ML021070279

License Amendment Request for One-Time Extension of Containment Integrated Leakage Rate Test Interval

ML021070279
Person / Time
Site:	Cook
Issue date:	04/11/2002
From:	Joseph E Pollock Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP:NRC:2612, RG-1.174
Download: ML021070279 (51)

Text

Indiana Michigan Power Company 500 Circle Drive Buchanan, Ml 49107 1373 INDIANA MICHIGAN POWER April 11, 2002 AEP:NRC:2612 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001

SUBJECT:

Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 License Amendment Request for One-time Extension of Containment Integrated Leakage Rate Test Interval

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant Unit 1 and Unit 2, proposes to amend Appendix A, Technical Specifications (TS), of Facility Operating Licenses DPR-58 and DPR-74. I&M proposes to revise the Surveillance Requirements for containment leakage rate testing in TS 4.6.1.2 to allow a one-time extension of the interval between integrated leakage rate tests (ILRTs) from 10 to 15 years.

The proposed amendment would provide savings in radiation exposure to personnel, cost, and critical path time during the 2003 refueling outages for Units 1 and 2.

The proposed amendment is risk-informed and follows the guidance in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis." In accordance with RG 1.174, I&M has performed an analysis showing that the increases in risk resulting from the proposed amendment are small and within established guidance. I&M has also determined that defense-in-depth principles will be maintained based on both risk and non risk based considerations. to this letter provides an oath and affirmation affidavit pertaining to the requested amendment. Enclosure 2 provides a detailed description and safety analysis to support the proposed amendment, including an evaluation of significant hazards considerations pursuant to 10 CFR 50.92(c) and an 7ý D17

U. S. Nuclear Regulatory Commission AEP:NRC:2612 Page 2 environmental assessment. Attachments 1A and 1B provide TS pages that are marked to show the proposed changes for Unit 1 and Unit 2, respectively.

Attachments2A and 2B provide TS pages with the proposed changes incorporated for Unit 1 and Unit 2, respectively. Attachment 3 provides a risk impact assessment for extending the ILRT interval. There are no new regulatory commitments made in this letter.

I&M requests approval of the proposed amendment by October 15, 2002, to support planning for the next Unit 2 refueling outage. Once approved, the amendment will be implemented within 45 days.

No pending amendment requests affect the TS pages that are submitted in this request. If any future submittals affect these TS pages, I&M will coordinate the changes to the pages with the NRC Project Manager to ensure proper TS page control when the associated license amendment requests are approved.

If you have any questions or require additional information, please contact Mr. Gordon P. Arent, Manager of Regulatory Affairs, at (616) 697-5553.

Sincerely, J. E. Pollock Site Vice President

/jen

Enclosures:

1 Affidavit 2 Evaluation of the Proposed Changes Attachments:

1A and 1B Technical Specification Pages Marked To Show Proposed Changes 2A and 2B Proposed Technical Specification Pages 3 Risk Impact Assessment for Extending Containment Type A Test Interval

U. S. Nuclear Regulatory Commission AEP:NRC:2612 Page 3 c: K. D. Curry J. E. Dyer MDEQ - DW & RPD NRC Resident Inspector R. Whale

Enclosure 1 to AEP:NRC:2612 AFFIRMATION I, Joseph E. Pollock, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.

American Electric Power Service Corporation J. E. Pollock Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THI_1L DAY OF .fi _L. ,2002 My Commission Expires / Z_/J! _

to AEP:NRC:2612 Page I Application for Amendment Technical Specification 4.6.1.2, Surveillance Requirements for Containment Leakage Rate Testing

1.0 DESCRIPTION

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (I&M), the licensee for Donald C.

Cook Nuclear Plant (CNP) Unit 1 and Unit 2, proposes to amend Appendix A, Technical Specifications (TS), of Facility Operating Licenses DPR-58 and DPR-74. I&M proposes to revise the Surveillance Requirements for containment leakage rate testing in TS 4.6.1.2 to allow a one time extension of the interval between integrated leakage rate tests (ILRTs) from 10 to 15 years. The proposed amendment would provide savings in radiation exposure to personnel, cost, and critical path time during the 2003 refueling outages for Units 1 and 2.

The proposed amendment is risk-informed and follows the guidance in Regulatory Guide (RG) 1.174 (Reference 1). In accordance with RG 1.174, I&M has performed an analysis showing that the increases in risk resulting from the proposed amendment are small and within established guidance. I&M has also determined that defense-in-depth principles will be maintained based on both risk and non-risk based considerations.

2.0 PROPOSED CHANGE

The proposed change would add a note to Unit 1 and Unit 2 TS 4.6.1.2 stating that the ILRT (Type A test) frequency specified in the Nuclear Regulatory Commission (NRC)-endorsed industry guideline (NEI 94-01) as "...at least once per 10 years based on acceptable performance history" is changed to "...at least once per 15 years based on acceptable performance history."

The note would also state that the change applies only to the interval following the previous ILRT for the respective unit. Attachments 1A and 1B provide TS pages that are marked to show the proposed changes for Unit 1 and Unit 2, respectively. Attachments 2A and 2B provide TS pages with the proposed changes incorporated for Unit 1 and Unit 2, respectively.

3.0 BACKGROUND

The CNP Unit 1 and Unit 2 containment buildings are reinforced concrete structures with a steel liner plate capable of withstanding a design internal pressure of 12 pounds per square inch gage.

The containment systems include an ice condenser capable of rapidly absorbing the energy released by a loss of coolant accident. The containment buildings and associated systems are described in detail in Chapter 5 of the CNP Updated Final Safety Analysis Report (UFSAR).

to AEP:NRC:2612 Page 2 Current Requirements TS 4.6.1.2 currently requires that leakage rate testing of the containment be performed in accordance with 10 CFR 50, Appendix J, Option B, except as modified by NRC-approved exemptions, and in accordance with RG 1.163 (Reference 2). Regulatory Position C.1 of RG 1.163 states that licensees should establish test intervals based upon the criteria in Section 11.0 of NEI 94-01 (Reference 3). Section 11.0 of NEI 94-01 references Section 9.0 which allows ILRTs to be performed at a frequency of once per 10 years if the calculated leakage rate for two consecutive previous tests is less than 1.0 La. La is defined in the CNP Unit 1 and Unit 2 TS as 0.25 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa, which is 12.0 psig.

The CNP Unit 1 and Unit 2 containments have met this criterion and therefore qualify for the 10 year frequency.

Section 9.0 of NEI 94-01, however, also allows a 15-month extension of the ILRT test interval

"...in cases where refueling schedules have been changed to accommodate other factors." I&M considers that the extended Unit 1 and Unit 2 shutdown that began in 1997 and ended in 2000 changed the subsequent refueling schedules to such an extent that the 15-month extension allowance may be applied to both units. Since the ILRTs were last completed for Unit 1 and Unit 2 on October 1, 1992, and May 12, 1992, respectively, the current due dates for the next ILRTs are January 1, 2004, and August 12, 2003, respectively. Compliance with these due dates would require that ILRTs be performed for Unit 1 and Unit 2 during their 2003 refueling outages.

Basis for Current Requirements The limitations on containment leakage rates in the TS 3.6.1.2 Limiting Condition for Operation ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure. As an added conservatism, the Action for TS 3.6.1.2 limits the measured overall integrated leakage rate to less than or equal to 0.75 La during performance of periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.

The performance-based ILRT requirements of Option B of 10 CFR 50, Appendix J, provide an alternative to the 3 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 1 test per 10 years if certain performance criteria are met. The basis for the 1 test per 10-year frequency is described in Section 11.0 of NEI 94-01, which states that NUREG-1493 (Reference 4) 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 to AEP:NRC:2612 Page 3 are documented in Electric Power Research Institute (EPRI) report TR-104285 (Reference 5).

The EPRI study determined a reduction in the frequency of ILRTs from 3 tests per 10 years to 1 test per 10 years would result in an incremental risk contribution of 0.035 percent. This value is comparable to the range of risk increases (0.002 percent to 0.14 percent) presented in NUREG 1493 for the same frequency reduction. Additionally, NUREG-1493 described the increase in risk resulting from an even lower frequency, 1 test per 20 years, as "imperceptible."

Reason for Requesting Amendment Extension of the ILRT interval from 10 years to 15 years would eliminate the need to perform an ILRT for both CNP units during their 2003 outages. This would save a total of approximately 0.5 person rem exposure. This would also result in an estimated monetary savings of $200,000 per test, and save an estimated 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> of critical path time per outage, at $20,000 per hour.

The total monetary savings for both units would therefore be approximately $2.08 million. I&M is requesting this license amendment to obtain these personnel exposure and monetary savings.

4.0 TECHNICAL ANALYSIS

The proposed amendment would authorize a one-time extension of the ILRT interval from 10 years to 15 years for CNP Units 1 and 2. The proposed amendment is supported by both risk and non-risk based considerations.

Risk Based Considerations An analysis was performed for I&M on the risk from changing the containment ILRT frequency from once per 10 years to once per 15 years. The analysis was performed in accordance with NEI 94-01 guidelines and the NRC guidance in RG 1.174. The methodology used was similar to that presented in EPRI TR-104285 and NUREG-1493, and incorporated revised guidance and additional information from EPRI and NEI (References 6 and 7). The analysis used a simplified bounding analysis approach to evaluate the change in risk associated with increasing the ILRT interval from 10 to 15 years, by examining the CNP Individual Plant Examination (IPE)

(Reference 8) plant specific accident sequences in which the containment remains intact or the containment is impaired. The analysis is provided as Attachment 3 to this letter.

The change in plant risk was evaluated by quantifying the change in the predicted person-rem/year and the change in Large Early Release Frequency (LERF) that would result from the proposed ILRT interval extension. To accomplish this, specific accident sequences in the IPE were identified and frequencies were determined where appropriate. The sequences were as follows:

to AEP:NRC:2612 Page 4

" Core damage sequences in which the containment remains intact (EPRI TR-104285 Class 1 sequences).

" Core damage sequences in which containment integrity is impaired due to random isolation failures of components other than those subject to Type B and C testing as specified in 10 CFR 50, Appendix J (EPRI TR-104285 Class 3 sequences, which were further divided into Class 3A, small containment leak sequences, and Class 3B, large containment leak sequences).

" Core damage sequences in which containment integrity is impaired due to isolation pathways left open following post-maintenance testing (EPRI TR-104285, Class 6 sequences).

" Accident sequences involving containment failure induced by severe accident phenomena (EPRI TR-104285 Class 7 sequences), containment bypass (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 Class 4 and 5 sequences would be impacted by changes in Type B and C test intervals, rather than changes in the ILRT intervals. Therefore, these were not evaluated further.

The following steps were then performed:

"* The base-line risk in terms of frequency per reactor-year for accident classes 1, 2, 3a, 3b, 6, 7, and 8 was quantified.

"* Person-rem doses (population dose) per reactor-year were calculated for accident classes 1, 2, 3a, 3b, 6, 7, and 8.

"* The risk impact of changing the ILRT frequency from 10 to 15 years was evaluated.

