Text

Request for License Amendment Technical Specifications 6.8.4.k and TS Surveillance Requirement (SR) 4.6.1.6.1
	ML051050123
Person / Time
Site:	Harris
Issue date:	04/06/2005
From:	Scarola J; Progress Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	HNP-05-002
	Download: ML051050123 (85)
	v • d • e

James Scarola C Progress Energy Vice President Harris Nuclear Plant Progress Energy Carolinas. Inc.

APR 0 6 20Q5 Serial: HNP-05-002 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.6.1 Ladies and Gentlemen:

In accordance with the Code of Federal Regulations, Title 10, Part 50.90, Carolina Power and Light Company (CP&L) doing business as Progress Energy Carolinas, Inc., requests a license amendment for the Harris Nuclear Plant (HNP) to Technical Specifications (TS) 6.8.4.k, "Containment Leakage Rate Testing Program" and TS Surveillance Requirement (SR) 4.6.1.6.1, "Containment Vessel Surfaces." The proposed amendment would modify the TS to allow for a one-time extension of the containment Type A test interval from once in 10 years to once in 15 years.

HNP has re-evaluated the risk basis for the previously approved TS change and has determined that a one-time interval extension of 15 years is justified. The extension of the Type A test from once in 10 years to once in 15 years is consistent with extensions recently granted to other licensees. provides the description, background, and technical analysis for the proposed change to the Technical Specifications. details, in accordance with 10 CFR 50.91(a), the basis for HNP's determination that the proposed change to the Technical Specifications does not involve a significant hazards consideration. provides the proposed Technical Specification change. provides the revised Technical Specification page. provides a copy of calculation HNP-F/PSA-0066, "Evaluation of Risk Significance of ILRT Extension."

P.O.

Box 165 New Hill, NC 27562 6

T > 919.362.2502 F > 919.362.2095

HNP-05-002 Page 2 With respect to this proposed change there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite and there is no significant increase in individual or cumulative occupational radiation exposure.

The proposed change to the Technical Specifications meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental assessment or environmental impact statement is required for approval of this application.

In accordance with 10 CFR 50.91(b), HNP is providing the State of North Carolina with a copy of the proposed license amendment. HNP requests that the proposed amendment be issued prior to October 14, 2005 to support HNP Refueling Outage (RFO)-13, which is scheduled for April 8, 2006.

This document contains no new Regulatory Commitment.

Please refer any question regarding this submittal to Mr. Dave Corlett at (919) 362-3137.

I declare, under penalty of perjury, that the attached information is true and correct (Executed on APR 0 6 2005 ).

Sincerely, JS/jpy Attachments:

1. Description, Background, and Technical Analysis

2. 10 CFR 50.92 Evaluation

3. Proposed Technical Specification Change

4. Revised Technical Specification Page

5. HNP-F/PSA-0066, "Evaluation of Risk Significance of ILRT Extension" c:

Mr. R. A. Musser, NRC Senior Resident Inspector Ms. B. 0. Hall, N.C. DENR Section Chief Mr. C. P. Patel, NRC Project Manager Dr. W..D. Travers, NRC Regional Administrator

Attachment I to SERIAL: HNP-05-002 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. I DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.

6.1 DESCRIPTION

. BACKGROUND. AND TECHNICAL ANALYSIS Description HNP proposes to revise Technical Specifications (TS) 6.8.4.k, "Containment Leakage Rate Testing Program" to incorporate a one-time extension of the Type A test from once in 10 years to once in 15 years, which is consistent with extensions recently granted to other licensees. The proposed change will allow the Type A test to be performed within 15 years of the most recent Type A test that was performed in May 1997. The next Type A test is currently scheduled to be performed during Refueling Outage (RFO)-13 (Spring 2006). The proposed change will require performance of the next HNP Type A test no later than May 23, 2012.

In addition, HNP proposed to revise the wording of TS Surveillance Requirement (SR) 4.6.1.6.1, "Containment Vessel Surfaces," to specify that additional visual inspections are performed in accordance with Subsections IWE (Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Plants) and IWL (Requirements for Class CC Concrete Components of Light-Water Cooled Plants) of the American Society of Mechanical Engineers (ASME)Section XI Code, which provide the requirements approved by the NRC for containment vessel inspections. This proposed change does not change the number of the visual inspections of the containment exposed accessible interior surface (i.e., metal liner) from three inspections in a 1 0-year interval. This proposed change may change the number of the visual inspections of the containment exposed accessible exterior surface (i.e., concrete), depending on the scheduling of IWL inspections (i.e., from three inspections to two inspections in a 1 0-year interval), but this number of inspections corresponds with the schedule required by the IWL program approved by the ASME Section XI Code and the NRC. Subsections IWE and IWL of the ASME Section XI Code provide the latest industry standards, structured examination methodology and schedule, and specific acceptance criteria to ensure that the structural integrity and leak-tightness of the HNP containment are maintained. This proposed change is intended to specify the approved program to perform these additional visual inspections and to remove reference to an extended 1 0-year interval.

Page Al-l of 8

Attachment I to SERIAL: HNP-05-002 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. I DOCKET NO. 50400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.6.1 Proposed Change HNP proposes to revise Technical Specifications (TS) 6.8.4.k, "Containment Leakage Rate Testing Program" and TS Surveillance Requirement (SR) 4.6.1.6.1, "Containment Vessel Surfaces."

The current HNP TS 6.8.4.k requires that the next Type A test be performed within 10 years from the performance of the previous Type A test based on the successful performance of the previous Type A tests. Therefore, the current HNP TS 6.8.4.k requires that the next Type A test be performed by May 23, 2007. The proposed change would revise TS 6.8.4.k to allow the next Type A test for HNP to be performed "no later than May 23, 2012." This proposed change is intended to incorporate the one-time extension of the Type A test from once in 10 years to once in 15 years, which is consistent with extensions recently granted to other licensees.

The current HNP TS SR 4.6.1.6.1 requires that in addition to the visual inspection of the accessible interior and exterior surfaces of the containment vessel prior to a Type A test, additional inspections shall be performed during two other refueling outages before the next Type A test if the interval for the Type A test has been extended to 10 years. The proposed change would revise TS SR 4.6.1.6.1 to specify that the additional inspections are to be done in accordance with Subsections IWE and IWL of the ASME Section XI Code, which provide the requirements approved by the NRC for containment vessel inspections. This proposed change does not change the number of the visual inspections of the containment exposed accessible interior surface (i.e., metal liner) from three inspections in a 10-year interval. This proposed change may change the number of the visual inspections of the containment exposed accessible exterior surface (i.e., concrete), depending on the scheduling of IWL inspections (i.e., from three inspections to two inspections in a 10-year interval), but this number of inspections corresponds with the schedule required by the IWL program approved by the ASME Section XI Code and the NRC. Subsections IWE and IWL of the ASME Section XI Code provide the latest industry standards, structured examination methodology and schedule, and specific acceptance criteria to ensure that the structural integrity and leak-tightness of the HNP containment are maintained.

This proposed change is intended to specify the program intended to perform additional visual inspections and to remove reference to an extended 10-year interval.

As described within the Technical Justification section of this letter, the bases for the proposed change are the satisfactory results from previous tests and inspections, combined with the re-evaluation of the risk basis for the previously approved TS change.

Page Al-2 of 8 to SERIAL: HNP-05-002 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. I DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.6.1

Background

Containment structure testing is intended to assure the leak-tight integrity of the containment structure under all design basis conditions. Conservative design and construction have led to very few containment Type A tests exceeding the leak test acceptance criteria. The NRC has extended the allowable Type A test period from three times in 10 years to once in 10 years based on past successful tests. NUREG-1493, "Performance-Based Containment Leak-Test Program,"

which supported that change, also states that test periods of up to 20 years would lead to an imperceptible increase in risk.

The currently approved Harris Nuclear Plant (HNP) TS have established a program to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. This program shall be in conformance with the NRC Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, with the exception that the program is only applicable to Type A testing. Type B and Type C testing shall continue to be conducted in accordance with the original commitment to 10 CFR 50 Appendix J, Option A.

NRC Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program,"

dated September 1995 endorses Nuclear Energy Institute (NEI) guidance document NEI 94-01, Revision 0 (July 1995), "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," as a method acceptable to the NRC staff for complying with the performance-based Appendix J, Option B. RG 1.163 provides four exceptions to the guidance in NEI 94-01, Revision 0. Exception 1 discusses the test interval for Type A tests. The RG states that ANSI/ANS 56.8-1994, "Containment System Leakage Testing Requirements," test intervals are not performance-based. Therefore, licensees intending to comply with 10 CFR Part 50 Appendix J, Option B for Type A test intervals must comply with Section 11.0 of NEI 94-01, which refers the licensee to Sections 9 and 10 of that document.

NEI 94-01 Section 9.2.3, "Extended Test Intervals," discusses Type A tests. This section states that Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per 10 years based on acceptable performance history. Acceptable performance history is defined as completion of two consecutive periodic Type A tests where the calculated performance leakage rate was less than 1.0 La. Elapsed time between the first and last tests in a series shall be at least 24 months.

Exception 3 discusses the visual examination of accessible internal and exterior surfaces of the containment system for structural problems. Exception 3 further states, "These examinations should be conducted prior to initiating a Type A test, and during two other refueling outages before the next Type A test if the interval for the Type A test has been extended to 10 years, in order to allow for early uncovering of evidence of structural deterioration."

Page Al-3 of 8 to SERIAL: HNP-05-002 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.6.1 Background (continued)

The other two exceptions in RG 1.163 are not pertinent to the discussion of Type A test frequencies, but instead involve Type B and Type C testing, which is not part of this license amendment request.

The preoperational Type A test for the HNP containment structure was successfully performed in February 1986. Three consecutive periodic Type A tests have been satisfactorily completed at HNP on October 25, 1989; September 21, 1992 and May 23, 1997 (see listing in Technical Justification).

With the two most recent successful Type A tests, and a greater than 24 months of elapsed time between the two tests, HNP currently has a test interval of once every 10 years. The current 10-year interval for the completion of the next HNP Type A test ends May 23, 2007. The next Type A test is currently scheduled to be performed during Refueling Outage (RFO)-13 (Spring 2006).

Page Al-4 of 8 to SERIAL: HNP-05-002 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. I DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.6.1 Technical Justification Summary of Test and Inspection Programs Satisfactory results from previous Type A tests at Harris Nuclear Plant (HNP) as well as continued satisfactory results from Type B and Type C Local Leak Rate Tests and containment inspections support the proposed one-time extension of the containment Type A test interval.

The HNP reactor containment vessel (CV) will continue to be inspected under the requirements of the HNP programs for Subsections IWE and IWL of the American Society of Mechanical Engineers (ASME) Code,Section XI. The existing Type B and Type C containment penetration testing program will continue to be performed in accordance with 10 CFR 50 Appendix J, Option A.

Further justification is based on research documented in NUREG-1493 that, generically, very few potential containment leakage paths fail to be identified by Type B and Type C tests. In fact, an analysis of 144 ILRT results, including 23 failures, found that no failures were due to containment liner breach. The NUREG concluded that reducing the Type A (ILRT) testing frequency to one per twenty years would lead to an imperceptible increase in risk. Also, HNP has evaluated the risk significance of the ILRT extension and determined that the risk impact is very minor. This evaluation is provided as Attachment 5, "HNP-F/PSA-0066, Evaluation of Risk Significance of ILRT Extension."

Previous Type A Test Results The preoperational Type A test for the HNP containment structure was successfully performed in February 1986. Three consecutive periodic Type A tests have been satisfactorily completed at HNP with each test passing the as-found acceptance criteria. The design basis containment leak rate limit (La) is 0.1% weight per day. The table below provides the test data summary.

As Found Data (Total Time Method) for HNP ILRTs Performed at Peak Accident Pressure

(% weight per day)

Date Laml 95% UCL2 Corrections 3 As Found 95% UCL Leakage Rate 2-25-19864 0.0559 0.0719 0.0008 0.0727 10-25-1989 0.0406 0.0472 0.0008 0.0480 9-21-1992 0.0354 0.0456 0.0158 0.0614 5-23-1997 0.0285 0.0588 0.0004 0.06685 Page Al-5 of 8 to SERIAL: HNP-05-002 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.6.1 Technical Justification (continued)

Notes:

1 -

Calculated Leakage Rate 2 - Upper Confidence Limit calculated at 95 percent probability 3 - Correction for penetration paths not exposed to ILRT pressure and correction for net free volume change due to water level changes 4 - Preoperational Test 5 - Includes additions of 0.0041% weight per day associated with repair of leak identified at reduced pressure and 0.0035% weight per day for valves/penetrations which were repaired and/or adjusted prior to the ILRT Subsection IWE and IWL Program Results The HNP Subsection IWE and IWL programs are fully implemented, and expedited examinations for the first period (9/9/1998 - 9/8/2001) of the program interval (9/9/1998 -

9/8/2008) have been completed. General visual examination of 100% of the accessible surfaces of the CV liner (pressure boundary) and the reinforced concrete exterior (structural integrity) were conducted between 2000 and 2001 in accordance with the 1992 Edition (with 1992 Addenda) of the ASME Code for Subsections IWE and IWL. Those examinations are summarized below:

May 2000 (RFO-9), General visual examination of the liner, mechanical and electrical penetrations, personnel and emergency airlocks, equipment hatch, valve chambers, sumps and moisture barrier. No recordable conditions.

October 2000 - September 2001, General visual examination of the reinforced concrete exterior, including the dome exterior and main steam tunnel area. No recordable conditions.

In accordance with IWE-1240, "Surface Areas Requiring Augmented Examination," an engineering evaluation has been developed to determine areas that might require augmented examinations. No areas exist that are currently categorized as Examination Category E-C for augmented examinations.

The IWE and IWL program examinations have demonstrated that the structural integrity and leak-tightness of the HNP containment have not been compromised.

Page A1-6 of 8 to SERIAL: HNP-05-002 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.6.1 Technical Justification (continued)

The proposed change would revise TS SR 4.6.1.6.1 to specify that the additional inspections are to be done in accordance with Subsections IWE and IWL of the ASME Section XI Code, which provide the requirements approved by the NRC for containment vessel inspections. This proposed change does not change the number of the visual inspections of the containment exposed accessible interior surface (i.e., metal liner) from three inspections in a 10-year interval.

This proposed change may change the number of the visual inspections of the containment exposed accessible exterior surface (i.e., concrete), depending on the scheduling of IWL inspections (i.e., from three inspections to two inspections in a 10-year interval), but this number of inspections corresponds with the schedule required by the IWL program approved by the ASME Section XI Code and the NRC. Subsections IWE and IWL of the ASME Section XI Code provide the latest industry standards, structured examination methodology and schedule, and specific acceptance criteria to ensure that the structural integrity and leak-tightness of the HNP containment are maintained.

The IWE and IWL examinations in combination with Type B and Type C testing provide a high degree of assurance that degradation of the containment structure will be detected and corrected before it can produce a containment leak path or impact structural integrity.

Re-Evaluation of Plant-Specific Risk Basis A plant-specific risk assessment has been performed using guidance provided from the following documents:

Nuclear Energy Institute (NEI) 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 0, July 26, 1995 Electric Power Research Institute (EPRI) Topical Report TR-1 04285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," August 1994 NRC Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessments in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," July 1998 Page A1-7 of 8

Attachment I to SERIAL: HNP-05-002 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.6.1 Technical Justification (continued) to this letter provides the HNP calculation HNP-F/PSA-0066, "Evaluation of Risk Significance of ILRT Extension", Revision 0, which evaluated the risk impact associated with the proposed change. The following conclusions are summarized from the completed risk assessment:

The increase in Large Early Release Frequency (LERF) from the one-time extension of the Type A test interval from 10 years to 15 years is 2.566E-8/year, or 0.0016%. The calculated increase in LERF due to potential containment liner corrosion of 2.47E-9/year (attributable to a change in test frequency from three times in 10 years to one time in 15 years) bounds the increase from 10 years to 15 years. Therefore, the total combined increase in LERF, due to a change in test interval from once in 10 years to one time in 15 years is 2.81E-8/year.

The increase in LERF from the test interval of three times in 10 years to one time in 15 years is 7.68E-8/year, or 0.0048%. The calculated increase in LERF due to potential containment liner corrosion, considering a change in test frequency from three times in 10 years to one time in 15 years, is 2.47E-9/year. Therefore, the increase in LERF, due to both a change in test interval from three times in 10 years to one time in 15 years and potential containment liner corrosion, is 7.93E-8/year.

RG 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines "very small changes" in risk as resulting in increases of Core Damage Frequency (CDF) below 10-6/year and increases in LERF below 107/year. Since the containment Type A testing does not impact CDF, the relevant criterion is LERF.

