident Class 2 given a 3-in-10 years ILRT interval

=

1.50E-08/ry (Table 2)

=

Frequency of EPRI accident Class 3a given a 3-in-10 years ILRT interval

=

9.75E-07/ry (Table 2)

=

Frequency of EPRI accident Class 3b given a 3-in-10 years ILRT interval

=

9.75E-08/ry (Table 2)

=

Frequency of EPRI accident Class 7 given a 3-in-10 years ILRT interval

=

9.38E-06ry (Table 2)

=

Frequency of EPRI accident Class 8a given a 3-in-10 years ILRT interval

=

2.37E-07/ry (Table 2)

=

Frequency of EPRI accident Class 8b given a 3-in-10 years ILRT interval

=

1.83E-06/ry (Table 2)

=

Frequency of EPRI accident Class 8c given a 3-in-10 years ILRT interval 17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 18 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed I Risk impact Assessment for Extending Containment Type A Test Interval

=

2.62E-06ry (Table 2)

For Unit 2, Class1 2Frequency Class_23Frequency Class_3abFrequency Class 3b Frequency Class_7_Frequency Class_8aFrequency Class_8bFrequency Class 8c-Frequency Therefore, for Unit 1, Class_ 1_DoseRate Class 2 DoseRate Class 3a_DoseRate Class 3b DoseRate Class_7_DoseRate Class 8a DoseRate Class 8b DoseRate Class 8c DoseRate Therefore, for Unit 2, Class_ _DoseRate Class 2 DoseRate Class 3a DoseRate Class 3b DoseRate Class 7 DoseRate Class 8a DoseRate Class 8bDoseRate Class_8cDoseRate

=

Frequency of EPRI accident Class 1 given a 3-in-10 years ILRT interval

=

3.03E-05/ry (Table 2)

=

Frequency of EPRI accident Class 2 given a 3-in-10 years ILRT interval

=

1.55E-08/ry (Table 2)

=

Frequency of EPRI accident Class 3a given a 3-in-10 years ILRT interval

=

8.42E-07/ry (Table 2)

=

Frequency of EPRI accident Class 3b given a 3-in-10 years ILRT interval

=

8.42E-08/ry (Table 2)

=

Frequency of EPRI accident Class 7 given a 3-in-1 0 years ILRT interval

=

1.58E-05/ry (Table 2)

=

Frequency of EPRI accident Class 8a given a 3-in-10 years ILRT interval

=

2.37E-07/ry (Table 2)

=

Frequency of EPRI accident Class 8b given a 3-in-10 years ILRT interval

=

1.91 E-06/ry (Table 2)

=

Frequency of EPRI accident Class 8c given a 3-in-10 years ILRT interval

=

2.66E-06/ry (Table 2) 3.86E+03 1.13E+05 3.86E+04 1.35E+05 1.39E+05 1.13E+06 1.88E+05 1.39E+05 3.86E+03 1.13E+05 3.86E+04 1.35E+05 1.39E+05 1.13E+06 1.88E+05 1.39E+05 3.50E-05 1.50E-08 9.75E-07 9.75E-08 9.38E-06 2.37E-07 1.83E-06 2.62E-06 3.03E-05 1.55E-08 8.42E-07 8.42E-08 1.58E-05 2.37E-07 1.91 E-06 2.66E-06 1.35E-01 1.70E-03 3.76E-02 1.32E-02 1.30E+00 2.68E-01 3.44E-01 3.64E-01 1.17E-01 1.76E-03 3.25E-02 1.14E-02 2.20E+00 2.68E-01 3.59E-01 3.69E-01 (person-rem/ry)

(person-rem/ry)

Tables 4a and 4b summarize the resulting baseline population dose rates by EPRI accident classes.

17670-0001 PB ILRT Rev 3.doc

kSCIENTECH.

PAGE: 19 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 161 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 11.1.5 Step 5 - Change in Probability of Detectable Leakage This step calculates the change in probability of leakage detectable only by ILRT, and associated frequency for the new surveillance intervals of interest. Note that with increases in the ILRT surveillance interval, the size of the postulated leak path and the associated leakage rate are assumed not to change, however the probability of leakage detectable only by ILRT does increase.

According to NUREG-1493 (Reference 11) and the EPRI Interim Guidance (Reference 4), the calculation of the change in the probability of a pre-existing ILRT-detectable containment leakage is based on the relationship that relaxation of the ILRT interval results in increasing the average time that a pre-existing leak would exist undetected. Specifically, the relaxation of the Type A ILRT interval from 3-in-10 years to 1-in-10 years will increase the average time that a leak detectable only by an ILRT goes undetected from 182 to 603 months, a factor of 3.33 increase (60/18). Therefore, the change in probability of leakage due to the ILRT interval extension is calculated by applying a multiplier factor determined by the ratio of the average times of undetection for the two ILRT interval cases.

From Section 7.0 "Input Information", the calculated pre-existing ILRT detectable leakage probabilities based on 3 in-1 0 years ILRT frequency is 0.027 for small pre-existing leakage (EPRI accident class 3a) and 0.0027 for large pre-existing leakage (EPRI accident class 3b).

Point Beach has been operating under a 1-in-10 years ILRT testing frequency consistent with the performance-based Option B of 10 CFR Part 50, Appendix J. As a result, the baseline leakage probabilities, (which are based on a 3-in-1 0 years ILRT frequency) must be revised to reflect the current 1-in-10 years PB ILRT testing frequency. This is performed as follows:

Probciass_3a_10 = Probclass_3a* [ Survtest1 0/18]

Probcoass 3b 10 = Probclass_3b * [ Survtestjo/I 8]

Where:

Probciass_3alo = probability of small pre-existing containment liner leakage given a 1-in-10 years ILRT frequency.

Probc1 ass_3b_10 = probability of large pre-existing containment liner leakage given a 1-in-10 years ILRT frequency.

Probciass 3a = probability of small pre-existing containment liner leakage given a 3-in-10 years ILRT frequency = 0.027 [Section 7.0]

Probciass_3b = probability of large pre-existing containment liner leakage given a 3-in-10 years ILRT frequency = 0.0027 [Section 7.0]

Survtestj 0 = surveillance interval of interestl2 = 10 years*12months/year/2 = 60 months Therefore, Probciass _3a_10 = 0.027 * [60/18] = 0.09 Probcoass 3b 10 = 0.0027 [60/18] = 0.009 2 One half of the test interval for 3 tests in 10 years is approximately 18 months.

3 One half of the test interval for 1 test in 10 years is 60 months.

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PAGE: 20 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 161 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Similarly, the pre-existing ILRT detectable leakage probabilities for the 1-in-1 5-1/2 years ILRT frequency being analyzed by PB are calculated as follows:

Probclass_3a_15 Probctass_3a

Survtest1 5 / 18 Probclass_3b_15 Probclass_3b

Survtest1 5 / 18 Where:

Probciass 3a 15 probability of small pre-existing containment liner leakage given a I-in-15-1/2 years ILRT frequency.

Probciass 3b 15 = probability of large pre-existing containment liner leakage given a 1-in-15-1/2 years ILRT frequency.

Probc~ass 3a = probability of small pre-existing containment liner leakage given a 3-in-1 0 years ILRT frequency = 0.027 [Section 7.0]

Probciass_3b = probability of large pre-existing containment liner leakage given a 3-in-1 0 years ILRT frequency = 0.0027 [Section 7.0]

Survtest1 5 = surveillance interval of interest/2 = 15-1/2 years*12months/year/2 = 93 months Therefore, Probclass_3a_15 = 0.027 * [93 / 18] = 0.1395 Probcjass_3b_15 = 0.0027 *[ 93 /18] = 0.01395 Given the above revised leakage probabilities, the frequencies of the EPRI accident classes calculated in Step 1 also need to be revised to reflect the change in leakage probabilities.

As previously stated, Type A tests impact only Class 1 and Class 3 sequences. Therefore, EPRI accident Class 1 frequency changes are calculated similar to Step 1, and the other EPRI classes (2, 7, and 8) remain the same.

Revised Frequency of EPRI Class 3a Sequences.

Consistent with EPRI Interim Guidance (Reference 4), the frequency per reactor year for this category is calculated as:

Class_3aFrequency1 o = Probciass_3a_10 * [CDFTotal - CDFindep]

Class_3aFrequency _5 = Probclass_3a_15 * [CDFTotal - CDFIndep]

Where:

Class_3aFrequency 10 = frequency of small pre-existing containment liner leakage given a 1 -in-10 years ILRT interval Class_3aFrequencys15 = frequency of small pre-existing containment liner leakage given a 1-in 1/2 years ILRT interval 17670-0001 PB ILRT Rev 3.doc

.SCIENTECH.

PAGE: 21 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 161 P

161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Probcass 3a10 l-probability of small pre-existing containment liner leakage given a 1-in-1 0 years ILRT frequency = 0.09 [See above write-up]

Probciass 3a 15 = probability of small pre-existing containment liner leakage given a 1-in-1 5-1/2 years ILRT frequency = 0.1395 [See above write-up]

CDF-rotai ul = PB U1 PSA Li core damage frequency = 5.02E-05/ry [See step 1 write-up]

CDFTota U2 = PB U2 PSA Li core damage frequency = 5.18E-05/ry CDFIndep ul = CDF for those individual Unit 1 sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences = 1.41 E-05/yr [See step 1 write-up]

CDFIndep U2 = CDF for those individual Unit 2 sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences 2.06E-05/yr [See step 1 write-up]

Therefore, for Unit 1, Class_3aFrequency_10 = 0.09 * (5.02E-05/yr - 1.41 E-05/yr) = 3.25E-06/ry Class_3aFrequency 15 = 0.1395 * (5.02E-05/yr-1.41 E-05/yr) = 5.04E-06/ry For Unit 2, Class_3aFrequency1l0 = 0.09 * (5.18E-05/yr - 2.06E-05/yr) = 2.81 E-06fry Class_3aFrequency_15 = 0.1395 * (5.18E-05/yr-2.06E-05/yr) = 4.35E-06/ry Frequency of EPRI Class 3b Sequences.

Consistent with EPRI Interim Guidance (Reference 4), the frequency per reactor year for this category is calculated as:

Class_3b Frequency_10 = Probciass_3bo10 [CDFTota l-CDFIndep}

Class_3b Frequency_15 = Probciass_3b_15 [CDFTotaI - CDFIndep}

Where:

Class_3bFrequency_10 = frequency of large pre-existing containment liner leakage given a 1 -in-10 years ILRT interval Class_3bFrequency 1 5 = frequency of large pre-existing containment liner leakage given a 1-in-1 5-1/2 years ILRT interval Probciass_3b_10 = probability of large pre-existing containment liner leakage given a 1-in-1 0 years ILRT frequency = 0.009 [See above write-up]

Probciass_3b_5 = probability of large pre-existing containment liner leakage given a 1-in-1 5-1/2 years ILRT frequency = 0.01395 [See above write-up]

CDFTotal ui = PB Ul PSA Li core damage frequency = 5.02E-05/ry [See step 1 write-up]

CDFTotal U2 = PB U2 PSA Li core damage frequency = 5.18E-05/ry CDFindep U1 = CDF for those individual Unit I sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences = 1.41 E-05/yr [See step 1 write-up]

17670-0001 PB ILRT Rev 3.doc

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PAGE: 22 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval COFindep U2 = CDF for those individual Unit 2 sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences = 2.06E-05/yr [See step 1 write-up]

Therefore, for Unit 1, Class_3bFrequency_10 = 0.009 * (5.02E-05/yr - 1.41 E-05/yr) = 3.25E-07/ry Class_3bFrequency_15 = 0.01395 * (5.02E-05/yr - 1.41 E-05/yr) = 5.04E-07/ry For Unit 2, Class_3bFrequency 10 = 0.009 * (5.18E-05/yr - 2.06E-05/yr) = 2.81 E-07/ry Class_3bFrequencyIs = 0.01395 * (5.18E-05/yr - 2.06E-05/yr) = 4.35E-07/ry 11.1.6 Sten 6 -Population Dose Rate for New ILRT Interval This step, per the EPRI Interim Guidance (Reference 4), calculates the population dose rate for the new surveillance intervals of interest by multiplying the population dose (Table 3) by the frequency for each of the eight EPRI's accident classes (Table 2). In addition, sum the accident class dose rates to obtain the total dose rate. Per the EPRI Interim Guidance (Reference 4), EPRI accident Classes 4 (Small Isolation Failure - failure-to-seal, Type B test), 5 (Small Isolation Failure - failure-to-seal, Type C test), and 6 (Containment Isolation Failures, dependent failures, personnel errors) are not affected by ILRT frequency and are not analyzed as part of this risk assessment. Therefore no selections of population dose estimates are. made for these accident classes. The calculation for a 1-in-10 years ILRT interval is as follows:

The calculation for a 1-in-10 years ILRT interval is as follows:

Class 1 DoseRate-10 Class 2 DoseRatelo Class_3a_ DoseRatelo Class 3b_ DoseRate-lO Class 7 DoseRate-lo Class 8a_ DoseRate.O1 Class 8b_ DoseRate-10 Class_8c_ DoseRate.1o Class_1_Dose Class 2 Dose.

Class_3aDose Class 3b Dose Class_7_Dose Class 8a Dose Class_8bDose Class_8cDose

/

Class_1_Frequencylo Class 2_Frequency1 o Class_3aFrequency1 o Class 3bFrequency1 o Class 7 Frequency1 0 Class 8aFrequency10 Class_8bFrequency1 o Class_8cFrequency1 o Where Class_1_DoseRatelo Class 2 DoseRate-10 Class 3a_ DoseRate-lO Class 3b_ DoseRate-10 Class 7 DoseRate-lO Class 8_a DoseRate-10 Class 8b_ DoseRate-lO Class_8c_ DoseRate-io EPRI accident Class 1 dose rate given a 1-in-10 years ILRT interval EPRI accident Class_2_dose rate given a 1-in-10 years ILRT interval EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval EPRI accident Class_7_dose rate given a 1-in-10 years ILRT interval EPRI accident Class 8a dose rate given a 1-in-10 years ILRT interval EPRI accident Class_8b dose rate given a 1-in-10 years ILRT interval EPRI accident Class_8c dose rate given a 1-in-10 years ILRT interval ClasslDose Class_2_Dose

=

EPRI accident Class_1_dose

=

EPRI accident Class_2_dose

= 3.86E+03 person-rem (Table 3)

=

1.13E+05 person-rem (Table 3) 17670-0001 PB ILRT Rev 3.doc

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PAGE: 23 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class 3a Dose

=

EPRI accident Class 3a dose

=

3.86E+04 person-rem (Table 3)

Class 3b Dose

=

EPRI accident Class_3b dose

=

1.35E+05 person-rem (Table 3)

Class 7 Dose

=

EPRI accident Class 7 dose

=

1.39E+05 person-rem (Table 3)

Class_8a Dose

=

EPRI accident Class 8a dose

=

1.13E+06 person-rem (Table 3)

Class 8b Dose

=

EPRI accident Class 8b dose

=

1.88E+05 person-rem (Table 3)

Class_8cDose

=

EPRI accident Class_8c dose

=

1.39E+05 person-rem (Table 3)

For Unit 1, Class1 2Frequency,0 Class_2aFrequency 10 Class_3abFrequencylo Class 3b Frequencylo Class_7_Frequencylo Class 8a Frequencylo Class_8bFrequencylo Class_8c Frequencylo For Unit 2, Class 1 Frequency1 o Class 2 Frequencylo Class_3aFrequency1 o Class_3bFrequencylo Class 7 Frequency1 o Class_8aFrequency1 o Class_8b Frequency1 o 17670-0001 PB ILRT Rev 3.doc

=

Frequency of EPRI accident Class 1 given a 1-in-10 years ILRT interval

=

3.25E-05/ry (Table 5a)

=

Frequency of EPRI accident Class 2 given a 1-in-10 years ILRT interval

=

1.15E-08/ry (Table 5a)

=

Frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval

=

3.25E-06/ry (Table 5a)

=

Frequency of EPRI accident Class 3b given a 1-in-10 years ILRT interval 3.25E-07/ry (Table 5a)

Frequency of EPRI accident Class 7 given a 1-in-10 years ILRT interval

=

9.38E-06/ry (Table 5a)

=

Frequency of EPRI accident Class 8a given a 1-in-10 years ILRT interval

=

2.37E-07/ry (Table 5a)

=

Frequency of EPRI accident Class 8b given a 1-in-10 years ILRT interval

=

1.83E-06/ry (Table 5a)

=

Frequency of EPRI accident Class 8c given a 1-in-10 years ILRT interval 2.62E-06/ry (Table 5a)

= Frequency of EPRI accident Class 1 given a 1-in-10 years ILRT interval

= 2.81 E-05/ry (Table 5b)

= Frequency of EPRI accident Class 2 given a 1-in-1 0 years ILRT interval

= 1.19E-08/ry (Table 5b)

= Frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval

= 2.81E-06/ry (Table 5b)

= Frequency of EPRI accident Class 3b given a 1-in-10 years ILRT interval

= 2.81 E-07/ry (Table 5b)

= Frequency of EPRI accident Class 7 given a 1-in-10 years ILRT interval

= 1.58E-05/ry (Table 5b)

= Frequency of EPRI accident Class 8a given a 1-in-10 years ILRT interval

= 2.37E-07/ry (Table 5b)

= Frequency of EPRI accident Class 8b given a 1-in-10 years ILRT interval

= 1.91E-06/ry (Table 5b)

SCIENTECH.

PAGE: 24 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 161 P

161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class_8cFrequencylo Therefore, for Unit 1, Class 1 DoseRate_10 Class 2 DoseRate-10 Class 3a_DoseRate_10 Class 3b_DoseRat-10 Class 7 DoseRate-lo Class 8a_DoseRate-10 Class 8b_DoseRate-10 Class_8cDoseRate-10 Therefore, for Unit 2, Class_1_DoseRate40 Class 2 DoseRate-10 Class 3a_DoseRate-10 Class 3b_DoseRate-1O Class_8c_DoseRate-lO Class 8a_DoseRatei0 Class 8b DoseRate-10 Class_8c_DoseRate-10

= Frequency of EPRI accident Class 8c given a 1-in-10 years ILRT interval

= 2.66E-06/ry (Table 5b) 3.86E+03 1.13E+05 3.86E+04 1.35E+05 1.39E+05 1.13E+06 1.88E+05 1.39E+05 3.86E+03 1.13E+05 3.86E+04 1.35E+05 1.39E+05 1.13E+06 1.88E+05 1.39E+05

k

k 3.25E-05 1.15E-08 3.25E-06 3.25E-07 9.38E-06 2.37E-07 1.83E-06 2.62E-06 2.81 E-05 1.19E-08 2.81 E-06 2.81 E-07 1.58E-05 2.37E-07 1.91 E-06 2.66E-06 1.26E-01 1.30E-03 1.25E-01 4.39E-02 1.30E+00 2.68E-01 3.44E-01 3.64E-01 1.08E-01 1.35E-03 1.08E-01 3.79E-02 2.20E+00 2.68E-01 3.59E-01 3.69E-01 (person-rem/ry)

(person-rem/ry)

k

k The calculation for a 1-in-1 5-1/2 years ILRT interval is as follows:

Class 1 DoseRate-15 Class 2 DoseRatel15 Class 3a_ DoseRate15 Class 3b_ DoseRate-15 Class 7 DoseRate-15 Class 8a_ DoseRate-15 Class 8b_ DoseRate_15 Class_8c_ DoseRate-15 Class_1_Dose Class_2_Dose Class_3aDose Class_3b-Dose Class_7_Dose Class_8aDose Class_8bDose Class_8cDose Class_1_Frequency 15 Class_2_Frequency1 5 Class_3a_Frequency1 5 Class' 3bFrequency1 5 Class 7_Frequency1 5 Class_8aFrequency 15 Class_8bbFrequency1 5 Class_8cFrequency1 5 Where Class 1 DoseRat.15 Class 2 DoseRate-15 Class 3a_ DoseRate15 Class 3b_ DoseRate-15 Class 7 DoseRate-15 Class 8a_ DoseRate-15 Class 8b_ DoseRat,15 ClassBc DoseRate-15 EPRI accident Class 1 dose rate given a 1-in-15-1/2 years ILRT interval EPRI accident Class_2_dose rate given a 1-in-15-1/2 years ILRT interval EPRI accident Class 3a dose rate given a 1-in-1 5-1/2 years ILRT interval EPRI accident Class 3b dose rate given a 1-in-1 5-1/2 years ILRT interval EPRI accident Class 7_dose rate given a 1-in-15-1/2 years ILRT interval EPRI accident Class 8a dose rate given a 1-in-1 5-1/2 years ILRT interval EPRI accident Class 8b dose rate given a 1-in-1 5-1/2 years ILRT interval EPRI accident Class_8c dose rate given a 1-in-1 5-1/2 years ILRT interval Class_1_Dose Class 2 Dose Class 3a Dose Class 3b Dose Class_7_Dose EPRI accident ClassIdose EPRI accident Class 2 dose EPRI accident Class 3a dose EPRI accident Class_3b dose EPRI accident Class_7_dose 3.86E+03 person-rem (Table 3) 1.13E+05 person-rem (Table 3) 3.86E+04 person-rem (Table 3) 1.35E+05 person-rem (Table 3) 1.39E+05 person-rem (Table 3) 17670-0001 PB ILRT Rev 3.doc

kSCIENTECH.

PAGE: 25 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 16 5

161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class 8a Dose

=

EPRI accident Class 8a dose

=

1.13E+06 person-rem (Table 3)

Class 8b Dose

=.EPRI accident Class_8b dose

=

1.88E+05 person-rem (Table 3)

Class_8cDose

=

EPRI accident Class_8c dose

=

1.39E+05 person-rem (Table 3)

For Unit 1, Class1 2Frequency,5 Class_23Frequency 15 Class_3aFrequency1 5 Class 3b Frequency1 5 Class 7 Frequencyls Class_8aFrequency1 s Class_8bFrequency,5 Class_8cFrequency1 5 For Unit 2, Class 1 Frequencyl 5 Class 2 Frequency1 5 Class_3aFrequency1 5 Class_3bFrequency1 5 Class 7 Frequency 15 Class_8aFrequency 15 Class_8bFrequency,5 Class_8cFrequency15 17670-0001 PB ILRT Rev 3.doc

=

Frequency of EPRI accident Class 1 given a 1-in-1 5-1/2 years ILRT interval

=

3.06E- 05/ry (Table 6a)

=

Frequency of EPRI accident Class 2 given a 1-in-1 5-1/2 years ILRT interval

=

1.15E-08/ry (Table 6a)

=

Frequency of EPRI accident Class 3a given a I-in-15-1/2 years ILRT interval

=

5.04E-06/ry (Table 6a)

=

Frequency of EPRI accident Class 3b given a 1-in-1 5-1/2 years ILRT -interval

=

5.04E-07/ry (Table 6a)

=

Frequency of EPRI accident Class 7 given a 1-in-1 5-1/2 years ILRT interval

=

9.38E-06/ry (Table 6a)

=

Frequency of EPRI accident Class 8a given a 1-in-1 5-1/2 years ILRT interval

=

2.37E-07/ry (Table 6a)

=

Frequency of EPRI accident Class 8b given a 1-in-1 5-1/2 years ILRT interval

=

1.83E-06/ry (Table 6a)

=

Frequency of EPRI accident Class 8c given a 1-in-1 5-1/2 years ILRT interval

=

2.62E-06/ry (Table 6a)

= Frequency of EPRI accident Class 1 given a 1-in-1 5-1/2 years ILRT interval

= 2.64E-05/ry (Table 6b)

= Frequency of EPRI accident Class 2 given a 1-in-1 5-1/2 years ILRT interval

= 1.19E-08/ry (Table 6b)

= Frequency of EPRI accident Class 3a given a 1-in-1 5-1/2 years ILRT interval

= 4.35E-06/ry (Table 6b)

= Frequency of EPRI accident Class 3b given a 1-in-1 5-1/2 years ILRT interval

= 4.35E-07/ry (Table 6b)

= Frequency of EPRI accident Class 7 given a 1-in-1 5-1/2 years ILRT interval

= 1.58E-05/ry (Table 6b)

= Frequency of EPRI accident Class 8a given a 1-in-1 5-1/2 years ILRT interval

= 2.37E-07/ry (Table 6b)

= Frequency of EPRI accident Class 8b given a 1-in-1 5-1/2 years ILRT interval

= 1.91 E-06/ry (Table 6b)

= Frequency of EPRI accident Class 8c given a 1-in-1 5-1/2 years ILRT interval

= 2.66E-06/ry (Table 6b)

S CIENTECH PAGE: 26 OF

[

OSCIENTECH, CLIENT: Nuclear Management Company BY: E. A. Krantz 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval '1 Therefore, for Unit 1, Class 1 DoseRate15 Class 2 DoseRate-15 Class_3a_DoseRate-15 Class 3b DoseRate.I5 Class 7 DoseRate-15 Class_8a_DoseRate-15 Class 8b_DoseRate-15 Class_8cDoseRate-.15 Therefore, for Unit 2, ClassIDoseRate-15 Class_2_DoseRate-15 Class_3a_DoseRate.15 Class 3b_DoseRate-15 Class_7_DoseRate-15 Class_8a_DoseRate_15 Class_8b_DoseRate-15 Class_8cDoseRate-15 3.86E+03 1.13E+05 3.86E+04 1.35E+05 1.39E+05 1.13E+06 1.88E+05 1.39E+05 3.86E+03 1.13E+05 3.86E+04 1.35E+05 1.39E+05 1.13E+06 1.88E+05 1.39E+05 3.06E-05 1.15E-08 5.04E-06 5.04E-07 9.38E-06 2.37E-07 1.83E-06 2.62E-06 2.64E-05 1.19E-08 4.35E-06 4.35E-07 1.58E-05 2.37E-07 1.91 E-06 2.66E-06 1.18E-01 1.30E-03 1.94E-01 6.80E-02 1.30E+00 2.68E-01 3.44E-01 3.64E-01 1.02E-01 1.35E-03 1.68E-01 5.88E-02 2.20E+00 2.68E-01 3.59E-01 3.69E-01 (person-rem/ry)

(person-rem/ry)

Tables 5a and 5b summarize the resulting population dose rates by EPRI accident classes for a 10 year test interval. Tables 6a and 6b summarize the resulting population dose rates by EPRI accident classes for a 15-1/2 year test interval 11.1.7 Step 7 - Change in Population Dose Rate Due to New ILRT Interval This step, per the EPRI Interim Guidance (Reference 4) calculates the percentage of the total dose rate attributable to EPRI accident Classes 3a and 3b (those accident classes affected by change in ILRT surveillance interval) and the change in this resulting dose rate from the base dose rate attributable to changes in ILRT surveillance interval.

Based on the results summarized in Tables 5a and 5b, for the current PB 1-in-10 years ILRT interval, the percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b is calculated as follows:

PrctTD10 = percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b given a 1-in-10 years ILRT interval, which is calculated using the following equation:

PrctTD 1 o = [(Class_3a_DoseRate-10 + Class_3b DoseRate-lo)/Total_DoseRate-lo]

100%

Class 3a_DoseRate-10 = EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= 1.25E-01 [Table 5a] for Unit 1 Class 3b_DoseRate-lo = EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= 4.39E-02 [Table 5a]

Total DoseRate_1o = Total dose rate for all EPRI's classes given a 1-in-10 years ILRT interval

= 2.58 [Table 5a]

Therefore, 17670-0001 PB ILRT Rev 3.doc

,SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 27 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval PrctTDo10 = [(1.25E-01 + 4.39E-02)/ 2.581

100% = 6.57%

For Unit 2, Class 3a_DoseRate-10 = EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= 1.08E-01 [Table 5b]

Class 3b_DoseRate-lo = EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= 3.79E-02 [Table 5b]

Total DoseRatelo = Total dose rate for all EPRI's classes given a 1-in-10 years ILRT interval

= 3.45 [Table 5b]

Therefore, PrctTDo10 = [(1.08E-01+ 3.79E-02)/ 3.45]

100% = 4.24%

The percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b based on the proposed 1-in-1 5-1/2 years ILRT interval is calculated as follows:

Prct TD15 = percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b given a 1-in-1 5-1/2 years ILRT interval, which is calculated using the following equation:

PrctTD15 = [(Class_3a_DoseRate-15 + Class_3b DoseRate.15)/TotalDoseRate15]

100%

For Unit 1:

Class 3a_DoseRate-lS = EPRI accident Class 3a dose rate given a 1-in-1 5-1/2 years ILRT interval 1.94E-01 [Table 6a]

Class_3b_DoseRaate-5 = EPRI accident Class 3b dose rate given a 1-in-1 5-1/2 years ILRT interval

= 6.80E-02 [Table 6a]

TotalDoseRatel5 = Total dose rate for all EPRI's classes given a 1-in-1 5-1/2 years ILRT interval

= 2.66 [Table 6a]

Therefore, PrctTD15 = [(1.94E-01 + 6.80E-02)/ 2.66]

100% = 9.86%

For Unit 2, Class 3aDoseRate1s = EPRI accident Class 3a dose rate given a 1-in-1 5-1/2 years ILRT interval

= 1.68E-01 [Table 6b]

Class 3bDoseRate-15 = EPRI accident Class 3b dose rate given a 1-in-15-1/2 years ILRT interval

= 5.88E-02 [Table 6b]

TotalDoseRates5 = Total dose rate for all EPRI's classes given a 1-in-1 5-1/2 years ILRT interval 3.52 [Table 6b]

Therefore, Prct.TD15 = [(1.68E-01 + 5.88E-02)/ 3.52]

100% = 6.44%

17670-0001 PB ILRT Rev 3.doc

,)SCIENTECH.

PAGE: 28 OF CLIENT: Nuclear Management Company BY: E. A. Krantz P61 y

161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Based on the above results, the changes from the 1-in-10 years to 1-in-1 5-1/2 years dose rate is as follows for Unit 1:

Increase10. 15 = [(TotalDoseRate Total_DoseRate.lo)/ TotalDoseRatelo ] *100%

Where:

Increase10. 15 = percent change from 1-in-10 years ILRT interval to 1-in-15-1/2 years ILRT interval Total DoseRate_1s = Total dose rate for all EPRI's classes given a 1 -in-1 5-1/2 years ILRT interval

= 2.66 (person-rem/ry) [Table 6a]

Total DoseRate-o = Total dose rate for all EPRI's classes given a 1-in-10 years ILRT interval

= 2.58 (person-rem/ry) [Table 5a]

Therefore, Increase10.15 = [(2.66-2.58)/ 2.58 1 *100% = 3.32%

For Unit 2, Total DoseRatels = Total dose rate for all EPRI's classes given a 1-in-1 5-1/2 years ILRT interval

= 3.52 (person-rem/ry) [Table 6b]

TotalDoseRate-lo = Total dose rate for all EPRI's classes given a 1-in-10 years ILRT interval

= 3.45 (person-rem/ry) [Table 5b]

Therefore, Increase1 0. 15 = [(3.52-3.45)/ 3.45 ] *100% = 2.14%

11.1.8 Step 8 - Change in LERF Due to New ILRT Interval This step, per EPRI Interim Guidance (Reference 4) evaluates the increase in the Large Early Release Frequency (LERF) due to extending the ILRT test interval from that corresponding to 3 tests in 10 years to that corresponding to 1 test in 15-1/2 years and from a 10 year interval to a 15-1/2 year interval.

