NRC 2007-0063, License Amendment Request 256 One-time Extension of Containment Integrated Leakage Rate Test Interval

License Amendment Request 256 One-time Extension of Containment Integrated Leakage Rate Test Interval

ML072910053
Person / Time
Site:	Point Beach
Issue date:	10/12/2007
From:	Mccarthy J Florida Power & Light Energy Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2007-0063
Download: ML072910053 (217)

Text

FPL Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241 FPL En;:r.gd;

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October 12, 2007 NRC 2007-0063 10 CFR 50.90 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 License Amendment Request 256 One-Time Extension of Containment Integrated Leakage Rate Test Interval Pursuant to 10 CFR 50.90, FPL Energy Point Beach, LLC (FPLE-PB) hereby requests to amend Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP),

Units 1 and 2, respectively. FPLE-PB proposes to revise Technical Specification (TS) 5.5.15 "Containment Leakage Rate Testing Program," for Units 1 and 2. The proposed change would allow a one-time interval extension of no more than 5 years for the Type A, Integrated Leakage Rate Test (ILRT). FPLE-PB has evaluated the proposed change in accordance with 10 CFR 50.92 and concluded that the change involves no significant hazards consideration.

The proposed amendment is risk-informed and follows the guidance in Regulatory Guide (RG) 1. I 74, "An approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Bases," Revision 1, November 2002. FPLE-PB has performed an analysis showing that the increase in risk resulting from the proposed amendment is small and within established guidance. FPLE-PB has also determined that defense-in-depth principles will be maintained based on both risk and other considerations. contains the description and analysis of this application. Enclosure 2 contains a markup of the affected TS page. Enclosure 3 contains the calculation of the risk impact assessment for extending the containment Type A test interval.

FPLE-PB requests approval of the proposed license amendment by March 1, 2008, with the amendment being implemented within 30 days. The lead-time for approval is necessary to allow FPLE-PB adequate time for the mobilization of the qualified contractors to perform the ILRT for the April 2008 Unit 2 refueling outage if the application is not approved.

This submittal has been reviewed by the Plant Operations Review Committee. This application contains no new commitments or revisions to existing commitments.

An FPL Group company

Document Control Desk Page 2 In accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated Wisconsin Official.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on October 12,2007 Very truly yours, Site Vice President Enclosures cc: Regional Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW

ENCLOSURE I LICENSE AMENDMENT REQUEST 256 POINT BEACH NUCLEAR PLANT ONE-TIME EXTENSION OF CONTAINMENT INTEGRATED LEAKAGE RATE TEST INTERVAL EVALUATION OF THE PROPOSED CHANGE I.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION

3. TECHNICAL EVALUATION

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirementslcriteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES 13 Pages Follow

1.0

SUMMARY

DESCRIPTION This License Amendment Request is submitted to revise the Point Beach Nuclear Plant (PBNP)

Units 1 and 2 Technical Specifications (TS) Surveillance Requirements for containment integrated leakage rate testing contained in TS 5.5.15.a. This revision would allow a one-time extension of the interval between reactor containment vessel integrated leakage rate tests (ILRTs) from 10 to 15 years.

The proposed amendment is risk-informed and follows the guidance in Regulatory Guide (RG) 1. I 74, "An approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis". In accordance with RG 1.174, FPL Energy Point Beach, LLC (FPLE-PB) has performed an analysis showing that the increase in risk resulting from the proposed amendment is small and within established guidance.

2.0 DETAILED DESCRIPTION The proposed license amendments would revise TS 5.5.15.a to allow a one-time interval extension of no more than 5 years for the Type A ILRT on both units. This revision is a one-time exception to the 1O-year frequency of the performance-based leakage rate testing program for Type A tests as defined by Nuclear Energy Institute (NEI) 94-01, Revision 0, "Industry Guideline For Implementing Performance-Based Option of Title 10 CFR 50, Appendix J", and endorsed by 10 CFR 50, Appendix J, Option B.

The proposed one-time exception being requested is to TS Surveillance Requirement 5.5.15.a to perform an ILRT at a frequency of up to 10 years, with an allowance for a 15-month extension. The requested exception is to allow the ILRT to be performed within 15 years from the last ILRT. The specific changes to the TS are provided in Enclosure 2.

2.1 Backnround The PBNP reactor containment system is a right cylinder with a flat base slab and a shallow domed roof. A % inch thick welded ASTM A-442 steel liner is attached to the inside face of the concrete shell to insure a high degree of leak tightness. The base liner is installed on top of the structural slab and is covered with concrete. The structure provides biological shielding for both normal and accident situations. The internal containment net free volume is 1,000,000 cubic feet, and its associated engineered safety features systems are capable of withstanding a design internal pressure of 60 pounds per square inch gauge and a temperature of 286°F. The engineered safety features for containment include containment spray and the air recirculation cooling systems, which are used to ensure that containment does not exceed its design value of 60 pounds per square inch gauge at 286°F. The containment systems and engineered safety features are described in detail in Chapters 5 and 6, respectively, of the PBNP Final Safety Analysis Report (FSAR).

2.2 Current Requirements TS 5.5.15.a requires that a program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in RG 1.163, "Performance-Based Containment Leak-Test Program," dated Page 1 of 13

September 1995. Regulatory Position C.l of RG 1.I63 states that licensees should establish test intervals based upon the criteria in Section 11.0 of NEI 94-01. Section 11.0 of NEI 94-01 references Section 9.0 which allows ILRTs to be performed at a frequency of one per 10 years if the calculated leakage rate for two consecutive previous tests are less than 1.0 La. La is defined in PBNP TS 5.5.15.c as 0.40 percent of containment air weight per day at the peak accident pressure, Pa,of 60.0 psig. The PBNP reactor containment vessels have met this criterion and therefore, qualify for the 10-year frequency.

2.3 Basis for Current Requirements The maximum allowable containment leakage rate, Laat Pashall be 0.4 percent of containment air weight per day, specified in TS 5.5.15.a. The Containment Leakage Rate Testing program implements the leakage rate testing of containment as required by 10 CFR 50 Appendix J Option B. TS 5.5.15.d limits as-left Type A leakage. This leakage shall be 4.75 La.

The performance-based ILRT requirements of Option B of 10 CFR 50, Appendix J, provide an alternative to the three tests per 10-year frequency specified by the prescriptive requirements of Option A of 10 CFR 50, Appendix J. As documented in RG 1.163, the NRC has endorsed NEI 94-01 as providing acceptable methods for complying with the requirements of Option B of 10 CFR 50 Appendix J. NEI 94-01 specifies an ILRT frequency of one test per 10 years if certain performance criteria are met. The basis for the one test per 10-year frequency is described in Section II.O of NEI 94-01, which states that NUREG-1493, "Performance-Based Containment Leak-Test Program", provides the technical basis to support rulemaking that established Option B. That basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. NEI undertook a similar study, the results of which are documented in Electric Power Research Institute (EPRI) report TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," dated August 1994. The EPRI study determined a reduction in the frequency of ILRTs from three tests per 10 years to one test per 10 years would result in an incremental risk contribution of approximately 0.035 percent. This value is comparable to the range of risk increases (0.02 percent to 0.14 percent) presented in NUREG-1493 for the same frequency reduction.

2.4 Reason for Requestinn Amendment Extension of the ILRT interval from 10 years to 15 years would eliminate the need to perform an ILRT for PBNP Units 1 and 2 during the 2008 outages. This would save approximately 0.4 person-rem exposure per unit. FPLE-PB is requesting this license amendment to obtain this personnel exposure reduction.

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3.0 TECHNICAL EVALUTION The proposed amendment would authorize a one-time extension of the ILRT interval from 10 years to 15 years for PBNP. The proposed amendment is supported by both risk and non-risk considerations.

3.1 Risk Assessment An evaluation was performed to assess the risk impact of a one-time extension of the currently allowed containment Type A ILRT frequency from 10 years to 15-112 years.

The risk assessment follows the guidelines from NEI 94-01 and the NRC regulatory guidance, as outlined in RG 1.174, on the use of PRA findings and risk insights in support of the request to change PBNP's licensing basis. This methodology is similar to that presented in EPRl TR-104285 and NUREG-1493.

The potential impact of age-related corrosion of the steel containment vessel on the risk associated with extending the ILRT interval has also been determined. The methodology used for this analysis is similar to the assessments performed for Calvert Cliffs Nuclear Power Plant (CCNPP), and subsequently used in other submittals including those for Comanche Peak and D. C. Cook. The details of this assessment are contained in Enclosure 3.

3.1.2 Input Information The risk assessment utilizes input of population doses for containment failure modes provided in the Environmental Report Operating License Renewal Stage, Point Beach Nuclear Plant, Units 1 and 2, dated February 2004. The updated PRA total Core Damage Frequency (CDF) and frequency of various release categories are based upon the calculations done for the Environmental Report. Data from NUREG-1493 and the EPRl Interim Guidance were used to calculate the probabilities of a liner leak size.

Point Beach PRA Model Revision 3.17 was used for the risk impact assessment. A draft version of PRA Model Revision 3 underwent a peer review in June of 2001. The peer review team concluded that the model was of sufficient quality to support risk-informed applications supported by deterministic analyses, provided the Level A and Level B observations were addressed. Level A observations and Level B observations pertinent to Large Early Release Frequency (LERF) analyses have been resolved.

3.1.3 Results 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 one-in-I 0 year test interval to a one-in-I 5-112 year interval, was found to be 2.2 percent (0.15 person-remlyr) for Unit 1 and 1.2 percent (0.1 person-remlyr) for Unit 2. Given the low total risk to the public, these values are not significant increases in risk.

The realistic combined internallexternal events contribution to Large Early Release LERF at PBNP is 6.90E-06 for Unit 1 and 7.05E-06 for Unit 2 as described in the analysis provided in

Appendix C of Enclosure 3. The change in the realistic combined internal eventslexternal events LERF associated with increasing the ILRT interval at PBNP is 2.95E-07 for Unit 1 and 2.51 E-07 for Unit 2. Because RG 1.I74 defines small changes in LERF as below 1E-OGlyr, an increase in the ILRT interval at PBNP represents a small change in plant risk from the realistic LERF perspective. Similarly, the change in realistic values of LERF of 6.70E-07 for Unit 1 and 5.71E-071yr for Unit 2 falls into Region II, Small Change in Risk, of the acceptance guidelines in RG 1. I 74.

The change in conditional containment failure probability due to the proposed change in ILRT frequency is 0.63 percent for Unit 1 and 0.53 percent for Unit 2. This change is small compared to the total containment failure probability.

The cumulative impact of the proposed change in ILRT frequency is an increase in integrated risk of 0.30 person-remlyr or 4.4 percent of the baseline risk for Unit I , and 0.223 person-remlyr or 2.8 percent of the baseline risk for Unit 2. These cumulative changes are small.

The impact of age-related corrosion of the steel containment is very small on each of the risk measures associated with the extension of the Type A ILRT test frequency. This conclusion remains valid even including consideration of corrosion.

3.2 Other Considerations Consistent with the defense-in-depth philosophy provided in RG 1.I 74, FPLE-PB has assessed other considerations relevant to the proposed amendment. These are discussed below.

3.2.1 ILRT History TS 5.5.15.a requires the measurement of the containment leakage rate. TS 5.5.15.d limits as-left Type A leakage, this leakage shall be 4.75 La. The results of past Type A tests for PBNP are provided below. The more conservative as-left acceptance criteria is listed with the worst case as-found leakage. The current method for leakage determination is the mass point 95 percent upper confidence level (UCL) estimate of leakage rate. The pre-operational tests reported the tests results as a calculated point-to-point leak rate. The results of Type A tests performed at PBNP have met as-found acceptance criteria except for the Unit 1 April 1987 test.

The as-found results were 0.241 percent of containment air weight per day and the as-left results were 0.086 percent of containment air weight per day. Corrections were made during this test for packing leakage. These results demonstrate a history of satisfactory performance for both leak tightness and structural integrity of the containment vessel.

The pre-operational tests and the most recent test for each unit (1997) were conducted at the design pressure of 60 psig. The remainder of the tests were performed at a reduced pressure of 30 psig (half the design pressure) with acceptance criteria established for the reduced pressure test based upon the full pressure acceptance criteria. Performance of a reduced pressure test was allowable per the original Appendix J. In 1995, Appendix J was revised to provide Option B, which does not allow reduced pressure testing. PBNP implemented the amended regulation via license amendments 169 and 173 for Units 1 and 2, respectively dated October 9, 1996, as corrected via NRC letter dated November 13, 1996. These license amendments removed the performance of reduced pressure ILRT testing from the PBNP TS and required testing under design basis loss-of-coolant accident (LOCA) containment peak pressure only.

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As-left acceptance criteria is listed in the tables below relative to the test leakage limit, where:

La = Design Basis Accident Leakage Rate at Accident Conditions (60 psig, 286 OF)

Lp = Maximum Allowable Test Leakage Rate at Test Conditions (60 psig, 80 OF);

LT = Maximum Allowable Test Leakage Rate at Reduced Test Pressure (30 psig, 80 OF).

Unit I Date Test Pressure "As-Found" Leak Rate "As-Left" Acceptance

(~sig) (Wt. % 1 day) Criteria (Wt. % I day)

October 1997 60 0.0465 0.300 (0.75La)

April 1993 30 0.072 0.212 (0.75LT)

April 1990 30 0.067 0.212 (0.75LT)

April 1987 30 0.241 0.212 (0.75LT)

April 1984 30 0.195 0.212 (0.75LT)

October 1981 30 0.0680 0.212 (0.75LT)

October 1977 30 -0.0179 0.212 (0.75LT)

April 1974 30 0.1254 0.212 (0.75LT)

• June 1970 60 0.0235 0.289 (0.75Lp)

• June 1970 30 0.0824 0.204 (0.75LT)

Pre-operational test

-Unit 2 Date Test Pressure "As-Found" Leak Rate "As-Left" Acceptance

(~sig) (Wt. % I day) Criteria (Wt. % I day)

March 1997 60 0.1087 0.300 (0.75La)

October 1992 30 0.101 0.201 (0.75LT)

September 1989 30 0.060 0.201 (0.75LT)

October 1986 30 0.040 0.201 (0.75LT)

April 1982 30 0.0715 0.201 (0.75LT)

March 1978 30 0.0930 0.201 (0.75LT)

October 1974 30 -0.0050 0.201 (0.75LT)

• March 1971 60 0.0129 0.289 (0.75Lp)

• March 1971 30 0.00863 0.204 (0.75LT)

• Pre-operational test Page 5 of 13

3.2.2 Local Leakaqe Rate Testing (LLRT)

As documented in NUREG-1493, industry experience has shown that most ILRT failures result from leakage that is detectable by local leakage rate testing (Type B and C testing as defined in 10 CFR 50, Appendix J). The PBNP local leak rate test (LLRT) requirements contained in the Containment Leakage Rate Testing Program are not affected by this proposed amendment.

Specific testing frequencies for the Appendix J local leak rate tests are reviewed prior to every refueling outage (18-month cycle). An outage scope document is issued to document local leak rate test periodicity and to ensure all pre-maintenance and post-maintenance testing is complete. The post-outage report provides a written record of extended testing interval changes and reasons for the changes based upon testing results and maintenance history.

Based on the above measures, the LLRT program will provide continuing assurance that the most likely sources of leakage will be identified and repaired.

3.2.3 Containment Inservice Inspection Program (IWE)

PBNP has established a containment inservice inspection program in accordance with 10 CFR 50.55a for Class MC components. The second IWE inspection interval has been developed in accordance with the requirements of the 2001 Edition with the 2003 Addenda of the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE, as modified by 10 CFR 50.55a. The scope of the program includes all the accessible pressure retaining containment surface areas including: Containment vessel liner surfaces and integral attachments, surfaces requiring augmented examination, mechanicaI/electricalpenetrations, moisture barriers, pressure retaining bolting and Appendix J tested IWE components. The period prior to the requested 5-year extension period was concurrent with the first 10-year inspection interval, which was established from September 9, 1996, to September 9, 2006, but was extended to September 9, 2007, as permitted by IWA-2430(d),

1992 Edition11992 Addenda. The first 10-year inspection interval was conducted in accordance with the ASME IWE 1992 Edition with the 1992 Addenda. The second 10-year inspection interval, from September 2007 through September 2016, will apply Subsection IWE, 2001 Edition with the 2003 Addenda.

Each 10-year inspection interval consists of three examination periods. A visual examination of interior and exterior containment vessel surface areas is required each period by ASME Section XI and is implemented by the PBNP IWE Containment lnspection Program. The containment inservice inspection program is not affected by the proposed amendment for Appendix J testing. Therefore, even though the proposed amendment will extend the ILRT testing frequency for 5 additional years, the IWE inspections will supplement the requirements of RG 1.163. General visual examinations will be conducted in two refueling outages during the requested ILRT extension period before the next Type A test. These examinations and requirements will continue to provide assurance that degradation of the containment will be detected and corrected before it can result in a leakage path.

3.2.4 Approved Alternatives to Subsection IWE Requirements For the second IWE inservice inspection interval, there are no relief requests or other NRC approved alternatives being implemented.

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3.2.5 IWE Examination Cateqory E-C, Item No. E4.11 and E4.12 - Containment Surfaces Requiring Auqmented Examination This category includes IWE component areas selected for augmented examination because of known existing degraded conditions. Surface areas likely to experience accelerated degradation and aging require augmented examination. In addition, interior containment surfaces that are subject to excessive wear causing a loss of protective coatings, deformation or material loss are also examined. Examination methods are detailed visual examinations (VT-1) and ultrasonic testing (UT).

3.2.6 Containment Inspection History Based on previous PBNP Units 1 and 2 containment inspections, areas were identified that require augmented inspections. A summary of the specific cases is as follows:

Units 1 and 2 Containment Horizontal Liner Plate at El. 6'-6":

Access to the El. 6'-6" horizontal liner plate is achieved through seven core drilled holes through the El. 8' floor in Unit 1 and four core drilled holes through the El. 8' floor in Unit 2.

These core drilled holes were installed in 1988 as a result of standing water discovered on the El. 8' of containment. They provided access to monitor the corrosion rate of the horizontal liner plate. Corrosion probes were originally installed in two core drilled holes for each unit and conductivity was measured to monitor corrosion rate. Caulking was installed as a moisture barrier at the same time to prevent water from accessing the horizontal liner plate at El. 6' 6". The moisture barriers and liners are examined in accordance with the IWE program.

Units 1 and 2 Mechanical Penetrations:

Corrosion and damaged coatings were observed on 14 pipe penetrations on Unit 1 and 10 pipe penetrations on Unit 2 during the IWE pre-service inspections. Augmented examinations were performed on each penetration until the coatings on the penetrations were repaired during subsequent refueling outages. Coatings on the Unit 1 penetrations were repaired during U1R25 and on Unit 2 during U2R24. During the period of augmented inspections, the containment boundary remained intact.

Units 1 and 2 Containment Sump A:

Units 1 and 2 containment sump A, which is below the reactor vessel, has horizontal liner plates under the concrete floor, vertical liner plates on the walls and a horizontal liner plate on a portion of the ceiling. Due to indications adjacent to the inaccessible liner plate underneath the concrete floor, the inaccessible area was required to be examined and evaluated. Access to the liner plates below the concrete is provided by the installation of core drilled holes. The liner plates below grade were examined and indications noted were determined to be acceptable. However, due to the repeated wetting and drying that occurs in the area, the inaccessible areas will continue to be examined in accordance with IWE-1241(a).

During the drilling operations for the core holes in Unit 1, a drill bit contacted the floor liner and gouged the liner. The indication was evaluated and determined not to affect the integrity of the containment liner plate.

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Liner Plates 1CP-130 ( U l ) and 2CP-129 (U2):

These liner plates were observed to have gouges deeper than 10 percent of the nominal plate thickness. As such, they were designated for augmented examinations as required by IWE-2420(b). Indications on both liner plates were evaluated and determined to be acceptable.

Units 1 and 2, Personnel Access Airlock:

During an examination of the personnel air lock for Units 1 and 2, standing water was discovered below the deck plates on the floors of the air locks. The condition was evaluated and determined to be acceptable. However, since wetted conditions were identified, the wetted area for both units was placed on an augmented examination plan as required by IWE-1241(a).

3.2.7 Containment Penetration Bellows Information Notice 92-20, "Inadequate Local Leak Rate Testing," dated March 3, 1992, stated that problems exist with testing of stainless steel containment penetration bellows. Specifically, in-leakage through such bellows may not be readily detectable by LLRTs. The testing deficiency can occur if the test tap pressurizes between the two sheets of bellows materials.

The PBNP mechanical penetrations sealed with a bellows arrangement are located outside of containment and are not subjected to containment pressure. The portion of the mechanical penetrations inside containment that provide the containment pressure boundary is tested during the Type A test by removing the mechanical penetration plug outside containment to ensure full differential test pressure across the mechanical penetration weld to the liner plate (e.g., single passive barrier).

3.2.8 Maintenance Rule The containment isolation function of limiting the release of radioactive fission products following an accident has been classified as high risk significant and its condition is monitored pursuant to 10 CFR 50.65 in accordance with the PBNP Maintenance Rule program. Operability of the containment isolation equipment is ensured by compliance with TS Sections 3.6 and 5.5. The proposed amendment affects only the ILRT requirements and does not impact the PBNP Maintenance Rule program.

3.3 Conclusions This proposed amendment requests a one-time containment ILRT interval extension for Unit 1 and for Unit 2 from 10 years to 15 years. FPLE-PB has demonstrated through a risk assessment and deterministic considerations that the containment for each unit will continue to perform its safety function following issuance of the proposed TS change. Since the containment safety function will continue to be provided, operation of the PBNP with this revised Technical Specification will continue to protect the health and safety of the public.

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4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirementsicriteria 4.1 . I Title 10 Code of Federal Requlations Part 50 Appendix J Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix J, Option B requires that a Type A test be conducted at a periodic interval based upon historical performance of the overall containment system. The Point Beach Nuclear Plant (PBNP) Technical Specification (TS) 5.5.15, "Containment Leakage Rate Testing Program," requires that leakage rate testing be performed as required by 10 CFR 50, Appendix J, Option B, as modified by approved exemptions, and in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995. This Regulatory Guide endorses, with certain exceptions, Nuclear Energy Institute (NEI) Report NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," dated July 26, 1995.

A Type A test is an overall (integrated) leakage rate test of the containment structure.

NEI 94-01 specifies an initial test interval of 48 months, but allows an extended interval of 10 years, based upon two consecutive successful tests. There is also a provision for extending the test interval an additional 15 months in certain circumstances. The most recent two Type A tests at PBNP have been successful, enabling the current interval to be 10 years.

FPL Energy Point Beach, LLC (FPLE-PB) is requesting a change to TS 5.5.15.a which would add a one-time exception from the guidance contained in RG 1.I63 and NEI 94-01, Revision 0, regarding the Type A test interval for PBNP. Specifically, the proposed TS change allows a one-time extension of the containment integrated leak rate test interval from 10 years to 15 years.

The technical analysis for the proposed license amendment is based on risk related and non-risk related considerations. A risk analysis was performed which demonstrated that the increases in estimated person-rem and containment large early release frequency are consistent with guidance provided in RG 1.174, and NUREG-1493. FPLE-PB has also demonstrated that defense-in-depth would be provided by the low increase in the conditional containment failure probability, and by non-risk based considerations such as the containment integrated leakage rate test results, containment inspection history, the ongoing local leakage rate testing and IWE inservice inspection programs.

4.1.2 Regulatory Guide 1.174, "An Approach for Usina Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensins Bases" RG 1.174 provides an acceptable method for licensees to use in assessing the nature and impact of licensing basis changes when the licensee chooses to support the changes with risk information. FPLE-PB has performed a probabilistic risk assessment using the guidance of RG 1. I 74 to support the proposed TS change which allows a one-time extension of the containment integrated leakage rate test interval from 10 years to 15 years. The applicable guidance in RG 1.I74 is provided as an acceptable change in the annual large early release frequency increase and the total large early release frequency. The increase in the large early release frequency resulting from the proposed extension was determined to be within the guidelines published in RG 1.I74 when the containment integrated leakage rate test interval is Page 9 of 13

extended to 15 years one time for each unit. Thus, the proposed TS changes meet the guidance of RG 1. I 74, which provides a basis for approval.

4.2 Precedent The NRC has approved one-time extensions of the ILRT interval to 15 years based on risk and non-risk based considerations to a number of licensees. This submittal is similar to Prairie Island Nuclear Generating Plant and Kewaunee Nuclear Power Plant, which both received the one-time 5-year extensions. Like PBNP, the risk assessment performed for Kewaunee used the guidelines of NEI 94-01, the methodology used in EPRl TR-104285, the EPRl "Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals," dated November 2001, and the regulatory guidance from RG 1.I 74.

4.3 Significant Hazards Consideration FPLE-PB has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92(c) as discussed below:

1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No This license amendment proposes to revise the Technical Specifications (TS) to allow for the one-time extension of the containment integrated leakage rate test interval from 10 to 15 years. The containment vessel function is to mitigate consequences of an accident. There are no design basis accidents initiated by a failure of the containment leakage mitigation function. The extension of the containment integrated leakage rate test interval will not create an adverse interaction with other systems that could result in initiation of a design basis accident. Therefore, the probability of occurrence of an accident previously evaluated is not significantly increased.

The potential consequences of the proposed change have been quantified by analyzing the changes in risk that would result from extending the containment integrated leakage rate test interval from 10 to 15 years. The increase in risk in terms of person-rem per year within 50 miles resulting from design basis accidents was estimated to be of a magnitude that NUREG-1493 indicates is very small. FPLE-PB has also analyzed the increase in risk in terms of the frequency of large early releases from accidents. The increase in the large early release frequency resulting from the proposed extension was determined to be within the guidelines published in RG 1. I 74. Additionally, the proposed change maintains defense-in-depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation. FPLE-PB has determined that the increase in conditional containment failure probability from reducing the containment integrated leakage rate test frequency from one test per 10 years to one test per 15 years would be small.

