ML061740567

Proposed Risk-Informed Technical Specifications Change Five-Year Extension of Type a Test Interval (LBDCR 06-MP3-010)

ML061740567
Person / Time
Site:	Millstone
Issue date:	06/14/2006
From:	Grecheck E Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
06-315
Download: ML061740567 (52)

Text

Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, Virginla 23060 W h Address: www.dom.com June 14, 2006 U.S. Nuclear Regulatory Commission Serial No.06-315 Attention: Document Control Desk NSS&UDF RO Washington, D.C. 20555 Docket No. 50-423 License No. NPF-49 DOMINION NUCLEAR CONNECTICUT. INC. (DNC)

MILLSTONE POWER STATION UNIT 3 Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) requests an amendment to Facility Operating License Number NPF-49 in the form of a change to the technical specifications for Millstone Power Station Unit 3 (MPS3). The proposed change will permit a one-time, five-year extension of the ten-year performance-based Type A test interval established in NEI 94-01, "Nuclear Energy Institute Industry Guideline For Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 0, dated July 26, 1995. The risk assessment methodology used to support this amendment is based on EPRl's "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," which was developed for the Nuclear Energy Institute in December 2003.

This change has been prepared in accordance with the guidance provided in Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk Informed Decisions on Plant Specific Changes to the Licensing Basis." A discussion of the proposed change and the associated supporting risk assessment are included in Attachments 1 and 2 of this letter, respectively. A mark-up of Technical Specification 6.8.4.fJ "Containment Leakage Rate Testing Program," is provided in Attachment 3.

The retyped technical specification page is provided in Attachment 4.

The proposed amendment does not involve a significant impact on public health and safety arid does not involve a significant hazards consideration pursuant to the provision!; of 10 CFR 50.92 (see Significant Hazards Consideration in Attachment 1).

The Site Operations Review Committee has reviewed and concurred with the determinations.

To permit:effective Cycle 12 planning, DNC is requesting NRC staff review and approval of the proposed change by February 2007. Once approved, the amendment will be implemented within 60 days.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Page 2 of 3 Should you have any questions or require additional information, please contact Mr. Paul R. Willoughby at (804) 273-3572.

Very truly yours, Eugene St. Grecheck Vice President - Nuclear Support Services Attachments (4)

1. Evaluation of Proposed License Amendment

2. Risk Impact Assessment

3. Mark-up of Technical Specifications

4. Retyped Technical Specifications Page Commitments made in this letter: None cc: U.S. Nuclear Regulatory Commission Region I 47i5 Allendale Road King of Prussia, PA 19406-1415 V. Nerses Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8C2 Rolckville, MD 20852-2738 S. M. Schneider NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Page 3 of 3 COMMONWEALTH OF VIRGINIA )

COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Eugene S. Grecheck, who is Vice President -

Nuclear Support Services, of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this 14" day of 9- , 2006.

My Commission Expires: h l ~ a 3t/ . a d 8 .

(SEAL)

Serial No.06-315 Docket No. 50-423 ATTACHMENT 1 PROPOSED RISK-INFORMED TECHNICAL SPECIFICATIONS CHANGE FIVE-YEAR EXTENSION OF TYPE A TEST INTERVAL

{LBDCR 06-MP3-010)

EVALUATION OF PROPOSED LICENSE AMENDMENT DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 1 Page 1 of 13 EVALUATION OF PROPOSED LICENSE AMENDMENT DESCRIPTION PRtOPOSED CHANGE BA,C:KGROUND 3.1 10 CFR 50, Appendix J, Option B Requirements 3.2 Reason for Proposed Amendment TEC:HNICAL ANALYSIS 4.1 Plant Specific Risk Assessment for the Extended ILRT Test Interval 4.2 10 CFR 50 Appendix J, Option B Integrated Leak Test Information 4.3 Plan Operational Performance 4.4 IW E/IWL lnservice Inspection (ISI) Program and Activities to Support ILRT 4.5 Safety Related Porous Concrete Groundwater Sump (Underdrain System Sump) 4.6 Containment Liner Corrosion Sensitivity Analysis 4.7 Fuel Transfer Bellows NC) SIGNIFICANT HAZARDS CONSIDERATION EhlVlRONMENTAL CONSIDERATION

7.0 REFERENCES

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 1 Page 2 of 13 1.0 DESCR 'IOP Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) requests a change to the surveillance requirements referenced in Section 4.6.1 of the Millstone Power Station Unit 3 technical specifications for the containment structure. The proposed change will permit Millstone Power Station Unit 3 (MPS3) a one-time, five-year extension to the requirement of NEI 94-01 (Reference 1) which specifies performance of an integrated leak rate test (ILRT) at a frequency of up to ten years with allowance for a fifteen-month extension.

A mark-ulp of Technical Specifications page 6.17 incorporating the proposed change to Technical Specification 6.8.4.f) "Containment Leakage Rate Testing Program," is provided in Attachment 3 and the retyped page is provided in Attachment 4. The wording in MPS3 Technical Specification 4.6.1.2 will remain the same.

It has been determined that the change qualifies for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(~)(9). Therefore, no environmental impact statement or environmental assessment is needed in connection with the a.pproval of the proposed change.

2.0 PFIOPOSED CHANGE The proposed change will modify the first paragraph of Technical Specification 6.8.4.f as follows:

Current A program shall be established to implement the leakage rate testing on the cointainment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions*. This program shall be in accordance with the guidelines contained in Regulatory Guide (RG) 1.I 63, "Performance-Based Containment Leak-Test Program," dated September 1995.

A program shall be established to implement the leakage rate testing on the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions*. This program shall be in accordance with the guidelines contained in RG 1.I 63, "Performance-Based Camtainment Leak-Test Program," dated September 1995, as modified by the following exception to NEI 94-01, Rev. 0, "Industry Performance-Based Option of 10 CFR Part 50, Appendix J": The first Type A test performed after the January 6, 1998 Type A test shall be performed no later than January 6, 2013.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 1 Page 3 of 13

3.0 BACKGROUND

The MPS3 current ten-year Type A test interval ends on January 6, 2008. In order to meet the interval requirements of NEI 94-01, this test must be performed during either the spring 2007 refueling outage, or using the fifteen-month extension provision, during the fall 2008 refueling outage.

The proposed amendment to the TS takes a one-time exception to the 10-year frequency of the performance-based leakage rate testing program for Type A test as documen'ted by NEI 94-01. The exception will allow ILRT testing within 15 years from the last IL.RT which was performed in January 1998.

The proposed changes discussed within this license amendment request are similar to license amendments issued to Millstone Power Station Unit 2 License No. DPR-65 (Amendment No. 285) on April 6, 2005, Surry Power Station Unit 1 License No. DPR-32 (Amendment No.233) on December 16, 2002, and to North Anna Power Station Unit 1 License No. NPF-4 (Amendment No. 234) on December 31, 2002, and Vermont Yankee Nuclear Power Station License No. DPR-28 (Amendment No. 227) on August 31,2005.

3.1 10 CFR 50, Appendix J, Option B Requirements The testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage through the containment, including systems and components that penetrate the containment, does not exceed the allowable leakage values specified in the technical specifications. The limitation of containment leakage provides assurance that the containment would perform its design function following an accident up to and including the plant design basis accident.

10 CFR !SO, Appendix J, was revised, effective October 26,1995, to allow licensees to choose containment leakage testing under Option A "Prescriptive Requirements" or Option B "Performance-Based Requirements." Amendment 186 (Reference 2) was issued to Millstone Power Station Unit 3 to permit implementation of 10 CFR 50, Appendix J, Option B. Amendment 186 modified Technical Specification Section 4.6.1 and Technical Specification 6.8.4 which require testing in accordance with the Containment Leakage Testing Program and RG 1.I 63 (Reference 3), respectively. RG

1. I 63 specifies a method acceptable to the NRC for complying with Option B by approving the use of NEI 94-01 and ANSIIANS 56.8-1 994 (Reference 4), subject to several regulatory positions in the guide.

The adoption of the Option B performance-based containment leakage rate testing program did not alter the basic method by which Appendix J leakage rate testing is performed, but it did alter the frequency of measuring primary containment leakage in Type A, B and C tests. Frequency is based upon an evaluation which looks at the "as

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 1 Page 4 of 13 found" leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained. The changes to the Type A test frequency did not result in an increase in containment leakage. Similarly, this proposed change to the Type A test frequency will not result in an increase in containment leakage.