The risk impact 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 containment could in fact result in large release due to failure to detect a pre-existing leak during the relaxation period. For this evaluation only Class 3 sequences have the potential to result in large releases if pre-existing leaks were present. Class 1 sequences are not considered as potential large release pathways because for these sequences the containment remains intact. Therefore, the containment leak rate is expected to be small (less than 2 La). A larger leak rate would imply an impaired containment, such as classes 2, 3, 6 and 7.

Late releases are excluded regardless of the size of the leak because late releases are, by definition, not a LERF event. At the same time, sequences in the PB PSA (Reference 13), which result in large releases (e.g., large isolation valve failures), are not impacted because a LERF will occur regardless of the presence of a pre-existing leak. Therefore, the frequency of accident Class 3b sequences (Tables 5a/b and 6a/b) is used as the LERF for Point Beach.

The affect on the LERF risk measure due to the proposed ILRT interval extension is calculated as follows for Unit 1:

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 29 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TheLERF frequencies of interest are:

Class_3bFrequency = 9.75E-08 [Table 4a]

Class_3bFrequencylo = 3.25E-07 [Table 5a]

Class_3b_Frequency 15 = 5.04E-07 {Table 6a]

Therefore, ALERF1o15S = the change in LERF from 1-in-10 years ILRT interval to 1-in-1 5-1/2 years ILRT interval

= Class 3bFrequency1 5 - Class 3b Frequency1 o

= 5.04E-07/ry - 3.25E-07/ry

= 1.79E-07/ry It should be noted that if the risk increase is measured from the original 3-in-1 Oyears ILRT interval, the increase in LERF is as follows:

ALERF 3_15 the change in LERF from 3-in-10 years ILRT interval to 1-in-15-1/2 years ILRT interval

= Class_3bFrequency1 5 - Class 3b Frequency

= 5.04E-07/ry - 9.75E-08/ry

= 4.06E-07/ry For Unit 2:

The LERF frequencies of interest are:

Class_3bFrequency = 8.42E-08 [Table 4b]

Class_3bFrequencyo.= 2.81E-07 [Table 5b]

Class_3bFrequency 15 = 4.35E707{Table 6b]

Therefore, ALERFO.*15 = the change in LERF from 1-in-10 years ILRT interval to 1-in-15-1/2 years ILRT interval

= Class_3bFrequency 1 5 - Class 3b Frequencylo

= 4.35E-07/ry - 2.81 E-07/ry

= 1.54E-07/ry It should be noted that if the risk increase is measured from the original 3-in-10years ILRT interval, the increase in LERF is as follows:

ALERF 3_15 = the change in LERF from 3-in-10 years ILRT interval to 1-in-1 5-1/2 years ILRT interval

= Class 3bFrequency,5 - Class 3b Frequency

= 4.35E-07/ry - 8.42E-08/ry

= 3.51 E-07/ry Regulatory Guide 1.174 (Reference 9) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 (Reference 9) defines small changes in risk as resulting in increases of core damage frequency (CDF) greater than 1 E-06 but below 1 E-17670-0001 PB ILRT Rev 3.doc

(.3SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 30 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 05/ry and increases in LERF greater that I E-07 but below 1 E-06/ry. Since the ILRT does not impact CDF, the relevant risk metric is LERF.

The internal events contribution to LERF at Point Beach is 2.11 E-06 for Unit 1 and 2.19E-06 for Unit 2 (Section 11.1.1); these values of LERF allow some increases in risk per Figure 4 of Regulatory Guide 1.174. The change in LERF associated with increasing the ILRT interval at Point Beach from the currently allowed 1-in-10 years to 1-in-15-1/2 years is 1.79E-07 for Unit 1 and 1.54E-07 for Unit 2.

Because Reference 9 defines small changes in LERF as below 1 E-06/ry, increasing the ILRT interval at Point Beach from the currently allowed 1-in-10 years to 1-in-I 5-1/2 years represents a small change in plant risk from the LERF perspective. Similarly, the change in realistic values of LERF for moving from 3-in-10 years ILRT interval to 1-in-15-1/2 years of 4.06E-07 for Unit I and 3.51E-07/ry for Unit 2 falls into Region II, Small Change in Risk, of the acceptance guidelines in NRC Regulatory Guide 1.174.

11.1.9 Step 9 - Impact on Conditional Containment Failure Probability (CCFP)

This step, per the EPRI Interim Guidance (Reference 4) calculates the change in conditional containment failure probability (CCFP). The CCFP risk metric ensures and shows that the proposed change in ILRT interval is consistent with the defense-in-depth philosophy described in Regulatory Guide 1.174 (Reference 9)4.

In this calculation, the change in CCFP relates to the impact of the ILRT on both early (LERF) and late radionuclide releases. Based on the EPRI Interim Guidance (Reference 4), CCFP consists of all those accident sequences resulting in a radionuclide release other than the intact containment state for EPRI accident Class 1, and small failures state for EPRI accident Class 3a. In addition, the CCFP is conditional given a severe core damage accident. The CCFP is calculated by the following equation:

CCFP = 1-[IntactContainmentFrequency/TotalCDF]

OR CCFP

[1 - (Class 1 Frequency + Class_3a Frequency)/ CDFTotal]

100%

Where; Class 1 Frequency = Frequency per year of EPRI accident Class 1.

Class_3a_ Frequency = Frequency per year of EPRI accident Class 3a.

CDFTotal = PB Core Damage Frequency For the 1-in-10 years ILRT interval, CCFP 10 = [1 - (Class 1 Frequency10 + Class_3a Frequencylo)/ CDFTotaI j

100%

Where; CCFP1o = Conditional containment failure probability given 1-in-10 years ILRT interval 4 The defense-in-depth philosophy is maintained as a reasonable balance among prevention of core damage, containment failure and consequence mitigation.

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 31 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class_ 1 Frequencylo = Frequency per year of EPRI accident Class 1 given 1-in-10 years ILRT interval Class 3a_ Frequencylo = Frequency per year of EPRI accident Class 3a given 1-in-10 years ILRT interval For the 1-in-1 5-1/2 years ILRT interval, CCFP 15 = [1 - (Classl_ Frequency1 5 + Class 3a Frequency 15)/ CDFTotal]

100%

Where; CCFP15 = Conditional containment failure probability given 1-in-1 5-1/2 years ILRT interval Class 1 Frequency15 = Frequency per year of EPRI accident Class 1 given Iqin-15-1/2 years ILRT interval Class_3a_ Frequency 15 = Frequency per year of EPRI accident Class 3a given 1-in-1 5-1/2 years ILRT interval For Unit 1, the frequencies of interest are:

CDFTotal = 5.02E-05 [Table 4a]

Class 1 Frequency = Frequency per year of EPRI accident Class 1.

= 3.50E-05/ry [Table 4a]

Class 3a_ Frequency = Frequency per year of EPRI accident Class 3a.

= 9.75E-07/ry [Table 4a]

Class 1 Frequencyl0 = Frequency per year of EPRI accident Class 1 given 1-in-10 years ILRT interval

= 3.25E-05/ry [Table 5a]

Class 3a_ Frequency1 o = Frequency per year of EPRI accident Class 3a given 1-in-10 years ILRT interval

= 3.25E-06/ry [Table 5a]

Class 1_ Frequency, 5 = Frequency per year of EPRI accident Class 1 given 1-in-1 5-1/2 years ILRT interval

= 3.06E-05/ry [Table 6a]

Class 3a_ Frequency, 5 = Frequency per year of EPRI accident Class 3a given 1-in-1 5-1/2 years ILRT interval

= 5.04E-06/ry[Table 6a]

Therefore, CCFP = [1 - (Class 1 Frequency + Class_3a Frequency)/ CDFrot*]

100%

= [1-(3.50E-05 + 9.75E-07)/5.02E-05]

100%

= 28.25%

CCFP1 o = [1 - (Class 1 Frequencylo + Class 3a Frequency10 )/ CDFTotai]

100%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

PAGE: 32 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 161 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval

= [1 - (3.25E-05 + 3.25E-06)/ 5.02E-05]

100%

= 28.70%

CCFP15 = [1 - (Class 1 Frequency 15 + Class 3a Frequency 15)/ CDFTotal]

100%

= [1 - (3.06E-05 + 5.04E-06)/ 5.02E-05]

100%

= 29.06%

For Unit 2, the frequencies of interest are:

CDFTotaI = 5.18E-05 Fable 4b]

Class 1 Frequency = Frequency per year of EPRI accident Class 1.

= 3.03E-05/ry [Table 4b]

Class 3a_ Frequency = Frequency per year of EPRI accident Class 3a.

= 8.42E-07/ry [Table 4b]

Class 1 Frequencylo = Frequency per year of EPRI accident Class 1 given 1-in-10 years ILRT interval

= 2.81 E-05/ry [Table 5b]

Class 3a_ Frequency1 o = Frequency per year of EPRI accident Class 3a given 1-in-10 years ILRT interval

= 2.81 E-06/ry [Table 5b]

Class 1 Frequency15 = Frequency per year of EPRI accident Class 1 given 1-in-1 5-1/2 years ILRT interval

= 2.64E-05/ry [Table 6b]

Class 3a_ Frequency 15 Frequency per year of EPRI accident Class 3a given 1-in-1 5-1/2 years ILRT interval

= 4.35E-06/ry [Table 6b]

Therefore, CCFP = [1 - (Class_ _ Frequency + Class_3a_ Frequency)/ CDFTotal]

100%

= [1-(3.03E-05 + 8.42E-07)/5.18E-05]

100%

= 39.96%

CCFP10 = [1 - (Classl_ Frequencylo + Class 3a Frequencylo)/ CDFTotal]

100%

= [1 - (2.81 E-05 + 2.81E-06)/ 5.18E-05 ]

100%

40.34%

CCFP 15 = [1 - (Class 1 Frequency15 + Class_3a Frequencys)/ CDFTota 2 J

100%

= [1 - (2.64E-05 + 4.35E-06)/ 5.18E-05 ]

100%

= 40.63%

The change in CCFP due to the ILRT interval going from that corresponding to 3 tests in 10 years to that corresponding to 1 test in 10 years is as follow:

ACCFP3 o 10 = CCFP10 - CCFP 17670-0001 PB ILRT Rev 3.doc

,SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 33 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval For Unit 1:

ACCFP 3.10 = 28.70% - 28.25% = 0.45%

For Unit 2:

ACCFP 3.10 = 40.34% - 39.96% = 0.38%

The change in CCFP due to the ILRT interval going from that corresponding to 1 test in 10 years to that corresponding to 1 test in 15-1/2 years is as follow:

ACCFP1IO15 = CCFP15 - CCFP1 o For Unit 1:

ACCFP 10.15 = 29.06%- 28.70% = 0.36%

For Unit 2:

ACCFP10o15 = 40.63% - 40.34% = 0.29%

The change in CCFP due to the ILRT interval going from that corresponding to 3 tests in 10 years to that corresponding to 1 test in 15-1/2 years is as follow:

ACCFP 3_15 = CCFPI5 - CCFP For Unit 1:

ACCFP3SI 5 = 29.06% - 28.25% = 0.81%

For Unit 2 ACCFP3a15 = 40.63% - 39.96% = 0.67%

These changes of approximately 1% or less are not significant.

17670-0001 PB ILRT Rev 3.doc

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PAGE: 34 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 161 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 11.2 External Events and Internal Flooding Impacts External events (fire, seismic, other external) and internal flooding for the Point Beach plant have not been included in the PSA. Per Reference 15 the core damage frequency contribution by these events are, however, estimated to be 3.6E-05 per reactor year. This section summarizes the impact on this ILRT risk assessment of including this contribution to the overall core damage frequency.

The purpose of the external events evaluation is to determine whether there are any unique insights or important quantitative information that explicitly impact the risk assessment results when considering only internal events.

The quantitative consideration of external hazards is discussed in more detail in Appendix A of this report. The combined internal/external events contribution to LERF at Point Beach is 3.64E-06 (2.11 E-06 + 1.53E-06) for Unit 1 and 3.72E-06 (2.19E-06 + 1.53E-06) for Unit 2 (Section 11.1.1/Table A-4);

these values of LERF permit changes to be made that result in increases in LERF per Figure 4 of Regulatory Guide 1.174. The change in the combined internal events/external events LERF associated with increasing the ILRT interval at Point Beach from the currently allowed 1-in-10 years to 1-in-1 5-1/2 years is 3.08E-07 for Unit 1 and 2.62E-07 for Unit 2. Because Reference 9 defines small changes in LERF as below 1 E-06/ry, increasing the ILRT interval at Point Beach from the currently allowed 1-in-10 years to 1-in-15-1/2 years represents a small change in plant risk. Similarly, the change in realistic values of LERF for moving from 3-in-10 years ILRT interval to 1-in-15-1/2 years of 7.OOE-07 for Unit 1 and 5.97E-07/ry for Unit 2 falls into Region II, Small Change in Risk, of the acceptance guidelines in NRC Regulatory Guide 1.174.

Other salient results from Appendix A, found the combined internal and external events increase in risk for those accident sequences influenced by Type A testing, compared with the total integrated plant risk, given the change from a 1-in-10 years test interval to a 1-in-1 5-1/2 years test interval, to be 4.4% or 0.2 person-rem/ry for Unit 1 (2.1% or 0.1 person-rem/ry for Unit 2). In addition, the change in the combined internal and external events conditional containment failure probability from 1-in-10 years to 1-in-1 5-1/2 years is 0.36% for Unit 1 and 0.30% for Unit 2. A change in CCFP of less than 1%

is not significant from a risk perspective.

Therefore, incorporating external event accident sequence results into this analysis does not change the conclusion of an internal-events only risk assessment (i.e., increasing the PB ILRT interval from 10 to 15-1/2 years is an acceptable plant change from a risk perspective). This result is expected, because the proposed ILRT interval extension impacts plant risk in a very specific and limited way.

11.3 Containment Liner Corrosion Risk Impact Recently, the NRC issued a series of Requests for Additional Information (RAIs) in response to the onetime relief requests for the ILRT surveillance interval submitted by various licensees. One of the PAls related to the risk assessment performed in this report is provided below.

Request for Additional Information:

Inspections of reinforced and steel containments at some facilities (e.g., North Anna, Brunswick D.C. Cook, and Oyster Creek) have indicated degradation from the uninspectable (embedded) side of the steel shell and liner of primary containments. The major uninspectable areas of the Mark I containment are the vertical portion of the drywell shell and part of the shell sandwiched between the drywell floor and the basemat. Please discuss what programs are used to monitor their conditions. Also, address how potential leakage due to age-related degradation from these uninspectable areas are factored into the risk assessment in support of the requested interval extension.

17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 35 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval The impact of the risk assessment portion of the above RAls is summarized in this section (refer to Appendix B for further details).

The containment liner corrosion analysis utilizes the referenced Calvert Cliffs Nuclear Power Plant assessment (Reference 17) to estimate the likelihood and risk-implication of degradation-induced leakage occurring and going undetected in visual examinations during the extended test interval. It should be noted that the Calvert Cliffs analysis was performed for a concrete cylinder and dome containment with a steel liner and the PB containment is similar.

Consistent with the Calvert Cliffs analysis, the following issues are addressed:

Differences between the containment basemat and the cylinder/dome liner The historical cylinder/dome steel shell 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 Consistent with Calvert Cliffs analysis (Reference 17), the following six steps are performed:

1) Determine the historical liner flaw likelihood.

2) Determine aged adjusted liner flaw likelihood.

3) Determine the increase in flaw likelihood between 3, 10 and 15 years.

4) Determine the likelihood of containment breach given liner flaw.

5) Determine the visual inspection detection failure.

6) Determine the likelihood of non-detected containment leakage:

In addition to these steps, the following three steps are added to evaluate risk-implication of containment liner corrosion:

7) Evaluate the risk impact in terms of population dose rate and percentile change for the interval cases.

8) Evaluate the risk impact in terms of LERF.

9) Evaluate the change in conditional containment failure probability.

The quantitative consideration of the containment liner corrosion analysis is discussed in more detail in Appendix B of this report. As can be seen from Appendix B, including corrosion effects in the ILRT assessment would not alter the conclusions from the original internal events analysis. That is, the change in LERF from extending the interval to 15-1/2 years from the current 10-year requirement due to consideration of corrosion is estimated to be 7.75E-09 for Unit 1 (6.69E-09/ry for Unit 2). This value is below the NRC Regulatory Guide 1.174 limit of 1 E-07/ry. Therefore, because Regulatory Guide 1.174 defines very small changes in LERF as below I E-07/ry, increasing the ILRT interval at PBNP from the currently allowed 1-in-10 years to 1-in-1 5-1/2 years and taking into consideration the likelihood of a containment liner flaw due to corrosion represents only a small increase in risk.

Additionally, the dose increase is estimated to be 1.05E-03 person-rem/ry for Unit 1 (9.04E-04 person-rem/ry for Unit 2), and the conditional containment failure probability increase is estimated to be less than 0.02% at each unit. Both of these increases are also considered to be small. As a result, the ILRT interval extension is considered to have a minimal impact on plant risk (including age-adjusted corrosion impacts), and is therefore acceptable.

In addition, a series of parametric sensitivity studies (discussed in more detail in Appendix B of this report) regarding the potential age related corrosion effects on the containment steel liner also predict that even with conservative assumptions, the conclusions from the original internal events analysis would not change.

17670-0001 PB ILRT Rev 3.doc

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PAGE: 36 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 161 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 12.0 COMPUTER INPUT AND OUTPUT NONE 13.0

SUMMARY

OF RESULTS The effects of internal hazard risk on ILRT risk are shown in Table 7a/b. The combined internal and external events effect on the ILRT risk is shown in Table A-9. This table combines the results of Table 4a/b with the results depicted in Table A-8a/b.

Appendix B provides an assessment of the sensitivity of the above results to age-related corrosion of the containment shell. The above major results are. repeated below along with the results if the impact of age-related corrosion is included.

14.0 CONCLUSION

S:

The conclusions regarding the change in plant risk associated with extension of the Type A ILRT test frequency from one test in ten years to one test in fifteen and a half years, based on the results in Section 13 and Appendix A, are as follows:

The combined internal and external events increase in risk for those accident sequences influenced by Type A testing, compared with the total integrated plant risk, given the change from a 1-in-10 years test interval to a 1-in-15-1/2 years test interval, is found to be 4.4% for Unit 1 and 2.1% for Unit 2.

Given the low total risk to the public, these values are not significant increases in risk.

The combined internal/external events contribution to LERF at Point Beach is 3.64E-06 (2.11 E-06 +

1.53E-06) for Unit 1 and 3.72E-06 (2.19E-06 + 1.53E-06) for Unit 2 (Section 11.1.1/Table A-4); these values of LERF permit changes to be made that result in increases in LERF per Figure 4 of Regulatory Guide 1.174. The change in the combined internal events/external events LERF associated with increasing the ILRT interval at Point Beach from the currently allowed 1-in-10 years to 1-in-1 5-1/2 years is 3.08E-07 for Unit 1 and 2.62E-07 for Unit 2. According to Reference 9, increasing the ILRT interval at Point Beach from the currently allowed 1-in-10 years to 1-in-15-1/2 years represents a small change in plant risk. Similarly, the change in LERF for moving from 3-in-10 years ILRT interval to 1-in-15-1/2 years of 7.OOE-07 for Unit 1 and 5.97E-07/ry for Unit 2 falls into Region II, Small Change in Risk, of the acceptance guidelines in NRC Regulatory Guide 1.174.

The change in the combined internal and external events conditional containment failure probability from 1-in-10 years to 1-in-1 5-1/2 years is 0.36% and 0.30% respectively for Units 1 and 2. A change in CCFP of less than 1% is insignificant from a risk perspective.

The impact of age-related corrosion of the steel containment has a negligible or very small impact on each of the risk measures associated with the extension of the Type A ILRT test frequency. The above conclusions remain valid even including consideration of corrosion.

17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company

_BY:

E. A. Krantz PAGE: 37 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table 1-Detailed Description for the Eight Accident Classes as defined by EPRI TR-104285 Class Detailed Description 1

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. The allowable leakage rates (La), are typically 0.1 weight percent of containment volume per day for PWRs.(all measured at Pa, calculated peak containment pressure related to the design basis accident). Changes to leak rate testing frequencies do not affect this classification.

2 Containment isolation failures (as reported in the IPEs) include those accidents in which the pre-existing leakage is due to failure to isolate the containment. These include those that are dependent on the core damage accident in progress (e. g.,

initiated by common cause failure or support system failure of power) and random failures to close a containmernt path.

Changes in Appendix J testing requirements do not impact these accidents.

3 Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal (i. e.,

provide a leak-tight containment) is not dependent on the sequence in progress. This accident class is applicable to sequences involving ILRTs (Type A tests) and potential failures not detectable by LLRTs.

4 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.

5 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 4 isolation failures, but is applicable to sequences involving Type C tests and their potential failures.

6 Containment isolation failures include those leak paths not identified by the LLRTs. The type of penetration failures considered under this class includes those covered in the plant test and maintenance requirement or verified by in service inspection and testing (ISI/IST) program. This failure to isolate is not typically identified in LLRT. Changes in Appendix J LLRT test intervals do not impact this class of accidents.

7 Accidents involving containment failure induced by severe accident phenomena. Changes in Appendix J testing requirements do not impact these accidents.

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

17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 38 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE 2 - Containment Frequency Measures for a Given Accident Class Class Description Unit 1 Freq. -

% of Unit Unit 2 Freq. -

% of Unit 2 per yr.

I CDF per yr.

CDF 1

No Containment Failure 3.50E-05 69.8%

3.03E-05 58.4%

2 Large Containment Isolation Failure (Failure-To-Close) 1.50E-08 0.0%

1.55E-08 0.0%

3a Small Isolation Failures (Liner Breach) 9.75E-07 1.9%

8.42E-07 1.6%

3b Large Isolation Failures (Liner Breach) 9.75E-08 0.2%

8.42E-08 0.2%

4 Small Isolation Failure - Failure-To-Seal (Type B test) 0.00E+00 0.0%

0.OOE+00 0.0%

5 Small Isolation Failure - Failure-To-Seal (Type C Test) 0.OOE+00 0.0%

0.OOE+00 0.0%

6 Containment isolation Failures (Dependent failures, Personnel Errors) 0.OOE+00 0.0%

0.OOE+00 0.0%

7 Severe Accident Phenomena Induced Failure 9.38E-06 18.7%

1.58E-05 30.5%

8a Containment Bypassed (ISLOCA) 2.37E-07 0.5%

2.37E-07 0.5%

8b Containment Bypassed (Early SGTR) 1.83E-06 3.6%

1.91 E-06 3.7%

8c Containment Bypassed (Late SGTR) 2.62E-06 5.2%

2.66E-06 5.1%

All Containment Event Tree (CET) Endstates 5.02E-05 100.0%

5.18E-05 100.0%

17670-0001 PB ILRT Rev 3.doe

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CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 39 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE 3 - Conditional Person-Rem Measures for a Given Accident Class Class Description Person-Rem C__ass_

(50-miles) 1 No Containment Failure 3.86E+03 2

Large Containment Isolation Failure (Failure-To-Close) 1.13E+05 3a Small Isolation Failures (Liner Breach) 3.86E+04 3b Large Isolation Failures (Liner Breach) 1.35E+05 4

Small Isolation Failure - Failure-To-Seal (Type B test) 5 Small Isolation Failure - Failure-To-Seal (Type C Test) 6 Containment isolation Failures (Dependent failures, Personnel Errors) 7 Severe Accident Phenomena Induced Failure 1.39E+05 8a Containment Bypassed (ISLOCA) 1.13E+06 8b Containment Bypassed (Early SGTR) 1.88E+05 8c Containment Bypassed (Late SGTR) 1.39E+05 17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 40 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE 4a - Unit 1 Baseline Mean Consequence Measures for a Given Accident Class Unit I CasDsrponPerson-Rem Unit 1~ Frequency Person-Rem/yr Class Description (50-miles)

- per yr.

(50-miles) 1 No Containment Failure 3.86E+03 3.50E-05 1.35E-01 2

Large Containment Isolation Failure (Failure-To-Close) 1.13E+05 1.50E-08 1.70E-03 3a Small Isolation Failures (Liner Breach) 3.86E+04 9.75E-07 3.76E-02 3b Large Isolation Failures (Liner Breach) 1.35E+05 9.75E-08 1.32E-02 4

Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+00 O.OOE+00 5

Small Isolation Failure - Failure-To-Seal (Type C Test)

O.OOE+00 O.OOE+00 6

Containment isolation Failures (Dependent failures, O.OOE+00 Personnel Errors)

O.OOE+00 7

Severe Accident Phenomena Induced Failure 1.39E+05 9.38E-06 1.30E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 2.37E-07 2.68E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.83E-06 3.44E-01 8c Containment Bypassed (Late SGTR) 1.39E+05 2.62E-06 3.64E-01 Core All Containment Event Tree (CET) Endstates Damage 5.02E-05 2.47E+00 17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company BY: E. A. Krantz I PAGE: 41 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE 4b - Unit 2 Baseline Mean Conse uence Measures for a Given Accident Class Unit 2 Class Description Person-Rem Unit 2 Frequency Person-Rem/yr (50-miles)

- per yr.

(50-miles) 1 No Containment Failure 3.86E+03 3.03E-05 1.17E-01 2

Large Containment Isolation Failure (Failure-To-Close) 1.13E+05 1.55E-08 1.76E-03 3a Small Isolation Failures (Liner Breach) 3.86E+04 8.42E-07 3.25E-02 3b Large Isolation Failures (Liner Breach) 1.35E+05 8.42E-08 1.14E-02 Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+00 O.OOE+00 5

Small Isolation Failure - Failure-To-Seal (Type C Test)

O.OOE+00 O.OOE+00 6

Containment isolation Failures (Dependent failures, O.00E+0 O.OOE+00 Personnel Errors) 7 Severe Accident Phenomena Induced Failure 1.39E+05 1.58E-05 2.20E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 2.37E-07 2.68E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.91 E-06 3.59E-01 8c Containment Bypassed (Late SGTR) 1.39E+05 2.66E-06 3.69E-01 Core 11 Containment Event Tree (CET) Endstates Damage 5.18E-05 3.35E+00 17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company T BY: E. A. Krantz PAGE: 42 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE 5a Mean Consequence Measures for 10 - Year Test Interval for a Given Accident Class Unit 1 Person-Person-Class Description Rem Frequency Remlyr (50-miles)

-per yr (50-miles) 1 No Containment Failure 3.86E+03 3.25E-05 1.26E-01 2

Large Containment Isolation Failure (Failure-To-Close) 1.13E+05 1.15E-08 1.30E-03 3a Small Isolation Failures (Liner Breach) 3.86E+04 3.25E-06 1.25E-01 3b Large Isolation Failures (Liner Breach) 1.35E+05 3.25E-07 4.39E-02 4

Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+O0 O.OOE+00 5

Small Isolation Failure - Failure-To-Seal (Type C Test)

O.OOE+00 0.OOE+00 6

Containment isolation Failures (Dependent failures, Personnel Errors)

O.OOE+00 O.OOE+00 7

Severe Accident Phenomena Induced Failure 1.39E+05 9.38E-06 1.30E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 2.37E-07 2.68E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.83E-06 3.44E-01 8c Containment Bypassed (Late SGTR) 1.39E+05 2.62E-06 3.64E-01 All Containment Event Tree (CET) Endstates 5.02E-05 2.58E+00 17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 43 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE 5b Mean Consequence Measures for 10 - Year Test Interval for a Given Accident Class Unit 2 Person-Person-Class Description Rem Frequency Rem/yr (50-miles)

-per yr (50-miles) 1 No Containment Failure 3.86E+03 2.81E-05 1.08E-01 2

Large Containment Isolation Failure (Failure-To-Close) 1.13E+05 1.19E-08 1.35E-03 3a Small Isolation Failures (Liner Breach) 3.86E+04 2.81E-06 1.08E-01 3b Large Isolation Failures (Liner Breach) 1.35E+05 2.81E-07 3.79E-02 4

Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+00 O.OOE+00 5

Small Isolation Failure - Failure-To-Seal (Type C Test)

O.OOE+O0 O.OOE+00 6

Containment isolation Failures (Dependent failures, Personnel Errors)

O.OOE+00 O.OOE+00 7

Severe Accident Phenomena Induced Failure 1.39E+05 1.58E-05 2.20E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 2.37E-07 2.68E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.91 E-06 3.59E-01 8c Containment Bypassed (Late SGTR) 1.39E+05 2.66E-06 3.69E-01 All Containment Event Tree (CET) Endstates 5.18E-05 3.45E+00 17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company 7BY:

E. A. Krantz PAGE: 44 OF 161 FILE NO. 17670-0001, Rev. 3 1 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE 6a - Mean Consequence Measures for 15-1/2 - Year Test Interval for a Given Accident Class Unit 1 Person-

Person-Class Description Rem Frequency Rem/yr (50-miles)

-per yr (50-miles) 1 No Containment Failure 3.86E+03 3.06E-05 1.18E-01 2

Large Containment Isolation Failure (Failure-To-Close) 1.13E+05 1.15E-08 1.30E-03 3a Small Isolation Failures (Liner Breach) 3.86E+04 5.04E-06 1.94E-01 3b Large Isolation Failures (Liner Breach) 1.35E+05 5.04E-07 6.80E-02 4

Small Isolation Failure - Failure-To-Seal (Type B test) 0.OOE+00 0.00E+00 5

Small Isolation Failure - Failure-To-Seal (Type C Test) 0.00E+00 0.00E+00 6

Containment isolation Failures (Dependent failures, Personnel Errors) 0.OOE+00 0.OOE+00 7

Severe Accident Phenomena Induced Failure 1.39E+05 9.38E-06 1.30E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 2.37E-07 2.68E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.83E-06 3.44E-01 8c Containment Bypassed (Late SGTR) 1.39E+05 2.62E-06 3.64E-01 All Containment Event Tree (CET) Endstates 5.02E-05 2.66E+00 17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company IBY: E. A. Krantz I PAGE: 45 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval I

TABLE 6b - Mean Consequence Measures for 15-1/2 - Year Test Interval for a Given Accident Class Unit 2 Person-Frequency Person-Class Description Rem Fre y

Rem/yr (50-miles)

-per yr (50-miles) 1 No Containment Failure 3.86E+03 2.64E-05 1.02E-01 2

Large Containment Isolation Failure (Failure-To-Close) 1.13E+05 1.19E-08 1.35E-03 3a Small Isolation Failures (Liner Breach) 3.86E+04 4.35E-06 1.68E-01 3b Large Isolation Failures (Liner Breach) 1.35E+05 4.35E-07 5.88E-02 4

Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+00 O.OOE+00 5

Small Isolation Failure - Failure-To-Seal (Type C Test)

O.OOE+00 O.OOE+00 6

Containment isolation Failures (Dependent failures, Personnel Errors)

O.OOE+00 O.OOE+00 7

Severe Accident Phenomena Induced Failure 1.39E+05 1.58E-05 2.20E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 2.37E-07 2.68E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.91 E-06 3.59E-01 8c Containment Bypassed (Late SGTR) 1.39E+05 2.66E-06 3.69E-01 All Containment Event Tree (CET) Endstates 5.18E-05 3.52E+00 17670-0001 PB ILRT Rev 3.doc

(SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 46 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE 7a - Effect of Internal Events Hazard Risk on PB ILRT Risk Assessment (Unit 1)

Dose Rate as a Function of ILRT Interval (Person-Rem/ry)

Class Description Baseline Current Proposed 3-per-10 1-per-10 years 1-per-15-1/2 years years ILRT ILRT ILRT 1

No Containment Failure 1.35E-01 1.26E-01 1.18E-01 2

Large Containment Isolation Failure (Failure-To-Close) 1.70E-03 1.30E-03 1.30E-03 3a Small Isolation Failures (Liner Breach) 3.76E-02 1.25E-01 1.94E-01 3b Large Isolation Failures (Liner Breach) 1.32E-02 4.39E-02 6.80E-02 4

Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+00 O.OOE+00 O.OOE+00 5

Small Isolation Failure-Failure-To-Seal (Type C Test)

O.OOE+00 O.OOE+00 O.OOE+00 Containment isolation Failures (Dependent failures, Personnel Errors)

O.OOE+00 0.OOE+00 O.OOE+00 7

Severe Accident Phenomena Induced Failure 1.30E+00 1.30E+00 1.30E+00 8a Containment Bypassed (ISLOCA) 2.68E-01 2.68E-01 2.68E-01 8b Containment Bypassed (Early SGTR) 3.44E-01 3.44E-01 3.44E-01 8c Containment Bypassed (Late SGTR) 3.64E-01 3.64E-01 3.64E-01 Totals 2.47 2.58 2.66 17670-0001 PB ILRT Rev 3.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE 7b - Effect of Internal Events Hazard Risk on PB ILRT Risk Assessment (Unit 2)

Dose Rate as a Function of ILRT Interval (Person-Rem/ry)

Class Description Baseline Current Proposed 3-per-10 1-per-10 years 1-per-15-1/2 years years ILRT ILRT ILRT 1

No Containment Failure 1.17E-01 1.08E-01 1.02E-01 2

Large Containment Isolation Failure (Failure-To-Close) 1.76E-03 1.35E-03 1.35E-03 3a Small Isolation Failures (Liner Breach) 3.25E-02 1.08E-01 1.68E-01 3b Large Isolation Failures (Liner Breach) 1.14E-02 3.79E-02 5.88E-02 4

Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+00 O.OOE+00 O.OOE+00 5

Small Isolation Failure-Failure-To-Seal (Type C Test)

O.OOE+00 O.OOE+00 0.OOE+00 Containment isolation Failures (Dependent failures, Personnel Errors)

O.OOE+00 O.OOE+00 0.OOE+00 7

Severe Accident Phenomena Induced Failure 2.20E+00 2.20E+00 2.20E+OO 8a Containment Bypassed (ISLOCA) 2.68E-01 2.68E-01 2.68E-01 8b Containment Bypassed (Early SGTR) 3.59E-01 3.59E-01 3.59E-01 8c Containment Bypassed (Late SGTR) 3.69E-01 3.69E-01 3.69E-01 Totals 3.35 3.45 3.52 17670-0001 PB ILRT Rev 3,doc

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval ANALYSIS FILE: 17670-0001, Rev. 3, Appendix A A.1.0 CLIENT FPL Energy - Point Beach Nuclear Power Plant A.2.0 TITLE External Events Assessment During an Extension of ILRT Interval A.3.0 AUTHOR E. A. Krantz A.4.0 PURPOSE The purpose of this calculation is to assess the impact on the risk assessment of extending the ILRT interval from 10 to 15-1/2 years due to including the risk contribution of external events and internal flooding.