Based on the guidance of RG 1.174, the change in LERF associated with an ILRT interval increase from 10 to 15 years at HNP constitutes a "very small change" in risk. HNP's risk assessment has demonstrated that increasing the ILRT interval from 10 to 15 years is non-risk significant.

Page Al-8 of 8 to SERIAL: HNP-05-002 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.6.1 10 CFR 50.92 NO SIGNIFICANT HAZARDS EVALUATION A written evaluation of the significant hazards consideration of a proposed license amendment is required by 10 CFR 50.92. Harris Nuclear Plant (HNP) has evaluated the proposed amendment and determined that it involves no significant hazards consideration.

According to 10 CFR 50.92, a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

I. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety The basis for this determination is presented below.

Proposed Change Carolina Power & Light (CP&L) Company doing business as Progress Energy Carolinas, Inc., is proposing a change to the Appendix A, Technical Specifications (TS), of Facility Operating License No. NPF-63, for the Harris Nuclear Plant (HNP). This change will revise the requirements of TS 6.8.4.k, "Containment Leakage Rate Testing Program," to incorporate a one-time extension to the 1 0-year interval for the performance-based leakage rate testing program for Type A tests specified by Nuclear Energy Institute (NEI) 94-01, "Industry Guideline For Implementing Performance-Based Option of 10 CFR Part 50, Appendix J,"

Revision 0, July 26, 1995, and endorsed by 10 CFR Part 50, Appendix J, Option B. The proposed change will allow the Type A test to be performed within 15 years of the most recent Type A test that was performed in May 1997. The proposed change will require performance of the next HNP Type A test no later than May 23, 2012. In addition, a change to the wording of TS Surveillance Requirement (SR) 4.6.1.6.1, "Containment Vessel Surfaces," will specify that additional visual inspections are performed in accordance with Subsections IWE and IWL of the ASME Section XI Code. This proposed change is intended to specify the approved program to perform additional visual inspections and to remove reference to an extended 1 0-year interval.

Page A2-1 of 4 to SERIAL: HNP-05-002 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. I DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.6.1 Basis This change does not involve a significant hazards consideration for the following reasons:

1. The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change to HNP TS 6.8.4.k and TS SR 4.6.1.6.1 provide a one-time extension of the containment Type A test interval from 10 years to 15 years and specifies that additional visual inspections are done in accordance with Subsections IWE and IWL of the ASME Section XI Code. The existing 10-year test interval is based on past test performnance. The proposed TS change does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment vessel is designed to provide a leak-tight barrier against the uncontrolled release of radioactivity to the environment in the unlikely event of postulated accidents. As such, the containment vessel is not considered as the initiator of an accident. Therefore, the proposed TS change does not involve a significant increase in the probability of an accident previously evaluated.

The proposed change involves only a one-time change to the interval between containment Type A tests. Type B and C leakage testing will continue to be performed at the intervals specified in 10 CFR Part 50, Appendix J, Option A, as required by the HNP TS. As documented in NUREG-1493, "Performance-Based Containment Leakage-Test Program," industry experience has shown that Type B and C containment leak rate tests have identified a very large percentage of containment leak paths, and that the percentage of containment leak paths that are detected only by Type A testing is very small. In fact, an analysis of 144 integrated leak rate tests, including 23 failures, found that none of the failures involved a containment liner breach. NUREG-1493 also concluded, in part, that reducing the frequency of containment Type A testing to once per 20 years results in an imperceptible increase in risk. The HNP test history and risk-based evaluation of the proposed extension to the Type A test interval supports this conclusion. The design and construction requirements of the containment vessel, combined with the containment inspections performed in accordance with the American Society of Mechanical Engineers (ASME) Code,Section XI, and the Maintenance Rule (10 CFR 50.65) provide a high degree of assurance that the containment vessel will not degrade in a manner that is detectable only by Type A testing. Therefore, the proposed TS change does not involve a significant increase in the consequences of an accident previously evaluated.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Page A2-2 of 4 to SERIAL: HNP-05-002 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. I DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.6.1 Basis (Continued)

2. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change to HNP TS 6.8.4.k and TS SR 4.6.1.6.1 provide a one-time extension of the containment Type A test interval to 15 years and specifies that additional visual inspections are done in accordance with Subsections IWE and IWL of the ASME Section XI Code. The existing 1 0-year test interval is based on past test performance.

The proposed change to the Type A test interval does not result in any physical changes to HNP. In addition, the proposed test interval extension does not change the operation of HNP such that a failure mode involving the possibility of a new or different kind of accident from any accident previously evaluated is created.

Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed amendment does not involve a significant reduction in a margin of safety.

The proposed change to HNP TS 6.8.4.k and TS SR 4.6.1.6.1 provide a one-time extension of the containment Type A test interval from 10 years to 15 years and specifies that additional visual inspections are done in accordance with Subsections IWE and IWL of the ASME Section XI Code. The existing 10-year test interval is based on past test performance. The NUREG-1493 study of the effects of extending containment leak rate testing found that a 20 year extension for Type A testing resulted in an imperceptible increase in risk to the public. NUREG-1493 found that, generically, the design containment leak rate contributes a very small amount to the individual risk and that the decrease in Type A testing frequency would have a minimal affect on this risk since most potential leak paths are detected by Type B and C testing. The proposed change involves only a one-time extension of the interval for containment Type A testing; the overall containment leak rate specified by the HNP TS is being maintained. Type B and C testing will continue to be performed at the frequency required by the HNP TS. The regular containment inspections being performed in accordance with the ASME Code,Section XI, and the Maintenance Rule (10 CFR 50.65) provide a high degree of assurance that the containment will not degrade in a manner that is only detectable by Type A testing. In addition, a plant-specific risk evaluation has demonstrated that the one-time extension of the Type A test interval from 10 years to 15 years results in a very small increase in risk for those accident sequences influenced by Type A testing.

Therefore, this change does not involve a significant reduction in a margin of safety.

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

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CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintainod at a level consistent with the acceptance criteria in Specification 4.6.1.6.1.

APPLICABILITY:

MODES 1. 2. 3. and 4.

ACTION:

With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.6.1 Containment Vessel Surfaces. The structural integrity of the exposed accessible interior and exterior surfaces of the containment vessel, including the liner plate, shall be determined, during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2). by a visual inspection of these surfaces.

This inspection shall be performed prio to the Type A containment leakage rate test to verify no apparent changes in appearanco d Gradat n. Addito I c

_ho intr.A LE for tho Typo A tc3t Ez hben Oextonfe tA 19 eAtor.

4.6.1.6.2 Reports.

Any abnormal degradation of the containment vessel structure detected during the above required inspections shall be reported to td

)

the Commission in a Special Report pursuant to Specification 6.9.2 within 15 days. This report shall include a description of the condition of the concrete, the inspection procedure. the tolerances on cracking. and the c rretive actions taken.

'rda vioseote'oe h S e*3 ov" IWE av, AJWL. a the tAe )

c,;jer g; t

rA Fre'ore Vet 'en CoAe.., 'eL 4

c cv XI.

SHEARON HARRIS - UNIT 1 3/4 6-8 Amendment No.l

ADMINISTRATIVE CONTROLS I

PROCEDURES AND PROGRAMS (Continued)

k. Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54 (o) and 10 CFR 50 Appendix J. Option B. as modified by aproved exemptions. This program shall be in conformance with the NRC Regulatory Guide 1.163.

'Performance-Based Containment Leak-Test Program," dated September 1995. with the following exceptior joted'

1) The above Containment Leakage Rate Teting Program is only applicable to Type A testing. Type B and C testing shall continue to be conducted in accordance with the original commitment to 10 CFR 50 Appendix J. Option A.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident is 41.8 psig. The calculated l peak containment internal pressure related to the design basis main steam line break is 41.3 psig.

Pa will be assumed to be 41.8 psig for the purpose of containment testing in accordance with this Technical Specification.

(4)

The maximum allowable containment leakage rate. La at Pa. shall be 0.1 % of containment air weight per day.

The containment overall leakage rate acceptance criterion is < 1.0 La.

During the first unit startup following testing in accordance with this program. the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests. and < 0.75 La for Type A tests.

The provisions of Surveillance Requirement 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program. However, test frequencies specified in this Program may be extended consistent with the guidance provided in Nuclear Energy Institute (NEI) 94-01. "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J." as endorsed by Regulatory Guide 1.163. Specifically. NIE 94-01 has this provision for test frequency extension:

1) Consistent with standard scheduling practices for Technical Specifications Required Surveillances. intervals for recommended Type A testing may be extended by up to 15 months.

This option should be used only in cases where refueling schedules have been changed to accommodate other factors.

The provisions of Surveillance Requirement 4.0.3 are applicable to

52) rhe. ;;r<t Tyafe. P..

4zCt per~orrzeca Jct aches M> '2-, '997 th onyt ainment Leakg atv e Testaingd Progrm IRQ vnrcyt>So SHEARON HARRIS - UNIT 1 6-19c Amendment No. 1 I

to SERIAL: HNP-05-002 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. I DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.6.1 REVISED TECHNICAL SPECIFICATION PAGE Page A4-1 of 3

CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.1.

APPLICABILITY: MODES 1, 2. 3. and 4.

ACTION:

With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.6.1 Containment Vessel Surfaces. The structural integrity of the exposed accessible interior and exterior surfaces of the containment vessel.

including the liner plate, shall be determined, during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2). by a visual inspection of these surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance or other abnormal degradation. Additional inspections shall be conducted in accordance with Subsections IWE and IWL of the ASME Boiler and Pressure Vessel Code,Section XI.

4.6.1.6.2 Reports. Any abnormal degradation of the containment vessel structure detected during the above required inspections shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2 within 15 days.

This report shall include a description of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective actions taken.

SHEARON HARRIS - UNIT 1 3/4 6-8 Amendment No.

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

k. Containment LeakaQe Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54 (o) and 10 CFR 50 Appendix J. Option B. as modified by a pproved exemptions. This program shall be in conformance with the NRC Regulatory Guide 1.163.

'Performance-Based Containment Leak-Test Program," dated September 1995. with the following exceptions noted:

1) The above Containment Leakage Rate Testing Program is only applicable to Type A testing. Type B and C testing shall continue to be conducted in accordance with the original commitment to 10 CFR 50 Appendix J. Option A.

2) The first Tpe A test performed after the May 23. 1997 Type A test shall be performed no later than May 23, 2012.

3) Visual examination of the containment system shall be in accordnace with Specificaiton 4.6.1.6.1.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident is 41.8 psig.

The calculated peak containment internal pressure related to the design basis main steam line break is 41.3 psig.

Pa will be assumed to be 41.8 psig for the purpose of containment testing in accordance with this Technical Specification.

The maximum allowable containment leakage rate, La at Pa. shall be 0.1 % of containment air weight per day.

The containment overall leakage rate acceptance criterion is < 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and < 0.75 La for Type A tests.

The provisions of Surveillance Requirement 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program. However, test frequencies specified in this Program may be extended consistent with the guidance provided in Nuclear Energy Institute (NEI) 94-01. "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J." as endorsed by Regulatory Guide 1.163. Specifically NE I 94-01 has this provision for test frequency extension:

1) Consistent with standard scheduling practices for Technical Specifications Required Surveillances, intervals for recommended Type A testing may be extended by up to 15 months. This option should be used only in cases where refueling schedules have been changed to accommodate other factors.

The provisions of Surveillance Requirement 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

SHEARON HARRIS - UNIT 1 6-19c Amendment No.

to SERIAL: HNP-05-002 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. I DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATIONS (TS) 6.8.4.K AND TS SURVEILLANCE REQUIREMENT (SR) 4.6.1.6.1 HNP-F/PSA-0066, "EVALUATION OF RISK SIGNIFICANCE OF ILRT EXTENSION" Page A5-1 of 80

SYSTEM #

CALC. SUB-TYPE PRIORITY CODE QUALITY CLASS N/A N/A 4

E NUCLEAR GENERATION GROUP HNP-F/PSA-0066 (CALCULATION #)

EVALUATION OF RISK SIGNIFICANCE OF ILRT EXTENSION (Title including structures, systems, components)

LI BNP UNIT__

LI CR3 Z HNP LRNP L NES Wf ALL APPROVAL REV PREPARED BY REVIEWED BY SUPERVISOR Sigr~ture Signature Sig 8nature kkt~m4m!ACa 6o1v -kft..q s4,Q;^VK NamName Name 0

Bruce A. Morgen Bradley W. Dolan Steven A. Laur Date Date Date 04106104 04106/04 4-1_

04__do (For Vendor Calculations)

Vendor Vendor Document No.

Owners Review By Date.

CALCULATION NO.HNP-F/PSA-0066 REV. 0 PAGE 2 of 9 LIST OF EFFECTIVE PAGES PAGE REV PAGE REV ATTACHMENTS Cover 0

Number Rev Number of Pacies 2-9 0

1 0

51 2

0 3

3 0

3 4

0 13 AMENDMENTS Letter Rev Number of Pages

CALCULATION NO.HNP-F/PSA-0066 REV. 0 PAGE 3 of 9 REVISION

SUMMARY

Rev. #

Revision Summary (list of ECs incorporated) 0 Revision 0 evaluates the risk significance of extending the ILRT test interval at the Harris Nuclear Plant (HNP) to 15 years, using the method developed for the Crystal River 3 (CR3)

ILRT extension. This calculation also evaluates the risk due to postulated concealed containment liner corrosion at HNP, using a method obtained from a relevant Calvert Cliffs RAI response.

CALCULATION NO.HNP-F/PSA-0066 REV. 0 PAGE 4 of 9 DOCUMENT INDEXING TABLE Document ID Number Function Relationship to Calc.

Action Type (e.g., Calc No., Dwg.

(i.e. IN for design (e.g. design input, assumption basis, (specify if Doc. Services (e.g. CALC, DWG No., Equip. Tag No.,

inputs or references; reference, document affected by or Config. Mgt. to Add, TAG.

CEDUR, Procedure No.,

OUT for affected results)

Deleted or Retain) (e.g.,

SOFTWARE)

Software name and documents)

CM Add, DS Delete)

SOFTWARE)____

version)

CALC HNP-F/PSA-0001 IN Used in the calculation DS Retain LICENSING HO-930142 IN Individual Plant Examination DS Retain (IPE) Submittal, August 1993.

CALC HNP-F/PSA-0067 IN Used in the calculation DS Retain

CALCULATION NO.HNP-F/PSA-0066 REV. 0 PAGE 5 of 9 RECORD OF LEAD REVIEW p

1 Design HNP-F/PSA-0066 Re RISK SIGNIFICANCE OF ILRT EXTENSION vision 0

The signature below of the Lead Reviewer records that:

the review indicated below has been performed by the Lead Reviewer; appropriate reviews were performed and errors/deficiencies (for all reviews performed) have been resolved and these records are included in the design package; the review was performed in accordance with EGR-NGGC-0003.

O] Design Verification Review El Design Review E] Alternate Calculation E Qualification Testing El Engineering Review E

Owner's Review o] Special Engineering Review E] YES 3 N/A Other Records are attached.

Bradlev W. Dolan / Tt. 66t q

b dttm god I

I PSA Discioline 04106104 Lead Reviewer Date Item No.

Deficiency Resolution None.

FORM EGR-NGGC40003-2-5 This form Is a GA Record when completed and Included with a completed design package. Owners Reviews may be processed as stand alone GA records when Owner's Review Is completed.

EGR-NGGC-0003 l

Rev9 9 l

CALCULATION NO.HNP-F/PSA-0066 REV. 0 PAGE 6 of 9 RECORD OF INTERDISCIPLINARY REVIEWS Design HNP-F/PSA-0066

[] Design Verification Review Z

Engineering Review El Design Review a] Alternate Calculation j Qualification Testing Revision 0

[] Owner's Review 13Special Engineering Review Review: 1. References used to calculate liner surface area

2. Procedure used to pick the test pressure

3. Assumption made that only 97% of the dome and liner can be accessed for visual inspection. This is intended to cover the fuel transfer canal and obstructed or high dose areas.

4. Assumption that the sump liner can be ignored for calculating the surface area of the liner.

5. Assumption that 1 psia Is added to test pressure for conservatism.

l.i ~A"tL*IC4

.Jnhn KAlIv ME M.

'.-o Date Concurrent Reviewer (orint/sian)

Discioline Item No.

Deficiency Resolution 1

Two significant digits are displayed in Table 1 Display altered to show appropriate level of for the ILRT test pressure and upper end and significant figures (no digits to the right of the lower end pressures.

This is not consistent decimal place).

with rounding employed in the calculation for transforming from psig to psia.