" The change in risk in terms of LERF was determined. 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 result in a large release due to failure to detect a pre-existing leak during the relaxation period. The determination of impact on the LERF was, therefore, based on Class 3B sequences since they have the potential to result in large releases.

"* The change in conditional containment failure probability was determined for the proposed and cumulative changes to the ILRT interval.

The analysis results are summarized below:

to AEP:NRC:2612 Page 5

1. The increase in the total integrated plant risk (person rem/year within 50 miles) resulting from reducing the ILRT frequency from 1 test per 10 years to 1 test per 15 years was estimated to be 0.025 percent. The increase in the total integrated plant risk resulting from reducing the ILRT frequency from 3 tests per 10 years to 1 test per 15 years was estimated to be 0.06 percent. These increases in risk are reasonable when compared to the range of increase, 0.02 to 0.14 percent, estimated in NUREG-1493 for reducing the ILRT frequency from 3 tests per 10 years to the 1 test per 10 years allowed by Option B to 10 CFR 50, Appendix J. As noted above under "Basis for Current Requirements," NUREG-1493 indicates that such increases are "imperceptible."

2. The increase in the LERF resulting from reducing the ILRT frequency from 1 test per 10 years to 1 test per 15 years was estimated to be 3.4 x 10 8/year. The increase in the LERF resulting from reducing the ILRT frequency from 3 tests per 10 years to 1 test per 15 years was estimated to be 8.2 x 10-8/year. RG 1.174 provides guidelines for acceptable changes in LERFs resulting from proposed changes to a plant's licensing basis. The LERF increases calculated for this proposed amendment are below the LERF acceptance guideline of less than 10-7 given in RG 1.174. RG 1.174 also provides guidelines for acceptable changes in the core damage frequencies (CDFs) resulting from a proposed change to a plant's licensing basis. However, the CDF would not be affected by a change to the ILRT interval. Therefore, the change in LERF is the sole RG 1.174 guideline applicable to this proposed change.

3. The increase in conditional containment failure probability from reducing the ILRT frequency from 1 test per 10 years to 1 test per 15 years was estimated to be 0.47 percent.

The increase in conditional containment failure probability from reducing the ILRT frequency from 3 tests per 10 years to 1 test per 15 years was estimated to be 1.1 percent.

These small changes demonstrate that, consistent with the defense-in-depth guidelines provided in RG 1.174, the proposed amendment would preserve a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation.

In reviewing similar amendment requests from other licensees, the NRC has noted the potential for degradation on the uninspectable (embedded) side of the containment liner. The NRC has questioned how potential leakages due to age-related degradation were considered in risk assessments for extending the ILRT interval. The potential for containment leakage was explicitly included in the CNP risk assessment. By definition, the intact containment cases, EPRI Class 1, include a leakage term that is independent of the source of the leak. Similarly, Classes 3a qne 3b model the potential leakage impact of the ILRT interval extension. These cases include the potential that the leakage is due to a containment liner failure. The assessment results show that, even with the potential to have an undetected containment flaw or leak path, the increase in risk is small.

to AEP:NRC:2612 Page 6 The above results demonstrate that the increases in risk and LERF resulting from the proposed amendment are within established guidelines and that defense-in-depth principles would be maintained.

Non-Risk Based Considerations Consistent with the defense-in-depth philosophy provided in RG 1.174, I&M has assessed non risk based considerations relevant to the proposed amendment. These are discussed below.

ILRT History TS 4.6.1.2 requires measurement of the containment leakage rate. TS 3.6.1.2 establishes the limit for the measured overall integrated containment leakage rate as 0.75 La (i.e., 0.1875 weight percent) of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa. The table below provides a history of the ILRT results for Units 1 and 2.

Unit 1 ILRT History Unit 2 ILRT History Completion Leakage in Completion Leakage in Date weight percent Date weight percent per day* per day*

10/1/92 0.0792 5/12/92 0.0344 6/4/89 0.1046 2/13/89 0.0454 8/30/85 0.0168 6/16/84 0.0387 7/30/81 0.0582 5/4/81 0.0575 5/29/78 0.0654 10/2/77 0.0337 11/24/74 0.1604

• 95 percent upper confidence limit using Total Time method identified in 10 CFR 50, Appendix J, Section III.A.3(a).

The test summary reports for the above ILRTs indicate that, with one potential exception, there were no test failures resulting from containment pressure boundary leakage. The one potential exception occurred during the Unit 1, July 30, 1981, ILRT. During the initial pressurization phase of the test, a flange leak, a packing leak, and a valve vent leak were repaired without first quantifying the leakages. Therefore, it cannot be stated with certainty that the containment would have met the leakage rate criteria stated in the TS in its as-found condition. However, the other ILRT results indicate that the containments have consistently met the specified leakage criteria in their as-found condition.

to AEP:NRC:2612 Page 7 Local Leakage Rate Testing 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 CNP local leakage rate testing requirements are unaffected by this proposed amendment. The local leakage rate testing program will, therefore, provide continuing assurance that the most likely sources of leakage will be identified and repaired.

Containment Inservice Inspection Program CNP has established a containment inservice inspection program that implements the requirements for examination and testing of ASME Section XI and 10 CFR 50.55a Class MC and Class CC components. This program was developed in accordance with the requirements of the 1992 Edition with the 1992 Addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Division 1, Subsections IWE and IWL, as modified by NRC final rulemaking of 10 CFR 50.55a, published in the Federal Register on August 8, 1996. The scope of the program includes the containment liner (including penetrations), fuel transfer tube, equipment access hatch, personnel air locks, and concrete containment structure. The first ten-year inspection interval has been established from September 9, 1998, to September 9, 2008. The containment inservice inspection program is unaffected by the proposed amendment, and will continue to provide a high degree of assurance that any degradation of the containment will be detected and corrected before it can result in a leakage path.

Approved Alternatives to Subsection IWE and IWL Requirements There are no alternatives to Subsection IWE and IWL requirements approved for CNP that credit the performance of integrated leakage rate testing. There is one approved (Reference 9) alternative, CISIR-01, that credits local leakage rate testing (LLRT). As noted in the NRC review of CISIR-01, the pressure retaining capability of seals and gaskets is verified by the performance of LLRTs once each inspection period. LLRT requirements are unaffected by the proposed amendment.

Containment Inspection History ISI program examinations performed on both units since their most recent ILRTs include visual examinations of the liners pursuant to Subsection IWE, which were conducted from 1996 through 2000. With the exception of the conditions described below, all results were within the established acceptance criteria and there were no areas that required augmented examination in accordance with IWIE Subsection 1240. ISI examinations have also been performed during the course of repair and replacement activities.

to AEP:NRC:2612 Page 8

" In 1998, an inspection of the Unit 1 containment identified corrosion and pitting of the steel liner plate along the moisture barrier seal near the containment cylinder base at elevation 598 feet, 9-3/8 inches. Similar conditions were found in Unit 2. Based on a detailed engineering analysis, I&M concluded that the structural integrity of the containments to withstand normal operating, design basis accident, and severe accident loads was not affected by the as-found condition of the liner. The leak tight integrity of the containments was not impaired, and the liners, as-found, would have fulfilled their function as effective leak-tight barriers. This condition was documented in an I&M letter to the NRC dated March 3, 2000 (Reference 10). Modifications were made to the floor-liner seal to prevent further degradation of the liner. This issue was included in a special inspection pursuant to NRC Manual Chapter 0350, "Staff Guidelines for Restart Approval." Closure of the issue was documented in an NRC Inspection Report dated January 19, 2000 (Reference 11).

" In 1999, an inspection of the containment liner identified an apparent weld repair of the liner plate. Surface preparation to allow further inspection dislodged repair material exposing a hole through the liner plate. The hole was circular in appearance with a diameter of approximately 3/16" on the exterior surface and 3/4" on the interior surface. It appeared that the liner hole resulted from an inadequate repair of a hole drilled in error during plant construction. After the damaged liner plate section was cut out, a piece of wood, determined to be the handle of a wire brush, was found embedded in the concrete. Some minor corrosion was noted on the concrete side of the liner plate in the area of the embedded wire brush. This condition was also determined to be construction-related. The affected area of containment was restored to an acceptable design configuration. The repair was vacuum box tested and subjected to an LLRT. This condition was initially reported pursuant to 10 CFR 50.73 in LER 2000-001-00 (Reference 12). Additional evaluation of the condition determined that the containment structure would have performed its safety related function during a design basis accident in the as-found condition, and the LER was retracted by LER 2000-001-01 (Reference 13).

ISI program examinations performed on both units since their most recent ILRTs also included visual examinations of the containment exterior concrete pursuant to Subsection IWL, which were conducted during 2001. These examinations did not reveal any observations that could potentially affect the structural integrity or the calculated design safety margins of the containments.

Containment Penetration Bellows In reviewing similar amendment requests from other licensees, the NRC has noted that stainless steel containment penetration bellows have been found to be susceptible to trans-granular stress corrosion cracking. As documented in NRC Information Notice 92-20 (Reference 14), leakage through such bellows is not readily detectable by LLRTs. I&M evaluated the conditions to AEP:NRC:2612 Page 9 documented in the Information Notice and determined that the bellows at CNP are not part of the containment isolation barrier.

Maintenance Rule The containment function of limiting the release of radioactive fission products following an accident has been classified as highly safety significant and its condition is monitored pursuant to 10 CFR 50.65 in accordance with the CNP Maintenance Rule program. The monitoring sources include subsection IWE liner inspections, LLRTs, and the results of TS 4.6.1.1 surveillance requirements pertaining to isolation valve, equipment hatch, and airlock status. None of these monitoring sources is affected by the proposed statement.

5.0 Remulatory Safety Analysis 5.1 No Significant Hazards Consideration I&M has evaluated whether or not a significant hazards consideration is involved with the proposed change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?

Response: No Probability of Occurrence of an Accident Previously Evaluated The proposed change to extend the ILRT interval from 10 to 15 years does not affect any accident initiators or precursors. The containment liner function is purely mitigative. There is no design basis accident that is initiated by a failure of the containment leakage mitigation function. The extension of the ILRT will not create any adverse interactions 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.

Consequences of an Accident Previously Evaluated The potential consequences of the proposed change have been quantified by analyzing the changes in risk that would result from extending the ILRT 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 imperceptible. I&M has also analyzed the increase in risk in terms of the frequency of large early releases from accidents. The to AEP:NRC:2612 Page 10 increase in the large early release frequency resulting from the proposed extension was determined to be within the guidelines published in Regulatory Guide 1.174.

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. I&M has determined that the increase in conditional containment failure probability from reducing the ILRT frequency from 1 test per 10 years to 1 test per 15 years would be small. Continued containment integrity is also assured by the history of successful ILRTs, and the established programs for local leakage rate testing and inservice inspections which are unaffected by the proposed change. Therefore, the consequences of an accident previously analyzed are not significantly increased.

In summary, the probability of occurrence and the consequences of an accident previously evaluated are not significantly increased.