Continued containment integrity is also assured by the history of successful containment integrated leakage rate tests, and the established programs for local leakage rate testing and IWE inservice inspections which are not affected by the proposed change.

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Therefore, the probability of occurrence or the consequences of an accident previously analyzed are not significantly increased.

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change to extend the containment integrated leakage rate test interval from 10 to 15 years does not create any new or different accident initiators or precursors. The length of the containment integrated leakage rate test interval does not affect the manner in which any accident begins. The proposed change does not create any new failure modes for the containment and does not affect the interaction between the containment and any other system. Thus, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No The risk-based margins of safety associated with the containment integrated leakage rate test are those associated with the estimated person-rem per year, the large early release frequency and the conditional containment failure probability. FPLE-PB has quantified the potential effect of the proposed change on these parameters and determined that the effect is not significant. The non-risk-based margins of safety associated with the containment integrated leakage rate test are those involved with its structural integrity and leak tightness. The proposed change to extend the containment integrated leakage rate test interval from 10 to 15 years does not adversely affect either of these attributes. The proposed change only affects the frequency at which these attributes are verified. Therefore, the proposed change does not involve a significant reduction in margin of safety.

Based on the above, FPLE-PB concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions In conclusion, based on the considerations discussed above, ( I ) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 EVIRONMENTAL CONSIDERATION The proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(~)(9).Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis,"

Revision 1, dated November 2002

2. Nuclear Energy Institute document NEI 94-01,"Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," dated July 26, 1995

3. Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995

4. NUREG-1493, "Performance-Based Containment Leak-Test Program," dated September 1995

5. Electric Power Research Institute report TR-104285 "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," dated August 1994

6. EPRl document "Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals," dated November 2001

7. Letter from NRC to Kewaunee Nuclear Power Plant, "Kewaunee Nuclear Power Plant -

Issuance of Amendment (TAC No. MB9907), dated April 6,2004 (ML040340168)

8. NRC Information Notice 92-20, "Inadequate Local Leak Rate Testing," dated March 3, 1992

9. Letter from TXU Energy to NRC, "Comanche Peak Steam Electric Station (CPSES),

Docket Nos. 50-445 and 50-446, Response to Request for Additional lnformation Regarding License Amendment Request (LAR) 01-14, Revision to Technical Specification (TS) 5.5.16 Containment Leakage Rate Testing Program," dated June 12,2002 (ML021750053)

Page 120f 13

10. Letter from NRC to Nuclear Management Company, LLC, "Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendment Re: One Time Extension of Containment lntegrated Leak Rate Test (TAC Nos. MC9272 and MC9273)," dated October 2,2006 (ML062400005)

11. Letter from Constellation Nuclear to NRC "Calvert Cliffs Nuclear Power Plant Unit No. 1:

Docket No. 50-317, Response to Request for Additional lnformation Concerning the License Amendment Request for a One-Time lntegrated Leak Rate Test Exception" dated March 27, 2002 (ML020920100)

12. Letter from Indiana Michigan Power Company to NRC, "Donald C. Cook Nuclear Plant Units 1 and 2, Response to Nuclear Regulatory Commission Request for Additional lnformation Regarding the License Amendment for a One-Time Extension of Containment lntegrated Leakage Rate Test Interval," dated November 11, 2002 (ML023170524)

13. 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, dated February 2004 Page 130f 13

ENCLOSURE 2 LICENSE AMENDMENT REQUEST 256 POINT BEACH NUCLEAR PLANT ONE-TIME EXTENSION OF CONTAINMENT INTEGRATED LEAKAGE RATE TEST INTERVAL PROPOSED TECHNICAL SPECIFICATION CHANGE PAGE MARKUPS 1 Page Follows

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Safety Function Determination Program (SFDP) (continued)

A loss of safety function exists when, assuming no concurrent single failure, and assuming no concurrent loss of offsite power or loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LC0 in which the loss of safety function exists are required to be entered.

When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.15 Containment Leakage Rate Testing Program

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.I 63, "Performance-Based Containment Leak-Test Program," dated September 1995 as modified bv the followina exce~tion to NFI 94-01. Rev. 0, Industry Guideline for lm~lementinaPerformance-Based O ~ t i o nof 10 CFR 50. A ~ ~ e n dJ.i xSection 9.2.3. to allow the followina:

(i) T he f irst Unit 1 Tv ~ Ae test ~erformedafter October 7. 1997. shalI be gerformed bv October 7.2012.

(ii) The first Unit 2 T v ~ Ae test Derformed after March 31. 1997. shall be gerformed bv March 31.201 2.

b. The peak design containment internal accident pressure, Pa, is 60 psig.

c. The maximum allowable containment leakage rate, La at Pa, shall be 0.4% of containment air weight per day.

Point Beach 5.5-14 Unit 1 - Amendment No. 2.26 Unit 2 - Amendment No. 2%

ENCLOSURE 3 LICENSE AMENDMENT REQUEST 256 POINT BEACH NUCLEAR PLANT ONE-TIME EXTENSION OF CONTAINMENT INTEGRATED LEAKAGE RATE TEST INTERVAL RISK ASSESSMENT 198 Pages Follow

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 1 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Nuclear Management Company Point Beach Nuclear Power Plant RISK IMPACT ASSESSMENT FOR EXTENDING CONTAINMENT TYPE A TEST INTERVAL Analysis File 17670-0001, Rev. 2 September 2007 Scientech, a Curtiss-Wright Flow Control Company Idaho Falls, Idaho

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 2 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table of Contents Description Page No.

1.0 CLIENT 4 2.0 TITLE 4 3.0 AUTHOR 4 4.0 PURPOSE 4 5.0 INTENDED USE OF ANALYSIS RESULTS 4 6.0 TECHNICAL APPROACH 4 7.0 INPUT INFORMATION 6

8.0 REFERENCES

7 9.0 MAJOR ASSUMPTIONS 8 10.0 IDENTIFICATION OF COMPUTER CODES 8 11.0 DETAILED ANALYSIS 9 12.0 COMPUTER INPUT AND OUTPUT 36 13.0

SUMMARY

OF RESULTS 36

14.0 CONCLUSION

S 36 APPENDIX A 49 APPENDIX B 84 APPENDIX C 166 17670-0001 PB ILRT Rev 2.doc

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval List of Tables Table 1 - Detailed Description for the Eight Accident Classes as defined by EPRI TR-104285 38 Table 2 - Containment Frequency Measures for a Given Accident Class 39 Table 3 - Conditional Person-Rem Measures for a Given Accident Class 40 Table 4a - Unit 1 Baseline Mean Consequence Measures for a Given Accident Class 41 Table 4b - Unit 2 Baseline Mean Consequence Measures for a Given Accident Class 42 Table 5a - Mean Consequence Measures for 10 - Year Test Interval for a Given Accident Class Unit 1 43 Table 5b - Mean Consequence Measures for 10 - Year Test Interval for a Given Accident Class Unit 2 44 Table 6a - Mean Consequence Measures for 15-1/2 - Year Test Interval for a Given Accident Class Unit 1 45 Table 6b - Mean Consequence Measures for 15-1/2 - Year Test Interval for a Given Accident Class Unit 2 46 Table 7a - Effect of Internal Events Hazard Risk on PB ILRT Risk Assessment (Unit 1) 47 Table 7b - Effect of Internal Events Hazard Risk on PB ILRT Risk Assessment (Unit 2) 48 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 4 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval ANALYSIS FILE: 17670-0001, Rev. 2 1.0 CLIENT Nuclear Management Company - 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 used to 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 plants 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 11] 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 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 5 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval increasing the ILRT Type A interval from 10 to 15-1/2 years by using core damage and containment failure frequency information from the most recent update of the PBNP PRA [Reference 13].

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-104285 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-104285 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 2.doc

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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-15-1/2 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 the application of the conservative modeling approach provided in NUREG-1570 (Reference 19). This approach yields high values for frequency of ISGTR, a LERF contributor. Due to the conservative nature of the conditional probability of ISGTR given the combination of high pressure core damage and dry steam generator from this approach, a sensitivity analysis was conducted using a more realistic conditional probability estimate. This sensitivity is attached as Appendix C.

2. Population Doses for containment failure modes. Provided from Applicants 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-4) 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 Jefferys non-informative prior distribution would be appropriate for the 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 7 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class 3b distribution. (The rationale for using the Jefferys non-informative prior distribution was discussed in Reference 4.)

Prob(Class 3b) = Failure probability = (# of failures (0) + 1/2)/(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-104285, Risk Assessment of Revised Containment Leak Rate Testing Intervals August 1994.

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

17670-0001 PB ILRT Rev 2.doc

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SUBJECT:

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

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.

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

15. Applicants 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. Point Beach 2007 Level 2 Model.

19. U.S. Nuclear Regulatory Commission, Risk Assessment of Severe Accident-Induced Steam Generator Tube Rupture, NUREG-1570, March 1998.

9.0 MAJOR ASSUMPTIONS:

1. The containment leakage for Class 1 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 2.doc

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SUBJECT:

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 1 - Quantify the baseline frequency per reactor year for each of the eight 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 13 provides the following results of the latest PBNP PRA update. Also included are the accident classes corresponding to the PBNP release categories (RCs) used in Reference 15.

PBNP Source Term Description EPRI Unit 1 Unit 2 Category Accident Frequencies Frequencies Class 1 Intact 1 2.83E-05 2.44E-05 2 Late Basemat 7b 2.31E-07 2.13E-07 3 Late SGTR 8c 2.62E-06 2.66E-06 4 Late Rupture 7a 7.31E-06 1.23E-05 5 Early Liner 7c 2.24E-08 2.06E-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 1.15E-05 1.19E-05 Total Internal Events CDF 5.02E-05 5.18E-05 Total Internal Events LERF 1.18E-05 1.22E-05 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 1 La. The frequency for these sequences is determined as follows:

Class_1_Frequency = NCF - Class_3a_Frequency - Class_3b_Frequency Where:

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

= 2.83E-05/yr (Unit 1) [From table above for STC1]

= 2.44E-05/yr (Unit 2) [From table above for STC1]

Class_3a_Frequency = Frequency of small pre-existing containment liner leakage

= 7.63E-07/yr (Unit 1) [See below]

= 6.59E-07/yr (Unit 2) [See below]

Class_3b_Frequency = Frequency of large pre-existing containment liner leakage

= 7.63E-08/yr (Unit 1) [See below]

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SUBJECT:

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

= 6.59E-08/yr (Unit 2) [See below]

Therefore.

Class_1_Frequency = 2.83E-05/yr - 7.63E-07/yr - 7.63E-08/yr = 2.74E-05/yr (Unit 1)

Class_1_Frequency = 2.44E-05/yr - 6.59E-07/yr - 6.59E-08/yr = 2.37E-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_3a_Frequency = ProbClass3a * (CDFTotal - CDFIndep)

Where, Class_3a_Frequency = Frequency of EPRI Class 3a given a 3-in-10 years ILRT interval ProbClass3a = Probability of small pre-existing containment liner leakage

= 0.027 [Section 7]

CDFTotal = PB Core Damage Frequency

= 5.02E-05/yr (Unit 1) [Table above]

and

= 5.18E-05/yr (Unit 2) [Table above]

CDFIndep = 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 = 7.31E-06/yr

• EPRI Class 7b = 2.31E-07/yr

• EPRI Class 7c = 2.24E-08/yr

• EPRI Class 8a = 2.37E-07/yr 17670-0001 PB ILRT Rev 2.doc

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SUBJECT:

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

• EPRI Class 8b = 1.15E-05/yr

• EPRI Class 8c = 2.62E-06/yr

= 2.19E-05/yr (Unit 1) and

• EPRI Class 2 = 1.55E-08/yr

• EPRI Class 7a = 1.23E-05/yr

• EPRI Class 7b = 2.13E-07/yr

• EPRI Class 7c = 2.06E-08/yr

• EPRI Class 8a = 2.37E-07/yr

• EPRI Class 8b = 1.19E-05/yr

• EPRI Class 8c = 2.66E-06/yr

= 2.74E-05/yr (Unit 2)

Therefore, Class_3a_Frequency = 0.027 * (5.02E-05/yr - 2.19E-05/yr) = 7.63E-07 (Unit 1) and

= 0.027 * (5.18E-05/yr - 2.74E-05/yr) = 6.59E-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_3b_Frequency = 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]

CDFTotal = PB Core Damage Frequency

= 5.02E-5/yr (Unit 1) [Table above]

and

= 5.18E-5/yr (Unit 2) [Table above]

CDFIndep = 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 = 7.31E-06/yr

• EPRI Class 7b = 2.31E-07/yr

• EPRI Class 7c = 2.24E-08/yr

• EPRI Class 8a = 2.37E-07/yr

• EPRI Class 8b = 1.15E-05/yr

• EPRI Class 8c = 2.62E-06/yr

= 2.19E-05/yr (Unit 1) and

• EPRI Class 2 = 1.55E-08/yr

• EPRI Class 7a = 1.23E-05/yr

• EPRI Class 7b = 2.13E-07/yr 17670-0001 PB ILRT Rev 2.doc

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SUBJECT:

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

• EPRI Class 7c = 2.06E-08/yr

• EPRI Class 8a = 2.37E-07/yr

• EPRI Class 8b = 1.19E-05/yr

• EPRI Class 8c = 2.66E-06/yr

= 2.74E-05/yr (Unit 2)

Therefore, Class_3b_Frequency = 0.0027 * (5.02E-05/yr - 2.19E-05/yr) = 7.63E-08 (Unit 1) and

= 0.0027 * (5.18E-05/yr - 2.74E-05/yr) = 6.59E-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.

• Class 7a - Late containment rupture.

• Class 7b - Late Basemat failure.

• Class 7c - Early liner failure.

Class_7a_Frequency = Frequency of late containment rupture

= 7.31E-06 (Unit 1) [From above table]

and

= 1.23E-05 (Unit 2) [From above table]

Class_7b_Frequency = Frequency of late Basemat failure

= 2.31E-07 (Unit 1) [From above table]

and

= 2.13E-07 (Unit 2) [From above table]

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class_7c_Frequency = Frequency of early liner failure

= 2.24E-08 (Unit 1) [From above table]

and

= 2.06E-08 (Unit 2) [From above table]

Therefore, the Class 7 frequencies are:

Class_7_Frequency = 7.56E-06 (Unit 1) and

= 1.26E-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_8a_Frequency = Frequency of ISLOCA

= 2.37E-07 (Unit 1) [From above table]

and

= 2.37E-07 (Unit 2) [From above table]

Class_8b_Frequency = Frequency of SGTRE

= 1.15E-05 (Unit 1) [From above table]

and

= 1.19E-05 (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 17670-0001 PB ILRT Rev 2.doc

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Interim Guidance (Reference 4) recommends containment leakage rates of 10La 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.

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:

• 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 1

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.

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Population 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 [Table 3]

= 3.86E+03 person-rem 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 [Table 3]

= 1.13E+05 person-rem 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 10La and 35La (Reference 4), respectively, times the population dose of EPRI accident Class 1.

Therefore, Class_3a_Dose = 10

• Class_1_Dose Class_3b_Dose = 35

• Class_1_Dose Class_3a_Dose = 10

• 3.86E+03 person-rem Class_3b_Dose = 35

• 3.86E+03 person-rem Class_3a_Dose = 3.86E+04 person-rem Class_3b_Dose = 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 2.doc

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Population 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 [Table 3]

= 1.39E+05 person-rem 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_8a_Dose = 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_8b_Dose = 1.88E+03 person-sv*100 person-rem / person-sv [Table 3]

= 1.88E+05 person-rem 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 ruptures) is based on the Point Beach release category "Late SGTR (see table above). Therefore, Class_8c_Dose = 1.39E+03 person-sv*100 person-rem / person-sv [Table 3]

= 1.39E+05 person-rem 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-10 year ILRT frequency, the calculated baseline results reflect a 3-per-10 year ILRT surveillance frequency.

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class_1_DoseRate = Class_1_Dose

• Class_1_Frequency Class_2_DoseRate = Class_2_Dose

• Class_2_Frequency Class_3a_DoseRate = Class_3a_Dose

• Class_3a_Frequency Class_3b_DoseRate = Class_3b_Dose

• Class_3b_Frequency Class_7_DoseRate = Class_7_Dose

• Class_7_Frequency Class_8a_DoseRate = Class_8a_Dose

• Class_8a_Frequency Class_8b_DoseRate = Class_8b_Dose

• Class_8b_Frequency Class_8c_DoseRate = Class_8c_Dose

• Class_8c_Frequency Where Class_1_DoseRate = EPRI accident Class_1_dose rate given a 3-in-10 years ILRT interval Class_2_DoseRate = EPRI accident Class_2_dose rate given a 3-in-10 years ILRT interval Class_3a_DoseRate = EPRI accident Class_3a_dose rate given a 3-in-10 years ILRT interval Class_3b_DoseRate = EPRI accident Class_3b_dose rate given a 3-in-10 years ILRT interval Class_7_DoseRate = EPRI accident Class_7_dose rate given a 3-in-10 years ILRT interval Class_8a_DoseRate = EPRI accident Class_8a_dose rate given a 3-in-10 years ILRT interval Class_8b_DoseRate = EPRI accident Class_8b_dose rate given a 3-in-10 years ILRT interval Class_8c_DoseRate = EPRI accident Class_8c_dose rate given a 3-in-10 years ILRT interval Class_1_Dose = EPRI accident Class_1_dose = 3.86E+03 person-rem (Table 3)

Class_2_Dose = EPRI accident Class_2_dose = 1.13E+05 person-rem (Table 3)

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_8c_Dose = EPRI accident Class_8c_dose = 1.39E+05 person-rem (Table 3)

For Unit 1, Class_1_Frequency = Frequency of EPRI accident Class 1 given a 3-in-10 years ILRT interval

= 2.74E-05/ry (Table 2)

Class_2_Frequency = Frequency of EPRI accident Class 2 given a 3-in-10 years ILRT interval

= 1.50E-08/ry (Table 2)

Class_3a_Frequency = Frequency of EPRI accident Class 3a given a 3-in-10 years ILRT interval

= 7.63E-07/ry (Table 2)

Class_3b_Frequency = Frequency of EPRI accident Class 3b given a 3-in-10 years ILRT interval

= 7.63E-08/ry (Table 2)

Class_7_Frequency = Frequency of EPRI accident Class 7 given a 3-in-10 years ILRT interval

= 7.56E-06ry (Table 2)

Class_8a_Frequency = Frequency of EPRI accident Class 8a given a 3-in-10 years ILRT interval

= 2.37E-07/ry (Table 2)

Class_8b_Frequency = Frequency of EPRI accident Class 8b given a 3-in-10 years ILRT interval

= 1.15E-05/ry (Table 2)

Class_8c_Frequency = Frequency of EPRI accident Class 8c given a 3-in-10 years ILRT interval 17670-0001 PB ILRT Rev 2.doc

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SUBJECT:

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

= 2.62E-06ry (Table 2)

For Unit 2, Class_1_Frequency = Frequency of EPRI accident Class 1 given a 3-in-10 years ILRT interval

= 2.37E-05/ry (Table 2)

Class_2_Frequency = Frequency of EPRI accident Class 2 given a 3-in-10 years ILRT interval

= 1.55E-08/ry (Table 2)

Class_3a_Frequency = Frequency of EPRI accident Class 3a given a 3-in-10 years ILRT interval

= 6.59E-07/ry (Table 2)

Class_3b_Frequency = Frequency of EPRI accident Class 3b given a 3-in-10 years ILRT interval

= 6.59E-08/ry (Table 2)

Class_7_Frequency = Frequency of EPRI accident Class 7 given a 3-in-10 years ILRT interval

= 1.26E-05/ry (Table 2)

Class_8a_Frequency = Frequency of EPRI accident Class 8a given a 3-in-10 years ILRT interval

= 2.37E-07/ry (Table 2)

Class_8b_Frequency = Frequency of EPRI accident Class 8b given a 3-in-10 years ILRT interval

= 1.19E-05/ry (Table 2)

Class_8c_Frequency = Frequency of EPRI accident Class 8c given a 3-in-10 years ILRT interval

= 2.66E-06/ry (Table 2)

Therefore, for Unit 1, Class_1_DoseRate = 3.86E+03

• 2.74E-05 = 1.06E-01 (person-rem/ry)

Class_2_DoseRate = 1.13E+05

• 1.50E-08 = 1.70E-03 (person-rem/ry)

Class_3a_DoseRate = 3.86E+04

• 7.63E-07 = 2.94E-02 (person-rem/ry)

Class_3b_DoseRate = 1.35E+05

• 7.63E-08 = 1.03E-02 (person-rem/ry)

Class_7_DoseRate = 1.39E+05

• 7.56E-06 = 1.05E+00 (person-rem/ry)

Class_8a_DoseRate = 1.13E+06

• 2.37E-07 = 2.68E-01 (person-rem/ry)

Class_8b_DoseRate = 1.88E+05

• 1.15E-05 = 2.16E+00 (person-rem/ry)

Class_8c_DoseRate = 1.39E+05

• 2.62E-06 = 3.64E-01 (person-rem/ry)

Therefore, for Unit 2, Class_1_DoseRate = 3.86E+03

• 2.37E-05 = 9.15E-02 (person-rem/ry)

Class_2_DoseRate = 1.13E+05

• 1.55E-08 = 1.76E-03 (person-rem/ry)

Class_3a_DoseRate = 3.86E+04

• 6.59E-07 = 2.54E-02 (person-rem/ry)

Class_3b_DoseRate = 1.35E+05

• 6.59E-08 = 8.91E-03 (person-rem/ry)

Class_7_DoseRate = 1.39E+05

• 1.26E-05 = 1.74E+00 (person-rem/ry)

Class_8a_DoseRate = 1.13E+06

• 2.37E-07 = 2.68E-01 (person-rem/ry)

Class_8b_DoseRate = 1.88E+05

• 1.19E-05 = 2.24E+00 (person-rem/ry)

Class_8c_DoseRate = 1.39E+05

• 2.66E-06 = 3.69E-01 (person-rem/ry)

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

17670-0001 PB ILRT Rev 2.doc

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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 2 3 goes undetected from 18 to 60 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-10 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-10 years ILRT frequency) must be revised to reflect the current 1-in-10 years PB ILRT testing frequency. This is performed as follows:

ProbClass_3a_10 = ProbClass_3a* [ Survtest10/18]

ProbClass_3b_10 = ProbClass_3b * [ Survtest10/18]

Where:

ProbCIass_3a_10 = probability of small pre-existing containment liner leakage given a 1-in-10 years ILRT frequency.

ProbCIass_3b_10 = probability of large pre-existing containment liner leakage given a 1-in-10 years ILRT frequency.

ProbClass_3a = probability of small pre-existing containment liner leakage given a 3-in-10 years ILRT frequency = 0.027 [Section 7.0]

ProbClass_3b = probability of large pre-existing containment liner leakage given a 3-in-10 years ILRT frequency = 0.0027 [Section 7.0]

Survtest10 = surveillance interval of interest/2 = 10 years*12months/year/2 = 60 months Therefore, ProbClass _3a_10 = 0.027 * [ 60/18] = 0.09 ProbClass_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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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-15-1/2 years ILRT frequency being analyzed by PB are calculated as follows:

ProbClass_3a_15 = ProbClass_3a

• Survtest15 / 18 ProbClass_3b_15 = ProbClass_3b

• Survtest15 / 18 Where:

ProbClass_3a_15 = probability of small pre-existing containment liner leakage given a 1-in-15-1/2 years ILRT frequency.

ProbClass_3b_15 = probability of large pre-existing containment liner leakage given a 1-in-15-1/2 years ILRT frequency.