FORM EGR-NGOC-0003-3-5 This form Is a CA Record when completed and included with a completed design package. Owner's Reviews may be processed as stand alone OA records when Owner's Review is completed EGR-NGGC-0003 I

Rev 9 l

CALCULATION NO.HNP-F/PSA-0066 REV. 0 PAGE 7 of 9 TABLE OF CONTENTS Page No.

List of Effective Pages.........................

2 Revision Summary.........................

3 Document Indexing Table.........................

4 Record of Lead Review.........................

5 Record of Interdisciplinary Reviews.........................

6 Table of Contents.........................

7 Purpose.........................

8 References.........................

8 Body of Calculation.........................

8 Conclusions.........................

9 Attachments

1.

RSC REPORT 04-03

2.

OWNERS REVIEW OF RSC 04-03

3.

RISK FROM CONCEALED LINER CORROSION

4.

CALVERT CLIFFS METHOD

CALCULATION NO.HNP-F/PSA-0066 REV. 0 PAGE 8 of 9 Purpose This calculation evaluates the risk significance of extending the ILRT test interval at the Harris Nuclear Plant (HNP) to 15 years, using the method developed for the Crystal River 3 (CR3) ILRT extension. This calculation also evaluates the risk due to postulated concealed containment liner corrosion at HNP, using a method obtained from a relevant Calvert Cliffs RAI response. A sensitivity evaluation to certain conservative assumptions is also provided.

References

1.

HNP-93-835, -Shearon Harris Nuclear Power Plant Docket No. 50-400/License No. NFP-63, Submittal of Individual Plant Examination (IPE)", August 20,1993.

2.

HNP-F/PSA-0001, "HNP Probabilistic Safety Assessment Model", Revision 5.

3.

HNP-F/PSA-0067, 'Estimate of 50 Mile Population Dose from Design Basis Containment Leakage Following a Core Melt Accident", Revision 0.

4.

RSC 04-03, "Evaluation of Risk Significance of ILRT Extension", Revision 0, April 2004 (Attachment 1)

5.

Constellation Nuclear, Calvert Cliffs Nuclear Power Plant, Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, 3/27/02.

(Provided as )

6.

EST-210, 'Periodic Containment Integrated Leak Rate Testing (Type A Test)",

Revision 11.

7.

HNP-F/PSA-0059, -HNP PRA - Appendix L - Summary Document", Revision 2.

8.

(Drawing CAR-216)8-G-0228, 'CONTAINMENT BLDG - CONTAINMENT LINER-SH 1 UNIT 1", Revision 13.

9.

Letter, Jan F. Lucas to USNRC, "Supplement to Amendment Request Regarding One-Time Extension of Containment Type A Test Interval", Serial: RNP-RA/03-0121, October 13, 2003.

Body of Calculation The Probabilistic Safety Assessment (PSA) model does not provide plant design basis information nor is the PSA model used to modify design outputs. Therefore, no design inputs are used.

The inputs to and assumptions for the ILRT evaluation are documented in the vendor report. The inputs to and assumptions for the evaluation of concealed containment liner corrosion are documented in Attachment 3.

Acceptance of the Attachment 1 vendor report is documented by the enclosed Owners Review, shown as Attachment 2.

CALCULATION NO.HNP-F/PSA-0066 REV. 0 PAGE 9 of 9 Conclusions The risk impact of the proposed extension of the ILRT test interval as documented herein is very minor, by a number of measures.

Reg. Guide 1.174 provides guidance by determining the risk impact of specific plant changes.

It defines very small changes in risk as those resulting in increases in core damage frequency (CDF) below 1 E-6/yr and increases in LERF below 1 E-7/yr. Since the ILRT does not impact CDF, the relevant metric is LERF.

Referring to the information in Table 1 of Attachment 1, the increase in LERF due to extending the test interval from the current once-per-10 year basis is 2.56E-8/yr. This change is very small.

Referring to the information in Table 1 of Attachment 3, the increase in LERF due to potential concealed corrosion is 2.47E-9/yr. This change is very small. The combined increase in LERF is 2.81 E-8/yr and is very small.

The sensitivity evaluation documented herein (using a refined definition for sequences contributing to LERF due to the ILRT interval) demonstrates that the above risk impact contains a substantial amount of conservatism.

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Progress Energy RSC 04-03 Harris Nuclear Plant Probabilistic Safety Assessment C,.........

1,; zt,

-3 Evaluation of Risk Significance of ILRT Extension Revision 0 April 2004 Principal Analyst Ricky Summitt

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension RSC Document Configuration Control Form FORM NO.: RSC-RPT-STD99-04, Rev. 7

Title:

Evaluation of Risk Significance of ILRT Extension Report Number: RSC 04-03 Revision Preparer Reviewer/Date Approver/Date Original Issue R. Summitt JDM/3-19-2004 RLS/4-5-2004 I

E I

2 I

3

/

Abstract (brief statement of purpose): Develop a risk informed analysis of the impact of extending the Harris Nuclear Plant ILRT.

List of keywords for document retrieval (not more than three): Level 3 Analysis, ILRT, Harris Nuclear Plant RSC Project Number:

RSC 04-01 Project Folder on RSC Server:

HNP ILRT Extension Document Name and Last Saved:

RSC 04-03 Ri.doc / April 05, 2004 Software and Version Used:

Word (doc) Pre-XP, Version XP (bold all that apply)

Excel (xis): Pre-XP, Version XP Access (mdb): Version 97, Version 2000, Version XP Visio Version (vsd)': 2002 CAFTA: Version 3.2b, 5.0 EOOS Version 2.5, Version 32-bit MAAP BWR Version 3.0B R9, R1 1, Version 4.0 MAAP PWR Version 18, 19,20, Version 4.0 RSC Software: PRAMS, SIP, TIFA, FATCAT, DBM INC RSC STD RPT R8

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significancc of ILRT Extension Report Review and Resolution Form FORM NO.: RSC-RPT-RVROO-02, Rev. 8 Verification and Review Method: (Place X to Left of Method Utilized)

X Detailed Review Alternative Calculation Other (list below)

Review Not Required Describe Other Review Method:

General Documentation Requirements Acceptable Reviewer Comments (Yes or No)

Introduction provides summary of

purpose, scope, and principle tasks required to meet objective Yes Methodology - description of process and supporting methodology that is sufficient to understand approach and to support peer Yes review Analysis and Results detailed documentation of the implementation of the methodology and task steps that may be Yes supported by report appendices and includes intermediate and final results Conclusions and Recommendations concise presentation of results of the analysis that answers the objectives of the Yes study and should include any important assumptions and/or findings Editorial Review: Yes Spell Checked, Grammar Checked, Tables and Figures Checked and Section Titles Checked IRSC RSC STD RPT R8

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Report Review and Resolution Form FORM NO.: RSC-RPT-RVROO-02, Rev. 8 RSC Internal Comment Resolution Reviewer Comment Resolution of Comment

1. Editorial comments included in Corrected identified editorial errors.

markup.

2. Check marked data in Table 1 to see Values corrected.

if the marked data is correct.

3. Check 3a and 3b dose on page 21.

Listed doses are correct.

.1.

Editorial or illustrative comments are attached to this review sheet to complete the review package.

BSC RSC STD RPT R8

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Table of Contents Section Page 1.0 PURPOSE.......................................................

1I 1.1

SUMMARY

OF THE ANALYSIS......................................................

1 1.2

SUMMARY

OF RESULTS/CONCLUSIONS......................................................

1 2.0 DESIGN INPUTS........................................................

4 3.0 ASSUMPTIONS........................................................

7 4.0 CALCULATIONS........................................................

8 4.1 CALCULATIONAL STEPS........................................................

8 4.2 SUPPORTING CALCULATIONS.......................................................

10 4.3 ALTERNATIVE LERF CALCULATION USING REFINED DATA 22 4.4 REVISED LERF BASED ON INTACT CONTAINMENT.

24 4.5 REVISED TYPE A LERF EXCLUDING PREEXISTING LERF AND NON-LERF CASES......................................................

25

5.0 REFERENCES

28 Appendix A.......................................................

I RSC 04-03 i

Printed: 04/06/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension List of Tables Table Page Table 1 Summary of Risk Impact on Extending Type A ILRT Test Frequency................

................. 2 Table 2 HNP Plant Damage States..................................................................:.....................................4 Table 3 Release Category Radionuclide Percentage Release and Total Person-Rem 7

Table 4 Containment Failure Classifications (from Reference 6)

.0 Table 5 HNP PSA RC Grouping to EPRI Classes (Reference 6).1 Table 6 Baseline Risk Profile (three tests per ten years).15 Table 7 Current Risk Profile (one test per ten years)

.17 Table 8 Proposed Risk Profile (once per fifteen years).19 Table 9 Impact on LERF due to Extended Type A Testing Intervals.21 Table 10 Impact on Conditional Containment Failure Probability due to Extended Type A Testing Intervals..........................................................

22 Table 11 LERF Release Category Contributions..........................................................

23 Table 12 Class 3b Contributions Using Adjusted CDF..........................................................

24 Table 13 Comparison of Class 3b, Class 7 and Class 8 Population Doses........................................ 24 Table 14 Class 3b Contributions Using Intact Containment Component Only................................. 25 Table 15 Contributing PDSs with CSET States A, D or G..........................................................

26 Table 16 Class 3b Contributions Using Adjusted CDF.....................................

..................... 28 RSC 04-03 ii Printed: 04/06/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension 1.0 PURPOSE The purpose of this calculation is to evaluate the risk of extending the Type A Integrated Leak Rate Test (ILRT) interval beyond the current 10 years required by 10 CFR 50, Appendix J at the Shearon Harris Nuclear Plant.

1.1

SUMMARY

OF THE ANALYSIS 10 CFR 50, Appendix J allows individual plants to extend Type A surveillance testing requirements and to provide for performance-based leak rate testing. This report documents a risk-based evaluation of the proposed change of the integrated leak rate test (ILRT) interval for the Shearon Harris Nuclear Plant (HNP). The proposed change would impact testing associated with the current surveillance test for Type A leakage (procedure EST-210)'. No change to Type B or Type C testing is proposed at this time.

The evaluation for HNP is consistent with similar assessments performed for the Indian Point 3 (IP3) plant23 and for the Crystal River 3 (CR3) plant4 which were approved by the NRC. This assessment utilizes the guidelines set forth in NEI 94-015, the methodology used in EPRI TR-1042856 and the regulatory guidance on the use of Probabilistic Risk Assessment (PRA) findings in support of a licensee request to a plant's licensing basis, RG 1.1747.

The calculation evaluates the risk associated with various ILRT intervals as follows:

3 years - Historical testing data interval (3 tests per 10 years) 10 years - Current testing interval for HNP (1 test per 10 years) 15 years - Proposed extended test interval, similar to IP3 request (1 test per 15 years)

The analysis utilizes the recent HNP probabilistic safety assessment (PSA) results8. The PSA was initially developed for the HNP individual plant examination (IPE)9 to estimate the baseline core damage and plant damage classes. Several updates to the HNP level I analysis and to the level 2 information have been incorporated since the IPE.

The release category and person-rem information are based on design basis leakage evaluations and extrapolation of the release category information using a modeling framework that develops the person-rem estimates based on the relative release fractions of radionuclides. This approach has been presented in other licensing submittals (Reference 10). The approach is consistent with and similar to the method used in the CR3 submittal (Reference 4). The framework is described in Appendix A.

1.2

SUMMARY

OF RESULTS/CONCLUSIONS The specific results are summarized in Table 1 below. The Type A contribution to LERF is defined as the contribution from Class 3b.

RSC 04-03 I

Printed 04/06/04

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Table 1 Summary of Risk Impact on Extending Type A ILRT Test Frequency RiskI Impact for rs i

0-Risk Impact for 15-(baseline) years (current years requirement) otal Integrated Risk (Person-Rem/yr) 3.21127E+2 3.21137E+2 3.21142E+2 Type A Testing Risk (Person-Rem/yr) 1.094E-1 1.203E-1 1.258E-1

% Total Risk (Type A / Total) 0.034%

0.037%

0.039%

Type A LERF (Class 3b) (per year) 5.12E-7 5.63E-7 5.89E-7 Changes due to extension from 1 Oyears (current)

A Risk from current (Person-rem/yr) 5.13E-3

% Increase from current (A Risk Total Risk) 0.0016%

A LERF from current (per year) 2.566E-8 A CCFP from current 0.104%

Changes due to extension from 3 years (baseline),Ac --

A Risk from baseline (Person-rcm/yr) 1.54A-2

% Increase from baseline (A Risk Total Risk) 0.0048%

A LERF from baseline (peryear) 7.68E-8 CCFP from baseline 0.312%

RSC 04-03 2

Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Based on the analysis and available data the following is stated:

The person-rem/year increase in risk contribution from extending the ILRT test frequency from the current once-per-ten-year interval to a once-per-fifteen year interval is 5.13E-3 person-rem/yr.

The risk increase in LERF from extending the ILRT test frequency from the current once-per-10-year interval to a once-per-1S year interval is 2.56E-8/yr.

The change in CCFP from the current once-per-10-year interval to a once-per-15 year interval is 0.104%.

The change in Type A test frequency from once-per-ten-years to once-per-fifteen-years increases the risk impact on the total integrated plant risk by only 0.0016%. Also, the change in Type A test frequency from the original three-per-ten-years to once-per-fifteen-years increases the risk only 0.0048% Therefore, the risk impact when compared to other severe accident risks is negligible.

Regulatory Guide (RG) 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines very small changes in risk as any change resulting in an incremental increase in CDF below 10-6/yr and an incremental increase in LERF below 107/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 a once-per-ten-years to a once per-fifteen-years is 2.56E-8/yr.

Since guidance in RG 1.174 defines very small changes in LERF as below 1047/yr, increasing the ILRT interval from 10 to 15 years is therefore considered non-risk significant. In addition, the change in LERF resulting from a change in the Type A ILRT test interval from a three-per-ten-year interval to a once per-fifteen-year interval is 7.68E-8/yr. This calculated increase in LERF is considered very small since it falls below the RG 1.174 guidance criterion.

RG 1.174 also encourages the use of risk analysis techniques to help ensure and show that the proposed change is consistent with the defense-in-depth philosophy. Consistency with defense-in-depth philosophy is maintained by demonstrating that the balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation. The change in conditional containment failure probability was estimated to be 0.104% for the proposed change and 0.312% for the cumulative change of going from a test interval of 3 in 10 years to 1 in 15 years. These changes are small and they demonstrate that the defense-in-depth philosophy is maintained.

RSC 04-03 3

Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension 2.0 DESIGN INPUTS The HNP PSA is a non-safety related tool and is intended to provide "best estimate" results that can be used as an input to the risk informed decision process.

The current HNP PSA (Reference 8) is an update to the IPE, which was a NRC submittal of the PSA provided in response to requests from Generic Letter 88-20. The PSA is not considered as design basis information.

The inputs for this calculation come from the information documented in the HNP PSA and the more recent update (Reference 8).

The total core damage frequency (CDF) for HNP is 2.46E-5/yr. There are 82 plant damage states that contribute to the CDF. The top 20 capture over 95% of the total CDF and they are summarized in Table 2.

Table 2 HNP Plant Damage States Damage State

- RepresentativeSequence Frequency (/yr) lop Loss of offsite power, RCP seal LOCA at diesel failure time, no 5.38E-6 RCS makeup.

iP Loss of offsite power, failure of AFW after battery depletion, RCP 5.16E-6 seal LOCA at 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . Containment spray and fans are failed.

2A Loss of feedwatcr, failure of AFW, and operator fails to accomplish 3.04E-6 feed-and-bleed cooling. The containment is isolated and both containment sprays and fans function.

X16C SGTR with failure of safety injection and one SRV associated with 2.54E-6 the faulted steam generator fails open. This results in a large bypass sequence.

4A Loss of an ac bus and failure of both AFW and high-pressure 2.52E-6 recirculation. Containment isolation occurs and the containment sprays and fans are functioning.

7A ATWS occurs with overpressure due to insufficient moderator 9.78E-7 feedback. Containment isolation occurs and the containment sprays and fans are functioning.

B16B SGTR, failure of SI and shutdown cooling with a cycling SRV on 8.92E-7 the faulted steam generator. The releases through the steam generator SRV results in a small bypass.

17A Sl LOCA occurs with failure of recirculation due to loss of RHR 5.68E-7 pumps. Containment isolation occurs and the containment sprays and fans are functioning.

13A S2 LOCA with failure of recirculation. Containment isolation 4.35E-7 occurs and the containment sprays and fans arc functioning.

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ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Table 2 (continued)

HNP Plant Damage States Damage State

Representative Sequence Frequency (/yr) 15A S2 LOCA occurs without injection. Containment isolation occurs 3.85E-7 and the containment sprays and fans are functioning.