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

Response: No The proposed change to extend the ILRT interval from 10 to 15 years does not create any new or different accident initiators or precursors. The length of the ILRT 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. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The risk-based margins of safety associated with the containment ILRT are those associated with the estimated person-rem per year, the large early release frequency, and the conditional containment failure probability. I&M 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 ILRT are those involved with its structural integrity and leak tightness. The proposed change to extend the ILRT 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.

to AEP:NRC:2612 Page I1I Therefore, the proposed changes do not involve a significant reduction in margin of safety.

In summary, based upon the above evaluation, I&M has concluded that the proposed changes involve 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.

5.2 Applicable Regulatory Regquirements/Criteria 5.2.1 Regulations TS 4.6.1.2 currently requires that leakage rate testing of the containment be performed in accordance with 10 CFR 50, Appendix J, Option B, except as modified by NRC-approved exemptions, and in accordance with RG 1.163.

Regulatory Position C.1 of RG 1.163 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, as described above under "Background,"

would require that ILRTs be performed for Unit 1 and Unit 2 within 10 years plus 15 months from the date of their last performance.

I&M is proposing a license amendment that would modify TS 4.6.1.2 to allow a one-time extension of this interval to 15 years. The techmical analysis for the proposed license amendment is based on risk related and non-risk related considerations. A risk analysis was performed for I&M showing that the increases in estimated person-rem and LERF are consistent with guidance provided in RG 1.174 and NUREG-1493. I&M 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 ILRT and containment inspection history, and the ongoing LLRT and ISI programs. The technical analysis provides the basis for I&M's determination that the proposed amendment does not involve significant hazards considerations as described in 10 CFR 50.92.

No other regulations or TS will be affected by the proposed amendment.

5.2.2 UFSAR UFSAR Section 5.0, "Containment System," provides licensing basis information for the Unit 1 and Unit 2 containments. Subsection 5.7.3 describes periodic leakage rate testing of the containment. This subsection states that periodic leakage testing of the containment is performed in accordance with the TS, 10 CFR 50, Appendix J, Option B, and Regulatory Guide 1.163. The validity of to AEP:NRC:2612 Page 12 the statement is unaffected by the proposed amendment since the proposed extension will only apply to the current ILRT intervals and will not alter the accuracy of the statements as descriptions of normal requirements. Additionally, the proposed amendment does not affect any other aspect of the ILRT, such as test methodology, pressure, or acceptance criteria.

UFSAR Section 14, "Safety Analysis," for Unit 1 and Unit 2 provides descriptions of the licensing basis accident analyses for the respective units, including the relevant parameters for the analyses. The proposed amendment only involves the ILRT interval and does not affect any parameters, such as pressure or leakage rate, that can affect the results of these analyses.

In conclusion, based on the considerations discussed above, (1) 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.

6.0 Environmental Considerations I&M has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. I&M has determined that 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(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared concerning the proposed amendment.

7.0 Precedent Licensing Actions The NRC has approved one-time extensions of the ILRT interval to 15 years based on risk and non-risk based considerations for Waterford Steam Electric Station, Unit 3 (Reference 15), Peach Bottom Atomic Power Station, Unit 3 (Reference 16), Crystal River Nuclear Plant, Unit 3 (Reference 17), and Indian Point 3 Nuclear Power Plant (Reference 18).

to AEP:NRC:2612 Page 13 8.0 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," dated July 1998.

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

3. Nuclear Energy Institute document NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 26, 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. EPRI document "Interim Guidance for Performing Risk impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals,"

Rev. 4, dated November 2001.

7. NEI memo, "One-Time Extension of Containment Integrated Leak Rate Test Interval Additional Information," dated November 30, 2001.

8. Letter from E. E. Fitzpatrick, I&M, to NRC Document Control Desk, "Individual Plant Examination - Response to NRC Audit Concerns and Request for Additional Information,"

dated October 26, 1995.

9. Letter from C. M Craig, NRC, to R. P. Powers, I&M, "Donald C. Cook Nuclear Power Plant, Units 1 and 2 - Safety Evaluation for Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii) for Containment Inspections (TAC NO. MB1249 AND MB1250),

dated July 13, 2001.

10. Letter from M. W. Rencheck, I&M, to NRC Document Control Manager, transmitting LER 315/1998-011-03, "LER Retraction, Containment Liner Pitting," dated March 8, 2000.

11. Letter from J. A. Grobe, NRC, to R. P. Powers, I&M, "NRC Inspection Report 50-315/99026(DRS); 50-316/99026(DRS)," dated January 19, 2000.

to AEP:NRC:2612 Page 14

12. Letter from M. W. Rencheck, I&M, to NRC Document Control Manager, transmitting LER 316/2000-001-00: "Through-Liner Hole Discovered in Containment Liner," dated February 16, 2000.

13. Letter from J. E. Pollock, I&M, to NRC Document Control Manager, transmitting LER 316/2000-001-01: "Through-Liner Hole Discovered in Containment Liner," dated March 16, 2001.

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

15. Letter from N. Kalyanam, NRC, to J. E. Venable, Entergy Operations Incorporated, "Waterford Steam Electric Station, Unit 3, - Issuance of Amendment Re: Integrated Leakage Rate Testing Interval Extension (TAC No. MB2461)," dated February 14, 2002.

16. Letter from J. P. Boska, NRC, to 0. D. Kingsley, Exelon Generation Company, "Peach Bottom Atomic Power Station, Unit 3 - Issuance of Amendment Re: Extension of the Containment Integrated Leak Rate Test (TAC No. MB2094)," dated October 4, 2001.

17. Letter from J. M. Goshen, NRC, to D. E. Young, Florida Power Corporation, "Crystal River Unit 3 - Issuance of Amendment Regarding Containment Leakage Rate Testing Program (TAC No. MB 1349)," dated August 30, 2001.

18. Letter from G. F. Wunder, NRC, to M. Kansler, Entergy Nuclear Operations, Incorporated, "Indian Point Nuclear Generating Unit 3 - Issuance of Amendment Re: Frequency of Performance-Based Leakage Rate Testing (TAC No. MB 0178)," dated April 17, 2001.

Attachment lA to AEP:NRC:2612 TECHNICAL SPECIFICATIONS PAGES MARKED TO SHOW PROPOSED CHANGES REVISED PAGES UNIT 1 3/4 6-2

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.6 CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of < La, 0.25 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa, 12.0 psig, and

b. A combined leakage rate of < 0.60 La for all penetrations and valves subject to Types B and C tests when pressurized to Pa.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With either (a) the measured overall integrated containment leakage rate exceeding 0.75 La, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 La, restore the overall integrated leakage rate to < 0.75 L, and the combined leakage rate for all penetrations and valves subject to Types B and C tests to < 0.60 La prior to increasing the Reactor Coolant System temperature above 200 0F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 Perform leakage rate testing in accordance with 10 CFR 50 Appendix J Option B, except as modified by NRC-approved exemptions, and Regulatory Guide 1.163, dated September 1995.

• See Notes 1 afd2

a. Each containment air lock shall be verified to be in compliance with the requirements of Specification 3.6.1.3.

b. The provisions of Specification 4.0.2 are not applicable.

Notes.

• j* A one-time exception to the requirement to perform post-modification Type A testing is allowed for the steam generators and associated piping, as components of the containment barrier. For this case, ASME Section XI leak testing will be used to verify the leak tightness of the repaired or modified portions of the containment barrier. Entry into MODES 3 and 4 following the extended outage that commenced in 1997 may be made to perform this testing.

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 6-2 AMENDMENT -18,4460,1-7, -96, 2A9, 248

Attachment 1B to AEP:NRC:2612 TECHNICAL SPECIFICATIONS PAGES MARKED TO SHOW PROPOSED CHANGES REVISED PAGES UNIT 2 3/4 6-2

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.6 CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of _<La, 0.25 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa, 12 psig, and

b. A combined leakage rate of < 0.60 La for all penetrations and valves subject to Types B and C tests when pressurized to Pa.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With either (a) the measured overall integrated containment leakage rate exceeding 0.75 La, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 La, restore the overall integrated leakage rate to < 0.75 La and the combined leakage rate for all penetrations and valves subject to Types B and C tests to < 0.60 L, prior to increasing the Reactor Coolant System temperature above 2000F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 Perform leakage rate testing in accordance with 10 CFR 50 Appendix J Option B, except as modified by NRC-approved exemptions, and Regulatory Guide 1.163, dated September 1995.

Sce Noteý 1.

a. Each containment air lock shall be verified to be in compliance with the requirements of Specification 3.6.1.3.

b. The provisions of Specification 4.0.2 are not applicable.

Nutte:s, COOK NUCLEAR PLANT-UNIT 2 Page 3/4 6-2 AMENDMENT 44:2, 3, M--,229

Attachment 2A to AEP:NRC:2612 PROPOSED TECHNICAL SPECIFICATIONS PAGES REVISED PAGES UNIT 1 3/4 6-2

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.6 CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of _<La, 0.25 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa, 12.0 psig, and

b. A combined leakage rate of _ 0.60 La for all penetrations and valves subject to Types B and C tests when pressurized to Pa.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With either (a) the measured overall integrated containment leakage rate exceeding 0.75 La, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 La, restore the overall integrated leakage rate to < 0.75 La and the combined leakage rate for all penetrations and valves subject to Types B and C tests to

• 0.60 La prior to increasing the Reactor Coolant System temperature above 2000 F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 Perform leakage rate testing in accordance with 10 CFR 50 Appendix J Option B, except as modified by NRC-approved exemptions, and Regulatory Guide 1.163, dated September 1995.

See Notes 1 and 2.

a. Each containment air lock shall be verified to be in compliance with the requirements of Specification 3.6.1.3.

b. The provisions of Specification 4.0.2 are not applicable.

Notes:

I A one-time exception to the requirement to perform post-modification Type A testing is allowed for the steam generators and associated piping, as components of the containment barrier. For this case, ASME Section XI leak testing will be used to verify the leak tightness of the repaired or modified portions of the containment barrier. Entry into MODES 3 and 4 following the extended outage that commenced in 1997 may be made to perform this testing.

2 The Type A testing frequency specified in NEI 94-01, Revision 0, Paragraph 9.2.3, as "...at least once per 10 years based on acceptable performance history" is modified to be "...at least once per 15 years based on acceptable performance history." This change applies only to the interval following the Type A test performed in October 1992.

COOK NUCLEAR PLANT-UNIT I Page 3/4 6-2 AMENDMENT 48, 4-60, U87,196, 209,248,

Attachment 2B to AEP:NRC:2612 PROPOSED TECHNICAL SPECIFICATIONS PAGES REVISED PAGES UNIT 2 3/4 6-2

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.6 CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of < La, 0.25 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at P,, 12 psig, and

b. A combined leakage rate of < 0.60 La for all penetrations and valves subject to Types B and C tests when pressurized to Pa.

APPLICABILITY: MODES 1,2, 3 and 4.

ACTION:

With either (a) the measured overall integrated containment leakage rate exceeding 0.75 La, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 La, restore the overall integrated leakage rate to _<0.75 La and the combined leakage rate for all penetrations and valves subject to Types B and C tests to < 0.60 La prior to increasing the Reactor Coolant System temperature above 2000F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 Perform leakage rate testing in accordance with 10 CFR 50 Appendix J Option B, except as modified by NRC-approved exemptions, and Regulatory Guide 1.163, dated September 1995.