3.2 Reason for Proposed Amendment The frequency interval for testing allowed by NEI 94-01 is based upon a generic evaluation documented in NUREG-1493 (Reference 5). NUREG-1493 made the following observations with regard to extending the test frequency:

"Reducing the Type A (ILRT) testing frequency to one per twenty years was fo~indto lead to an imperceptible increase in risk. The estimated increase in risk is small because ILRTs identify only a few potential leakage paths that cannot be identified by Type B and C testing, and the leaks that have been found by Type A tests have been only marginally above the existing requirements. Given the insensitivity of risk to containment leakage rate, and the small fraction of leakage detected solely by Type A testing, increasing the interval between ILRT testing hald minimal impact on public risk."

"While Type B and C tests identify the vast majority (greater than 95%) of all potential leakage paths; performance-based alternatives are feasible without significant risk impacts. Since leakage contributes less than 0.1 percent of overall risk under existing requirements, the overall effect is very small."

Exceptior~sto the requirements of RG 1.163, are allowed by 10 CFR 50, Appendix J, Option B,Section V.B, "Implementation," which states, "The Regulatory Guide or other implementing document used by a licensee or applicant for an operating license, to develop a performance based leakage-testing program must be included, by general reference, in the plant technical specifications. The submittal for technical specification revisions must contain justification, including supporting analyses, if the licensee chooses to deviate from methods approved by the Commission and endorsed in a regulatory guide." Since exceptions meeting the stated requirements are permitted, license amendment requests satisfying these requirements do not require an exemption to Option B.

4.0 TECHNICAL ANALYSIS

4.1 Plant Specific Risk Assessment for the Extended ILRT Test lnterval The surveillance frequency for Type A testing in NEI 94-01 is at least once per ten years based on an acceptable performance history (i.e., two consecutive periodic Type A tests at least 24 months apart where the calculated leakage rate was less than 1.0 La) and considera.tion of the performance factors in NEI 94-01, Section 11.3. Based on the

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 1 Page 5 of 13 October '1993 and January 1998 ILRTs, the current interval for MPS3 is once every ten years.

A risk assessment was performed in accordance with the guidelines set forth in NEI 94-01, the methodology used in EPRI TR-104285 (Reference 6), the NEI Interim Guidance (Reference 7), and the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) findings and risk insights in support of a licensee request for changes to a plant's licensing basis, RG 1.174 (Reference 8). The risk impact assessment is provided in Attachment 2 of this letter.

4.1.1 Method of Analysis A simplified bounding analysis approach was used for evaluating the change in risk associated with increasing the interval for performing the Type A test from ten years to fifteen years.

The Type! A test measures the containment air mass and calculates the leakage from the change in mass over time. Likewise, this approach is used in the analyses presented in EPRI TR-104285, NUREG-1493, and the NEI lnterim Guidance. The analysis performed examines plant specific accident sequences in which the containment integrity remains intact or the containment is impaired. Specifically, the following ,were considered:

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

Core damage sequences in which containment integrity is impaired due to random isolation failures of plant components other than those associated with Type B or Type C test components (e.g., a liner breach or steam generator manway leakage [EPRI TR-104285 Class 3 sequences]). Type B tests measure component leakage across pressure retaining boundaries (e.g., gaskets, expansion bellows and air locks). Type C tests measure component leakage ratles across containment isolation valves.

Core damage sequences in which containment integrity is impaired due to containment isolation failures of pathways left 'opened' following a plant post-maintenance test (e.g., a valve failing to close following a valve stroke test

[EPRI TR-104285 Class 6 sequences]).

Accident sequences involving containment failure induced by severe accident phenomena (EPRI TR-104285 Class 7 sequences), containment bypassed (EPRI TR-104285 Class 8 sequences) and large containment isolation failures (EPRI TR-104285 Class 2 sequences). Small containment isolation

'failure-to-seal' events (EPRI TR-104285 Class 4 and 5 sequences) were not

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 1 Page 6 of 13 accounted for in this evaluation. These sequences are impacted by changes in Type B and C test intervals, not changes in the Type A test interval.

Based 011 the above sequences considered, the following conclusions are made regarding the plant risk associated with extending the Type A ILRT test frequency from ten years to fifteen years:

RG 1.I74 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.I74 defines very small changes in risk as resulting in increases of CDF below 10'6/yr and increases in LERF below 10'~1yr.

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 3.1 x 1 0 - ~ /based

~r on internal events. RG 1.I 74 states that when the calculated increase in LERF is in the range of 10-~/yrto I ~ ~ ~ applications l ~ r , will be considered if it can be shown that the total LERF is less than 10-~/yr. Since the total LERF for the 15-year metric is 6.3~10'~/yr,then the proposed change is considered acceptable.

The increase in the total integrated plant risk is defined here by person-remlyear increases for those accident sequences influenced by Type A testing. The one-time change to the Type A test interval from ten years to fifteen years increases the ILRT dose rate by 2.1 %. This change in dose rate is due to the conservative assumption made in the source term release fraction calculation.

The change in conditional containment failure probability (CCFP) is calculated to de~monstratethe impact on 'defense-in-depth.' For the current ten-year ILRT interval, the contribution of sequences involving containment failure for the ten-year interval is 50.2%. For the proposed fifteen-year interval, the contribution of sequences involving containment failure increased to 50.7%. Therefore, the AC8CFPlo-15is found to be 0.5%. This represents a small change in the MPS3 containment defense-in-depth.

4.2 10 CFR 50 Appendix J, Option B Integrated Leak Test Information A Type A test can detect containment leakage due to a loss of structural capability. All other sources of containment leakage detected in Type A test analyses can be detected by the Type B and C tests.

Previous Type A tests confirmed that the Millstone Power Station Unit 3 reactor containment structure has extremely low leakage and represents an insignificant potential risk contributor to increased containment leakage. The increased leakage is minimized by continued Type B and Type C testing for penetrations with direct

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 1 Page 7 of 13 communication with containment atmosphere. Also, the inservice inspection (ISI) program and maintenance rule program require periodic inspection of the interior and exterior of the containment structure to identify degradation.

The results for the last two Type A tests are reported in the following table for MPS3:

As Found Leakage Rate(*) Limit(**) Test Pressure Date WTO/ddav WTO/ddav (osial October 12, 1993 0.1 333 0.65 39 January 6, 1998 0.1 158 0.3 39

• This is the leakage attributable to the leakage integrity of the containment str~ucture. It is calculated as the sum of the Type A upper confidence limit (UCL) anld As-Left minimum pathway leakage for all Type B and Type C pathways in service, isolated and not lined up in their test position prior to performing the test.

• • The performance criteria for Type A test allowable leakage is less than 1.0 La 4.3 Plant Operational Performance During power operation, control room instrumentation provides constant indication of containment pressure. With the containment at subatmospheric conditions, if pressure rises, an alarm annunciates advising conditions are approaching the limits allowed by the technical specifications. Additionally, any significant degradation would become evident by the inability to maintain containment vacuum or by excessive operation of the containment vacuum pumps. This monitoring of the containment pressure equates to continuo~~s on-line monitoring of the containment leakage during operation.

4.4 IW EIIWL lnservice Inspection (ISI) Program and Activities to Support ILRT The current regulatory requirement mandated by 10 CFR 50.55a requires licensees to implement a containment inspection program in accordance with the rules and requirements of the 1992 Edition through the 1992 Addenda of ASME Section XI, Subsections IWE and IWL, as amended in the regulation. DNC implemented the Containment IS1 Program in accordance with these rules at each of its two operating nuclear units. The regulatory requirement allows five years for the implementation of the first period inspections. In consideration of these rules, the Initial Period (First Period) for the performance of Containment IS1 began on September 9, 1996 and ended on September 9, 2001. The subsequent periods (IWE) comply with the normal period requirements of four years for the second period and three years for the third period of inspection program B of ASME Section XI. The subsequent IWL intervals are repeated every five years. The proposed frequency extension of ILRT requirements would have no effect upon these requirements. The regulation requires the general visual

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 1 Page 8 of 13 examination, IWE Category E-A, be conducted each inspection period during the interval in addition to the Code requirement that is to be completed just prior to the Type A test. This general visual examination requirement is similar to the visual requirement of Appendix J. The general visual examination requirement conducted each period will be maintained during the extended ILRT period beyond the normal code required ten-year interval. No Code requirement (IWE, Category E-A) will be affected by the ILRT period ex1:ension.

The follovving relief requests were reviewed to assess the effect, if any, resulting from the propomsed ILRT period extension:

Relief Requests RR-El and RR-L1 requested relief from Section XI of the ASME Code, 1992 Edition, 1992 Addenda, for all IWE and IWL zones, respectively.