B3B SGTR occurs with failure of RCS depressurization and loss of SI

2.54E-7 source. The releases through the steam generator SRV results in a small bypass.

3P Similar to lOP except operators fail to depressurize the RCS.

2.52E-7 8G Loss of one emergency bus and failure of cooling to the RCPs 2.34E-7 resulting in a seal LOCA. Failure of all injection and shutdown cooling.

IA Similar to 2A with a pressurizer PORV opening late and failing to 1.67E-7 reclose. Containment isolation occurs and the containment sprays and fans are functioning.

3A Transient-induced LOCA with failure of long term cooling due to 1.52E-7 the failure of depressurization and cooldown of the RCS. A failure of the RHR system may also occur.

5A Sl LOCA occurs with failure of recirculation and cooldown.

1.34E-7 Containment isolation occurs and the containment sprays and fans are functioning.

4P SI LOCA occurs with failure of heat removal and no feed-and-1.32E-7 bleed cooling. Isolation of the containment is successful, but the containment sprays and fans are failed.

12A Medium LOCA occurs with a failure of recirculation. Containment 1.17E-7 isolation occurs and the containment sprays and fans are functioning.

IQ Similar to I P except that the isolation failure is small.

1.09E-7 IOQ Similar to lOP except that the isolation failure is small.

I.051E-7 Other PDS 1.04E-6 Contributors Total

-2.46E-5 RSC 04-03 5s Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension In order to develop the person-rem dose associated with the plant damage state, it is necessary to associate each plant damage state with an associated release of radionuclides and from this information to calculate the associated dose.

The IP3 submittal (Reference 2) utilizes a multiplication factor to adjust the design basis leakage value (Lo) that is based on generic information that relates dose to leak size. The CR3 submittal (Reference 4) utilized plant-specific dose estimates based on the predicted level 2 analysis results.

The HNP PSA (Reference 8) contains the necessary information to convert the plant damage classes to release categories. The plant damage classes are mapped to one of the fourteen release categories. In addition, the fraction of intact containment cases is determined using the split fraction information documented in Reference 8.

Since the HNP PSA contains the necessary release fraction information, an approach similar to the CR3 submittal is utilized that better reflects the specific release conditions for HNP. The HNP PSA (Reference 8) release categories are defined by the release fraction of major radionuclides. These are extrapolated to dose using the approach presented in Appendix A with the exception of the intact containment dose. The intact containment dose is based on the licensing design basis leakage rate and is developed in Reference 11. The release category dose information is presented in Table 3.

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ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Tablc 3 Release Category Radionuclide Percentage Release and Total Person-Rcm iRelease Noble Gas' Iodine' Cesium' Tellurium' Strontium' Person-.-

.Category Frequency

(%) :

()

(%)

(%j Rem.

IC-I 1.35E-5 NA2 NA NA NA NA 3.26E+33 RC-I 3.57E-7 100 0.162 0.553 0

I.9E-5 1.86E+6 RC-IA 1.03E-8 100 1.8E-4 1.7E-4 3.7E-6 6.5E-8 1.50E+6 RC-1B 6.1 1E-7 100 1.27 1.89 0

1.7E-5 3.08E+6 RC-IBA 5.07E-8 100 1.8E-4 1.7E-4 7.8E-2 2.11E-2 1.51E+6 RC-2 2.52E-8 100 0.846 1

0 1.1E-5 2.42E+6 RC-2B 6.53E-8 100 6.62 5.5 3.3E-2 3.8E-3 7.56E+6 RC-3 2.27E-7 100 9.2E-2 8.9E-2 1.6E-6 5.3E-5 1.59E+6 RC-3B 3.79E-8 100 0.185 0.186 0

3.3E-5 1.69E+6 RC4 0

100 1.94 1.62 0

3.6E-5 3.28E+6 RC-4C 5.05E-8 100 2.41 2.13 0

2.9E-4 3.77E+6 RC-5 0

100 77.5 80.8 0

10 8.12E+7 RC-5C 3.78E-6 100 77.5 80.8 0

10 8.12E+7 RC-6 1.33E-6 100 0.021 0.063 7.8E-3 2.1E-3 1.54E+6 RC-7 4.53E-6 100 0.21 0.63 7.8E-2 2.1 E-2 1.92E+6

1. Contributing fission product groups are discussed in Appendix A.

2. Release fractions not necessary for this calculation.

3. Intact containment representing design basis leakage (developed in Reference 11).

Other inputs to this calculation include ILRT test data from NUREG-149312 and the EPRI report (Reference 6). They are referenced in the body of the calculation where it is appropriate.

3.0 ASSUMPTIONS

1. The maximum containment leakage for EPRI Class I (Reference 6) sequences is I La (Type A acceptable leakage) because a new Class 3 has been added to account for increased leakage due to a change in Type A inspection frequency.

2. The maximum containment leakage for Class 3a (References 2 and 4) sequences is 10 La based on the previously approved methodology (References 2 and 3).

3. The maximum containment leakage for Class 3b sequences is 35 L4 based on the previously approved methodology (References 2 and 3).

4. Class 3b is conservatively categorized as a LERF leakage rate based on the previously approved methodology (References 2 and 3).

5. Containment leakage due to EPRI Classes 4 and 5 are considered negligible based on the RSC 04-03 7

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ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension previously approved methodology (References 2 and 3).

6. The containment releases do not vary significantly with time.

7. The containment releases for EPRI 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 to or greater than those represented by Type A leakage.

8. Because EPRI Class 8 sequences are containment bypass sequences, potential releases progress directly to the environment. Therefore, the status of the containment structure will not impact the release magnitude of this class.

9. Several release categories contribute to the person-rem estimates for Classes 6, 7 and 8.

The presented person-rem value for these classes is based on a frequency weighted average of each release category person-rem.

4.0 CALCULATIONS This calculation applies the HNP PSA release category information in terms of frequency and person-rem to estimate the changes in risk due to increasing the ILRT test interval. The changes in risk are assessed consistent with the previously approved methodology used by Indian Point 3 (References 2 and 3) and Crystal River 3 (Reference 4).

This approach is similar to that presented in EPRI TR-104285 (Reference 6) and NUREG-1493 (Reference 12). Namely, the analysis performed examined the HNP PSA plant specific results in which the containment integrity remains intact or the containment is impaired. The detailed calculations performed to support this report were of a level of mathematical significance necessary to calculate the results

-recorded. However, the tables and illustrational calculation steps presented may present rounded values to support readability.

4.1 CALCULATIONAL STEPS The analysis is based on guidance provided in Reference 6 and uses risk metrics presented in Reference 7 to evaluate the impact of a proposed change on plant risk. References 2 and 4 utilize several measures in their assessments. These measures are change in release frequency, change in risk as defined by the change in person-rem, the change in LERF and the change in the conditional containment failure probability.

Reference 7 also lists the change in core damage frequency as a measure to be considered. Since the testing addresses the ability of the containment to maintain its function, the proposed change has no measurable impact on core damage frequency. Therefore, this attribute remains constant and has no risk significance in this evaluation.

The overall process is outlined below:

Define baseline plant damage classes and person-rem estimates

Calculate baseline Type A leakage estimate to define the analysis baseline

Modify Type A leakage estimate to address extension of the Type A test frequency RSC 04-03 8

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ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Compare analysis metrics to estimate the impact and significance of the increase related to those metrics The first step in the analysis is to define the baseline plant damage classes and person-rem dose measures.

Plant damage state information is developed using the HNP PSA (Reference 8) results. The plant damage state information and the results of the containment analysis are used to define the sequences. The population person-rem dose estimates for each key plant damage classes are based on the application of the method described in Appendix A and design basis information (Reference 11).

The product of the person-rem for the plant damage classes and the frequency of the plant damage state is used to estimate the annual person-rem for the plant damage state. Summing these estimates produces the annual person-rem dose based on the sequences defined in the PSA.

The PSA plant damage state definitions consider isolation failures due to Type B and Type C faults and examine containment challenges occurring after core damage and/or reactor vessel failure. These sequences are grouped into plant damage classes. Bypass, isolation failures and phenomena-related containment failures are identified using the plant damage state information.

Once identified, the sequence is classified by release category definitions specified in Reference

6. With this information developed, the PSA baseline model is completed.

The second step expands the baseline model to address Type A leakage.

The PSA did not directly address Type A (liner-related) faults and this contribution must be added.to provide a complete baseline. In order to define leakage that can be linked directly to the Type A testing, it is important to limit the assessed failures to those exclusively identified by Type A testing.

Reference 6 provides the estimate for the probability of a leakage contribution that could only be identified by Type A testing based on industry experience. This probability is then used to adjust the intact containment category of the HNP PSA to develop a baseline model including Type A leakage events.

The release, in terms of person-rem, is developed based on information contained in Reference 6 and is estimated as a leakage increase relative to allowable release La defined as part of the ILRT.

The predicted probability of Type A leakage is then modified to address the expanded time between testing. This is accomplished by a ratio of the existing testing interval and the proposed test interval. This assumes a constant failure rate and that the failures are randomly dispersed during the interval between the test.

The change due to the expanded interval is calculated and reported in terms of the change in release due to the expanded testing interval, the change in the population person-rem and the change in large early release frequency.

The change in the conditional containment failure probability is also developed. From these comparisons, a conclusion is drawn as to the risk significance of the proposed change.

Using this process, the following were performed:

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ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension

1. Map the release categories into the 8 release classes defined by the EPRI Report (Reference 6).

2. Calculate the Type A leakage estimate to define the analysis baseline.

3. Calculate the Type A leakage estimate to address the current inspection frequency.

4. Modify the Type A leakage estimates to address extension of the Type A test interval.

5. Calculate increase in risk due to extending Type A inspection intervals.

6. Estimate the change in LERF due to the Type A testing.

7. Estimate the change in conditional containment failure probability due to the Type A testing.

4.2 SUPPORTING CALCULATIONS Step 1: Map the Level 3 release categories into the 8 release classes defined by the EPRI Report EPRI Report TR-104285 (Reference 6) defines eight (8) release classes as presented in Table 4.

Table 4 Containment Failure Classifications (from Reference 6)

'Failure Classification Description Intcrpretation for Assigning HNP Release Category I

Containment remains intact with Intact containment bins containment initially isolated 2

Dependent failure modes or common Isolation faults that are related to a loss of cause failures power or other isolation failure mode that is not a direct failure of an isolation component 3

Independent containment isolation Isolation failures identified by Type A testing failures due to Type A related failures 4

Independent containment isolation Isolation failures identified by Type B testing failures due to Type B related failures 5

Independent containment isolation Isolation failures identified by Type C testing failures due to Type C related failures 6

Other penetration failures Other faults not previously identified 7

Induced by severe accident phenomena Early containment failure sequences as a result of hydrogen bum or other early phenomena 8

Bypass Bypass sequence or SGTR RSC 04-03 10 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Table 5 presents the HNP release category mapping for these eight accident classes. Person-rem per year is the product of the frequency of occurrence and the associated person-rem.

Table 5 HNP PSA RC Grouping to EPRI Classes (Reference 6)

-Person-Class Description Release Category Frequency Person-Rem Reim'yr I

No containment failure IC-1 1.35E-5 3.26E+3 4.414E-2 2

Large containment None 0.OOE+O' 0.OOE+0 O.OOE+0 isolation failures 3a Small isolation failures None O.OOE+0 3.26E+4 O.OOE+0 (liner breach) 3b Large isolation failures None O.OOE+0 1.14E+5 O.OOE+O (liner breach) 4 Small isolation failures -

None O.OOE+O O.OOE+O O.OOE+0 failure to seal (type B)

Small isolation failures-O.OOE+O O.OOE+O 0.OOE+O failure to seal (type C)

Containment isolation 6

failures (dependent failure, RC-3, RC-3B 2.64E-7 1.60E+062 4.241E-1 personnel errors)

Severe accident 7

phenomena induced failure All other RCs 6.88E-6 1.95E+06 2 1.340E+1 (Early and Late)

RC-4, RC-4C, 8

Containment bypass RC-5, RC-5C, 3.92E-6 7.84E+072 3.072E+2 RC-2, RC-2B Total 2.46E-5

-3.2102E+02

1. No results. Contributions are below quantification truncation value.

2. Value represents a frequency weighted average of contributing release category source terms.

Step 2: Calculate tde Type A leakage estimate to define the analysis baseline (3 year test interval)

RSC 04-03 11 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension As displayed in Table 5 the HNP PSA did not identify any release categories specifically associated with EPRI Classes 2, 3, 4, or 5. Therefore each of these classes must be evaluated for applicability to this study.

Class 3:

Containment failures in this class are due to leaks such as liner breaches that could only be detected by performing a Type A ILRT.

Reference 3 states that a review of experience data finds that Type A testing identified only 4 leakage events of the 144 events identified. Thus about 3% (0.028) of containment leakage events are identified by the ILRT. The remaining events were identified by LLRT (Type B and C testing) and are not included in the analysis.

This probability, however, is based on conducting three tests over a 10-year period. One test per ten-years is the interval currently employed at HNP (Reference 1) and the probability (0.028) must be adjusted to reflect this difference.

The probability of liner failure is divided into two size ranges to maintain an approach consistent with the previously approved methodology (References 2 and 3). Class 3a is defined as a small liner breach and Class 3b is defined as a large liner breach.

Calculation of Class 3b Probability To calculate the probability that a liner leak will be large (Class 3b), use was made of the data presented in NUREG-1493 (Reference 12). One data set found iri NUREG-1493 reviewed 144 ILRTs. The largest reported leak rate from those 144 tests was 21 times the allowable leakage rate (La). Since a leakage of 21 La does not constitute a large release, no large releases have.

occurred based on the 144 ILRTs reported in NUREG-1493.

To estimate the failure probability given that no failures have occurred, a conservative estimate is obtained from the 95th percentile of the x2 distribution. This is consistent with the Indian Point 3 (Reference 2) and Crystal River 3 (Reference 4) templates.

In statistical theory, the x2 distribution can be used for statistical testing, goodness-of-fit tests (See Reference 13). The x2 distribution is really a family of distributions, which range in shape from that of the exponential to that of the normal distribution.

Each distribution is identified by the degrees of freedom, v. For time-truncated tests (versus failure-truncated tests), an estimate of the probability of a large leak using the x2 distribution can be calculated using the following equation:

_ 2 (2F +2, a) p(a) =

2N (Eq. 1) where: N is the number of events, F is the number of events (faults) of interest, a is the percentile distribution (typically assumed to be the 95%-tile). The result of 2F+2 defines the degree of freedom.

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ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Given that there have been no large leaks (n = 0, therefore v =2) in 144 events (N = 144) the value of X2(2, 0.05) is equal to 5.99. Solving for the 95th percentile estimate of the probability of a large leak yields approximately 0.021 as presented below:

z2(20 05) 99 PCIX3 2 (,0

= 0.0208 (Eq. 2) 2.144 288 Calculation of Class 3a Probability The data presented in NUREG-1493 (Reference 12) is also used to calculate the probability that a liner leak will be small (Class 3a). It is reported in the report that 23 of 144 tests had allowable leak rates in excess of 0L.a However, of the 23 events that exceeded the test requirements, only 4 were found by an ILRT, the others were found by Type B and Type C testing or were due to errors in the testing alignment.

Therefore, a best estimate for the probability of leakage is -0.03 (4-in-144).

The Class 3a probability is estimated using the conservative X2 distribution approach described previously.

This is consistent with the approach taken in References 2, 3 and 4.

The X2 distribution is calculated by F=4 (number of small leaks) and N=144 (number of events) which yields a solution as shown below:

X 2(10,0.05) 18,307 PCIass3A = ~2(,.5 837= 0.0636 (Eq. 3) 2.144 288 Therefore, the 95th.percentile estimate of the probability of a small leak (Class 3a) is calculated as 0.064.

The probability of liner failures must then be multiplied by an appropriate accident frequency to determine the Class 3a and Class 3b frequencies. The IP3 (Reference 2) and CR3 (Reference 4) submittals utilized the entire core damage frequency when developing the contributions for Classes 3a and 3b and then adjusted the Class 1 contribution.

This is somewhat conservative since it maximizes the possible contributions due to the extension of the ILRT testing interval by including other large release sequences such as containment bypass and late overpressure. This approach is maintained for the HNP analysis, in order to be consistent with the approved methodology.