A.5.0 INTENDED USE OF ANALYSIS RESULTS The results of this calculation will be used to indicate the sensitivity of the risk associated with the extension in the ILRT interval to including the risk contribution estimates of external events and internal flooding. This analysis supports the regulatory submittal for obtaining NRC approval to extend the Integrated Leak Rate Test (ILRT) interval at PBNP from 10 years to 15-1/2 years.

A.6.0 TECHNICAL APPROACH A.6.1 Internal Flooding A.6.1.1 Internal Flooding Methodoloqy The internal flooding analysis was performed in May 1993 and is documented in Reference Al. The following steps describe the process used:

1. All areas of the plant were evaluated to select those which, if a flooding event were postulated to occur and all equipment located in the area were disabled, a plant trip would result and at least one accident mitigating component would fail. Potential scenarios addressed in this analysis consisted of those resulting from leaks/ruptures of piping/gaskets, valves, pumps, expansion joints, tanks and heat exchangers. The effects of equipment submergence due to the accumulation of water, as well as the effects from spray, dripping, and steam damage were addressed.

2. The plant areas selected in this manner were subjected to a screening procedure to exclude those areas which were clearly seen to have an insignificant contribution to plant risk if all equipment located within those areas was disabled by a flood. The contribution of such events is accounted for, or enveloped by, accident sequences already modeled in the internal events portion of the PSA. The plant areas selected were termed Point Beach's flood zones, and these zones received further treatment in this analysis.

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval

3.

A survey of each flooding zone was performed to identify each potential flooding source located within the zone. The heights of accident mitigating components above floor level were also recorded.

4.

To evaluate the potential effects of disabling equipment due to submergence, an estimate of anticipated flow rates due to leakage from flooding sources was performed. An evaluation was then made to determine whether any particular flood source could result in the accumulation of flood water. Submergence was said to occur if the flooding source has a flow rate sufficient to cause accumulation in the room (that is, it exceeds the drainage capacity of the room) and has sufficient volumetric capacity to fill the room to the height necessary to flood the equipment and cause a plant trip. Also, an evaluation was conducted to determine the potential for initiation of floods due to maintenance activities and the potential for damage of equipment due to spraying, dripping and steam blanketing.

5.

For scenarios involving equipment submergence, the likelihood that operators can intervene in the progression of the flood and either isolate the leak or divert the flood accumulation and thereby prevent a plant trip was estimated from the predicted time windows available between the onset of the flood and the damage of the accident mitigating systems in question. These time windows were estimated from the rate of accumulation and the critical volume of the zone in question for floods involving equipment submergence. In the case of spraying, dripping or steam blanketing, equipment damage was assumed to be instantaneous. In the event that a flood fails to be isolated and a plant trip occurs with accident mitigating systems disabled, any potential recovery of disabled accident mitigating systems was also modeled.

A.6.1.2 Internal Floodinq General Assumptions The following are general assumptions used in the analysis. Assumptions which were specific to a step in the analysis are provided in Reference Al.

1.

The void fraction, or fraction of a flood zone's volume which is void, was estimated to be 80 percent for each zone except Zone 2 (CSRINon-Vital Switchgear Room) and the vital switchgear room which were assigned void fractions of 0.7.

2.

The friction factor (K) for floor drain screens was assumed to be 16. This accounted for both the available flow area through the screen (about half of the screen's total area) and the friction losses of flow passing through them. The friction factor for pipes was taken to be 1 for conservatism, and that of door and equipment hatch gaps to be 2.5 based on engineering judgment.

3.

The hollow metal (fire) doors located in and on the boundaries of the flood zones were assumed to fail open when a flood height of 4 feet is reached if they're hinged to swing out from where the flood originates (due to the failure of the latching mechanism). For doors hinged to swing in toward where the flood originates, (since the door jamb provides support) the door is assumed to fail open when a flood height of 8 feet is reached.

4.

The flow through floor drains located in Zone 5 (IA Compressor/EDG Rooms) was assumed to not be limited by the capacity of the sump pumps to which these drains are connected. Since the calculated flow rates through the floor drains far exceeds the capacity of the sump pumps to which they are connected, it is assumed that the excess flow will flow out through other floor drains connected to these sumps. These drains are installed in the Unit 2 turbine building which has significant drainage capacity; therefore, no damage to equipment located in this building was assumed to occur.

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CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 50 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A.6.1.3 Internal Flooding Analysis Results The results of the internal flooding evaluation for Point Beach Nuclear Power plant are summarized as follows:

Table A-I. Internal Flooding Results Flood-Plant Area Where Flooding Primary Equipment Damaged PSA Assumed Core ing Originates Flooding Equipment Damage Zone Source Damage Frequency Frequency (l/yr)

(1/yr) 1 Auxiliary Building Service Loss of CCW, SI, RHR, and 4.6E-06 4.6E-06 for Units Water Charging in Units 1 and 2 1 and 2 Equipment for RCP seal cooling and RCS injection disabled.

2 Cable Spreading Room I Service 1B03, 1B04, 1A01, 1A02, 1.5E-06 1.5E-06 for Units Non-vital Switchgear Room Water 2B03, 2B04, 2A01,2A02, 1 and 2 No credit D11,D12,D13,D14 for Alt. Shutdown Equipment 3

Unit 1 Turbine Building or Service Auxiliary Feedwater, Main 6.4E-06

< 1.0E-07 Note 1.

Diesel Generator G1 Room Water and Feedwater Circ. Water 3

Unit 1 Turbine Building or Service Vital Switchgear (1A03, 1.1E-06

< 1.OE-07 Note 1.

Diesel Generator GI Room Water and 1A04, 1A05, 1A06, 2A04, Circ. Water 2A05, 2A06, D01, D02),

Auxiliary Feedwater 4

Auxiliary Feedwater Pump Service Vital Switchgear (1A03, 1.9E-07 1.9E-07 Area Water 1A04, 1A05, 1A06, 2A04, No credit for Alt.

2A05, 2A06, D01, D02),

Shutdown Auxiliary Feedwater Equipment.

4 Auxiliary Feedwater Pump Service Auxiliary Feedwater 2.2E-06

< 1.OE-07 Note 2.

Area Water 5

Diesel Generator Rooms, Service Loss of instrument air 1.7E-05

< 1.OE-07 Instrument Air Compressor Water Note 3.

Room. 2/6 service water pumps left. Success AFW, CH 6

Water Intake Facility Circ. Water Loss of SW and Fire 1.5E-05 4.5E-06 (Service Water, Circulating and Protection 0.3 failure Water and Fire Water)

Firewater probability F&B with AFW from hotwells or water treatment Notes:

1.

2.

3.

Random failure of systems not affected by flooding in this area is 6.2E-05.

Random failure of systems not affected by flooding in this area is 1.0E-02.

Random failure of systems not affected by flooding in this area is 1.OE-03.

Summing the core damage contributions identified in the last column of the table indicates a total CDF contribution due to internal flooding of 1.1 E-05. This information is used to provide insight into the impact of internal flooding and external events on the conclusions of this ILRT risk assessment.

A.6.2 External Events 17670-0001 PB ILRT Rev 3.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval This appendix discusses the risk-implication associated with external hazards in support of the PBNP Integrated Leak Rate Testing (ILRT) interval extension risk assessment.

In response to Generic Letter 88-20, Supplement 4 (Reference A6), PB submitted an Individual Plant Examination of External Events (IPEEE) in June 1995 (Reference A2). The IPEEE was a review of external hazard risk (i.e., seismic, fires, high winds, external flooding, etc) to identify potential plant vulnerabilities and to understand severe accident risks. The results of the PB IPEEE are therefore used in this risk assessment to provide a comparison of the effect of external hazards when extending the current 1-in-10 years to 1-in-1 5-1/2 years Type A I LRT interval.

A.6.2.1 PB IPEEE Seismic Analysis Seismic Analysis Methodology The Point Beach SPSA is documented in the Point Beach IPEEE (Reference A2) and was developed in accordance with the guidance provided in NUREG-1407 (Reference A7) and NUREG/CR-2300, "PRA Procedure Guide - A guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants," (Reference A8). Seismic event tree models were developed to address the failure of structures and components during a seismic event. The structure of seismic event trees reflects a partitioning of seismic failure and non-seismic failure mechanisms. The non-seismic failures of systems/functions are explicitly included as event tree tops in the model. Failure of containment safeguards systems (i.e., fan coolers, containment isolation) were included in the model to address scenarios leading to significant early release during a seismic event.

The enhancements recommended in Appendix 1 to Generic Letter 88-20, Supplement 4 (Reference A6) were implemented as part of the development of the SPSA. The major inputs to the SPSA development were the results and insights obtained from plant walkdown activities. The walkdown process was implemented using guidance provided in the SQUG GIP, "Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment," (Reference A9) and EPRI NP-6041 -SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," (Reference A10). Structures and. components that did not have capacities significantly greater than the Review Level Earthquake' (RLE) of 0.3g were explicitly modeled in the SPSA (i.e., unique basic event names were assigned). The failure of the remaining screened items were included as part of a surrogate top event in the event tree model. Spatial interaction basic events were developed to account for spatial interactions such as failure of block walls on SPSA components. Seismic correlated basic events were developed to account for the common cause failure of identical equipment on the same elevation. Probability distributions for post-earthquake human error events were developed to account for increased likelihood of human errors as a function of seismic hazard level. A relay chatter evaluation was performed in accordance with the requirements of NUREG-1407.

Key Assumptions in the Seismic Analysis The SPSA model was developed using the following key assumptions; (judgments of seismic "weakness" made prior to the walkdowns were confirmed during the walkdowns):

The instrument air system (lAS) was assumed to be unavailable to support active functions of air-operated equipment such as pressurizer power operated relief valves (PORVs), steam generator (SG) atmospheric steam dump valves (ASDVs), etc. However, failure of the IAS was not assumed to preclude the opening of the pressurizer PORVs caused by primary system pressurization. The IAS was judged to be seismically "weak" prior to the walkdown due to long lengths of IA piping running throughout the plant.

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval The gas turbine was assumed to be unavailable as an alternate AC power source. The gas turbine was judged to be seismically "weak" prior to the walkdown due to its non-seismic fuel oil supply (i.e., unanchored fuel oil storage tanks).

Reactor coolant pump (RCP) seal injection cooling from the chemical and volume control system (CVCS) was assumed to be unavailable. Only thermal barrier cooling via the component cooling water (CCW) System was assumed to be available for RCP seal cooling. The CVCS is not safety-related and not seismic class I (except for containment isolation purposes) and was judged to be seismically "weak" prior to the walkdown.

Normal charging to the primary system was not credited. The CVCS was judged to be seismically "weak" prior to the walkdown.

The power conversion system (PCS) was assumed to be unavailable due to the assumed loss of IAS, which is required to maintain main steam isolation valves open and to control feedwater regulating valves.

Recovery of off-site power was not credited due to the high likelihood of long-term loss of off-site power during a seismic event. However, the negative impact of offsite power being available with the PCS unavailable was considered (e.g., Anticipated Transient Without Scram (ATWS)).

Primary system depressurization can be accomplished using the pressurizer PORVs and/or auxiliary spray. However, this capability was not credited in the SPSA due to the assumed loss of IAS and CVCS.

Primary cooldown can be accomplished via the secondary system using one auxiliary feedwater (AFW) train and local/manual operation of the SG ASDV for the associated SG. However, no credit was taken for primary cooldown/depressurization due to the assumed loss of lAS.

Firewater makeup to the condensate storage tanks (CSTs).for long-term primary/secondary heat removal was not credited. The firewater equipment was judged to be seismically "weak" prior to the walkdown due to long lengths of fire water piping running throughout the plant.

Low pressure injection during small break LOCA (SLOCA) event is not credited because RCS cooldown and depressurization is not credited (as above).

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Results of the Seismic Analysis Table A-2. Seismic Results Contribution Of Individual Accident Sequences Towards The Total Plant Risk Seismic Sequence Mean Annual Core Percent of Total Damage Frequency PBO-2 8.06E-06 61.70 PBO-4 2.12E-06 16.24 PBO-3 9.56E-07 7.32 PB2-3 4.33E-07 3.32 TR33.1.4-15 3.40E-07 2.61 PBI-1 1.93E-07 1.48 All other sequences 9.98E-07 7.33 Total 1.31 E-050 100.00 Where:

Sequence PBO-2 Sequence PBO-4 Sequence PBO-3 Sequence PB2-3 Sequence TR3A-15 This sequence corresponds to failure of the cable trays inside the cable spreading room (which cause a loss of indication and control in the control room, but no losses of power to essential equipment) in conjunction with the failure of the operators to achieve safe shutdown from the independent, remote shutdown panels. This sequence contributes roughly 62 percent of the total CDF for the Point Beach plant.

This sequence results from failure of the Point Beach surrogate element. This sequence contributes roughly 16 percent of the total CDF for the Point Beach plant.

This sequence corresponds to failure of the cable trays outside the cable spreading room (since these trays do carry power cables to essential equipment, therefore their failure is assumed to go directly to core melt). This sequence contributes roughly 7 percent of the total CDF for the Point Beach plant.

This sequence results from failure of the fuel oil supply to the EDGs. This sequence contributes roughly 3 percent of the total CDF for the Point Beach plant.

This sequence results from failure of the AFW system. For this sequence to occur, off-site power must still be available and the AC buses, DC system, and.

120V AC buses must all still be available. Additionally, in this sequence, LOCA, SGTR nor any steamline break (inside or outside) occur. Failure of the AFW system is dominated by failure of all of the Level Transmitters for the Condensate Storage Tank (this removes one of the operator cues that the AFW suction source must be switched to SW) in conjunction with the failure of the operators to align the AFW suction to the backup supply from SW. This sequence contributes roughly 3 percent of the total CDF for the Point Beach plant.

5 Section 3.3 of Reference A-2 states that the seismic event CDF is 1.31 E-05 for PB. It also indicates that upgrades pending at the time of the analysis would reduce the contribution to 1.1 E-5. Section 8.1.1 indicates the dependence of the CDF result on the seismic hazard curve used; it indicates that the 1.3.1E-05 is based upon the LLNL curve but that the CDF would be 1.40E-05 if the EPRI PB-specific curve was used.

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Sequence PB1-1 This sequence results from failure of the 120V AC instrument buses Y01, Y02, Y03, and Y04 where off-site power is available and results from a block wall failure. This sequence contributes roughly 1.5 percent of the total CDF for the Point Beach plant.

The core damage information in the table above is used in this appendix to provide insight into the impact of external hazards risk on the conclusions of this ILRT risk assessment.

A.6.2.2 PB IPEEE Fire Analysis Fire Analysis Methodology The Fire analysis performed for the PB IPEEE submittal (Reference A2) used the EPRI Fire Induced Vulnerability Evaluation (FIVE) Methodology (Reference A5). The fire PRA analysis entailed the identification of critical areas of vulnerability, the calculation of fire initiation frequencies, the identification of fire-induced initiating events and their impact on systems, the disabling of critical safety functions, and potential fire-induced containment failure. Based on this examination, the Core Damage Frequency from internal fires was estimated to be 5.1 E-05/year. This was subsequently revised per Reference Al 1 to 1.24E-05/ry In general, no significant fire concerns were discovered in the Point Beach Nuclear Plant Fire Analysis.

The dominant contributors to fire-induced core damage are fires in the gas turbine building, diesel generator rooms, non-vital switchgear room, vital switchgear room, monitor tank room, control room, the cable spreading room, and the auxiliary feedwater pump room. Point Beach Nuclear Plant meets all the requirements of 10 CFR 50, Appendix R (other than exemptions approved by the NRC), and an additional equipment failure or human error in addition to the equipment damage caused by the postulated fires is necessary for core melt to occur.

Key Assumptions in the Fire Analysis The following are the key assumptions used in the analysis.

Offsite power is available unless disabled by the fire. Offsite power means offsite power is available to the safeguards 4160 VAC switchgear.

The mission time for this analysis is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Fire barriers remain intact and will contain fires of less than or equal to rated duration.

All automatic fire suppression systems are sized to effectively mitigate the maximum size fire.

All cables of interest are non-IEEE-383 rated with polyethylene (PE) insulation.

Information obtained from CARDS and the FPER is current and valid.

Fire Analysis Results The PB IPEEE submittal (Reference A-2) for the fire induced core damage scenarios and the associated frequency results were reviewed in support of this assessment. The result is judged to be conservative because of limited data and conservative fire propagation and mitigate assumptions. The CDF results for all the compartments, which were quantitatively evaluated, are provided in Table A-3.

This information is used in Section A5.0 of this appendix to provide insight into the impact of external hazard risk on the conclusions of this ILRT risk assessment.

17670-0001 PB ILRT Rev 3.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A.6.2.3 PB IPEEE Other External Events This analysis was an update of a previously performed Point Beach PSA for the TAP A-45, "Shutdown Decay Heat Removal Requirements," issue. All credible external events were addressed. Specifically examined in the other external events analysis are external flooding, aircraft accidents, severe winds, ship impact accidents, nearby industrial facility accidents, and gas turbine missiles, No vulnerabilities were identified that require additional detailed quantification of any of these accidents or events. It is therefore concluded that the effects from any of the other external events described here are not a significant concern for Point Beach Nuclear Plant.

Therefore, these other external event hazards are not included in this appendix and are expected not to impact the conclusions of this ILRT risk assessment.

A.7.0 INPUT INFORMATION

1.

The CDF contribution due to Internal Flooding from Reference Al, 1.1 E-05/yr

2.

The CDF contribution due to Seismic events from Reference A2, 1.31 E-05/yr

3.

The CDF contribution due to Fire events per Reference Al 1, 1.24E-05/yr

4.

The CDF contribution due to Other external events from Reference A2 is negligible.

5.

The calculations of the main body of this document.

6.

The distribution of release frequencies for the EPRI accident classes shown in Table 2.

A.

8.0 REFERENCES

Al.

"Internal Flooding Analysis", Point Beach Nuclear Power Plant Units 1 & 2 Probabilistic Safety Assessment, Section 6.0, Revision 0, Wisconsin Electric Power Company, May, 1993 A2.

Point Beach Nuclear Plant Individual Plant Examination of External Events for Severe Accident Vulnerabilities, Summary Report, Wisconsin Electric Company, June 30, 1995.

A3.

J. Haugh, J. M. Gisclon, W. Parkinson, K. Canavan, "Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals", Rev. 4, EPRI, November, 2001.

A4.

Regulatory Guide 1.174, "An Approach for using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis", Revision 1, November 2002.

A5.

"Fire Induced Vulnerability Evaluation Methodology, (FIVE)," prepared for EPRI, September 1991.

A6.

Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR50.54(f)", U.S. NRC, November 23, 1988.

A7.

NUREG 1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," U.S. NRC, June 1991.

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A8.

NUREG/CR-2300, "PRA Procedures Guide," U.S. NRC, January 1983.

A9.

SQUG GIP, "Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment," Revision 2, February 14, 1992.

Al 0.

EPRI NP-6041 -SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin,"

Revision 1, August 1991.

Al 1.

Personal communication, P. Knoespel/E. Krantz, August 11, 2006.

A.9.0 MAJOR ASSUMPTIONS:

1. Because the internal flooding, seismic and fire risk assessment did not report accident progression releases consistently with the internal events analysis, for the purpose of this report the EPRI accident classes for internal flooding and external events will be based on percent contribution for the accident class frequencies for Internal Events as presented in Table 2 of the main body of this report.

2.

It is assumed that the distribution of the internal flooding and external events (IF/EE) contributions to core damage frequency among the source term categories is similar to that of internal events.

A.10.0 IDENTIFICATION OF COMPUTER CODES None used.

A.11.0 DETAILED ANALYSIS:

In this analysis the PB internal flooding and IPEEE external events information presented in Section A.6 is used to calculate the following, in accordance with the NEI Interim Guidance (Reference A-3):

Evaluate the risk impact for the new surveillance intervals of interest.

Evaluate the internal flooding and external hazard risk impact in terms of LERF.

Evaluate the internal flooding external hazard change in conditional containment failure probability.

A.11.1 Estimate Level 2 Release Frequencies due to Internal Flooding and External Events It is assumed that the distribution of the internal flooding and external events (IF/EE) contributions to core damage frequency will be similar to that of internal events. The percent contribution of the total CDF to each accident class is provided in Table 2 of the main body of this report. The total contribution to CDF from IF/EE is 1.1 E-05/yr (internal flooding) + 1.31 E-05/yr (seismic) + 1.24E-05/yr (fire) = 3.65E-05/yr.

Table A-4 provides the results of distributing the internal flooding and external events CDF contributions to the EPRI accident classes.

A.1 1.2 Risk Impact for the New Surveillance Intervals 17670-0001 PB ILRT Rev 3.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval This step calculates the percentage of the total dose rate attributable to EPRI Accident Classes 3a and 3b (those accident classes affected by change in ILRT surveillance interval) and the change in this result dose rate from the base dose rate attributable to changes in ILRT surveillance interval..

The change in population dose rate is calculated as outline in Step 7 (Section 11.7) of the main body of this report. The results of these calculations are presented in Tables A-5a/b, A-6 and A-7. Tables A-5a/b provides the dose rates for a 3-in-10 years ILRT interval. Table A-6 provides the dose rates for a 1-in-10 years ILRT interval. Table A-7 provides the dose rates for a 1-in-15-1/2 years ILRT interval.

Based on the results summarized in Table A-6 and those presented in Table 5a and 5b (of the main body of the report), for the current PB 1-in-1 0 years ILRT interval, the percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b is calculated as follows:

PerCtdComb-lO Where PerCtdcomb-lO

= [(Class_3a_Dosecomb-lo + Class_3b Dosecomb-lO)/ Total_Dosecomblo]*100%

= Combined internal and external events percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b given a 1-in-10 years ILRT interval combined internal and external events EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= Class_3aDoselternal.lo + Class_3 a DoseExternal-1o

= combined internal and external events EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= Class_3bDoseinternallo + Class_3 b DoseExternal-lo Class_3 a Dosecmb-lO Class_3b DosecomblO Class_ 3a Doseinterna~io internal events EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= 1.25E-01 for Unit 1 (Table 5a)

And 1.08E-01 for Unit 2 (Table 5b)

Class_ 3bDoselnternaiio = internal events EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= 4.39E-02 for Unit 1 (Table 5a)

And

= 3.79E-02 for Unit 2 (Table 5b)

Class_ 3 aDoseExternal.io And Class_3 b DoseExternai-lo And

= internal flooding/external events EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= 9.07E-02 for Unit 1 (Table A-6)

= 7.59E-02 for Unit 2 (Table A-6)

= internal flooding/external events EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= 3.18E-02 for Unit 1 (Table A-6)

= 2.66E-02 for Unit 2 (Table A-6) 17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 58 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Total_Doseconb-lO Total_Doseinternai-lo And Total_DoseEtera,-lO And

= total combined internal and external events dose rate for all EPRI classes given a 1-in-10 years ILRT interval TotaiDoselnterna,-10 + TotalDoseExternal-10

= total internal events dose rate for all EPRI classes given a 1-in-1 0 years ILRT interval

= 2.58 person-rem/yr for Unit 1 (Table 5a)

= 3.45 person-rem/yr for Unit 2 (Table 5b)

= total IF/external events dose rate for all EPRI classes given a 1-in-10 years ILRT interval

= 1.86 person-rem/yr for Unit 1 (Table A-6)

= 2.42 person-rem/yr for Unit 2 (Table A-6)

= [(Class_3aDosecomb.10 + Class_3b Dosecomb-lO)/ Total_Dosecomb_10]*1 00%

= [({1.25E-01 + 9.07E-02} + {4.39E-02 + 3.18E-02})/(2.58 + 1.86)]

100%

= 6.6% for Unit 1

= [({1.08E-01 + 7.59E-02} + {3.79E-02+ 2.66E-02})/ (3.45 + 2.42)]

100%

= 4.2% for Unit 2 Therefore, PerCtdComb.10 And The percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b based on the proposed 1-in-1 5-1/2 years ILRT interval is calculated as follows:

PerCtdCombl5 Where PerCtdComb.15 Class_3 a Dosecmb-15 Class_3b Dosecmb-15 Class_3 a Doselnternal.15 And Class_3 b Doseinternal-15

= [(Class_3aDosecomb-15 + Class_3b Dosecomb-15)/ Total_Dosecombl15]*l00%

= Combined internal and external events percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b given a 1-in-1 5-1/2 years ILRT interval

= combined internal and external events EPRI accident Class 3a dose rate given a 1-in-1 5-1/2 years ILRT interval

= Class_3aDoseinternal-15 + Class_3 a DoseExternal_15

= combined internal and external events EPRI accident Class 3b dose rate given a 1-in-15-1/2 years ILRT interval

= Class_3bDoselnternal-15 + Class_3 b DOSeEternai-15

= internal events EPRI accident Class 3a dose rate given a 1-in-1 5-1/2 years ILRT interval

= 1.94E-01 for Unit 1 (Table 6a)

= 1.68E-01 for Unit 2 (Table 6b)

= internal events EPRI accident Class 3b dose rate given a 1-in-1 5-1/2 years ILRT interval 17670-0001 PB ILRT Rev 3.doe

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 59 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval And Class_ 3 a DoseExteal.15 And Class_ 3 b DoseExterna.15 And Total_Dosecomb-15 Total _Doseinternais5 And Total_DoseEternal-5 And

= 6.80E-02 for Unit 1 (Table 6a)

= 5.88E-02 for Unit 2 (Table 6b)

= internal flooding/external events EPRI accident Class 3a dose rate given a 1-in-15-1/2 years ILRT interval

= 1.94E-01 for Unit I (Table A-7)

= 1.18E-01 for Unit 2 (Table A-7)

= internal flooding/external events EPRI accident Class 3b dose rate given a 1-in-15-1/2 years ILRT interval

= 4.92E-02 for Unit 1 (Table A-7)

= 4.12E-02 for Unit 2 (Table A-7)

= total combined internal and external events dose rate for all EPRI classes given a 1-in-1 5-1/2 years ILRT interval TotalDosejnternai-15 + Total DoseExtemal.15

= total internal events dose rate for all EPRI classes given a 1-in-1 5-1/2 years ILRT interval

= 2.66 person-rem/yr for Unit 1 (Table 6a)

= 3.52 person-rem/yr for Unit 2 (Table 6b)

= total IF/external events dose rate for all EPRI classes given a 1-in-15-1/2 years ILRT interval

= 1.97 person-rem/yr for Unit 1 (Table A-7)

= 2.47 person-rem/yr for Unit 2 (Table A-7)

Therefore, PerCtdcomb-15

= [(Class_3aDosecomb-15 + Class_3b Dosecomb-15)/ Total_Dosecomb-15}*1 00%

= [({1.94E-01 + 1.94E-01 } + {6.80E-02 + 4.92E-02 })/(2.66 + 1.97)]

100%

= 10.9% for Unit 1 And

= [(1.68E-01 + 1.18E-01 } + {5.88E-02 + 4.12E-02 })/(3.52 + 2.47)]

100%

= 6.4% for Unit 2 Based upon the above results, the combined internal and external events changes from the 1-in10 years to 1-in 15-1/2 years dose rate is as follows:

IncrCtdcomb_10-15

= [(Total_Dosecombr_

- TotalDosecombl0) / TotalDosecomb-lo]

100%

Where:

IncrCtdCombo0l5

= combined internal and external events percent change from 1-in-10 years ILR1I Total_Dosecomb_15 interval to 1-in-15-1/2 years ILRT interval

= Total combined internal and external events dose rate for all EPRI classes given a 1-in 15-1/2 years ILRT interval 17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 60 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Total__DoseinternaIml5 And Total_DoseExterna-15 And Total_Dosecomb.iO TotalDoseintemai0o And Total_DoseExterna1o0 And Therefore, IncrCtdComb_10-15

= TotalDoselnterna1-1s + Total_DoseExternal-15

= total internal events dose rate for all EPRI classes given a 1-in-1 5-1/2 years ILRT interval

= 2.66 person-rem/yr for Unit 1 (Table 6a)

= 3.52 person-rem/yr for Unit 2 (Table 6b)

= total IF/external events dose rate for all EPRI classes given a 1-in-1 5-1/2 years ILRT interval

= 1.97 person-rem/yr for Unit 1 (Table A-7)

= 2.47 person-rem/yr for Unit 2 (Table A-7)

= Total combined internal and external events dose rate for all EPRI classes given a 1-in-10 years ILRT interval

= Total Doseinternal-i1 + Total_DoseExternal-lo

= total internal events dose rate for all EPRI classes given a 1-in-10 years ILRT interval

= 2.58 person-rem/yr for Unit 1 (Table 5a)

= 3.45 person-rem/yr for Unit 2 (Table 5b)

= total IF/external events dose rate for all EPRI classes given a 1-in-10 years ILRT interval

= 1.86 person-rem/yr for Unit I (Table A-6) 2.42 person-rem/yr for Unit 2 (Table A-6)

= [(Total_Dosecomb_15 - TotalDosecomb.o) I TotalDosecomb i0]

100%

= [({2.66 + 1.97}- {2.58 + 1.86})/(2.58 + 1.86}]* 100%

= 4.4% for Unit 1 And

= [({3.52 + 2.47} - {3.45 + 2.42}) / (3.45 + 2.42)]* 100%

= 2.1% for Unit 2 The above increase in risk on the total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15-1/2 years test interval, is found to be 4.4% for Unit 1 and 2.1% for Unit 2. These values are not significant increases in risk.