The relief permits the use of the rules provided in the ASME Code Section XI, 19138 Edition, Subsections IWE and IWL for Class MC and Class CC examinations required to be performed under the expedited containment examination rules of 10 CFR 50.55a(g)(6)(ii)(B). The NRC letter dated April 21, 2000, granted this relief to Millstone Unit 3 (Reference 9). The proposed ILRT pe~riodextension only affects the length of time between Type A testing. The type or method of examinations is not changed and, therefore, the relief request rernains valid and unaffected by the proposed change.

Relief Request RR-E2 requested relief from Section XI of the ASME Code, 1998 Edition, IWE-2500(b)(l) which requires detailed visual examination of both sides of an accessible surface and IWE(b)(2) which requires ultrasonic thickness measurements. DNC's relief request proposed the use of detailed visual examination on the accessible surface areas supplemented by volumetric examination as specified as part of the engineering evaluation of each E-C category surface. The NRC letter dated November 14, 2000, granted this relief to Millstone Power Station Unit 3 (Reference 10). The proposed ILRT frequency extension affects Type A testing only. The examination methods authorized by the NRC remain unchanged. As a result, the relief request remains valid and is unaffected by the proposed change.

DNC Englineering performs IWEIIWL IS1 inspection activities in support of the required Type A (ILRT) test. There will be no change to the schedule for these inspections due to the extension of the Type A test interval. The activities that assure continued containment integrity include:

During the March 2001 and April 2004 refueling outages, DNC performed IWE General Visual examinations of the Containment Metal Liner (IWE - MC component). All accessible surface areas were examined including the area of the liner behind the 112-inch by 112-inch moisture barrier seal. Some localized rust and surface anomalies were detected, most associated near or at attachments to the upper uncoated dome liner area or behind the moisture

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 1 Page 9 of 13 bavrier seal. Repairs scheduled for a number of Examination Category E-C (accelerated degradation) classified items were made during the past two refueling outages. They included repairs to the moisture barrier seal and recoating of some of the liner surface area at selected localized rust areas. A nu~nberof new E-C items, most located in the upper dome area, were identified in the April 2004 examination. They were conservatively classified as E-C to the next inspection period and are expected to be reclassified the next inspection.

ln~~pections of the containment liner are performed during the interval between ILRTs. The extension of the ILRT period will not affect the inspections.

Thle performance-based ILRT program guidance (NEI 94-01 and Regulatory Guide 1.163) requires a minimum of three inspections of the accessible portions of ,the inside and outside of the containment structure to assess the condition of the containment structure during the ten year interval. Engineering personnel performed these inspections to an owner-defined program up through the March 2001 refueling outage. Certified Level II VT-1 and VT-3 visual examiners performed these inspections under the supervision of the Responsible Engineer during the April 2004 refueling outage. Any identified discrepancies noted in the liner, penetrations or concrete are documented and dispositioned in accordance with the appropriate Codeldesign requirements. These inspections are conducted using a mixture of direct and remote examination techniques.

Thle accessible portions of the containment liner are inspected during each of the three periods in the ten-year inspection interval as required by ASME Code, Se'ction IWE. These inspections are performed by qualified personnel, and any identified discrepancies are documented and dispositioned in accordance with ASME Section XI requirements. These inspections are conducted using a mixture of direct and remote examination techniques.

Coating inspections are performed each outage on accessible portions of the containment liner by engineering personnel. Any identified discrepancies in the coating or liner are documented and dispositioned in accordance with the appropriate design standards.

The above visual inspections of the containment have proven to be effective in identifying degradation of either the interior liner or the exterior concrete surface.

4.5 Safety Related Porous Concrete Groundwater Sump (Underdrain System Sump)

There is a porous concrete groundwater sump and non-safety related sump pump located in the engineered safety features building that collects (via an underdrain and porous concrete media) any significant amount of groundwater seepage which has circumverlted the waterproof membrane. The sump protects the containment steel liner from hydrlostatic loading. The electric sump pump is powered from a non-safety related

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 1 Page 10 of 13 electrical circuit which derives power from a safety-related electrical bus, providing greater assurance of a reliable energy source. (FSAR Section 3.8.1.6.4 contains information on the waterproof membrane, and FSAR Section 9.3.3.2.4.1 contains additional details on the sump pump.)

4.6 Containment Liner Corrosion Sensitivity Analysis An undetected through-wall hole in both the concrete and the liner, at approximately the same location would have to be postulated to be a LERF contributor. Furthermore, both leak paths would have to exist long enough for the pathways to grow sufficiently such that the release would be large enough to be considered a LERF contributor. As a result of the liner and concrete inspections, the likelihood of an undetected through-wall path from the containment atmosphere to the environment for even a very small leak is considered to be remote. The likelihood of occurrence of an undetected through-wall path becomes even smaller as the assumed leak size increases. A sensitivity analysis has been performed to estimate the impact of failure from a defect initiated between the containment wall and the liner. This sensitivity analysis used historical data to establish flaw likelihood. Given the assumed liner flaw, the containment fragility analysis is used to estimate the probability of breaching the containment at the design pressure. Finally, the likelihood of visual detection failure is assessed and included in the analysis. The product of these terms is the likelihood of non-detected containment leakage, which was calculated for both the containment cylinder and the basemat in the sensitivity analysis.

The product of this likelihood and the non-large early release frequency is the increase in LERF due to non-detected containment leakage. The key calculations and assumptions in the sensitivity analysis are located in Attachment 2.

4.7 Fuel Transfer Tube Bellows The MPS3 fuel transfer tube consists of a sleeve welded to the containment liner and attached 'to the transfer tube by means of a bellows connection. The area between the tube and the sleeve is provided with a test connection for testing the bellows seal connection. The fuel transfer tube blind flange is double-gasketed and can be pressurized for Type B leakage rate testing each refueling outage.

5.0 NC) SIGNIFICANT HAZARDS CONSIDERATION The proposed revision to the Millstone Power Station Unit 3 Technical Specifications permits a one-time extension to the current interval for Type A testing. The current test interval of ten years, which is based on the standard of good past performance, would be extended on a one-time basis to fifteen years from the last Type A test for Millstone Power Station Unit 3.

Dominion Nuclear Connecticut, Inc. (DNC) has evaluated whether or not a Significant Hazards Consideration is involved with the proposed changes by addressing the three standards set forth in 10 CFR 50.92(c) as discussed below.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 1 Page 11 of 13 Criterion I:

Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed extension to Type A testing cannot increase the probability of an accident previously evaluated since extension of the containment Type A testing is not a physical plant modification that could alter the probability of accident occurrence nor, is it an activity or modification that by itself could lead to equipment failure or accident initiation.

The proposed one-time, five-year extension to Type A testing does not result in a significant increase in the consequences of an accident as documented in NUREG-1493. The NUREG notes that very few potential containment leakage paths are not identified by Type B and C tests. It concludes that even reducing the Type A (ILRT) testing frequency to once per twenty years leads to an imperceptible increase in risk.

DNC provides a high degree of assurance through indirect testing and inspection that the containment will not degrade in a manner detectable only by Type A testing. The last two Type A tests identified containment leakage within acceptance criteria, indicating a very leak-tight containment. Inspections required by the ASME Code are also performed in order to identify indications of containment degradation that could affect leak-tightness. Separately, Type B and C testing required by Technical Specifical:ions, identifies any containment opening from design penetrations, such as valves, that would otherwise be detected by a Type A test. These factors establish that a one-time, five-year extension to the Millstone Power Station Unit 3 Type A test interval will not represent a significant increase in the consequences of an accident.

Criterion 2:

Does the proposed license amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed revision to Technical Specifications adds a one-time extension to the current interval for Type A testing for Millstone Power Station Unit 3. The current test interval o,f ten years, based on past performance, would be extended on a one-time basis to fifteen years from the last Type A test. The proposed extension to Type A testing doles not create the possibility of a new or different type of accident since there are no physical changes being made to the plant and there are no changes to the operation of the plant that could introduce a new failure.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 1 Page 12 of 13 Criterion :3:

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The proposed revision to Millstone Power Station Unit 3 Technical Specifications adds a one-time extension to the current interval for Type A testing. The current test interval of ten years, based on past performance, would be extended on a one-time basis to fifteen years from the last Type A test for Millstone Power Station Unit 3. 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 CDF below 10-

'lyr and increases in LERF below 10'~lyr. 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 3.1 x 10-~/yr, based on internal events. RG 1.174 states that when the calculated increase in LERF is in the ranlge of 10'~lyrto 10'~/yr,applications will be considered if it can be shown that the total LEFT is less than 10-~/yr. Since the total LERF for the 15-year metric is 6 . 3 x 1 0 - ~ / ~then r , the change is considered acceptable. Increasing the ILRT interval from ten to fifteen years is, therefore, considered non-risk significant and will not significantly reduce the margin of safety. The NUREG-1493 generic study of the effects of extending containment leakage testing found that a 20-year interval in Type A leakage testing resulted in an imperceptible increase in risk to the public. NUREG-1493 generically concludes that the design containment leakage rate contributes about 0.1 percent of the overall risk. Decreasing the Type A testing frequency would have a minimal affect on this risk since 95% of the Type A detectable leakage paths would already ble detected by Type B and C testing.