Therefore the frequency of a Class 3b failure is calculated as:

FREQclass3b = PROB,,...b x CDF = 0.0208 x 2.46E-5/yr = 5.12E-7/yr (Eq. 4)

Therefore the frequency of a Class 3a failure is calculated as:

FREQciass3a = PROBL,,.

x CDF = 0.0636 x 2.46E-5/yr = 1.56E-6/yr (Eq. 5)

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A1TACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Class 4:

This group consists of all core damage accident accidents for which a failure-to-seal containment isolation failure of Type B test components occurs. By definition, these failures are dependent on Type B testing, and Type A testing will not impact the probability. Therefore this group is not evaluated any further, consistent with the approved methodology.

Class 5:

This group consists of all core damage accident accidents for which a failure-to-seal containment isolation failure of Type C test components occurs. By definition, these failures are dependent on Type C testing, and Type A testing will not impact the probability. Therefore this group is not evaluated any further, consistent with the approved methodology.

Class 6:

The Class 6 group is comprised of isolation faults that occur as a result of the accident sequence progression. The leakage rate is not considered large by the PSA definition and therefore it is placed into Class 6 to represent a small isolation failure and identified in Table 6 as Class 6.

FREQCLsS6 =2.64E-7/yr Class 1:

Although Type A testing does not directly impact the frequency of this class, the PSA did not model Class 3 failures, and the frequency for Class I should be reduced by the.estimated frequencies in the new Class 3a and Class 3b in order to preserve the total CDF. The revised Class 1 frequency is therefore:

FREQcIass, = FREQciassl - (FREQclass3a + FREQclass3b)

(Eq. 6a):

FREQ.I~y 1 = 1.35E-5/yr - (1.56E-6/yr + 5.12E-7/yr) 1.146E-5/yr (Eq. 6b)

Class 2:

The Table 6 does not identify any contribution to Class 2.

Class 7:

The frequency of Class 7 is the sum of those release categories identified in Table 6 as Class 7.

FREQClSS7 = 6.88E-6/yr Class 8:

The frequency of Class 8 is the sum of those release categories identified in Table 6 as Class 8.

FREQCLIsS8 = 3.92E-6/yr RSC 04-03 14 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Table 6 summarizes the above information by the EPRI defined classes. This table also presents dose exposures calculated using the methodology described in Appendix A. For Class 1, 3a and 3b, the person-rem is developed based on the design basis assessment of the intact containment (Reference 11). The Class 3a and 3b doses are represented as 1OL, and 35La respectively.

Table 6 also presents the person-rem frequency data determined by multiplying the failure class frequency by the corresponding dose exposure.

Table 6 Baseline Risk Profile (three tests per ten years)

.Class -

Description-Fiequency.

Person-rem.

Person-rem Person-rem

(/r#(rom

.(rom La fr calculation) factors)

No containment failure 1.15E-5 3.26E+3 2 3.737E-2 2

Large containment O.OOE+O NA O.OOE+O O.OOE+O isolation failures 3a Small isolation failures 1.56E-6 3.26E+43 5.098E-2 (liner breach) 3b Large isolation failures 5l2E7

.14E+5 4

5.840E-2 (liner breach)

Small isolation failures-O.OOE-O failure to seal (type B) 5 Small isolation failures -

O.OOE+O NA O.OOE+O O.OOE+O failure to seal (type C)

Containment isolation 6

failures (dependent failure, 2.64E-7 1.60E+65 4.24 1E-I personnel errors)

Severe accident 7

phenomena induced failure 6.88E-6 1.95E+6 5 1.340E+1 (early and late)

Containment bypass 3.92E-6 7.84E+75 3.072E+2 Total 2.46E-5 3.21127E+2 I.

2.

3.

4.

5.

From 1 able i using the metnod presented in Appenaix A.

I La dose value calculated in Reference 11.

10 times. L,.

35 times La.

Value represents a frequency weighted average of contributing release category source terms.

RSC 04-03 15 Printed 4/6t2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significancc of ILRT Extension The percent risk contribution due to Type A testing is as follows:

%Risk.,sE =[( Class3a.E + Class3b.E) / TotalsE] x 100 (Eq. 7a)

Where:

Class3a.sE = Class 3a person-rem/year = 5.098E-2 person-rem/year Class3bas, = Class 3b person-rem/year = 5.840E-2 person-rem/year Total., = total person-rem year for baseline interval = 3.21127E+2 person-rem/year (Table 6)

%Risk.E = [(5.098E-2 + 5.840E-2) / 3.21127E+2] x 100 = 0.034%

(Eq. 7b)

Step 3: Calcuilate the Tjpe A leakage estimate to address the current inspection interval The current surveillance testing requirements as proposed in NEI 94-01 (Reference 5) for Type A testing and allowed by 10 CFR 50, Appendix J 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 L.OLa).

According to NUREG-1493 (Reference 12), extending the Type A ILRT interval from 3-in-10 ycars 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. Multiplying the testing interval by 0.5 and multiplying by 12 to convert from "years" to "months" calculates the average time for an undetected condition to exist.

Since ILRTs only detect about 3% of leaks (4/144) that arc not detected by other local tests, the increase for a 10-yr ILRT interval is the ratio of the average time for a failure to detect for the increased ILRT test interval (60 months) to the baseline average time for a failure to detect of 18 months (i.e., 0.03 x 60/18 = 0.10). References 2 and 4 indicate this is a 10% increase in the likelihood of a Type A leak.

Risk Impact due to 10-year Test Interval Based on the previously approved methodology (References 2 and 3), the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences. Consistent with Reference 2 and 4 the risk contribution is determined by multiplying the Class 3 accident frequency by the increase in the probability of leakage (1.1 x Class 3 baseline). The results of this calculation are presented in Table 7 below.

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ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of 1LRT Extension Table 7 Current Risk Profile (one test per ten years)

Class Description Frequency (/yr)

Person-rem Person-rem (/yr)

I No containment failure' 1.13E-5 3.26E+3 3.669E-2 2

Large containment isolation O.OOE+O O.OOE+O O.OOE+O failures 3a Small isolation failures (liner 1.72E-6 3.26E+4 5.608E-2 breach) 3b Large isolation failures (liner 5.63E-7 1.14E+5 6.424E-2 breach) 4 Small isolation failures - failure O.OOE+O O.OOE+O O.OOE+O to scal (type B)

Small isolation failures - failure O.OOE+O O.OOE+O O.OOE+O to seal (type C)

Containment isolation failures 6

(dependent failure. personnel 2.64E-7 1.60E+6 4.241E-1 errors) 7 Severe accident phenomena 6.88E-6 1.95E+6 1.340E+l i

induced failure (early and late) 8 Containment bypass 3.92E-6 7.84E+7 3.072E+2 Total 2.46E-S 3.21137E+2

1. The PSA frequency of Class I has been reduced by the amount of increase in the frequency of Class 3a and Class 3b in order to preserve total CDF.

2. Refer to notes for Table 6.

Using the same methods as for the baseline and the data in Table 7 the percent risk contribution due to Type A testing is as follows:

%Riskio =[(Class3a,0 + Class3b,0) / Total30] x 100 (Eq. 8a)

Where:

Class3a10 = Class 3a person-rem/year = 5.608E-2 person-rem/year Class3b,0 = Class 3b person-rem/year = 6.424E-2 person-rem/year Total,0 = total person-rem year for current 10-year interval = 3.21137E+2 person-rem/year (Table 7)

%Risk,. = [(5.608E-2 + 6.424E-2) / 3.21137E+2] x 100 = 0.037%

(Eq. 8b)

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ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension The percent risk increase (A%Risk,.) due to a ten-year ILRT over the baseline case is as follows:

A%Risk,. = [(Total,. - TotalB.sE) / Total.,SE] x 100.0 (Eq. 9a)

Where:

TotalbE = total person-rem/year for baseline interval = 3.21127+2 person-rem/year (Table 6)

Total, = total person-rem/year for 10-year interval = 3.21137E+2 person-rem/year (Table 7)

A%Risk, 0 = [(3.21137E+2 - 3.21127E+2) / 3.21127E+2] x 100.0 = 0.0032%

(Eq. 9b)

Step 4: Calculate the Type A leakage estimate to address extended inspection intervals If the test interval is extended to I in 15 years, the average time that a leak detectable only by an ILRT test goes undetected increases to 90 months (0.5 x 15 x 12). For a 15-yr-test interval, the result is the ratio (0.03 x 90/18) of the exposure times. Thus, increasing the ILRT test interval from 10 years to 15 years results in a proportional increase in the overall probability of leakage.

The approach for developing the risk contribution for a 15-year interval is the same as that for the 10-year interval. References 2 and 4 indicate that the increase is a 50% increase from that for the 10-year interval or a 15% increase from the baseline. Different values are provided for the probability of leakage. In addition, the containment leakage used for the 10-year test interval for Class 3 is used in the 15-year interval evaluation (1.15 x Class 3 baseline). The results for this calculation are presented in Table 8.

RSC 04-03 18 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Table 8 Proposed Risk Profile (once per fifteen years)

Class Description Frequency (/yr).

Person-rem 2 Person-rem (/yr)

No containment failure' 1.12E-5 3.26E+3 3.635E-2 2

Large containment isolation O.OOE+0 O.OOE+O O.OOE+O failures 3a Small isolation failures (liner 1.80E-6 3.26E+4 5.863E-2 3a breach) 3b Large isolation failures (liner 5.89E-7 1.14E+5 6.716E-2 breach) 4 Small isolation failures - failure O.OOE+0 O.OOE+O O.OOE+0 to seal (type B) 5 Small isolation failures - failure O.OOE+0 O.OOE+O O.OOE+0 to seal (type C)

Containment isolation failures 2.64E-7 1.60E4 6 4.241 E-l 6

(dependent failure, personnel errors)

Severe accident phenomena 6.88E-6 1.95E+6 1.340E+*

induced failure (early and late) 8 Containment bypass 3.92E-6 7.84E+7 3.072E+2 Total 2.46E-5 3.21142E+2

1. The PSA frequency of Class I has been reduced by the amount of increase in the frequency of Class 3a and Class 3b in order to preserve total CDF.

2. Refer to notes for Table 6.

Using the same methods as for the baseline, and the data in Table 10 the percent risk contribution due to Type A testing is as follows:

%Risk,, =[( Class3a,5 + Class3b1,) / Total,,] x 100 (Eq. lOa)

Where:

Class3a., = Class 3a person-rem/year= 5.863E-2 person-rem/year Class3b, = Class 3b person-rem/year = 6.716E-2 person-rem/year Totalis = total person-rem year for 15-year interval = 3.21142E+2 person-rem/year (Table 8)

%Risk., = [(5.863E-2 + 6.716E-2) / 3.21142E+2] x 100 = 0.039%

(Eq. lOb)

RSC 04-03 19 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension The percent risk increase (A%Risk,,) due to a fifteen-year ILRT over the baseline case is as follows:

A%Risk,, = [(Totalb, - TotalBASE) / Totals,] x 100.0 (Eq. II a)

Where:

TotalBASE = total person-rem/year for baseline interval = 3.21127E+2 person-rem/year (Table 6)

Totals = total person-rcm/year for 15-year interval = 3.21142E+2 person-rem/year (Table 8)

A%Risk1, = [(3.21142E+2 - 3.21127E+2) / 3.21127E+2] x 100.0 = 0.0048%

(Eq. I lb)

Step 5: Calcuelate increase in risk due to extending Type A inspection intervals Based on the previously approved methodology (Reference 2 and 4), the percent increase in risk (in terms of person-rem/yr) of these associated specific sequences is computed as follows.

%Risk,.,, =[(PER-REM,, - PER-REM,.) / PER-REM,.] x 100 (Eq. 12a)

Where:

PER-REM,. = person-rem/year of ten years interval (see Table 7, classes 1, 3a and 3b)

= 1.570E-1 person-rem/yr PER-REM1,= person-rem/year of fifteen years interval (Table 8, classes 1, 3a and 3b)

= 1.621E-1 person-'rem/yr

%Risk,..l, = [(1.621E 1.570E-1) / 1.570E-1] x 100 = 3.27%

(Eq. 12b)

The percent increase on the total integrated plant risk for these accident sequences is computed as follows.

%TotalIo-1s = [(Total,, - Total,.) /Total 1.] x 100 (Eq. 13a)

Where:

Totalio = total person-rem/year for 10-year interval

= 3.21137E+2 person-rem/year (Table 7)

TotalI5 = total person-remn/year for 15-year interval

= 3.21142E+2 person-rem/year (Table 8)

% Totalho-is = [(3.21142E+2 - 3.21137E+2) /3.21137E+2] x 100 = 0.0016%

(Eq. 13b)

RSC 04-03 20 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Step 6: Calculate the change in Risk in terms of Large Early Release Frequency (LERF)

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 a larger release due to failure to detect a preexisting leak during the relaxation period.

From References 2 and 4, the Class 3a dose is assumed to be 10 times the allowable intact containment leakage, La (or 32,600 person-rem) and the Class 3b dose is assumed to be 35 times La (or 11,400 person-rem).

The dose equivalent for allowable leakage (La) is developed in Reference 11. This compares to a historical observed average of twice La. Therefore, the estimate is somewhat conservative.

Based on the previously approved methodology (References 2 and 4), only Class 3 sequences have the potential to result in large releases if a preexisting leak were present.

Class 1 sequences are not considered as potential large release pathways because for these sequences the containment remains intact. Therefore, the containment leak rate is expected to be small (less than 2La). A larger leak rate would imply an impaired containment, such as Classes 2, 3, 6 and 7.

Late releases are excluded regardless of the size of the leak because late releases are, by definition, not a LERF event. At the same time, sequences in the HNP PSA (Reference 8),

resulting in large releases, are not impacted because a LERF will occur regardless of the presence of a pre-existing leak. Therefore, the change in the frequency of Class 3b. sequences is used as the increase in LERF for HNP, and the change in LERF can be determined by the differences. References 2 and 4 identify that Class 3b is considered to be the contributor to LERF. Table 9 summarizes the results of the LERF evaluation that Class 3b is indicative of a LERF sequence.

Table 9 Impact on LERF due to Extended Type A Testing Intervals TILRT Inspection Interval Years (baseline) 10Years 15 Years--

Class 3b (Type A LERF) 5.12E-7/yr 5.63E-7/yr 5.89E-7/yr ALERF (IO year baseline) 2.56E-8/yr ALERF (3 year baseline) 7.68E-8/yr RG 1.174 (Reference 7) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. The RG 1.174 guidance defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 1E-6/yr and increases in LERF below IE-7/yr. Since the ILRT does not impact CDF, the relevant metric is LERF. Calculating the increase in LERF requires determining the impact of the ILRT interval on the leakage probability.

RSC 04-03 21 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Since guidance in Reg. Guide 1.174 defines very small changes in LERF as below 1.OE-7/yr, increasing the ILRT interval to 15 years (2.56E-8/yr) is non-risk significant. It should be noted that if the risk increase is measured from the original 3-in-10-year interval, the increase in LERF is 7.68E-8/yr, which is also below the l.OE-7/yr screening criterion in RG 1. 174.

Step 7: Calculate the change in Conditional Containment Failuire Probability (CCFP)

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

CCFP = I [-

rf(nf)]

(Eq. 14)

Where f(nc]) is the frequency of those sequences which result in no containment failure. This frequency is determined by summing the Class I and Class 3a results, and CDF is the total frequency of all core damage sequences.

Therefore the change in CCFP for this analysis is the CCFP using the results for 15 years (CCFP,,) minus the CCFP using the results for 10 years (CCFPI). This can be expressed by the following:

ACCFP015 = CCFP,5 - CCFP 1 o (Eq. 15)

Using the data previously developed the change in CCFP from the current testing interval is calculated and presented in Table 10.

Table 10 Impact on Conditional Containment Failure Probability due to Extended Type A Testing Intervals

]LRTlIispectionInterval 3 Y (baselin 10 Years 15 Years ftncf) (/yr) 1.303E-5 1.298E-5 1.295E-5 J(ncf)/CDF 0.5295 0.5274 0.5264 CCFP 0.4705 0.4726 0.4736 ACCFP (3 year baseline) 0.208%

0.312%

ACCFP (10 year baseline) 0.104%

4.3 ALTERNATIVE LERF CALCULATION USING REFINED DATA The LERF development using the guidance in Reference 2 assumes that all sequences would be impacted by the potential for Type A leakage. However, the presence or absence of a Type A RSC 04-03 22 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension leak would not significantly alter those accident sequences that are already LERF contributors.

This alternative calculation excludes existing LERF sequences from the estimation of Type A leakage and presents the results of this refined assessment.

From Table 6 the baseline Class 3b frequency is 5.12E-7/yr and is derived by multiplying the total CDF by the probability of Type A leakage. The HNP analysis already includes LERF contributions from both interfacing systems LOCA and steam generator tube rupture events.

From Reference 8, their contributions to LERF are captured in release categories RCA, RC-4C, RC-5 and RC-5C. Early containment failures are also taken from RC-2 and RC-2B.