Based upon the above results, the combined internal and external events changes from the 3-in-10 years to 1-in-1 5-1/2 years dose rate is as follows:

IncrCtdcomb_3-15

= [(Total_Dosecombl15 - TotalDosecomb) / TotalDosecormb]

100%

Where:

IncrCtdcomb_3-15

= combined internal and external events percent change from 3-in-10 years ILRT interval to 1-in-15-1/2 years ILRT interval 17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 61 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Total_Dosecomb-15 TotalDoseinternal-15 And Total_DoseExternal-15 And Total_Dosecomb TotalDoseinternal And Total_DoseExternal And DeltaDosecomb 3-15

= Total combined internal and external events dose rate for all EPRI classes given a 1-in-15-1/2 years ILRT interval

= Total_Doselnterneai 5 + Total_DoseExternal.15

= total internal events dose rate for all EPRI classes given a 1-in-1 5-1/2 years ILRT interval

= 2.66 person-rem/yr for Unit 1 (Table 6a)

= 3.52 person-rem/yr for Unit 2 (Table 6b)

= total IF/external events dose rate for all EPRI classes given a 1-in-1 5-1/2 years ILRT interval

= 1.97 person-rem/yr for Unit 1 (Table A-7)

= 2.47 person-rem/yr for Unit 2 (Table A-7)

= Total combined internal and external events dose rate for all EPRI classes given a 3-in-10 years ILRT interval

= Total_Doseinternal + Total_DoseExternal

= total internal events dose rate for all EPRI classes given a 3-in-1 0 years ILRT interval 2.47 person-rem/yr for Unit 1 (Table 4a)

= 3.35 person-rem/yr for Unit 2 (Table 4b)

= total IF/external events dose rate for all EPRI classes given a 3-in-10 years ILRT interval

= 1.78 person-rem/yr for Unit 1 (Table A-5a)

= 2.35 person-rem/yr for Unit 2 (Table A-5b)

= (TotalDosecomb_15 - TotalDosecomb)

= ({2.66 + 1.97} - {2.47 + 1.78})

= 0.38 for Unit 1 And

= ({3.52 + 2.47} - {3.35 + 2.35})

= 0.29 for Unit 2 Therefore, IncrCtdComb 3.15 And

= [(Total_Dosecom.b15 - TotalDosecomb) / TotalDosecomb]

100%

= [({2.66 + 1.97} - {2.47 + 1.78}) / {2.47 + 1.78}]* 100%

= 9.0% for Unit I

= [({3.52 + 2.47} - {3.35 + 2.35}) / {3.35 + 2.35}1]* 100%

= 5.0% for Unit 2 17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 62 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval The above increase in risk on the total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 3-in-10 years test interval to a 1-in-15-1/2 years test interval, is found to be 9.0% for Unit 1 and 5.0% for Unit 2. These values are not significant increases in risk.

A.1 1.3 Evaluate the External Events Hazard Risk Impact in Terms of LERF This step, per the EPRI Interim Guidance (Reference A3) calculates the change in the large early release frequency with extending the ILRT interval from 1-in-10 years to 1-in-1 5-1/2-years.

The combined internal and external events affect on the LERF risk measure due to the proposed ILRT interval extension is calculated as follows:

ALERFcombinedlO-15 = Class-3 bCombinedI5 - Class-3 bCombinedlO Where:

ALERFcombinedlO-15 ClaSS-3 bcomhbned15 Class-3 bInterna -15 And Class-3 bEternal-15 And Class-3 bCombinedlO

= the combined internal and external events change in LERF from 1-in-10 years ILRT interval to 1-in-1 5-1/2 years ILRT interval

= the combined internal and external frequency of EPRI accident Class 3b given a 1-in-1 5-1/2 years ILRT Interval

= Class-3 binternai -15 + Class-3 bExternaHl5

= internal events frequency of EPRI accident Class 3b given a 1-in-1 5-1/2 years ILRT Interval

= 5.04E-07 for Unit 1 (Table 6a)

= 4.35E-07 for Unit 2 (Table 6b)

= External events frequency of EPRI accident Class 3b given a 1-in-1 5-1/2 years ILRT Interval

= 3.64E-07 for Unit 1 (Table A-7)

= 3.05E-07 for Unit 2 (Table A-7)

= the combined internal and external frequency of EPRI accident Class 3b given a 1-in-10 years ILRT Interval

=Class-3bnternai -10 + ClaSS-3 bExternal-10 Class-3 binternal -10 And Class-3bE,tmral-10

= internal events frequency of EPRI accident Class 3b given a 1-in-10 years ILRT Interval

= 3.25E-07/ry for Unit 1 (Table 5a)

= 2.81 E-07/ry for Unit 2 (Table 5b)

= External events frequency of EPRI accident Class 3b given a 1-in-1 0 years ILRT Interval

= 2.35E-07/ry for Unit 1 (Table A-6) 17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 63 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval And

= 1.97E-O7/ry for Unit 2 (Table A-6)

Therefore, ALERFCombinedlO-15 And

= Class-3 bCombined15 - Class-3 bCombinedlO

= (5.04E-07 + 3.64E-07) - (3.25E-07 + 2.35E-07)

= 3.08E-07/ry for Unit 1

= (4.35E-07 + 3.05E-07) - (2.81 E-07 + 1.97E-07)

= 2.62E-07/ry for Unit 2 ALERFcombined 3-15 = Class-3bcombined15 - Class-3bcombined Where:

ALERFcombined 3-15 Class-3bcormbined15 Class-3binternal -is And Class-3bExtemal-15 And Class-3 bCombinedlO Class-3 blnternal And

= the combined internal and external events change in LERF from 3-in-10 years ILRT interval to 1-in-15-1/2 years ILRT interval

= the combined internal and external frequency of EPRI accident Class 3b given a 1-in-1 5-1/2 years ILRT Interval

= Class-3 binternai -15 + Class-3 bExternal-15

= internal events frequency of EPRI accident Class 3b given a 1-in-15-1/2 years ILRT Interval

= 5.04E-07 for Unit 1 (Table 6a)

= 4.35E-07 for Unit 2 (Table 6b)

= External events frequency of EPRI accident Class 3b given a 1-in-1 5-1/2 years ILRT Interval

= 3.64E-07 for Unit 1 (Table A-7)

= 3.05E-07 for Unit 2 (Table A-7)

= the combined internal and external frequency of EPRI accident Class 3b given a 3-in-10 years ILRT Interval

= Class-3blnternal + Class-3 bExternal

= internal events frequency of EPRI accident Class 3b given a 3-in-10 years ILRT Interval

= 9.75E-08/ry for Unit 1 (Table 4a)

= 8.42E-08/ry for Unit 2 (Table 4b) 17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 64 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class-3 bExternal

= External events frequency of EPRI accident Class 3b given a 3-in-10 years ILRT Interval

= 7.05E-08/ry for Unit I (Table A-4)

And

= 5.90E-08/ry for Unit 2 (Table A-4)

Therefore, ALERFcombined3-15

= C[ass-3 bcombined15 - Class-'3 bCombined

= (5.04E-07 + 3.64E-07) - (9.75E-08 + 7.05E-08)

= 7.00E-07/ry for Unit 1 And

= (4.35E-07 + 3.05E-07) - (8.42E-08 + 5.90E-08)

= 5.97E-07/ry for Unit 2 The risk acceptance criteria of Regulatory Guide 1.174 as previously discussed in Section 11.1.8, Step 8 of this report, is used here to assess the ILRT interval extension. Regulatory Guide 1.174, "An Approach for Using PRA in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Reference A4), provides NRC recommendations for using risk information in support of applications requesting changes to the licensing basis of the plant.

The combined internal/external events contribution to LERF at Point Beach is 3.64E-06 (2.11 E-06 +

1.53E-06) for Unit 1 and 3.72E-06 (2.19E-06 + 1.53E-06) for Unit 2 (Section 11.1.1/Table A-4); these values of LERF permit changes to be made that result in -increases in LERF per Figure 4 of Regulatory Guide 1.174. The change in the combined internal events/external events LERF associated with increasing the ILRT interval at Point Beach from the currently allowed 1-in-10 years to 1-in-1 5-1/2 years is 3.08E-07 for Unit 1 and 2.62E-07 for Unit 2. Because Reference 9 defines small changes in LERF as below 1 E-06/ry, increasing the ILRT interval at Point Beach from the currently allowed 1-in-l0 years to 1-in-1 5-1/2 years represents a small change in plant risk. Similarly, the change in realistic values of LERF for moving from 3-in-1 0 years ILRT interval to 1-in-1 5-1/2 years of 7.00E-07 for Unit 1 and 5.97E-07/ry for Unit 2 falls into Region II, Small Change in Risk, of the acceptance guidelines in NRC Regulatory Guide 1.174.

A.1 1.4 Evaluate the External Events Hazard Change in Conditional Containment Failure Probability This step calculates the change in conditional containment failure probability (CCFP).

Similar to Step 9 (Section 11.1.9) of this report, the change in CCFP reflects the impact of the ILRT on both early (LERF) and late radionuclide releases. Therefore, CCFP consists of all those accident sequences resulting in a radionuclide release other that the intact containment state for EPRI accident Class 1, and small failure states for EPRI accident Class 3a. In additional, the CCFP is conditional given a severe core damage accident. The change in CCFP is calculated by the following equation:

CCFP = [1 - (Class 1 Frequency + Class_3a_ Frequency)/ CDFTotaI]

100%

For the combined internal and external events 1-in-10 years ILRT interval:

CCFPcomblo = [1 - (Class-1 comb-10 + Class_3a Comb-1O)/ CDFTotaj-Comb]

100%

Where; 17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 65 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class_1 Comb-10 Class_1 Internal-10 And Class-1 External-10 And Class_3a Comb-lO Class_3 ainternal-lo And Class_3a External-10 And CD FTotaIComb And

= combined internal and external events frequency of EPRI accident Class I given a 1-in-10 years ILRT interval

= Classl Internal-10 + Class 1 External-10

= internal events frequency of EPRI accident Class 1 given a 1-in-10 years ILRT interval

= 3.25E-05/ry for Unit 1 (Table 5a)

= 2.81 E-05/ry for Unit 2 (Table 5b)

= External events frequency of EPRI accident Class I given a 1-in-10 years ILRT interval

= 2.35E-05/ry for Unit 1 (Table A-6)

= 1.97E-05/ry for Unit 2 (Table A-6)

= combined internal and external events frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval

= Class_3a Internal-10 + Class_3a External-i0

= internal events frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval

= 3.25E-06/ry for Unit 1 (Table 5a)

= 2.81 E-06/ry for Unit 2 (Table 5b)

= External events frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval

= 2.35E-06/ry for Unit 1 (Table A-6)

= 1.97E-06/ry for Unit 2 (Table A-6)

= PB combined internal and external events CDF

= 5.02E-05/ry + 3.63E-05/ry for Unit 1 (Tables 4a and A-6)

= 8.65E-05/ry for Unit 1

= 5.18E-05/ry + 3.63E-05/ry for Unit 2 (Tables 4b and A-6)

= 8.81 E-05/ry for Unit 2

= [1 - (Classl Comb-iC + Class_3a Comb-10)/ CDFTotal-Comb]

100%

= [1-({3.25E-05 + 2.35E-05} + {3.25E-06 + 2.35E-06})/ 8.65E-05]

100%

= 28.70% for Unit 1

= [1-({2.81E-05 + 1.97E-05} + {2.81E-06 + 1.97E-06})/ 8.81 E-05]

100%

= 40.33% for Unit 2 Therefore, CCFPComblC And For the combined internal and external events 1-in-1 5-1/2 years ILRT interval:

CCFPComb15 = [1 - (Class_1 Comb-15 + Class_3a Comb-15)/ CDFTotal-comb ]

100%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 66 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Where; Class_1 Comb-15 Class_1 Internal-15 And Class_1 External-15 And Class_3a Comb-15 Class_ 3 alnternal-15 And Class_3a External-15 And CDFTotal-Comb And

= combined internal and external events frequency of EPRI accident Class 1 given a 1-in-1 5-1/2 years ILRT interval

= Class 1 Internal-15 + Class 1 External-15

= internal events frequency of EPRI accident Class 1 given a 1-in-15-1/2 years ILRT interval

= 3.06E-05/ry for Unit 1 (Table 6a)

= 2.64E-05/ry for Unit 2 (Table 6b)

= External events frequency of EPRI accident Class 1 given a 1-in-1 5-1/2 years ILRT interval

= 2.07E-05/ry for Unit 1 (Table A-7)

= 1.85E-05/ry for Unit 2 (Table A-7)

= combined internal and external events frequency of EPRI accident Class 3a given a 1-in-1 5-1/2 years ILRT interval

= Class_3a Internal-15 + Class_3a External-15

= internal events frequency of EPRI accident Class 3a given a 1-in-1 5-1/2 years ILRT interval

= 5.04E-06/ry for Unit 1 (Table 6a)

= 4.35E-06/ry for Unit 2 (Table 6b)

External events frequency of EPRI accident Class 3a given a 1-in-1 5-1/2 years ILRT interval

= 5.04E-06/ry for Unit 1 (Table A-7)

= 3.05E-06/ry for Unit 2 (Table A-7)

= PB combined internal and external events CDF

= 5.02E-05/ry + 3.63E-05/ry for Unit 1 (Tables 4a and A-6)

= 8.65E-05/ry for Unit 1

= 5.18E-05/ry + 3.63E-05/ry for Unit 2 (Tables 4b and A-6)

= 8.81 E-05/ry for Unit 2

= [1 - (Class_1 Comb-is + Class_3a Comb-15)/ CDFTotal-Comb]

100%

= [1-({3.06E-05 + 2.07E-05} + {5.04E-06 + 5.04E-06})/ 8.65E-05]

100%

= 29.06% for Unit 1

= [1-({2.64E-05 + 1.85E-05} + {4.35E-06 + 3.05E-06})/ 8.81 E-05]

100%

= 40.63% for Unit 2 Therefore, CCFPConbl5 And 17670-0001 PB ILRT Rev 3.doc

SSCIENTECHE CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 67 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Therefore, the change in the combined internal and external events conditional containment failure probability from 1-in-10 years to 1-in-i 5-1/2 years is:

ACCFPCombined 10-15

= CCFPCombined 15 - CCFPCombined 10 29.06% - 28.70% = 0.36% for Unit 1 And

= 40.63% - 40.33% = 0.30% for Unit 2 This change in CCFP of less than 1% is insignificant from a risk perspective.

For the combined internal and external events 3-in-10 years ILRT interval:

CCFPComb = [1 - (Class_1 Comb + Class_3a Comb)/ CDFTotai-Comb]

100%

Where; Class_1 Comb Class_1 Frequency And Class_1 External And Class_3a Comb Class_3a Frequency And Class_3a External And CDFTotal-Comb

= combined internal and external events frequency of EPRI accident Class 1 given a 3-in-10 ILRT interval

= Class_1 _Frequency1 + Class-1 External

= internal events frequency of EPRI accident Class 1 given a 3-in-10 years ILRT interval

= 3.50E-05/ry for Unit I (Table 4a)

= 3.03E-05/ry for Unit 2 (Table 4b)

= External events frequency of EPRI accident Class 1 given a 3-in-10 years ILRT interval 2.53E-05/ry for Unit 1 (Table A-4)

= 2.12E-05/ry for Unit 2 (Table A-4) combined internal and external events frequency of EPRI accident Class 3a given a 3-in-10 years ILRT interval

= Class_3aFrequency + Class_3a External

= internal events frequency of EPRI accident Class 3a given a 1-in-1 5-1/2 years ILRT interval

= 9.75E-07/ry for Unit 1 (Table 4a)

= 8.42E-07/ry for Unit 2 (Table 4b)

= External events frequency of EPRI accident Class 3a given a 1-in-1 5-1/2 years ILRT interval

= 7.05E-07/ry for Unit 1 (Table A-4)

= 5.90E-07/ry for Unit 2 (Table A-4)

= PB combined internal and external events CDF

= 5.02E-05/ry + 3.63E-05/ry for Unit 1 (Tables 4a and A-6)

= 8.65E-05/ry for Unit 1 17670-0001 PB ILRT Rev 3.doc

.SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 68 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval And

= 5.18E-05/ry + 3.63E-05/ry for Unit 2 (Tables 4b and A-6)

= 8.81 E-05/ry for Unit 2 Therefore, CCFPComb

= [1 - (Classl Comb + Class_3a Comb)/ CDFTotaC-corb ]

100%

= [1-({3.50E-05 + 2.53E-05} + {9.75E-07 + 7.05E-07})/ 8.65E-05]

100%

= 28.25% for Unit I And

= [1-({3.03E-05 + 2.12E-05} + {8.42E-07 + 5.90E-07})/ 8.81 E-05]

100%

= 39.95% for Unit 2 ACCFPcombined 3-15

= CCFPCombined 15 - CCFPCombined

= 29.06% - 28.25% = 0.81% for Unit 1 And

= 40.63% - 39.95% = 0.68% for Unit 2 A.12.OCOMPUTER INPUT AND OUTPUT None A.13.O

SUMMARY

OF RESULTS The effects of external hazard risk on ILRT risk are shown in Tables A-8a/b for Unit 1 and 2, respectively.

The combined internal and external events effect on the ILRT risk is shown in Tables A-9a/b for Unit2 1 and 2, respectively. This table combines the results of Tables 4a/b with the results depicted in Tables A-8a/b.

A.

14.0 CONCLUSION

S This appendix discusses the risk-implication associated with external hazards in support of the Point Beach Integrated Leak Rate Testing (ILRT) interval extension risk assessment. The following conclusions are derived from this evaluation

1. The combined internal/external events contribution to LERF at Point Beach is 3.64E-06 (2.11E-06 +

1.53E-06) for Unit 1 and 3.72E-06 (2.19E-06 + 1.53E-06) for Unit 2 (Section 11.1.1/Table A-4); these values of LERF permit changes to be made that result in increases in LERF per Figure 4 of Regulatory Guide 1.174. The change in the combined internal events/external events LERF associated with increasing the ILRT interval at Point Beach from the currently allowed 1-in-10 years to 1-in-1 5-1/2 years is 3.08E-07 for Unit 1 and 2.62E-07 for Unit 2. According to Reference 9, increasing the ILRT interval at Point Beach from the currently allowed 1-in-10 years to 1-in-15-1/2 years represents a small change in plant risk. Similarly, the change in values of LERF for moving from 3-in-10 years ILRT interval to 1-in 1/2 years of 7.OOE-07 for Unit 1 and 5.97E-07/ry for Unit 2 falls into Region II, Small Change in Risk, of the acceptance guidelines in NRC Regulatory Guide 1.174.

17670-0001 PB ILRT Rev 3.doc

'SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 69 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval

2. The combined internal and external events increase in risk for those accident sequences influenced by Type A testing, compared with the total integrated plant risk, given the change from a 1-in-10 years test interval to a 1-in-15-1/2 years test interval, is found to be 4.4% for Unit 1 and 2.1% for Unit 2. Given the low total risk to the public, these values are not significant increases in risk.

3. The change in the combined internal and external events conditional containment failure probability from 1-in-10 years to 1-in-1 5-1/2 years is 0.36% and 0.30% respectively for Units 1 and 2. A change in CCFP of less than 1% is insignificant from a risk perspective.

17670-0001 PB ILRT Rev 3.doc

.SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 70 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed I Risk impact Assessment for Extending Containment Type A Test Interval Table A-3 Point Beach Dominant IPEEE Fire Events - Core Damage Frequency Results Compartment Event Description CoF

_______________________Contribution Compartment 151 Containment The primary contributors are the charging pump cables for Unit 1 which are routed through the 3.56E-09 Spray & Safety Injection Pump area. No credit is taken for fire spray doors which would have isolated this compartment from Room the adjacent compartments.

Compartment 156 MCC IB-32 The primary contributor is the Unit 1 LCV-1 12B valve. Valve is located in this compartment.

8.07E-07 Room Outside Unit 1 Charging Without credit for manual bypass valve there is a high probability of a seal LOCA because seal Pump Rooms cooling is lost. This probability will be reduced once credit is taken for the boric acid storage tanks, LCV-112B manual bypass, Lesson Plan 2361, and Operating Procedure OP-5B. No credit is taken for fire spray doors which would have isolated this compartment from the adjacent compartments.

Compartment 166 MCC 2B-32 The primary contributor is the Unit 2 LCV-1 12B valve. Valve is located in this compartment.

1.07E-06 Room Outside Unit 2 Charging Without credit for manual bypass valve there is a high probability of a seal LOCA because seal Pump Rooms cooling is lost. This probability will be reduced once credit is taken for the boric acid storage tanks, LCV-112B manual bypass, Lesson Plan 2361, and Operating Procedure OP-5B. No credit is taken for fire spray doors which would have isolated this compartment from the adjacent compartments.

Compartment 187 Monitor Tank High initiating event frequency because of the large number of cables routed in this 4.86E-06 Room Auxiliary Operator's compartment and the number of adjacent compartments. There is automatic detection but no Station automatic suppression in this compartment. MSIVs, atmospheric steam dumps, auto start on 2/3 AFW pumps and the pressurizer PORVs are also affected.

Compartment 245 Unit 1 High initiating event frequency because there is a 250 gallon oil-filled transformer in the room.

3.50E-07 Electrical Equipment Room The oil would spread unobstructed across the floor, if there were a leak. Fire would disable one steam generator atmospheric dump valve Unit 1 CV-2015.

Compartment 246 Unit 2 High initiating event frequency because there is a 250 gallon oil-filled transformer in the room.

2.40E-07 Electrical Equipment Room The oil would spread unobstructed across the floor, if there were a leak. Fire would disable one steam generator atmospheric dump valve Unit 2 CV-2015.

Compartment 319 Non-Vital Would disable undervoltage auto start circuit for all auxiliary feedwater pumps. Modification 3.70E-06 Switchgear Room request in process which will permit auto start for auxiliary feedwater pumps if fire occurs in this area.

Compartment 681 Gas Turbine The initiating event frequency is high because the gas turbine fire frequency is high. The third 2.04E-05 17670-0001 PB ILRT Rev 3.doe

S(( SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 71 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Building diesel generator which has been installed, but not modeled and the fourth diesel generator which is scheduled to be installed by the end of 1995 will provide additional sources of alternate AC power.

Compartment 304 Auxiliary Any fire in this compartment with failure of fire suppression is assumed to fail all auxiliary 1.04E-07 Feedwater Pump Room feedwater pumps. This leads to core damage since decay heat removal fails.

Compartment 305 Vital Loss of both diesel generators results from a fire in this area. The initiating event frequency is 2.51 E-06 Switchgear Room high because of the amount of cable and number of electrical cabinets in this room. The third diesel generator which has been installed, but not modeled and the fourth diesel generator with their new vital switchgear in a different area will reduce the core damage frequency from the vital switchgear room fire.

Compartment 308 Diesel High initiating event frequency because diesel generator is located in the room. Failure rate 5.52E-06 Generator Room G01 dominated by loss of GO1. The third diesel generator which has been installed, but not modeled and the fourth diesel generator with their new vital switchgear in a different area will reduce the core damage frequency from a fire in this compartment.

Compartment 309 Diesel High initiating event frequency because diesel generator is located in the room. Failure rate 5.84E-06 Generator Room G02 dominated by loss of G02. The third diesel generator which has been installed, but not modeled and the fourth diesel generator with their new vital switchgear in a different area will reduce the core damage frequency from a fire in this compartment.

Compartment 318 Cable High initiating event frequency because there are four oil-cooled transformers in this room. Each 2.63E-06 Spreading Room transformer contains more than 200.gallons of oil which is free to spread across the floor if there is a leak. Failure rate is due to operator failing to properly align alternate shutdown switchgear.

Compartment 326 Control Room High initiating event frequency because of the large number of electrical cabinets and electrical 4.58E-06 components. Failure rate is due to operators failing to properly shut the plant down from remote shutdown panels.

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 72 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE A Distribution of Internal Flooding and External Events CDF to Accident Class PB Unit 1 IFIEE

% of Unit 1 Unit 2 IFIEE

% of Unit 2 Class Description STC Freq.

IF/EE CDF Freq.

IF/EE CDF (per yr)

(per yr) 1 No Containment Failure 1

2.53E-05 69.81%

2.12E-05 58.42%

2 Large Containment Isolation Failure (Failure-To-Close) 6 1.09E-08 0.03%

1.09E-08 0.03%

3a Small Isolation Failures (Liner Breach) 7.05E-07 1.94%

5.90E-07 1.62%

3b Large Isolation Failures (Liner Breach) 7.05E-08 0.19%

5.90E-08 0.16%

.4 Small Isolation Failure - Failure-To-Seal (Type B test) 0.00E+00 0.0%

0.OOE+00 0.0%

5 Small Isolation Failure - Failure-To-Seal (Type C Test) 0.00E+00 0.0%

0.OOE+00 0.0%

6 Containment isolation Failures (Dependent failures, Personnel 0.OOE+00 0.OOE+00 Errors) 0.0%

0.0%

7a Severe Accident Phenomena Induced Failure - Late Rupture 4

6.55E-06 18.04%

1.09E-05 29.92%

7b Severe Accident Phenomena Induced Failure - Late Basemat 2

2.13E-07 0.59%

1.90E-07 0.52%

7c Severe Accident Phenomena Induced Failure -Early Liner 5

2.12E-08 0.06%

1.89E-08 0.05%

8a Containment Bypassed (ISLOCA) 7 1.71 E-07 0.47%

1.66E-07 0.46%

8b Containment Bypassed (Early SGTR) 8 1.32E-06 3.65%

1.34E-06 3.69%

8c Containment Bypassed (Late SGTR) 3 1.90E-06 5.22%

1.86E-06 5.12%

All Containment Event Tree (CET) Endstates 3.63E-05 100.0%

3.63E-05 100.0%

LERF Frequency (STCs 5, 6, 7, 8) 1.53E-06 1.53E-06 17670-0001 PB ILRT Rev 3.doc

SCIENTECH.

CLIENT: Nuclear Management Company 7

BY: E.A. Krantz PAGE: 73 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE A-5a - Mean Consequence Measures for IFIEE for 3-in-10 Years ILRT Interval Unit I IF/EE Unit 1 Unit I Class Description Person-Rem F

(50-miles)

Frequency - per Person-Rem/yr yr.

(50-miles) 1 No Containment Failure 3.86E+03 2.53E-05 9.78E-02 2

Large Containment Isolation Failure (Failure-To-Close) 1.13E+05 1.09E-08 1.23E-03 3a Small Isolation Failures (Liner Breach) 3.86E+04 7.05E-07 2.72E-02 3b Large Isolation Failures (Liner Breach) 1.35E+05 7.05E-08 9.52E-03 4

Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+00 O.OOE+00 O.OOE+00 5

Small Isolation Failure - Failure-To-Seal (Type C Test)

O.OOE+00 O.OOE+00 O.OOE+00 6

Containment isolation Failures (Dependent failures, Personnel Errors)

O.OOE+00 O.OOE+00 O.OOE+00 7

Severe Accident Phenomena Induced Failure 1.39E+05 6.78E-06 9.43E-01 8a Containment Bypassed (ISLOCA) 1.13E+06 1.71 E-07 1.94E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.32E-06 2.49E-01 8c Containment Bypassed (Late SGTR) 1.39E+05 1.90E-06 2.64E-01 All Containment Event Tree (CET) Endstates 3.63E-05 1.78E+00 17670-0001 PB ILRT Rev 3.doc

.(( SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 74 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE A-5b - Mean Consequence Measures for IFIEE for 3-in-10 Years ILRT Interval Unit 2 Unit 2 Class Description Person-Rem (50-IF/EE Unit 2 Person-miles)

Frequency - per yr.