In summiary, DNC concludes that the proposed amendment does not represent a Significant Hazards Consideration under the standards set forth in 10 CFR 50.92(c).

6.0 EhlVlRONMENTAL CONSIDERATION DNC has determined that the proposed amendment would change requirements with respect to use of a facility component located within the restricted area, as defined by 10 CFR 20, or an inspection or surveillance requirement. DNC has evaluated the proposed change and has determined that the change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released off site, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51 .22(c)(!3). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 1 Page 13 of 13

1. NEl 94-01, "Nuclear Energy Institute Industry Guideline For Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 0, July 26, 19!35.

2. NFlC letter to Millstone Unit 3 Issuing Technical Specification Amendment 186, dated November 2, 2000, to implement the requirements of 10 CFR 50, Appendix J, Option B for performance-based primary reactor containment leatkage testing.

3. Regulatory Guide 1.I 63, "Performance-Based Containment Leak-Test Program,"

September 1995.

4. American National Standard ANSIIANS 56.8-1994, "Containment System Le'akage Testing Requirements."

5. NUREG-1493, "Performance-Based Containment Leak-Test Program," Final Report, September 1995.

6. EF'RI TR-104285, "Risk Assessment of Revised Containment Leak Rate Testing Intervals," dated August 1994.

7. Interim Guidance for Performing Risk Impact Assessments In Support of One-Tirne Extensions for Containment Integrated Leak Rate Tests for Surveillance Intervals, Dated November 2001.

8. Regulatory Guide 1.I 74, "An Approach for Using Probabilistic Risk Assessment In Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

dated July 1998.

9. USNRC letter from James W. Clifford to S.E. Scace, "Safety Evaluation for Alternative Associated With the Use of Subsection IW E and IWL of the ASME Code for Containment Inspection, Millstone Nuclear Power Station, Unit Nos. 2 and 3 (TAC Nos. MA5332 and MA5338)," dated April 21,2000.

10. USNRC letter from James W. Clifford to S.E. Scace, "Safety Evaluation of Relief Request RR-E2 for the Containment Inspection Program, Millstone Nuclear Power Station, Unit Nos. 2 and 3 (TAC Nos. MB0164 and MB0165)," dated November 14,2000.

Serial No.06-315 Docket No. 50-423 ATTACHMENT 2 PROPOSED RISK-INFORMED TECHNICAL SPECIFICATIONS CHANGE FIVE-YEAR EXTENSION OF TYPE A TEST INTERVAL JLBDCR 06-MP3-010)

RISK IMPACT ASSESSMENT DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 1 of 30 RISK IMPACT ASSESSMENT Table of Contents

1. Purpose........................................................................................................................... 2 2 . Summary ......................................................................................................................... 2 3 . References ...................................................................................................................... 2 4 . Assumptions .................................................................................................................... 4 5 . Method of Calculation...................................................................................................... 4 6 . Boldy of Calculation ......................................................................................................... 6 7 . Design Review .............................................................................................................. 28 8 . Enclosure ...................................................................................................................... 28 -A: MP3 Frequency and Dose Data................................................................. 30 List alf Tables Table 1 . Mean Containment Frequencies Measures .Given Accident Class ...................11 Table 2 . Person .Rem Measures .Given Accident Class .............................................. 12 Table 3 . Baseline Mean Person-Rem Measures .Given Accident Class .........................12 Table 4 . Mean Consequence Measures for 10-Year Test Interval .Given Accident Class 14 Table 5 - Mean Consequence Measures for 15-Year Test Interval - Given Accident Class 16 Table 6 - Evaluated Impact of Containment Leak Size on Containment Leak Rate .......... 18 Table 7 - Liner Corrosion Base Case ................................................................................ 24 Table 8 - Summary of Results ............................. ........................................................... 27 Table A MPS3 Frequency and Dose Data ................................................................... 30 List of Figures Figure 1 . Fractional Impact on Risk Associated with Containment Leak Rates................19

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 2 of 30

1. Purpose Provide a risk impact assessment on extending the plant's integrated leak rate test (ILRT) interval from ten to fifteen years. The risk assessment is performed in accordance with the guidel~inesset forth in NEI 94-01 [I], the NEI interim guidance [21], the methodology used in EPRl reports [2, 221 and the NRC regulatory guidance RG 1. I74 [3]. The RG provides use of Probabilistic Risk Assessment (PRA) findings and risk insights in support of a licensee request for changes to a plant's licensing basis.

2. Summary In October 26, 1995, the NRC revised 10 CFR 50, Appendix J. The revision to Appendix J allowed individual plants to select containment leakage testing under Option A "Prescriptive Requirements" or Option B "Performance-Based Requirements." The Millstone Unit 3 Nuclear Power Station (MPS3) selected the requirements under Option B as its testing program [4].

The surveillance testing requirements as proposed in NEI 94-01 [ I ] for Type A testing is at least once per 10 years based on an acceptable performance history (defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage was less than 1La).

The Millstone Unit 3 current ten-year Type A test interval ends on January 6, 2008. The propo:sed amendment to the TS takes a one-time exception to the 10-year frequency of the performance-based leakage rate testing program for Type A test as documented in NEI 94-

01. The exception will allow ILRT testing within 15 years from the last ILRT which was performed in January 1998.

This calculation will provide a risk impact assessment on extending the plant's integrated leak rate test (ILRT) interval from ten to fifteen years. The risk assessment will be performed in accordance with the guidelines set forth by NEI [ I , 211, the methodology used by EPRI [2, 221 and the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) findings and risk insights in support of a licensee request for changes to a plan~t'slicensing basis, RG 1.174 [3].

In addition, the results and findings from the Millstone Individual Plant Examination (IPE)

[5], the level 2 data from the license renewal SAMA [23] and the revised model [10,17] are used for this risk assessment calculation.

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

2) RF:-Report, EPRl TR-104285, "Risk Assessment of Revised Containment Leak Rate Testing Intervals," August 1994.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 3 of 30

3) RF-Report, Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis," July 1998.

4) RF-Procedure, Surveillance Procedure, SP31103, "Containment lntegrated Leak Rate Test, Type A," Revision 3, 1998.

5) RF-Report, "Millstone Unit No. 3 lndividual Plant Examination For Severe Accident Vulnerabilities," August 1990.

6) RF-Calc., PRA02NQA-01895S3 Revision 1, "MACCS2 model for Millstone Unit 3 Level 3 Application," August 2004.

7) RF-Report, NUREG-1493, "Performance-Based Containment Leak-Test Program," July 19!35.

8) RF-Report, United States Nuclear Regulatory Commission, "Individual Plant Examination: Submittal Guidance," NUREG-1335, August 1989.

9) RF-Report, Z.T. Mendoza, et al., "Generic Framework for lndividual Plant Examination (IPE) Backend (Level 2) Analysis, Volume 1 - Main Report and Volume 3 - BWR Implementation Guidelines," prepared by SAlC International, Inc., Electrical Power Research Institute, NSAC-159, EPRI PR3114-29, 1991.

10) RF-Calc., PRA02YQA-01822S3, Revision 0, "Millstone 3 PRA Model," October 2002.

1 I)RF-Report, ERC:25212-ER-04-0015, "MP3 Containment Risk Significant Valve Review",

3-17-2004.

12) RF-Calc., Indian Point 3, IP3-CALC-VC-03357 Revision 0, "Risk lmpact Assessment of Extending Containment Type A Test Interval," 1-4-01.

13) RF-Report, Patrick D. T. OIConnor, "Practical Reliability Engineering," John Wiley &

Sons, 2nd Edition, 1985.

14) RF-Report, Burns, T.J., "lmpact of Containment Building Leakage on LWR Accident Risk," Oak Ridge National Laboratory, NUREGICR-3539, April 1984.

15) RF-Report, United States Nuclear Regulatory Commission, Reactor Safety Study, WASH-1400, October 1975.

16)RF-Calc., Florida Power Calculation, F-01-0001, Revision 2, "Evaluation of Risk Significance of ILRT Extension," 6-19-01.

17) RFr-Calc., PRA02NQA-01941S3, Revision 2, "MP3 SAMA lmpact on Containment Release Frequencies," August 2004.