All of these contributions are currently considered LERF sequences and their contributions are listed in Table 11.

Table 11 LERF Release Category Contributions

- Release Category Frequency Qyr)

LERF Cumulative Frequency (/yr)

RCA4 l

l RC-4C 5.05E-8 5.05E-8 RC-5 E

5.05E-8 RC-5C 3.78E-6 3.83E-6 RC-2 2.52E-8 3.85E-6 RC-2B 6.53E-8 1

3.92E-6

1. Contribution is less than truncation.

These release categories arc considered LERF contributors regardless of the potential for Type A leakage and can be excluded from the baseline CDF as presented in Equation 16.

Adjusted CDF = 2.46E-5/yr - 3.92E-6/yr = 2.07E-5/yr (Eq. 16)

The Class 3b frequency is then determined by the following equation:

FREQc.ass3b = PROBI.c.,b x CDF = 0.0208 x 2.07E-5/yr = 4.30E-7/yr (Eq. 17)

This can then be extrapolated using the methods presented earlier to determine the 10-year and 15-year contributions and to generate adjusted LERF values as presented in Table 12.

RSC 04-03 23 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Table 12 Class 3b Contributions Using Adjusted CDF Interval Frequency (/yr)

Delta Frequency from Prior Period

(/yr)

Baseline 4.30E-7 10-year (current)

.4.73E-7 4.30E-8 15-year 4.95E-7 2.15E-8 As the results indicate, the value still are below the threshold for a significant change in LERF and the inclusion of contributions related to existing LERF sequences does not impact the analysis conclusions.

4.4 REVISED LERF BASED ON INTACT CONTAINMENT Reference 14 utilized only the intact containment contribution when calculating the Class 3 contributions.

The analysis was based on the fact that the overall dose for any other case involving an impaired containment would be bounded'by the existing dose rate such that the predicted Type A dose would be inconsequential. This is supported when the predicted doses are compared. Table 13 provides a comparison of the Type 3b dose to the Class 7 and Class 8 d6ses.

I Table 13 Comparison of Class 3b, Class 7 and Class 8 Population Doses

. Case Population Dose' Normalized Dose to Class 3b

rson-rein)

Class 3b 1.14E+5 1.0 Class 7 1.95E+6 17.0 Class 8 7.84E+7 688.0 As the table indicates, the expected dose from the impaired containment is at least 17 times that from the Class 3b release. Only the intact containment case would experience an increase in consequence due to Type A leakage.

If this approach is utilized, the probability of Class 3b leakage (0.0208) is multiplied by the intact containment contribution (1.35E-5/yr) to generate the Class 3b frequency. This value is then extrapolated to the 10-year and 15-year cases using previously described multiplication factors.

The results are then compared using the methods described earlier for determining the change in LERF. The results are summarized in Table 14.

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ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Table 14 Class 3b Contributions Using Intact Containment Component Only Interval Frequency (/yr)

Delta Frequency from Prior Period

(/yr)

Baseline 2.816E-07 10-year (current) 3.098E-07 2.817E-08 15-year 3.239E-07 1.408E-08 If only the intact containment contribution is considered, the results indicate that the LERF frequency increases for both the baseline (3 tests per 10 years) and the current interval (1 test per 10 years) cases. However, neither case exceeds the numeric requirement from Reference 7. This conclusion again supports the extension of the ILRT testing interval.

4.5 REVISED TYPE A LERF EXCLUDING PREEXISTING LERF AND NON-LERF CASES Reference 14 indicates that the estimation of the impact to LERF can be refined by excluding that frequency already defined as LERF (see Section 4.3) and the frequency contribution from those sequences that would not result in a large early release regardless of the presence of a Type A failure due to scrubbing provided by containment sprays.

The LERF contribution was presented in Table 11 as 3.92E-6/yr.

This calculation examines the additional benefit of removing accident sequences that would have adequate scrubbing to preclude LERF releases.

The assessment assumes that containment spray must be available for both injection and recirculation to ensure scrubbing of released radionuclides that are released during initial reactor vessel failure and subsequent releases from radionuclides released from the RCS after vessel failure. The end state (plant damage state) must also be an intact containment state since the unisolated containment states are already considered by the LERF fraction.

A review of the HNP containment safeguards event tree (CSET) identifies that this condition can be met for PDS states "A", "G" and "D" only.

PDSs with endstates "A" and "G" are associated with continuous operation of the containment sprays. PDS "D" is representative of a sequence where containment spray recirculation fails.

The contribution to this sequence predominantly involves the loss of heat removal to the sprays and a failure of pressure control. The containment spray would continue to function based on analysis developed for the PSA (Reference 8) and could maintain the scrubbing function.

Containment pressure control would be maintained by the containment fan coolers.

Therefore, the assumption is made those contributions to PDS "D" can also be excluded from the calculation of LERF since containment spray would be available to scrub any release. Table 15 lists the contributing PDSs that are from one of the three PDSs identified above.

RSC 04-03 25 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Table 15 Contributing PDSs with CSET States A, D or G PDS Frequency (/yr) CDF Contribution 2A 3.04E-6 12.4%

4A 2.52E-6 22.6%

7A 9.78E-7 26.6%

17A 5.68E-7 28.9%

13A 4.35E-7 30.7%

15A 3.85E-7 32.2%

8G 2.34E-7 33.2%

IA 1.67E-7 33.8%

3A 1.52E-7 34.5%

5A 1.34E-7 35.0%

12A 1.17E-7 35.5%

lOG 9.51E-8 35.9%

4D 8.1 OE-8 36.2%

14A 7.04E-8 36.5%

2D 6.25E-8 36.7%

1OA 5.81 E-8 37.0%

8A 5.78E-8 37.2%

2G 4.67E-8 37.4%

RSC 04-03 26 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Table 15 (continued)

Contributing PDSs with CSET States A, D or G PDS Frequency (/yr). CDF Contribution 4G 3.46E-8 37.5%

IA 3.21E-8 37.7%

3G 2.82E-8 37.8%

9A 1.62E-8 37.9%

7D 6.84E-9 37.9%

17D 7.04E-10 37.9%

15D 4.13E-10 37.9%

13D 1.06E-10 37.9%

The first ten PDS contributions in Table 15 represent 35% of the total CDF. Since all involve either state "A" or "G", the operation of the containment sprays is assured and encompassed system failures would not result in a release of radionuclides sufficient to be classified as LERF.

The total frequency of PDSs that will not result in LERF release is 9.32E-6/yr. The adjusted CDF for calculating the Class 3b frequency is then calculated as shown in Equation 18 by removing both the LERF sequences and those PDSs that cannot result in a LERF release given a Type A failure.

Adjusted CDF = 2.46E-5/yr - (3.92E-6/yr + 9.32E-6/yr) = 1.14E-5/yr (Eq. 18)

The Class 3b frequency calculation is presented in Equation 19:

FREQc~aSS3b = PROBC,,-b x CDF = 0.0208 x 1.14E-5/yr = 2.36E-7/yr (Eq. 19)

The LERF result is then extrapolated to 10 years and 15 years and the differences calculated in Table 16.

RSC 04-03 27 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension Table 16 Class 3b Contributions Using Adjusted CDF Interval Frequency (/yr)

Delta Frequency from Prior Period(/yr)

Baseline 2.364E-7 10-year (current) 2.601E-7 2.364E-8 15-year 2.719E-7 1.18213-8 As the table indicates, the change from the current testing interval (10 years) to the proposed interval (15 years) meets the 1.OE-7/yr criterion for an insignificant increase in risk.

The increase from the baseline case (3 years) is also below the 1.OE-7/yr criterion. The combined value is still less than the numeric criterion. This alternative calculation supports the argument that the ILRT extension does not pose a significant increase in risk.

5.0 REFERENCES

1. Shearon Harris Nuclear Power Plant. Plant Operating Manual. Volume 6. Part 6.

Engineering Surveillance Test. Periodic Containment Integrated Leak Rate Testing (TvLe A Test), Rev. 11, Progress Energy, EST-210.

2.

Indian Point 3 Nuclear Power Plant, "Supplemental Information Regarding Proposed Change to Section 6.14 of the Administrative Section of the Technical Specification",

Entergy, IPN-01-007, January 18, 2001.

3.

Indian Point Nuclear Generating Unit No.3 - Issuance of Amendment Re: Frcquency of Performance-Based Leakage Rate Testing (TAC NO. MB0178), United States Nuclear Regulatory Commission, April 17, 2001.

4.

Evaluation of Risk Significance of ILRT Extension, Revision 2, Florida Power Corporation, F-0l -0001, June 2001.

5.

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

6.

Gisclon, J. M., et al, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, Electric Power Research Institute, TR-1 04285, August 1994.

7. An Approach for Using Probabilistic Risk Assessment in Risk-Informed decisions on Plant-Specific Changes to the Licensing Basis, U.S. Nuclear Regulatory Commission (USNRC), Regulatory Guide 1.174, July 1998.

RSC 04-03 28 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension

8.

Updated Individual Plant Examination Probabilistic Safety Assessment Model, Rev 5, Progress Energy, HNP-F/PSA-OOO, November 2003.

9.

Oliver, R. E., et al, Harris Nuclear Plant Individual Plant Examination Submittal, Carolina Power and Light Company, August 1993.

10. Summitt, R., Assessment of Safety Benefit for Installation of a Generator Disconnect Switch at Robinson, Rev. 0, Ricky Summitt Consulting (RSC), Inc., RSC 98-19, June 1998.

11. Estimate of 50-mile Population Dose from Design Basis Containment Leakage Following a Core Melt Accident, Rev. 0, Progress Energy, HNP-F/PSA-0067 March 2004.

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

13. PRA Procedures Guide - A Guide to the Performance of ProbabilistiRisk Assessments for Nuclear Power Plants American Nuclear Society and the National Institute of Electrical and Electronic Engineers, NUREG/CR-2300, January 1983.

14. Summitt, R., Robinson Nuclear Plant Probabilistic Safety Assessment Risk Significance of ILRT Extension based on NEI Guidance, Rev. 0, RSC, Inc., RSC 02-12, February 2002.

RSC 04-03 29 Printed 4/6/2004

ATTACHMENT 1 TO HNP-F/PSA-0066 REV. 0 Evaluation of Risk Significance of ILRT Extension APPENDIX A Surrogate Person-Rem Methodology (RSC 01-44)

RSC 04-03 A.1 Printed 4/6/2004

CALCULATION NO.HNP-FIPSA-0066 REV. 0 ATTACHMENT 2, PAGE 1 ATTACHMENT 2 - OWNERS REVIEW OF VENDOR REPORT Design RSC 04-03 Revision 0

EVALUATION OF RISK SIGNIFICANCE OF ILRT EXTENSION The signature below of the Lead Reviewer records that:

the review indicated below has been performed by the Lead Reviewer; appropriate reviews were performed and errors/deficiencies (for all reviews performed) have been resolved and these records are included in the design package; the review was performed in accordance with EGR-NGGC-0003.

5 Design Verification Review a

Engineering Review 0

Owner's Review o Design Review O Altemate Calculation E Qualification Testing 5 Special Engineering Review_

o YES E N/A Other Records are attached.

Steven L. Mabe *LM4L, Hi k Aft PSA 04106/04 Lead Reviewer Discipline Date Item No.

Deficiency Resolution 1

Include Appendix A on Page i, Table of The appendix has been added to the table of Contents.

contents.

2 Page 1, Section 1.1, 4 paragraph, last Suggested change included.

sentence, reword to read as follows: " Several updates to the HNP Level 1 analysis and to the Level 2 Information have been incorporated since the IPE."

3 Page 3, 5. bullet, last sentence, reword as Suggested change Included.

follows: 'This calculated increase in LERF is considered very small since It falls below the RG 1.174 guidance criterion.'

4 Pages 4 and 5, Section 2.0, Table 2, for A review of the reference was made and the PDS Damage States 1 OP, 3P, and 3A, the representative sequences have been changed to description of the Representative Sequence match.

does not seem to correlate to that found in the current HNP PSA documentation in Table 7.20 in Section 7 for these Damage States (reference HNP-FIPSA-0031, Rev. 1).

I EGR-NGGC-0003 Rev 9 I

CALCULATION NO.HNP-F/PSA-0066 REV. 0 ATTACHMENT 2, PAGE 2 Design RSC 04-03 Revision 0

EVALUATION OF RISK SIGNIFICANCE OF ILRT EXTENSION 5

Page 7, Table 3, the Tellurium values for The datasheet double count has been corrected.

Release Category RC-1A, RC-1BA, RC-2B, The change has no significant impact on the RC-3, and RC-6 do not match the Excel reported analysis.

Spreadsheet values. It appears the spreadsheet values 'double-count' Te2 in the

'Calculate Dose' sheet; for example, for RC-1A, cell F16=D33+L33

=[D40+L40]+L40

=[0+3.7E-6]+3.7E-6

=7.4E-6, which is double the correct value which is shown in Table 3.

6 Page 7, Table 3, the Tellurium value for RC-7 The typographical error has been corrected. The appears it should be 7.8E-2 per Table 5 of analysis data was correct.

Section 9 (reference HNP-F/PSA-0033, Rev. 1) 7 Page 9, sentence prior to the numbered items Suggested change included.

should read: 'Using this process, the following I were performed:'

8 It is not clear how the Person-Rem values were The selected values were based on engineering arrived at for Class 6, 7, and 8 in Table 5 on judgment based on the expected doses from the page 11. For example, Class 6 includes contributors and their frequency. As a result of the Release Category RC-3 and RC-3B but the comment the analysis has been changed to reflect Person-Rem value corresponds to RC-3B a frequency-weighted value for each of the value from the Excel Spreadsheet "Calculate classes. A note has been added to the table and Dose' Tab under 'Equation Value" (cell H 14).

the calculations added to the spreadsheet. In Class 7 includes "All other RCs"; however, the addition, assumption #9 has been added to the text Person-Rem corresponds to RC-1 B in the s/s.

related to the calculation process. Although the Class 8 includes RC-4, RC-4C, RC-5, RC-5C, total change is small and does not directly impact RC-2, and RC-2B; however, the Person-Rem Class 3a and 3b, the results were also corresponds to RC-5C in the s/s. Can a better recalculated.

explanation be presented of how these Person-Rem values were determined?

9 There are several places in the report where A review of the document indicates that all cases of the allowable leakage rate is shown as La; La appear to be the character 'L" in the report. It however, many of the abbreviations look like la; may have been based on printer resolution but this although this may show up due to the printer cannot be determined.

utilized.

10 In Note 1 to Table 7 on page 17, consider Suggested change included.

wording the note as "The PSA frequency of Class 1 has been reduced by the amount of increase in the frequency of Class 3a and Class 3b in order to preserve total CDF.' This comment also applies to Note I of Table 8 on page 19.

I EGR-NGGC-0003 I

Rev 9

CALCULATION NO.HNP-F/PSA-0066 REV. 0 ATTACHMENT 2, PAGE 3 Design RSC 04-03 Revision 0

EVALUATION OF RISK SIGNIFICANCE OF ILRT EXTENSION 11 The report includes values obtained from the The level of significance provided in the report is Excel Spreadsheet provided with the report consistent with other submittals we have which contain more significant figures than the performed. However, due to the very small report itself (Table values are one example).

changes additional significance was needed in the Another example, Equation 9b on page 18 calculation to obtain results. To indicate this, the contains the Person-Rem/year values from (as following text has been added to Section 4.0:

stated immediately prior to Eq. 9b) Tables 6 and 7; however, these values are obviously

'The detailed calculations performed to support this from the Excel Spreadsheet with more report were of a level of mathematic significance significant numbers. A general statement necessary to calculate the results recorded.

should be included in the report at an However, the tables and illustrational calculation appropriate location to address the fact that steps presented may be rounded values to support Table and/or report numerical values are based readability."

on the Excel Spreadsheet values and this may cause some inconsistent results due to "rounding". For example, Table 16 shows a Delta Frequency of 1.1 8E-8 for the 15-year interval; however, if the frequency values in the table are used the Delta frequency would be 1.20E-8 [2.72E 2.60E-8].

12 In Table 14, page 25, the Baseline Frequency A value from the spreadsheet was placed into the is shown as 2.82E-7; however, the product of incorrect location in the report table. The error has 0.0208 x 1.15E-7 is 2.39E-7. If 2.39E-7 is been corrected.

correct, then the other table values of Frequency and Delta Frequency may be incorrect.

13 Page 27, 2' paragraph following Table 15, 15 Suggested change included.

sentence, considering wording as 'The total frequency of PDSs that will not result in LERF release is 9.32E-6/yr."

14 Section 5.0, page 29, item #8, should replace Suggested change included.

"date not provided' with "November 2003".