ProbClass_3a = probability of small pre-existing containment liner leakage given a 3-in-10 years ILRT frequency = 0.027 [Section 7.0]

ProbClass_3b = probability of large pre-existing containment liner leakage given a 3-in-10 years ILRT frequency = 0.0027 [Section 7.0]

Survtest15 = surveillance interval of interest/2 = 15-1/2 years*12months/year/2 = 93 months Therefore, ProbClass_3a_15 = 0.027 * [ 93 / 18] = 0.1395 ProbClass_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_3a_Frequency_10 = ProbClass_3a_10 * [CDFTotal - CDFIndep]

Class_3a_Frequency_15 = ProbClass_3a_15 * [CDFTotal - CDFIndep]

Where:

Class_3a_Frequency_10 = frequency of small pre-existing containment liner leakage given a 1 -in-10 years ILRT interval Class_3a_Frequency_15 = frequency of small pre-existing containment liner leakage given a 1-in 1/2 years ILRT interval 17670-0001 PB ILRT Rev 2.doc

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval ProbClass_3a_10 = probability of small pre-existing containment liner leakage given a 1-in-10 years ILRT frequency = 0.09 [See above write-up]

ProbClass_3a_15 = probability of small pre-existing containment liner leakage given a 1-in-15-1/2 years ILRT frequency = 0.1395 [See above write-up]

CDFTotal U1 = PB U1 PSA L1 core damage frequency = 5.02E-05/ry [See step 1 write-up]

CDFTotal U2 = PB U2 PSA L1 core damage frequency = 5.18E-05/ry CDFIndep U1 = CDF for those individual Unit 1 sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences = 2.19E-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 = 2.74E-05/yr [See step 1 write-up]

Therefore, for Unit 1, Class_3a_Frequency_10 = 0.09 * (5.02E-05/yr - 2.19E-05/yr) = 2.54E-06/ry Class_3a_Frequency_15 = 0.1395 * (5.02E-05/yr - 2.19E-05/yr) = 3.94E-06/ry For Unit 2, Class_3a_Frequency_10 = 0.09 * (5.18E-05/yr - 2.74E-05/yr) = 2.20E-06/ry Class_3a_Frequency_15 = 0.1395 * (5.18E-05/yr - 2.74E-05/yr) = 3.41E-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 = ProbClass_3b_10 [CDFTotal - CDFIndep}

Class_3b Frequency_15 = ProbClass_3b_15 [CDFTotal - CDFIndep}

Where:

Class_3b_Frequency_10 = frequency of large pre-existing containment liner leakage given a 1 -in-10 years ILRT interval Class_3b_Frequency_15 = frequency of large pre-existing containment liner leakage given a 1-in 1/2 years ILRT interval ProbClass_3b_10 = probability of large pre-existing containment liner leakage given a 1-in-10 years ILRT frequency = 0.009 [See above write-up]

ProbClass_3b_15 = probability of large pre-existing containment liner leakage given a 1-in-15-1/2 years ILRT frequency = 0.01395 [See above write-up]

CDFTotal U1 = PB U1 PSA L1 core damage frequency = 5.02E-05/ry [See step 1 write-up]

CDFTotal U2 = PB U2 PSA L1 core damage frequency = 5.18E-05/ry CDFIndep U1 = CDF for those individual Unit 1 sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences = 2.19E-05/yr [See step 1 write-up]

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SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 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 = 2.74E-05/yr [See step 1 write-up]

Therefore, for Unit 1, Class_3b_Frequency_10 = 0.009 * (5.02E-05/yr - 2.19E-05/yr) = 2.54E-07/ry Class_3b_Frequency_15 = 0.01395 * (5.02E-05/yr - 2.19E-05/yr) = 3.94E-07/ry For Unit 2, Class_3b_Frequency_10 = 0.009 * (5.18E-05/yr - 2.74E-05/yr) = 2.20E-07/ry Class_3b_Frequency_15 = 0.01395 * (5.18E-05/yr - 2.74E-05/yr) = 3.41E-07/ry 11.1.6 Step 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_1_Dose

• Class_1_Frequency10 Class_2_ DoseRate-10 = Class_2_Dose

• Class_2_Frequency10 Class_3a_ DoseRate-10 = Class_3a_Dose

• Class_3a_Frequency10 Class_3b_ DoseRate-10 = Class_3b_Dose

• Class_3b_Frequency10 Class_7_ DoseRate-10 = Class_7_Dose

• Class_7_Frequency10 Class_8a_ DoseRate-10 = Class_8a_Dose

• Class_8a_Frequency10 Class_8b_ DoseRate-10 = Class_8b_Dose

• Class_8b_Frequency10 Class_8c_ DoseRate-10 = Class_8c_Dose

• Class_8c_Frequency10 Where Class_1_DoseRate-10 = EPRI accident Class_1_dose rate given a 1-in-10 years ILRT interval Class_2_ DoseRate-10 = EPRI accident Class_2_dose rate given a 1-in-10 years ILRT interval Class_3a_ DoseRate-10 = EPRI accident Class_3a_dose rate given a 1-in-10 years ILRT interval Class_3b_ DoseRate-10 = EPRI accident Class_3b_dose rate given a 1-in-10 years ILRT interval Class_7_ DoseRate-10 = EPRI accident Class_7_dose rate given a 1-in-10 years ILRT interval Class_8a_ DoseRate-10 = EPRI accident Class_8a_dose rate given a 1-in-10 years ILRT interval Class_8b_ DoseRate-10 = EPRI accident Class_8b_dose rate given a 1-in-10 years ILRT interval Class_8c_ DoseRate-10 = EPRI accident Class_8c_dose rate given a 1-in-10 years ILRT interval Class_1_Dose = EPRI accident Class_1_dose = 3.86E+03 person-rem (Table 3)

Class_2_Dose = EPRI accident Class_2_dose = 1.13E+05 person-rem (Table 3) 17670-0001 PB ILRT Rev 2.doc

PAGE: 23 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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_8c_Dose = EPRI accident Class_8c_dose = 1.39E+05 person-rem (Table 3)

For Unit 1, Class_1_Frequency10 = Frequency of EPRI accident Class 1 given a 1-in-10 years ILRT interval

= 2.55E-05/ry (Table 5a)

Class_2_Frequency10 = Frequency of EPRI accident Class 2 given a 1-in-10 years ILRT interval

= 1.15E-08/ry (Table 5a)

Class_3a_Frequency10 = Frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval

= 2.54E-06/ry (Table 5a)

Class_3b_Frequency10 = Frequency of EPRI accident Class 3b given a 1-in-10 years ILRT interval

= 2.54E-07/ry (Table 5a)

Class_7_Frequency10 = Frequency of EPRI accident Class 7 given a 1-in-10 years ILRT interval

= 7.56E-06/ry (Table 5a)

Class_8a_Frequency10 = Frequency of EPRI accident Class 8a given a 1-in-10 years ILRT interval

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

Class_8b_Frequency10 = Frequency of EPRI accident Class 8b given a 1-in-10 years ILRT interval

= 1.15E-05/ry (Table 5a)

Class_8c_Frequency10 = Frequency of EPRI accident Class 8c given a 1-in-10 years ILRT interval

= 2.62E-06/ry (Table 5a)

For Unit 2, Class_1_Frequency10 = Frequency of EPRI accident Class 1 given a 1-in-10 years ILRT interval

= 2.20E-05/ry (Table 5b)

Class_2_Frequency10 = Frequency of EPRI accident Class 2 given a 1-in-10 years ILRT interval

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

Class_3a_Frequency10 = Frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval

= 2.20E-06/ry (Table 5b)

Class_3b_Frequency10 = Frequency of EPRI accident Class 3b given a 1-in-10 years ILRT interval

= 2.20E-07/ry (Table 5b)

Class_7_Frequency10 = Frequency of EPRI accident Class 7 given a 1-in-10 years ILRT interval

= 1.26E-05/ry (Table 5b)

Class_8a_Frequency10 = Frequency of EPRI accident Class 8a given a 1-in-10 years ILRT interval

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

Class_8b_Frequency10 = Frequency of EPRI accident Class 8b given a 1-in-10 years ILRT interval

= 1.19E-05/ry (Table 5b) 17670-0001 PB ILRT Rev 2.doc

PAGE: 24 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class_8c_Frequency10 = Frequency of EPRI accident Class 8c given a 1-in-10 years ILRT interval

= 2.66E-06/ry (Table 5b)

Therefore, for Unit 1, Class_1_DoseRate-10 = 3.86E+03

• 2.55E-05 = 9.83E-02 (person-rem/ry)

Class_2_DoseRate-10 = 1.13E+05

• 1.15E-08 = 1.30E-03 (person-rem/ry)

Class_3a_DoseRate-10 = 3.86E+04

• 2.54E-06 = 9.81E-02 (person-rem/ry)

Class_3b_DoseRate-10 = 1.35E+05

• 2.54E-07 = 3.43E-02 (person-rem/ry)

Class_7_DoseRate-10 = 1.39E+05

• 7.56E-06 = 1.05E+00 (person-rem/ry)

Class_8a_DoseRate-10 = 1.13E+06

• 2.37E-07 = 2.68E-01 (person-rem/ry)

Class_8b_DoseRate-10 = 1.88E+05

• 1.15E-05 = 2.16E+00 (person-rem/ry)

Class_8c_DoseRate-10 = 1.39E+05

• 2.62E-06 = 3.64E-01 (person-rem/ry)

Therefore, for Unit 2, Class_1_DoseRate-10 = 3.86E+03

• 2.20E-05 = 8.50E-02 (person-rem/ry)

Class_2_DoseRate-10 = 1.13E+05

• 1.19E-08 = 1.35E-03 (person-rem/ry)

Class_3a_DoseRate-10 = 3.86E+04

• 2.20E-06 = 8.48E-02 (person-rem/ry)

Class_3b_DoseRate-10 = 1.35E+05

• 2.20E-07 = 2.97E-02 (person-rem/ry)

Class_8c_DoseRate-10 = 1.39E+05

• 1.26E-05 = 1.74E+00 (person-rem/ry)

Class_8a_DoseRate-10 = 1.13E+06

• 2.37E-07 = 2.68E-01 (person-rem/ry)

Class_8b_DoseRate-10 = 1.88E+05

• 1.19E-05 = 2.24E+00 (person-rem/ry)

Class_8c_DoseRate-10 = 1.39E+05

• 2.66E-06 = 3.69E-01 (person-rem/ry)

The calculation for a 1-in-15-1/2 years ILRT interval is as follows:

Class_1_DoseRate-15 = Class_1_Dose

• Class_1_Frequency15 Class_2_ DoseRate-15 = Class_2_Dose

• Class_2_Frequency15 Class_3a_ DoseRate-15 = Class_3a_Dose

• Class_3a_Frequency15 Class_3b_ DoseRate-15 = Class_3b_Dose

• Class_3b_Frequency15 Class_7_ DoseRate-15 = Class_7_Dose

• Class_7_Frequency15 Class_8a_ DoseRate-15 = Class_8a_Dose

• Class_8a_Frequency15 Class_8b_ DoseRate-15 = Class_8b_Dose

• Class_8b_Frequency15 Class_8c_ DoseRate-15 = Class_8c_Dose

• Class_8c_Frequency15 Where Class_1_DoseRate-15 = EPRI accident Class_1_dose rate given a 1-in-15-1/2 years ILRT interval Class_2_ DoseRate-15 = EPRI accident Class_2_dose rate given a 1-in-15-1/2 years ILRT interval Class_3a_ DoseRate-15 = EPRI accident Class_3a_dose rate given a 1-in-15-1/2 years ILRT interval Class_3b_ DoseRate-15 = EPRI accident Class_3b_dose rate given a 1-in-15-1/2 years ILRT interval Class_7_ DoseRate-15 = EPRI accident Class_7_dose rate given a 1-in-15-1/2 years ILRT interval Class_8a_ DoseRate-15 = EPRI accident Class_8a_dose rate given a 1-in-15-1/2 years ILRT interval Class_8b_ DoseRate-15 = EPRI accident Class_8b_dose rate given a 1-in-15-1/2 years ILRT interval Class_8c_ DoseRate-15 = EPRI accident Class_8c_dose rate given a 1-in-15-1/2 years ILRT interval Class_1_Dose = EPRI accident Class_1_dose = 3.86E+03 person-rem (Table 3)

Class_2_Dose = EPRI accident Class_2_dose = 1.13E+05 person-rem (Table 3)

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) 17670-0001 PB ILRT Rev 2.doc

PAGE: 25 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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_8c_Dose = EPRI accident Class_8c_dose = 1.39E+05 person-rem (Table 3)

For Unit 1, Class_1_Frequency15 = Frequency of EPRI accident Class 1 given a 1-in-15-1/2 years ILRT interval

= 2.39E- 05/ry (Table 6a)

Class_2_Frequency15 = Frequency of EPRI accident Class 2 given a 1-in-15-1/2 years ILRT interval

= 1.15E -08/ry (Table 6a)

Class_3a_Frequency15 = Frequency of EPRI accident Class 3a given a 1-in-15-1/2 years ILRT interval

= 3.94E -06/ry (Table 6a)

Class_3b_Frequency15 = Frequency of EPRI accident Class 3b given a 1-in-15-1/2 years ILRT interval

= 3.94E -07/ry (Table 6a)

Class_7_Frequency15 = Frequency of EPRI accident Class 7 given a 1-in-15-1/2 years ILRT interval

= 7.56E-06/ry (Table 6a)

Class_8a_Frequency15 = Frequency of EPRI accident Class 8a given a 1-in-15-1/2 years ILRT interval

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

Class_8b_Frequency15 = Frequency of EPRI accident Class 8b given a 1-in-15-1/2 years ILRT interval

= 1.15E-05/ry (Table 6a)

Class_8c_Frequency15 = Frequency of EPRI accident Class 8c given a 1-in-15-1/2 years ILRT interval

= 2.62E-06/ry (Table 6a)

For Unit 2, Class_1_Frequency15 = Frequency of EPRI accident Class 1 given a 1-in-15-1/2 years ILRT interval

= 2.07E-05/ry (Table 6b)

Class_2_Frequency15 = Frequency of EPRI accident Class 2 given a 1-in-15-1/2 years ILRT interval

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

Class_3a_Frequency15 = Frequency of EPRI accident Class 3a given a 1-in-15-1/2 years ILRT interval

= 3.41E-06/ry (Table 6b)

Class_3b_Frequency15 = Frequency of EPRI accident Class 3b given a 1-in-15-1/2 years ILRT interval

= 3.41E-07/ry (Table 6b)

Class_7_Frequency15 = Frequency of EPRI accident Class 7 given a 1-in-15-1/2 years ILRT interval

= 1.26E-05/ry (Table 6b)

Class_8a_Frequency15 = Frequency of EPRI accident Class 8a given a 1-in-15-1/2 years ILRT interval

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

Class_8b_Frequency15 = Frequency of EPRI accident Class 8b given a 1-in-15-1/2 years ILRT interval

= 1.19E-05/ry (Table 6b)

Class_8c_Frequency15 = Frequency of EPRI accident Class 8c given a 1-in-15-1/2 years ILRT interval

= 2.66E-06/ry (Table 6b) 17670-0001 PB ILRT Rev 2.doc

PAGE: 26 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Therefore, for Unit 1, Class_1_DoseRate-15 = 3.86E+03

• 2.39E-05 = 9.24E-02 (person-rem/ry)

Class_2_DoseRate-15 = 1.13E+05

• 1.15E-08 = 1.30E-03 (person-rem/ry)

Class_3a_DoseRate-15 = 3.86E+04

• 3.94E-06 = 1.52E-01 (person-rem/ry)

Class_3b_DoseRate-15 = 1.35E+05

• 3.94E-07 = 5.32E-02 (person-rem/ry)

Class_7_DoseRate-15 = 1.39E+05

• 7.56E-06 = 1.05E+00 (person-rem/ry)

Class_8a_DoseRate-15 = 1.13E+06

• 2.37E-07 = 2.68E-01 (person-rem/ry)

Class_8b_DoseRate-15 = 1.88E+05

• 1.15E-05 = 2.16E+00 (person-rem/ry)

Class_8c_DoseRate-15 = 1.39E+05

• 2.62E-06 = 3.64E-01 (person-rem/ry)

Therefore, for Unit 2, Class_1_DoseRate-15 = 3.86E+03

• 2.07E-05 = 7.98E-02 (person-rem/ry)

Class_2_DoseRate-15 = 1.13E+05

• 1.19E-08 = 1.35E-03 (person-rem/ry)

Class_3a_DoseRate-15 = 3.86E+04

• 3.41E-06 = 1.31E-01 (person-rem/ry)

Class_3b_DoseRate-15 = 1.35E+05

• 3.41E-07 = 4.60E-02 (person-rem/ry)

Class_7_DoseRate-15 = 1.39E+05

• 1.26E-05 = 1.74E+00 (person-rem/ry)

Class_8a_DoseRate-15 = 1.13E+06

• 2.37E-07 = 2.68E-01 (person-rem/ry)

Class_8b_DoseRate-15 = 1.88E+05

• 1.19E-05 = 2.24E+00 (person-rem/ry)

Class_8c_DoseRate-15 = 1.39E+05

• 2.66E-06 = 3.69E-01 (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:

Prct_TD10 = 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:

Prct_TD10 = [(Class_3a_DoseRate-10 + Class_3b_DoseRate-10)/Total_DoseRate-10]

• 100%

Class_3a_DoseRate-10 = EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= 9.81E-02 [Table 5a] for Unit 1 Class_3b_DoseRate-10 = EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= 3.43E-02 [Table 5a]

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

= 4.08 [Table 5a]

Therefore, 17670-0001 PB ILRT Rev 2.doc

PAGE: 27 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Prct_TD10 = [(9.81E-02 + 3.43E-02)/ 4.08]

• 100% = 3.25%

For Unit 2, Class_3a_DoseRate-10 = EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= 8.48E-02 [Table 5b]

Class_3b_DoseRate-10 = EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= 2.97E-02 [Table 5b]

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

= 4.82 [Table 5b]

Therefore, Prct_TD10 = [(8.48E-02+ 2.97E-02)/ 4.82]

• 100% = 2.37%

The percentage contribution to total dose rate from EPRI' s accident Classes 3a and 3b based on the proposed 1-in-15-1/2 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-15-1/2 years ILRT interval, which is calculated using the following equation:

Prct_TD15 = [(Class_3a_DoseRate-15 + Class_3b_DoseRate-15)/Total_DoseRate-15]

• 100%

For Unit 1:

Class_3a_DoseRate-15 = EPRI accident Class 3a dose rate given a 1-in-15-1/2 years ILRT interval

= 1.52E-01 [Table 6a]

Class_3b_DoseRate-15 = EPRI accident Class 3b dose rate given a 1-in-15-1/2 years ILRT interval

= 5.32E-02 [Table 6a]

Total_DoseRate-15 = Total dose rate for all EPRI's classes given a 1-in-15-1/2 years ILRT interval

= 4.14 [Table 6a]

Therefore, Prct_TD15 = [(1.52E-01 + 5.32E-02)/ 4.14]

• 100% = 4.95%

For Unit 2, Class_3a_DoseRate-15 = EPRI accident Class 3a dose rate given a 1-in-15-1/2 years ILRT interval

= 1.31E-01 [Table 6b]

Class_3b_DoseRate-15 = EPRI accident Class 3b dose rate given a 1-in-15-1/2 years ILRT interval

= 4.60E-02 [Table 6b]

Total_DoseRate-15 = Total dose rate for all EPRI's classes given a 1-in-15-1/2 years ILRT interval

= 4.88 [Table 6b]

Therefore, Prct_TD15 = [(1.31E-01+ 4.60E-02)/ 4.88]

• 100% = 3.64%

17670-0001 PB ILRT Rev 2.doc

PAGE: 28 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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-15-1/2 years dose rate is as follows for Unit 1:

Increase10-15 = [(Total_DoseRate Total_DoseRate-10)/ Total_DoseRate-10 ] *100%

Where:

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

= 4.14 (person-rem/ry) [Table 6a]

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

= 4.08 (person-rem/ry) [Table 5a]

Therefore, Increase10-15 = [(4.14- 4.08)/ 4.08 ] *100% = 1.64%

For Unit 2, Total_DoseRate-15 = Total dose rate for all EPRI' s classes given a 1-in-15-1/2 years ILRT interval

= 4.88 (person-rem/ry) [Table 6b]

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

= 4.82 (person-rem/ry) [Table 5b]

Therefore, Increase10-15 = [(4.88- 4.82)/ 4.82 ] *100% = 1.20%

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 2.doc

PAGE: 29 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

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

Class_3b_Frequency = 7.63E-08 [Table 4a]

Class_3b_Frequency10 = 2.54E-07 [Table 5a]

Class_3b_Frequency15 = 3.94E-07 {Table 6a]

Therefore, LERF10-15 = the change in LERF from 1-in-10 years ILRT interval to 1-in-15-1/2 years ILRT interval

= Class_3b_Frequency15 - Class_3b_Frequency10

= 3.94E-07/ry - 2.54E-07/ry

= 1.40E-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:

LERF3-15 = the change in LERF from 3-in-10 years ILRT interval to 1-in-15-1/2 years ILRT interval

= Class_3b_Frequency15 - Class_3b_Frequency

= 3.94E-07/ry - 7.63E-08/ry

= 3.18E-07/ry For Unit 2:

The LERF frequencies of interest are:

Class_3b_Frequency = 6.59E-08 [Table 4b]

Class_3b_Frequency10 = 2.20E-07 [Table 5b]

Class_3b_Frequency15 = 3.41E-07{Table 6b]

Therefore, LERF10-15 = the change in LERF from 1-in-10 years ILRT interval to 1-in-15-1/2 years ILRT interval

= Class_3b_Frequency15 - Class_3b_Frequency10

= 3.41E-07/ry - 2.20E-07/ry

= 1.21E-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:

LERF3-15 = the change in LERF from 3-in-10 years ILRT interval to 1-in-15-1/2 years ILRT interval

= Class_3b_Frequency15 - Class_3b_Frequency

= 3.41E-07/ry - 6.59E-08/ry

= 2.75E-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 1E-06 but below 1E-17670-0001 PB ILRT Rev 2.doc

PAGE: 30 OF CLIENT: Nuclear Management Company BY: E. A. Krantz 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 05/ry and increases in LERF greater that 1E-07 but below 1E-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 1.18E-05 for Unit 1 and 1.22E-05 for Unit 2 (Section 11.1.1); these values of LERF would be in Region 1 of Figure 4 of Regulatory Guide 1.174 and would not allow any changes that would increase risk. However, these are conservative LERF values as discussed in Appendix C. More realistic values for LERF are 4.01E-06 for Unit 1 and 4.15E-06 for Unit 2 as described in the sensitivity analysis provided in Appendix C. The change in the realistic values of 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 1.71E-07 for Unit 1 and 1.48E-07 for Unit 2. Because Reference 9 defines small changes in LERF as below 1E-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 from the realistic 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 3.89E-07 for Unit 1 and 3.36E-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 4

Guide 1.174 (Reference 9) .

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-[Intact_Containment_Frequency/Total_CDF]

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, CCFP10 = [1 - (Class_1_ Frequency10 + Class_3a_ Frequency10)/ CDFTotal ]

• 100%

Where; 4

The defense-in-depth philosophy is maintained as a reasonable balance among prevention of core damage, containment failure and consequence mitigation.

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval CCFP10 = Conditional containment failure probability given 1-in-10 years ILRT interval Class_1_ Frequency10 = Frequency per year of EPRI accident Class 1 given 1-in-10 years ILRT interval Class_3a_ Frequency10 = Frequency per year of EPRI accident Class 3a given 1-in-10 years ILRT interval For the 1-in-15-1/2 years ILRT interval, CCFP15 = [1 - (Class_1_ Frequency15 + Class_3a_ Frequency15)/ CDFTotal ]

• 100%

Where; CCFP15 = Conditional containment failure probability given 1-in-15-1/2 years ILRT interval Class_1_ Frequency15 = Frequency per year of EPRI accident Class 1 given 1-in-15-1/2 years ILRT interval Class_3a_ Frequency15 = Frequency per year of EPRI accident Class 3a given 1-in-15-1/2 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.

= 2.74E-05/ry [Table 4a]

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

= 7.63E-07/ry [Table 4a]

Class_1_ Frequency10 = Frequency per year of EPRI accident Class 1 given 1-in-10 years ILRT interval

= 2.55E-05/ry [Table 5a]

Class_3a_ Frequency10 = Frequency per year of EPRI accident Class 3a given 1-in-10 years ILRT interval

= 2.54E -06/ry [Table 5a]

Class_1_ Frequency15 = Frequency per year of EPRI accident Class 1 given 1-in-15-1/2 years ILRT interval

= 2.39E-05/ry [Table 6a]

Class_3a_ Frequency15 = Frequency per year of EPRI accident Class 3a given 1-in-15-1/2 years ILRT interval

= 3.94E-06/ry[Table 6a]

Therefore, CCFP = [1 - (Class_1_ Frequency + Class_3a_ Frequency)/ CDFTotal ]

• 100%

= [1-(2.74E-05 + 7.63E-07)/5.02E-05]

• 100%

= 43.85%

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval CCFP10 = [1 - (Class_1_ Frequency10 + Class_3a_ Frequency10)/ CDFTotal ]

• 100%

= [1 - (2.55E-05 + 2.54E-06)/ 5.02E-05 ]

• 100%

= 44.21%

CCFP15 = [1 - (Class_1_ Frequency15 + Class_3a_ Frequency15)/ CDFTotal ]

• 100%

= [1 - (2.39E-05 + 3.94E-06)/ 5.02E-05 ]

• 100%

= 44.48%

For Unit 2, the frequencies of interest are:

CDFTotal = 5.18E-05 [Table 4b]

Class_1_ Frequency = Frequency per year of EPRI accident Class 1.

= 2.37E-05/ry [Table 4b]

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

= 6.59E-07/ry [Table 4b]

Class_1_ Frequency10 = Frequency per year of EPRI accident Class 1 given 1-in-10 years ILRT interval

= 2.20E-05/ry [Table 5b]

Class_3a_ Frequency10 = Frequency per year of EPRI accident Class 3a given 1-in-10 years ILRT interval

= 2.20E-06/ry [Table 5b]

Class_1_ Frequency15 = Frequency per year of EPRI accident Class 1 given 1-in-15-1/2 years ILRT interval

= 2.07E-05/ry [Table 6b]

Class_3a_ Frequency15 = Frequency per year of EPRI accident Class 3a given 1-in-15-1/2 years ILRT interval

= 3.41E-06/ry [Table 6b]

Therefore, CCFP = [1 - (Class_1_ Frequency + Class_3a_ Frequency)/ CDFTotal ]

• 100%

= [1-(2.37E-05 + 6.59E-07)/5.18E-05]

• 100%

= 52.97%

CCFP10 = [1 - (Class_1_ Frequency10 + Class_3a_ Frequency10)/ CDFTotal ]

• 100%

= [1 - (2.20E-05 + 2.20E-06)/ 5.18E-05 ]

• 100%

= 53.27%

CCFP15 = [1 - (Class_1_ Frequency15 + Class_3a_ Frequency15)/ CDFTotal ]

• 100%

= [1 - (2.07E-05 + 3.41E-06)/ 5.18E-05 ]

• 100%

= 53.50%

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:

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval CCFP3-10 = CCFP10 - CCFP For Unit 1:

CCFP3-10 = 44.21% - 43.85% = 0.36%

For Unit 2:

CCFP3-10 = 53.27% - 52.97% = 0.30%

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:

CCFP10-15 = CCFP15 - CCFP10 For Unit 1:

CCFP10-15 = 44.48%- 44.21% = 0.27%

For Unit 2:

CCFP10-15 = 53.50% - 53.27% = 0.23%

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:

CCFP3-15 = CCFP15 - CCFP For Unit 1:

CCFP3-15 = 44.48% - 43.85% = 0.63%

For Unit 2 CCFP3-15 = 53.50% - 52.97% = 0.53%

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

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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 2.03E-05 (1.18E-05 + 8.51E-06) for Unit 1 and 2.07E-05 (1.22E-05 + 8.54E-06) for Unit 2 (Section 11.1.1); these values of LERF would be in Region 1 of Figure 4 of Regulatory Guide 1.174 and would not allow any changes that would increase risk. However, these are conservative LERF values as discussed in Appendix C.