18) RF-Memo., NSE-06-014, "MP3 ILRT 5-Year Exemption Support," March 23, 2006.

19) RF-Report., Calvert Cliffs Nuclear Power Plant, Letter from Mr. Charles H. Cruse to NFlC Document Control Desk, "Response to Request for Additional Information Co~ncerningthe License Amendment Request for a One-Time lntegrated Leak Rate Test Extension," dated March 27, 2002.

20) RF-Calc., SM-1237, Revision 0, "Surry and North Anna Containment Isolation Modeling," April 20, 2000.

21) RF-Report, "Interim Guidance for Performing Risk lmpact Assessments In Support of One-Time Extensions for Containment lntegrated Leakage Rate Test Surveillance Intiervals," Developed for NEI by EPRI, November 2001.

22) RF-Report, EPRI TR-109325, "Risk lmpact Assessment of Extended Leak Rate Testing Intlervals," December 2003.

23) RF-Report, "Application for Renewed Operating Licenses, Appendix E- Environmental Report, Millstone Power Station, Units 2 and 3," January 2004.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 4 of 30

4. Assumptions The ILIPS3 leakage rate (La) acceptance criteria is defined as:

La= 0.3 percent by weight of containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at calculated peak pressure (Pa).

1. Containment leak rates greater than lLa, but less than 35 La, indicate an impaired containment. Break openings of greater than 0.1-inch and less than 0.78-inch in diameter are considered as small leak rate releases. Break openings of greater than 0.78-inch diameter are considered as large leak rate releases.

2. Containment leak rates greater than 35 La indicate a containment breach. This leak rate is considered "large".

3. Containment leak rates less than l L a indicate an intact containment. This leak rate is considered as "negligible".

4. The maximum containment leakage for Class 1 sequences is 1La.

5. The maximum containment leakage for Class 2 sequences is 35La.

6. The maximum containment leakage for Class 3a sequences is 1OLa.

7. The maximum containment leakage for Class 3b sequences is 35La.

8. The maximum containment leakage for Class 6 sequences is 35La.

9. Because Class 8 sequences are containment bypass sequences, potential releases are directly to the environment. Therefore, the containment structure will not impact the release magnitude.

10.Containment leakage due to Classes 4 and 5 are considered negligible based on the previously approved methodology [I 21.

11.The containment releases are not impacted with time.

12.The containment releases for Classes 2, 7 and 8 are not impacted by the ILRT Type A Test frequency. These classes already include containment failure with release colnsequences equal or greater than those impacted by Type A.

5. Method of Calculation A simlplified bounding analysis approach for evaluating the change in risk associated with increasing the interval from 10-years-to-15-years for Type A test was used. Type A test measures the containment air mass and calculates the leakage from the change in mass over time. This approach is similar to that presented in EPRl TR-104285 [2], NUREG-1493

[7] and the NEI interim guidance report [21]. Namely, the analysis performed examined the MPS3 IPE [5] plant specific accident sequences in which the containment

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 5 of 30 integrity remains intact or the containment is impaired. Specifically, the following were considered:

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

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

Type C test measures component leakage rates across containment isolation valves.

Core damage sequences in which containment integrity is impaired due to containment isolation failures of pathways left "opened" following a plant post-maintenance test. For example, a valve failing to close following a valve stroke test (EPRI TR-104285 Class 6 sequences).

Accident sequences involving containment failure induced by severe accident phenomena (EPRI TR-104285 Class 7 sequences), containment bypassed (EPRI TF1-104285 Class 8 sequences) and large containment isolation failures (EPRI TR-104285 Class 2 sequences). Small containment isolation "failure-to-seal" events (EIPRI TR-104285 Class 4 and 5 sequences) were not accounted for in this evaluation.

These sequences are impacted by changes in Type B and C test intervals, not changes in the Type A test interval.

The siteps taken to perform this risk assessment evaluation are as follows:

Step 1 Quantify the baseline risk in terms of frequency per reactor year for each of the eight accident classes presented in Table 1. Map the Level 3 release categories into 8 release classes defined by the EPRI Report [2]. See Table A-1 of Attachment A.

Step 2 - Develop baseline plant specific person-rem dose (population dose) per reactor year for each of the eight accident classes evaluated in EPRI TR-104285 [2].

Step I3 Evaluate risk impact of extending Type A test interval from 10-to-15 years.

Step 4 - Determine the change in risk in terms of Large Early Release Frequency (LERF) in accordance with RG 1.174 [3].

Step 5 Evaluate the risk impact in terms of ALERF.

Step ti - Determine impact on conditional containment failure probability.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 6 of 30

6. Body of Calculation Step 1 Quantify the baseline risk in terms of frequency per reactor year.

This step involves the review of the MPS3 IPE [5] containment event tree (CET). The CET characterizes the response of the containment to important severe accident sequences.

The CET used in this evaluation is based on important phenomena and systems-related event!; identified in NUREG-1335 [8] and NSAC-159, Volume 2 [9] and on plant features that in~fluencethe phenomena.

As previously described, the extension of the Type A interval does not influence those accident progressions that involve large containment isolation failures, Type B or Type C testing or containment failure induced by severe accident phenomena. As a result, the CET containment isolation model was reviewed for applicable isolation failures and their impact on the overall plant risk.

A review of the containment isolation valves reported in Reference 11 was made. The five issues associated with containment isolation in NUREG-1335 [8] were examined. These issues are:

(1) The identity of pathways that could significantly contribute to containment isolation failure.

(2) The signals required to automatically isolate the containment penetration.

(3) The potential generating signals for all initiating events.

(4) The examination of testing and maintenance procedures.

(5) The quantification of each containment isolation mode.

The containment isolation valves in Reference 11 screened out lines less than 2 inches in diameter. An Expert Panel subcommittee representing Maintenance Rule, Engineering and Opera.tions, evaluated the containment isolation valves at MPS3. The Expert Panel determined that a containment penetration size of 2 inches or less was considered to be non-risk significant.

This ILRT evaluation considers lines between 0.1 inches and 2.0 inches as potential candidates for containment leakage. This is used for the EPRl containment failure classilications. This group consists of all core damage accident progression bins for which a pre-existing leakage in the containment structure (i.e., containment liner) exists. The containment leakage for these sequences can be either small (1La to 35 La ) or large (>35 La).

Externla1 event contributors for all accident classes have been considered in this risk assessment. It has been concluded that external events have little impact on the LERF due to the extension of the containment leak rate testing. See further discussion in Section 6 of this calculation.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 7 of 30 The Level 3 release categories were mapped into 8 release classes (See Table A-1 in Attachment A) as defined in the EPRl Report [2]. These EPRl containment failure classifications are listed below.

EPRl Containment Failure Classifications

- - 1 Classl Containment remains intact including accident sequences that do not lead to containment failure in the long term. The release of fission products (and attendant consequences) is determined by the maximum allowable leakage rate values La, under Appendix J for that plant. The allowable leakage rates (La), are typically 0.1 weight percent of containment volume per day for PWR's (e.g. MPS3 measured at Pa, calculated peak contai~nmentpressure related to the design basis accident). Changes to leak rate testing frequencies do not affect this classification.

- - 2 Classl Containment isolation failures (as reported in the IPEs) include those accidents in which the pre-existing leakage is due to failure to isolate the containment.

These include those that are dependent on the core damage accident in progress (e.g.

initiated by common cause failure or support system failure of power) and random failures to close a containment path. Changes in Appendix J testing requirements do not impact these accidents.

- - 3 Class; Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal (i. e., provide a leak-tight containment) is not dependent on the sequence in progress. This accident class is applicable to sequences involving ILRTs (Type A tests) and potential failures not detectable by LLRTs.

- - 4 Class; Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal is not dependent on the sequence in progress. This class is similar to Class 3 isolation failures, but is applicable to sequences involving Type B tests and their potential failures. These are the Type B- tested components that have isolated but exhibit excessive leakage.

- - 5 Class' Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal is not dependent on the sequence in progress. This class is similar to Class 4 isolation failures, but is applicable to sequences involving Type C tests and their potential failures.

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

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

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 8 of 30 Class;8 Accidents in which the containment is bypassed (either as an initial condition or induced by phenomena) are included in class 8. Changes in Appendix J testing requirements do not typically impact these accidents, particularly for PWRs.

The frequencies for the above eight classes are calculated below. The Class 3-6 frequencies are calculated first since these values are needed to determine the Class 1 frequency.

Class 3 Sequences. This group consists of all core damage accident progression bins for which a pre-existing leakage in the containment structure (i.e., containment liner) exists.