15 Section 5.0, page 29, item #10, 'Harris' should Suggested change included.

be replaced with 'Robinson" 16 Section 5.0, page 29, item #11, 'CP&L, HNP-Suggested change included.

PSA-0050, November 2001" should be replaced by"Progress Energy, HNP-F/PSA-0067, March 2004".

FORM EGR-NGGC-0003-3-5 This form is a QA Record when completed and included with a completed design package. Owner's Reviews may be processed as stand alone QA records when Owner's Review is completed.

I EGR-NGGC-0003 Rev9 9 l

CALCULATION NO.HNP-F/PSA-0066 REV. 0 ATTACHMENT 3, PAGE 1 ATTACHMENT 3 - RISK FROM CONCEALED LINER CORROSION References See main section of the calculation.

Design Inputs The evaluation of risk due to containment liner corrosion does not provide plant design basis information nor is the evaluation used to modify design outputs. Therefore, no design inputs are used.

The inputs to the evaluation are documented in the attached report and its references.

Assumptions This calculation applies an analytical method developed by Calvert Cliffs in response to an NRC question about concealed containment liner corrosion (Reference 5).

Assumptions associated with that method are provided in Reference 5. Information from the updated HNP PSA Model is used in this analysis. The PSA model and its associated assumptions are described in References 1 and 2.

The applicability of the Calvert Cliffs assumptions to the HNP analysis is discussed below:

ASSUMPTION

^

^ BASIS FOR APPLICABILITY TO HNP A. Zero basemat corrosion failures are evaluated Typical PRA assumption for cases with zero actual as if 0.5 failures.

failures. Industry-wide data is employed. That data is applicable to HNP.

B. Success data limited to period since 10 CFR Industry-wide data is employed.

That data is 50.55 requirements to inspect.

applicable to HNP.

C. Liner flaw likelihood doubles every five years.

Calvert Cliffs sensitivities addressing range of doubling period bracket this assumption and are applicable to HNP, as well.

D. Likelihood of liner breach is a function of Calvert Cliffs sensitivities addressing range of failure pressure.

probabilities at high and low range of pressure are applicable to HNP, as well.

E. Basemat leakage assumption.

HNP basemat (liner thickness and placement of concrete) is similar to Calvert Cliffs.

F. Visual detection likelihoods.

Calvert Cliffs sensitivities addressing range of detection likelihoods are applicable to HNP.

G.

Non-detectable containment over-This conservative assumption avoids need for detailed pressurization failures are assumed to be LERF.

analysis of containment failure timing and operator recovery actions.

Calculations This analysis provides information previously requested by the NRC for ILRT extension evaluations at the other Progress Energy nuclear plants.

An analytical approach employed by Calvert Cliffs to estimate risk due to concealed containment corrosion is used, with adaptations for the Harris Nuclear Plant described below. Discussion of the method and its underlying assumptions is provided in Reference 5. Table 1 below documents the application of the method to HNP.

CALCULATION NO.HNP-F/PSA-0066 REV. 0 ATTACHMENT 3, PAGE 2 Step 0 is added to show the calculation of inaccessible areas of the containment liner.

To account for areas near the fuel transfer tube and other obstructed locations, only 97%

of the cylinder walls and dome is assumed to be accessible for inspection. A sensitivity calculation assuming an approximate 5,000 sq. ft. sump liner did not affect the conclusion. Liner dimensions are taken from Reference 8.

Step 1 is updated from the Calvert Cliffs analysis to account for two additional failures recognized as applicable by the NRC (Reference 9). The applicable period since the 10 CFR 50.55a requirement is now 7.5 years (Sep-96 to Mar-04).

Steps 2 and 3 are calculated with this updated information. An additional sensitivity to increase the number of failures by one (to a total of five) is included, to address emerging information from the March 2004 Brunswick Nuclear Plant liner inspection.

Plant specific information is input into Step 4. The upper end pressure (153 psig) is taken from the HNP IPE containment overpressure capacity for the limiting failure mode, basemat shear (Reference 1) and converted to psia. The ILRT test pressure (44 psig) is taken from Reference 6 and converted to psia with I psia added for conservatism. The Step 6 likelihood of non-detected containment leakage is weighted by the accessible and inaccessible percentage of the liner, calculated in Step 0.

The internal events CDF is taken from the current PSA model (Reference 2). Portions of CDF that are already LERF or that can never become LERF and thus are unaffected by the liner corrosion are taken from Reference 4 (Attachment 1). Reference 7 provided the input to the vendor report for plant damage states.

The Calvert Cliffs analysis provides a number of sensitivity calculations.

Those calculations are illustrative of the impact of the assumptions and are not repeated here.

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

Conclusions If all non-detectable containment leakage events are considered to be LERF, then the increase in LERF associated with containment liner corrosion based on going from an ILRT frequency of three times per ten years to once per fifteen years is 2.47E-9, which is a very small contribution.

If an additional failure is included in the historical liner flaw likelihood, the increase in LERF is 3.21 E-9, which remains a very small contribution. If sequences that are already LERF and sequences that can never become LERF are excluded from the calculation, then the increase in LERF associated with containment liner corrosion is 1.14E-9. These sensitivities demonstrate the substantial amount of conservatism in the above calculation.

CALCULATION NO.HNP-F/PSA-0066 REV. 0 ATTACHMENT 3, PAGE 3 Table I Step Item Inputs ContainmnentCylinder and Dome Basemat 0

Percent Cylinder Accessible for Inspection 97.0%

Percent Dome Accessible for Inspection 100.0%

Dome Surface Area (sq ft) 26,546 26,546 0

Cylinder Surface Area (sq ft) 65,345 63,385 1,960 Drywell Floor Surface Area (sq ft) 13,273 13,273 Vertical Sides of Sump 0

0 Total Surface Area (sq ft) 105,165 89,931 1,960 13,273 Percentage Total 85.5%

1.9%

12.6%

Percentage Accessible (to weight Step 6) 97.9%

2.1%

Historic Liner Flaw Likelihood Failure Data: Containment location specific Succcess Data: Based on 70 steel-lined containments and 7.5 years since the 10CFR50.55a requirement for periodic visual inspections of containment 1

surfaces. (4 failures in 7.5 yr. assume 0.5 failure for basemat in 7.5 yr) 7.619E-03 7.619E-03 9.524E-04 2

Age Adjusted Liner Flaw Likelihood (15 yr avg) 9.44E-03 9.44E-03 1.18E-03 3

Increase in Flaw Likelihood between 3 and 15 Years 12.25%

12.25%

1.63%

Upper End Pressure (100% likelihood), psia 168 Lower End Pressure (0.1% likelihood), psia 20 Test Pressure (psia) 60 Slope (m) 4.67E-02 Intercept (b) 3.93E-04 4

Likelihood of Breach in Containment Given Liner Flaw 0.65%

0.65%

0.06%

5 Visual Inspection Detection Failure Likelihood 10%

100%

6 Likelihood of Non-Detected Containment Leakage 0.00792%

0.0792%

0.00105%

Total Likelihood of Non-Detected Containment Leakage (weighted) 0.0100%

Internal Events CDF (MOR 2003) 2.47E-05 Already LERF (RC-2,-2B,-4C,-5C) 3.92E-06 Never go to LERF (PDS Endstates A, G and D) 9.632-06 Non-LERF CDF - internal 1.14E-05 Increase in LERF due to Liner Corrosion (Internal Events Only) 2.47E-09 Increase in LERF due to Liner Corrosion (Non-LERF CDF - internal) 1.14E-09

Charles H. Cruse Vice President Nuclear Energy Constellation I

Nuclear Calvert Cliffs Nuclear Power Plant A Member of the Constellation Energy Group ATTACHMENT 4 TO HNP-F/PSA-0066 REV. 0 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410 495-4455 March 27, 2002 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

SUBJECT:

Document Control Desk Calvert Cliffs Nuclear Power Plant Unit No. 1; Docket No. 50-317 Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension

REFERENCES:

(a)

Telephone Conferences between Ms. D. J. Moeller, et al. (CCNPP) and Ms. D. M. Skay, et al., dated March 1, March 7, March 14, and March 19, 2002, same subject (b)

Letter from Mr. C. H. Cruse (CCNPP) to NRC Document Control Desk, dated January 31, 2002, "License Amendment Request:

One-Time Integrated Leakage Rate Test Extension" (c)

Letter from Mr. C. H. Cruse (CCNPP) to NRC Document Control Desk, dated November 19, 2001, "License Amendment Request: Revision to the Containment Leakage Rate Testing Program Technical Specification to Support Steam Generator Replacement" This letter provides the information requested in a series of teleconferences (Reference a) and supplements the information provided in Reference (b).

Specifically, we were asked to provide information addressing how the potential leakage due to age-related degradation mechanisms were factored into the risk assessment for our requested Integrated Leakage Rate Test (ILRT) one-time extension. In addition, we are submitting a correction to the marked-up pages originally provided in Reference (b). This information does not change the conclusions of the significant hazards determination provided in Reference (b).

REQUESTED CHANGE The final Technical Specification pages are included in Attachment (1). In Reference (b), the term "exempted" was used in the marked-up version of the Technical Specification pages. The correct term that should have been used was "excepted."

The final Technical Specification pages reflect this correction. This correction should also be applied to the change requested in Reference (c).

ATTACHMENT 4 TO HNP-F/PSA-0066 REV. 0 Document Control Desk March 27, 2002 Page 2 SUPPLEMENTAL INFORMATION Structural Design Walls The Containment Structure is a post-tensioned, reinforced concrete cylinder and dome connected to and supported by a massive reinforced concrete slab (basemat). The liner plate is 14-inch thick and is attached and anchored to the containment concrete structure. The concrete vertical wall thickness is 3-% feet. The concrete dome thickness is 3-14 feet. Since the concealed side of the liner plate is in contact with the concrete, leakage requires a localized transmission path connecting a breach in the containment concrete with a flaw in the liner.

Floor The containment basemat is a 10-foot thick base slab that was constructed monolithically with steel sections (H or W sections) laid out to match the liner plate joints and embedded such that one flange surface was flush with the finished concrete. The liner plates were then laid out on top of these sections and welded. The liner plates are full penetration welded to each other with a gap of sufficient thickness to allow the root of the weld to partially penetrate the embedded steel. This provides a segmented area under the floor liner plates where free communication from one area to the other is heavily constrained.

After welding was complete, the welds themselves were covered with channel sections (leak chases), seal welded to the plates, and ported to allow pressure testing of the liner welds. The floor liner plates were oiled and the interior slab was poured with the test connections left in place to provide for future weld testing during ILRTs.

The liner plates under the interior slab are in contact with the concrete on both sides except for a small area at the leak chases and at the edge of the concrete where an expansion material was used. Since concrete acts to protect steel in contact with it, we feel that there is little likelihood of corrosion occurring in the floor liner plates. During replacement of the moisture barrier, the area directly behind the old barrier material was determined to be the area most affected by corrosion. This area was evaluated on both units and has been incorporated into an augmented examination population required by the American Society of Mechanical Engineers (ASME) Code.

Inspectable Area Approximately 85 percent of the interior surface of the liner is accessible for visual inspections. The 15 percent that is inaccessible for visual inspections includes the fuel transfer tube and area under the containment floor.

Liner Corrosion Events Two events of corrosion that initiated from the non-visible (backside) portion of the containment liner have occurred in the industry. These events are summarized below:

On September 22, 1999, during a coating inspection at North Anna Unit 2, a small paint blister was observed and noted for later inspection and repair. Preliminary analysis determined this to be a through-wall hole. On September 23, a local leak rate test was performed and was well below the allowable leakage. The corrosion appeared to have initiated from a 4"x4"x6' piece of lumber embedded in the concrete.

ATTACHMENT 4 TO HNP-F/PSA-0066 REV. 0 Document Control Desk March 27, 2002 Page 3 An external inspection of the North Anna Containment Structures was performed in September 2001. This inspection (using the naked eye, binoculars, and a tripod-mounted telescope) found several additional pieces of wood in both Unit 1 and Unit 2 Containments. No liner degradation associated with this wood was discovered.

On April 27, 1999, during a visual inspection of the Brunswick 2 drywell liner, two through-wall holes and a cluster of five small defects (pits) in the drywell shell were discovered. The through-wall holes were believed to have been started from the coated (visible side). The cluster of defects was caused by a worker's glove embedded in the concrete.

Calvert Cliffs Inspection Program To help assure continued containment integrity, the containment liners at Calvert Cliffs Nuclear Power Plant (CCNPP) are examined in accordance with the requirements of ASME Boiler and Pressure Vessel (B&PV) Code Section XI, Subsection IWE (as amended and modified by 10 CFR 50.55a) and the plant Protective Coatings Program, both as a natural consequence of maintenance activities and as planned events. Each will be discussed separately.

During the course of maintenance activities requiring repairs to the containment liner plate coatings, ASME XI Subsection IWE requires visual exams to evaluate the condition of the liner plate. Typically, these repairs are done to correct blisters, peeling, flaking, delamination, and mechanical damage of the coating system of the liner. To date, there have been over 500 exams of this nature (one repair generates multiple exams) performed at CCNPP since the requirements of Subsection IWE were imposed with no indication of liner base metal degradation.

The safety-related Protective Coatings Program at CCNPP requires a walkdovwn of the containment interior be performed at the beginning of each refueling outage to determine areas requiring repair. This walkdown, performed by engineering personnel, maintenance personnel, and National Association of Corrosion Engineers (NACE)-trained coatings examiners, looks at accessible coated structures in the Containment as well as the liner.

Repair of items found on these walkdowns is then planned, staged, and performed, with any postponement of repairs beyond the current outage requiring engineering approval. Liner coating repairs are witnessed and documented at the beginning stage and upon completion by a Certified Non-Destructive Examination (NDE) Examiner. This is to allow proper assessment of the cause of the damage prior to repair and to document the as-left condition. The specific goal of this approach is to identify any indication of liner damage. As stated above, over 500. documented exams have shown no evidence of liner degradation.

Scheduled inservice inspection (ISI) exams are performed in accordance with the scheduling requirements of the ASME Section XI, Subsection IVE, and 10 CFR 50.55a.

These documents require visual examination of essentially 100% of the containment liner accessible surface area once per ]SI period (three in ten years). This exam is performed and documented by Certified NDE Examiners during the outage and/or before an ILRT.

This exam is performed both directly and remotely, depending upon the accessibility to the various areas.

Remote exams are performed with binoculars to provide a clear view of all areas. To date, this exam has been performed twice on Unit I and once on Unit 2 with no recordable indications of liner plate degradation.

ATTACHMENT 4 TO HNP-F/PSA-0066 REV. 0 Document Control Desk March 27, 2002 Page 4 Several areas were identified on both units as candidate areas for Augmented Examination, in accordance with IWE-1241. These included areas beneath the liner to floor slab moisture barriers, potential ponding areas at structural steel attachments, and several areas with photographic evidence of dark areas. Further evaluation of these areas yielded the following conclusions:

No ponding areas were evident either as being presently wet or by the presence of watermarks.

The dark areas were identified in both cases to be insulation at a penetration.

The area beneath the moisture barrier on both units showed degradation that required engineering evaluation. The area beneath the moisture barrier was found to suffer from scaling, rust, and pitting. Areas visually representative of the worst of these were selected for detailed examination and documented using a combination of ultrasonic thickness measurement, pit depth measurement, and detailed visual examination.

These areas are now designated as Augmented Examination in accordance with Subsection IWE, and are subject to repeat examination once per ISI period as required by Subsection IWE.

The bolting examinations required by Table IWE-2500-1, Category E8.10 and E8.20, are performed during preventive maintenance activities of certain components.

These maintenance activities are scheduled to support replacement of the seals and gaskets used in the component connections.

Additionally, some of these connections are routinely used during outages, and the examination and testing of these connections is performed to re-establish containment integrity at the end of the outage.

Any parts (except for seals and gaskets, which are exempt) that are replaced are subject to compliance with our Repair and Replacement Program and receive the appropriate inspections at that time.

Non-destructive examination examiner qualifications are governed by Calvert Cliffs procedure MP-3-105, "Qualification of Non-Destructive Examination Personnel and Procedures." This procedure requires documenting the necessary experience, training, visual acuity, and certifications in accordance with American National Standards Institute/American Society for Nondestructive Testing CP-189.

Additionally the CCNPP coating examiners are NACE trained.

Effectiveness of the CCNPP inspection programs is judged to be high. This is based on the use of both NACE and CP-189-certified examiners for the different exams that are conducted. The depth that is provided by this approach yields a level of redundancy due to the differing focus of each examination.