Rem/yr (50-miles) 1 No Containment Failure 3.86E+03 2.12E-05 8.19E-02 2

Large Containment Isolation Failure (Failure-To-Close) 1.13E+05 1.09E-08 1.23E-03 3a Small Isolation Failures (Liner Breach) 3.86E+04 5.90E-07 2.28E-02 3b Large Isolation Failures (Liner Breach) 1.35E+05 5.90E-08 7.97E-03 4

Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+00 O.OOE+00 O.OOE+00 5

Small Isolation Failure - Failure-To-Seal (Type.C Test)

O.OOE+00 O.OOE+00 O.OOE+O0 6

Containment isolation Failures (Dependent failures, Personnel Errors)

O.OOE+00 O.OOE+00 O.OOE+00 7

Severe Accident Phenomena Induced Failure 1.39E+05 1.11 E-05 1.54E+00 8a Containment Bypassed (ISLOCA).

1.13E+06 1.66E-07 1.87E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.34E-06 2.52E-01 8c Containment Bypassed (Late SGTR) 1.39E+05 1.86E-06 2.59E-01 All Containment Event Tree (CET) Endstates 3.63E-05 2.35E+00 17670-0001 PB ILRT Rev 3.doc

-(SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 75 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE A Dose Rate by Accident Class for 1-in-10 Years ILRT Interval (IF/Ext Events)

Uniti1 Unit 2 Person-IF/EE Unit I Person-IF/EE Unit 2 Uerton Class Description Rem (50-Frequency -

Rem/yr Frequency -

Rem/yr miles) per yr.

(50-miles) per yr.

(50-miles) 1 No Containment Failure 3.86E+03 2.35E-05 9.08E-02 1.97E-05 7.60E-02 2

Large Containment Isolation Failure (Failure-To-Close) 1.13E+05 1.09E-08 1.23E-03 1.09E-08 1.23E-03 3a Small Isolation Failures (Liner Breach) 3.86E+04 2.35E-06 9.07E-02 1.97E-06 7.59E-02 3b Large Isolation Failures (Liner Breach) 1.35E+05 2.35E-07 3.18E-02 1.97E-07 2.66E-02 4

Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 5

Small Isolation Failure - Failure-To-Seal (Type C Test)

O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 6

Containment isolation Failures (Dependent failures, O.OOE+00 Personnel Errors)

O.0OE+00 O.OOE+00 O.OOE+00 O.OOE+00 7

Severe Accident Phenomena Induced Failure 1.39E+05 6.78E-06 9.43E-01 1.11 E-05 1.54E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 1.71 E-07 1.94E-01 1.66E-07 1.87E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.32E-06 2.49E-01 1.34E-06 2.52E-01 8c Containment Bypassed (Late SGTR) 1.39E+05 1.90E-06 2.64E-01 1.86E-06 2.59E-01 All Containment Event Tree (CET) Endstates 3.63E-05 1.86E+00 3.63E-05 2.42E+00 17670-0001 PB ILRT Rev 3.doc

(C SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 76 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE A Dose Rate by Accident Class for 1-in-1 5-1/2 Years ILRT Interval (IF/Ext Events)

Unit 1 Unit 2 Person-IF/EE Unit 1 Person-IF/EE Unit 2 Person Person-Person-Class Description Rem (50-Frequency -

Rem/yr Frequency -

Rem/yr miles) per yr.

(50-miles) per yr.

(50-miles) 1 No Containment Failure 3.86E+03 2.07E-05 8.OOE-02 1.85E-05 7.14E-02 2

Large Containment Isolation Failure (Failure-To-Close) 1.13E+05 1.09E-08 1.23E-03 1.09E-08 1.23E-03 3a Small Isolation Failures (Liner Breach) 3.86E+04 5.04E-06 1.94E-01 3.05E-06 1.18E-01 3b Large Isolation Failures (Liner Breach) 1.35E+05 3.64E-07 4.92E-02 3.05E-07 4.12E-02 4

Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 5

Small Isolation Failure - Failure-To-Seal (Type C Test)

O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 6

Containment isolation Failures (Dependent failures, O.OOE+00 Personnel Errors) 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 7

Severe Accident Phenomena Induced Failure 1.39E+05 6.78E-06 9.43E-01 1.11 E-05 1.54E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 1.71 E-07 1.94E-01 1.66E-07 1.87E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.32E-06 2.49E-01 1.34E-06 2.52E-01 8c Containment Bypassed (Late SGTR) 1.39E+05 1.90E-06 2.64E-01 1.86E-06 2.59E-01 All Containment Event Tree (CET) Endstates 3.63E-05 1.97E+00 3.63E-05 2.47E+00 17670-0001 PB ILRT Rev 3.doc

S C

SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 77 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE A-8a - Effect of External Events Hazard Risk on PB ILRT Risk Assessment (Unit 1)

Dose Rate as a Function of ILRT Interval (Person-Rem/ry)

Class Description Baseline Current Proposed 3-per-10 1-per-10 years 1-per-15-1/2 years years ILRT ILRT ILRT 1

No Containment Failure 9.78E-02 9.08E-02 8.OOE-02 2

Large Containment Isolation Failure (Failure-To-Close) 1.23E-03 1.23E-03 1.23E-03 3a Small Isolation Failures (Liner Breach) 2.72E-02 9.07E-02 1.94E-01 3b Large Isolation Failures (Liner Breach) 9.52E-03 3.18E-02 4.92E-02 Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+00 O.OOE+00 O.OOE+00 5

Small Isolation Failure-Failure-To-Seal (Type C Test)

O.OOE+00 O.OOE+00 O.OOE+00 6

Containment isolation Failures (Dependent failures, Personnel Errors)

O.OOE+00 O.OOE+00 O.OOE+00 7

Severe Accident Phenomena Induced Failure 9.43E-01 9.43E-01 9.43E-01 8a Containment Bypassed (ISLOCA) 1.94E-01 1.94E-01 1.94E-01 8b Containment Bypassed (Early SGTR) 2.49E-01 2.49E-01 2.49E-01 8c Containment Bypassed (Late SGTR) 2.64E-01

.2.64E-01 2.64E-01 Totals 1.78 1.86 1.97 17670-0001 PB ILRT Rev 3.doc

(CSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 78 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE A-8b - Effect of External Events Hazard Risk on PB ILRT Risk Assessment (Unit 2)

Dose Rate as a Function of ILRT Interval (Person-Rem/ry)

Class Description Baseline Current Proposed 3-per-10 1-per-10 years 1-per-15-1/2 years years ILRT ILRT ILRT 1

No Containment Failure 8.19E-02 7.60E-02 7.14E-02 2

Large Containment Isolation Failure (Failure-To-Close) 1.23E-03 1.23E-03 1.23E-03 3a Small Isolation Failures (Liner Breach) 2.28E-02 7.59E-02 1.18E-01 3b Large Isolation Failures (Liner Breach) 7.97E-03 2.66E-02 4.12E-02 Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+00 O.OOE+00 O.OOE+00 Small Isolation Failure - Failure-To-Seal (Type C Test)

O.OOE+00 O.OOE+00 O.OOE+00 6

Containment isolation Failures (Dependent failures, Personnel Errors)

O.OOE+00 0.OOE+00 O.OOE+00 7

Severe Accident Phenomena Induced Failure 1.54E+00 1.54E+00 1.54E+00 8a Containment Bypassed (ISLOCA) 1.87E-01 1.87E-01 1.87E-01 8b Containment Bypassed (Early SGTR) 2.52E-01 2.52E-01 2.52E-01 8c Containment Bypassed (Late SGTR) 2.59E-01 2.59E-01 2.59E-01 Totals 2.35 2.42 2.47 17670-0001 PB ILRT Rev 3.doc

I)SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 79 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed J Risk impact Assessment for Extending Containment Type A Test Interval TABLE A-9a - Effect of Internal and External Events Risk on PB ILRT Risk Assessment (Unit 1)

Dose Rate as a Function of ILRT Interval (Person-Rem/ry)

Class Description Baseline Current Proposed 3-per-1 0 1 -per-1 0 years 1 -per-1 5-1/2 years years ILRT ILRT ILRT 1

No Containment Failure 2.33E-01 2.16E-01 1.98E-01 2

Large Containment Isolation Failure (Failure-To-Close) 2.93E-03 2.53E-03 2.53E-03 3a Small Isolation Failures (Liner Breach) 6.48E-02 2;16E-01 3.89E-01 3b Large Isolation Failures (Liner Breach) 2.27E-02 7.56E-02 1.17E-01 Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+O0 O.OE+00 O.OOE+00 Small Isolation Failure - Failure-To-Seal (Type C Test)

O.OOE+00 O.OOE+00 O.OOE+00 6

Containment isolation Failures (Dependent failures, Personnel Errors)

O.OOE+00 O.OOE+00 O.OOE+00 7

Severe Accident Phenomena Induced Failure 2.25E+00 2.25E+00 2.25E+00 8a Containment Bypassed (ISLOCA) 4.61 E-01 4.61 E-01 4.61 E-01 8b Containment Bypassed (Early SGTR) 5.93E-01 5.93E-01 5.93E-01 8c Containment Bypassed (Late SGTR) 6.28E-01 6.28E-01 6.28E-01 Totals 4.25 4.44 4.63 17670-0001 PB ILRT Rev 3.doc

(OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 80 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE A-9b - Effect of Internal and External Events Risk on PB ILRT Risk Assessment (Unit 2)

Dose Rate as a Function of ILRT Interval (Person-Rem/ry)

Class Description; Baseline Current Proposed 3-per-10 1-per-10 years 1-per-15-1/2 years years ILRT ILRT ILRT 1

No Containment Failure 1.99E-01 1.84E-01 1.73E-01 2

Large Containment Isolation Failure (Failure-To-Close) 2.99E-03 2.58E-03 2.58E-03 3a Small Isolation Failures (Liner Breach) 5.53E-02 1.84E-01 2.86E-01 3b Large Isolation Failures (Liner Breach) 1.93E-02 6.45E-02 9.99E-02 Small Isolation Failure - Failure-To-Seal (Type B test)

O.OOE+00 O.OOE+00 0.OOE+00 Small Isolation Failure-Failure-To-Seal (Type C Test) 0.OOE+00 0.OOE+00 0.OOE+00 Containment isolation Failures (Dependent failures, Personnel Errors) 0.OOE+00 0.OOE+00 0.OOE+00 7

Severe Accident Phenomena Induced Failure 3.73E+00 3.73E+00 3.73E+00 8a Containment Bypassed (ISLOCA) 4.55E-01 4.55E-01 4.55E-01 8b Containment Bypassed (Early SGTR) 6.11 E-01 6.11 E-01 6.11 E-01 8c Containment Bypassed (Late SGTR) 6.28E-01 6.28E-01 6.28E-01 Totals 5.70 5.86 5.99 17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Companyi BY: E. A. Krantz PAGE: 81 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval ANALYSIS FILE: 17670-0001, Rev. 3, Appendix B B.1.0 CLIENT FPL Energy - Point Beach Nuclear Power Plant B.2.0 TITLE Effect of Age-Related Degradation on Risk Informed/Risk Impact Assessment for Extending Containment Type A Test Interval B.3.0 AUTHOR E. A. Krantz B.4.0 PURPOSE Inspections of reinforced and steel containments at some facilities (e.g., North Anna, Brunswick, D.C.

Cook, and Oyster Creek) have indicated degradation from the inaccessible side of the steel shell and liner of primary containments. As a result of these inaccessible areas, a potential increase in risk due to liner leakage caused by age-related degradation mechanisms may occur when extending the current 1-in-1 0 years to 1-in-1 5-1/2 years Type A Integrated Leak Rate Testing (ILRT) interval.

The purpose of this calculation is to assess the effect of age-related degradation of the containment on the risk impact for extending the Point Beach Nuclear Plant (PBNP) Integrated Leak Rate Test (ILRT or Containment Type A test) interval from 10 to 15-1/2 years.

B.5.0 INTENDED USE OF ANALYSIS RESULTS The results of this calculation will be used to indicate the sensitivity of the risk associated with the extension in the ILRT interval to potential age-related degradation of the containment shell to support obtaining NRC approval to extend the Integrated Leak Rate Test (ILRT) interval at PBNP from 10 years to 15-1/2 years.

B.6.0 TECHNICAL APPROACH This present analysis shows the sensitivity of the assessment results of the risk impact of extending the PBNP Type A test interval to age-related liner corrosion.

The prior assessment included the increase in containment leakage for EPRI Containment Failure Class 3 leakage pathways that are not included in the Type B or Type C tests. These classes (3a and 3b) include the potential for leakage due to flaws in the containment shell. The impact of increasing the ILRT interval for these classes included the probability that a flaw would occur and be detected by the Type A test that was based on historical data. Since the historical data includes all known failure events, the resulting risk impact inherently includes that due to age-related degradation.

The present analysis is intended to provide additional assurance that age-related liner corrosion will not change the conclusions of the prior assessment. The methodology used for this analysis is similar to the assessments performed for Calvert Cliffs Nuclear Power Plant (CCNPP - Reference B1), Comanche Peak Steam Electric Station (CPSES - Reference B2), D. C. Cook (CNP - Reference B3) and St. Lucie (SL - Reference B4) in response to requests for additional information (RAIs) from the NRC staff. The CCNPP, CPSES and CNP extension request submittals have been approved by the NRC.

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Consistent with the Calvert Cliffs analysis, the following issues are addressed:

Differences between the containment basemat and the cylinder/dome liner The historical cylinder/dome steel shell flaw likelihood due to concealed corrosion (uninspectable)

The impact of aging The corrosion leakage dependency on containment pressure The likelihood that visual inspections will be effective at detecting a flaw Similar to the approach described in Reference Bi, this calculation uses the following steps with PBNP values utilized where appropriate:

Step 1 - Determine a corrosion-related flaw likelihood Historical data will be used to determine the annual rate of corrosion flaws for the containment. The significantly lower potential for corrosion in the freestanding PBNP containment will be included.

Step 2 - Determine an age-adjusted flaw likelihood The historical flaw likelihood will be assumed to double every 5 years. The cumulative likelihood of a flaw is then determined as a function of ILRT interval.

Step 3 - Determine the change in flaw likelihood for an increase in inspection interval The increase in the likelihood of a flaw due to age-related corrosion over the increase in time interval between tests is-then determined from the results of Step 2.

Step 4 - Determine the likelihood of a breach in containment given a flaw For there to be a significant leak from the containment, the flaw must lead to a gross breach of the containment. The likelihood of this occurring is determined as a function of pressure and evaluated at the PBNP ILRT pressure.

Step 5 - Determine the likelihood of failure to detect a flaw by visual inspection The likelihood that the visual inspection will fail to detect a flaw will be determined considering the portion of the containment that is uninspectable at PBNP as well as an inspection failure probability.

Step 6 - Determine the likelihood of non-detected containment leakage due to the increase in test interval The likelihood that the increase in test interval will lead to a containment leak not detected by visual examination is then determined as the product of the increase in flaw likelihood due to the increased test interval (Step 3), the likelihood of a breach in containment (Step 4) and the visual inspection non-detection likelihood (Step 5). The results of the above for the two regions of the containment (inspectable and uninspectable) are then added to get the total increased likelihood of non-detected containment leakage due to age-related corrosion resulting from the increase in ILRT interval.

Step 7-Determine the risk impact in terms of population dose rate and percent increase due to the increase in test interval This step calculates the change in population dose rate for EPRI accident Class 3b (all non-detectable containment failures are considered to result in large early releases), the change in percentage of the 17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 83 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval total dose rate attributable to liner corrosion and the change in this result dose rate from the base dose rate attributable to changes in ILRT surveillance interval.

The change in population dose rate is calculated as outlined in Section 11.1.7 (Step 7), of this risk assessment.

Step 8 - Determine the risk impact in terms of LERF and percent increase due to the increase in test interval This step calculates the change in the large early release frequency with extending the ILRT intervals from 1-in-10 years to 1-in-15-1/2 years given the inclusion of a postulated liner corrosion flaw failure.

Step 9-Determine the risk impact in terms of change in conditional containment failure probability and percent increase due to the increase in test interval This step calculates the change in conditional containment failure probability (CCFP). Similar to Section 11.1.9 Step 9 of this risk assessment, the change in CCFP relates to the impact of the ILRT on both early (LERF) and late radionuclide releases. Therefore, CCFP consists of all those accident sequences resulting in a radionuclide release other than the intact containment state for EPRI accident Class 1 and small failures stated for EPRI accident Class 3a. In addition, the CCFP is conditional given a severe core damage accident.

B.7.0 INPUT INFORMATION

1.

General methodology and generic results from the Calvert Cliffs assessment of age-related liner degradation (Reference B1).

2.

The PBNP ILRT test pressure of 60.7 to 61.0 psig (Reference B5 and Reference B6).

3.

The number of steel-lined containments is 70 (Reference B1).

4.

PBNP containment failure pressure of 140 psig (Reference B7). This is a 95 % confidence level best estimate of containment ultimate failure pressure.

B.

8.0 REFERENCES

B1.

"Calvert Cliffs Nuclear Power Plant Unit No. 1; Docket No. 50-317, Response to Request for

,Additional Information Concerning the License Amendment Request for a One-time Integrated Leakage Rate Test Extension," Constellation Nuclear letter to USNRC, March 27, 2002.

B2.

"Comanche Peak Steam Electric Station (CPSES), Docket Nos. 50-445 and 50-446, Response to Request for Additional Information Regarding License Amendment Request (LAR) 01-14 Revision to Technical Specification (TS) 5.5.16 Containment Leakage Rate Testing Program,"

TXU Energy letter to USNRC, June 12, 2002.

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval I B3.

"Donald C. Cook Nuclear Plant Units 1 and 2, Response to Nuclear Regulatory Commission Request for Additional Information Regarding the License Amendment Request for a One-time Extension of Integrated Leakage Rate Test Interval," Indiana Michigan Power Company, November 11, 2002.

B4._

"St. Lucie Units 1 and 2, Docket Nos. 50-335 and 50-389, Proposed License Amendments, Request for Additional Information Response on Risk-Informed One Time Increase in Integrated Leak Rate Test Surveillance Interval,' Florida Power & Light Company letter to USNRC, December 13, 2003.

B5.

"Containment Building Integrated Leak Rate Test Unit 1", ORT 17, Rev. 5, PBNP.

B6.

"Containment Building Integrated Leak Rate Test Unit 2", ORT 17, Rev. 5, PBNP.

B7.

Point Beach Nuclear Plant Individual Plant Examination Of External Events For Severe Accident Vulnerabilities Summary Report, June 30, 1995, Wisconsin Electric Power Company.

B8.

"Containment Liner Through Wall Defect due to Corrosion," Licensee Event Report, LER-NA2-99-02, North Anna Nuclear Power Station Unit 2.

B9.

"Brunswick Steam Electric Plant, Units 1 and 2, Dockets 50-325 and 50-324/License Nos. DPR-71 and DPR-62, Response to Request for Additional Information Regarding Request for License Amendments -Frequency of Performance Based Leakage Rate Testing," CP&L letter to USNRC, February 5, 2002.

B10.

"IE Information Notice No. 86-99: Degradation Of Steel Containments," USNRC, December 8, 1986.

B11.

"PRA Procedures Guide," NUREG/CR-2300, December 1982.

B12.

Regulatory Guide 1.174, "An Approach for using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis", Revision 1, November 2002.

E.9.0 MAJOR ASSUMPTIONS:

1. As indicated in the NRC's RAIs (References B3 and B4, for example) there have been 4 instances of age-related corrosion leading to holes in steel containment liners or shells. Three of these instances (Cook -Reference B3, North Anna - Reference B8 and Brunswick - Reference B9) were in concrete containments with steel liners and due to foreign material imbedded in the concrete in contact with the steel liner. The fourth instance (Oyster Creek - Reference B10) was in a freestanding steel containment and occurred in an area where sand fills the gap between the steel shell and the surrounding concrete and was attributed to water accumulating in this sand. This data is considered to represent a corrosion induced failure but is assumed to be not applicable to PBNP because of the difference in containment type.

2.

The visual inspection data are conservatively limited to 5.5 years reflecting the time from September 1996, when 10 CFR 50.55a started requiring visual inspection, through March 2002, the cutoff date for this analysis. Additional success data were not used to limit the aging impact of this corrosion issue, even though inspections were being performed prior to September 1996 (and after March 2002) and there is no evidence that liner corrosion issues were identified. (Step 1) 17670-0001 PB ILRT Rev 3.doc

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CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval

3.

As in Reference B1, the containment flaw likelihood is assumed to double every 5 years. This is included to address the increased likelihood of corrosion due to aging. (Step 2)

4. The likelihood of a significant breach in the containment due to a corrosion induced localized flaw is a function of containment pressure. At low pressure, a breach is very unlikely. Near the nominal failure point, a breach is expected. As in Reference B1, anchor points of 0.1% chance of cracking near the flaw at 20 psia and 100% chance at the failure pressure (140 psig for PBNP from Reference B7) are assumed with logarithmic interpolation between these two points. (Step 4)

5.

In general, the likelihood of a breach in the lower head region of the containment occurring, and this breach leading to a large release to the atmosphere, is less than that for the cylindrical portion of the containment. The assumption discussed in item 4 above is, however, conservatively applied to the lower head region of the containment, as well as to the cylindrical portions.

6.

All non-detected containment overpressure leakage events are assumed to be large early releases.

7.

The interval between ILRTs at the original frequency of 3 tests in 10 years is taken to be 3 years.

8.

Consistent with Reference B1, a half failure is assumed for basemat concealed (uninspectable) liner corrosion due to the lack of identified failures.

9. Consistent with Reference B1, the likelihood of the containment atmosphere reaching the outside atmosphere given a liner flaw exists was estimated as a function of the pressure inside the containment.

10. Consistent with Reference B1, a 0.05 (5%) visual inspection detection failure likelihood (given the flaw is visible) and a total detection failure likelihood of 0.10 (10%) was used.

11. Consistent with Reference B1, 1.0 (100%) visual inspection detection failure likelihood given the flaw is located in an inaccessible (uninspectable) area of the liner/basemat was assumed.

12. Consistent with Reference B1, leakage through the Basemat is 10 times less likely than through other sections of the containment structure.

B.10.0 IDENTIFICATION OF COMPUTER CODES None used.

B.11.0 DETAILED ANALYSIS:

B.11.1 Step 1 - Determine a corrosion-related flaw likelihood This step calculates historical liner flaw likelihood consistent with the Calvert Cliffs methodology. This value for Point Beach consists of the accessible (inspectable) portion of the containment cylinder and dome and the inaccessible (uninspectable) portion of the containment basemat.

The accessible portion of the containment cylinder and dome flaw likelihood is determined as follows:

CCDF = NFail2 / (NPlants

TExpo) 17670-0001 PB ILRT Rev 3.doc

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval The inaccessible portion of the containment basemat liner flaw likelihood is determined as follows:

CBMF = NFailia / (NPlants

TExpo)

Where:

CCDF = accessible portion of the containment cylinder and dome flaw CBMF = inaccessible portion of the containment basemat flaw likelihood NFaila-number of industry events due to liner corrosion = 3 (Section B9.0)

NFailj, = number of industry events due basemat corrosion = 0.5 (Section B9.0)

NPlants = number of steel-lined containments = 70 (Section B7.0)

TExpo = time exposure since issuing of 10CFR50.55a = 5.5 years (Section B9.0)

Therefore, CCDF = 3 / (70

5.5) = 7.79E-03/yr CBMF = 0.5 / (70

5.5) = 1.30E-03/yr The above results are comparable to those documented in Table B-4.

B.11.2 Step 2-Determine an age-adjusted liner flaw likelihood Per the Calvert Cliffs methodology (Reference 81), the aged adjustment liner flaw likelihood is calculated for a 15-1/2-year interval given that the failure rate doubles every 5 years (Section B9.0) or increases 14.9 % per year. In addition, the average for the 5th to 10th year was set to the historical failure*

calculated in Step 1.

The results, based on an iterative process that satisfies the above conditions are presented in Table B-I.

B.1 1.3 Step 3 - Determine the change in flaw likelihood for an increase in inspection interval This step calculates the increase in flaw likelihood at 3-in-10 years interval (or 1-in-3 years), 1-in-10 years interval, and 1-in-1 5-1/2 years interval, per the Calvert Cliffs methodology (Reference B1). The results of Step 2 are used to generate these values as follows:

Accessible (inspectable) portion of the containment cylinder and dome, CCDFlaW3..10 CCDFlawl-10 CCDFIaw 1_15

= ECCDFRateii for i=1 to 3

= ECCDFRateii for i=1 to 10

= ECCDFRateii for i=1 to 15-1/2 Inaccessible (uninspectable) portion of the containment Basemat, CB3MFlaw 3_10 CB3MFIaw 1_10 CBMFlawl-15

= ZCBMFRateii for i=1 to 3

= ZCBMFRateii for i=1 to 10

= ZCBMFRateii for i=1 to 15-1/2 Where 17670-0001 PB ILRT Rev 3.doc

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval CCDFlaw3 10

= increase in flaw likelihood at 3-in-10 years test interval given accessible portion of the containment CCDFlaw1*10

= increase in flaw likelihood at 1-in-lO years test interval given accessible portion of the containment CCDFlaw1.> 5

= increase in flaw likelihood at 1-in-15-1/2 years test interval given accessible portion of the containment CBMFlaw310

= increase in flaw likelihood at 3-in-10 years test interval given inaccessible portion of the containment CBMFlaw 1._ 0

= increase in flaw likelihood at 1-in-10 years test interval given inaccessible portion of the containment CBMFlaw 1.

5

= increase in flaw likelihood at 1-in-15-1/2 years test interval given inaccessible portion of the containment CCDFRateii

=age adjusted liner flaw likelihood, given accessible portion of the containment CBMFRateii

=age adjusted liner flaw likelihood, given inaccessible portion of the containment Therefore, CCDFlaw310

= 1.1%

CCDFlaw1.1 0

= 6.2%

CCDFlaw1_15

= 16.8%

CBMFlaw 3. 10

= 0.2%

CBMFlawl_10

= 1.0%

CBMFlawl_15

= 2.8%

The results are documented in Table B-2 B.11.4 Step 4 - Determine the likelihood of a breach in containment given a liner flaw The likelihood of a breach in containment given a liner flaw is based on the Calvert Cliffs methodology (Reference B1 ) with a Point Beach specific value for the upper-end pressure failure (100% likelihood) taken from Reference B7. A containment pressure of 140 psig (154.7 psia) corresponds with the 100%

probability of failure. The lower-end pressure failure (0.1% likelihood) is set at 20 psia, consistent with Reference BI. Per the Calvert Cliffs methodology, the containment failure probability (FP) versus containment pressure (P) is assumed to be an equation of the form:

CCDFP(P) b

e m*p Where:

CCDFP(P) = containment cylinder and dome failure probability given containment liner breach m = slope of the containment failure probability 17670-0001 PB ILRT Rev 3.doc

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval b = intercept of the containment failure probability p = containment pressure, psia The two anchor points of 0.1% at 20 psia and 100% at 154.7 psia provide sufficient information to solve for the slope m, and the intercept b, as follows:

Slope m, m = (LN (100%) - LN (0.1%)) / (Upper Pressure-Lower Pressure) m = (LN (1.0) - LN (0.001)) (154.7-20) m = 5.13E-02 Intercept b, b = CCDFP(100%) e e M.P b'= 1 / e 5.13E-02 154.7 b = 3.58E-04 The Point Beach ILRT test pressure is of 61 psig (or 75.7 psia) (References B5 and B6). Based on this pressure the likelihood of containment breach in the liner is:

CCDFP(75.7 psia)

= 3.58E-04

5.13E'02. 75.7

= 0.0174 or 1.74%

For the basemat, the failure probability is set to one-tenth of the failure probability for cylinder and dome, or 0.174%. (See Section B9.0).

Based on the above equation, containment liner breach and drywell floor intermediate values for FP are calculated and presented in Table B-3 and Figure B-i.

B.11.5 Step 5-Determine the likelihood of failure to detect a flaw by visual inspection The visual inspection detection failure likelihood for the accessible area of the containment cylinder and dome is set to 10%, consistent with the Calvert Cliffs analysis (Reference B1). This represents a 5%

(0.05) failure to identify a visual flaw and 5% (0.05) likelihood that the flaw is not visible.

Because the liner under the Basemat cannot be visually inspected, a visual detection failure likelihood of 100 % (1.0) is assigned, consistent with the Calvert Cliffs method.

The above results are documented in Table B-4.