The containment leakage for these sequences can be either small (1La to 35 La ) or large

(>35 La).

To calculate the probability that a liner leak will be large (Event CLASS-3B), use was made of the data presented in NUREG-1493 [7]. The data found in NUREG-1493 states that 144 ILRTs were conducted. The largest reported leak rate from those 144 tests was 21 times the allowable leakage rate (L,). Since 21 Ladoes not constitute a large release (refer to the write-up in Step 4), no large releases have occurred based on the 144 ILRTs reported in NUREG-1493 [7].

An iniprovement in the methodology used to determine the frequencies of leakages detectable only by ILRTs, classes 3a and 3b was made using the methods documented in Reference 21. The method utilized in the aforementioned utility submittals (discussed in Reference 21) involved using a 95% confidence of a x2 distribution of the noted ILRT failures (4 of 144 reported in NUREG-1493). Data collected recently by NEI from 91 nuclear power plants indicates that 38 plants have conducted ILRTs since 1/1/95, with only one failure (due to construction debris from a penetration modification). This would indicate that tlhe statistical information should be based on 51182. Rather than using the x2 distribution used previously, it has been considered more appropriate to utilize the mean (5118;!=0.027) for the class 3a distribution.

The Jeffreys non-informative prior distribution (Reference 21) was used for the class 3b distribution:

NumberojFailures(0) + 1 / 2 Failure Pr obability =

NumberoJTests(l82)+ 1 The nlumber of large failures is zero, so the class 3b probability is 0.51183=0.0027.

Compared with the previous revision of this calculation, the impact of the second improvement on the overall results is small, with larger impact for class 3b than for class 3a.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 9 of 30 The respective frequencies per year are determined as follows:

CLAS'S-3A-FREQUENCY = PROB,1ass-3a* CDF CLAS'S-3B-FREQUENCY = PROBclaSssb

• CDF Where:

PROE3,1as,sa= probability of small pre-existing containment liner leakage = 0.027 PROE3class-3b= probability of large pre-existing containment liner leakage = 0.0027 CDF := 2.88 x 10-5 /year from Reference 17 For this analysis the associated maximum containment leakage for class 3A is 1OLa and for class 3B is 35La.

Class; 4 Sequences. This group consists of all core damage accident progression bins 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 the probability will not be impacted by Type A testing. Because these failures are detected by Type B tests, this group is not evaluated any further, consistent with approved methodology.

Class; 5 Sequences. This group consists of all core damage accident progression bins 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 the probability will not be impacted by Type A testing. Because these failures are detected by Type C tests, this group is not evaluated any further, consistent with approved methodology.

Class; 6 Seauences. This group is similar to Class 2 and addresses additional failure mode:s not typically modeled in PRAs due to the low probability of occurrence. These are sequences that involve core damage accident progression bins for which a failure-to-seal containment leakage due to failure to isolate the containment occurs. These sequences are dominated by misalignment of containment isolation valves following a testlmaintenance evolution.

The low failure probabilities are based on the need for multiple failures, the presence of autorrlatic closure signals, and control room indication. Based on the purpose of this calculation, and the fact that this failure class is not impacted by Type A testing, no further evaluation is needed. This is consistent with the EPRl guidance.

To be consistent with the NEI interim guidance (Ref. 21) a frequency of zero is used for class 6 accidents.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 10 of 30 Class 1 Seauences. This group consists of all core damage accident progression bins for which the containment remains intact. The frequency per year for these sequences is 1.46 x lo-!' /year (Attachment A, Table A-1). For this analysis the associated maximum containment leakage for this group is lLa. The MPS3 IPE did not model Class 3 type failures, therefore this needs to be accounted for in the Class 1 accident class. Using Reference 21 methodology, the frequency for Class 1 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:

Class 2 Seauences. This group consists of all core damage accident progression bins for which a pre-existing leakage due to failure to isolate the containment occurs. The frequency for Class 2 is the sum of those release categories identified in Table A-1 as Class 2.

CLASS-2-FREQUENCY = 0.00 /year [Table A-1]

Class 7 Seauences. This group consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena (Early and Late Failures). The frequency of Class 7 is the sum of those release categories identified in Table A-1 as Class 7.

CLASS-7-FREQUENCY = 1.30 x lom5/ year Class 8 Seauences. This group consists of all core damage accident progression bins in which containment bypass occurs. The frequency of Class 8 is the sum of those release categories identified in Table A-1 as Class 8.

CLASS-&FREQUENCY = 1.22 x low6/ year Note: For this class the maximum release is not based on normal containment leakage, because the releases are released directly to the environment. Therefore, the containment structure will not impact the release magnitude.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 11 of 30 Table 1 Baseline Containment Freauencies Given Accident Class Class Description Frequency (per Rx-year)

No Containment Failure 1.37E-05 Larae Containment Isolation Failures (Failure-to-Close) 0.00E+00 Small Isolation Failures (Type A test) 7.78E-07 Large Isolation Failures (Type A test) 7.78E-08 Small Isolation Failure - Failure-to-Seal ( T v ~ e B test) Not Analvzed Small Isolation Failure - Failure-to-Seal (Type C test) Not Analyzed Containment Isolation Failures (dependent failures, Not Analyzed I I ~ersonnelerrors) I I y7 Severe Accident Phenomena Induced Failure (Early and 1.30E-05 Late Failures)

Containment Bypassed (SGTR & V-Sequence) 1.22E-06 EF: Core Damage All cET End states 2.88E-05 Step 2 - Develop baseline plant specific person-rem dose (population dose) per reactor vear.

Plant-specific MAAPlMACCS2 analysis was performed to evaluate the person-rem dose to the population, within a 50-mile radius from the MPS3 plant. The no containment failure Class 1 dose was used for Class 1 accident release as shown in Table A-1 in Attachment A. The Source Term Category M-4 containment isolation failure accident sequence has characteristics that are representative of an EPRl Class 2 containment leakage.

Using the total population dose for Class 1 accidents as the starting reference point, the Class 3 through 6 accidents are calculated below. The population dose is converted to the corresponding Class value using the appropriate dose multiplier as was used in Reference 12 to [predict the person-rem dose for accident classes 1 to 6 as follows. The dose for the Class 7 accidents was obtained by frequency weighting all the Class 7 dose values and dividing the sum of the products by the sum of the frequencies from Table A-1. The baseli~nedose results are shown below.

Class 1 = 1.65 x 1o4 person-rem Class 2 = 0.00E+00 person-rem Class 3a = 1.65 x 1o4 *I 0 = 1.65 x 1o5 person-rem Class 3b = 1.65 x 1o4 *35 = 5.78 x 10"erson-rem Class 4 = Not analyzed Class 5 = Not analyzed Class 6 = Not analyzed Class 7 = Cn(Freq x Dose)/CnFreq = 5.83 x 1o5 person-rem Class 8 = 1"(Freq x Dose)/C" Freq = 4.10 x 1o6 person-rem Class 8 sequences include containment bypass failures; as a result, the person-rem dose is not based on normal containment leakage. The releases for this class are expected to be

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 12 of 30 released directly to the environment. The Class 8 doses are frequency weighted as were done for class 7. The frequency weighted Class 8 dose from Table A-1 represent the sum of the dose for the Event-V and SGTR sequences.

The above values are summarized in Table 2 below.

Table 2 Baseline Person-Rem Measures Given Accident Class I s I Description 1 Person-Rem t-i- NOcontainment Failure (50-Miles) 1.65 x lo4 Large Containment lsolation Failures (Failure-to-Close) 0.00E+00 Small Isolation Failures (Type A test) 1.65 x l o 5 Large Isolation Failures (Type A test) 5.78 x 1o5 Small Isolation Failure - Failure-to-Seal (Type B test) Not Analyzed Small Isolation Failure - Failure-to-Seal (Type C test) Not Analyzed Other Isolation Failures (e.g., Dependent Failures) Not Analyzed Failure lnduced by Phenomena (Early and Late 5.83 x lo5 Failures)

Containment Bv~assed(SGTR & V-Seauence) 4.10 x l o 6 The above dose results, when combined with the frequency results presented in Table 1, yields the MPS2 baseline mean consequence measures for each accident class. These results are presented in Table 3 below.