See Note 1.

a. Each containment air lock shall be verified to be in compliance with the requirements of Specification 3.6.1.3.

b. The provisions of Specification 4.0.2 are not applicable.

Notes:

1 The Type A testing frequency specified in NEI 94-01, Revision 0, Paragraph 9.2.3, as "...at least once per 10 years based on acceptable performance history" is modified to be "...at least once per 15 years based on acceptable performance history." This change applies only to the interval following the Type A test performed in May 1992.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 6-2 AMENDMENT 4-42, 3, V)3, Z^-9

Attachment 3 to AEP:NRC:2612 Risk Impact Assessment for Extending Containment Type A Test Interval

CLIENT: American Electric Power Corp. BY: E. R. Schmidt PAGE: 1 OF 24 FILE NO. 17141-0007-A2, Rev. 1 CHECKED BY: T. A. Morgan Date: 2/27/02

SUBJECT:

DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test American Electric Power Nuclear Generating Group Donald C. Cook Nuclear Plant Units 1 and 2 RISK IMPACT ASSESSMENT FOR EXTENDING CONTAINMENT TYPE A TEST INTERVAL Analysis File 17141-0007-A2, Rev. 1 February 27,2002 Z " '

Prepared By: Date:

Date: 0(2 7 .

Reviewed By:

SCIENTECH, Inc.

Gaithersburg, Maryland

CLIENT: American Electric Power Corp. BY: E. R. Schmidt PAGE: 2 OF 24 FILE NO. 17141-0007-A2, Rev. 1 CHECKED BY: T. A. Morgan Date: 2/27/02

SUBJECT:

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

6 9.0 MAJOR ASSUMPTIONS 7 10.0 IDENTIFICATION OF COMPUTER CODES 7 11.0 DETAILED ANALYSIS 7 12.0 COMPUTER INPUT AND OUTPUT 16 13.0

SUMMARY

OF RESULTS 16

14.0 CONCLUSION

S 17 APPENDIX A 22 17141-0007-A2-R1 .doc

CLIENT: American Electric Power Corp. BY: E. R. Schmidt PAGE: 3 OF 24 FILE NO. 17141-0007-A2, Rev. 1 CHECKED BY: T. A. MorganI Date: 2/27/02

SUBJECT:

DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test List of Tables Table Ri - Major Results 16 Table R2 - Other Results 17 Table 1 A - Detailed Description for the Eight Accident Classes as defined by EPRI TR-1 04285 18 Table 1 - Mean Containment Frequency Measures for a Given Accident Class 19 Table 2 - Person-Rem Measures for a Given Accident Class 19 Table 3 - Baseline Mean Consequence Measures for a Given Accident Class 20 Table 4 - Mean Consequence Measures for 10 - Year Test Interval for a Given Accident Class 20 Table 5 - Mean Consequence Measures for 15 - Year Test Interval for a Given Accident Class 21 17141-0007-A2-RI .doc

CLIENT: American Electric Power Corp. BY: E. R. Schmidt I PAGE: 4 OF 24 FILE NO. 17141-0007-A2, Rev. 1 CHECKED BY: T. A. Morgan Date: 2/27/02

SUBJECT:

DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test ANALYSIS FILE: 17141-0007-A2, Rev. 1

1.0 CLIENT

American Electric Service Power Corporation -D.C. COOK Nuclear Plant Units 1 and 2

2.0 TITLE

Risk Informed/Risk Impact Assessment for Extending Containment Type A Test Interval

3.0 AUTHOR

Hassan Elrada (Rev. 0)

E. Robert Schmidt (Rev. 1)

4.0 PURPOSE

The purpose of this calculation is to assess the risk impact for extending the D.C.Cook Integrated Leak Rate Test (ILRT) interval from ten to fifteen 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 under Option A "Prescriptive Requirements" or Option B "Performance-Based Requirements". The DCCOOK Nuclear Power Plant units 1 and 2 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 1.001La. D.C.Cook will use this analysis to seek a one time exemption from 1 in 10 years test interval to 1 in 15 years test interval.

Revision 1 of this analysis file is identical to Revision 0 except for editorial and format changes and corrections.

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 (ILRT) from one in ten years to one in fifteen years.

6.0 TECHNICAL APPROACH The methodology used for this analysis is similar to the assessments performed for Crystal River 3 (CR3) [Reference 9] and Indian Point 3 (IP3) [Reference 7] with enhancement outlined in the EPRI Interim Guidance [Reference 10]. The CR3 and IP3 submittals have been approved by the NRC.

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

[Reference 31. This methodology is similar to that presented in EPRI TR104285 [Reference 2] and NUREG-1493 [Reference 5] and incorporates the revised guidance and additional information of References 10 and 11. It uses a simplified bounding analysis approach to evaluate the risk impact on increasing the ILRT Type A interval from 10 to 15 years by examining the D.C.Cook Individual Plant Examination (IPE) [Reference 4] plant specific accident sequences in which the containment integrity remains intact or the containment is impaired. Specifically, the following were considered:

• Core damage sequences in which the containment remains intact initially and in the long term (EPRI TR104285 Class 1 sequences).

17141-0007-A2-Ri .doc

CLIENT: American Electric Power Corp. BY: E. R. Schmidt ' PAGE: 5 OF 24 FILE NO. 17141-0007-A2, Rev. 1 CHECKED BY: T. A. Morgan Date: 2/27/02

SUBJECT:

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

" Core damage sequences in which containment integrity is impaired due to random isolation failures of plant components other than those associated with Type B or Type C test components.

For example, liner breach or steam generator manway leakage. (EPRI TR-1 04285 Class 3 sequences). Type B test measures component leakage across pressure retaining boundries (e.g.

gaskets, expansion bellows and air locks). Type C test measures component leakage rates across containment isolation valves.

"* Core damage sequences in which containment integrity is impaired due to containment isolation failures of pathways left 'opened' following a plant post-maintenance test. (For example, a valve failing 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-1 04285 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 provided in Table 1A of this report [Reference 2].

This calculation uses the following steps:

Step 1 - Quantify the base-line risk in terms of frequency per reactoryear for each of the eight accident classes presented in Table 1.

The D.C.Cook IPE [Reference 4], and NUREG-1493 [Reference 5] were used to provide data to evaluate the annual frequencies for Classes 1,2,3,6,7 and 8. These frequencies are evaluated in detail in Section 11.1 of this analysis. Table 1 summarizes the results of this step. Class 4 and 5 sequences were not quantified because they are not impacted by the Type A test interval and are small contributors to the total. The containment failure modes modeled in the D.C.Cook IPE were based on important phenomena and system related events identified in NUREG 1335 [Reference 6].

Step 2- Develop plant specific person-rem dose (populationdose) per reactoryear for each of the eight accident classes (See Table 2).

Reference 8 was used to assign person-rem to each of the classes described in Table 1A excluding Classes 4 and 5. Reference 8 is a calculation of the conditional person-rem dose to the population, within a 50-mile radius from the D.C.Cook plant. The total population dose in person-rem for each class is evaluated in detail in Section 11.2 of this analysis. Table 2 summarizes the results of this step.

Step 3 - Evaluaterisk impact of extending Type A test interval from 10 to 15 years.

This step evaluates potential increase in the population dose release due to extending the ILRT test interval from 3-in-10-year to 1-in-10 year and to 1-in-15-year. Section 11.3 of this calculation contains the detailed evaluation of this step. Section 13.0 and Tables 3, 4 and 5 summarize the results of this step.

Step 4 - Determine the change in risk in terms of Large Early Release Frequency (LERF)in Accordance with R.G. 1.174 [Reference 3].

This step evaluates the increase in the Large Early Release Frequency (LERF) due to extending the ILRT test interval from 3 in 10 year test interval to 1 in 15 year test interval and from 1 in 10 year to 1 in 15 year test intervals. Section 11.4 of this calculation contains the detailed evaluation of this step while Section 13.0 summarizes the result of this step.

17141-0007-A2-RI.doc

CLIENT: American Electric Power Corp. BY: E. R. Schmidt PAGE: 6 OF 24 FILE NO. 17141-0007-A2, Rev. 1 CHECKED BY: T. A. Morgan Date: 2/27/02

SUBJECT:

DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Step 5 - Determine the change in the ConditionalContainment FailureProbabilityfor the proposedand cumulative changes of Type A test interval.

This step evaluates increase in the Conditional Containment Failure Probability (CCFP) due to extending the ILRT test interval from one test interval to another. CCFP is defined as:

[1 - (Frequency Class1 + Frequency Class3a)/Core Damage Frequency (CDF)]. The changes in CCFP are evaluated in detail in Section 11.5 while Section 13.0 summarizes the results of this step.

7.0 INPUT INFORMATION

1. IPE total Core Damage Frequency (CDF) and frequency of various release categories from D.C.Cook Individual Plant Examination (IPE) [Reference 4] as summarized in Appendix A to this calculation.

2. Dose Rates for the eight classes. Provided by P.J. Fulford, "D.C.Cook Year 2000 Offsite Dose Assessment, Calculation # 17141-0007-Al", dated 11/13/01 [Reference 81.

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

3. Regulatory Guide 1.174, "An Approach for using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis" July 1998.

4. American Electric Power Corp., "Donald C. Cook Nuclear Plants Units 1 & 2 Individual Plant Examination, Revision 1", Transmitted to the NRC by letter from E. E. Fitzpatrick, Indiana Michigan Power Company, to NRC Document Control Desk, dated October 26, 1995.

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

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

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

January 18, 2001.

8. P.J. Fulford, "D.C.Cook Year 2000 Offsite Dose assessment, Calculation # 17141-0007-Al", dated 11/13/01.

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

June 20, 2001.

10. 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.

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

17141-0007-A2-Ri.doc

CLIENT: American Electric Power Corp. BY: E. R. Schmidt I PAGE: 7 OF 24 FILE NO. 17141-0007-A2, Rev. 1 CHECKED BY: T. A. Morgan Date: 2/27/02

SUBJECT:

DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test 9.0 MAJOR ASSUMPTIONS:

1. Containment leaks rates less than twice the allowable leak rate (La) or 2 La indicates an intact containment. This leak rate is considered as "negligible".

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

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

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

5. Because Class 8 sequences are containment bypass sequences (e.g. Steam Generator Tube Rupture - SGTR, Isolation 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.

11.0 DETAILED ANALYSIS:

11.1 Step 1 - Quantify the base-lined risk in terms of 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 2. Except for Class 1 and Class 7, the equations used in this quantification are very similar to those used in the Indian Point Unit 3 (1P3) Calculation [Reference 7].

Class 1 and Class 7 were evaluated based on the Crystal River Unit 3 (CR3) Calculation [Reference 9]

where the term Cl (CI is the sum of the frequencies for Classes 3a, 3b, and 6) is deducted from Class 1 as shown below. In the IP3 Calculation Reference 7], the term Cl was deducted from Class 7.