More realistic values for the combined internal/external events LERF are 6.90E-06 for Unit 1 and 7.05E-06 for Unit 2 as described in the sensitivity analysis provided in Appendix C. 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-15-1/2 years is 2.95E-07 for Unit 1 and 2.51E-07 for Unit 2. Because Reference 9 defines small changes in LERF as below 1E-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 from the realistic 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 6.70E-07 for Unit 1 and 5.71E-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-15-1/2 years test interval, to be 1.64% or 0.06 person-rem/ry for Unit 1 (1.2% or 0.06 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-15-1/2 years is 0.27% for Unit 1 and 0.23% 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 (RAls) in response to the onetime relief requests for the ILRT surveillance interval submitted by various licensees. One of the RAls 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 17670-0001 PB ILRT Rev 2.doc

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

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 6.07E-09 for Unit 1 (5.24E-09/ry for Unit 2). This value is below the NRC Regulatory Guide 1.174 limit of 1E-07/ry. Therefore, because Regulatory Guide 1.174 defines very small changes in LERF as below 1E-07/ry, increasing the ILRT interval at PBNP from the currently allowed 1-in-10 years to 1-in-15-1/2 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 8.19E-04 person-rem/ry for Unit 1 (7.08E-04 person-rem/ry for Unit 2), and the conditional containment failure probability increase is estimated to be approximately 0.01%. 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.

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

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 2.2% (0.15 person-rem/yr) for Unit 1 and 1.2% (0.1 person-rem/yr) 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 2.03E-05 (1.18E-05 +

8.51E-06) for Unit 1 and 2.07E-05 (1.22E-05 + 8.54E-06) for Unit 2 (Section 11.1.1); these values of LERF would be in Region 1 of Figure 4 of Regulatory Guide 1.174 and would not allow any changes that would increase risk. However, these are conservative LERF values as discussed in Appendix C.

More realistic values for the combined internal/external events LERF are 6.90E-06 for Unit 1 and 7.05E-06 for Unit 2 as described in the sensitivity analysis provided in Appendix C. 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-15-1/2 years is 2.95E-07 for Unit 1 and 2.51E-07 for Unit 2. Because Reference 9 defines small changes in LERF as below 1E-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 from the realistic 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 6.70E-07 for Unit 1 and 5.71E-07/ry for Unit 2 falls into Region II, Small Change in Risk, of the acceptance guidelines in NRC Regulatory Guide 1.174.

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval The change in conditional containment failure probability due to the change in ILRT frequency from 3 tests in 10 years to 1 test in 15-1/2 years is 0.63% for Unit 1 and 0.53% for Unit 2 is small compared to the total containment failure probability. The cumulative impact of the change in ILRT frequency from 3 tests in 10 years to 1 test in 15-1/2 years is an increase in integrated risk of 0.30 person-rem/yr for Unit 1 (0.223 person-rem/yr for Unit 2) or 4.4% of the baseline risk for Unit 1 (2.8% for Unit 2). All of these cumulative changes are small and considered acceptable.

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.

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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 containment 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.

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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. 1 CDF per yr. CDF 1 No Containment Failure 2.74E-05 54.6% 2.37E-05 45.8%

2 Large Containment Isolation Failure (Failure-To-Close) 1.50E-08 0.0% 1.55E-08 0.0%

3a Small Isolation Failures (Liner Breach) 7.63E-07 1.5% 6.59E-07 1.3%

3b Large Isolation Failures (Liner Breach) 7.63E-08 0.2% 6.59E-08 0.1%

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

5 Small Isolation Failure - Failure-To-Seal (Type C Test) 0.00E+00 0.0% 0.00E+00 0.0%

6 Containment isolation Failures (Dependent failures, Personnel Errors) 0.00E+00 0.0% 0.00E+00 0.0%

7 Severe Accident Phenomena Induced Failure 7.56E-06 15.1% 1.26E-05 24.2%

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

8b Containment Bypassed (Early SGTR) 1.15E-05 22.9% 1.19E-05 23.0%

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%

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE 3 - Conditional Person-Rem Measures for a Given Accident Class Person-Rem Class Description (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 2.doc

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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 1 Person-Rem Unit 1 Frequency Class Description Person-Rem/yr (50-miles) - per yr.

(50-miles) 1 No Containment Failure 3.86E+03 2.74E-05 1.06E-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 7.63E-07 2.94E-02 3b Large Isolation Failures (Liner Breach) 1.35E+05 7.63E-08 1.03E-02 4 Small Isolation Failure - Failure-To-Seal (Type B test) -- 0.00E+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, -- 0.00E+00 Personnel Errors) 0.00E+00 7 Severe Accident Phenomena Induced Failure 1.39E+05 7.56E-06 1.05E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 2.37E-07 2.68E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.15E-05 2.16E+00 8c Containment Bypassed (Late SGTR) 1.39E+05 2.62E-06 3.64E-01 Core All Containment Event Tree (CET) Endstates 5.02E-05 3.99E+00 Damage 17670-0001 PB ILRT Rev 2.doc

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

(50-miles) 1 No Containment Failure 3.86E+03 2.37E-05 9.15E-02 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 6.59E-07 2.54E-02 3b Large Isolation Failures (Liner Breach) 1.35E+05 6.59E-08 8.91E-03 4 Small Isolation Failure - Failure-To-Seal (Type B test) -- 0.00E+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, -- 0.00E+00 0.00E+00 Personnel Errors) 7 Severe Accident Phenomena Induced Failure 1.39E+05 1.26E-05 1.74E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 2.37E-07 2.68E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.19E-05 2.24E+00 8c Containment Bypassed (Late SGTR) 1.39E+05 2.66E-06 3.69E-01 Core All Containment Event Tree (CET) Endstates 5.18E-05 4.75E+00 Damage 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 43 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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-Frequency Class Description Rem Rem/yr

-per yr (50-miles) (50-miles) 1 No Containment Failure 3.86E+03 2.55E-05 9.83E-02 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 2.54E-06 9.81E-02 3b Large Isolation Failures (Liner Breach) 1.35E+05 2.54E-07 3.43E-02 4 Small Isolation Failure - Failure-To-Seal (Type B test) -- 0.00E+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) -- 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 1.39E+05 7.56E-06 1.05E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 2.37E-07 2.68E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.15E-05 2.16E+00 8c Containment Bypassed (Late SGTR) 1.39E+05 2.62E-06 3.64E-01 All Containment Event Tree (CET) Endstates 5.02E-05 4.08E+00 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 44 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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-Frequency Class Description Rem Rem/yr

-per yr (50-miles) (50-miles) 1 No Containment Failure 3.86E+03 2.20E-05 8.50E-02 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 2.20E-06 8.48E-02 3b Large Isolation Failures (Liner Breach) 1.35E+05 2.20E-07 2.97E-02 4 Small Isolation Failure - Failure-To-Seal (Type B test) -- 0.00E+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) -- 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 1.39E+05 1.26E-05 1.74E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 2.37E-07 2.68E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.19E-05 2.24E+00 8c Containment Bypassed (Late SGTR) 1.39E+05 2.66E-06 3.69E-01 All Containment Event Tree (CET) Endstates 5.18E-05 4.82E+00 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 45 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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-Frequency Class Description Rem Rem/yr

-per yr (50-miles) (50-miles) 1 No Containment Failure 3.86E+03 2.39E-05 9.24E-02 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 3.94E-06 1.52E-01 3b Large Isolation Failures (Liner Breach) 1.35E+05 3.94E-07 5.32E-02 4 Small Isolation Failure - Failure-To-Seal (Type B test) -- 0.00E+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) -- 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 1.39E+05 7.56E-06 1.05E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 2.37E-07 2.68E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.15E-05 2.16E+00 8c Containment Bypassed (Late SGTR) 1.39E+05 2.62E-06 3.64E-01 All Containment Event Tree (CET) Endstates 5.02E-05 4.14E+00 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 46 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

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

-per yr (50-miles) (50-miles) 1 No Containment Failure 3.86E+03 2.07E-05 7.98E-02 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 3.41E-06 1.31E-01 3b Large Isolation Failures (Liner Breach) 1.35E+05 3.41E-07 4.60E-02 4 Small Isolation Failure - Failure-To-Seal (Type B test) -- 0.00E+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) -- 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 1.39E+05 1.26E-05 1.74E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 2.37E-07 2.68E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 1.19E-05 2.24E+00 8c Containment Bypassed (Late SGTR) 1.39E+05 2.66E-06 3.69E-01 All Containment Event Tree (CET) Endstates 5.18E-05 4.88E+00 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 47 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 1.06E-01 9.83E-02 9.24E-02 Large Containment Isolation Failure (Failure-2 To-Close) 1.70E-03 1.30E-03 1.30E-03 3a Small Isolation Failures (Liner Breach) 2.94E-02 9.81E-02 1.52E-01 3b Large Isolation Failures (Liner Breach) 1.03E-02 3.43E-02 5.32E-02 Small Isolation Failure - Failure-To-Seal 4

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

(Type C Test) 0.00E+00 0.00E+00 0.00E+00 Containment isolation Failures (Dependent 6

failures, Personnel Errors) 0.00E+00 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 1.05E+00 1.05E+00 1.05E+00 8a Containment Bypassed (ISLOCA) 2.68E-01 2.68E-01 2.68E-01 8b Containment Bypassed (Early SGTR) 2.16E+00 2.16E+00 2.16E+00 8c Containment Bypassed (Late SGTR) 3.64E-01 3.64E-01 3.64E-01 Totals 3.99 4.08 4.14 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 48 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

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 9.15E-02 8.50E-02 7.98E-02 Large Containment Isolation Failure (Failure-2 To-Close) 1.76E-03 1.35E-03 1.35E-03 3a Small Isolation Failures (Liner Breach) 2.54E-02 8.48E-02 1.31E-01 3b Large Isolation Failures (Liner Breach) 8.91E-03 2.97E-02 4.60E-02 Small Isolation Failure - Failure-To-Seal 4

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

(Type C Test) 0.00E+00 0.00E+00 0.00E+00 Containment isolation Failures (Dependent 6

failures, Personnel Errors) 0.00E+00 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 1.74E+00 1.74E+00 1.74E+00 8a Containment Bypassed (ISLOCA) 2.68E-01 2.68E-01 2.68E-01 8b Containment Bypassed (Early SGTR) 2.24E+00 2.24E+00 2.24E+00 8c Containment Bypassed (Late SGTR) 3.69E-01 3.69E-01 3.69E-01 Totals 4.75 4.82 4.88 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 49 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval ANALYSIS FILE: 17670-0001, Rev. 2, Appendix A A.1.0 CLIENT Nuclear Management Company - 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 Methodology The internal flooding analysis was performed in May 1993 and is documented in Reference A1. 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.

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 50 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

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 Flooding 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 A1.

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 (CSR/Non-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.

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 51 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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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-1. Internal Flooding Results Flood- Plant Area Where Flooding Primary Equipment Damaged PSA Assumed Core ing Originates Flooding Equipment Damage Zone Source Damage Frequency Frequency (1/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 / 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 < 1.0E-07 Note 1.

Diesel Generator G1 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 < 1.0E-07 Note 2.

Area Water 5 Diesel Generator Rooms, Service Loss of instrument air 1.7E-05 < 1.0E-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. Random failure of systems not affected by flooding in this area is 6.2E-5.

2. Random failure of systems not affected by flooding in this area is 1.0E-2.

3. Random failure of systems not affected by flooding in this area is 1.0E-3.

Summing the core damage contributions identified in the last column of the table indicates a total CDF contribution due to internal flooding of 1.1E-5. 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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 52 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

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-15-1/2 years Type A ILRT 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 (IAS) 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.

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 53 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

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 IAS.

• 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).

17670-0001 PB ILRT Rev 2.doc

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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 PBl-1 1.93E-07 1.48 All other sequences 9.98E-07 7.33 5

Total 1.31E-05 100.00 Where:

Sequence PBO-2 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.

Sequence PBO-4 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.

Sequence PBO-3 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.

Sequence PB2-3 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.

Sequence TR3A-15 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 1.31E-05 for PB. It also indicates that upgrades pending at the time of the analysis would reduce the contribution to 1.1E-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.

17670-0001 PB ILRT Rev 2.doc

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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 Y0l, 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 5.1E-05/year. This was subsequently revised per Attachment A-1 to 1.24E-5/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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 56 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

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 A1, 1.1E-05/yr

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

3. The CDF contribution due to Fire events from Attachment A-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

A1. 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.

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 57 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

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, A10. EPRI NP-6041-SL, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, 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 1.1E-05/yr (internal flooding) + 1.31E-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.11.2 Risk Impact for the New Surveillance Intervals 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.

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 58 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test 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-10 years ILRT interval, the percentage contribution to total dose rate from EPRI' s accident Classes 3a and 3b is calculated as follows:

Per_CtdComb-10 = [(Class_3a_DoseComb-10 + Class_3b_DoseComb-10)/ Total_DoseComb-10]*100%

Where Per_CtdComb-10 = Combined internal and external events percentage contribution to total dose rate from EPRIs accident Classes 3a and 3b given a 1-in-10 years ILRT interval Class_3a_DoseComb-10 = combined internal and external events EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= Class_3a_DoseInternal-10 + Class_3a_DoseExternal-10 Class_3b_DoseComb-10 = combined internal and external events EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= Class_3b_DoseInternal-10 + Class_3b_DoseExternal-10 Class_3a_DoseInternal-10 = internal events EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= 9.81E-02for Unit 1 (Table 5a)

And

= 8.48E-02for Unit 2 (Table 5b)

Class_3b_DoseInternal-10 = internal events EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= 3.43E-02for Unit 1 (Table 5a)

And

= 2.97E-02for Unit 2 (Table 5b)

Class_3a_DoseExternal-10 = internal flooding/external events EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= 7.10E-02 for Unit 1 (Table A-6)

And

= 5.95E-02 for Unit 2 (Table A-6)

Class_3b_DoseExternal-10 = internal flooding/external events EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= 2.48E-02for Unit 1 (Table A-6)

And

= 2.08E-02 for Unit 2 (Table A-6)

Total_DoseComb-10 = total combined internal and external events dose rate for all EPRI classes given a 1-in-10 years ILRT interval Total_DoseInternal-10 + Total_DoseExternal-10 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 59 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Total_DoseInternal-10 = total internal events dose rate for all EPRI classes given a 1-in-10 years ILRT interval

= 4.08 person-rem/yr for Unit 1 (Table 5a)

And

= 4.82 person-rem/yr for Unit 2 (Table 5b)

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

= 2.95 person-rem/yr for Unit 1 (Table A-6)

And

= 3.38 person-rem/yr for Unit 2 (Table A-6)

Therefore, Per_CtdComb-10 = [(Class_3a_DoseComb-10 + Class_3b_DoseComb-10)/ Total_DoseComb-10]*100%

= [({9.81E-02 + 7.10E-02} + {3.43E-02 + 2.48E-02})/(4.08 + 2.95)]

• 100%

= 3.2% for Unit 1 And

= [({8.48E-02+ 5.95E-02} + {2.97E-02+ 2.08E-02})/ (4.82 + 3.38)]

• 100%

= 2.4% for Unit 2 The percentage contribution to total dose rate from EPRI' s accident Classes 3a and 3b based on the proposed 1-in-15-1/2 years ILRT interval is calculated as follows:

Per_CtdComb-15 = [(Class_3a_DoseComb-15 + Class_3b_DoseComb-15)/ Total_DoseComb-15]*100%

Where Per_CtdComb-15 = Combined internal and external events percentage contribution to total dose rate from EPRIs accident Classes 3a and 3b given a 1-in-15-1/2 years ILRT interval Class_3a_DoseComb-15 = combined internal and external events EPRI accident Class 3a dose rate given a 1-in-15-1/2 years ILRT interval

= Class_3a_DoseInternal-15 + Class_3a_DoseExternal-15 Class_3b_DoseComb-15 = combined internal and external events EPRI accident Class 3b dose rate given a 1-in-15-1/2 years ILRT interval

= Class_3b_DoseInternal-15 + Class_3b_DoseExternal-15 Class_3a_DoseInternal-15 = internal events EPRI accident Class 3a dose rate given a 1-in-15-1/2 years ILRT interval

= 1.52E-01for Unit 1 (Table 6a)

And

= 1.31E-01 for Unit 2 (Table 6b)

Class_3b_DoseInternal-15 = internal events EPRI accident Class 3b dose rate given a 1-in-15-1/2 years ILRT interval

= 5.32E-02 for Unit 1 (Table 6a)

And

= 4.60E-02 for Unit 2 (Table 6b) 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 60 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class_3a_DoseExternal-15 = internal flooding/external events EPRI accident Class 3a dose rate given a 1-in-15-1/2 years ILRT interval

= 1.52E-01 for Unit 1 (Table A-7)

And

= 9.22E-02 for Unit 2 (Table A-7)

Class_3b_DoseExternal-15 = internal flooding/external events EPRI accident Class 3b dose rate given a 1-in-15-1/2 years ILRT interval

= 3.85E-02 for Unit 1 (Table A-7)

And

= 3.23E-02 for Unit 2 (Table A-7)

Total_DoseComb-15 = total combined internal and external events dose rate for all EPRI classes given a 1-in-15-1/2 years ILRT interval Total_DoseInternal-15 + Total_DoseExternal-15 Total_DoseInternal-15 = total internal events dose rate for all EPRI classes given a 1-in-15-1/2 years ILRT interval

= 4.14 person-rem/yr for Unit 1 (Table 6a)

And

= 4.88 person-rem/yr for Unit 2 (Table 6b)

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

= 3.04 person-rem/yr for Unit 1 (Table A-7)

And

=3.42 person-rem/yr for Unit 2 (Table A-7)

Therefore, Per_CtdComb-15 = [(Class_3a_DoseComb-15 + Class_3b_DoseComb-15)/ Total_DoseComb-15}*100%

= [({1.52E-01 + 1.52E-01 } + {5.32E-02 + 3.85E-02 })/(4.14 + 3.04)]

• 100%

= 5.5% for Unit 1 And

= [({1.31E-01 + 9.22E-02 } + {4.60E-02 + 3.23E-02 })/(4.88 + 3.42)]

• 100%

= 3.6% 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:

Incr_CtdComb_10-15 = [(Total_DoseComb Total_DoseComb-10) / Total_DoseComb-10]

• 100%

Where:

Incr_CtdComb_10-15 = combined internal and external events percent change from 1-in-10 years ILRT interval to 1-in-15-1/2 years ILRT interval Total_DoseComb-15 = Total combined internal and external events dose rate for all EPRI classes given a 1-in 15-1/2 years ILRT interval

= Total_DoseInternal-15 + Total_DoseExternal-15 Total_DoseInternal-15 = total internal events dose rate for all EPRI classes given a 1-in-15-1/2 years ILRT interval 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 61 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

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

= 4.14 person-rem/yr for Unit 1 (Table 6a)

And

= 4.88 person-rem/yr for Unit 2 (Table 6b)

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

= 3.04 person-rem/yr for Unit 1 (Table A-7)

And

= 3.42 person-rem/yr for Unit 2 (Table A-7)

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

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

= 4.08 person-rem/yr for Unit 1 (Table 5a)

And

= 4.82 person-rem/yr for Unit 2 (Table 5b)

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

= 2.95 person-rem/yr for Unit 1 (Table A-6)

And

= 3.38 person-rem/yr for Unit 2 (Table A-6)

Therefore, Incr_CtdComb_10-15 = [(Total_DoseComb Total_DoseComb-10) / Total_DoseComb-10]

• 100%

= [({4.14 + 3.04} - {4.08 + 2.95}) / (4.08 + 2.95}]* 100%

= 2.2% for Unit 1 And

= [({4.88 + 3.42} - {4.82 + 3.38}) / (4.82 + 3.38)]* 100%

= 1.2% 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 2.2% for Unit 1 and 1.2% 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-15-1/2 years dose rate is as follows:

Incr_CtdComb_3-15 = [(Total_DoseComb Total_DoseComb) / Total_DoseComb]

• 100%

Where:

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

= Total_DoseInternal-15 + Total_DoseExternal-15 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 62 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Total_DoseInternal-15 = total internal events dose rate for all EPRI classes given a 1-in-15-1/2 years ILRT interval

= 4.14 person-rem/yr for Unit 1 (Table 6a)

And

= 4.88 person-rem/yr for Unit 2 (Table 6b)

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

= 3.04 person-rem/yr for Unit 1 (Table A-7)

And

= 3.42 person-rem/yr for Unit 2 (Table A-7)

Total_DoseComb = 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_DoseInternal = total internal events dose rate for all EPRI classes given a 3-in-10 years ILRT interval

= 3.99 person-rem/yr for Unit 1 (Table 4a)

And

= 4.75 person-rem/yr for Unit 2 (Table 4b)

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

= 2.89 person-rem/yr for Unit 1 (Table A-5a)

And

= 3.33 person-rem/yr for Unit 2 (Table A-5b)

Delta_DoseComb 3-15 = (Total_DoseComb Total_DoseComb)

= ({4.14 + 3.04} - {3.99 + 2.89})

= 0.30 for Unit 1 And

= ({4.88 + 3.42} - {4.75 + 3.33})

= 0.22 for Unit 2 Therefore, Incr_CtdComb_3-15 = [(Total_DoseComb Total_DoseComb) / Total_DoseComb]

• 100%

= [({4.14 + 3.04} - {3.99 + 2.89}) / {3.99 + 2.89}]* 100%

= 4.4% for Unit 1 And

= [({4.88 + 3.42} - {4.75 + 3.33}) / {4.75 + 3.33}]* 100%

= 2.8% 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 3-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 2.8% for Unit 2. These values are not significant increases in risk.

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 63 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A.11.3 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-15-1/2-years.

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

LERFCombined10-15 = Class-3bCombined15 - Class-3bCombined10 Where:

LERFCombined10-15 = the combined internal and external events change in LERF from 1-in-10 years ILRT interval to 1-in-15-1/2 years ILRT interval Class-3bCombined15 = the combined internal and external frequency of EPRI accident Class 3b given a 1-in-15-1/2 years ILRT Interval

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

= 3.94E-07 for Unit 1 (Table 6a)

And

= 3.41E-07 for Unit 2 (Table 6b)

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

= 2.85E-07 for Unit 1 (Table A-7)

And

= 2.39E-07 for Unit 2 (Table A-7)

Class-3bCombined10 = the combined internal and external frequency of EPRI accident Class 3b given a 1-in-10 years ILRT Interval

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

= 2.54E-07/ry for Unit 1 (Table 5a)

And

= 2.20E-07/ry for Unit 2 (Table 5b)

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

= 1.84E-07/ry for Unit 1 (Table A-6)

And

= 1.54E-07/ry for Unit 2 (Table A-6)

Therefore, 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 64 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval LERFCombined10-15 = Class-3bCombined15 - Class-3bCombined10

= (3.94E-07 + 2.85E-07) - (2.54E-07 + 1.84E-07)

= 2.41E-07/ry for Unit 1 And

= (3.41E-07 + 2.39E-07) - (2.20E-07 + 1.54E-07)

= 2.06E-07/ry for Unit 2 LERFCombined 3-15 = Class-3bCombined15 - Class-3bCombined Where:

LERFCombined 3-15 = 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 Class-3bCombined15 = the combined internal and external frequency of EPRI accident Class 3b given a 1-in-15-1/2 years ILRT Interval

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

= 3.94E-07 for Unit 1 (Table 6a)

And

= 3.41E-07 for Unit 2 (Table 6b)

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

= 2.85E-07 for Unit 1 (Table A-7)

And

= 2.39E-07 for Unit 2 (Table A-7)

Class-3bCombined10 = the combined internal and external frequency of EPRI accident Class 3b given a 3-in-10 years ILRT Interval

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

= 7.63E-08/ry for Unit 1 (Table 4a)

And

= 6.59E-08/ry for Unit 2 (Table 4b)

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

= 5.52E-08/ry for Unit 1 (Table A-4)

And

= 4.62E-08/ry for Unit 2 (Table A-4) 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 65 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Therefore, LERFCombined3-15 = Class-3bCombined15 - Class-3bCombined

= (3.94E-07 + 2.85E-07) - (7.63E-08 + 5.52E-08)

= 5.48E-07/ry for Unit 1 And

= (3.41E-07 + 2.39E-07) - (6.59E-08 + 4.62E-08)

= 4.68E-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 2.03E-05 (1.18E-05 +

8.51E-6) for Unit 1 and 2.07E-05 (1.22E-05 + 8.54E-06) for Unit 2 (Section 11.1.1); these values of LERF would be in Region 1 of Figure 4 of Regulatory Guide 1.174 and would not allow any changes that would increase risk. However, these are conservative LERF values as discussed in Appendix C. More realistic values for the combined internal/external events LERF are 6.90E-06 for Unit 1 and 7.05E-06 for Unit 2 as described in the sensitivity analysis provided in Appendix C. 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-15-1/2 years is 2.95E-07 for Unit 1 and 2.51E-07 for Unit 2.

Because Reference 9 defines small changes in LERF as below 1E-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 from the realistic 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 6.70E-07 for Unit 1 and 5.71E-07/ry for Unit 2 falls into Region II, Small Change in Risk, of the acceptance guidelines in NRC Regulatory Guide 1.174.