Table 3 Baseline Mean Person-Rem Measures Given Accident Class Class Description Frequency Person-Rem Person-(per Rx-yr) (50-Miles) Remlyr (50-Miles) 1 No Containment Failure 1.37E-05 1.65 x 1o4 2.27E-01 2 L.arge Isolation Failures (Failure-to-Close) O.OOE+OO O.OOE+OO 0.00E+00 3a Small Isolation Failures (Type A test) 7.78E-07 1.65 x 10' 1.28E-01 3b L.arge Isolation Failures (Type A test) 7.78E-08 5.78 x 1o5 4.49E-02 4 Small Isolation Failure-to-Seal (Type B test) N/A NIA N/A 5 Small Isolation Failure-to-Seal (Type C test) N/A N/A NIA 6 Other Isolation Failures (e.g., Dependent NIA N/A NIA Failures) 7 Failure Induced by Phenomena (Early and 1.30E-05 5.83 x 1o5 7.57E+00 L.ate Failures) 8 Containment Bypassed (SGTR & V- 1.22E-06 4.10 x lo6 5.00E+00 Sequence)

Total Pd CET End States 2.88E-05 NIA 12.98

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 13 of 30 Based on the above values, using the same methodology as Reference 16, the baseline percent of total dose rate (DR) due to Type A testing is as follows:

Where:

~ S E3a person-redyear = 1.28 x I 0-' person-redyear C L A S S ~ ~ =~ class [Table 31 CLASS^^^^^^ = class 3b person-redyear = 4.49 x 1o - person-redyear

~ [Table 31

= total person-redyear for baseline interval = 12.98 person-redyear [Table 31

% of Total =[(I .28 x 10" + 4.49 x lo-*) 112.981x 100%

% of Total DRBAsE = 1.3%

Therefore, the baseline percent of total dose rate is 1.3%.

Step :3 Evaluate risk impact of extendina T V P A~ test interval from 10-to-15 vears.

The revised methodology in Reference 21 suggests using the following method. It is now believed that the multiplier should be a factor representing the change in probability of leakage. As stated in References 2 and 7 , relaxing the test interval from three in ten years to one in ten years increases the average time that a leak detectable only by an ILRT would go undetected from 18 (3 yrsI2) to 60 (10 yrs12) months. This is a factor of 60/18=3.333.

The blaseline dose associated with the ten-year interval was previously calculated using the percentage increase (lo%), or 1.1 times the baseline dose. Using the 3.33 multiplier would yield a slightly higher ten-year dose. For a 15-year test interval a factor of 90118 = 5 should be applied.

Risk Impact due to 10-year Test Interval As previously stated, Type A tests impact only Class 1 and Class 3 sequences. In addition, the increased probability of not detecting excessive leakage has no impact on the frequency of occurrence for Class 1 sequences. For Class 3 sequences, the release magnitude is not impacted by the change in test interval (a small or large liner opening remairis the same, even though the probability of not detecting the liner opening increases).

Thus, only the frequency of Class 3 sequences is impacted. Therefore, for Class 3 sequences, the risk contribution is determined by multiplying the Class 3 accident frequency by the increase in probability of leakage of 3.33.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 14 of 30 The increased leakage for the 10-year Class 3a and 3b frequencies are obtained by applying the 3.33 multiplier to the base values as shown below.

Class 3a = 7.78 x lo-' *3.33 = 2.59E-06 Class 3b = 7.78 x 1o - *3.33

~ = 2.59E-07 The f~requencyfor Class 1 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:

The results of these calculations are presented in Table 4 below.

Table 4 Mean Consequence Measures for 10-Year Test Interval Given Accident Class Description Frequency Person-Rem Person-Remlyr (per Rx-yr) (50-Miles) (50-Miles) ontainment Failure 1.18E-05 1.65 x lo4 1.94E-01 3 Isolation Failures (Failure-to Close) 0.00E+00 0.00E-tOO 0.00E+00 I Isolation Failures (Type A test) 2.59 E-06 1.65 x 1o5 4.27E-01 3 Isolation Failures (Type A test) 2.59E-07 5.78~10~ 1.50E-01 I Isolation Failure-to-Seal (Type B test) NIA NIA NIA I Isolation Failure-to-Seal (Type C N/A NIA N/A r Isolation Failures (e.g., Dependent NIA N/A NIA res) re Induced by Phenomena (Early and 1.30E-05 5.83 x 1o5 7.57E+00 Failures) ss (SGTR) 1.22E-06 4.10 x lo6 5.00E+00 i T End States 2.88E-05 NIA 13.34E+00 Based on the above values, the Type A 10-year test frequency percent of total dose rate (DR) for Class 3 is as follows:

Where:

CLASS3alo= class 3a person-remlyear = 4.27E-01 person-remlyear [Table 41 CLAS!33blo = class 3b person-redyear = 1.50E-01 person-remlyear [Table 41

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 15 of 30 Totallo = total person-redyear for 10-year interval = 13.34 person-redyear [Table 41 O/o of Total DRlo = [(4.27E-01+ 1.50E-01) / 13.341 x 100

% of Total DRlo = 4.3%

Therefore, the total 10-year test frequency ILRT interval percent of total dose rate due to Type A testing is 4.3%.

The A,% change in the 10-year ILRT DR from the baseline value is 4.3% - 1.3% = 3.0%.

The ten-vear dose rate change (due to a ILRT) over the baseline case is as follows:

DR Clhangelo = [En(Class 1,3a,3b) t o - Cn(Class 113a,3b)B,, ]

Where:

Cn(Class 1,3al3b),,B, = 0.40 person-redyear [Table 31 Cn(Class 113a,3b)lo = 0.77 person-redyear [Table 41 DR Changelo = [0.77 - 0.401 person-remlyear DR CIiangelo = 0.37 person-redyear Therefore, the ten-year dose rate change from the baseline case is 0.37 person-redyear.

Risk Impact due to 15-vear Test Interval The risk contribution for a 15-year interval is similar to the 10-year interval. The difference is in the increase in probability of leakage value. This increase in containment leakage is accounted for by using the multiplier 5 on the Class 3 frequencies.

The increased leakage for the 15 year Class 3a and 3b frequencies are obtained by applying the multiplier 5 to the base values as shown below.

Class 3a = 7.78 x lo-' *5 = 3.89E-06 Class 3b = 7.78 x 1 0 ' ~*5 = 3.89E-07

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 16 of 30 The frequency for Class 1 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 the reflore:

CLASS-1-FREQ = 1.03 x 1 lyear The results of this calculation are presented in Table 5 below.

Table 5 Mean Consequence Measures for 15-Year Test interval Given Accident Class Class I Description I Frequency I Person-Rem I Person-Remlyr ]

Isolation Failures (Type A test) 3.89E-07 5.78E+05 2.25E-01 Small Isolation Failure-to-Seal ( T v ~ eB test) N/A 0.00E+00 N/A 3

Small Isolation Failure-to-Seal (Type C N/A 0.00E+00 N/A Other Isolation Failures (e.g., Dependent NIA 0.00E+00 N/A Failures)

Failure Induced by Phenomena (Early and 1.30E-05 5.83E+05 7.57E+00 Late Failures)

B ass (SGTR) 1.22E-06 4.1OE+06 5.00E+00 CDF All CET End States 2.88E-05 N/A 13.61E+OO Based on the above values, the Type A 15-year test frequency percent of total dose rate (DR) flor Class 3 is as follows:

% of Total DR15= [(CLASS3a15+ CLASS3b15) 1 Totall5] x 100 Where:

CLASS3a15= class 3a person-redyear = 6.42E-01 person-redyear [Table 51 CLAS!S3b15= class 3b person-redyear = 2.25E-01 person-redyear [Table 51 Totall5; = total person-redyear for 10-year interval = 13.61 person-redyear [Table 51 O/O of Total DR15 = [(6.42E-01 + 2.25E-01) / 13.611 x 100

% of Total DR15= 6.4%

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 17 of 30 Therefore, the total 15-year test frequency ILRT interval percent of total dose rate due to Type A testing is 6.4%.

The A% change in the 15-year ILRT DR from the baseline value is 6.4% - 1.3% = 5.1%.

The A% change in the total dose rate between the ten-to-fifteen year interval due to Type A testinig is:

A% C I ~ a n g e=~% ~ .of~Total

~ DR15 - % of Total DR10= 6.4% - 4.3% = 2.1%

The fifteen-vear dose rate channe (due to a ILRT) over the baseline case is as follo~rs:

DR Change15 = [En(Class 1,3a,3b) 15 - En(Class 1,3a,3b),,B, ]

Where:

I n (Class 1,3a,3b) Base = 0.40 person-redyear [Table 31 En(C~~XSS 1,3a,3b) = 1.04 person-redyear [Table 51 DR Changel5 = [ I .04 - 0.401 person-redyear DR Changel5 = 0.64 person-redyear Therefore, the fifteen-year dose rate change from the baseline case is 0.64 person-redyear.