Class 3 was evaluated based on interim Guidance [Reference 10]. 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 1La. The frequency for these sequences is determined as follows:

Classh1 Frequency = NoContFailureFreq - Cl Where:

No-ContFailureFreq = 4.10E-05/yr [From Appendix A]

CI Class_3a Frequnency + Class 3bjFrequency + Class_6_Frequency

= 1.93E-06/yr + 1.93E-07/yr + 7.147E-08 /yr = 2.194E-06/yr

[These values are obtained from the Class 3 and 6 sequences sections below].

or Class 1 Frequency = 4.1 OE-05/yr - 2.194E-06/yr = 3.88E-05/yr 17141-0007-A2-R I.doc

CLIENT: American Electric Power Corp. BY: E. R. Schmidt PAGE: 8 OF 24 FILE NO. 17141-0007-A2, Rev. 1 CHECKED BY: T. A. Morgan Date: 2/27/02

SUBJECT:

DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Class 2 Sequences. This group consists of all core damage accident progression bins for which a pre-existing leakage due to failure to isolate the containment occurs. These sequences are dominated by failure to close of large, greater than 2 inch diameter, containment isolation valves. The frequency for these sequences is determined as follows:

Class 2 Frequency = 1.051 E-07/yr [From Appendix A]

Class 3 Sequences. This group consists of all core damage accident progression bins for which a pre-existing leakage in the containment structure (i.e. containment liner) exists. The containment leakage for these sequences can be either small (10 La for Class 3a) or large (35 La for Class 3b).

For this analysis, the question on containment analysis was modified to include the probability of a liner breach (due to excessive leakage) at the time of core damage. This class is divided to two classes (Class 3a and Class 3b). Class 3a is defined as small liner breach and Class 3b represents a large containment breach. Evaluation of these two classes is based on EPRI TR 104285 [Reference 2] and the Interim Guidance [Reference 10].

The frequency for this Class event is determined as follows:

Class 3a Frequency = Prob(Class 3a)*CDF Class_3bFrequency = Prob(Class 3b)*CDF Probabilityof Class 3a Event (Small ContainmentBreach) - ProbClass_3a To calculate the probability that a liner leak will be small (Class 3a), use was made of the data presented in NUREG-1493 [Reference 5] and the EPRI Interim Guidance [Reference 101. The NUREG-1493 states that 144 ILRTs were conducted. The data reported that 23 of 144 tests had allowable leak rates in excess of 1.OLa. However, of these 23 'failures,' only 4 were found by an ILRT.

The others were found by Type B and C testing or 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 large leak is calculated as 5/182 = 0.027 [Reference 10].

Therefore the frequency of release due to Class 3a failures is calculated as:

FREQciass 3a = PROBciass 3a x CDF = 0.027 x 7.147E-05/yr = 1.93E-06/yr Probabilityof Class 3b Event (Large ContainmentBreach) - ProbClass_3b To calculate the probability that a liner leak will be large (Class 3b), use was made of the data presented in NUREG-1493 [Reference 5] and new data presented by the EPRI Interim Guidance

[Reference 10]. 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 38 ILRTs and was discussed in EPRI Interim Guidance and this failure was not considered large.

Because no Class 3b failures have occurred in 182 ILRT tests, the EPRI Interim Guidance suggested that the Jeffery's non-informative prior distribution would be appropriate for the Class 3b distribution (The rational for using the Jeffery's non-informative prior distribution was discussed in Reference 10.)

Failure probability = (# of failures (0) + Y2)/(Number of tests (182) + 1)

The number of large failures is zero and the probability is 0.5/183 = 0.0027 17141-0007-A2-R l .doc

CLIENT: American Electric Power Corp. BY: E. R. Schmidt I PAGE: 9 OF 24 FILE NO. 17141-0007-A2, Rev. 1 CHECKED BY: T. A. Morgan Date: 2/27/02

SUBJECT:

DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Therefore the frequency of release due to Class 3b failures is calculated as:

FREQcjass 3 b = PROBcjass 3b x CDF = 0.0027 x 7.147E-05/yr = 1.93E-07/yr It should be noted that the above calculation is very conservative. Not all core damage progression will contribute to Class 3b failure. This point will be further discussed in LERF calculation section.

Class 4 Sequences. This group consists of all core damage accident progression bins for which a failure-to-seal containment isolation due to failure of Type B test components occurs. Because these failures are detected by Type B tests, this group is not evaluated further.

Class 5 Sequences. This group consists of all core damage accident progression bins for which a failure-to-seal containment isolation due to failure of Type C test components occurs. Because these failures are detected by Type C tests, this group is not evaluated further.

Class 6 Sequences. This group is similar to Class 2 and addresses additional failure modes not typically modeled in PRAs due to the low probability of occurrence. These are sequences that involve core damage accident progression bins for which a failure-to-seal containment leakage due to failure to isolate the containment occurs. These sequences are dominated by misalignment of containment isolation valves following a test/maintenance evolution. A conservative screening value of 1.OE-03 will be used to evaluate this class.

The low failure probabilities are based on the need for multiple failures, the presence of automatic closure signals, and control room indication. Based on the purpose of this calculation, and the fact that this failure class is not impacted by Type A testing, no further evaluation is needed. This is consistent with the EPRI guidance. However, in order to maintain consistency with the previously approved methodology (i.e.PROBClass6 > 0), a conservative screening value of 1.OE-03 will be used to evaluate this class.

The annual frequency for these sequences is determined as follows:

Class 6jFrequency = (Screening Value) *CDF Where:

Screening Value = 1.0 x 10-3 [Assumed Conservative Value]

And CDF = 7.147-05/yr Class 6 Frequency = 1.0E-03

• 7.147E-05/yr = 7.147E-08 /yr 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). For this analysis the associated maximum containment leakage for this group is 35La.

The annual frequency for these sequences is determined as follows:

Classj7jFrequency =CDF - (No-ContFailureFreq + Class_8_Frequency + Class_2_Frequency)

Where:

CDF = 7.147E-5/yr [From Appendix A]

No-ContFailureFreq =4.1 E-05/yr [From Appendix A]

Class 8 Frequency = 9.689E-06/yr [From Appendix A]

Class 2 Frequency = 1.051 E-07/yr [From Appendix A]

Classj7 Frequency = 7.147E-05/yr - (4.1 E-05/yr + 9.689E-06/yr + 1.051 E-07/yr) = 2.068E-05/yr 17141-0007-A2-R I .doc

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DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Class 8 Sequences. This group consists of all core damage accident progression bins in which containment bypass occurs. From above, the failure frequency for this class is:

Class 8jFrequency = 9.689E-06/yr. [From Appendix A]

Note for this class the maximum release is not based on normal containment leakage, because most of the releases are released 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 1.

11.2 Step 2 - Develop plant specific person-rem dose (population dose) per reactor year for each of the eight accident classes and quantify baseline risk In accordance with guidance given by Reference 2, This step develops D.C.Cook population dose and evaluates the baseline risk impact for the eight accident classes defined in the previous sections of this calculation.

2a) Characterize accident scenarios into major groups (eight classes).

(See Class one through eight sequences above) 2b) Develop plant specific person-rem dose (population dose) per reactor year.

Reference 8 documents an updated assessment of the D. C. Cook Power Plant off-site population dose consequences due to the accidental release of radiological materials resulting from several severe accident scenarios. This assessment utilizes a year 2000 population estimate and is an update in this respect of a previous Level III PRA consequence analysis prepared for the plant IPE preparation for a year 1980 region population distribution. The results are for a 50-mile radius region surrounding the plant.

This calculation uses the 200% of 1980 increase data values from Table 2 of Reference 8 for sequences SB01 81$ = 1.01+03 for Classes 1, 3a and 3b, LL08** = 3.84E+06 for Classes 2, 6 and 7 and SGR50** = 9.68E+6 for Class 8. These sequences and their source terms were those selected for the IPE Level III analysis and are defined in Reference 4. Sequence SBO181$ is a station blackout sequence with no containment failure. Sequence LL08** is a large LOCA sequence with early containment failure at the basemat. Sequence SGR50** is a steam generator containment bypass sequence with containment failure at the basemat. The resulting conditional population doses given the release are Class 1 = (1.01 E+03)

• 1.01La = 1.01 E+03 person-rem Class 2 = (3.84E+06) person-rem Class 3a = (1.01 E+03)

• 1OLa = 1.01 E+04 person- rem Class 3b = (1.01 E+03)

• 35La = 3.535E+04 person-rem Class 4 = Not Analyzed Class 5 = Not Analyzed Class 6 = (3.84E+06) person-rem Class 7 = (3.84E+06) person-rem Class 8 sequences involve containment bypass failures; as a result, the person-rem dose is not based on normal containment leakage. The releases for this class are expected to be released directly to the environment. Based on Table 2 of Reference 8, the value used is 9.68E+06 person-rem.

The above values are summarized in Table 2.

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DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test 2c) Calculate and Review Baseline Risk for Each Accident Class The baseline risk for each accident class is presented in Table 3. The baseline risk is defined as the product of the containment failure mode frequency and the conditional population dose. Table 3 is product of Tables 1 and 2. The ILRT baseline risk is based on the test interval of 3 in 10 years or about 1 in 3 years.

As mentioned in the method section of this calculation, only Classes 3a and 3b are impacted by the Type A ILRT test. Therefore, the percent risk contribution (%BaseRisk) for these classes is:

%BaseRisk = [( Class3aBase + Class3bBase) / Total-base)]

• 100 Where:

Class3aBase = 1.95 E-02 person-rem/year Class3bbase = 6.82 E-03 person-rem/year Class_3_BaseTotal = 1.95E-02 + 6.82E-03 = 0.0263 person-rem/yr Totalbase = 1.7393E+02 person-rem/year

%BaseRisk = [(1.95E-02 + 6.82E-03 ) / 1.7393E+02]

• 100

%BaseRisk = 0.015%

Therefore, the total baseline risk contribution of leakage, potentially impacted by ILRT test interval, representedby Class 3 accident scenariosis 0.0263 person-rem/yearor 0.015% of the total population exposure risk.

11.3 Step 3 - Evaluate risk impact of extending Type A test interval from 10 to 15 years.

Risk impact due to 10-year test interval According to NUREG-1493 [Reference 7], extending the Type A ILRT interval from 3 in 10 years to 1 in 10 years will increase the average time that a leak detectable only by an ILRT goes undetected from 18 to 60 months. The average time for undetection is calculated by multiplying the test interval by 0.5 and multiplying by 12 to convert from "years" to "months." The recent EPRI Guidance suggested use the factor of 3.33 (60/18) to estimate the increase of Class 3. Since, Type A tests impact only Class 3 sequences. This is very conservative for estimating the Class 3b and will be used here for population dose calculation only. Note that Class 3b will be used later in section 11.4 to estimate the change in the Large Early Release Frequency (LERF). In section 11.4 conservatism will be removed from the frequency of Class 3b. Also, as with the baseline case, the frequency of Class 1 has been reduced by the frequencies of Classes 3a, 3b, and Class 6 in order to preserve total CDF.

The results of this calculation are presented in Table 4.