A.11.4 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)/ CDFTotal ]

• 100%

For the combined internal and external events 1-in-10 years ILRT interval:

CCFPComb10 = [1 - (Class_1Comb-10 + Class_3a Comb-10)/ CDFTotal-Comb ]

• 100%

Where; Class_1 Comb-10 = combined internal and external events frequency of EPRI accident Class 1 given a 1-in-10 years ILRT interval

= Class_1 Internal-10 + Class_1 External-10 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 66 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class_1 Internal-10 = internal events frequency of EPRI accident Class 1 given a 1-in-10 years ILRT interval

= 2.55E-05/ry for Unit 1 (Table 5a)

And

= 2.20E-05/ry for Unit 2 (Table 5b)

Class_1 External-10 = External events frequency of EPRI accident Class 1 given a 1-in-10 years ILRT interval

= 1.84E-05/ry for Unit 1 (Table A-6)

And

= 1.54E-05/ry for Unit 2 (Table A-6)

Class_3a Comb-10 = 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-10 Class_3aInternal-10 = internal events frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval

= 2.54E-06/ry for Unit 1 (Table 5a)

And

= 2.20E-06/ry for Unit 2 (Table 5b)

Class_3a External-10 = External events frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval

= 1.84E-06/ry for Unit 1 (Table A-6)

And

= 1.54E-06/ry for Unit 2 (Table A-6)

CDFTotal-Comb = 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 And

= 5.18E-05/ry + 3.63E-05/ry for Unit 2 (Tables 4b and A-6)

= 8.81E-05/ry for Unit 2 Therefore, CCFPComb10 = [1 - (Class_1Comb-10 + Class_3a Comb-10)/ CDFTotal-Comb ]

• 100%

= [1-({2.54E-05 + 1.84E-05} + {2.54E-06 + 1.84E-06})/ 8.65E-5]

• 100%

= 44.20% for Unit 1 And

= [1-({2.20E-05 + 1.54E-05} + {2.20E-06 + 1.54E-06})/ 8.81E-5]

• 100%

= 53.27% for Unit 2 For the combined internal and external events 1-in-15-1/2 years ILRT interval:

CCFPComb15 = [1 - (Class_1Comb-15 + Class_3a Comb-15)/ CDFTotal-Comb ]

• 100%

Where; 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 67 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Class_1 Comb-15 = combined internal and external events frequency of EPRI accident Class 1 given a 1-in-15-1/2 years ILRT interval

= Class_1 Internal-15 + Class_1 External-15 Class_1 Internal-15 = internal events frequency of EPRI accident Class 1 given a 1-in-15-1/2 years ILRT interval

= 2.39E-05/ry for Unit 1 (Table 6a)

And

= 2.07E-05/ry for Unit 2 (Table 6b)

Class_1 External-15 = External events frequency of EPRI accident Class 1 given a 1-in-15-1/2 years ILRT interval

= 1.62E-05/ry for Unit 1 (Table A-7)

And

= 1.45E-05/ry for Unit 2 (Table A-7)

Class_3a Comb-15 = combined internal and external events frequency of EPRI accident Class 3a given a 1-in-15-1/2 years ILRT interval

= Class_3a Internal-15 + Class_3a External-15 Class_3aInternal-15 = internal events frequency of EPRI accident Class 3a given a 1-in-15-1/2 years ILRT interval

= 3.94E-06/ry for Unit 1 (Table 6a)

And

= 3.41E-06/ry for Unit 2 (Table 6b)

Class_3a External-15 = External events frequency of EPRI accident Class 3a given a 1-in-15-1/2 years ILRT interval

= 3.94E-06/ry for Unit 1 (Table A-7)

And

= 2.39E-06/ry for Unit 2 (Table A-7)

CDFTotal-Comb = 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-5/ry for Unit 1 And

= 5.18E-05/ry + 3.63E-05/ry for Unit 2 (Tables 4b and A-6)

= 8.81E-05/ry for Unit 2 Therefore, CCFPComb15 = [1 - (Class_1Comb-15 + Class_3a Comb-15)/ CDFTotal-Comb ]

• 100%

= [1-({2.39E-05 + 1.62E-05} + {3.94E-06 + 3.94E-06})/ 8.65E-5]

• 100%

= 44.48% for Unit 1 And

= [1-({2.07E-05 + 1.45E-05} + {3.41E-06 + 2.39E-06})/ 8.81E-5]

• 100%

= 53.50% for Unit 2 Therefore, the change in the combined internal and external events conditional containment failure probability from 1-in-10 years to 1-in-15-1/2 years is:

CCFPCombined 10-15 = CCFPCombined 15 - CCFPCombined 10 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 68 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

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

= 44.48% - 44.20% = 0.28% for Unit 1 And

= 53.50% - 53.27% = 0.23% 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_1Comb + Class_3a Comb)/ CDFTotal-Comb ]

• 100%

Where; Class_1 Comb = combined internal and external events frequency of EPRI accident Class 1 given a 3-in-10 ILRT interval

= Class_1_Frequencyl + Class_1 External Class_1_Frequency = internal events frequency of EPRI accident Class 1 given a 3-in-10 years ILRT interval

= 2.74E-05/ry for Unit 1 (Table 4a)

And

= 2.37E-05/ry for Unit 2 (Table 4b)

Class_1 External = External events frequency of EPRI accident Class 1 given a 3-in-10 years ILRT interval

= 1.98E-05/ry for Unit 1 (Table A-4)

And

= 1.66E-05/ry for Unit 2 (Table A-4)

Class_3a Comb = combined internal and external events frequency of EPRI accident Class 3a given a 3-in-10 years ILRT interval

= Class_3a_Frequency + Class_3a External Class_3a_Frequency = internal events frequency of EPRI accident Class 3a given a 1-in-15-1/2 years ILRT interval

= 7.63E-07/ry for Unit 1 (Table 4a)

And

= 6.59E-07/ry for Unit 2 (Table 4b)

Class_3a External = External events frequency of EPRI accident Class 3a given a 1-in-15-1/2 years ILRT interval

= 5.52E-07/ry for Unit 1 (Table A-4)

And

= 4.62E-07/ry for Unit 2 (Table A-4)

CDFTotal-Comb = 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-5/ry for Unit 1 And

= 5.18E-05/ry + 3.63E-05/ry for Unit 2 (Tables 4b and A-6)

= 8.81E-05/ry for Unit 2 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 69 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Therefore, CCFPComb = [1 - (Class_1Comb + Class_3a Comb)/ CDFTotal-Comb ]

• 100%

= [1-({2.74E-05 + 1.98E-05} + {7.63E-07 + 5.52E-07})/ 8.65E-5]

• 100%

= 43.85% for Unit 1 And

= [1-({2.37E-05 + 1.66E-05} + {6.59E-07 + 4.62E-07})/ 8.81E-5]

• 100%

= 52.97% for Unit 2 CCFPCombined 3-15 = CCFPCombined 15 - CCFPCombined

= 44.48% - 43.85% = 0.63% for Unit 1 And

= 53.50% - 52.97% = 0.53% for Unit 2 A.12.0 COMPUTER INPUT AND OUTPUT None A.13.0

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 The combined internal/external events contribution to LERF at Point Beach is 2.03E-05 (1.18E-05 +

8.51E-6) for Unit 1 and 2.07E-05 (1.22E-05 + 8.54E-06) for Unit 2 (Section 11.1.1); these values of LERF would be in Region 1 of Figure 4 of Regulatory Guide 1.174 and would not allow any changes that would increase risk. However, these are conservative LERF values as discussed in Appendix C. More realistic values for the combined internal/external events LERF are 6.90E-06 for Unit 1 and 7.05E-06 for Unit 2 as described in the sensitivity analysis provided in Appendix C. 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-15-1/2 years is 2.95E-07 for Unit 1 and 2.51E-07 for Unit 2.

Because Reference 9 defines small changes in LERF as below 1E-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 from the realistic 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 6.70E-07 for Unit 1 and 5.71E-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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 70 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 2.2% for Unit 1 and 1.2% 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-15-1/2 years is 0.28% and 0.23% 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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 71 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table A-3 Point Beach Dominant IPEEE Fire Events - Core Damage Frequency Results CDF Compartment Event Description 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 lB-32 The primary contributor is the Unit 1 LCV-112B 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-112B 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 Operators 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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 72 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 2.51E-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 G01. 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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 73 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 Unit 1 IF/EE Unit 2 IF/EE PB  % of Unit 1  % of Unit 2 Class Description Freq. Freq.

STC IF/EE CDF IF/EE CDF (per yr) (per yr) 1 No Containment Failure 1 1.98E-05 54.6% 1.66E-05 45.8%

2 Large Containment Isolation Failure (Failure-To-Close) 6 1.09E-08 0.0% 1.09E-08 0.0%

3a Small Isolation Failures (Liner Breach) - 5.52E-07 1.5% 4.62E-07 1.3%

3b Large Isolation Failures (Liner Breach) - 5.52E-08 0.2% 4.62E-08 0.1%

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

5 Small Isolation Failure - Failure-To-Seal (Type C Test) - 0.00E+00 0.0% 0.00E+00 0.0%

6 Containment isolation Failures (Dependent failures, Personnel - 0.00E+00 0.00E+00 Errors) 0.0% 0.0%

7a Severe Accident Phenomena Induced Failure - Late Rupture 4 7.31E-06 14.56% 1.23E-05 23.74%

7b Severe Accident Phenomena Induced Failure - Late Basemat 2 2.31E-07 0.46% 2.13E-07 0.41%

7c Severe Accident Phenomena Induced Failure -Early Liner 5 2.24E-08 0.04% 2.06E-08 0.04%

8a Containment Bypassed (ISLOCA) 7 1.71E-07 0.5% 1.66E-07 0.5%

8b Containment Bypassed (Early SGTR) 8 8.32E-06 22.9% 8.35E-06 23.0%

8c Containment Bypassed (Late SGTR) 3 1.90E-06 5.2% 1.86E-06 5.1%

All Containment Event Tree (CET) Endstates 3.63E-05 100.0% 3.63E-05 100.0%

LERF Frequency (STCs 5, 6, 7, 8) 8.51E-06 8.54E-06 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 74 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE A-5a - Mean Consequence Measures for IF/EE for 3-in-10 Years ILRT Interval Unit 1 IF/EE Unit 1 Unit 1 Person-Rem Class Description Frequency - per Person-Rem/yr (50-miles) yr. (50-miles) 1 No Containment Failure 3.86E+03 1.98E-05 7.65E-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 5.52E-07 2.13E-02 3b Large Isolation Failures (Liner Breach) 1.35E+05 5.52E-08 7.45E-03 4 Small Isolation Failure - Failure-To-Seal (Type B test) 0.00E+00 0.00E+00 0.00E+00 5 Small Isolation Failure - Failure-To-Seal (Type C Test) 0.00E+00 0.00E+00 0.00E+00 6 Containment isolation Failures (Dependent failures, Personnel Errors) 0.00E+00 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 1.39E+05 5.47E-06 7.60E-01 8a Containment Bypassed (ISLOCA) 1.13E+06 1.71E-07 1.94E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 8.32E-06 1.56E+00 8c Containment Bypassed (Late SGTR) 1.39E+05 1.90E-06 2.64E-01 All Containment Event Tree (CET) Endstates 3.63E-05 2.89E+00 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 75 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE A-5b - Mean Consequence Measures for IF/EE for 3-in-10 Years ILRT Interval Unit 2 Unit 2 Person-Rem (50- IF/EE Unit 2 Person-Class Description miles) Frequency - per yr. Rem/yr (50-miles) 1 No Containment Failure 3.86E+03 1.66E-05 6.41E-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 4.62E-07 1.78E-02 3b Large Isolation Failures (Liner Breach) 1.35E+05 4.62E-08 6.24E-03 4 Small Isolation Failure - Failure-To-Seal (Type B test) 0.00E+00 0.00E+00 0.00E+00 5 Small Isolation Failure - Failure-To-Seal (Type C Test) 0.00E+00 0.00E+00 0.00E+00 6 Containment isolation Failures (Dependent failures, Personnel Errors) 0.00E+00 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 1.39E+05 8.80E-06 1.22E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 1.66E-07 1.87E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 8.35E-06 1.57E+00 8c Containment Bypassed (Late SGTR) 1.39E+05 1.86E-06 2.59E-01 All Containment Event Tree (CET) Endstates 3.63E-05 3.33E+00 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 76 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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)

Unit 1 Unit 2 Person- IF/EE Unit 1 IF/EE Unit 2 Person- Person-Class Description Rem (50- Frequency - Frequency -

Rem/yr Rem/yr miles) per yr. per yr.

(50-miles) (50-miles) 1 No Containment Failure 3.86E+03 1.84E-05 7.11E-02 1.54E-05 5.95E-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 1.84E-06 7.10E-02 1.54E-06 5.95E-02 3b Large Isolation Failures (Liner Breach) 1.35E+05 1.84E-07 2.48E-02 1.54E-07 2.08E-02 4 Small Isolation Failure - Failure-To-Seal (Type B test) 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 Small Isolation Failure - Failure-To-Seal (Type C Test) 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 Containment isolation Failures (Dependent failures, 0.00E+00 Personnel Errors) 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 1.39E+05 5.47E-06 7.60E-01 8.80E-06 1.22E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 1.71E-07 1.94E-01 1.66E-07 1.87E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 8.32E-06 1.56E+00 8.35E-06 1.57E+00 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 2.95E+00 3.63E-05 3.38E+00 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 77 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE A Dose Rate by Accident Class for 1-in-15-1/2 Years ILRT Interval (IF/Ext Events)

Unit 1 Unit 2 Person- IF/EE Unit 1 IF/EE Unit 2 Person- Person-Class Description Rem (50- Frequency - Frequency -

Rem/yr Rem/yr miles) per yr. per yr.

(50-miles) (50-miles) 1 No Containment Failure 3.86E+03 1.62E-05 6.26E-02 1.45E-05 5.59E-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 3.94E-06 1.52E-01 2.39E-06 9.22E-02 3b Large Isolation Failures (Liner Breach) 1.35E+05 2.85E-07 3.85E-02 2.39E-07 3.23E-02 4 Small Isolation Failure - Failure-To-Seal (Type B test) 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 Small Isolation Failure - Failure-To-Seal (Type C Test) 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 Containment isolation Failures (Dependent failures, 0.00E+00 Personnel Errors) 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 1.39E+05 5.47E-06 7.60E-01 8.80E-06 1.22E+00 8a Containment Bypassed (ISLOCA) 1.13E+06 1.71E-07 1.94E-01 1.66E-07 1.87E-01 8b Containment Bypassed (Early SGTR) 1.88E+05 8.32E-06 1.56E+00 8.35E-06 1.57E+00 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 3.04E+00 3.63E-05 3.42E+00 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 78 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 7.65E-02 7.11E-02 6.26E-02 Large Containment Isolation Failure (Failure-2 To-Close) 1.23E-03 1.23E-03 1.23E-03 3a Small Isolation Failures (Liner Breach) 2.13E-02 7.10E-02 1.52E-01 3b Large Isolation Failures (Liner Breach) 7.45E-03 2.48E-02 3.85E-02 Small Isolation Failure - Failure-To-Seal 4

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

(Type C Test) 0.00E+00 0.00E+00 0.00E+00 Containment isolation Failures (Dependent 6

failures, Personnel Errors) 0.00E+00 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 7.60E-01 7.60E-01 7.60E-01 8a Containment Bypassed (ISLOCA) 1.94E-01 1.94E-01 1.94E-01 8b Containment Bypassed (Early SGTR) 1.56E+00 1.56E+00 1.56E+00 8c Containment Bypassed (Late SGTR) 2.64E-01 2.64E-01 2.64E-01 Totals 2.89 2.95 3.04 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 79 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 6.41E-02 5.95E-02 5.59E-02 Large Containment Isolation Failure (Failure-2 To-Close) 1.23E-03 1.23E-03 1.23E-03 3a Small Isolation Failures (Liner Breach) 1.78E-02 5.95E-02 9.22E-02 3b Large Isolation Failures (Liner Breach) 6.24E-03 2.08E-02 3.23E-02 Small Isolation Failure - Failure-To-Seal 4

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

(Type C Test) 0.00E+00 0.00E+00 0.00E+00 Containment isolation Failures (Dependent 6

failures, Personnel Errors) 0.00E+00 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 1.22E+00 1.22E+00 1.22E+00 8a Containment Bypassed (ISLOCA) 1.87E-01 1.87E-01 1.87E-01 8b Containment Bypassed (Early SGTR) 1.57E+00 1.57E+00 1.57E+00 8c Containment Bypassed (Late SGTR) 2.59E-01 2.59E-01 2.59E-01 Totals 3.33 3.38 3.42 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 80 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / 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-10 1-per-10 years 1-per-15-1/2 years years ILRT ILRT ILRT 1 No Containment Failure 1.82E-01 1.69E-01 1.55E-01 Large Containment Isolation Failure (Failure-2 To-Close) 2.93E-03 2.53E-03 2.53E-03 3a Small Isolation Failures (Liner Breach) 5.07E-02 1.69E-01 3.04E-01 3b Large Isolation Failures (Liner Breach) 1.78E-02 5.92E-02 9.17E-02 Small Isolation Failure - Failure-To-Seal 4

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

(Type C Test) 0.00E+00 0.00E+00 0.00E+00 Containment isolation Failures (Dependent 6

failures, Personnel Errors) 0.00E+00 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 1.81E+00 1.81E+00 1.81E+00 8a Containment Bypassed (ISLOCA) 4.61E-01 4.61E-01 4.61E-01 8b Containment Bypassed (Early SGTR) 3.73E+00 3.73E+00 3.73E+00 8c Containment Bypassed (Late SGTR) 6.28E-01 6.28E-01 6.28E-01 Totals 6.88 7.03 7.18 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 81 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 1.56E-01 1.44E-01 1.36E-01 Large Containment Isolation Failure (Failure-2 To-Close) 2.99E-03 2.58E-03 2.58E-03 3a Small Isolation Failures (Liner Breach) 4.33E-02 1.44E-01 2.24E-01 3b Large Isolation Failures (Liner Breach) 1.51E-02 5.05E-02 7.83E-02 Small Isolation Failure - Failure-To-Seal 4

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

(Type C Test) 0.00E+00 0.00E+00 0.00E+00 Containment isolation Failures (Dependent 6

failures, Personnel Errors) 0.00E+00 0.00E+00 0.00E+00 7 Severe Accident Phenomena Induced Failure 2.97E+00 2.97E+00 2.97E+00 8a Containment Bypassed (ISLOCA) 4.55E-01 4.55E-01 4.55E-01 8b Containment Bypassed (Early SGTR) 3.81E+00 3.81E+00 3.81E+00 8c Containment Bypassed (Late SGTR) 6.28E-01 6.28E-01 6.28E-01 Totals 8.08 8.20 8.30 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 82 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Attachment A-1 Email from Paul Knoespel to Ed Krantz:

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 83 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval From: Knoespel, Paul [Paul.Knoespel@nmcco.com]

Sent: Friday, August 11, 2006 12:02 PM To: Krantz, Ed

Subject:

1998 "New" Fire Frequency for Point Beach Ed, The total CDF due to fire events at Point Beach is 1.24E-05/yr. This is based on a 1998 study which used conditional core damage probabilities from the 1996 version (later known as Revision 2.0) of the internal events PRA model. Ill let you know if I come up with some reasoning why the fire CDF should (or should not) result in a similar distribution of releases.

Paul Paul D. Knoespel Site PRA Lead Point Beach Nuclear Plant Nuclear Management Company, LLC 920-755-7508 paul.knoespel@nmcco.com 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 84 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval ANALYSIS FILE: 17670-0001, Rev. 2, Appendix B B.1.0 CLIENT Nuclear Management Company - 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-10 years to 1-in-15-1/2 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.

17670-0001 PB ILRT Rev 2.doc

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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

• 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 B1, 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 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 2.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 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.

17670-0001 PB ILRT Rev 2.doc

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SUBJECT:

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

B.9.0 MAJOR ASSUMPTIONS:

1. As indicated in the NRCs 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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 88 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 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 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 potion of the containment cylinder and dome and the inaccessible portion of the containment basemat.

The accessible portion of the containment cylinder and dome flaw likelihood is determined as follows:

CCDF = NFaila / (NPlants

• TExpo) 17670-0001 PB ILRT Rev 2.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)

NFailja = 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 B1), 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-1.

B.11.3 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-15-1/2 years interval, per the Calvert Cliffs methodology (Reference B1). The results of Step 2 are used to generate these values as follows:

Accessible portion (visible areas) of the containment cylinder and dome, CCDFlaw3-10 = CCDFRateii for i=1 to 3 CCDFlaw1-10 = CCDFRateii for i=1 to 10 CCDFlaw1-15 = CCDFRateii for i=1 to 15-1/2 Inaccessible portion of the containment Basemat, CBMFlaw3-10 = CBMFRateii for i=1 to 3 CBMFlaw1-10 = CBMFRateii for i=1 to 10 CBMFlaw1-15 = CBMFRateii for i=1 to 15-1/2 Where 17670-0001 PB ILRT Rev 2.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-10 years test interval given accessible portion of the containment CCDFlaw1-15 = increase in flaw likelihood at 1-in-15-1/2 years test interval given accessible portion of the containment CBMFlaw3-10 = increase in flaw likelihood at 3-in-10 years test interval given inaccessible portion of the containment CBMFlaw1-10 = increase in flaw likelihood at 1-in-10 years test interval given inaccessible portion of the containment CBMFlaw1-15 = 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, CCDFlaw3-10 = 1.1%

CCDFlaw1-10 = 6.2%

CCDFlaw1-15 = 16.8%

CBMFlaw3-10 = 0.2%

CBMFlaw1-10 = 1.0%

CBMFlaw1-15 = 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 B1. Per the Calvert Cliffs methodology, the containment failure probability (FP) versus containment pressure (P) is assumed to be an equation of the form:

m*p CCDFP(P) = b

• e Where:

CCDFP(P) = containment cylinder and dome failure probability given containment liner breach 17670-0001 PB ILRT Rev 2.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval m = slope of the containment failure probability 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)) I (154.7-20) m = 5.13E-02 Intercept b, m*p b = CCDFP(100%) / e 5.13E-02

• 154.7 b=1/e b = 3.58E-04 The Point Beach ILRT test pressure is of 61psig (or 75.7 psia) (References B5 and B6). Based on this pressure the likelihood of containment breach in the liner is:

5.13E-02

• 75.7 CCDFP(75.7 psia) = 3.58E-04

• e

= 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-1.