Rigor of the examinations is provided by compliance with our Protective Coatings, NDE, and ISI programs. The coatings program controls the initial walkdown and focuses on the condition of the safety-related Level I coatings. This effort provides an initial assessment of the gross liner condition. In addition, the NDE Program provides a CP-l 89 certified examiner when preparation is started on each area to be repaired. This is done to verify the condition of the base metal as the defective coating is removed.

As noted previously, this activity has resulted in over 500 documented examinations with no indications of liner deterioration.

Further, the ISI Program for Subsections IWE and IWL requires examination of the accessible portions of the liner once per period. This exam is conducted using a mixture of direct and remote examination techniques. Both units have been examined completely through these joint programs at least one time each with no defects noted. We will perform an additional Subsection IWE visual exam during the 2004 Unit I refueling outage.

ATTACHMENT 4 TO HNP-F/PSA-0066 REV. 0 Document Control Desk March 27, 2002 Page 5 Liner Corrosion Analysis The following approach was used to determine the change in likelihood, due to extending the ILRT, of detecting liner corrosion. This likelihood was then used to determine the resulting change in risk. The following issues are addressed:

Differences between the containment basemat and the containment cylinder and dome; The historical liner flaw likelihood due to concealed corrosion; The impact of aging; The liner corrosion leakage dependency on containment pressure; and The likelihood that visual inspections will be effective at detecting a flaw.

Assumptions A. A half failure is assumed for basemat concealed liner corrosion due to the lack of identified failures.

(See Table 1, Step 1.)

B.

The success data was limited to 5.5 years to reflect the years since September 1996 when 10 CFR 50.55a started requiring visual inspection. Additional success data was not used to limit the aging impact of this corrosion issue, even though inspections were being performed prior to this date and there is no evidence that liner corrosion issues were identified. (See Table 1, Step 1.)

C.

The liner flaw likelihood is assumed to double every five years. This is based solely on judgment and is included in this analysis to address the increase likelihood of corrosion as the liner ages.

Sensitivity studies are included that address doubling this rate every 10 years and every two years.

(See Table 1, Steps 2 and 3, and Tables 5 and 6.)

D. The likelihood of the containment atmosphere reaching the outside atmosphere given a liner flaw exists is a function of the pressure inside the Containment. Even without the liner, the Containment is an excellent barrier. But as the pressure in Containment increases, cracks will form. If a crack occurs in the same region as a liner flaw, then the containment atmosphere can communicate to the outside atmosphere. At low pressures, this crack formation is extremely unlikely. Near the point of containment failure, crack formation is virtually guaranteed. Anchored points of 0.1% at 20 psia and 100% at 150 psia were selected. Intermediate failure likelihoods are determined through logarithmic interpolation. Sensitivity studies are included that decrease and increase the 20 psia anchor point by a factor of 10. (See Table 4 for sensitivity studies.)

E.

The likelihood of leakage escape (due to crack formation) in the basemat region is considered to be 10 times less likely than the containment cylinder and dome region. (See Table 1, Step 4.)

F.

A 5% visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 10% is used. To date, all liner corrosion events have been detected through visual inspection. (See Table 1, Step 5.) Sensitivity studies are included that evaluate total detection failure likelihoods of 5% and 15%. (See Table 4 for sensitivity studies.)

G.

All non-detectable containment over-pressurization failures are assumed to be large early releases.

This approach avoids a detailed analysis of containment failure timing and operator recovery actions.

ATTACHMENT 4 TO HNP-F/PSA-0066 REV. 0 Document Control Desk March 27, 2002 Page 6 Analysis Table 1 Liner Corrosion Base Case i

i

,i Containment Cylinder and t

Step i cDsiption W

I~~m iio Historical Liner Flaw Likelihood Events: 2 Events: 0 Failure Data: Containment location (Brunswick 2 and North Assume half a failure specific Anna 2)

Success Data: Based on 70 steel-lined 2/(70

5.5) = 5.2E-3 0.5/(70

5.5) = 1.3E-3 Containments and 5.5 years since the 10 CFR 50.55a requirement for periodic visual inspections of containment surfaces.

2 Aged Adjusted Liner Flaw Likelihood Year Failure Rate Year Failure Rate During 15-year interval, assumed failure I

2.1E-3 I

5.0E-4 rate doubles every five years (14.9%

avg 5-10 5.2I-3 avg 5 - 10 1.3E-3 increase per year). The average for 5 to I O1h year was set to the historical failure 15 1.4E-2 15 3.5E-3 rate. (See Table-5 for an example.)

15 year avg -=6.27E-3 15 year avg G 1.57E-3 3

Increase in Flaw Likelihood Between 3 and 15 years Uses aged adjusted liner flaw likelihood 8.7%

2.2%

(Step 2), assuming failure rate doubles every five years. See Tables 5 and 6.

4 Likelihood of Breach in Containment Pressure Likelihood Pressure Likelihood given Liner Flaw (psia) of Breach (psia) of Breach The upper end pressure is consistent 20 0.1%

20 0.01%

with the Calvert Cliffs Probabilistic Risk 64.7 (ILR1) 1.1%

64.7 (ILRT) 0.11%

Assessment (PRA) Level 2 analysis.

100 7.02%

100 0.7%

0.1 % is assumed for the lower end.

120 20.3%

120 2.0%

Intermediate failure likelihoods are 150 100%

150 10.0%

determined through logarithmically interpolation. The basemat is assumed to be 1/10 of the cylinder/dome analysis 5

Visual Inspection Detection Failure 10%

100%

Likelihood 5% failure to identify visual Cannot be visually flaws plus 5% likelihood that inspected.

the flaw is not visible (not through-cylinder but could be detected by ILRT)

All events have been detected through visual inspection.

5% visible failure detection is a conservative assumption.

ATTACHMENT 4 TO HNP-F/PSA-0066 REV. 0 Document Control Desk March 27, 2002 Page 7 Table I Liner Corrosion Base Case

--1 Containment Cylinder andI' Step l

Descijtiol'li,tl i

'D^C a etiiasema 6

Likelihood of Nton-Detected 0.0096%

0.0024%

oContainment Leakage (Steps3*4*5) 87%*1%*10%

2.2%*0.11%* 100%

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

Total Likelihood of Non-Detected Containment Leakage = 0.0096% + 0.0024% = 0.012%

The non-large early release frequency (LERF) containment over-pressurization failures for CCNPP Unit I are estimated at 8.6E-5 per year. This is based on the Revision 0 Unit I Model. This model includes both internal and external events. The external events portion of the model was recently finalized. External events represents 55% of the total core damage frequency (CDF) with fire being by far the largest external event contributor. The total CDF is 8.9E-5. This current CDF is used to re-generate the delta LERF/rem impacts for both the Crystal River (CR) method and Combustion Engineering Owners Group (CEOG) method. If all non-detectable containment leakage events are considered to be LERF, then the increase in LERF associated with the liner corrosion issue is:

Increase in LERF (ILRT 3 to 15 years) = 0.012%

8.6E-5 = 1E-8 per year.

Change in Risk The risk of extending the ILRT from 3 in 10 years to I in 15 years is small and estimated as being less than 1E-7. It is evaluated by considering the following elements:

1.

The risk associated with the failure of the Containment due to a pre-existing containment breach at the time of core damage (Class 3 events).

2.

The risk associated with liner corrosion that could result in an increased likelihood that containment over-pressurization events become LERF events.

3.

The likelihood that improved visual inspections (frequency and quality) will be effective in discovering liner flaws that could lead to LERF.

These elements are discussed in detail below.

ATTACHMENT 4 TO HNP-F/PSA-0066 REV. 0 Document Control Desk March 27, 2002 Page 8 Pre-existing Containment Breach The original submittal addressed Item 1. The submittal calculated the increase risk using a new CEOG methodology and a previously NRC-approved methodology. This supplement modifies, in Table 2, these values to reflect the recent update of the CCNPP Unit I PRA.

Table 2 Original Submitted with Updated Values

.;Method

.Acrease.Pe

i'-mly- 'i rcnta Inr ease

, i,;Inc rease; I'

linI.?e sn r-rexlyr.;.

CEOG Method 5.4E-8 236 0.3 6%

NRC Approved 2.9E-7 19.4 0.24%

Method The numerical results for the previously-approved methodology shows an LERF increase that is greater than IE-7. However, as noted in the original submittal, the calculated LERF would likely be lower than IE-7 if conservatisms associated with the modeling of the steam generator tube rupture sequences were removed (note that this improvement was not incorporated into the modified values). In addition, the steam generators for Unit I are being replaced and should further reduce this likelihood.

Liner Corrosion The original submittal also did not fully address the risk associated with liner corrosion. This supplement shows an additional small increase in LE-RF of 1E-8. Table 2 would be modified as follows:

Table 3 Updated Values with Corrosion Impact

'i

- Methtod-;

,l,.,

LERFInase P

IPecerntage Increase CEOG Method 5.4E-8 236 0.36%

CEOG Method with 6.4E-8 250 0.38%

Liner Corrosion NRC-Approved Method 2.9E-7 19.4 0.24%

NRC-Approved Method 3.0E-7 20.3 0.25%

with Liner Corrosion Visual Inspections The original submittal did not fully address the benefit of the Subsection IWE visual inspections. Visual inspections following the 1996 change in the ASME Code are believed to be more effective in detecting flaws. In addition, the flaws that are of concern for LERF are considerably larger than those of concern for successfully passing the ILRT. Integrated leakage rate test failures have occurred even though visual inspections have been performed. However, the recorded ILRT flaw sizes for these failed tests are much smaller than that for LERF. Therefore, it is likely that future inspections would be effective in detecting the larger flaws associated with a LERF.

ATTACHMENT 4 TO HNP-F/PSA-0066 REV. 0 Document Control Desk March 27, 2002 Page 9 An additional visual inspection is now planned for 2004 to further increase the likelihood for flaw detection.

Impact of Improved Visual Inspections The raw data for both the CEOG method and the NRC-approved method is contained in NUREG-1493.

This containment performance data is pre-1994.

An amendment to 10 CFR 50.55a became effective September 9, 1996. This amendment, by endorsing the use of Subsections IWE and INVL of Section XI of the ASME B&PV Code, provides detailed requirements for ISI of Containment Structures. Inspection (which includes examination, evaluation, repair, and replacement) of the concrete containment liner plate, in accordance with the 10 CFR 50.55a requirements, involves consideration of the potential corrosion areas. Although the improvement gained by this requirement varies from plant to plant, it is believed that this requirement makes the detection of flaws post-September 1996 much more likely than pre-September 1996 using visual inspections.

Visual inspection improvements directly reduce the delta LERF increases as calculated in the CEOG method and NRC-approved method. The CCNPP Unit I Containment was visually inspected in 2000 and 2002. The Unit I containment is scheduled for inspection in 2004. This increased inspection frequency further reduces the delta LERF as calculated by both the CEOG and NRC-approved methods.

Table 7 illustrates the benefit of visual inspection improvements on the delta LERF calculations:

If the improved inspections (additional inspection, improved effectiveness, and larger flaw size) were 90% effective in detecting the flaws in the visible regions of the containment (5% for failure to detect and 5% for flaw not detectable [not-through-wall]), then the increase ILRT LERF frequency could be reduced by 23.5%. See Table 7 for additional sensitivity cases. This would result in a LERF increase of less than I E-7 (without consideration of the LERF reduction due to PRA model improvements).

g ATTACHMENT 4 TO HNP-F/PSA-0066 REV. 0 Document Control Desk March 27, 2002 Page 10 Sensitivity Studies The following cases were developed to gain an understanding of the sensitivity of this analysis to the various key parameters.

Table 4 Liner Corrosion Sensitivity Cases ii.

.W

'r

!i.ji i!

i ij~sa~s scin-is

ijj, i

~~Conftainmnt iel"ti'; "

-itl

! j

,.]ili

' 1';.,'

~~2~..:h~IV lvi Non-V~isua di lThii

,:n!tit'j+I

!ak Ic'!

66 i I Age (Step,2);~J BIa 6h ifI1 1jL1:

J ii II ii Base Case Base Case Base Case Base Case Base Case Doubles every 5 years 1.1/0.11 10%

100%

IE-8 Doubles every 2 years Base Base Base 8E-8 Doubles every 10 years Base Base Base 5E-9 Base Base point 10 times Base Base 2E-9 lower (0.24/0.02)

Base Base point 10 times Base Base 51-8 higher (4.9/0.49)

Base Base 5%

Base 6E-9 Base Base 15%

Base IE-8 Lower Bound Doubles every 1O years Base point 10 times 10%

7E-II lower (0.24/0.02)T Upper Bound Double every2 years Base point 10 times 15%

100%

5E-7

ATTACHMENT 4 TO HNP-F/PSA-0066 REV. 0 Document Control Desk March 27, 2002 Page 11 Table 5 Flaw Failure Rate as a Function of Time

i;e, iilui::

tR e l, jhSuccess Rate..

0 1.79E-03 9.98E-01 I

2.05E-03 9.98E-O1 2

2.36E-03 9.98E-01 3

2.71E-03 9.97E-01 4

3.1 IE-03 9.97E-Ol 5

3.57E-03 9.96E-O1 6

4.1 OE-03 9.96E-O01 7

4.71E-03 9.95E-O1 8

5.41E-03 9.95E-01 9

6.22E-03 9.94E-Ol 10 7.14E-03 9.93E-Ol 11 8.20E-03 9.92E-01 12 9.42E-03 9.91E-01 13 1.08E-02 9.89E-0 I 14 1.24E-02 9.88E-0l 15 1.43E-02 9.86E-01 Table 6 Average Failure Rate

-Average,. -

.Avera' ge,

-,Yeirs,

Success Raitei !; F?,lFugire'Rate; 1 to 3 9.93E-1 0.71%

1 to 10 9.59E-1 4.06%

I to 15 9.06E-1 9.40%

A = 9.40% - 0.71% = 8.7% (delta between 1 in 3 years to 1 in 15 years)

ATTACHMENT 4 TO HNP-F/PSA-0066 REV. 0 Document Control Desk March 27, 2002 Page 12 Table 7 Benefit of Visual Inspection Improvements FactorI provement Reductio

.i CEOG w/Liner due to Yksja H i Delta

.IApoe~

IMt~

/ie Method WAit Coser

ALR MetoiI~,5 UAU~I~~

ethdDFt Crsonsiee JflSPLI' ions' 11 EeiI.

a LERF on Jee Itd i

r s

Pre-1996 Inspection 0%

3E-07 3E-07 SE-08 6E-08 Approach (Base Case)

Post-1996 with Visual 85%

4E-08 5E-08 8E-09 2E-08 Inspections Perfectly Accurate Post-1996 with Visual 80.8%

6E-08 7E-08 IE-08 2E-08 Inspections 95%

Accurate Post-1996 with Visual 76.5%

7E-08 8E-08 IE-08 2E-08 Inspections 95%

Accurate and 5%

chance of Undetectable Leakage Post-1996 with Visual 63.8%

IE-07 IE-07 2E-08 3E-08 Inspections 80%

accurate and a 5%

Chance of Undetectable Leakage Conclusion Considering increased frequency of visual inspections and the benefit of improved visual inspections post-1996, the increase in risk is considered to be less than IE-7 for LERF. Changes less than IE-7 are considered small per Regulatory Guide 1.174. The one-time extension of the ILRT interval from 3-in-10 years to I-in-1S years is considered an acceptable risk increase.

ATTACHMENT 4 TO HNP-F/PSA-0066 REV. 0 Document Control Desk March 27, 2002 Page 13 Should you have questions regarding this matter, we will be pleased to discuss them with you.

Very truly yours, STATE OF MARYLAND COUNTY OF CALVERT

TO WIT:

1, Charles H. Cruse, being duly sworn, state that I am Vice President - Nuclear Energy, Calvert Cliffs Nuclear Power Plant, Inc. (CCNPP), and that I am duly authorized to execute and file this License Amendment Request on behalf of CCNPP. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other CCNPP employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.

Subscribed and sworn before me a Nota nublic i

d for the State of Maryland and County of this

& day of A 'J

, 2002.

WITNESS my Hand and Notarial Seal:

B LUWCUJ L). iJULLLa Notary Public My Commission Expires:

£9 /D; /l6o Date CHC/DJM/dIm

Attachment:

(1)

Final Technical Specification Pages cc:

R. S. Fleishman, Esquire J. E. Silberg, Esquire Director, Project Directorate I-1, NRC D. M. Skay, NRC H. J. Miller, NRC Resident Inspector, NRC R. I. McLean, DNR

ML051050123

Text

6.1 DESCRIPTION

Background

SUMMARY

Title:

SUMMARY

SUMMARY

5.0 REFERENCES

SUMMARY

SUMMARY

5.0 REFERENCES

SUBJECT:

REFERENCES:

Attachment:

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