Based on the above values, the Type A 10-year test frequency percent risk contribution (%Risk_10) for Class 3 is as follows:

%Risk1 0 = [(Class3a-l 0 +Class3b1 0) / Total-1 01

• 100 Where:

Class3a_10 = 6.49E-02 person-rem/year 17141-0007-A2-Rl .doc

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DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Class3b_10 = 2.27E-02 person-rem/year Class3 10_total = 6.49E-02 + 2.27-02 = 0.0876 person-rem/year Total_10 = 1.7398E+02 person-rem/year

%Risk_10 = [(6.49E-02 + 2.27E-02 ) / 1.7398E+02 ]

• 100

%Risk_10 = 0.05%

Therefore, the total risk contributionof leakagefor Type A 10-Year ILRT intervalrepresented by Class 3 accidentscenarios is 0.0876 person-rem/yearor 0.05% of the totalpopulation risk.

Since the only change in risk is due to the change in Class 3, the percent risk increase due to extending the ILRT interval from 3 in 10 years (baseline case) to 1 in 10 years is evaluated as follows:

[(Total_10 - Total-base) / Totalbasel

• 100 =

[(Class3_10_total - Class_3_BaseTotal) / Total-base]

• 100 Where; Class_3_BaseTotal = 0.0263 person-rem/yr [From above]

Class3_10_total 0.0876 person-rem/year [From above]

Totalbase = 1.7393E+02 person-rem/year [From Table 3]

[(Class3 10 total - Class3_Basetotal) / Total-base]

• 100

= [(0.0876 - 0.0263 ) / 1.7393E+02 ] - 100 = (0.0613/1.7393E+02)

• 100 = 0.035 %

Therefore, The total risk increase due to extending the ILRT interval from 3 in 10 years (baseline case) to 1 in 10 years is 0.0613 person-rem/yearor 0.035% of the total population risk.

Risk Impact due to 15-year test interval The risk contribution for a 15-year interval is similar to the 10-year interval. The difference is in the increase in probability of leakage value. If the test interval is extended to 1 in 15 years, the mean time that a leak detectable only by an ILRT test goes undetected increases to 90 months (0.5

• 15

• 12).

Reference 11 suggested to use a factor of 5 (90/18) to account for the increased likelihood of fail to detect, which will be implemented here. As discussed previously, the factor of 5 is very conservative (especially for evaluating Class 3b for LERF). As with the baseline case, the PSA frequency of Class 1 has been reduced by the frequency of Class 3a, 3b, and Class 6 in order to preserve total CDF.The results for this calculation are presented in Table 5.

Based on the above values, the Type A 15-year test interval percent risk contribution (%Risk -15) for Class 3 is as follows:

%Risk_15 = [(Class3a-l 5 +Class3b_1 5) / Total1 5]

• 100 Where:

Class3a_15 = 9.74E-02 person-rem/year Class3b_15 = 3.41 E-02 person-rem/year Class3_15_total = 9.74E-02 + 3.41 E-02 = 0.1315 person-rem/year 17141-0007-A2-R1 .doc

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DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Total_15 = 1.7403E+02 person-rem/year

%Risk_15 = [(9.74E-02 + 3.41E-02 ) / 1.7403E+02]

• 100 [From Table 5]

%Risk_1 5 =0.076%

Therefore, the total risk contribution of leakage for Type A 15-year ILRT interval representedby Class 3 accidentscenarios is 0.1315 person-rem/yearor 0.076% of the total population risk.

The percent risk increase due to extending the ILRT interval from 3 in 10 years (baseline case) to 1 in 15 years is evaluated as follows:

[(Total_1 5 - Total-base) / Total-base]

• 100 =

[(Class3_15_total - Class_3_Base-Total) / Total-base]

• 100 Where; Class3_15 total = 0.1315 person-rem/year [From above]

Class_3_BaseTotal = 0.0263 person-rem/yr [From above]

Totalbase = 1.7393E+02 person-rem/year [From Table 3]

[(Class3 15 total - Class_3 BaseTotal) / Total-base]

• 100= [(0.1315 - 0.0263 )/ 1.7393E+02 ]

• 100 = (0.105/1.7393E+02)

• 100 = 0.06%

Therefore, the total risk increasedue to extending the ILRT interval from 3 in 10 years (baseline case) to 1 in 15 years is 0. 105 person-rem/yearor 0.06% of the total baseline populationrisk.

The percent risk increase in terms of person-rem/year from 1 in 10 years to 1 in 15 years test interval for Classes 3a and 3b is:

% Risk (10-1 5PR) =[(Class3 15_total) - (Class_3_10_Total) / (Class_3_10_Total)]*100 Where; Class3 15 total =0.1315 person-rem/year [From above]

Class_3_10_Total = 0.0876 person-rem/yr [From above]

% Risk (10-15PR) = [(0.1315 - 0.0876 ) /0.0876 ]

• 100 = 50.1%

The increase in person-rem/year for all accident classes from 1 in 10 years to 1 in 15 years test interval is:

[(person-rem(Class3)_15)- (person-rem(Class3)_10)] = 0.1315 - 0.0876 =0.0439 person rem The percent risk increase due to extending the ILRT interval from 1 in 10 years to 1 in 15 years is evaluated as follows:

[(Class3 15 total - Class_3_10_Total) / Total1 0]

• 100 Where; Class3_15_total = 0.1315 person-rem/year [From above]

Class_3 10_Total = 0.0876 person-rem/yr [From above]

Total_10 = 1.7398E+02 person-rem/year [From Table 4]

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

[(Class3 15 total - Class 3 10 Total) / Total_10]

• 100= [(0.1315 - 0.0876 )/ 1.7398E+02] 100 =

(0.0439 /1.7398E+02)

• 100 = 0.025%

Therefore, the total risk increasedue to extending the ILRT interval from 1 in 10 years (baselinecase) to I in 15 years is 0.0439 person-rem/yearor 0.025% of the total baseline population risk.

11.4 Step 4 - Determine the change in risk in terms of Large Early Release FreQuency (LERF)

This step evaluates the increase in the Large Early Release Frequency (LERF) due to extending the ILRT test interval from 3 in 10 years test interval to 1 in 15 years test interval and from 1 in 10 years to 1 in 15 years test intervals.

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 3b sequences, which have the potential to result in large releases if pre-existing leak were present, are impacted by the ILRT Type A test.

The previous methodology [References 7 and 9] employed for determining LERF (Class 3b frequency) involved multiplying the total CDF by the failure probability for this class (3b) of accident. This was done for simplicity and is conservative. However, some plant-specific accident classes leading to core damage are likely to include individual sequences that either may already (independently) cause a LERF or could never cause a LERF. For instance, the CR3 [Reference 9] evaluation assumption number 7 states that 'The containment releases for Classes 2, 6, 7, and 8 are not impacted by the ILRT Type A test frequency. These classes already include containment failure with release consequences equal or greater than those impacted by Type A."

In addition to the references mentioned above, improvements suggested in the EPRI Interim Guidance

[Reference 10] and in Reference 11 and its enclosures are implemented in the present evaluation.

Thus conservatism was removed from Class 3b in the evaluation of LERF by using the following equation:

Frequency 3b=(3b Failure probability)*(PCDFTypeA)

Where 3b Failure Probability =0.0027 [See Section 11.1]

and PCDFTypeA = portion of CDF that may be impacted by Type A leakage and contribute to LERF

= Total CDF - CDF of sequences that go to LERF irrespective of Type A Leakage

- CDF of sequences that cannot cause a LERF

= Total CDF - Sum of CDF for Classes 2, 6, 8 and early severe accident failures

- CDF no containment failure with containment spray before core damage 17141-0007-A2-R i .doc

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DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Where:

CDF for Classes 2, 6, and 8 from Section 11.1.

CDF for early containment failure due to severe accident phenomena = 1.955E-05 [See Appendix A]

CDF no containment failure with containment spray before core damage = (3.449E-05 [from Appendix A]) - Class 6

= 3.449E 7.147E-08 = 3.442E-05 In the latter, the Class 6 CDF is conservatively subtracted from the IPE based number. This class is not specifically included in the IPE evaluation and to maintain the correct total CDF, its value must be subtracted from the total no containment failure frequency. Similarly for Class 3a, while it cannot lead to a large early release (and therefore removed from PCDFTypeA), it also would need to be removed from the total no containment failure frequency. The net effect is that it would be subtracted and then added back in with no net impact on the results.

Therefore:

PCDFType A = 7.147E (1.051 E-07 + 7.147E-08 + 9.689E-06 + 1.955 E-05)

- 3.442E-05 = 7.63E-06 per year The Baseline LERF affected by ILRT = 7.63E-06

• 0.0027 = 2.06E-08 per year The 1 in 10 years LERF affected by ILRT = 2.06E-08

• 3.33 = 6.86E-08 per year The 1 in 15 years LERF affected by ILRT = 2.06E-08

• 5 = 1.03E-07 per year Change in LERF going from 3 in 10 years test interval to 1 in 15 years test interval =

1.03E 2.06E-08 = 8.2E-O8lyear Change in LERF going from 1 in 10 years test interval to 1 in 15 years test interval =

1.03E 6.86E-08 = 3.4E-O8year 11.5 Step 5 - Determine the change in the Conditional Containment Failure Probability (CCFP) for the proposed and cumulative changes of Type A test interval The change in Conditional Containment Failure Probability (CCFP) for the proposed and cumulative changes is estimated as follows:

1. Estimate the CCFP for each test interval (i.e .3 tests in ten years, 1 test in ten years, and 1 test in fifteen years)

2. Calculate the change in COEP between the test intervals.

The Conditional Containment Failure Probability (CCFP) can be defined as:

[1 - (Frequency Class 1 + Frequency Class 3a)/CDF]

Where Frequency Class1 = Frequency per year of No Containment Failure.

Frequency Class3a = Frequency per year of Small Isolation Failure.

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DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Using the above equation and the data from Table 1(i.e. Class 1 frequency is 3.881 E-05 , the Class 3a frequency is 1.93E-06 and CDF = 7.147E-05),

the CCFP for 3 tests in ten years =

1 - [(3.881 E-05 + 1.93E-06 )/7.147E-05] = 4.30E-01 Using the above equation and the data from Table 4 (i.e. Class 1 frequency is 3.386E-05 ,the Class 3a frequency is 6.426E-06 and CDF=7.147E-05),

the CCFP for 1 test in ten years =

1-[ (3.386E-05 +6.426E-06 )/ 7.147E-051 = 4.363E-01 Using the above equation and the data from Table 5 (i.e. Class 1 frequency is 3.032E-05, the Class 3a frequency is 9.648E-06 and CDF = 7.147E-05),

the CCFP for 1 test in fifteen years =

1-[(3.032E-05 + 9.648E-05 )/7.147E-05] = 4.41 E-01 The change in CCFPgoing from 3 in 10 years test interval to 1 in 15 years interval

=4.41E-01 -4.30E-01 = 0.011 or 1.1%

The change in CCFPgoing from 1 in 10 years test intervalto 1 in 15 years interval

= 4.41E 4.363E-01 = 0.0047 or 0.47 %

12.0 COMPUTER INPUT OUTPUT NONE 13.0

SUMMARY

OF RESULTS Table R1 below summarizes the major results.