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.11.6 Step 6 - Determine the likelihood of non-detected containment leakage due to the increase in test interval 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 92 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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, CCDLeak3-10 = CCDFlaw3-10

• CCDFPILRT

• CCDVisual CCDLeak1-10 = CCDFlaw1-10

• CCDFPILRT

• CCDVisual CCDLeak1-15 = CCDFlaw1-15

• 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 CCDLeak1-10 = 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 yrs test interval and accessible portion of the containment cylinder and dome CCDFlaw3-10 = increase in flaw likelihood at 3-in-10 years test interval given accessible portion of the containment cylinder and dome = 1.06% (0.0106) (Table B-2)

CBMFlaw1-10 = 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)

CBMFlaw1-15 = 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 (1.74%) (Step 4)

CCDVisual = visual inspection detection failure accessible portion of the containment cylinder and dome

= 0.1 (10%) (Step 5)

Therefore, CCDLeak3-10 = CCDFlaw3-10

• CCDFPILRT

• CCDVisual

= 0.0106

• 0.0174

• 0.1

= 1.85E-05 (.00185%)

CCDLeak1-10 = CCDFlaw1-10

• CCDFPILRT

• CCDVisual

= 0.062

• 0.0174

• 0.1

= 1.08E-04 (.0108%)

CCDLeak1-15 = CCDFlaw1-15

• CCDFPILRT

• CCDVisual

= 0.1678

• 0.0174

• 0.1

= 2.92E-04 (.0292%)

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 93 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Inaccessible portion of the liner (Basemat),

CBMLeak3-10 = CBMFlaw3-10

• CBMFPILRT

• CBMVisual CBMLeak1-10 = CBMFlaw1-10

• CBMFPILRT

• CBMVisual CBMLeak1-15 = CBMFlaw1-15

• CBMFPILRT

• CBMVisual Where:

CBMLeak3-10 = 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 CBMLeak1-15 = likelihood of non-detected containment leakage, given 1-in-15-1/2 years test interval and inaccessible portion of the containment basemat CBMFlaw3-10 = increase in flaw likelihood at 3-in-10 years test interval given inaccessible portion of the containment basemat = 0.18% (0.0018) (Table B-2)

CBMFlaw1-10 = 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)

CBMFlaw1-15 = 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, CBMLeak3-10 = CBMFlaw3-10

• CBMFPILRT

• CBMVisual

= 0.0018

• 0.00174

• 1.0

= 3.09E-06 (0.00031%)

CBMLeak1-10 = CBMFlaw1-10

• CBMFPILRT

• CBMVisual

= 0.0104

• 0.00174

• 1.0

= 1.80E-05 (0.0018%)

CBMLeak1-15 = CBMFlaw1-15

• 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, Total3-10 = CCDLeak3-10 + CBMLeak3-10 Total1-10 = CCDLeak1-10 + CBMLeak1-10 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 94 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Total1-15 = CCDLeak1-15 + CBMLeak1-15

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 Total1-15 = total likelihood of non-detected containment leakage due to corrosion, given 1-in-15-1/2 years test interval CCDLeak3-10 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and accessible portion of the containment cylinder and dome CCDLeak1-10 = 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 CBMLeak3-10 = 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 CBMLeak1-15 = likelihood of non-detected containment leakage, given 1-in-15-1/2 years test interval and inaccessible portion of the containment basemat Therefore, Total3-10 = CCDLeak3-10 + CBMLeak3-10

= 0.00185% + 0.00031%

= 0.00216%

Total1-10 = CCDLeak1-10 + CBMLeak1-10

= 0.0108% + 0.0018%

= 0.0126%

Total1-15 = CCDLeak1-15 + CBMLeak1-15

= 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 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 95 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A 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 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_Freq3-10 = (ProbClass_3b + Liner_3b_Incr3-10) * (CDFTotal - CDFIndep)

Liner_3b_Freq1-10 = (ProbClass_3b_10 + Liner_3b_Incr1-10) * (CDFTotal - CDFIndep)

Liner_3b_Freq1-15 = (ProbClass_3b_15 + Liner_3b_Incr1-15) * (CDFTotal - CDFIndep)

Where:

Liner_3b_Freq3-10 = frequency of EPRI Class 3b due to liner corrosion failure given a 3-in-10 years ILRT interval Liner_3b_Freq1-10 = frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-10 years ILRT interval Liner_3b_Freq1-15 = frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-15-1/2 years ILRT interval ProbClass_3b = probability of large pre-existing containment liner leakage given a 3-in-10 years ILRT frequency = 0.0027 [Section 7.0]

ProbClass_3a_10 = probability of small pre-existing containment liner leakage given a 1-in-10 years ILRT frequency = 0.09 [Section 11.1.5]

ProbClass_3a_15 = probability of small pre-existing containment liner leakage given a 1-in-15-1/2 years ILRT frequency = 0.1395 [Section 11.1.5]

CDFTotal U1 = PB U1 PSA L1 core damage frequency = 5.02E-05/ry [Section 11.1.1]

CDFTotal U2 = PB U2 PSA L1 core damage frequency = 5.18E-05/ry [Section 11.1.1]

CDFIndep U1 = CDF for those individual Unit 1 sequences that may independently cause a LERF, identified as EPRI Class 2, Class 7, and Class 8 sequences = 2.19E-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 = 2.74E-05/yr [Section 11.1.1]

Liner_3b_Incr3-10 = Total3-10

• EPRI_3b_Fraction Liner_3b_Incr1-10 = Total1-10

• EPRI_3b_Fraction Liner_3b_Incr1-15 = Total1-15

• EPRI_3b_Fraction Where Liner_3b_Incr3-10 = Liner corrosion increase in EPRI Class 3b given 3-in-10 years test interval Liner_3b_Incr1-10 = Liner corrosion increase in EPRI Class 3b given 1-in-10 years test interval Liner_3b_Incr1-15 = Liner corrosion increase in EPRI Class 3b given 1-in-15-1/2 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) 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 96 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Total1-10 = total likelihood of non-detected containment leakage due to corrosion, given 1-in-10 years test interval = 0.0126% (Table B-4)

Total1-15 = 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_Incr3-10 = Total3-10

• EPRI_3b_Fraction

= 0.00216%

• 1.0

= 0.00216%

Liner_3b_Incr1-10 = Total1-10

• EPRI_3b_Fraction

= 0.0126%

• 1.0

= 0.0126%

Liner_3b_Incr1-15 = Total1-15

• EPRI_3b_Fraction

= 0.0341%

• 1.0

= 0.0341%

Therefore, Liner_3b_Freq3-10 = (ProbClass_3b + Liner_3b_Incr3-10) * (CDFTotal - CDFIndep)

= (0.0027 + 0.00216%) * (5.02E-05/ry - 2.19E-05/ry)

= 7.69E-08/ry for Unit 1 And

= (0.0027 + 0.00216%) * (5.18E-05/ry - 2.74E-05/ry)

= 6.64E-08/ry for Unit 2 Liner_3b_Freq1-10 = (ProbClass_3b_10 + Liner_3b_Incr1-10) * (CDFTotal - CDFIndep)

= (0.009 + 0.0126%) * (5.02E-05/ry - 2.19E-05/ry)

= 2.58E-07/ry for Unit 1 And

= (0.009 + 0.0126%) * (5.18E-05/ry - 2.74E-05/ry)

= 2.23E-07/ry for Unit 2 Liner_3b_Freq1-15 = (ProbClass_3b_15 + Liner_3b_Incr1-15) * (CDFTotal - CDFIndep)

= (0.01395 + 0.0341%) * (5.02E-05/ry - 2.19E-05/ry)

= 4.04E-07/ry for Unit 2 And

= (0.01395 + 0.0341%) * (5.18E-05/ry - 2.74E-05/ry)

= 3.49E-07/ry for Unit 2 Increase to EPRI class 1 frequencies Liner_1_Freq3-10 = NCF- Class_3a_Frequency - Liner_3b_Freq3-10 Liner_1_Freq1-10 = NCF- Class_3a_Frequency_10 - Liner_3b_Freq1-10 Liner_1_Freq1-15 = NCF- Class_3a_Frequency_15 - Liner_3b_Freq1-15 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 97 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

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

Liner_1_Freq3-10 = frequency of EPRI Class 1 due to liner corrosion failure given a 3-in-10 years ILRT interval Liner_1_Freq1-10 = frequency of EPRI Class 1 due to liner corrosion failure given a 1-in-10 years ILRT interval Liner_1_Freq1-15 = frequency of EPRI Class 1 due to liner corrosion failure given a 1-in-15-1/2 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 = 7.63E-07 for Unit 1 And

= 6.59E-07 for Unit 2 Class_3a_Frequency_10 = frequency of small pre-existing containment liner leakage given a 1 -in-10 years ILRT interval (Section 11.1.5)

Class_3a_Frequency_10 = 2.54E-06/ry for Unit 1 And

= 2.20E-06/ry for Unit 2 Class_3a_Frequency_15 = frequency of small pre-existing containment liner leakage given a 1-in-15-1/2 years ILRT interval (Section 11.1.5)

Class_3a_Frequency_15 = 3.94E-06/ry for Unit 1 And

= 3.41E-06/ry for Unit 2 Liner_3b_Freq3-10 = frequency of EPRI Class 3b due to liner corrosion failure given a 3-in-10 years ILRT interval (from above)

= 7.69E-08/ry for Unit 1 And

= 6.64E-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)

= 2.58E-07/ry for Unit 1 And

= 2.23E-07/ry for Unit 2 Liner_3b_Freq1-15 = frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-15-1/2 years ILRT interval (from above)

= 4.04E-07/ry for Unit 2 And

= 3.49E-07/ry for Unit 2 NCF = Frequency in which containment leakage is at or below maximum allowable Technical Specification leakage.

= 2.83E-05/yr (Unit 1) (Section 11.1.1)

= 2.44E-05/yr (Unit 2) (Section 11.1.1)

Therefore, 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 98 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Liner_1_Freq3-10 = NCF- Class_3a_Frequency - Liner_3b_Freq3-10

=2.83E 7.63E 7.69E-08

=2.740E-05/ry for Unit 1 And

=2.44E 6.59E 6.64E-08

=2.37E-05/ry for Unit 2 Liner_1_Freq1-10 = NCF- Class_3a_Frequency_10 - Liner_3b_Freq1-10

=2.83E 2.54E 2.58E-07

=2.55E-05/ry for Unit 1 And

=2.44E 2.20E 2.23E-07

=2.20E-05/ry for Unit 2 Liner_1_Freq1-15 = NCF- Class_3a_Frequency_15 - Liner_3b_Freq1-15

=2.83E 3.94E 4.04E-07

=2.39E-05/ry for Unit 1 And

=2.44E 3.41E 3.49E-07

=2.07E-05/ry for Unit 2 The results of other pertinent calculations are presented below:

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 99 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval For 3-in-10 years, For Unit 1 Class Person-Rem Unit 1 Baseline Frequency Unit 1 Baseline Dose Rate 1 3.86E+03 2.74E-05 1.06E-01 2 1.13E+05 1.50E-08 1.70E-03 Corrosion 3a 3.86E+04 7.63E-07 Addition 2.94E-02 3b 1.35E+05 7.69E-08 6.10E-10 1.04E-02 4 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 8a 1.13E+06 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 8c 1.39E+05 2.62E-06 3.64E-01 Total 5.02E-05 3.99E+00 ILRT Dose Rate from 3a and 3b = 3.98E-02 person-rem/yr

% of Total = 1.00%

LERF from 3b = 7.69E-08 CCFP%Liner3-10 = 43.9%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 100 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval For Unit 2 Class Person-Rem Unit 2 Baseline Frequency Unit 2 Baseline Dose Rate 1 3.86E+03 2.37E-05 9.15E-02 2 1.13E+05 1.55E-08 1.76E-03 Corrosion 3a 3.86E+04 6.59E-07 Addition 2.54E-02 3b 1.35E+05 6.64E-08 5.27E-10 8.98E-03 4 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 8b 1.88E+05 1.19E-05 2.24E+00 8c 1.39E+05 2.66E-06 3.69E-01 Total 5.18E-05 4.75E+00 ILRT Dose Rate from 3a and 3b = 3.44E-02 person-rem/yr

% of Total = 0.72%

LERF from 3b = 6.64E-08 CCFP%Liner3-10 = 53.0%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 101 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval For 1-in-10 years, For Unit 1 Class Person-Rem Unit 1 1-in-10 Frequency Unit 1 1-in-10 Dose Rate 1 3.86E+03 2.55E-05 9.83E-02 2 1.13E+05 1.15E-08 1.30E-03 Corrosion 3a 3.86E+04 2.54E-06 Addition 9.81E-02 3b 1.35E+05 2.58E-07 3.55E-09 3.48E-02 4 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 8a 1.13E+06 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 8c 1.39E+05 2.62E-06 3.64E-01 Total 5.02E-05 4.08E+00 ILRT Dose Rate from 3a and 3b = 1.33E-01 person-rem/yr

% of Total = 3.26%

LERF from 3b = 2.58E-07 CCFP%Liner1-10 = 44.2%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 102 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 Unit 2 1-in-10 Dose Rate 1 3.86E+03 2.20E-05 8.50E-02 2 1.13E+05 1.19E-08 1.35E-03 Corrosion 3a 3.86E+04 2.20E-06 Addition 8.48E-02 3b 1.35E+05 2.23E-07 3.07E-09 3.01E-02 4 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 8b 1.88E+05 1.19E-05 2.24E+00 8c 1.39E+05 2.66E-06 3.69E-01 Total 5.18E-05 4.82E+00 ILRT Dose Rate from 3a and 3b = 1.15E-01 person-rem/yr

% of Total = 2.38%

LERF from 3b = 2.23E-07 CCFP%Liner3-10 = 53.3%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 103 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval For 1-in-15-1/2 years, For Unit 1 Unit 1 1-in-15-1/2 Unit 1 1-in-15-1/2 Dose Class Person-Rem Frequency Rate 1 3.86E+03 2.39E-05 9.23E-02 2 1.13E+05 1.15E-08 1.30E-03 Corrosion 3a 3.86E+04 3.94E-06 Addition 1.52E-01 3b 1.35E+05 4.04E-07 9.62E-09 5.45E-02 4 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 8a 1.13E+06 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 8c 1.39E+05 2.62E-06 3.64E-01 Total 5.02E-05 4.15E+00 ILRT Dose Rate from 3a and 3b = 2.07E-01 person-rem/yr

% of Total = 4.98%

LERF from 3b = 4.04E-07 CCFP%Liner1-15 = 44.5%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 104 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval For Unit 2 Unit 2 1-in-15-1/2 Unit 2 1-in-15-1/2 Dose Class Person-Rem Frequency Rate 1 3.86E+03 2.07E-05 7.98E-02 2 1.13E+05 1.19E-08 1.35E-03 Corrosion 3a 3.86E+04 3.41E-06 Addition 1.31E-01 3b 1.35E+05 3.49E-07 8.31E-09 4.71E-02 4 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 8b 1.88E+05 1.19E-05 2.24E+00 8c 1.39E+05 2.66E-06 3.69E-01 Total 5.18E-05 4.88E+00 ILRT Dose Rate from 3a and 3b = 1.79E-01 person-rem/yr

% of Total = 3.66%

LERF from 3b = 3.49E-07 CCFP%Liner3-10 = 53.5%

Based on the above results, the changes from the 1-in-10 years to 1-in-15-1/2 yrs dose rate is as follows:

IncreaseLiner10-15 = ((Tot-DoseRate-Liner15 - Tot-DoseRate-Liner10)/Tot-DoseRate-Liner10)

• 100%

Where IncreaseLiner10-15 = Percent change from 1-in-10 years ILRT interval to 1-in-15-1/2 years ILRT interval Tot-DoseRate-Liner15 = Total dose rate for all EPRI Classes given a 1-in-15-1/2 years ILRT interval

= 4.15 person-rem/yr for Unit 1 And

= 4.88 person-rem/yr for Unit 2 Tot-DoseRate-Liner10) = Total dose rate for all EPRI Classes given a 1-in-10 years ILRT interval

= 4.08 person-rem/yr for Unit 1 And 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 105 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

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

= 4.82 person-rem/yr for Unit 2 IncreaseLiner10-15 = ((Tot-DoseRate-Liner15 - Tot-DoseRate-Liner10)/Tot-DoseRate-Liner10)

• 100%

= ((4.15 - 4.08)/4.08

• 100% = 1.7% for Unit 1 And

= ((4.88 - 4.82)/4.82 *100% = 1.2% 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 1.7% for Unit1 and 1.2% 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-10 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:

LERFLiner10-15 = Liner_3b_Freq1 Liner_3b_Freq1-10

Where, LERFLiner10-15 = 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 = frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-10 years ILRT interval (Section B.11.7)

= 2.58E-07/ry for Unit 1 And

= 2.23E-07/ry for Unit 2 Liner_3b_Freq1-15 = frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-15-1/2 years ILRT interval (Section B.11.7)

= 4.04E-07/ry for Unit 1 And

= 3.49E-07/ry for Unit 2 Therefore, LERFLiner10-15 = Liner_3b_Freq1 Liner_3b_Freq1-10

= 4.04E-07/ry - 2.58E-07/ry

= 1.46E-07/ry for Unit 1 And

= 3.49E-07/ry - 2.23E-07/ry

= 1.26E-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 1.46E-07/ry in Unit1 and 1.26E-07 in Unit 2. These values are between the NRC Regulatory Guide 1.174 (Reference B12) values of 1E-7 and 1E-6/ry and are defined as a small increase in risk. Therefore, 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 106 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval increasing the ILRT interval at Point Beach from the currently allowed 1-in-10 years to 1-in-15-1/2 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:

LERFLiner3-15 = Liner_3b_Freq1 Liner_3b_Freq3-10

Where, LERFLiner3-15 = the change in LERF from the 3-in-10 years ILRT interval to the 1-in-15-1/2 years interval Liner_3b_Freq3-10 = frequency of EPRI Class 3b due to liner corrosion failure given a 3-in-10 years ILRT interval (Section B.11.7)

= 7.69E-08/ry for Unit 1 And

= 6.64E-08/ry for Unit 2 Liner_3b_Freq1-15 = frequency of EPRI Class 3b due to liner corrosion failure given a 1-in-15-1/2 years ILRT interval (Section B.11.7)

= 4.04E-07/ry for Unit 1 And

= 3.49E-07/ry for Unit 2 Therefore, LERFLiner3-15 = Liner_3b_Freq1 Liner_3b_Freq3-10

= 4.04E-07/ry - 7.69E-08/ry

= 3.27E-07/ry for Unit 1 And

= 3.49E-07/ry - 6.64E-08/ry

= 2.82E-07/ry for Unit 2 Similar to the LERFLiner10-15 result, the LERFLiner3-15 also represents only a small increase in plant risk.

B.11.9 Step 9 - Evaluate the Change 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:

CCFPLiner10-15 = CCFPLiner1 CCFPLiner1-10

Where, 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 107 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval CCFPLiner10-15 = the change in conditional containment failure probability from 1-in-10 years to 1-in 1/2 years given not-detected containment leakage.

CCFPLiner1-15 = conditional containment failure probability given 1-in-15-1/2 years ILRT interval and potential non-detected containment leakage

= 44.5% for Unit 1 (Step 7)

And

= 53.5% for Unit 2 CCFPLiner1-10 = conditional containment failure probability given 1-in-10 years ILRT interval and potential non-detected containment leakage

= 44.2% for Unit 1 (Step 7)

And

= 53.3% for Unit 2 Therefore, CCFPLiner10-15 = CCFPLiner1 CCFPLiner1-10

= 44.5% - 44.2% = 0.3% for Unit 1 And

= 53.5% - 53.3% = 0.2% 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 Sensitivity 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

• 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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 108 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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) (with corrosion)

ILRT Dose Rate from 3a and 3b 3.98E-02 1.33E-01 2.07E-01 Increase over original 8.24E-05 4.80E-04 1.30E-03

% of Total Dose 1.00% 3.26% 4.98%

Increase over original 0.00% 0.01% 0.03%

Delta Dose Rate from 3a and 3b (10 to 15-1/2 yr) 7.37E-02 Increase over original 8.19E-04 LERF from 3b 7.69E-08 2.58E-07 4.04E-07 Increase over original 6.10E-10 3.55E-09 9.62E-09 Delta LERF (10 to 15-1/2 yr) 1.46E-07 Increase over original 6.07E-09 CCFP% 43.9% 44.2% 44.5%

Increase over original 0.001% 0.003% 0.02%

Delta CCFP% (10 to 15-1/2 yr) 0.3%

Increase over original 0.012%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 109 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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) (with corrosion)

ILRT Dose Rate from 3a and 3b 3.44E-02 1.15E-01 1.79E-01 Increase over original 7.13E-05 4.15E-04 1.12E-03

% of Total Dose 0.72% 2.38% 3.66%

Increase over original 0.00% 0.01% 0.02%

Delta Dose Rate from 3a and 3b (10 to 15-1/2 yr) 6.37E-02 Increase over original 7.08E-04 LERF from 3b 6.64E-08 2.23E-07 3.49E-07 Increase over original 5.27E-10 3.07E-09 8.31E-09 Delta LERF (10 to 15-1/2 yr) 1.26E-07 Increase over original 5.24E-09 CCFP% 53.0% 53.3% 53.5%

Increase over original 0.0010% 0.003% 0.013%

Delta CCFP% (10 to 15-1/2 yr) 0.2%

Increase over original 0.010%

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 non-risk 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 6.07E-09 for Unit 1 (5.24E-09/ry for Unit 2). This is below the Regulatory Guide 1.174 acceptance criteria threshold of 1E-07/yr.

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 110 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 8.19E-04 person-rem/ry for Unit 1 (7.08E-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 0.012% for Unit 1 and 0.01% 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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 111 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table B1 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 1.00E+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 8.11E-03 9.92E-01 1.35E-03 9.99E-01 9 9.32E-03 9.91E-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 1.41E-02 9.86E-01 2.36E-03 9.98E-01 13 1.62E-02 9.84E-01 2.71E-03 9.97E-01 14 1.87E-02 9.81E-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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 112 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 113 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 114 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 F a i l u r e P r o b a b i l i ty 0.6000 Containment Liner 0.4000 Basemat 0.2000 0.0000 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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 115 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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-15.5 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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 116 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 2.74E-05 3.86E+03 1.06E-01 2.55E-05 3.86E+03 9.83E-02 2.39E-05 3.86E+03 9.23E-02 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 7.63E-07 3.86E+04 2.94E-02 2.54E-06 3.86E+04 9.81E-02 3.94E-06 3.86E+04 1.52E-01 3b 7.69E-08 1.35E+05 1.04E-02 2.58E-07 1.35E+05 3.48E-02 4.04E-07 1.35E+05 5.45E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 7.56E-06 1.39E+05 1.05E+00 7.56E-06 1.39E+05 1.05E+00 7.56E-06 1.39E+05 1.05E+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 1.15E-05 1.88E+05 2.16E+00 1.15E-05 1.88E+05 2.16E+00 1.15E-05 1.88E+05 2.16E+00 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 0.00E+00 3.99E+00 5.02E-05 0.00E+00 4.08E+00 5.02E-05 0.00E+00 4.15E+00 ILRT Dose Rate from 3a and 3b 3.98E-02 1.33E-01 2.07E-01 Increase over original 8.24E-05 4.80E-04 1.30E-03

% of Total 1.00% 3.26% 4.98%

Increase over original 0.00% 0.01% 0.03%

Delta Dose Rate from 3a and 3b (10 to 15 yr) 7.37E-02 Increase over original 8.19E-04 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 117 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval LERF from 3b 7.69E-08 2.58E-07 4.04E-07 Increase over original 6.10E-10 3.55E-09 9.62E-09 Delta LERF (10 to 15 yr) 1.46E-07 Increase over original 6.07E-09 CCFP% 43.9% 44.2% 44.5%

Increase over original 0.001% 0.003% 0.02%

Delta CCFP% (10 to 15 yr) 0.3%

Increase over original 0.012%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 118 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 2.37E-05 3.86E+03 9.15E-02 2.20E-05 3.86E+03 8.50E-02 2.07E-05 3.86E+03 7.98E-02 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 6.59E-07 3.86E+04 2.54E-02 2.20E-06 3.86E+04 8.48E-02 3.41E-06 3.86E+04 1.31E-01 3b 6.64E-08 1.35E+05 8.98E-03 2.23E-07 1.35E+05 3.01E-02 3.49E-07 1.35E+05 4.71E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.26E-05 1.39E+05 1.74E+00 1.26E-05 1.39E+05 1.74E+00 1.26E-05 1.39E+05 1.74E+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 1.19E-05 1.88E+05 2.24E+00 1.19E-05 1.88E+05 2.24E+00 1.19E-05 1.88E+05 2.24E+00 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 0.00E+00 4.75E+00 5.18E-05 0.00E+00 4.82E+00 5.18E-05 0.00E+00 4.88E+00 ILRT Dose Rate from 3a and 3b 3.44E-02 1.15E-01 1.79E-01 Increase over original 7.13E-05 4.15E-04 1.12E-03

% of Total 0.72% 2.38% 3.66%

Increase over original 0.00% 0.01% 0.02%

Delta Dose Rate from 3a and 3b (10 to 15 yr) 6.37E-02 Increase over original 7.08E-04 LERF from 3b 6.64E-08 2.23E-07 3.49E-07 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 119 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Increase over original 5.27E-10 3.07E-09 8.31E-09 Delta LERF (10 to 15 yr) 1.26E-07 Increase over original 5.24E-09 CCFP% 53.0% 53.3% 53.5%

Increase over original 0.0010% 0.003% 0.013%

Delta CCFP% (10 to 15 yr) 0.2%

Increase over original 0.010%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 120 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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% 6.10E-10 3.55E-09 9.62E-09 1.46E-07 5 years 0.174% Basemat Sensitivities Doubles Base Base Base 1.75E-10 2.96E-09 2.32E-08 1.60E-07 every 2 years Doubles Base Base Base 9.09E-10 3.94E-09 7.98E-09 1.44E-07 every 10 years Base Base 5% Base 3.49E-10 2.03E-09 5.50E-09 1.43E-07 Base Base 15% Base 8.72E-10 5.08E-09 1.37E-08 1.48E-07 Base 0.451% Liner, Base Base 1.58E-10 9.21E-10 2.49E-09 1.41E-07 0.0451% Basemat Base 6.72% Liner, Base Base 2.36E-09 1.37E-08 3.71E-08 1.63E-07 0.672% Basemat Lower Bound Doubles 0.451% Liner, 5% 10% 1.35E-11 5.83E-11 1.18E-10 1.40E-07 every 10 0.0451% Basemat years Upper Bound Doubles 6.72% Liner, 15% 100% 9.65E-10 1.63E-08 1.28E-07 2.51E-07 every 2 years 0.672% Basemat 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 121 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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% 5.27E-10 3.07E-09 8.31E-09 1.26E-07 years 0.174% Basemat Sensitivities Doubles every 2 Base Base Base 1.51E-10 2.56E-09 2.01E-08 1.38E-07 years Doubles every 10 Base Base Base 7.86E-10 3.40E-09 6.89E-09 1.24E-07 years Base Base 5% Base 3.01E-10 1.76E-09 4.75E-09 1.24E-07 Base Base 15% Base 7.53E-10 4.39E-09 1.19E-08 1.28E-07 Base 0.451% Liner, Base Base 1.37E-10 7.96E-10 2.16E-09 1.22E-07 0.0451% Basemat Base 6.72% Liner, Base Base 2.04E-09 1.19E-08 3.21E-08 1.41E-07 0.672% Basemat Lower Bound Doubles every 10 0.451% Liner, 5% 10% 1.16E-11 5.04E-11 1.02E-10 1.21E-07 years 0.0451% Basemat Upper Bound Doubles every 2 6.72% Liner, 15% 100% 8.34E-10 1.41E-08 1.11E-07 2.17E-07 years 0.672% Basemat 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 122 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 123 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table of Contents A1.0 Introduction 118 A2.0 Sensitivity Case 1 - Flaw Rate Doubles Every 2 Years 119 A3.0 Sensitivity Case 2 - Flaw Rate Doubles Every 10 Years 124 A4.0 Sensitivity Case 3 - 5%Visual Inspection Failures 129 A5.0 Sensitivity Case 4 - 15%Visual Inspection Failures 134 A6.0 Sensitivity Case 5 - Containment Breach Base Point 10 Times Lower 139 A7.0 Sensitivity Case 6 - Containment Breach Base Point 10 Times Higher 144 A8.0 Sensitivity Case 7 - Lower Bound 149 A9.0 Sensitivity Case 8 - Upper Bound 154 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 124 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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

• 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

• 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 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 125 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A2.0 Sensitivity Case 1 - Flaw Rate Doubles Every 2 Years Containment From Estimated Change Cylinder/Dome Basemat 1 to 3 years 0.31% 0.05%

1 to 10 years 5.17% 0.86%

1 to 15-1/2 years 40.50% 6.75%

Other Assumptions Containment Breach 1.740% 0.174%

Visual Inspection Failures 10.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 126 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 0.00062% 0.00062%