B.1 1.6 Step 6-Determine the likelihood of non-detected containment leakage due to the increase in test interval 17670-0001 PB ILRT Rev 3.doc

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Per the Calvert Cliffs methodology (Reference B1), the likelihood of a non-detected containment leakage is calculated by multiplying the results of Steps 3, 4, and 5. This yields the following:

Accessible portion of the containment cylinder and dome, CCDLeak 3 o 10 = CCDFlaw 3_10

CCDFPILRT

CCDVisual CCDLeak1.1 0 = CCDFlaw 11 0

CCDFPILRT

CCDVisual CCDLeakj_15 = CCDFlawls15

CCDFPILRT

CCDVisual Where:

CCDLeak3 10 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and accessible portion of the containment cylinder and dome CCDLeakl 10 = likelihood of non-detected containment leakage, given 1-in-10 years test interval and accessible portion of the containment cylinder and dome CCDLeakj1 5 = likelihood of non-detected containment leakage, given 1-in-1 5-1/2 yrs test interval and accessible portion of the containment cylinder and dome CCDFIaw31 0 = increase in flaw likelihood at 3-in-1 0 years test interval given accessible portion of the containment cylinder and dome = 1.06% (0.0106),(Table B-2)

CBMFlaw 110 = increase in flaw likelihood at 1-in-10 years test interval given accessible portion of the containment cylinder and dome = 6.20% (0.062) (Table B-2)

CBMFlawl-1 5 = increase in flaw likelihood at 1-in-15 years test interval given accessible portion of the containment cylinder and dome =,16.78% (0.1678) (Table B-2)

CCDFPILRT = likelihood of containment breach at ILRT test pressure (75.7 psia) given liner flaw and accessible portion of the containment cylinder and dome = 0.0174 (1.7.4%) (Step 4)

CCDVisual = visual inspection detection failure accessible portion of the containment cylinder and dome 0.1 (10%) (Step 5)

Therefore, CCDLeak 3_10

= CCDFlaw3_10

CCDFPILRT

CCDVisual

= 0.0106

0.0174

0.1

= 1.85E-05 (.00185%)

CCDLeakl.10

= CCDFIawl.10

CCDFPILRT

CCDVisual

= 0.062

0.0174

0.1

= 1.08E-04 (.0108%)

CCDLeakl-15

= CCDFlawl_15

CCDFPILRT

CCDVisual

= 0.1678

0.0174

0.1

= 2.92E-04 (.0292%)

Inaccessible portion of the liner (Basemat),

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval CBMLeak 3_10 = CBMFlaw 3. 10

CBMFPILRT

CBMVisual CBMLeak1_j0 = CBMFlawllo

CBMFPILRT

CBMVisual CBMLeakl. 15 = CBMFlawl1 5

CBMFPILRT

CBMVisual Where:

CBMLeak 3-10 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and inaccessible portion of the containment basemat CBMLeakjj>

1 = likelihood of non-detected containment leakage, given 1-in-10 years test interval and inaccessible portion of the containment basemat CBMLeakl1 5 = likelihood of non-detected containment leakage, given 1-in-1 5-1/2 years test interval and inaccessible portion of the containment basemat CBMFlaw 3-o0 = increase in flaw likelihood at 3-in-1 0 years test interval given inaccessible portion of the containment basemat = 0.18% (0.0018) (Table B-2)

CBMFlawlo10 = increase in flaw likelihood at 1-in-10 years test interval given inaccessible portion of the containment basemat = 1.04% (0.0104) (Table B-2)

CBMFlawl_15 = increase in flaw likelihood at 1-in-15 years test interval given inaccessible portion of the containment basemat = 2.80% (0.0280) (Table B-2)

CBMFPILRT = likelihood of containment breach at ILRT test pressure (75.7 psia) given liner flaw and inaccessible portion of the containment basemat = 0.00174 (0.174%) (Step 4)

CBMVisual = visual inspection detection failure inaccessible portion of the containment basemat = 1.0 (100%) (Step 5)

Therefore, CBMLeak 3.10

= CBMFlaw 3 10

CBMFPiLRT

CBMVisual

= 0.0018

0.00174

1.0

= 3.09E-06 (0.00031%)

CBMLeakj.10

= CBMFlaw1.10

CBMFPILRT

CBMVisual

= 0.0104

0.00174

1.0

= 1.80E-05 (0.0018%)

CBMLeakl 15

= CBMFlawl. 15

CBMFPILRT

CBMVisual

= 0.0280

0.00174

1.0

= 4.88E-05 (0.00488%)

The total likelihood of non-detected containment leakage due to corrosion is, Total3 o 10 = CCDLeak3_10 + CBMLeak 3_10 Total_10 = CCDLeakl 10 + CBMLeak1_j0 Total 1 5 = CCDLeak 1 _j5 + CBMLeakl_15 17670-0001 PB ILRT Rev 3,doc

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval

Where, Total3. 10 = total likelihood of non-detected containment leakage due to corrosion, given 3-in-10 years test interval Total1_10 = total likelihood of non-detected containment leakage due to corrosion, given 1-in-10 years test interval Totall 15 = total likelihood of non-detected containment leakage due to corrosion, given 1-in-1 5-1/2 years test interval CCDLeak31 0 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and accessible portion of the containment cylinder and dome CCDLeakj_1 0 = likelihood of non-detected containment leakage, given 1-in-10 years test interval and accessible portion of the containment cylinder and dome CCDLeak1.15 = likelihood of non-detected containment leakage, given 1-in-15-1/2 years test interval and accessible portion of the containment cylinder and dome CBMLeak 3.10 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and inaccessible portion of the containment basemat CBMLeak1.10 = likelihood of non-detected containment leakage, given 1-in-10 years test interval and inaccessible portion of the containment basemat CBMLeakl.1 5 = likelihood of non-detected containment leakage, given 1-in-1 5-1/2 years test interval and inaccessible portion of the containment basemat Therefore, Total3_10

= CCDLeak 3. 10 + CBMLeak 3_10

= 0.00185% + 0.00031%

= 0.00216%

Total*.10

= CCDLeakl.10 + CBMLeak1. 0

= 0.0108% + 0.0018%

= 0.0126%

Total1 _15

= CCDLeak_1 5 + CBMLeakl.15

= 0.0292% + 0.00488%

= 0.0341%

The above results are documented in Table B-4.

B. 11.7 Step 7 - Evaluate the Risk Impact in Terms of Population Dose Rate and Percentile Change for the Interval Cases This step calculates the change in population dose rate for EPRI accident Class 3b (all non-detectable containment failures are considered to result in large early releases), the change in percentage of the 17670-0001 PB ILRT Rev 3.doc

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval total dose rate attributable to liner corrosion and the change in this result dose rate from the base dose rate attributable to changes in ILRT surveillance interval.

The change in population dose rate is calculated as outlined in Step 7 of the main body of this risk assessment.

Increase to EPRI class 3b frequencies Liner 3b Freq 3_

10 = (Probciass_3b + Liner 3b lncr3_10) * (CDFTotal - CDFindep)

Liner 3b Freq1.10 = (Probclass_3b_10 + Liner 3b_lncr 1 0o) * (CDFTotal - CDFindep)

Liner_3bFreql-15 = (Probclass-_3b_15 + Liner_3b_ ncr.. 15) * (CDFTotal - CDFIndep)

Where:

Liner_3bFreq3.10 = frequency of EPRI Class 3b due to liner corrosion failure given a 3-in-10 years ILRT interval Liner 3b Freq1 o 10 = frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-10 years ILRT interval Liner 3b Freq_15 = frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-1 5-1/2 years ILRT interval Probciass_3b = probability of large pre-existing containment liner leakage given a 3-in-10 years ILRT frequency = 0.0027 [Section 7.0]

Probcoass 3a 10 = probability of small pre-existing containment liner leakage given a 1-in-1 0 years ILRT frequency = 0.09 [Section 11.1.5]

Probc1 ass 3a 2 15 = probability of small pre-existing containment liner leakage given a 1-in-1 5-1/2 years ILRT frequency= 0.1395 [Section 11.1.5]

CDFTotal u1 = PB U1 PSA Li core damage frequency = 5.02E-05/ry [Section 11.1.1]

CDFTotai U2 = PB U2 PSA Li core damage frequency = 5.18E-05/ry [Section 11.1.11 CDFlndep UL = CDF for those individual Unit 1 sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences = 1.41 E-05/yr [Section 11.1.1]

CDFIndep U2 = CDF for those individual Unit 2 sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences = 2.06E-05/yr [Section 11.1.1]

Liner 3b Incr3.10 = Total3. 10

EPRI_3bFraction Liner 3b Incro 10 = Total1 o 10

EPRI_3bFraction Liner_3bIncr.

15 = Total1. 15

EPRI_3bFraction Where Liner 3b lncr3_10 = Liner corrosion increase in EPRI Class 3b given 3-in-10 years test interval Liner 3b Incr.o 10 = Liner corrosion increase in EPRI Class 3b given 1-in-10 years test interval Liner 3bIncrl 15 = Liner corrosion increase in EPRI Class 3b given 1-in-1 5-1/2 years test interval Total3 10 = total likelihood of non-detected containment leakage due to corrosion, given 3-in-10 years test interval = 0.00216% (Table B-4)

Total1.10 = total likelihood of non-detected containment leakage due to corrosion, given 1-in-10 years test interval = 0.0126% (Table B-4) 17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 93 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Totall.15 = total likelihood of non-detected containment leakage due to corrosion, given 1-in-15 years test interval = 0.0341% (Table B-4)

EPRI 3b Fraction = fraction of containment failures due to liner corrosion and considered to result in large early releases = 100% (Section 9.0)

Therefore, Liner_3b lncr3. 10 Liner_3b lncr 1 _1 0 Liner 3b Incrl-

= Total3.1 0

EPRI_3bFraction

= 0.00216%

1.0

= 0.00216%

= Total1.10

EPRI_3bFraction

= 0.0126%

1.0

= 0.0126%

= Total1 _15

EPRI_3bFraction

= 0.0341%

1.0

= 0.0341%

Therefore, Liner_3b Freq 3-10 And Liner_3b Freq1_16 And Liner_3b Freql. 15 And

= (ProbcIaSS 3b + Liner_3b Incr3.10) * (CDFTotal - CDFIndep)

= (0.0027 + 0.00216%) * (5.02E-05/ry - 1.41 E-05/ry)

= 9.82E-08/ry for Unit 1

= (0.0027 + 0.00216%) * (5.18E-05/ry - 2.06E-05/ry)

= 8.48E-08/ry for Unit 2

= (Probc 1 ass_3b_10 + Liner 3b Incr_1 o) * (CDFTotaI - CDFIndep)

= (0.009 + 0.0126%) * (5.02E-05/ry-1.41 E-05/ry)

= 3.29E-07/ry for Unit 1

= (0.009 + 0.0126%) * (5.18E-05/ry - 2.06E-05/ry)

= 2.84E-07/ry for Unit 2 (Probciass 3b_15 + Liner 3b Incr1.15) * (CDFmota - CDFindep)

= (0.01395 + 0.0341%) * (5.02E-05/ry - 1.41 E-05/ry)

= 5.16E-07/ry for Unit 2

= (0.01395 + 0.0341%) * (5.18E-05/ry - 2.06E-05/ry)

= 4.46E-07/ry for Unit 2 Increase to EPRI class 1 frequencies LinerI Freq 3-10 = NCF-Class_3aFrequency - Liner 3bFreq3.1o Liner 1 Freq1.10 = NCF-Class_3aFrequency_10 - Liner_3bFreq1.1 0 Liner I Freq_1. 5 = NCF-Class_3aFrequency_15 - Liner_3bFreql. 15 Where:

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 94 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed [ Risk impact Assessment for Extending Containment Type A Test Interval Liner_1_Freq 3_10 = frequency of EPRI Class 1 due to liner corrosion failure given a 3-in-10 years ILRT interval Liner_1_Freq 11 0 = frequency of EPRI Class 1 due to liner corrosion failure given a 1-in-10 years ILRT interval Liner_1_Freqj_15 = frequency of EPRI Class I due to liner corrosion failure given a 1 -in-1 5-1/2 years ILRT interval Class 3a Frequency

= Frequency of IPRI Class 3a given a 3-in-10 years ILRT interval (Section 11.1.1)

Class 3a Frequency

= 9.75E-07 for Unit 1 And

= 8.42E-07 for Unit 2 Class_3a Frequency 1 0 = frequency of small pre-existing containment liner leakage given a 1 -in-10 years ILRT interval (Section 11.1.5)

Class 3a Frequency 1 0 = 3.25E-06/ry for Unit 1 And

= 2.81 E-06/ry for Unit 2 Class_3a Frequency_15 = frequency of small pre-existing containment liner leakage given a 1-in-1 5-1/2 years ILRT interval (Section 11.1.5)

Class 3a Frequency1j

= 5.04E-06/ry for Unit 1 And

= 4.35E-06/ry for Unit 2 Liner_3b Freq3_j0

= frequency of EPRI Class 3b due to liner corrosion failure given a 3-in-10 years ILRT interval (from above)

= 9.82E-08/ry for Unit 1 And

= 8.48E-08/ry for Unit 2 Liner_3b Freq1 -10

= frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-10 years ILRT interval (from above)

= 3.29E-07/ry for Unit 1 And

= 2.84E-07/ry for Unit 2 Liner_3b Freq,1 5

= frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-1 5-1/2 years ILRT interval (from above)

= 5.16E-07/ry for Unit 2 And

= 4.46E-07/ry for Unit 2 NCF

= Frequency in which containment leakage is at or below maximum allowable Technical Specification leakage.

= 3.61 E-05/yr (Unit 1)

(Section 11.1.1)

= 3.12E-05/yr (Unit 2)

(Section 11.1.1)

Therefore, Liner lFreq3.10

= NCF-Class_3a Frequency - Liner 3b Freq3-1 0

=3.61 E 9.75E 9.82E-08 17670-0001 PB ILRT Rev 3.doc

.SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 95 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval And

=3.50E-05/ry for Unit 1

=3.12E 8.42E 8.48E-08

=3.03E-05/ry for Unit 2 Linerl_1Freq1_10 And Liner

_ Freq1.15 And

= NCF-Class_3a Frequency_10 - Liner_3bFreq1_10

=3.61 E 3.25E 3.29E-07

=3.25E-05/ry for Unit 1

=3.12E 2.81 E 2.84E-07

=2.81 E-05/ry for Unit 2

= NCF-Class_3a Frequency 15 - Liner_3bFreql.1 5

=3.61 E 5.04E 5.16E-07

=3.06E-05/ry for Unit 1

=3.12E 4.35E 4.46E-07

=2.64E-05/ry for Unit 2 The results of other pertinent calculations are presented below:

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 96 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval For 3-in-10 years, For Unit 1 Class Person-Rem 1

3.86E+03 2

1.13E+05 3a 3.86E+04 3b 1.35E+05 4

0.OOE+00 5

O.OOE+00 6

0.OOE+00 7

1.39E+05 8a 1.13E+06 8b 1.88E+05 8c 1.39E+05 Total Unit I Baseline Frequency 3.50E-05 1.50E-08 9.75E-07 9.82E-08 0.OOE+00 0.OOE+00 0.OOE+00 9.38E-06 2.37E-07 1.83E-06 2.62E-06 5.02E-05 Corrosion Addition 7.80E-10 Unit 1 Baseline Dose Rate 1.35E-01 1.70E-03 3.76E-02 1.33E-02 0.OOE+00 0.OOE+00 0.OOE+00 1.30E+00 2.68E-01 3.44E-01 3.64E-01 2.47E+00 5.09E-02 person-rem/yr 2.06%

9.82E-08 28.3%

ILRT Dose Rate from 3a and 3b =

% of Total =

LERF from 3b =

CCFP%Liner 3_10 =

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 97 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval For Unit 2 Class Person-Rem Unit 2 Baseline Frequency 1

3.86E+03 3.03E-05 2

1.13E+05 1.55E-08 3a 3.86E+04 8.42E-07 3b1 1.35E+05 8.48E-08 4

O.OOE+00 O.OOE+00 5

O.OOE+00 O.OOE+0O 6

O.OOE+00 O.OOE+00 7

1.39E+05 1.58E-05 8a 1.13E+06 2.37E-07 8b 1.88E+05 1.91 E-06 8c 1.39E+05 2.66E-06 Total 5.18E-05 Corrosion Addition 6.74E-1 0 Unit 2 Baseline Dose Rate 1.17E-01 1.76E-03 3.25E-02 1.15E-02 O.OOE+00 O.OOE+00 0.QOE+00 2.20E+00 2.68E-01 3.59E-01 3.69E-01 3.35E+00 4.40E-02 person-rem/yr 1.31%

8.48E-08 40.0%

ILRT Dose Rate from 3a and 3b =

% of Total =

LERF from 3b =

CCFP%Liner3 _,o =

17670-0001 PB JLRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 98 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval For 1-in-10 years, For Unit 1 Class Person-Rem 1

3.86E+03 2

1.13E+05 3a 3.86E+04 3b 1.35E+05 4

0.OOE+00 5

0.OOE+00 6

O.00E+00 7

1.39E+05 8a 1.13E+06 8b 1.88E+05 8c 1.39E+05 Total Unit 1 1-in-10 Frequency 3.25E-05 1.15E-08 3.25E-06 3.29E-07 0.OOE+00 O.OOE+00 0.00E+00 9.38E-06 2.37E-07 1.83E-06 2.62E-06 5.02E-05 Corrosion Addition 4.54E-09 Unit 1 1-in-10 Dose Rate 1.26E-01 1.30E-03 1.25E-01 4.45E-02 O.OOE+00 O.OOE+00 0.OOE+00 1.30E+00 2.68E-01 3.44E-01 3.64E-01 2.58E+00 1.70E-01 person-rem/yr 6.59%

3.29E-07 28.7%

ILRT Dose Rate from 3a and 3b =

% of Total =

LERF from 3b =

CCFP%Linerl 10 =

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 99 OF 161 FILE NO. 17670-0001, Re v. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval For Unit 2 Class Person-Rem Unit 2 1-in-10 Frequency 1

3.86E+03 2.81 E-05 2

1.13E+05 1.19E-08 3a 3.86E+04 2.81 E-06 3b 1.35E+05 2.84E-07 4

0.OOE+00 0.OOE+00 5

0.OOE+00 0.00E+00 6

0.OOE+00 0.00E+00 7

1.39EE+05 1.58E-05 8a 1.13E+06 2.37E-07 8b 1.88E+05 1.91 E-06 8c 1.39E+05 2.66E-06 Total 5.18E-05 Corrosion Addition 3.92E-09 Unit 2 1-in-10 Dose Rate 1.08E-01 1.35E-03 1.08E-01 3.84E-02

0.00E+00 0.OOE+00 0.OOE+00 2.20E+00 2.68E-01 3.59E-01 3.69E-01 3.45E+00 1.47E-01 person-rem/yr 4.26%

2.84E-07 40.3%

ILRT Dose Rate from 3a and 3b

% of Total.=

LERF from 3b CCFP%Liner3.10 17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 100 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval For 1-in-15-1/2 years, For Unit 1 Class Person-Rem I

3.86E+03 2

1.13E+05 3a 3.86E+04 3b 1.35E+05 4

0.00E+00 5

0.00E+00 6

0.00E+00 7

1.39E+05 8a 1.13E+06 8b 1.88E+05 8c 1.39E+05 Total Unit I 1-in-15-1/2 Frequency 3.06E-05 1.1 5E-08 5.04E-06 5.16E-07

0.00E+00 0.00E+00 0.OOE+00 9.38E-06 2.37E-07 1.83E-06 2.62E-06 5.02E-05 Corrosion Addition 1.23E-08 Unit 1 1-in-1 5-1/2 Dose Rate 1.18E-01 1.30E-03 1.94E-01 6.97E-02 0.00E+00 0.OOE+00 0.OOE+00 1.30E+00 2.68E-01 3.44E-01 3.64E-01 2.66E+00 2.64E-01 person-rem/yr 9.92%

5.16E-07 29.1%

ILRT Dose Rate from 3a and 3b =

% of Total LERF from 3b =

CCFP%Linerl 15 =

17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 101 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval For Unit 2 Class 1

2 3a 3b 4

5 6

7 8a 8b 8c Total Person-Rem 3.86E+03 1.13E+05 3.86E+04 1.35E+05 0.00E+00 0.OOE+00 0.OOE+00 1.39E+05 1.13E+06 1.88E+05 1.39E+05 Unit 2 1-in-15-1/2 Frequency 2.64E-05 1.19E-08 4.35E-06 4.46E-07 0.OOE+00 0.OOE+00 0.0OE+00 1.58E-05 2.37E-07 1.91 E-06 2.66E-06 5.18E-05 Corrosion Addition 1.06E-08 Unit 2. 1-in-1 5-1/2 Dose Rate 1.02E-01 1.35E-03 1.68E-01 6.02E-02 0.OOE+00 0.OOE+00 0.OOE+00 2.20E+00 2.68E-01 3.59E-01 3.69E-01 3.52E+00 2.28E-01 person-rem/yr 6.47%'

4.46E-07 40.7%

ILRT Dose Rate from 3a and 3b =

% of Total =

LERF from 3b=

CCFP%Liner 3 lo =

Based on the above results, the change from the 1-in-10 years to 1-in-15-1/2 yrs dose rate is as follows:

IncreaseLinerIlo-15 = ((Tot-DoseRateLiner15 - Tot-DoseRate-Linerlo)/Tot-DoseRate-Lnerlo)

100%

Where lncreaseLinerlo.15 Tot-DoseRate-Liner15 And Tot-DoseRate-Linerl0 And

= Percent change from 1-in-10 years ILRT interval to 1-in-1 5-1/2 years ILRT interval

= Total dose rate for all EPRI Classes given a 1-in-1 5-1/2 years ILRT interval

= 2.66 person-rem/yr for Unit 1

= 3.52 person-rem/yr for Unit 2

= Total dose rate for all EPRI Classes given a 1-in-10 years ILRT interval

= 2.58 person-rem/yr for Unit 1 17670-0001 PB.ILRT Rev 3.doc

.'SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 102 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval

= 3.45 person-rem/yr for Unit 2 IncreaseLinerl0-15 And

= ((Tot-DoseRate-Liner15 - Tot-DoseRate-Linerl o)/Tot-DoseRate-Linerl0)

100%

= ((2.66 - 2.58)/2.58

100% = 3.4% for Unit 1

= ((3.52 - 3.45)/3.45 *100% = 2.2% for Unit 2 The above increase in risk on the total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15-1/2 years test interval, is found to be 3.4% for Unit1 and 2.2% for Unit 2. These values can be considered to be a small increase in risk.

B.11.8 Step 8-Evaluate the Risk Impact in Terms of LERF This step calculates the change in the large early release frequency with extending the ILRT interval from 1-in-1 0 years to 1-in-15-1/2 years given the inclusion of a postulated liner corrosion flaw failure.

The affect on the LERF risk measure due to liner corrosion flaw is calculated as follows:

ALERFLinerIO-15

= Liner_3bFreql 15 - Liner 3bFreqj.1 0

Where, ALERFLinerlO-15 = the change in LERF from the 1-in-10 years ILRT interval to the 1-in-15-1/2 years interval Liner_3b Freq1_10 And Liner_3b Freqj.15 And

= frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-10 years ILRT interval (Section B. 11.7)

= 3.29E-07/ry for Unit 1 2.84E-07/ry for Unit 2

= frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-1 5-1/2 years ILRT interval (Section B.1 1.7) 5.16E-07/ry for Unit 1

= 4.46E-07/ry for Unit 2 Therefore, ALERFLinerilo-15 And

= Liner_3bFreq_1 15 - Liner 3bFreq1 _

10

= 5.16E-07/ry - 3.29E-07/ry

= 1.86E-07/ry for Unit 1

= 4.46E-07iry - 2.84E-07/ry

= 1.61 E-07/ry for Unit 2 Based on this result, the inclusion of corrosion effects in the ILRT assessment would not change the previous conclusions of this calculation (see the main body of this calculation). That is, the change in LERF from extending the interval to 15-1/2 years from the current 10 years requirement is estimated to be about 1.86E-07/ry in Unit1 and 1.61 E-07 in Unit 2. These values are between the NRC Regulatory Guide 1.174 (Reference B12) values of 1 E-07 and 1 E-06/ry and are defined as a small increase in risk.

17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 103 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Therefore, increasing the ILRT interval at Point Beach from the currently allowed 1-in-10 years to 1-in 1/2 years and taking into consideration the likelihood of a containment liner flaw due to corrosion is only a small increase in overall plant risk.

Similarly, the change in LERF from the original 3-in-10 years interval is calculated as follows:

ALERFLiner3-15

= Liner_3bFreql 15 - Liner 3bFreq3_

10

Where, ALERFLiner 3-15 = the change in LERF from the 3-in-10 years ILRT interval to the 1-in-1 5-1/2 years interval Liner_3b Freq3.10 And Liner 3b Freq,-5 And

= frequency of EPRI Class 3b due to liner corrosion failure given a 3-in-10 years ILRT interval (Section B. 11.7)

= 9.82E-08/ry for Unit 1

= 8.48E-08/ry for Unit 2

= frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-1 5-1/2 years ILRT interval (Section B. 11.7)

= 5.16E-07/ry for Unit 1

= 4.46E-07/ry for Unit 2 Therefore, ALERFLiner3-15 And

= Liner_3bFreq 1.>rs-Liner 3bFreq3 10

= 5.16E-07/ry - 9.82E-08/ry 4.18E-07/ry for Unit 1

= 4.46E-07/ry -'8.48E-08/ry

= 3.61 E-07/ry for Unit 2 Similar to the ALERFLinerlO-15 result, the ALERFLiner3 -15 also represents only a small increase in plant risk.

B.11.9 Step 9 - Evaluate the Chanqe in Conditional Containment Failure Probability This step calculates the change in conditional containment failure probability (CCFP). Similar to Section 11.1.9 of this risk assessment, the change in CCFP relates the impact of the ILRT on both early (LERF) and late radionuclide releases. Therefore, CCFP consists of all those accident sequences resulting in a radionuclide release other that the intact containment state for EPRI accident Class 1 and small failures state for EPRI accident Class 3a. In addition, the CCFP is conditional given a severe core damage accident. Therefore, the change in the conditional containment failure probability from 1-in-10 years to 1-in-15-1/2 years is:

ACCFPLiner10-15 = CCFPLiner1i CCFPLinerl-10

Where, 17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 104 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07 SUJBJECT: Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval ACCFPLjnerIl-15 = the change in conditional containment failure probability from 1-in-10 years to 1-in-1 5-1/2 years given not-detected containment leakage.

CCFPLunerl-1 5

= conditional containment failure probability given 1 -in-1 5-1/2 years ILRT interval and potential non-detected containment leakage

= 29.1% for Unit 1 (Step 7)

And

= 40.7% for Unit 2 CCFPLiner-lo

= conditional containment failure probability given 1-in-10 years ILRT interval and potential non-detected containment leakage

= 28.7% for Unit 1 (Step 7)

And

= 40.3% for Unit 2 Therefore, ACCFPLiner1o-15

= CCFPLiner1 CCFPLinerl-1O

= 29.1% - 28.7% = 0.4% for Unit 1 And

= 40.7% - 40.3% = 0.4% for Unit 2 This change in CCFP of less that 1% is insignificant from a risk perspective.

The results of Steps 7, 8 and 9 of this ILRT assessment including the potential impact from non-detected containment leakage scenarios assuming that 100% of the leakages result in EPRI class 3b are shown in Table B-5a/b.

B. 11.9 - Steel Shell Corrosion Sensitivit, Additional sensitivity cases were also developed to gain an understanding of the sensitivity of this analysis to the various key parameters. The sensitivity cases are as follows:

Sensitivity Case 1 - Flaw rate doubles every 2 years Sensitivity Case 2 - Flaw rate doubles every 10 years Sensitivity Case 3 - 5% Visual inspection failures Sensitivity Case 4 - 15% Visual inspection failures Sensitivity Case 5 - Containment breach base point 10 times lower Sensitivity Case 6 - Containment breach base point 10 times higher 0

Sensitivity Case 7 - Flaw rate doubles every 10 years, containment breach base point 10 times lower, 5% visual inspection failures and 10% EPRI accident Class 3b are LERF (Lower bound)

Sensitivity Case 8 - Flaw rate doubles every 2 years, containment breach base point 10 times higher, 15% visual inspection failures and 100% EPRI accident Class 3b are LERF (upper bound)

The above sensitivities cases used the methodology presented in Steps 2B to 9B. These steps were accomplished in an EXCEL spreadsheet. The results are provided in Attachment A.

These results are summarized in Table B-6.

17670-0001 PB ILRT Rev 3.doe

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 105 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval B.12.0 COMPUTER INPUT AND OUTPUT None B.13.0

SUMMARY

OF RESULTS The following table summarizes the impact of corrosion on the major results of the ILRT extension analysis from the body of this analysis.

For Unit 1:

1-in-15-1/2 3-in-10 Years 1-in-10 Years Years ILRT ILRT Interval ILRT Interval Interval (with corrosion)

(with corrosion)

ILRT Dose Rate from 3a and 3b 5.09E-02 1.70E-01 2.64E-01 Increase over original 1.05E-04 6.14E-04 1.66E-03

% of Total Dose 2.06%

6.59%

9.92%

Increase over original 0.00%

0.02%

0.06%

Delta Dose Rate from 3a and 3b (10 to 15-1/2 yr) 9.42E-02 Increase over original 1.05E-03 LERF from 3b 9.82E-08 3.29E-07 5.16E-07 Increase over original 7.80E-10 4.54E-09 1.23E-08 Delta LERF (10 to 15-1/2 yr) 1.86E-07 Increase over original 7.75E-09 CCFP%

28.3%

28.7%

29.1%

Increase over original 0.002%

0.004%

0.02%

Delta CCFP% (10 to 15-1/2 yr) 0.4%

Increase over original 0.015%

17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 106 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval For Unit 2:

1-in-15-1/2 3-in-10 Years 1-in-10 Years Years ILRT ILRT Interval ILRT Interval Interval (with corrosion)

(with corrosion)

ILRT Dose Rate from 3a and 3b 4.40E-02 1.47E-01 2.28E-01 Increase over original 9.10E-05 5.30E-04 1.43E-03

% of Total Dose 1.31%

4.26%

6.47%

Increase over original 0.00%

0.01%

0.04%

Delta Dose Rate from 3a and 3b (10 to 15-1/2 yr) 8.13E-02 Increase over original 9.04E-04 LERF from 3b 8.48E-08 2.84E-07 4.46E-07 Increase over original 6.74E-10 3.92E-09 1.06E-08 Delta LERF (10 to 15-1/2 yr) 1.61E-07 Increase over original 6.69E-09 CCFP%

40.0%

40.3%

40.7%

Increase over original 0.0013%

0.003%

0.016%

Delta CCFP% (10 to 15-1/2 yr) 0.4%

Increase over original 0.013%

B.

14.0 CONCLUSION

S This appendix provides a sensitivity evaluation of considering potential containment liner corrosion impacts within the structure of the ILRT interval extension risk assessment. The evaluation yields the following conclusions:

1. The impact of including age-adjusted corrosion effects in the ILRT assessment has minimal impact on plant risk and is, therefore, acceptable.

2. The change in LERF, taking into consideration the likelihood of a containment liner flaw due to age-adjusted corrosion, is not significant from a risk perspective. Specifically, for extending the interval to 15-1/2 years from the current 10 years requirement, the change in LERF due to including corrosion is estimated to be 7.75E-09 for Unit 1 (6.69E-09/ry for Unit 2). This is below the Regulatory Guide 1.174 acceptance criteria threshold of 1 E-07/yr.

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 107 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval

3. The age-adjusted corrosion impact in dose increase is estimated to be 1.05E-03 person-rem/ry for Unit 1 (9.04E-04 person-rem/ry for Unit 2) from the current 1-in-10 years interval.

4. The age-adjusted corrosion impact on the conditional containment failure probability increase is estimated to be 0.015% for Unit 1 and 0.013% for Unit 2.

5. A series of parametric sensitivity studies regarding potential age related corrosion effects on the containment steel liner also demonstrated minimal impact on plant risk.