Table RI- Major Results Test Interval Extended From 3 in 10 years to From 1 in 10 years 1 in 15 years to 1 in 15 years Total person-rem/year increase (See Section 11.3) 0.105 0.0439 The percentage increase person-rem/year risk (See Section 11.3) 0.06% 0.025%

Change in LERF (See Section 11.4) 8.2E-08 3.4E-08 Change in the Conditional Containment Failure Probability (See Section 11.5) 1.1% 0.47%

Other results are shown in the Table R2 below. It shows (for example), The change in Type A test frequency from once per ten years to once per fifteen years increases the total integrated plant risk for those accident sequences influenced by Type A testing is only 0.0434 (i.e. 0.1315 - 0.0876 = 0.0439 person-rem/year .

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DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Table R2 - Other Results Class Risk Impact Baseline 3 in 10 years 1 in 10 years 1 in 15 years 3a and 3b. These 0.015% of integrated 0.05 % of integrated 0.076% of integrated classes are impacted by value based on 1OLa for value based on l0La for value based on 10La for Type A test Class 3a and 35La for Class 3a and 35La for Class 3a and 35La for Class 3b, which is Class 3b, which is Class 3b, which is equivalent to: equivalent to: equivalent to:

0.0263 person-rem/year 0.0876 person-rem/year 0.1315 person-rem/year Total Integrated Risk 173.93 person-rem/year 173.98 person-rem/year 174.03person-rerm/year The sensitivity of the above results to changes in the Level III results used in this evaluation have been considered. As indicated in Reference 8 the consequences for LOCA and Steam Generator Tube rupture sequences are higher for sensitivity study cases run with containment failure in the upper compartment rather than the base case assumption of failure at the basemat. If the higher numbers were used in the present analysis, the absolute value of the change in person-rem/year would remain the same while the percentage contribution to the total person-rem/year would decrease.

Also, as indicated in Reference 8 there is some uncertainty as to the accuracy of the original IPE population distribution. Two-hundred percent of the 1980 results were judged to be a conservative bounding estimate for the year 2000. If, however, the true consequences are higher than this, the total risk, as well as the incremental risk for the extended ILRT interval, will increase by the same proportion while the percentage contribution and percentage increase due to the requested change will remain the same.

14.0 CONCLUSION

S:

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

The change in Type A test frequency from once per ten years to once per fifteen years increases the total integrated plant risk for those accident sequences influenced by Type A testing by only 0.0439 person-rem/year. This increase in person-rem/year is negligible when compared to other accident risks.

Reg.Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Very small changes in risk are defined in Reg. Guide 1.174 as increases of CDF below 1.OE-06/yr or increases in LERF of less than 1E-07/yr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from once per 10 years to once per 15 years is 3.4E-08/yr. Since guidance in Reg. Guide 1.174 defines very small changes in LERF as below 1.OE-7/yr, increasing the ILRT interval from 10 to 15 years is therefore considered very small and non-risk significant.

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DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval Table 1 A- Detailed Description for the Eight Accident Classes as defined by EPRI TR-1 04285 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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DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE 1 - Mean Containment Frequency Measures for a Given Accident Class Class Description Frequency/yr.

1 No Containment Failure 3.881E-05

_2 Large Containment Isolation Failure (Failure-To-Close) 1.051 E-07 3a Small Isolation Failures (Liner Breach) 1.930E-06 3b Large Isolation Failures (Liner Breach) 1.930E-07 4 Small Isolation Failure - Failure-To-Seal (Type B test) NA 5 Small Isolation Failure - Failure-To-Seal (Type C Test) NA 6 Containment isolation Failures (Dependent failures, Personnel Errors) 7.147E-08 7 Severe Accident Phenomena Induced Failure (Early and Late Failures) 2.068E-05 8 Containment Bypassed (SGTR) 9.689E-06 ore Damaae All Containment Event Tree (CET) Endstates 7.147E-05 TABLE 2 - Conditional Person-Rem Measures for a Given Accident Class Class Description Person-Rem (50-miles) 1 No Containment Failure 1.01 E+03 2 -Large Containment Isolation Failure (Failure-To-Close) 3.84E+06 3a Small Isolation Failures (Liner Breach) 1.01 E+04 3b Large Isolation Failures (Liner Breach) 3.54E+04 4 _ Small Isolation Failure - Failure-To-Seal (Type B test) N/A 5 Small Isolation Failure - Failure-To-Seal (Type C Test) N/A 6 Containment isolation Failures (Dependent failures, Personnel Errors) 3.84E+06 7 Severe Accident Phenomena Induced Failure (Early and Late Failures) 3.84E+0 8 __ ontainment Bypassed (SGTR) 9.68E+06 17141-0007-A2-Ri.doc

CLIENT: American Electric Power Corp. BY: E. R. Schmidt IPAGE: 20 OF 24 FILE NO. CHECKED BY: T. A. Morgan Date: 2/27/02

SUBJECT:

DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE 3 - Baseline Mean Consequence Measures for a Given Accident Class Person-Rem Person-Rem/yr Class Description ]Frequency/yr (50-miles) 50-miles) 1 No Containment Failure .881E-05 1.01OE+03 3.919E-02 2 _ Large Containment Isolation Failure (Failure-To-Close) 1.051 E-07 3.840E+06 4.036E-01 3a Small Isolation Failures (Liner Breach) 1.930E-06 1.01 0E+04 1.949E-02 3b Large Isolation Failures (Liner Breach) 1.930E-07 3.535E+04 6.821 E-03 4 Small Isolation Failure - Failure-To-Seal (Type B test) NA N/A NA 5 Small Isolation Failure - Failure-To-Seal (Type C Test) NA N/A NA 6 Containment isolation Failures (Dependent failures, Personnel Errors) 7.147E-08 3.840E+06 .744E-01 Severe Accident Phenomena Induced Failure (Early and Late 7 _.068E-05 Failures) 3.840E+06 .940E+01 8 _ ontainment Bypassed (SGTR) 0.689E-06 9.680E+06 9.379E+01 11CET End states .147E-05 .7393E+02 TABLE 4 Mean Consequence Measures for 10 - Year Test Interval for a Given Accident Class Person Person-Rein Rem/yr Class Description Frequency/yr (50-miles) (50-miles) 1 No Containment Failure 3.386E-05 1.01 OE+03 3.420E-02 2 -Large Containment Isolation Failure (Failure-To-Close) 1.051 E-07 .840E+06 4.036E-01 3a Small Isolation Failures (Liner Breach) 6.426E-06 1.01OE+04 6.490E-02 3b Large Isolation Failures (Liner Breach) 6.426E-07 3.535E+04 2.272E-02 4 _ Small Isolation Failure - Failure-To-Seal (Type B test) NA N/A N/A 5 Small Isolation Failure - Failure-To-Seal (Type C Test) NA N/A N/A Containment isolation Failures (Dependent failures, Personnel Errors) 7.147E-08 3.840E+06 2.744E-01 Severe Accident Phenomena Induced Failure (Early and Late 7 Failures) .068E-05 3.840E+06 7.940E+01 8 Containment Bypassed (SGTR) 9.689E-06 9.680E+06 9.379E+01 CDF All CET Endstates .147E-05 1.7398E+02 17141-0007-A2-R 1.doc

CLIENT: American Electric Power Corp. BY: E. R. Schmidt I PAGE: 21 OF 24 FILE NO.I CHECKED BY: T. A. Morgan Date: 2/27/02

SUBJECT:

DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Interval TABLE 5 - Mean Consequence Measures for 15 - Year Test Interval for a Given Accident Class Person-Rem Person-Rem/yr Class Description Frequency/yr 50-miles) (50-miles) 1 No Containment Failure .032E-05 1.010E+03 3.062E-02 2 Large Containment Isolation Failure (Failure-To-Close) 1.051 E-07 3.840E+06 4.036E-01 3a 3mall Isolation Failures (Liner Breach) .648E-06 1.01 0E+04 9.745E-02 3b Large Isolation Failures (Liner Breach) .648E-07 3.535E+04 3.411 E-02 4 3mall Isolation Failure - Failure-To-Seal (Type B test) NA N/A N/A 5 Small Isolation Failure - Failure-To-Seal (Type C Test) NA N/A N/A 6 Containment isolation Failures (Dependent failures, Personnel Errors) 7.147E-08 3.840E+06 2.744E-01 7 Severe Accident Phenomena Induced Failure (Early and Late Failures) 2.068E-05 3.840E+06 7.940E+01 ontainment Bypassed (SGTR)

C__ 9.689E-06 9.680E+06 P.379E+01 CDF AII CET End States 7.147E-05 1.7403E+02 17141-0007-A2-R l.doc

CLIENT: American Electric Power Corp. I BY: E. R. Schmidt I PAGE: 22 OF 24 FILE NO. 17141-0007-A2 CHECKED BY: T. A. Morgan Date: 2/27/02

SUBJECT:

DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test I APPENDIX A CORE DAMAGE FREQUENCY FROM THE D. C. COOK IPE The core damage frequency associated with the various containment failure modes used to determine the frequency of the EPRI accident classes was obtained from the listing of dominant sequences and their end states provided in Table 4.6-1 of Reference 4. This information is included in the following Table A- 1. The last letter of the end state is the release category and definitions of these codes are given in Table 4.6-4 of Reference 4. The release codes that contribute to the EPRI accident classes are as follows:

Release Category Definition EPRI Class A, S No Containment Failure (NCF) 1*

E, F, G, U Containment Isolation Failures 2 H, I, J, V Early Containment Failure due to Severe 7*

Accident Phenomena B, C, D, V Containment Bypasses 8

• EPRI Class frequency is different from these values due to the adjustments discussed in the body of the calculation.

Also of interest are the no containment failure sequences with successful containment spray at the time of core damage since they will not result in LERF even if excess containment leakage occurs. These sequences are those with end state codes that end in either A or S (no containment failure) with the first part of the end state code beginning with an A or an S (a LOCA) and containing HR or LR (spray injection and recirculation). (From Figure 4.4-1 of Reference 4)

Table A-I lists the dominant sequences and their frequencies and then separately identifies the frequencies of the sequences included in the above classes.

The CDF for each of the above classes:

No Containment Failure = 4.1OOE-05 per year Isolation Failure = 1.051E-07 per year Early Containment Failure due to Severe Accident Phenomena = 1.955E-05 per year Containment Bypasses = 9.689E-06 per year Total CDF = 7.147E-05 per year No Containment Failure with Containment Spray = 3.449E-05 per year.