Unit 1 Unit 2 Person- Unit 1 Baseline Unit 2 Baseline Rem (50- Baseline Dose Baseline Dose Class miles) Frequency Rate Frequency Rate 1 3.86E+03 2.74E-05 1.06E-01 2.37E-05 9.15E-02 2 1.13E+05 1.50E-08 1.70E-03 1.55E-08 1.76E-03 Corrosion Corrosion 3a 3.86E+04 7.63E-07 Addition 2.94E-02 6.59E-07 Addition 2.54E-02 3b 1.35E+05 7.64E-08 1.75E-10 1.03E-02 6.61E-08 1.51E-10 8.93E-03 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 3.99E+00 5.18E-05 4.75E+00 ILRT Dose Rate from 3a and 3b = 3.98E-02 3.44E-02

% of Total = 1.00% 0.72%

LERF from 3b 7.64E-08 6.61E-08 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 127 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

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

=

CCFP%Liner= 43.85% 52.98%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 128 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-10 years Unit 1 Unit 2 Increase to 3a Frequency 0 0 Increase to 3b Frequency 0.0105% 0.0105%

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) 1 3.86E+03 2.55E-05 9.83E-02 2.20E-05 8.50E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 2.54E-06 Addition 9.81E-02 2.20E-06 Addition 8.48E-02 3b 1.35E+05 2.57E-07 2.96E-09 3.47E-02 2.22E-07 2.56E-09 3.00E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.08E+00 5.18E-05 4.82E+00 ILRT Dose Rate from 3a and 3b = 1.33E-01 1.15E-01

% of Total = 3.26% 2.38%

LERF from 3b = 2.57E-07 2.22E-07 CCFP%Liner= 44.21% 53.27%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 129 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 0.0822% 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) ILRT) ILRT) 1 3.86E+03 2.39E-05 9.23E-02 2.07E-05 7.98E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 3.94E-06 Addition 1.52E-01 3.41E-06 Addition 1.31E-01 3b 1.35E+05 4.17E-07 2.32E-08 5.64E-02 3.61E-07 2.01E-08 4.87E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.15E+00 5.18E-05 4.88E+00 ILRT Dose Rate from 3a and 3b = 2.08E-01 1.80E-01

% of Total = 5.03% 3.69%

LERF from 3b = 4.17E-07 3.61E-07 CCFP%Liner= 44.53% 53.54%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 130 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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) 7.56E-02 6.53E-02 3 to 15-1/2 Increase (Person-rem/ry) 1.69E-01 1.46E-01 10 to 15-1/2 Delta-LERF 1.60E-07 1.38E-07 3 to 15-1/2 Delta-LERF 3.41E-07 2.95E-07 10 to 15-1/2 Delta-CCFP 0.3% 0.3%

3 to 15-1/2 Delta-CCFP 0.7% 0.6%

3 to 10 Delta-LERF from Corrosion 2.79E-09 2.41E-09 10 to 15-1/2 Delta-LERF from Corrosion 2.03E-08 1.75E-08 Increase in LERF (ILRT 3-to-15-1/2 years) from Corrosion 2.30E-08 1.99E-08 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 131 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A3.0 Sensitivity Case 2 - Flaw Rate Doubles Every 10 Years Containment From Estimated Change Cylinder/Dome Basemat 1 to 3 years 1.58% 0.27%

1 to 10 years 6.86% 1.15%

1 to 15-1/2 years 13.91% 2.33%

Other Assumptions Containment Breach 1.740% 0.174%

Visual Inspection Failures 10.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 132 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 0.00322% 0.00322%

Person- Unit 1 Unit 1 Unit 2 Unit 2 Rem (50- Baseline Baseline Baseline Baseline Class miles) Frequency Dose Rate Frequency Dose Rate 1 3.86E+03 2.74E-05 1.06E-01 2.37E-05 9.15E-02 2 1.13E+05 1.50E-08 1.70E-03 1.55E-08 1.76E-03 Corrosion Corrosion 3a 3.86E+04 7.63E-07 Addition 2.94E-02 6.59E-07 Addition 2.54E-02 3b 1.35E+05 7.72E-08 9.09E-10 1.04E-02 6.67E-08 7.86E-10 9.01E-03 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 3.99E+00 5.18E-05 4.75E+00 ILRT Dose Rate from 3a and 3b

= 3.99E-02 3.45E-02

% of Total = 1.00% 0.73%

LERF from 3b = 7.72E-08 6.67E-08 CCFP%Liner= 43.85% 52.98%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 133 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-10 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-10 10 yrs Class miles) yrs ILRT) yrs ILRT) yrs ILRT) ILRT) 1 3.86E+03 2.55E-05 9.83E-02 2.20E-05 8.50E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 2.54E-06 Addition 9.81E-02 2.20E-06 Addition 8.48E-02 3b 1.35E+05 2.58E-07 3.94E-09 3.49E-02 2.23E-07 3.40E-09 3.01E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.08E+00 5.18E-05 4.82E+00 ILRT Dose Rate from 3a and 3b = 1.33E-01 1.15E-01

% of Total = 3.26% 2.38%

LERF from 3b = 2.58E-07 2.23E-07 CCFP%Liner= 44.21% 53.27%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 134 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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-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) ILRT) ILRT) 1 3.86E+03 2.39E-05 9.23E-02 2.07E-05 7.98E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 3.94E-06 Addition 1.52E-01 3.41E-06 Addition 1.31E-01 3b 1.35E+05 4.02E-07 7.98E-09 5.43E-02 3.47E-07 6.89E-09 4.69E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.15E+00 5.18E-05 4.88E+00 ILRT Dose Rate from 3a and 3b = 2.06E-01 1.78E-01

% of Total = 4.98% 3.65%

LERF from 3b = 4.02E-07 3.47E-07 CCFP%Liner= 44.50% 53.51%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 135 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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) 7.34E-02 6.34E-02 3 to 15-1/2 Increase (Person-rem/ry) 1.67E-01 1.44E-01 10 to 15-1/2 Delta-LERF 1.44E-07 1.24E-07 3 to 15-1/2 Delta-LERF 3.25E-07 2.81E-07 10 to 15-1/2 Delta-CCFP 0.3% 0.2%

3 to 15-1/2 Delta-CCFP 0.6% 0.5%

3 to 10 Delta-LERF from Corrosion 3.03E-09 2.62E-09 10 to 15-1/2 Delta-LERF from Corrosion 4.04E-09 3.49E-09 Increase in LERF (ILRT 3-to-15-1/2 years) from Corrosion 7.07E-09 6.11E-09 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 136 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A4.0 Sensitivity Case 3 - 5%Visual Inspection Failures Containment From Estimated Change Cylinder/Dome Basemat 1 to 3 years 1.06% 0.18%

1 to 10 years 6.20% 1.04%

1 to 15-1/2 years 16.78% 2.80%

Other Assumptions Containment Breach 1.740% 0.174%

Visual Inspection Failures 5.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 137 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 0.00123% 0.00123%

Person- Unit 1 Unit 1 Unit 2 Unit 2 Rem (50- Baseline Baseline Baseline Baseline Class miles) Frequency Dose Rate Frequency Dose Rate 1 3.86E+03 2.74E-05 1.06E-01 2.37E-05 9.15E-02 2 1.13E+05 1.50E-08 1.70E-03 1.55E-08 1.76E-03 Corrosion Corrosion 3a 3.86E+04 7.63E-07 Addition 2.94E-02 6.59E-07 Addition 2.54E-02 3b 1.35E+05 7.66E-08 3.49E-10 1.04E-02 6.62E-08 3.01E-10 8.95E-03 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 3.99E+00 5.18E-05 4.75E+00 ILRT Dose Rate from 3a and 3b

= 3.98E-02 3.44E-02

% of Total = 1.00% 0.72%

LERF from 3b = 7.66E-08 6.62E-08 CCFP%Liner= 43.85% 52.98%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 138 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-10 years Unit 1 Unit 2 Increase to 3a Frequency 0 0 Increase to 3b Frequency 0.0072% 0.0072%

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) 1 3.86E+03 2.55E-05 9.83E-02 2.20E-05 8.50E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 2.54E-06 Addition 9.81E-02 2.20E-06 Addition 8.48E-02 3b 1.35E+05 2.56E-07 2.03E-09 3.46E-02 2.21E-07 1.76E-09 2.99E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.08E+00 5.18E-05 4.82E+00 ILRT Dose Rate from 3a and 3b = 1.33E-01 1.15E-01

% of Total = 3.26% 2.38%

LERF from 3b = 2.56E-07 2.21E-07 CCFP%Liner= 44.21% 53.27%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 139 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 0.0195% 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) ILRT) ILRT) 1 3.86E+03 2.39E-05 9.23E-02 2.07E-05 7.98E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 3.94E-06 Addition 1.52E-01 3.41E-06 Addition 1.31E-01 3b 1.35E+05 4.00E-07 5.50E-09 5.40E-02 3.45E-07 4.75E-09 4.67E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.14E+00 5.18E-05 4.88E+00 ILRT Dose Rate from 3a and 3b = 2.06E-01 1.78E-01

% of Total = 4.97% 3.65%

LERF from 3b = 4.00E-07 3.45E-07 CCFP%Liner= 44.49% 53.51%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 140 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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) 7.33E-02 6.34E-02 3 to 15-1/2 Increase (Person-rem/ry) 1.66E-01 1.44E-01 10 to 15-1/2 Delta-LERF 1.43E-07 1.24E-07 3 to 15-1/2 Delta-LERF 3.23E-07 2.79E-07 10 to 15-1/2 Delta-CCFP 0.3% 0.2%

3 to 15-1/2 Delta-CCFP 0.6% 0.5%

3 to 10 Delta-LERF from Corrosion 1.68E-09 1.45E-09 10 to 15-1/2 Delta-LERF from Corrosion 3.47E-09 3.00E-09 Increase in LERF (ILRT 3-to-15-1/2 years) from Corrosion 5.15E-09 4.45E-09 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 141 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval A5.0 Sensitivity Case 4 - 15%Visual Inspection Failures Containment From Estimated Change Cylinder/Dome Basemat 1 to 3 years 1.06% 0.18%

1 to 10 years 6.20% 1.04%

1 to 15-1/2 years 16.78% 2.80%

Other Assumptions Containment Breach 1.740% 0.174%

Visual Inspection Failures 15.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 142 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 1 Unit 1 Unit 2 Unit 2 Rem (50- Baseline Baseline Baseline Baseline Class miles) Frequency Dose Rate Frequency Dose Rate 1 3.86E+03 2.74E-05 1.06E-01 2.37E-05 9.15E-02 2 1.13E+05 1.50E-08 1.70E-03 1.55E-08 1.76E-03 Corrosion Corrosion 3a 3.86E+04 7.63E-07 Addition 2.94E-02 6.59E-07 Addition 2.54E-02 3b 1.35E+05 7.71E-08 8.72E-10 1.04E-02 6.67E-08 7.53E-10 9.01E-03 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 3.99E+00 5.18E-05 4.75E+00 ILRT Dose Rate from 3a and 3b

= 3.99E-02 3.45E-02

% of Total = 1.00% 0.73%

LERF from 3b = 7.71E-08 6.67E-08 CCFP%Liner= 43.85% 52.98%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 143 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-10 years Unit 1 Unit 2 Increase to 3a Frequency 0 0 Increase to 3b Frequency 0.0180% 0.0180%

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) 1 3.86E+03 2.55E-05 9.83E-02 2.20E-05 8.50E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 2.54E-06 Addition 9.81E-02 2.20E-06 Addition 8.48E-02 3b 1.35E+05 2.59E-07 5.08E-09 3.50E-02 2.24E-07 4.39E-09 3.03E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.08E+00 5.18E-05 4.82E+00 ILRT Dose Rate from 3a and 3b = 1.33E-01 1.15E-01

% of Total = 3.27% 2.39%

LERF from 3b = 2.59E-07 2.24E-07 CCFP%Liner= 44.21% 53.28%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 144 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 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) ILRT) ILRT) 1 3.86E+03 2.39E-05 9.23E-02 2.07E-05 7.98E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 3.94E-06 Addition 1.52E-01 3.41E-06 Addition 1.31E-01 3b 1.35E+05 4.08E-07 1.37E-08 5.51E-02 3.52E-07 1.19E-08 4.76E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.15E+00 5.18E-05 4.88E+00 ILRT Dose Rate from 3a and 3b = 2.07E-01 1.79E-01

% of Total = 5.00% 3.67%

LERF from 3b = 4.08E-07 3.52E-07 CCFP%Liner= 44.51% 53.52%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 145 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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) 7.40E-02 6.40E-02 3 to 15-1/2 Increase (Person-rem/ry) 1.67E-01 1.45E-01 10 to 15-1/2 Delta-LERF 1.48E-07 1.28E-07 3 to 15-1/2 Delta-LERF 3.31E-07 2.86E-07 10 to 15-1/2 Delta-CCFP 0.3% 0.2%

3 to 15-1/2 Delta-CCFP 0.7% 0.5%

3 to 10 Delta-LERF from Corrosion 4.21E-09 3.63E-09 10 to 15-1/2 Delta-LERF from Corrosion 8.66E-09 7.49E-09 Increase in LERF (ILRT 3-to-15-1/2 years) from Corrosion 1.29E-08 1.11E-08 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 146 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 Containment From Estimated Change Cylinder/Dome Basemat 1 to 3 years 1.06% 0.18%

1 to 10 years 6.20% 1.04%

1 to 15-1/2 years 16.78% 2.80%

Other Assumptions Containment Breach 0.451% 0.045%

Visual Inspection Failures 10.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 147 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 0.00056% 0.00056%

Person- Unit 1 Unit 1 Unit 2 Unit 2 Rem (50- Baseline Baseline Baseline Baseline Class miles) Frequency Dose Rate Frequency Dose Rate 1 3.86E+03 2.74E-05 1.06E-01 2.37E-05 9.15E-02 2 1.13E+05 1.50E-08 1.70E-03 1.55E-08 1.76E-03 Corrosion Corrosion 3a 3.86E+04 7.63E-07 Addition 2.94E-02 6.59E-07 Addition 2.54E-02 3b 1.35E+05 7.64E-08 1.58E-10 1.03E-02 6.61E-08 1.37E-10 8.92E-03 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 3.99E+00 5.18E-05 4.75E+00 ILRT Dose Rate from 3a and 3b

= 3.98E-02 3.44E-02

% of Total = 1.00% 0.72%

LERF from 3b = 7.64E-08 6.61E-08 CCFP%Liner= 43.85% 52.98%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 148 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-10 years Unit 1 Unit 2 Increase to 3a Frequency 0 0 Increase to 3b Frequency 0.0033% 0.0033%

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) 1 3.86E+03 2.55E-05 9.83E-02 2.20E-05 8.50E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 2.54E-06 Addition 9.81E-02 2.20E-06 Addition 8.48E-02 3b 1.35E+05 2.55E-07 9.21E-10 3.45E-02 2.21E-07 7.96E-10 2.98E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.08E+00 5.18E-05 4.82E+00 ILRT Dose Rate from 3a and 3b = 1.33E-01 1.15E-01

% of Total = 3.25% 2.38%

LERF from 3b = 2.55E-07 2.21E-07 CCFP%Liner= 44.20% 53.27%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 149 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 0.0088% 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) ILRT) ILRT) 1 3.86E+03 2.39E-05 9.23E-02 2.07E-05 7.98E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 3.94E-06 Addition 1.52E-01 3.41E-06 Addition 1.31E-01 3b 1.35E+05 3.97E-07 2.49E-09 5.36E-02 3.43E-07 2.16E-09 4.63E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.14E+00 5.18E-05 4.88E+00 ILRT Dose Rate from 3a and 3b = 2.06E-01 1.78E-01

% of Total = 4.96% 3.64%

LERF from 3b = 3.97E-07 3.43E-07 CCFP%Liner= 44.49% 53.51%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 150 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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) 7.31E-02 6.32E-02 3 to 15-1/2 Increase (Person-rem/ry) 1.66E-01 1.43E-01 10 to 15-1/2 Delta-LERF 1.41E-07 1.22E-07 3 to 15-1/2 Delta-LERF 3.20E-07 2.77E-07 10 to 15-1/2 Delta-CCFP 0.3% 0.2%

3 to 15-1/2 Delta-CCFP 0.6% 0.5%

3 to 10 Delta-LERF from Corrosion 7.63E-10 6.60E-10 10 to 15-1/2 Delta-LERF from Corrosion 1.57E-09 1.36E-09 Increase in LERF (ILRT 3-to-15-1/2 years) from Corrosion 2.34E-09 2.02E-09 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 151 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 Containment From Estimated Change Cylinder/Dome Basemat 1 to 3 years 1.06% 0.18%

1 to 10 years 6.20% 1.04%

1 to 15-1/2 years 16.78% 2.80%

Other Assumptions Containment Breach 6.715% 0.671%

Visual Inspection Failures 10.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 152 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 1 Unit 1 Unit 2 Unit 2 Rem (50- Baseline Baseline Baseline Baseline Class miles) Frequency Dose Rate Frequency Dose Rate 1 3.86E+03 2.74E-05 1.06E-01 2.37E-05 9.15E-02 2 1.13E+05 1.50E-08 1.70E-03 1.55E-08 1.76E-03 Corrosion Corrosion 3a 3.86E+04 7.63E-07 Addition 2.94E-02 6.59E-07 Addition 2.54E-02 3b 1.35E+05 7.86E-08 2.36E-09 1.06E-02 6.80E-08 2.04E-09 9.18E-03 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 3.99E+00 5.18E-05 4.75E+00 ILRT Dose Rate from 3a and 3b

= 4.01E-02 3.46E-02

% of Total = 1.00% 0.73%

LERF from 3b = 7.86E-08 6.80E-08 CCFP%Liner= 43.86% 52.98%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 153 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-10 years Unit 1 Unit 2 Increase to 3a Frequency 0 0 Increase to 3b Frequency 0.0486% 0.0486%

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) 1 3.86E+03 2.55E-05 9.82E-02 2.20E-05 8.49E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 2.54E-06 Addition 9.81E-02 2.20E-06 Addition 8.48E-02 3b 1.35E+05 2.68E-07 1.37E-08 3.62E-02 2.32E-07 1.19E-08 3.13E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.08E+00 5.18E-05 4.82E+00 ILRT Dose Rate from 3a and 3b = 1.34E-01 1.16E-01

% of Total = 3.29% 2.41%

LERF from 3b = 2.68E-07 2.32E-07 CCFP%Liner= 44.23% 53.29%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 154 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 0.1315% 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) ILRT) ILRT) 1 3.86E+03 2.39E-05 9.22E-02 2.07E-05 7.97E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 3.94E-06 Addition 1.52E-01 3.41E-06 Addition 1.31E-01 3b 1.35E+05 4.31E-07 3.71E-08 5.82E-02 3.73E-07 3.21E-08 5.03E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.15E+00 5.18E-05 4.89E+00 ILRT Dose Rate from 3a and 3b = 2.10E-01 1.82E-01

% of Total = 5.07% 3.72%

LERF from 3b = 4.31E-07 3.73E-07 CCFP%Liner= 44.55% 53.56%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 155 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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) 7.60E-02 6.57E-02 3 to 15-1/2 Increase (Person-rem/ry) 1.70E-01 1.47E-01 10 to 15-1/2 Delta-LERF 1.63E-07 1.41E-07 3 to 15-1/2 Delta-LERF 3.53E-07 3.05E-07 10 to 15-1/2 Delta-CCFP 0.3% 0.3%

3 to 15-1/2 Delta-CCFP 0.7% 0.6%

3 to 10 Delta-LERF from Corrosion 1.14E-08 9.82E-09 10 to 15-1/2 Delta-LERF from Corrosion 2.34E-08 2.02E-08 Increase in LERF (ILRT 3-to-15-1/2 years) from Corrosion 3.48E-08 3.01E-08 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 156 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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)

Containment From Estimated Change Cylinder/Dome Basemat 1 to 3 years 1.58% 0.27%

1 to 10 years 6.86% 1.15%

1 to 15-1/2 years 13.91% 2.33%

Other Assumptions Containment Breach 0.451% 0.045%

Visual Inspection Failures 5.0% 100.0%

EPRI Class 3a Fraction 90.0% 90.0%

EPRI Class 3b Fraction 10.0% 10.0%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 157 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 0.00043% 0.00043%

Increase to 3b Frequency 0.00005% 0.00005%

Person- Unit 1 Unit 1 Unit 2 Unit 2 Rem (50- Baseline Baseline Baseline Baseline Class miles) Frequency Dose Rate Frequency Dose Rate 1 3.86E+03 2.74E-05 1.06E-01 2.37E-05 9.15E-02 2 1.13E+05 1.50E-08 1.70E-03 1.55E-08 1.76E-03 Corrosion Corrosion 3a 3.86E+04 7.63E-07 Addition 2.94E-02 6.59E-07 Addition 2.54E-02 3b 1.35E+05 7.63E-08 1.35E-11 1.03E-02 6.59E-08 1.16E-11 8.91E-03 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 3.99E+00 5.18E-05 4.75E+00 ILRT Dose Rate from 3a and 3b

= 3.97E-02 3.44E-02

% of Total = 1.00% 0.72%

LERF from 3b = 7.63E-08 6.59E-08 CCFP%Liner= 43.85% 52.97%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 158 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-10 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) 1 3.86E+03 2.55E-05 9.83E-02 2.20E-05 8.50E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 2.54E-06 Addition 9.81E-02 2.20E-06 Addition 8.48E-02 3b 1.35E+05 2.54E-07 5.83E-11 3.44E-02 2.20E-07 5.04E-11 2.97E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.08E+00 5.18E-05 4.82E+00 ILRT Dose Rate from 3a and 3b = 1.32E-01 1.15E-01

% of Total = 3.25% 2.37%

LERF from 3b = 2.54E-07 2.20E-07 CCFP%Liner= 44.20% 53.27%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 159 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 0.00377% 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) ILRT) ILRT) 1 3.86E+03 2.39E-05 9.23E-02 2.07E-05 7.98E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 3.94E-06 Addition 1.52E-01 3.41E-06 Addition 1.32E-01 3b 1.35E+05 3.94E-07 1.18E-10 5.32E-02 3.41E-07 1.02E-10 4.60E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.14E+00 5.18E-05 4.88E+00 ILRT Dose Rate from 3a and 3b

= 2.05E-01 1.78E-01

% of Total = 4.96% 3.64%

LERF from 3b = 3.94E-07 3.41E-07 CCFP%Liner= 44.48% 53.50%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 160 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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) 7.29E-02 6.30E-02 3 to 15-1/2 Increase (Person-rem/ry) 1.66E-01 1.43E-01 10 to 15-1/2 Delta-LERF 1.40E-07 1.21E-07 3 to 15-1/2 Delta-LERF 3.18E-07 2.75E-07 10 to 15-1/2 Delta-CCFP 0.3% 0.2%

3 to 15-1/2 Delta-CCFP 0.6% 0.5%

3 to 10 Delta-LERF from Corrosion 4.49E-11 3.88E-11 10 to 15-1/2 Delta-LERF from Corrosion 5.98E-11 5.17E-11 Increase in LERF (ILRT 3-to-15-1/2 years) from Corrosion 1.05E-10 9.05E-11 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 161 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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)

Containment From Estimated Change Cylinder/Dome Basemat 1 to 3 years 0.31% 0.05%

1 to 10 years 5.17% 0.86%

1 to 15-1/2 years 40.50% 6.75%

Other Assumptions Containment Breach 6.715% 0.671%

Visual Inspection Failures 15.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 162 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 0.00342% 0.00342%

Person- Unit 1 Unit 1 Unit 2 Unit 2 Rem (50- Baseline Baseline Baseline Baseline Class miles) Frequency Dose Rate Frequency Dose Rate 1 3.86E+03 2.74E-05 1.06E-01 2.37E-05 9.15E-02 2 1.13E+05 1.50E-08 1.70E-03 1.55E-08 1.76E-03 Corrosion Corrosion 3a 3.86E+04 7.63E-07 Addition 2.94E-02 6.59E-07 Addition 2.54E-02 3b 1.35E+05 7.72E-08 9.65E-10 1.04E-02 6.68E-08 8.34E-10 9.02E-03 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 3.99E+00 5.18E-05 4.75E+00 ILRT Dose Rate from 3a and 3b

= 3.99E-02 3.45E-02

% of Total = 1.00% 0.73%

LERF from 3b = 7.72E-08 6.68E-08 CCFP%Liner= 43.85% 52.98%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 163 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 1-in-10 years Unit 1 Unit 2 Increase to 3a Frequency 0 0 Increase to 3b Frequency 0.0579% 0.0579%

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) 1 3.86E+03 2.54E-05 9.82E-02 2.20E-05 8.49E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 2.54E-06 Addition 9.81E-02 2.20E-06 Addition 8.48E-02 3b 1.35E+05 2.71E-07 1.63E-08 3.66E-02 2.34E-07 1.41E-08 3.16E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.08E+00 5.18E-05 4.83E+00 ILRT Dose Rate from 3a and 3b = 1.35E-01 1.16E-01

% of Total = 3.30% 2.41%

LERF from 3b = 2.71E-07 2.34E-07 CCFP%Liner= 44.23% 53.30%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 164 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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 0.4532% 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) ILRT) ILRT) 1 3.86E+03 2.38E-05 9.19E-02 2.06E-05 7.94E-02 2 1.13E+05 1.15E-08 1.30E-03 1.19E-08 1.35E-03 Corrosion Corrosion 3a 3.86E+04 3.94E-06 Addition 1.52E-01 3.41E-06 Addition 1.31E-01 3b 1.35E+05 5.22E-07 1.28E-07 7.05E-02 4.51E-07 1.11E-07 6.10E-02 4 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 6 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 1.39E+05 7.56E-06 1.05E+00 1.26E-05 1.74E+00 8a 1.13E+06 2.37E-07 2.68E-01 2.37E-07 2.68E-01 8b 1.88E+05 1.15E-05 2.16E+00 1.19E-05 2.24E+00 8c 1.39E+05 2.62E-06 3.64E-01 2.66E-06 3.69E-01 Core Damage 5.02E-05 4.16E+00 5.18E-05 4.90E+00 ILRT Dose Rate from 3a and 3b = 2.23E-01 1.92E-01

% of Total = 5.35% 3.93%

LERF from 3b = 5.22E-07 4.51E-07 CCFP%Liner= 44.74% 53.72%

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 165 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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) 8.79E-02 7.60E-02 3 to 15-1/2 Increase (Person-rem/ry) 1.83E-01 1.58E-01 10 to 15-1/2 Delta-LERF 2.51E-07 2.17E-07 3 to 15-1/2 Delta-LERF 4.45E-07 3.84E-07 10 to 15-1/2 Delta-CCFP 0.5% 0.4%

3 to 15-1/2 Delta-CCFP 0.9% 0.7%

3 to 10 Delta-LERF from Corrosion 1.54E-08 1.33E-08 10 to 15-1/2 Delta-LERF from Corrosion 1.12E-07 9.65E-08 Increase in LERF (ILRT 3-to-15-1/2 years) from Corrosion 1.27E-07 1.10E-07 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 166 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval ANALYSIS FILE: 17670-0001, Rev. 2, Appendix C C.1.0 CLIENT Nuclear Management Company - Point Beach Nuclear Power Plant C.2.0 TITLE Impact of Realistic ISGTR Conditional Probability Estimate on LERF/Delta LERF Results for Internal Events C.3.0 AUTHOR E. A. Krantz C.4.0 PURPOSE The purpose of this appendix is to describe a sensitivity analysis performed on the risk analysis for extending the Type A Integrated Leak Rate Test (ILRT) interval from ten to fifteen and a half years. This sensitivity analysis was performed to estimate the impact of refining the modeling of Induced Steam Generator Tube Rupture (ISGTR) from the conservative analysis approach described in NUREG-1570 (Reference C1) to a more realistic modeling approach using an Accident Progression Event Tree (APET).