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 108 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval I

Table B-1 Flaw Failure Rate as a Function of Time CCDF Accessible Area CBMF Inaccessible Year Failure Rate Success Rate Failure Rate Success Rate 0

2.67E-03 9.97E-01 4.46E-04 1.OOE+00 1

3.07E-03 9.97E-01 5.12E-04 9.99E-01 2

3.52E-03 9.96E-01 5.89E-04 9.99E-01 3

4.05E-03 9.96E-01 6.77E-04 9.99E-01 4

4.65E-03 9.95E-01 7.77E-04 9.99E-01 5

5.35E-03 9.95E-01 8.93E-04 9.99E-01 6

6.14E-03 9.94E-01 1.03E-03 9.99E-01 7

7.06E-03 9.93E-01 1.18E-03 9.99E-01 8

8.11 E-03 9.92E-01 1.35E-03 9.99E-01 9

9.32E-03 9.91 E-01 1.56E-03 9.98E-01 10 1.07E-02 9.89E-01 1.79E-03 9.98E-01 11 1.23E-02 9.88E-01 2.06E-03 9.98E-01 12 1.41 E-02 9.86E-01 2.36E-03 9.98E-01 13 1.62E-02 9.84E-01 2.71 E-03 9.97E-01 14 1.87E-02 9.81 E-01 3.12E-03 9.97E-01 15 2.14E-02 9.79E-01 3.58E-03 9.96E-01 15 1/2 2.30E-02 9.77E-01 3.85E-03 9.96E-01 17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 109 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table B-2 Flaw Failure as a Function of Test Interval CCDF Accessible Area CBMF Inaccessible Years Failure Rate Success Rate Failure Rate Success Rate 3-in-10 1.06%

9.89E-01 0.18%

9.98E-01 1-in-10 6.20%

9.38E-01 1.04%

9.90E-01 1 -in-15-1/2 16.78%

8.32E-01 2.80%

9.72E-01 17670-0001 PB ILRT Rev 3.doc

S(aSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz IPAGE: 110 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table B-3 Point Beach Containment Failure Probability Given Containment Liner Flaw Pressure (psia)

Containment Liner Failure Probability Basemat Failure Probability 0

0.0004 0.0000 10 0.0006 0.0001 20 0.0010 0.0001 30 0.0017 0.0002 40 0.0028 0.0003 50 0.0047 0.0005 60 0.0078 0.0008 70 0.0130 0.0013 80 0.0217 0.0022 90 0.0362 0.0036 100 0.0605 0.0061 110 0.1011 0.0101 120 0.1688 0.0169 130 0.2820 0.0282 140 0.4710 0.0471 150 0.7866 0,0787 160 1.3139 0.1314 17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 111 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Figure B-1 Point Beach Containment Failure Probability Given Containment Liner Flaw Point Beach Containment Failure Probability Given Containment Liner Flaw 1.0000

'0.8000

--a-Containment Liner

"-0.4000

-,- Basemat 0.2000 ca 0.0000 f-:--

0 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160

'Containment Pressure, psia 17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 112 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table B-4 Point Beach Containment Liner Corrosion Base Case Step Description Accessible Area of Liner Basemat 1

Historical Steel Shell Flaw Likelihood 7.78E-03 1.30E-03 2

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

3.07E-03 1

5.12E-04 (Reference Table B1) 5-15 7.78E-03 5-15 1.30E-03 15.5 2.30E-02 15.5 3.85E-03 3

Increase in Flaw Likelihood at 3, 10, and 15 years 3-in-10 1.06%

3-in-10 0.18%

1-in-10 6.20%

1-in-10 1.04%

1-in-1 5.5 16.78%

1-in-15.5 2.80%

Likelihood of Breach in Containment Given Steel Likelihood of Pressure Likelihood of 4

Shell Flaw Pressure (psia)

Breach (psia)

Breach 20 0.0010 20 0.0001 (Reference Table B3) 75.7 (ILRT) 0.01740 75.7 (ILRT) 0.00174 130 0.2820 130 0.0282 140 0.4710 140 0.0471 154.7 1.0000 154.7 0.1000 5

Visual Inspection Detection Failure Likelihood 0.1 10%

1 100%

Likelihood of Non-Detected Containment Leakage 6

(Steps 3

4

5) 3-in-10 0.00185%

3-in-10 0.00031%

1-in-10 0.0108%

1-in-10 0.00180%

1-in-15.5 0.0292%

1-in-15.5 0.00488%

Total Likelihood of Non-Detected Containment Leakage 3-in-10 0.00216%

1-in-10 0.0126%

1-in-15.5 0.0341%

17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 113 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table B-5a Impact of Containment Steel Linder Corrosion on Point Beach ILRT Intervals (Unit 1)

Base Case 3 Years Extend to 10 Years Extend to 15 Years CDF Per-REM CDF Per-REM CDF Per-REM EPRI Class (Per RY)

Per-REM (Per RY)

(Per RY)

Per-REM (Per RY)

(Per RY)

Per-REM (Per RY) 1 3.50E-05 3.86E+03 1.35E-01 3.25E-05 3,86E+03 1.26E-01 3.06E-05 3.86E+03 1.18E-01 2

1.50E-08 1.13E+05 1.70E-03 1.15E-08 1.13E+05 1.30E-03 1.15E-08 1.13E+05 1.30E-03 3a 9.75E-07 3.86E+04 3.76E-02 3.25E-06 3.86E+04 1.25E-01 5.04E-06 3.86E+04 1.94E-01 3b 9.82E-08 1.35E+05 1.33E-02 3.29E-07 1.35E+05 4.45E-02 5.16E-07 1.35E+05 6.97E-02 4

0.00E+00 O.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.00E+00 0.00E+00 0.OOE+00 5

O.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 6

0.00E+00 O.OOE+00 O.OOE+00 0.00E+00 0,00E+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 7

9.38E-06 1.39E+05 1.30E+00 9.38E-06, 1.39E+05 1.30E+00 9.38E-06 1.39E+05 1.30E+00 8a 2.37E-07 1.13E+06 2.68E-01 2.37E-07 1.13E+06 2.68E-01 2.37E-07 1.13E+06 2.68E-01 8b 1.83E-06 1.88E+05 3.44E-01 1.83E-06 1.88E+05 3.44E-01 1.83E-06 1.88E+05 3.44E-01 8c 2.62E-06 1.39E+05 3.64E-01 2.62E-06 1.39E+05 3.64E-01 2.62E-06 1.39E+05 3.64E-01 Total 5.02E-05 0.OOE+00 2.47E+00 5.02E-05 0.OOE+00 2.58E+00 5.02E-05 0.00E+00 2.66E+00 ILRT Dose Rate from 3a and 3b 5.09E-02 1.70E-01 2.64E-01 Increase over original 1.05E-04 6.14E-04 1.66E-03

% of Total 2.06%

6.59%

9.92%

Increase over original 0.00%

0.02%

0.06%

Delta Dose Rate from 3a and 3b (10 to 15 yr) 9.42E-02 Increase over original 1.05E-03 17670-0001 PB ILRT Rev 3.doc

SC SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz T PAGE: 114 OF 161 FILE NO. 17670-0001, Rev.3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval LERF from 3b 9.82E-08 3.29E-07 5.16E-07 Increase over original 7.80E-10 4.54E-09 1.23E-08 Delta LERF (10 to 15 yr) 1.86E-07 Increase over original 7.75E-09 CCFP%

28.3%

28.7%

29.1%

Increase over original 0.002%

0.004%

0.02%

Delta CCFP% (10 to 15 yr) 0.4%

Increase over original 0.015%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 115 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table B-5b Impact of Containment Steel Linder Corrosion on Point Beach ILRT Intervals (Unit 2)

Base Case 3 Years Extend to 10 Years Extend to 15 Years CDF Per-REM CDF Per-REM CDF Per-REM EPRI Class (Per RY)

Per-REM (Per RY)

(Per RY)

Per-REM (Per RY)

(Per RY)

Per-REM (Per RY) 1 3.03E-05 3.86E+03 1.17E-01 2.81E-05 3.86E+03 1.08E-01 2.64E-05 3.86E+03 1.02E-01 2

1.55E-08 1.13E+05 1.76E-03 1.19E-08 1.13E+05 1.35E-03 1.19E-08 1.13E+05 1.35E-03 3a 8.42E-07 3.86E+04 3.25E-02 2.81E-06 3.86E+04 1.08E-01 4.35E-06 3.86E+04 1.68E-01 3b 8.48E-08 1.35E+05 1.15E-02 2.84E-07 1.35E+05 3.84E-02 4.46E-07 1.35E+05 6.02E-02 4

O.OOE+00 0.00E+00 O.OOE+00 0.00E+00 O.OOE+00 0.OOE+00 O.00E+00 0.OOE+00 0.OOE+00 5

0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.00E+00 0.OOE+00 0.OOE+00 6

0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 7

1.58E-05 1.39E+05 2.20E+00 1.58E-05 1.39E+05 2.20E+00 1.58E-05 1.39E+05 2.20E+00 8a 2.37E-07 1.13E+06 2.68E-01 2.37E-07 1.13E+06 2.68E-01 2.37E-07 1.13E+06 2.68E-01 8b 1.91 E-06 1.88E+05 3.59E-01 1.91 E-06 1.88E+05 3.59E-01 1.91 E-06 1.88E+05 3.59E-01 8c 2.66E-06 1.39E+05 3.69E-01 2.66E-06 1.39E+05 3.69E-01 2.66E-06 1.39E+05 3.69E-01 Total 5.18E-05 0.OOE+00 3.35E+00 5.18E-05 0.00E+00 3.45E+00 5.18E-05 O.00E+00 3.52E+00 ILRT Dose Rate from 3a and 3b 4.40E-02 1.47E-01 2.28E-01 Increase over original 9.10E-05 5.30E-04 1.43E-03

% of Total 1.31%

4.26%

6.47%

Increase over original 0.00%

0.01%

0.04%

Delta Dose Rate from 3a and 3b (10 to 15 yr) 8.13E-02 Increase over original 9.04E-04 17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 116 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval LERF from 3b 8.48E-08 2.84E-07 4.46E-07 Increase over original 6.74E-1 0 3.92E-09 1.06E-08 Delta LERF (10 to 15 yr) 1.61 E-07 Increase over original 6.69E-09 CCFP%

40.0%

40.3%

40.7%

Increase over original 0.0013%

0.003%

0.016%

Delta CCFP% (10 to 15 yr) 0.3%

Increase over original 0.013%

17670-0001 PB ILRT Rev 3.doe

S( SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 117 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table B-6a Sensitivity Analysis Summary for Containment Steel Liner Corrosion (Unit 1)

Age Containment Visual Likelihood LERF LERF LERF Total LERF (Step 2)

Breach Inspection Flaw is Increase Increase Increase Increase (Step 4)

& Non-LERF (EPRI From From From ILRT Visual Class 3b)

Corrosion Corrosion Corrosion (1-Extension Flaws (3-in-10 (1-in-10 in-15-1/2 (10-in-15-1/2 (Step 5) years) years) years) years)

Base Case Base Case Base Case Base Case Base Case Base Case Base Case Base Case Doubles every 1.74% Liner, 10%

100%

5 years 0.174% Basemat 7.80E-10 4.54E-09 1.23E-08 1.86E-07 Sensitivities Doubles Base Base Base every 2 years 2.24E-10 3.79E-09 2.97E-08 2.05E-07 Doubles Base Base Base every 10 years 1.16E-09 5.03E-09 1.02E-08 1.84E-07 Base Base 5%

Base 4.46E-10 2.60E-09 7.03E-09 1.83E-07 Base Base 15%

Base 1.11E-09 6.49E-09 1.76E-08 1.90E-07 Base 0.451% Liner, Base Base 0.0451% Basemat 2.02E-10 1.18E-09 3.19E-09 1.81 E-07 Base 6.72% Liner, Base Base 0.672% Basemat 3.01 E-09 1.75E-08 4.75E-08 2.09E-07 Lower Bound Doubles 0.451% Liner, 5%

10%

every 10 0.0451% Basemat years 1.72E-11 7.45E-11 1.51 E-10 1.79E-07 Upper Bound Doubles 6.72% Liner, 15%

100%

every 2 years 0.672% Basemat I.23E-09 2.09E-08 1.64E-07 3.21E-07 17670-0001 PB ILRT Rev 3.doc

SjSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 118 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table B-6b Sensitivity Analysis Summary for Containment Steel Liner Corrosion (Unit 2)

Age Containment Visual Likelihood LERF LERF LERF Total LERF (Step 2)

Breach Inspection Flaw is Increase Increase Increase Increase (Step 4)

& Non-LERF (EPRI From From From ILRT Visual Class 3b)

Corrosion Corrosion Corrosion (1-Extension Flaws (3-in-10 (1-in-10 in-15-1/2 (10-in-15-1/2 (Step 5) years) years) years) years)

Base Case Base Case Base Case Base Case Base Case Base Case Base Case Base Case Doubles every 5 1.74% Liner, 10%

100%

years 0.174% Basemat 6.74E-10 3.92E-09 1.06E-08 1.61 E-07 Sensitivities Doubles every 2 Base Base Base years 1.93E-10 3.27E-09 2.56E-08 1.77E-07 Doubles every 10 Base Base Base years 1.OOE-09 4.34E-09 8.80E-09 1.59E-07 Base Base 5%

Base 3.85E-10 2.24E-09 6.07E-09 1.58E-07 Base Base 15%

Base 9.62E-10 5.60E-09 1.52E-08 1.64E-07 Base 0.451% Liner, Base Base 0.0451% Basemat 1.75E-10 1.02E-09 2.75E-09 1.56E-07 Base 6.72% Liner, Base Base 0.672% Basemat 2.60E-09 1.51 E-08 4.1OE-08 1.80E-07 Lower Bound Doubles every 10 0.451% Liner, 5%

10%

years 0.0451% Basemat 1.49E-11 6.44E-11 1.30E-10 1.54E-07 Upper Bound Doubles every 2 6.72% Liner, 15%°100%

years 0.672% Basemat 1.06E-09 1.80E-08 1.41 E-07 2.78E-07 17670-0001 PB ILRT Rev 3.doc

.SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 119 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Attachment A to Appendix B Point Beach Risk Impact of Containment Liner Corrosion During an Extension of the ILRT Interval Results 17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 120 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment-for Extending Containment Type A Test Interval Table of Contents A1.0 Introduction 121 A2.0 Sensitivity Case 1 - Flaw Rate Doubles Every 2 Years 122 A3.0 Sensitivity Case 2 - Flaw Rate Doubles Every 10 Years 127 A4.0 Sensitivity Case 3-5%Visual Inspection Failures 132 A5.0 Sensitivity Case 4 - 15%Visual Inspection Failures 137 A6.0 Sensitivity Case 5 - Containment Breach Base Point 10 Times Lower 142 A7.0 Sensitivity Case 6 - Containment Breach Base Point 10 Times Higher 147 A8.0 Sensitivity Case 7-Lower Bound 152 A9.0 Sensitivity Case 8 - Upper Bound 157 17670-0001 PB ILRT Rev 3.doc

.SCIENTECH.

CLIENT: Nuclear Management Company

-BY: E. A. Krantz PAGE: 121 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A1.0 Introduction This attachment presents the results of the Point Beach risk impact of containment liner corrosion during an extension of the ILRT interval. Eight sensitivity cases were examined. These are:

Sensitivity Case 1 - Flaw rate doubles every 2 years Sensitivity Case 2 - Flaw rate doubles every 10 years Sensitivity Case 3 - 5% Visual inspection failures 0

Sensitivity Case 4 - 15% Visual inspection failures Sensitivity Case 5 - Containment breach base point 10 times lower 0

Sensitivity Case 6 - Containment breach base point 10 times higher 0

Sensitivity Case 7 - Flaw rate doubles every 10 years, containment breach base point 10 times lower, 5% visual inspection failures and 10% EPRI accident Class 3b are LERF (Lower bound)

Sensitivity Case 8 - Flaw rate doubles every 2 years, containment breach base point 10 times higher, 15% visual inspection failures and 100% EPRI accident Class 3b are LERF (upper bound)

The EXCEL spreadsheet results are presented in the following sections.

17670-0001 PB ILRT Rev 3.doc

-SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 122 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A2.0 Sensitivity Case 1 - Flaw Rate Doubles Every 2 Years From Estimated Change 1 to 3 years 1 to 10 years 1 to 15-1/2 years Other Assumptions Containment Breach Visual Inspection Failures EPRI Class 3a Fraction EPRI Class 3b Fraction Containment Cylinder/Dome 0.31%

5.17%

40.50%

Basemat 0.05%

0.86%

6.75%

1.740%

10.0%

0.0%

100.0%.

0.174%

100.0%

0.0%

100.0%

17670-0001 PB ILRT Rev 3.doc

S-SCIENTECH.

CLIENT: Nuclear Management Company BY7

E. A. Krantz FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval PAGE: 123 OF 161 Date: 12/06/07 3-in-10 years Increase to 3a Frequency Increase to 3b Frequency

,Class 1

2 3a 3b 4

5 6

7 8a 8b 8c Core Damage Unit 1 0

0.00062%

Unit 2 0

0.00062%

Person-Rem (50-miles) 3.86E+03 1.13E+05 3.86E+04 1.35E+05 0.OOE+00 0.OOE+00 0.OOE+00 1.39E+05 1.13E+06 1.88E+05 1.39E+05 Unit 1 Baseline Frequency 3.50E-05 1.50E-08 9.75E-07 9.77E-08 0.OOE+00 0.OOE+00

0.00E+00 9.38E-06 2.37E-07

.1.83E-06 2.62E-06 5.02E-05 Corrosion Addition 2.24E-10 Unit I Baseline Dose Rate 1.35E-01 1.70E-03 3.76E-02 1.32E-02 0.OOE+00 0.OOE+00 O.OOE+00 1'.30E+00 2.68E-01 3.44E-01 3.64E-01 2.47E+00 5.08E-02 2.06%

9.77E-08 28.25%

Unit 2 Baseline Frequency 3.03E-05 1.55E-08 8.42E-07 8.44E-08 0.OOE+00 0.OOE+00 0.OOE+00 1.58E-05 2.37E-07 1.91 E-06 2.66E-06 5.18E-05 Corrosion Addition 1.93E-10 Unit 2 Baseline Dose Rate 1.17E-01 1.76E-03 3.25E-02 1.14E-02 0.OOE+00 0.OOE+00 0.OOE+00 2.20E+00 2.68E-01 3.59E-01 3.69E-01 3.35E+00 4.39E-02 1.31%

8.44E-08 39.96%

ILRT Dose Rate from 3a and 3b =

% of Total =

LERF from 3b =

CCFP%Liner=

17670-0001 PB ILRT Rev 3.doc

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CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 124 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed Risk impact Assessment for Extending Containment Type A Test Interval 1-in-lO years Unit 1 Unit 2 Increase to 3a Frequency 0

0 Increase to 3b Frequency 0.0105%

0.0105%

Unit I Unit 1 Unit 2 Unit 2 Dose Person-Frequency Dose Rate Frequency Rate (1-in-Rem (50-(1-in-10 (1-in-10 (1-in-10 10 yrs Class miles) yrs ILRT) yrs ILRT) yrs ILRT)

ILRT) 1 3.86E+03 3.25E-05 1.26E-01 2.81E-05 1.08E-01 2

1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 3.25E-06 Addition 1.25E-01 2.81E-06 Addition 1.08E-01 3b 1.35E+05 3.29E-07 3.79E-09 4.44E-02 2.84E-07 3.27E-09 3.83E-02 4

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 5

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 6

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.58E+00 5.18E-05 3.45E+00 ILRT Dose Rate from 3a and 3b =

1.70E-01 1.47E-01

% of Total 6.59%

4.25%

LERF from 3b =

3.29E-07 2.84E-07 CCFP%Liner=

28.71%

40.34%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company 7

BY: E. A. Krantz PAGE: 125 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07 SU7BJECT: Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-15-1/2 years Unit 1 Unit 2 Increase to 3a Frequency 0

0 Increase to 3b Frequency 0.0822%

0.0822%

Unit 1 Unit 1 Unit 2 Frequency Dose Rate Frequency Unit 2 Dose Person-(1-in (1-in (1-in Rate (1-in-Rem (50-1/2 yrs 1/2 yrs 1/2 yrs 15-1/2 yrs Class miles)

ILRT)

ILRT) 1 3.86E+03 3.05E-05 1.18E-01 2.64E-05 1.02E-01 2

1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 5.04E-06 Addition 1.94E-01 4.35E-06 Addition 1.68E-01 3b 1.35E+05 5.33E-07 2.97E-08 7.20E-02 4.61 E-07 2.56E-08 6.22E-02 4

0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 5

O.OOE+00 O.OOE+O0 O.OOE+00 0*OOE+00 O.OOE+00 6

O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05

.2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.67E+00 5.18E-05 3.52E+00 ILRT Dose Rate from 3a and 3b =

2.66E-01 2.30E-01

% of Total =

9.99%

6.53%

LERF from 3b =

5.33E-07 4.61 E-07 CCFP%Liner=

29.11%

40.68%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 126 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Other Pertinent Risk Metrics Unit 1 Unit 2 10 to 15-1/2 Increase (Person-rem/ry) 9.66E-02 8.34E-02 3 to 15-1/2 Increase (Person-rem/ry) 2.16E-01 1.86E-01 10 to 15-1/2 Delta-LERF 2.05E-07 1.77E-07 3 to 15-1/2 Delta-LERF 4.36E-07 3.76E-07 10 to '15-1/2 Delta-CCFP 0.4%

0.3%

3 to 15-1/2 Delta-CCFP 0.9%

0.7%

3 to 10 Delta-LERF from Corrosion 3.56E-09 3.08E-09 10 to 15-1/2 Delta-LERF from Corrosion 2.59E-08 2.24E-08 Increase in LERF (ILRT 3-to-1 5-1/2 years) from Corrosion 2.94E-08 2.54E-08 17670-0001 PB ILRT Rev 3.doc

S'SCIENTECH.

CLIENT: Nuclear Management Company 7

7BY: E. A. Krantz PAGE: 127 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A3.0 Sensitivity Case 2 - Flaw Rate Doubles Every 10 Years From Estimated Change 1 to 3 years 1 to 10 years I to 15-1/2 years Other Assumptions Containment Breach Visual Inspection Failures EPRI Class 3a Fraction EPRI Class 3b Fraction Containment Cylinder/Dome 1.58%

6.86%

13.91%

Basemat 0.27%

1.15%

2.33%

1.740%

10.0%

0.0%

100.0%

0.174%

100.0%

0.0%

100.0%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company

] BY: E. A. Krantz PAGE: 128 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 3-in-10 years Unit 1 Unit 2 Increase to 3a Frequency Increase to 3b Frequency 0

0 0.00322%

0.00322%

Class 1

2 3a 3b 4

5 6

7 8a 8b 8c Core Damage Person-Rem (50-miles)

.3.86E+03 1.13E+05 3.86E+04 1.35E+05 0.OOE+00 0.OOE+00 O.OOE+00 1.39E+05 1.13E+06 1.88E+05 1.39E+05 Unit 1 Baseline Frequency 3.50E-05 1.50E-08 9.75Eý07 9.86E-08 0.OOE+00 0.00E+00 0.OOE+00 9.38E-06 2.37E-07 1.83E-06 2.62E-06 5.02E-05 Corrosion Addition 1.16E-09 Unit 1 Baseline Dose Rate 1.35E-01 1.70E-03 3.76E-02 1,33E-02 0.OOE+00 0.OOE+00 O.OOE+00 1.30E+00 2,68E-01 3.44E-01 3.64E-01 2.47E+00 5.09E-02 2.06%

9.86E-08 28.25%

Unit 2 Baseline Frequency 3.03E-05 1.55E-08 8.42E-07 8.52E-08 0.OOE+00 0.OOE+00 0.OOE+00 1.58E-05 2.37E-07 1.91 E-06 2.66E-06 5.18E-05 Corrosion Addition 1.OOE-09 3.25E-02 1.15E-02 0.OOE+00 0.OOE+00 0.OOE+00 2.20E+00 2.68E-01 3.59E-01 3.69E-01 3.35E+00 4.40E-02 1.31%

8.52E-08 39.96%

Unit 2 Baseline Dose Rate 1.17E-01 1.76E-03 ILRT Dose Rate from 3a and 3b

% of Total =

LERF from 3b CCFP.%Liner=

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 129 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-lO years Unit 1 Unit 2 Increase to 3a Frequency 0

.0 Increase to 3b Frequency 0.0139%

0.0139%

Unit 1 Unit 1 Unit 2 Unit 2 Dose Person-Frequency Dose Rate Frequency Rate (1-in-Rem (50-(1-in-10 (1-in-10 (1-in-lO 10 yrs Class miles) yrs ILRt) yrs ILRT) yrs ILRT)

ILRT) 1 3.86E+03 3.25E-05 1.26E-01 2.81 E-05 1.08E-01 2

1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 3.25E-06 Addition 1.25E-01 2.81E-06 Addition 1.08E-01 3b 1.35E+05 3.30E-07 5.03E-09 4.46E-02 2.85E-07 4.34E-09 3.85E-02 4

0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 5

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 6

0.OOE+00 O.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.58E+00 5.18E-05 3.45E+00 ILRT Dose Rate from 3a and 3b =

1.70E-01 1.47E-01

% of Total =

6.60%

4.26%

LERF from 3b =

3.30E-07 2.85E-07 CCFP%Liner=

28.71%

40.34%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 130 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-15-1/2 years Unit 1 Unit 2 Increase to 3a Frequency 0

0 Increase to 3b Frequency 0.0282%

0.0282%

Unit 1 Unit 1 Unit 2 Frequency Dose Rate Frequency Unit 2 Dose Person-(1-in-1 5-(1 -in-1 5-(1 -in-1 5-Rate (1-in-Rem (50-1/2 yrs 1/2 yrs 1/2 yrs 15-1/2 yrs Class miles)

ILRT)

ILRT) 1 3.86E+03 3.06E-05 1.18E-01 2.64E-05 1.02E-01 2

1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 5.04E706 Addition 1.94E-01 4.35E-06 Addition 1.68E-01 3b 1.35E+05 5.14E-07 1.02E-08 6.94E-02 4.44E-07 8.80E-09 5.99E-02 4

10.00E+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 5

0.00E+00 O.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 6

0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.00E+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.66E+00 5.18E-05 3.52E+00 ILRT Dose Rate from 3a and 3b 2.64E-01 2.28E-01

% of Total =

9.91%

6.47%

LERF from 3b=

5.14E-07 4.44E-07 CCFP%Liner=

29.08%

40.65%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz I PAGE: 131 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindredj Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Other Pertinent Risk Metrics 10 to 15-1/2 Increase (Person-rem/ry) 3 to 15-1/2 Increase (Person-rem/ry) 10 to 15-1/2 Delta-LERF 3 to 15-1/2 Delta-LERF 10 to 15-1/2 Delta-CCFP 3 to 15-1/2 Delta-CCFP Unit I 9.38E-02 2.13E-01 1.84E-07 4.15E-07 0.4%

0.8%

Unit 2 8.1OE-02 1.84E-01 1.59E-07 3.59E-07 0.3%

0.7%

3.34E-09 4.46E-09 7.80E-09 3 to 10 Delta-LERF from Corrosion 10 to 15-1/2 Delta-LERF from Corrosion Increase in LERF (ILRT 3-to-1 5-1/2 years) from Corrosion 3.87E-09 5.16E-09 9.03E-09 17670-0001 PB ILRT Rev 3.doc

.SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 132 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A4.0 Sensitivity Case 3 - 5%Visual Inspection Failures From Estimated Change 1 to 3 years 1 to 10 years 1 to 15-1/2 years Other Assumptions Containment Breach Visual Inspection Failures EPRI Class 3a Fraction EPRI Class 3b Fraction Containment Cylinder/Dome 1.06%

6.20%

16.78%

Basemat 0.18%

1.04%

2.80%

1.740%

5.0%

0.0%

100.0%

0.174%

100.0%

0.0%

100.0%

17670-0001 PB ILRT Rev 3.doc

OSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 133 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 3-in-10 years Unit 1 Unit 2 Increase to 3a Frequency Increase to 3b Frequency 0

0 0.00123%

0.00123%

Class 1

2 3a 3b 4

5 6

7 8a 8b 8c Core Damage Person-Rem (50-miles) 3.86E+03 1.13E+05 3.86E+04 1.35E+05 0.OOE+00 0.OOE+00 0.OOE+00 1.39E+05 1.13E+06 1.88E+05 1.39E+05 Unit 1 Baseline Frequency 3.50E-05 1.50E-08 9.75E-07 9.79E-08 0.OOE+00 0.OOE+00 0.OOE+00 9.38E-06 2.37E-07 1.83E-06 2.62E-06 5.02E-05 Corrosion Addition 4.46 E-10 Unit 1 Baseline Dose Rate 1.35E-01 1.70E-03 3.76E-02 1.32E-02 0.OOE+00 O.OOE+00 0.OOE+00 1.30E+00 2.68E-01 3.44E-01 3.64E-01 2.47E+00 5.08E-02 2.06%

9.79E-08 28.25%

Unit 2 Baseline Frequency 3.03E-05 1.55E-08 8.42E-07 8.46E-08 0.OOE+00 0.OOE+00 0.OOE+00 1.58E-05 2.37E-07 1.91 E-06 2.66E-06 5.18E-05 Corrosion Addition 3.85E-10 Unit 2 Baseline Dose Rate 1.17E-01 1.76E-03 3.25E-02 1.14E-02 0.OOE+00 0.00E+00 0.OOE+00 2.20E+00 2.68E-01 3.59E-01 3.69E-01 3.35E+00 ILRT Dose Rate from 3a and 3b

% of Total =

LERF from 3b =

CCFP%Liner=

4.39E-02 1.31%

8.46E-08 39.96%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 134 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-lO years Unit 1 Unit 2 Increase to 3a Frequency Increase to 3b Frequency

0.