17141-0007-A2-RI .doc

CLIENT: American Electric Power Corp. BY: E. R. Schmidt PAGE: 23 OF 24 IFILE NO. 17141-0007-A2 CHECKED BY: T. A. Morgan Date: 2/27/02

SUBJECT:

DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test TABLE A-i Seq. No. INITIATOR* END STATE* FREQ. CDF for no CDF for CDF for CDF for CDF for containment failure, Release Release Release Release containment spray Code A or S Codes E, F, Codes H, 1,J, Codes B3,C, operating at time of G, or U or V D, orT core damage 1 SLO SHR-S 2.77E-05 2.77E-05 2.77E-05 0.OOE+01 0.OOE+01 0.OOE+01 2 SGR GHC-C 8.77E-06 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 8.77E-06 3 SLO SHC-J 8.19E-06 0.OOE+01 0.OOE+01 0.OOE+01 8.19E-06 0.OOE+01 4 SLO SH-J 3.97E-06 0.OOE+01 0.OOE+01 0.OOE+01 3.97E-06 0.OOE+01 5 MILO SHR-S 2.67E-06 2.67E-06 2.67E-06 0.OOE+01 0.OOE+01 0.OOE+01 6 ATWV SHR-S 2.48E-06 2.48E-06 2.48E-06 0.OOE+01 0.OOE+01 0.OOE+01 7 ATVW THR-S 2.46E-06 2.46E-06 2.46E-06 0.OOE+01 0.OOE+01 0.OOE+01 8 ESW SLC-J 2.38E-06 0.OOE+01 0.OOE+01 0.OOE+01 2.38E-06 0.OOE+01 9 CCW THR-S 1.75E-06 0.OOE+01 1.75E-06 0.OOE+01 0.OOE+01 0.OOE+01 10 MILO SHC-J 1.11 E-06 0.OOE+01 0.OO11+01 0.OOE+01 1.11 E-06 0.OOE+01 11 CCW SIR-S 1.10E-06 1.10E-06 1.1OE-06 0.OOE+01 0.OOE+01 0.OOE+01 12 SBO THRI-J 8.99E-07 0.OOE+01 0.OOE+01 0.OOE+01 8.99E-07 0.OOE+01 13 LLO ALC-J 8.11 E-07 0.OOE+01 0.OOE+01 0.OOE+01 8. 11 E-07 0.OOE+01 14 SBO THR-S 7.58E-07 0.OOE+01 7.58E-07 0.OOE+01 0.OOE+01 0.OOE+01 15 SBO THWIF-M 6.57E-07 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 16 VDC THR-S 6.1 7E-07 0.OOE+01 6.17E-07 0.OOE+01 0.OOE+01 0.OOE+01 17 MILO SH--J 5.25E-07 0.OOE+01 0.OOE+01 0.OOE+01 5.25E-07 0.OOE+01 18 SLB THR-S 3.94E-07 0.OOE+01 3.94E-07 0.OOE+01 0.OOE+01 0.OOE+01 19 ESW THC-J 3.64E-07 0.OOE+01 0.OOE+01 0.OOE+01 3.64E-07 0.OOE+01 20 SGR GHC-T 3.49E-07 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 3.49E-07 21 LLO ALRI-J 3.43E-07 0.OOE+01 0.OOE+01 0.OOE+01 3.43E-07 0.OOE+01 22 VEF ALRI-J 3.OOE-07 0.OOE+01 0.OOE+01 0.OOE+01 3.OOE-07 0.OOE+01 23 LLO AIR-S 2.92E-07 2.92E-07 2.92E-07 0.OOE+01 0.OOE+01 0.OOE+011 24 ISL V-T 2.64E-07 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 2.64E-07 25 CCW THR-S 2.57E-07 0.00E+01 2.57E-07 0.OOE+01 0.OOE+011 0.OOE+01 26 MILO SLR-S 2.45E-07 2.45E-07 2.45E-07 0.OOE+01 0.OOE+01 0.OOE+01 27 SGR GHR-T 2.12E-07 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+011 2.12E-07 28 LLO AL 1.73E-07 0.OOE+01 0.00E+01 0.OOE+01 1.73E-07 0.OOE+01 29 COW THC-J 1.48E-07 0.OOE+011 0.OOE+01 0.OOE+01 1.48E-07 0.OOE+01 30 SLB THC-J 1.30E-07 0.OOE+01 0.OOE+01 0.OOE+01 1.30E-07 0.OOE+01 31 TRS THWIF-M 1.13E-07 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 32 SLO SHWIF-M 9.91 E-08 O.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 33 TRS THR-S 9.33E-08 0.OOE+01 9.33E-08 0.OOE+01 0.OOE+01 0.OOE+01 34 SLB THW-G 8.95E-08 0.OOE+01 0.0OE+01 8.95E-08 0.OOE+01 0.OOE+01 35 TRS THC-J 8.46E-08 0.OOE+01 0.OOE+01 0.OOE+01 8.46E-08 0.OOE+01 36 SLB THR-S 7.75E-08 0.OOE+01 7.75E-08 0.OOE+01 0.OOE+01 0.OOE+01 37 SBO THWI-M 7.02E-08 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 38 LSP THWIF-M 6.60E-08 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 39 TRA THR-S 6.01 E-08 0.OOE+011 6.01 E-08 0.OOE+01 0.OOE+01 0.OOE+01 40 SGR GHWIF-T 5.98E-08 0.OOE+01 0.OOE+01 0.OOE+011 0.OOE+01 5.98E-08 41 TRA THC-J 5.90E-08 0.OOE+011 0.OOE+01 0.OOE+01 5.90E-08 0.OOE+01 42 TRA THWIF-M 5.58E-08 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 43 VDC THWIF-M 5.08E-08 0.OOE+01 0.0OE+011 0.OOE+01 0.OOE+01 0.OOE+01 44 LSP THR-S 4.18E-08 0.OOE+01 4.18E-08 0.OOE+01 0.OOE+011 0.OOE+01 45 LSP THC-J 3.96E-08 0.OOE+01 0.OOE+01 0.OOE+011 3.96E-08 0.OOE+01 46 SGR GHR-C 2.24E-08 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 2.24E-08 47 MVLO SLC-J 1 .28E-08 0.OOE+01 0.OOE+01 0.OOE+01 1.28E-08 0.OOE+01 48 MILO SLIF-J 9.82E-09 0.OOE+01 0.OOE+01 0.OOE+01 9.82E-09 O.OOE+01 49 SLO SHR'-E 8.03E-09 0.OOE+01 0.OOE+01 8.03E-09 0.OOE+01 0.OOE+01 17141 -0007-A2-R L doc

CLIENT: American Electric Power Corp. BY: E. R. Schmidt PAGE: 24 OF 24 FILE NO. 17141-0007-A2 CHECKED BY: T. A. Morgan Date: 2/27/02

SUBJECT:

DCCOOK Risk-Informed / Risk impact Assessment for Extending Containment Type A Test Seq. No. INITIATOR* END STATE* FREQ. CDF for no CDF for CDF for CDF for CDF for containment failure, Release Release Release Release containment spray Code A or S Codes E, F, Codes H, I, J, Codes B, C, operating at time of G, or U or V D, or T core damage 50 SGR GHW-C 5.24E-09 0.00E+01 0.OOE+01 0.00E+01 0.00E+01 5.24E-09 51 LLO ALWIF-M 3.92E-09 O.00E+01 0.OOE+01 0.OOE+01 O.OOE+01 0.00E+01 52 SBO THC-J 3.51 E-09 O.00E+01 0.00E+01 0.OOE+01 3.51E-09 0.00E+01 53 SGR GHWIF-C 3.48E-09 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 3.48E-09 54 MLO SHWIF-M 3.29E-09 0.00E+01 0.OOE+01 0.OOE+01 0.00E+01 0.OOE+01 55 SGR GHC-T 2.97E-09 0.00E+01 0.OOE+01 0.OOE+01 0.OOE+01 2.97E-09 56 ESW SLR-S 2.45E-09 2.45E-09 2.45E-09 0.OOE+01 0.00E+01 0.OOE+01 57 SLO SHC'-U 2.38E-09 0.OOE+01 0.00E+01 2.38E-09 0.OOE+01 0.OOE+01 58 TRS THW-M 1.54E-09 0.OOE+01 0.OOE+01 O.OOE+01 0.OOE+01 0.OOE+01 59 SLO SHW-M 1.31E-09 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 60 MLO SHR'-E 7.74E-10 0.00E+01 0.OOE+01 7.74E-10 0.OOE+01 0.OOE+01 61 ATW SHR'-E 7.19E-10 0.OOE+01 0.OOE+01 7.19E-10 0.OOE+01 0.OOE+01 62 ATW THR'-E 7.13E-10 0.OOE+01 0.OOE+01 7.13E-10 0.OOE+01 0.OOE+01 63 ESW SLC'-U 6.90E-10 0.00E+01 0.OOE+01 6.90E-10 0.00E+01 0.OOE+01 64 LSP THW-M 5.29E-10 0.00E+01 0.00E+01 0.OOE+01 0.00E+01 0.OOE+01 65 ATW THWIF-M 5.21E-10 0.00E+01 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 66 CCW THR'-E 5.08E-10 0.00E+01 0.OOE+01 5.08E-10 0.OOE+01 0.OOE+01 67 ESW THC-J 4.62E-10 0.OOE+01 0.OOE+01 0.OOE+01 4.62E-10 0.00E+01 68 SGR GHC-C 4.29E-10 0.OOE+01 0.OOE+01 0.OOE+01 0.OOE+01 4.29E-10 69 ATW THC-J 4.04E-10 0.OOE+01 0.OOE+01 0.OOE+01 4.04E-10 0.OOE+01 70 MLO SHC'-U 3.22E-10 0.OOE+01 0.OOE+01 3.22E-10 0.OOE+01 0.OOE+01 71 CCW SLR'-E 3.19E-10 0.OOE+01 0.OOE+01 3.19E-10 0.OOE+01 0.00E+01 72 VEF ALR-S 3.00E-10 3.00E-10 3.OOE-10 0.00E+01 0.OOE+01 0.00E+01 73 SBO THRI'-U 2.61E-10 O.00E+01 0.00E+01 2.61E-10 0.OOE+01 0.OOE+01 74 LLO ALC'-U 2.35E-10 0.OOE+01 0.00E+01 2.35E-10 0.OOE+01 0.OOE+01 75 SBO THR'-E 2.20E-10 0.00E+01 0.OOE+01 2.20E-10 0.00E+01 0.OOE+01 76 ESW SHR-S 2.04E-10 2.04E-10 2.04E-10 0.OOE+01 0.OOE+01 0.00E+01 77 ESW SHC-J 2.04E-10 0.00E+01 0.OOE+01 0.OOE+01 2.04E-10 0.OOE+01 78 VDC THR'-E 1.79E-10 0.OOE+01 0.OOE+01 1.79E-10 0.00E+01 0.OOE+01 79 MLO SL-J 1.73E-10 0.OOE+01 0.OOE+01 0.OOE+01 1.73E-10 0.OOE+01 80 SGR GHW-T 1.23E-10 0.OOE+01 0.OOE+01 0.OOE+01 0.00E+01 1.23E-10 81 SLB THR'-E 1.14E-10 0.OOE+01 0.OOE+01 1.14E-10 0.00E+01 0.OOE+01 82 ESW THC'-U 1.06E-10 0.OOE+01 0.00E+01 1.06E-10 0.OOE+01 0.OOE+01 TOTAL FREQUENCY (PER YEAR) 7.1470E-05 3.4490E-05 4.0999E-05 1.0507E-07 1.9554E-05 9.6894E-06

• Detailed definitions of initiators and end state codes are provided in Reference 4. Definitions relevant to present application are discussed in text above.

17141-0007-A2-RI.doc