C.5.0 INTENDED USE OF ANALYSIS RESULTS The results of this analysis will be used to support the position that the Large Early Release Frequency (LERF) values described in the main body of this report are very conservative and do not accurately depict the frequency of these events for the Point Beach Nuclear Power Plant.

C.6.0 TECHNICAL APPROACH The method used in this analysis was to first estimate a more realistic conditional probability of steam generator tube rupture for accident sequences resulting in core damage while at high primary pressure and with a dry steam generator than that provided by Reference C1. This reference suggests that the value for this conditional probability is conservatively 0.25 which was applied in the most recent Level 2 analysis update. This approach, however, yields LERF results that are overly conservative and, in fact, results in calculated values for LERF for the two units in excess of 1E-5 per reactor year and would preclude making any changes to plant operation that would cause an increase in LERF per Regulatory Guide 1.174 (Reference C2).

Attachment C-1 describes the approach taken to determine a more realistic conditional probability of ISGTR for the sequences of interest.

C.7.0 INPUT INFORMATION

1. Attachment C-1 provides the conditional probability considered to be realistic for the Point Beach units.

2. The main body of this report contains the analysis of extending the ILRT test interval associated with internal events and Appendix A contains the analysis of extending the test interval for the combination of internal events and external events. These analyses 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 167 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval describe the process; this appendix provides the result of that process when the realistic conditional probability of ISGTR is used in the Level 2 analysis in place of the conservative value.

3. The STC frequencies resulting from the Level 2 analysis using the realistic estimation of the conditional probability of ISGTR is provided in Tables C-1and C-1b.

C.

8.0 REFERENCES

C1. U.S. Nuclear Regulatory Commission, Risk Assessment of Severe Accident-Induced Steam Generator Tube Rupture, NUREG-1570, March 1998.

C2. 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.

C3. Point Beach Level 2 PRA, 2007 Update, [documentation in progress, date to be provided]

C.9.0 MAJOR ASSUMPTIONS:

The assumptions related to this sensitivity are the same as for the original analysis.

C.10.0 IDENTIFICATION OF COMPUTER CODES None used.

C.11.0 DETAILED ANALYSIS:

This analysis was conducted using the same analysis approach as the original analysis. The only change was in the frequencies of the source term categories from the sensitivity Level 2 quantification. The source term frequencies associated with this sensitivity are provided in Table C-1a and C-1b along with the change from the original analysis (baseline).

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 168 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table C-1a Unit 1 Source Term Category Results Sensitivity Level 2 Baseline Level 2 Comparison of Source Term Results for Unit 1 Results for Unit 1 Sensitivity to Baseline Category Delta  %

Frequency  % of Total Frequency  % of Total Frequency Change 1 3.46E-05 68.8% 2.83E-05 56.3% 6.29E-06 12.5%

2 2.82E-07 0.6% 2.31E-07 0.5% 5.08E-08 0.1%

3 2.62E-06 5.2% 2.62E-06 5.2% 0.00E+00 0.0%

4 8.73E-06 17.4% 7.31E-06 14.6% 1.42E-06 2.8%

5 2.74E-08 0.1% 2.24E-08 0.0% 5.09E-09 0.0%

6 1.50E-08 0.0% 1.50E-08 0.0% 0.00E+00 0.0%

7 2.37E-07 0.5% 2.37E-07 0.5% 0.00E+00 0.0%

8 3.73E-06 7.4% 1.15E-05 22.9% -7.77E-06 -15.5%

Total 5.02E-05 100.0% 5.02E-05 100.0%

Table C-1b Unit 2 Source Term Category Results Sensitivity Level 2 Baseline Level 2 Comparison of Source Term Results for Unit 2 Results for Unit 2 Sensitivity to Baseline Category Delta  %

Frequency  % of Total Frequency  % of Total Frequency Change 1 2.98E-05 57.6% 2.44E-05 47.2% 5.41E-06 10.4%

2 2.60E-07 0.5% 2.13E-07 0.4% 4.67E-08 0.1%

3 2.66E-06 5.1% 2.66E-06 5.1% 0.00E+00 0.0%

4 1.49E-05 28.8% 1.23E-05 23.8% 2.59E-06 5.0%

5 2.52E-08 0.0% 2.06E-08 0.0% 4.67E-09 0.0%

6 1.55E-08 0.0% 1.55E-08 0.0% 0.00E+00 0.0%

7 2.37E-07 0.5% 2.37E-07 0.5% 0.00E+00 0.0%

8 3.87E-06 7.5% 1.19E-05 23.0% -8.05E-06 -15.5%

Total 5.18E-05 100% 5.18E-05 100%

C.12.0 COMPUTER INPUT AND OUTPUT None C.13.0

SUMMARY

OF RESULTS Using the STC frequencies for the sensitivity provided in Tables C-1a and C-1b results in the values for critical parameters associated with LERF shown in Table C-2. This table also 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 169 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval compares the values of these parameters for both the baseline case and the realistic sensitivity case.

Table C-2 LERF Related ILRT Analysis Results Comparison (Baseline/Realistic)

Parameter Unit 1 Values Unit 2 Values Baseline Realistic Baseline Realistic Sensitivity Sensitivity Internal Events LERF 1.18E-05 4.01E-06 1.22E-05 4.15E-06 LERF10-15 1.40E-07 1.71E-07 1.21E-07 1.48E-07 LERF3-15 3.18E-07 3.89E-07 2.75E-07 3.36E-07 Combined 2.03E-05 6.90E-06 2.07E-05 7.05E-06 Internal/External LERF LERFCombined10-15 2.41E-07 2.95E-07 2.06E-07 2.51E-07 LERFCombined 3-15 5.48E-07 6.70E-07 4.68E-07 5.71E-07 LERFLiner10-15 1.46E-07 1.78E-07 1.26E-07 1.54E-07 LERFLiner3-15 3.27E-07 4.00E-07 2.82E-07 3.45E-07 C.

14.0 CONCLUSION

S The purpose of this sensitivity was to determine the LERF-related ILRT analysis results using realistic estimates of the conditional probability of ISGTR rather than the conservative value given in Reference C1.

Table C-2 provides the parameters of interest resulting from this analysis. This table indicates that the changes in LERF associated with increasing the Type A ILRT interval at Point Beach from the currently allowed 1-in-10 years to 1-in-15-1/2 years are consistent with the values included in Region II of Figure 4 of Reference C2. Because Reference C2 defines small changes in LERF as below 1E-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 from the realistic LERF perspective. Similarly, the LERF for moving from 3-in-10 years ILRT interval to 1-in-15-1/2 years also falls into Region II, Small Change in Risk, of the acceptance guidelines in Reference C2.

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 170 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval ATTACHMENT C-1 Estimation of the Realistic Condition Probability of ISGTR 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 171 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

SUBJECT:

Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval C-1.1. Introduction This attachment describes the estimation of a conditional probability for Induced Steam Generator Tube Rupture (ISGTR) that is more plant-specific to Point Beach Nuclear Plant.

In addition, this attachment presents the results of the Level 2 sensitivity analysis which utilized this conditional probability.

C-1.1.1. Basis for New Estimation of ISGTR The Point Beach Nuclear Plant Level 2 PRA analysis described in this report employed a generic value for the conditional probability of an ISGTR. A generic conditional probability of 0.25, as estimated in NUREG-1570, was applied to frequencies of high RCS pressure/dry steam generator (high/dry) sequences to arrive at the frequency of an ISGTR. An ISGTR is a containment bypass event, and is generally considered to contribute to the Large Early Release Frequency (LERF).

Of the many input variables with conservative values that comprise the generic 0.25 conditional probability, two input variables were found that were particularly conservative (i.e., produced a larger ISGTR conditional probability value, and therefore an overly-conservative LERF frequency) for the Point Beach Nuclear Plant: 1) The probability of high/dry sequences occurring concurrent with an RCP seal LOCA was too high, and 2) The probability of pressure-induced steam generator tube ruptures (PI-SGTRs) was too high.

The bases for, and the calculations of, the new estimates for these probabilities is provided in the sections below.

C-1.2. New ISGTR Variable Probabilities C-1.2.1. High/Dry Sequences with Seal LOCA The Accident Progression Event Tree (APET) that was developed in NUREG-1570 to estimate the conditional probability of an ISGTR included a top event that questioned whether a high/dry core damage sequence occurred concurrent with an RCP seal LOCA. If an RCP seal LOCA occurred, then the APET questioned whether loop seal clearing took place (APET branches B1, B2, and B3). The probability of a temperature-induced SGTR (TI-SGTR) is estimated as unity (1.0) in any steam generator attached to an RCS loop with a cleared loop seal.

The probability that a high/dry core damage sequence occurred concurrent with an RCP seal LOCA was estimated as 0.211 in NUREG-1570, based on the Surry Plant PRA core damage results, which are dominated by a Station Blackout (SBO) event.

The Point Beach Nuclear Plant PRA core damage results are dominated by a Loss of Service Water event. The RCP seals can be adequately cooled by one of three positive displacement charging pumps, which do not rely on component cooling water/service water for cooling. Thus, for Point Beach, the probability that a high/dry core damage sequence occurs concurrent with an RCP seal LOCA is significantly less than the probability calculated using the Surry results.

Table C-1.1 lists the Point Beach Unit 2 Extended Level 1 Event Tree high/dry sequences with concurrent RCP seal LOCA. These events are SBO, Loss of Offsite Power (T1), and Loss of Service Water (TSW). Note that the Loss of Component Cooling Water (TCC) event also has 17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 172 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval sequences in which an RCP seal LOCA can occur concurrent with a dry steam generator.

However, these TCC sequences have very low frequencies and were not included in this assessment.

For an SBO event, the Table C-1.1 list includes all sequences with failure of auxiliary feedwater (AFW), all sequences wherein the core was uncovered at the time AC power was recovered, and all sequences where AC power was not recovered. All of these sequences will exhibit an RCP seal LOCA and would most likely have dry steam generators. The sum of these frequencies is 1.22E-7/yr.

For a T1 event, the Table C-1.1 list includes all sequences with failure of AFW. However, failure of RCP seal cooling is not guaranteed for these sequences, merely compromised. A 6

conservatively high failure probability of 0.1 will be applied to the Table C-1.1 T1 frequencies to account for failure of RCP seal cooling, which would result in an RCP seal LOCA. The sum of these frequencies, including the application of the 0.1 probability, is 4.60E-7/yr.

For a TSW event, the Table C-1.1 list includes all sequences with failure of charging injection.

Failure of charging injection, together with the loss of service water, results in an RCP seal LOCA. Note that failure of AFW is not guaranteed for these sequences, although AFW would be compromised by the loss of service water. It was conservatively assumed, however, that all these sequences included failure of AFW for this analysis. The sum of these frequencies is 4.28E-8/yr.

The combined frequency of these high/dry sequences with concurrent RCP seal LOCA is 6.25E-7/yr.

The Point Beach Unit 2 Level 2 plant damage diagram results presented in this report (Figure 5) provides a total high/dry frequency of 4.21E-5/yr.

Therefore, the probability of a high/dry sequence having a concurrent RCP seal LOCA is (6.25E-7)/(4.21E-5) = 0.015.

The Point Beach Unit 1 Level 1 and Level 2 results are very similar to the Unit 2 results.

Therefore, the estimated probability from above will be applied in the Unit 1 ISGTR estimation.

C-1.2.2. PI-SGTR The Accident Progression Event Tree (APET) that was developed in NUREG-1570 to estimate the conditional probability of an ISGTR utilized the probability of a pressure-induced SGTR (PI-SGTR) that was based on an NRC RES Branch-developed flaw distribution for steam generators with moderate degradation.

The probability of a PI-SGTR for a high/dry sequence was estimated in NUREG-1570 as 0.0549 per depressurized steam generator, based on the RES-developed flaw distribution.

The NRC recognized the sensitivity of the ISGTR results to the assumed flaw distribution. Two analyses were conducted in NUREG-1570 to address this sensitivity.

The first sensitivity analysis assumed the steam generator tubes were in pristine condition (i.e., all steam generator flaws were eliminated or the steam generators were replaced). The results of 6

The Point Beach RCP seal LOCA probability, for a T1 event, ranges from 0.004 to 0.02 depending on whether a diesel generator is available or the gas turbine is available.

17670-0001 PB ILRT Rev 2.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval this case are contained in NUREG-1570 Table 5.8 as Case 8. The frequency of a PI-SGTR is shown to be 0/yr. Note that the frequency of a TI-SGTR decreased approximately 30% for this case as well.

The second sensitivity analysis was based on an NRC NRR Branch-developed flaw distribution for steam generators for an average plant. NUREG-1570 Section 5.3.3 states that the pressure-induced tube failure probabilities using the NRR distribution are an order of magnitude lower than those on the basis of the RES distribution. Note that the probability of a TI-SGTR decreased as well.

The Point Beach Unit 1 steam generators were replaced in the mid-1980s and the Unit 2 steam generators were replaced in the mid-1990s. None of the generators have experienced any significant degradation. The current number of tubes plugged is provided below:

Point Beach Unit 1 Steam Generators (SGs)

SG A (3214 total tubes) - 4 tubes plugged = 0.1245%

SG B (3214 total tubes) - 6 tubes plugged = 0.1867%

Point Beach Unit 2 SGs SG A (3499 total tubes) - 0 tubes plugged = 0%

SG B (3499 total tubes) - 4 tubes plugged = 0.1143%

The insignificant amount of degradation experienced by the Point Beach replacement steam generators suggests that the PI-SGTR probability can be reduced.

The sensitivity analyses presented in NUREG-1570 would indicate a percent reduction in the probability of PI-SGTR ranges between 100% reduction (Case 8) and 90% reduction (NRR-developed flaw distribution) for steam generators with insignificant amounts of degradation.

A reduction of 95% (i.e., a multiplier of 0.05) will be applied to the probability of a PI-SGTR. Note that the insignificant amount of degradation experienced by the Point Beach replacement steam generators would result in a reduction in the probability of a TI-SGTR as well. Reduction in the probability of a TI-SGTR was conservatively ignored in this analysis.

17670-0001 PB ILRT Rev 2.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval C-1.3. Estimation of ISGTR Probability C-1.3.1. APET Fault Tree Development The Accident Progression Event Tree (APET) that was developed in NUREG-1570 was developed into a fault tree so that a new estimation of the ISGTR probability could be generated using the Point Beach plant-specific values generated in Sections C-1.2.1 and C-1.2.2, above.

Figure C-1.1 provides this fault tree. All APET RC-1 end states were captured in the fault tree.

Some APET branches were combined to reduce the size of the fault tree. For example, fault tree page 4 shows that, for APET branch C1, the conditional probability that valve leakage occurs in more than one steam generator (0.22) is combined with the conditional probability that 3-of-3 steam generators have valve leakage (0.0797), resulting in a probability of 0.0175 applied in the 7

fault tree . The probabilities shown in the Figure 1 fault tree are the same as those applied in the APET.

Figure C-1.2 presents the results of the quantified fault tree. The conditional probability of an ISGTR, using the APET probabilities, is 0.222. This result essentially matches the NUREG-1570 stated approximate ISGTR conditional probability of 1-in-4 for high/dry sequences.

C-1.3.2. Point Beach Conditional ISGTR Probability A sensitivity analysis was conducted using the quantified fault tree described in Section C-1.3.1 to develop the conditional ISGTR probability for Point Beach.

Figure C-1.2 presents the inputs to, and results of, the sensitivity study.

A Point Beach plant-specific value of 0.015, as determined in Section C-1.2.1, above, was substituted for the probability of a concurrent RCP seal LOCA. The complement value (1 - 0.015

= 0.985) was also substituted. This is shown in Figure C-1.2.

A Point Beach plant-specific multiplier of 0.05, as determined in Section C-1.2.2, above, was applied to the PI-SGTR probabilities, resulting in the substitute values shown in Figure C-1.2.

The complement values were also substituted. Note that the PI-SGTR probabilities for APET paths A1, A2, and A3 were combined with other probabilities to reduce the size of the Figure 1 fault tree, as described in Section C-1.3.1. Because of these combinations, and the resulting complement combinations, the PI-SGTR probabilities for APET paths A1, A2, and A3 were not modified for convenience. This will result in a slightly higher ISGTR conditional probability.

Figure C-1.2 shows that the Point Beach conditional ISGTR probability is 0.076.

C-1.4. Level 2 PRA ISGTR Sensitivity Results C-1.4.1. Unit 1 7

The APET was developed based on the Surry Plant. As such, it models three steam generators for potential depressurization, as well as PI-SGTR and TI-SGTR. The Point Beach plants have two steam generators per plant. A review of the APET indicates that the changes to the split fractions to account for two steam generators versus three steam generators would result in insignificant changes to the results, while eliminating the probability for PI-SGTR or TI-SGTR based on three steam generators would result in a slight decrease in the ISGTR probability (primarily due to the reductions in APET Path A3).

17670-0001 PB ILRT Rev 2.doc

CLIENT: Nuclear Management Company BY: E. A. Krantz PAGE: 175 OF 198 FILE NO. 17670-0001, Rev. 2 CHECKED BY: G.W. Kindred Date: 09/14/07

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval The Level 2 decomposition event tree (DET) ISGTR.DET, as described in Reference C3, was modified to incorporate the Point Beach conditional ISGTR probability of 0.076.

The Point Beach Unit 1 Level 2 model was then requantified.

Figure C-1.3 presents the requantified Unit 1 containment event tree (CET) results. Figure C-1.4 presents the requantified Unit 1 source term category (STC) results.

C-1.4.2. Unit 2 The Point Beach Unit 2 Level 2 model was requantified with the modified ISGTR.DET.

Figure C-1.5 presents the requantified Unit 2 CET results. Figure C-1.6 presents the requantified Unit 2 STC results.

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table C-1.1 Point Beach Unit 2 Extended Level 1 Event Tree High/Dry Sequences with Concurrent RCP Seal LOCA Extended Level 1 Sequence Frequency ESBO-28 0.00E+00 ESBO-29 4.08E-10 ESBO-30 2.18E-10 ESBO-31 7.04E-08 ESBO-32 5.91E-09 ESBO-33 7.17E-10 ESBO-34 0.00E+00 ESBO-35 0.00E+00 ESBO-36 1.31E-08 ESBO-37 0.00E+00 ESBO-38 2.21E-10 ESBO-45 0.00E+00 ESBO-46 0.00E+00 ESBO-47 0.00E+00 ESBO-48 7.20E-10 ESBO-49 6.04E-11 ESBO-50 7.33E-12 ESBO-51 0.00E+00 ESBO-52 0.00E+00 ESBO-53 1.91E-10 ESBO-54 0.00E+00 ESBO-55 3.96E-12 ESBO-76 0.00E+00 ESBO-77 0.00E+00 ESBO-78 0.00E+00 ESBO-79 1.65E-10 ESBO-80 0.00E+00 ESBO-81 0.00E+00 ESBO-82 0.00E+00 ESBO-83 2.91E-08 ESBO-84 0.00E+00 ESBO-85 9.73E-10 ESBO-86 0.00E+00 ET1-20 0.00E+00 ET1-21 0.00E+00 ET1-22 1.69E-07 ET1-23 0.00E+00 ET1-24 0.00E+00 17670-0001 PB ILRT Rev 2.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval ET1-25 0.00E+00 ET1-26 7.18E-09 ET1-27 0.00E+00 ET1-28 1.19E-07 ET1-29 8.44E-09 ET1-30 1.01E-09 ET1-31 9.58E-09 ET1-32 1.93E-09 ET1-33 7.59E-07 ET1-34 4.35E-09 ET1-35 3.46E-06 ET1-72 3.23E-11 ET1-73 4.41E-10 ET1-74 1.21E-09 ET1-75 0.00E+00 ET1-76 1.34E-09 ET1-77 5.85E-12 ET1-78 1.80E-09 ET1-79 3.51E-11 ET1-80 2.22E-09 ETSW-11 5.42E-09 ETSW-12 6.16E-10 ETSW-13 0.00E+00 ETSW-14 1.54E-09 ETSW-15 0.00E+00 ETSW-16 4.97E-09 ETSW-17 0.00E+00 ETSW-18 2.98E-08 ETSW-30 4.07E-10 17670-0001 PB ILRT Rev 2.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Figure C-1.1 17670-0001 PB ILRT Rev 2.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval 17670-0001 PB ILRT Rev 2.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Figure C-1.2

===================================================================

WinNUPRA 3.0 Production (SR-3) Licensed to: Mark Walz Sensitivity Analysis

===================================================================

EQN File: C:\TEMP\ISGTR.EQN BED File: C:\TEMP\ISGTR.BED Data Source: BED File Values Orig. Top Event Unavailability: 2.219e-001 New Top Event Unavailability: 7.590e-002 Change: -65.79 %

===================================================================

Event Name Probability New Prob.

===================================================================

0 OF 2 SG LEAK 6.300e-001 0 SG DEP 4.050e-001 1 OF 2 SG LEAK 8.850e-001 1 OF 3 SG LEAK 7.800e-001 1 SG DEP 5.950e-001 1 SG LP SEAL CLE 3.330e-001 1 SG NO SEAL CLE 6.670e-001 1 SG PI-SGTR 5.490e-002 2.745e-003 1 SG TI-SGTR 1.840e-002 2 OF 2 SG LEAK 1.150e-001 2 OF 3 SG LEAK 2.020e-001 2 SG LP SEAL CLE 6.670e-001 2 SG NO SEAL CLE 3.330e-001 2 SG PI-SGTR 1.070e-001 5.350e-003 2 SG TI-SGTR 3.650e-002 3 OF 3 SG LEAK 1.750e-002 3 SG DEP 3.570e-001 3 SG PI-SGTR 1.560e-001 7.800e-003 3 SG TI-SGTR 5.420e-002 A1 0 SG TI-SGTR 8.700e-003 A1 1 SG TI-SGTR 4.880e-002 A1 2 SG TI-SGTR 6.680e-002 A2 1 SG TI-SGTR 4.880e-002 A2 2 SG PI-SGTR 8.100e-002 A2 2 SG TI-SGTR 5.770e-002 A2 3 SG PI-SGTR 1.050e-001 A2 3 SG TI-SGTR 6.670e-002 A3 3 SG TI-SGTR 8.460e-002 B2-63 SG TI-SGTR 1.840e-001 B2-69 SG TI-SGTR 3.210e-001 B3 SG TI-SGTR 1.000e+000 C1 VALVE LEAKAGE 5.000e-001 INTACT 8.640e-001 NO 1 SG PI-SGTR 9.450e-001 9.973e-001 NO 2 SG PI-SGTR 8.930e-001 9.947e-001 NO 3 SG DEP 6.430e-001 NO 3 SG PI-SGTR 8.440e-001 9.922e-001 NO HOLDUP 100 5.140e-001 NO HOLDUP 103 3.400e-001 NO HOLDUP 105 3.520e-001 NO HOLDUP 69 5.210e-001 NO HOLDUP 72 5.000e-001 NO HOLDUP 74 5.140e-001 NO HOLDUP 77 3.330e-001 NO HOLDUP 79 3.520e-001 NO HOLDUP 98 5.050e-001 17670-0001 PB ILRT Rev 2.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval NO SEAL LOCA 7.890e-001 9.850e-001 SEAL LOCA 2.110e-001 1.500e-002 SORV 1.360e-001 VALVE LEAKAGE 3.700e-001 17670-0001 PB ILRT Rev 2.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Figure C-1.3 17670-0001 PB ILRT Rev 2.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Figure C-1.4 17670-0001 PB ILRT Rev 2.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Figure C-1.5 17670-0001 PB ILRT Rev 2.doc

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Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Figure C-1.6 17670-0001 PB ILRT Rev 2.doc