0 0.0072%

0.0072%

Class 1

2 3a 3b 4

5 6

7 8a 8b 8c Core Damage Person-Rem (50-miles) 3.86E+03 1.13E+05 3.86E+04 1.35E+05 0.OOE+00 0.OOE+00 0.OOE+00 1.39E+05 1.13E+06 Unit 1 Frequency (1-in-1 0 yrs ILRT) 3.25E-05 1.15E-08 3.25E-06 3.27E-07 0.OOE+00 0.OOE+00 0.OOE+00 9.38.E-06 2_37E-07 Corrosion Addition 2.60E-09 Unit 1 Dose Rate (1-in-10 yrs ILRT) 1.26E-01 1.30E-03 1.25E-01 4.42E-02 0.OOE+00 0.OOE+00 0.OOE+00 1.30E+00 2.68E-01 3.44E-01 3.64E-01 2.58E+00 1.70E-01 6.58%

3.27E-07 28.70%

Unit 2 Frequency (1-in-10 yrs ILRT) 2.81 E-05 1.19E-08 2.81 E-06 2.83E-07 0.OOE+00 0.00E+00 0.OOE+00 1.58E-05 2.37E-07 1.91 E-06 2.66E-06 5.18E-05 Corrosion Addition 2.24E-09 Unit 2 Dose Rate (1-in-10 yrs ILRT) 1.08E-01 1.35E-03 1.08E-01 3.82E-02 0.OOE+00 0.OOE+00 0.OOE+00 2.20E+00 2.68E-01 3.59E-01 3.69E-01 3.45E+00 1.88E+05 1.83E-06 1.39E+05 2.62E-06 5.02E-05 ILRT Dose Rate from 3a and 3b =

% of Total =

LERF from 3b CCFP%Liner=

1.47E-01 4.25%

2.83E-07 40.34%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 135 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1 -in-1 5-1/2 years Unit 1 Unit 2 Increase to 3a Frequency 0

0 Increase to 3b Frequency 0.0195%

0.0195%

Unit 1 Unit 1 Unit 2 Frequency Dose Rate Frequency Unit 2 Dose Person-(1-in (1-in -

-(1-in Rate (1-in-Rem (50-1/2 yrs 1/2 yrs 1/2 yrs 15-1/2 yrs Class miles)

ILRT)

ILRT) 1 3.86E+03 3.06E-05 1.18E-01 2.64E-05 1.02E-01 2

1.13E+05

-1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 5.04E-06 Addition 1.94E-01 4.35E-06 Addition 1.68E-01 3b 1.35E+05 5.11 E-07 7.03E-09 6.90E-02 4.41 E-07 6.07E-09 5.96E-02 4

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 5

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 6

0.OOE+00 0,00E+00 0.OOE+00 0.OOE+00 0.OOE+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E705 2.66E+00 5.18E-05 3.52E+00 ILRT Dose Rate from 3a and 3b =

2.63E-01 2.27E-01

% of Total =

9.89%

6.46%

LERF from 3b =

5.11E-07 4.41 E-07 CCFP%Liner=

29.07%

40.64%

17670-0001 PB ILRT Rev 3.doc

S(.SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 136 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Other Pertinent Risk Metrics 10 to 15-1/2 Increase (Person-rem/ry) 3 to 15-1/2 Increase (Person-rem/ry) 10 to 15-1/2 Delta-LERF 3 to 15-1/2 Delta-LERF 10 to 15-1/2 Delta-CCFP 3 to 15-1/2 Delta-CCFP Unit 1 9.37E-02 2.12E-01 1.83E-07 4.13E-07 0.4%

0.8%

Unit 2 8.09E-02 1.84E-01 1.58E-07 3.56E-07 0.3%

0.7%

1.86E-09 3.83E-09 5.68E-09 3 to 10 Delta-LERF from Corrosion 10 to 15-1/2 Delta-LERF from Corrosion Increase in LERF (ILRT 3-to-1 5-1/2 years) from Corrosion 2.15E-09 4.43E-09 6.58E-09 17670-0001 PB ILRT Rev 3.doc

S(SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 137 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A5.0 Sensitivity Case 4 - 15%Visual Inspection Failures From Estimated Change 1 to 3 years I to 10- years 1 to 15-1/2 years Other Assumptions Containment Breach Visual Inspection Failures EPRI Class 3a Fraction EPRI Class 3b Fraction Containment Cylinder/Dome 1.06%

6.20%

16.78%

Basemat 0.18%

1.04%

2.80%

1.740%

15.0%

0.0%

100.0%

0.174%

100.0%

0.0%

100.0%

17670-0001 PB ILRT Rev 3.doc

.SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 138 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 3-in-10 years Unit 1 Unit 2 Increase to 3a Frequency 0

0 Increase to 3b Frequency 0.00309%

0.00309%

Person-Unit I Unit 1 Unit 2 Unit 2 Rem (50-Baseline Baseline Baseline Baseline Class miles)

Frequency Dose Rate Frequency Dose Rate 1

3.86E+03 3.50E-05 1.35E-01 3.03E-05 1.17E-01 2

1.13E+05 1.50E-08 1.70E-03 1.55E-08.

1.76E-03 Corrosion Corrosion 3a 3.86E+04 9.75E-07 Addition 3.76E-02 8.42E-07 Addition 3.25E-02 3b 1.35E+05 9.86E-08 1.11E-09 1.33E-02 8.51E-08 9.62E-10 1.15E-02 4

O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 5

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 6

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.47E+00 5.18E-05 3.35E+00 ILRT Dose Rate from 3a and 3b 5.09E-02 4.40E-02

% of Total =

2.06%

1.31%

LERF from 3b =

9.86E-08 8.51 E-08 CCFP%Liner=

28.25%

39.96%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 139 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-lO years Unit 1 Unit 2 Increase to 3a Frequency Increase to 3b Frequency 0

0 0.0180%

0.0180%

Class 1

2 3a 3b 4

5 6

7 8a 8b 8c Core Damage Person-Rem (50-miles) 3.86E+03 1.13E+05 3.86E+04 1.35E+05 0.OOE+00 0.OOE+00 0.OOE+00 1.39E+05 1.13E+06 1.88E+05 1.39E+05 Unit I Frequency (1-in-1 0 yrs ILRT) 3.25E-05 1.15E-08 3.25E-06 3.31 E-07 0.OOE+00 0.OOE+00 0.OOE+00 9.38E-06 2.37E-07 1.83E-06 2.62E-06 5.02E-05 Corrosion Addition 6.49E-09 Unit 1 Dose Rate (1-in-10 yrs ILRT) 1.26E-01 1.30E-03 1.25E-01 4.48E-02 0.OOE+00 0.OOE+00 0.OOE+00 1.30E+00 2.68E-01 3.44E-01 3.64E-01 2.58E+00 1.70E-01 6.60%

3.31 E-07 28.71%

Unit 2 Frequency (1-in-1 0 yrs ILRT) 2.81 E-05 1.19E-08 2.81 E-06 2.86E-07 0.OOE+00 0.OOE+00 0.OOE+00 1.58E-05 2.37E-07 1.91 E-06 2.66E-06 5.18E-05 Corrosion Addition 5.60E-09 Unit 2 Dose Rate (1-in-10 yrs ILRT) 1.08E-01 1.35E-03 1.08E-01 3.87E-02 0.OOE+00 0.OOE+00 0.OOE+00 2.20E+00 2.68E-01 3.59E-01 3.69E-0.1 3.45E+00 ILRT Dose Rate from 3a and 3b =

% of Total =

LERF from 3b =

CCFP%Liner=

1.47E-01 4.26%

2.86E-07 40.34%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 140 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-15-1/2 years Unit 1 Unit 2 Increase to 3a Frequency 0

0 Increase to 3b Frequency 0.0486%

0.0486%

Unit 1.

Unit I Unit 2 Frequency Dose Rate Frequency Unit 2 Dose Person-(1-in (1-in (1-in Rate (1-in-Rem (50-1/2 yrs 1/2 yrs 1/2 yrs 15-1/2 yrs Class miles)

ILRT)

ILRT) 1 3.86E+03

.3.06E-05 1.18E-01 2.64E-05 1.02E-01 2

1.1,3E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 5.04E-06 Addition 1.94E-01 4.35E-06 Addition 1.68E-01 3b 1.35E+05 5.21 E-07 1.76E-08 7.04E-02 4.50E-07 1.52E-08 6.08E-02 4

0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 5

0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 6

0.00E+00 0.OOE+00 0.00E+00 0.OOE+00 0.00E+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E=06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.66E+00 5.18E-05 3.52E+00 ILRT Dose Rate from 3a and 3b =

2.65E-01 2.29E-01

% of Total =

9.94%

6.49%

LERF from 3b =

5.21 E-07 4.50E-07 CCFP%Liner=

29.09%

40.66%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 141 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Other Pertinent Risk Metrics 10 to 15-1/2 Increase (Person-rem/ry) 3 to 15-1/2 Increase (Person-rem/ry) 10 to 15-1/2 Delta-LERF 3 to 15-1/2 Delta-LERF 10 to 15-1/2 Delta-CCFP 3 to 15-1/2 Delta-CCFP Unit I 9.46E-02 2.14E-01 1.90E-07 4.23E-07 0.4%

0.8%

Unit 2 8.17E-02 1.85E-01 1.64E-07 3.65E-07 0.3%

0.7%

4.64E-09 9.56E-09 1.42E-08 3 to 10 Delta-LERF from Corrosion 10 to 15-1/2 Delta-LERF from Corrosion Increase in LERF (ILRT 3-to-1 5-1/2 years) from Corrosion 5.37E-09 1.11E-08 1.64E-08 17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company 77BY: E. A. Krantz PAGE: 142 OF 161 FILE NO. 17670-0001, Rev. 3I CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A6.0 Sensitivity Case 5 - Containment Breach Base Point 10 Times Lower From Estimated Change 1 to 3 years I to 10 years 1 to 15-1/2 years Other Assumptions Containment Breach Visual Inspection Failures EPRI Class 3a Fraction EPRI Class 3b Fraction Containment Cylinder/Dome 1.06%

6.20%

16.78%

Basemat 0.18%

1.04%

2.80%

0.451%

10.0%

0.0%

100.0%

0.045%

100.0%

0.0%

100.0%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY E. A. Krantz PAGE: 143 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 3-in-10 years Unit I Unit 2 Increase to 3a Frequency 0

0 Increase to 3b Frequency 0.00056%

0.00056,%

Person-Unit I Unit 1 Unit 2 Unit 2 Rem (50-Baseline Baseline Baseline Baseline Class miles)

Frequency Dose Rate Frequency Dose Rate 1

3.86E+03 3.50E-05 1.35E-01 3.03E-05 1.17E-01 2

1.13E+05 1.50E-08 1.70E-03 1.55E-08 1.76E-03 Corrosion Corrosion 3a 3.86E+04 9.75E-07 Addition 3.76E-02 8.42E-07 Addition 3.25E-02 3b 1.35E+05 9.77E-08, 2.02E-10 1.32E-02 8.43E-08 1.75E-10 1.14E-02 4

0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 5

0.OOE+00 0.OOE+00 0.00E+-00 0.OOE+00 0.OOE+00 6

0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 7

1.39E+05 9.38 E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07.

2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.47E+00 5.18E-05 3.35E+00 ILRT Dose Rate from 3a and 3b 5.08E-02 4.39E-02

% of Total 2.06%

1.31%

LERF from 3b =

9.77E-08 8.43E-08 CCFP%Liner=

28.25%

39.96%

17670-0001 PB ILRT Rev 3.doc

S'SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 144 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-lO years Increase to 3a Frequency Increase to 3b Frequency Unit 1 0

0.0033%

Class 1

2 3a 3b 4

5 6

7 8a 8b 8c Core Damage ILRT Dose Rate from 3a and 3b =

% of Total =

LERF from 3b CCFP%Liner=

Person-Rem (50-miles) 3.86E+03 1.13E+05 3.86E+04 1.35E+05 0.OOE+00 0.OOE+00 0.OOE+00 1.39E+05 1.13E+06 1.88E+05 1.39E+05 Unit 2 o

0.0033%

Unit 1 Frequency (1-in-10 yrs ILRT) 3.25E-05 1.15E-08 3.25E-06 3.26E-07 O.OOE+0O 0.OOE+00 O.OOE+00 9.38E-06 2.37E-07 1.83E-06 2.62E-06 5.02E-05 Corrosion Addition 1.18E-09 Unit 1 Dose Rate (1-in-10 yrs ILRT) 1.26E-01 1-.30E-03 1.25E-01 4.40E-02 0.OOE+00 0.OOE+00 0.OOE+00 1.30E+00 2.68E-01 3.44E-01 3.64E-01 2.58E+00 1.69E-01 6.58%

3.26E-07 28.70%

Unit 2 Frequency (1-in-10 yrs ILRT) 2.81 E-05 1.19E-08 2.81 E-06 2.82E-07 0.OOE+00 0.OOE+00 0.OOE+00 1.58E-05 2.37E-07 1.91 E-06 2.66E-06 5.18E-05 Corrosion Addition 1.02E-09 Unit 2 Dose Rate (1-in-10 yrs ILRT) 1.08E-01 1.35E-03 1.08E-01 3.80E-02 0.OOE+00 0.OOE+00 0.OOE+00 2.20E+00 2.68E-01 3.59E-01 3.69E-01 3.45E+00 1.46E-01 4.24%

2.82E-07 40.33%

17670-0001 PB ILRT Rev 3.doe

.SCIENTECH.

CLIENT: Nuclear Management Company 1 BY: E. A. Krantz PAGE: 145 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1 -in-1 5-1/2 years Unit I Unit 2 Increase to 3a Frequency 0

0 Increase to 3b Frequency 0.0088%

0.0088%

Unit 1 Unit 1 Unit 2 Frequency Dose Rate Frequency Unit 2 Dose Person-(1-in (1-in (1-in Rate (1-in-Rem (50-1/2 yrs 1/2 yrs 1/2 yrs 15-1/2 yrs Class miles)

ILRT)

ILRT) 1 3.86E+03 3.06E-05 1.18E-01 2.64E-05 1.02E-01 2

1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 5.04E-06 Addition 1.94E-01 4.35E-06 Addition 1.68E-01 3b 1.35E+05 5.07E-07 3.19E-09 6.85E-02 4.38E-07 2.75E-09 5.91E-02 4

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 5

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 6

0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.66E+00 5.18E-05 3.52E+00 ILRT Dose Rate from 3a and 3b =

2.63E-01 2.27E-01

% of Total =

9.87%

6.45%

LERF from 3b =

5.07E-07 4.38E-07 CCFP%Liner=

29.06%

40.63%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 146 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Other Pertinent Risk Metrics 10 to 15-1/2 Increase (Person-rem/ry) 3 to 15-1/2 Increase (Person-rem/ry) 10 to 15-1/2 Delta-LERF 3 to 15-1/2 Delta-LERF 10 to 15-1/2 Delta-CCFP 3 to 15-1/2 Delta-CCFP Unit 1 9.34E-02 2.12E-01 1.81 E-07 4.09E-07 0.4%

0.8%

Unit 2 8.06E-02 1.83E-01 1.56E-07 3.53E-07 0.3%

0.7%

8.42E-10 1.74E-09 2.58E-09 3 to 10 Delta-LERF from Corrosion 10 to 15-1/2 Delta-LERF from Corrosion Increase in LERF (ILRT 3-to-15-1/2 years) from Corrosion 9.75E-1 0 2.01 E-09 2.98E-09 17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz 7

PAGE: 147 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A7.0 Sensitivity Case 6 - Containment Breach Base Point 10 Times Higher From Estimated Change 1 to 3 years 1 to 10 years 1 to 15-1/2 years Other Assumptions Containment Breach Visual Inspection Failures EPRI Class 3a Fraction EPRI Class 3b Fraction Containment Cylinder/Dome 1.06%

6.20%

16.78%

Basemat 0.18%

1.04%

2.80%

6.715%

10.0%

0.0%

100.0%

0.671%

100.0%

0.0%

100.0%

17670-0001 PB ILRT Rev 3.doc

SC SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 148 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 3-in-10 years Unit 1 Unit 2 Increase to 3a Frequency 0

0 Increase to 3b Frequency 0.00834%

0.00834%

Person-Unit I Unit I Unit 2 Unit 2 Rem (50-Baseline Baseline Baseline Baseline Class miles)

Frequency Dose Rate Frequency Dose Rate 1

3.86E+03 3.50E-05 1.35E-01 3.03E-05 1.17E-01 2

1.13E+05 1.50E-08 1.70E-03 1.55E-08 1.76E-03 Corrosion Corrosion 3a 3.86E+04 9.75E-07 Addition 3.76E-02 8.42E-07 Addition 3.25E-02 3b 1.35E+05 1.OOE-07 3.01 E-09 1.36E-02 8.68E-08 2.60E-09 1.17E-02 4

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 5

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 6

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.47E+00 5.18E-05 3.35E+00 ILRT Dose Rate from 3a and 3b 5.12E-02 4.42E-02

% of Total 2.07%

1.32%

LERF from 3b =

1.OOE-07 8.68E-08 CCFP%Liner=

28.26%

39.96%

17670-0001 PB ILRT Rev 3.doc

S'SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 149 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1 -in-lO years Unit 1 Unit 2 Increase to 3a Frequency 0

0 Increase to 3b Frequency 0.0486%

0.0486%

Unit I Unit 1 Unit 2 Unit 2 Dose Person-Frequency Dose Rate Frequency Rate (1-in-Rem (50-(1-in-10 (1-in-10 (1-in-10 10 yrs Class miles) yrs ILRT) yrs ILRT) yrs ILRT)

ILRT) 1 3.86E+03 3.25E-05 1.26E-01 2.81 E-05 1.08E-01 2

1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 3.25E-06 Addition 1.25E-01 2.81E-06 Addition 1.08E-01 3b 1.35E+05 3.42E-07 1.75E-08 4.63E-02 2.96E-07 1.51 E-08 4.OOE-02 4

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 5

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 6

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.58E+00 5.18E-05 3.45E+00 ILRT Dose Rate from 3a and 3b =

1.72E-01 1.48E-01

% of Total =

6.66%

4.30%

LERF from 3b 3.42E-07 2.96E-07 CCFP%Liner=

28.73%

40.36%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 150 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1 -in-15-112 years Unit 1 Unit 2 Increase to 3a Frequency 0

0 Increase to 3b Frequency 0.1315%

0.1315%

Unit 1 Unit 1 Unit 2 Frequency Dose Rate Frequency Unit 2 Dose Person-(1-in (1-in (1-in Rate (1-in-Rem (50-1/2 yrs 1/2 yrs 1/2 yrs 15-1/2 yrs Class miles)

ILRT)

ILRT) 1 3.86E+03 3.05E-05 1.18E-01 2.64E-05 1.02E-01 2

1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 5.04E-06 Addition 1.94E-01 4.35E-06 Addition 1.68E-01 3b 1.35E+05 5.51E-07.

4.75E-08 7.44E-02 4.76E-07 4.1OE-08 6.43E-02 4

O.OOE+00 0.OOE+00 O.00E+00 0.OOE+00 0.OOE+00 5

0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6

0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.67E+00 5.18E-05 3.53E+00 ILRT Dose Rate from 3a and 3b =

2.69E-01 2.32E-01

% of Total =

10.08%

6.58%

LERF from 3b =

5.51 E-07 4.76E-07 CCFP%Liner=

29.15%

40.71%

17670-0001 PB ILRT Rev 3.doc

S'SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 151 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Other Pertinent Risk Metrics 10 to 15-1/2 Increase (Person-rem/ry) 3 to 15-1/2 Increase (Person-rem/ry) 10 to 15-1/2 Delta-LERF 3 to 15-1/2 Delta-LERF 10 to 15-1/2 Delta-CCFP 3 to 15-1/2 Delta-CCFP Unit 1 9.71 E-02 2.18E-01 2.09E-07 4.51 E-07 0.4%

0.9%

Unit 2 8.39E-02 1.88E-01 1.80E-07 3.89E-07 0.3%

0.7%

1.25E-08 2.58E-08 3.84E-08 3 to 10 Delta-LERF from Corrosion 10 to 15-1/2 Delta-LERF from Corrosion Increase in LERF (ILRT 3-to-15-1/2 years) from Corrosion 1.45E-08 2.99E-08 4.44E-08 17670-0001 PB ILRT Rev 3.doe

SSCIENTECH.

CLIENT: Nuclear Management Company

-BY:

E. A. Krantz PAGE: 152 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A8.0 Sensitivity Case 7 - Lower Bound (Flaw rate doubles every 10 years, containment breach base point 10 times lower, 5%visual inspection failures and.10% EPRI accident Class 3b are LERF)

From Estimated Change 1 to 3 years 1 to 10 years 1 to 15-1/2 years Other Assumptions Containment Breach Visual Inspection Failures EPRI Class 3a Fraction EPRI Class 3b Fraction Containment Cylinder/Dome 1.58%

6.86%

13.91%

Basemat 0.27%

1.15%

2.33%

0.451%.

5.0%

90.0%

10.0%

0.045%

100.0%

90.0%

10.0%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 153 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 3-in-10 years Unit 1 Unit 2 Increase to 3a Frequency Increase to 3b Frequency 0.00043%

0.00043%

0.00005%

Class 1

2 3a 3b 4

5 6

7 8a 8b 8c Core Damage Person-Rem (50-miles) 3.86E+03 1.13E+05 3.86E+04 1.35E+05 0.OOE+00 0.OOE+00 O.OOE+00 1.39E+05 1.13E+06 1.88E+05 1.39E+05 Unit I Baseline Frequency 3.50E-05 1.50E-08 9.75E-07 9.75E-08 0.OOE+00 0.OOE+00 0.OOE+00 9.38E-06 2.37E-07 1.83E-06 2.62E-06 5.02E-05 Corrosion Addition 1.72E-11 Unit 1 Baseline Dose Rate 1.35E-01 1.70E-03 3.76E-02 1.32E-02 0.OOE+00 0.OOE+00 0.O0E+00 1.30E+00 2.68E-01 3.44E-01 3.64E-01 2.47E+00 5.08E-02 2.06%

9.75E-08 28.25%

Unit 2 Baseline Frequency 3.03E-05 1.55E-08 8.42E-07 8.42E-08 0.00E+00 0.OOE+00 0.OOE+00 1.58E-05 2.37E-07 1.91 E-06 2.66E-06 5.18E-05 Corrosion Addition 1.49E-11 Unit 2 Baseline Dose Rate 1.17E-01 1.76E-03 3.25E-02 1.14E-02 0.OOE+00 0.OOE+00 0.OOE+00 2.20E+00 2.68E-01 3.59E-01 3.69E-01 3.35E+00 ILRT Dose Rate from 3a and 3b

% of Total =

LERF from 3b =

CCFP%Liner=

4.39E-02 1.31%

8.42E-08 39.96%

17670-0001 PB ILRT Rev 3.doe

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 154 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1 -in-lO years Unit 1 Unit 2 Increase to 3a Frequency 0.00186%

0.00186%

Increase to 3b Frequency 0.0002%

0.0002%

Unit 1 Unit 1 Unit 2 Unit 2 Dose Person-Frequency Dose Rate Frequency Rate (1-in-Rem (50-(1-in-10 (1-in-10 (1-in-10 10 yrs Class miles) yrs ILRT) yrs ILRT) yrs ILRT)

ILRT)

I 3.86E+03 3.25E-05 1.26E-01 2.81 E-05 1.08E-01 2

1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 3.25E-06 Addition 1.25E-01 2.81E-06 Addition 1.08E-01 3b 1.35E+05 3.25E-07 7.45E-1I 4.39E-02 2.81 E-07 6.44E-11 3.79E-02 4

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 5

0.OOE+00 0.OOE+0o 0.OOE+00 0.OOE+00 0.OOE+00 6

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.58E+00 5.18E-05 3.45E+00 ILRT Dose Rate from 3a and 3b 1.69E-01 1.46E-01

% of Total =

6.57%

4.24%

LERF from 3b =

3.25E-07 2.81 E-07 CCFP%Liner=

28.70%

40.33%

17670-0001 PB ILRT Rev 3.doe

S'SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 155 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-15-1/2 years Unit I Unit 2 Increase to 3a Frequency 0.00377%

0.00377%

Increase to 3b Frequency 0.0004%

0.0004%

Unit 1 Unit 1 Unit 2 Frequency Dose Rate Frequency Unit 2 Dose Person-(1-in (1-in (1*in Rate (1-in-Rem (50-1/2 yrs 1/2 yrs 1/2 yrs 15-1/2 yrs Class miles)

ILRT)

ILRT) 1 3.86E+03 3.06E-05 1.18E-01 2.64E-05 1.02E-01 2

1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 5.04E-06 Addition 1.94E-01 4.35E-06 Addition 1.68E-01 3b 1.35E+05 5.04E-07 1.51 E-10 6.80E-02 4.35E-07 1.30E-10 5.88E-02 4

0.00E+00 0.00E+00 0.00E+00 O.00E+00 0.OOE+00 5

0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 6

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.66E+00 5.18E-05 3.52E+00 ILRT Dose Rate from 3a and 3b 2.62E-01 2.27E-01

% of Total 9.86%

6.44%

LERF from 3b =

5.04E-07 4.35E-07 CCFP%Liner=

29.06%

40.63%

17670-0001 PB ILRT Rev 3.doc

.SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 156 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Other Pertinent Risk Metrics 10 to 15-1/2 Increase (Person-rem/ry) 3 to 15-1/2 Increase (Person-rem/ry) 10 to 15-1/2 Delta-LERF 3 to 15-1/2 Delta-LERF 10 to 15-1/2 Delta-CCFP 3 to 15-1/2 Delta-CCFP Unit 1 9.31 E-02 2.12E-01 1.79E-07 4.06E-07 0.4%

0.8%

Unit 2 8.04E-02 1.83E-01 1.54E-07 3.51 E-07 0.3%

0.7%

4.95E-1 1 6.61 E-1I 1.16E-10 3 to 10 Delta-LERF from Corrosion 10 to 15-1/2 Delta-LERF from Corrosion Increase in LERF (ILRT 3-to-15-1/2 years) from Corrosion 5.73E-1 1 7.65E-1 1 1.34E-10 17670-0001 PB ILRT Rev 3.doc

S( SCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz I PAGE: 157 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A9.0 Sensitivity Case 8 - Upper Bound (Flaw rate doubles every 2 years, containment breach base point 10 times higher, 15%visual inspection failures and 100% EPRI accident Class 3b are LERF)

From Estimated Change 1 to 3 years 1 to 10 years 1 to 15-1/2 years Other Assumptions Containment Breach Visual Inspection Failures EPRI Class 3a Fraction EPRI Class 3b Fraction Containment Cylinder/Dome 0.31%

5.17%

40.50%

Basemat 0.05%

0.86%

6.75%

6.715%

15.0%

0.0%

100.0%

0.671%

100.0%

0.0%

100.0%

17670-0001 PB ILRT Rev 3.doe

SSCIENTECH.

CLIENT: Nuclear Management Company

[ BY: E. A. Krantz 77PAGE: 158 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 3-in-10 years Increase to 3a Frequency Increase to 3b Frequency Class 1

2 Unit 1 0

0.00342%

Person-Rem (50-miles) 3.86E+03 1.13E+05 3.86E+04 1.35E+05 0.OOE+00 0.OOE+00 0.OOE+00 1.39E+05 1.13E+06 1.88E+05 1.39E+05 Unit 2 0

0.00342%

Unit 1 Baseline.

Frequency 3.50E-05 1.50E-08 9.75E-07 9.87E-08 0.OOE+00 O.OOE+00 O.OOE+O0 9.38E-06 2.37E.07 1.83E-06 2.62E-06 5.02E-05 3a 3b 4

5 6

7 8a 8b 8c Core Damage ILRT Dose Rate from 3a and 3b

% of Total =

LERF from 3b CCFP%Liner=

Corrosion Addition 1.23E-09 Unit 1 Baseline Dose Rate 1.35E-01 1.70E-03 3.76E-02 1.33E-02 0.OOE+00 0.OOE+00 0.OOE+00 1.30E+00 2.68E-01 3.44E-01 3.64E-01 2.47E+00 5.10E-02 2.06%

9.87E-08 28.25%

Unit 2 Baseline Frequency 3.03E-05 1.55E-08 8.42E-07 8.52E-08 0.OOE+00 0.OOE+00 0.OOE+00 1.58E-05 2.37E-07 1.91 E-06 2.66E-06 5.18E-05 Corrosion Addition 1.06E-09 Unit 2 Baseline Dose Rate 1.17E-01 1.76E-03 3.25E-02 1.15E-02 0.OOE+00 0.OOE+00 0.OOE+00 2.20E+00 2.68E-01 3.59E-01 3.69E-01 3.35E+00

.4.40E-02 1.31%

8.52E-08 39.96%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 159 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-lU years Unit 1 Unit 2 Increase to 3a Frequency 0

0 Increase to 3b Frequency 0.0579%

0.0579%

Unit I Unit 1 Unit 2 Unit 2 Dose Person-Frequency Dose Rate Frequency Rate (1-in-Rem (50-(1-in-10 (1-in-10 (1-in-10 10 yrs Class miles) yrs ILRT) yrs ILRT) yrs ILRT)

ILRT) 1 3.86E+03 3.25E-05 1.26E-01 2.81E-05 1.08E-01 2

1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 3.25E-06 Addition 1.25E-01 2.81E-06 Addition 1.08E-01 3b 1.35E+05 3.46E-07 2.09E-08 4.67E-02 2,99E-07 1.80E-08 4.03E-02 4

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 5

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 6

0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2,37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91 E-06 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.58E+00 5,18E-05 3.45E+00 ILRT Dose Rate from 3a and 3b =

1.72E-01 1.49E-01

% of Total =

6.67%

4.31%

LERF from 3b =

3.46E-07 2.99E-07 CCFP%Liner=

28.74%

40.37%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 160 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed Risk impact Assessment for Extending Containment Type A Test Interval 1-in-1 5-1/2 years Unit 1 Unit 2 Increase to 3a Frequency 0

0 Increase to 3b Frequency 0.4532%

0.4532%

Unit 1 Unit 1 Unit 2 Frequency Dose Rate Frequency Unit 2 Dose Person-(1-in (1-in (1-in Rate (1-in-Rem (50-1/2 yrs 1/2 yrs 1/2 yrs 15-1/2 yrs Class miles)

ILRT)

ILRT) 1 3.86E+03 3.04E-05 1.17E-01 2.63E-05 1.01 E-01 2

1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 5.04E-06 Addition 1.94E-01 4.35E-06 Addition 1.68E-01 3b 1.35E+05 6.67E-07 1.64E-07 9.01E-02 5.76E-07 1.41 E-07 7.78E-02 4

0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 5

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.O0E+00 6

0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 7

1.39E+05 9.38E-06 1.30E+00 1.58E-05 2.20E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.83E-06 3.44E-01 1.91E-06

- 3.59E-01 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 2.68E+00 5.18E-05 3.54E+00 ILRT Dose Rate from 3a and 3b =

2.84E-01 2.46E-01

% of Total 10.60%

6.94%

LERF from 3b 6.67E-07 5.76E-07 CCFP%Liner=

29.38%

40.90%

17670-0001 PB ILRT Rev 3.doc

SSCIENTECH.

CLIENT: Nuclear Management Company BY: E. A. Krantz 77PAGE: 161 OF 161 FILE NO. 17670-0001, Rev. 3 CHECKED BY: G.W. Kindred Date: 12/06/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Other Pertinent Risk Metrics 10 to 15-1/2 Increase (Person-rem/ry) 3 to 15-1/2 Increase (Person-rem/ry) 10 to 15-1/2 Delta-LERF 3 to 15-1/2 Delta-LERF 10 to 15-1/2 Delta-CCFP 3 to 15-1/2 Delta-CCFP Unit 1 1.12E-01 2.34E-01 3.21 E-07 5.68E-07 0.6%

1.1%

Unit 2 9.71 E-02 2.02E-01 2.78E-07 4.91 E-07 0.5%

0.9%

1.70E-08 1.23E-07 1.40E-07 3 to 10 Delta-LERF from Corrosion 10 to 15-1/2 Delta-LERF from Corrosion Increase in LERF (ILRT 3-to-15-1/2 years) from Corrosion 1.96E-08 1.43E-07 1.62E-07 17670-0001 PB ILRT Rev 3.doc

ML073520398

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