Step '4 - Determine the chancle in risk in terms of Larqe Earlv Release Frequency (LERF)

The one time extension of increasing the Type A test interval involves establishing the success criteria for a large release. These criteria are based on two prime issues:

1) The containment leak rate versus breach size, and

2) The impact on risk versus leak rate.

The containment leak size for the corresponding leak rate was calculated using the same methodology as in Reference 20. The effect of containment leak size on the containment leak rate is shown in Table 6. The leak rates were calculated using a containment design pressure of 45psig with the corresponding saturation temperature and a containment free volume of 2.26E6 ft3. It is to be noted that the MPS3 Technical Specification states that the maxirrum allowable leakage rate (La, at Pa =38.57psig) shall be 0.3% by weight of the containment air per day. In addition, Oak Ridge National Laboratory (ORNL) [14] completed a study evaluating the impact of leak rates on public risk using information from WASH-1400 [15] as the basis for its risk sensitivity calculations (see Figure 1).

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 18 of 30 Based upon the information in Table 6 and ORNL, it is judged that small leaks resulting from a severe accident (that are deemed not to dominate public risk) can be defined as those that change risk by less than 5%. This definition would include leaks of less than 35Wday. Based on the Table 6 data, a 35%/day containment leak rate equates to a diameter leak of approximately 0.8 inches. Therefore, this study defines small leakage as containment leakage resulting from an opening of 0.477 in2 or less, large leakage as greater than 35%/day and negligible leakage as 0.3% /day.

Table 6 Evaluated Impact of Containment Leak Size on Containment Leak Rate

[containment Leak Size I Approximate

- - Containment Leak Rate at Design Pressure Diameter Area Leak Rate inches) (in2) (Ydday)

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 19 of 30 Figure I Fractional Impact on Risk Associated with Containment Leak Rates [I 41 LEAKAGE RATE OF CONTAINMENTBUILDING, L ( M a y )

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 large release due to failure to detect a pre-existing leak during the relaxation period. For this evaluation only Class 3 sequences have the potential to result in large releases if a pre-existing leak were present. Class 1 sequences are not considered as potential large release pathways because for these sequences the contai~nmentremains intact. Therefore, the containment leak rate is expected to be small (less tlhan 2L,). A larger leak rate would imply an impaired containment, such as classes 2, 3, 6 arid 7. Since the ILRT does not impact CDF, the relevant metric is LERF.

Late releases are excluded regardless of the size of the leak because late releases are, by definiti~on,not a LERF event. At the same time, sequences in the MPS3 IPE [5],which result in large releases (e.g., large isolation valve failures), are not impacted because a LERF will occur regardless of the presence of a pre-existing leak. Therefore, the frequency of Class 38 sequences (Table 4) is used as the LERF for MPS3. This frequency, based on a ten-year test interval, is 2.59 xl0-'lyr.

Reg. Guide 1.174 [3] provides guidance for determining the risk impact of plant-specific changles to the licensing basis. Reg. Guide 1.I74 [3] states that when the calculated increase in LERF is in the range of per reactor year to per reactor year, applications will be considered only if it can be reasonably shown that the total LERF is less than per reactor year (Region 11). Calculating the increase in LERF requires determining the impact of the ILRT interval on the leakage probability.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 20 of 30 Step 5 Evaluate the Risk Impact in Terms of A LERF The A LERF from Base to once-per-10 years (10 year metrics) is calculated to be the difference between the Class 3b frequencies in Tables 3 and 4.

A LERF = Class 3b10- Class 3b~ase A LERF = 2.59E 7.78E-08 = 1.8E-7 The baseline total LERF for MPS3 has been calculated to be 3.17 x10-~/yrin Reference 10.

This A LERF increases the baseline LERF to 3.17 x10-7+ 1.81 XI oe7= 5.0 XI ~-~/yr.

The A. LERF from Base to once-per-15 years (15 year metrics) is calculated to be the difference between the Class 3b frequencies in Tables 3 and 5.

A LERF = 3.89E 7.78E-08 = 3.1 E-7 This A LERF increases the baseline LERF to 3.17 x10-7+ 3.1 1 XI oT7= 6.3 x10-~/yr.

The A LERF from once-per-10 years to once-per-15 years (5 year metrics) is calculated to be the difference between the Class 3b frequencies in Tables 4 and 5.

A LERF = 3.89E 2.59E-07 = 1.3E-7 The giuidance in Reg. Guide 1.I74 states that when the calculated increase in LERF is in the ralnge of per reactor year to per reactor year, a plications will be considered 9

only if it can be shown that the total LERF is less than 10- per reactor year. The total LERF for MPS3 has been calculated to be 3.17 x10-~/yrin Reference 10.

Since guidance in Reg. Guide 1.I74 defines small changes in LERF in the range of 10-~/yr, increasing the ILRT interval to 15 years is considered acceptable.

Step Ei - Determine Impact on Conditional Containment Failure Probability Another parameter that the NRC Guidance in Reg. Guide 1.I74 (Ref. 3) states can provide input into the decision making process is the consideration of change in the conditional containment failure probability (CCFP). The change in CCFP is indicative of the effect of the ILRT on all radionuclide releases, not just LERF. The CCFP can be calculated from the risk ca.lculationsperformed in this analysis.

In this assessment, based on the NEI Interim Guidance (Ref. 21), CCFP is defined such that containment failure includes all radionuclide release end states other than the intact state (EPRI Category 1) and small pre-existing leakages (EPRI Category 3a). The conditional part of the definition is conditional given a severe accident (i.e., core damage).

The CCFP percent for a given ILRT interval can be calculated using the following equation from Reference 21.

CCFP<I/, = [1-((Class1 Frequency + Class 3aFrequency)lTotal CDF)] x 100%

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 21 of 30 For the Base interval: The values are obtained from Table 3.

For the 10-year interval: The values are obtained from Table 4.

For the 15-year interval:

The 5-year change (10 to 15 years) in the conditional containment failure probability is:

The 10-year change in the conditional containment failure probability is:

The I!;-year change in the conditional containment failure probability is:

This 15-year change in CCFPo/,is slightly greater than 1 percent is considered to be small from a risk perspective.

Non-Inspected Linear Surface An alternative approach to show that the change in LERF meets the RG 1.I 74 acceptance guideline is to multiply the non-inspected area of the containment by the delta LERF from the 3-in-10 year interval to the 1-in-15 year interval.

The delta LERF is for Class 3b accidents from Tables 1 and 5 only:

ALERF = 3.89 x 10 7.78 x lo-* = 3.1 1 x lo-'

The non-inspected fraction has been calculated using dimensions from Reference 18 (see Attachment B).

Total area of liner = 104,321 ft2 Non-inspectedArea Total Nan-inspected area = 12,824 ft2

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 22 of 30 O/O Not Inspected = (12,82411 04,321) x100 = 12.3%

To account for additional containment liner surfaces that are not accessible inside contai~imentthe total non-inspected surface is rounded up to 13%.

The resulting change in LERF is calculated to be 0.13 x (3.1 1 x = 4.04 x Thus it has been independently shown that the change in LERF due to a 15-year ILRT meets the screening criterion in Reg. Guide 1.174.

External Event Sensitivity Analysis The Millstone Unit 3 IPE (Ref. 5) has some limited discussion pertaining to external events.

Table 1.4-2 in Ref. 5 shows percent contributions from internal, fire and seismic events for all release category frequencies. It is noteworthy to mention that for release category M4 which is the containment isolation failure, has a 89.9% contribution due to seismic, a 0.8%

due to fire and 9.4% contribution due to internal events. The internal events frequency for release category M4 is reported as zero in Table A-1 in Attachment A. It appears that external events would have the largest impact on the EPRl class 7 event for the ILRT evaluation.

For th~esevere accident mitigation alternatives analysis (SAMA) for the MPS3 license renewal (Ref. 23) a factor was used to account for the potential impact of external events.

The benefits of each SAMA were multiplied by a factor of 1.6 to account for the external events. This factor could be applied to the CDF used here to calculate the EPRl class 3a and 3b frequencies. Since class 3b represents a LERF then this multiplier would have the following effect on the ILRT analysis.

Baseline Class 3b frequency = 7.78E-08 x 1.6 = 1.24E-071yr 15 year Class 3b frequency = 3.89E-07 x 1.6 = 6.22E-071yr The external events A LERF from Base to the15-year test is This compares to the internal events Base to 15-year A LERF as 3.11 E-071yr.

Thus, it has been independently shown that with external events included the change in LERF due to a 15-year ILRT still meets the screening criterion in Reg. Guide 1.174. Since guidance in Reg. Guide 1.174 defines small changes in LERF in the range of 10-~/yr, increasing the ILRT interval to 15 years is considered acceptable.

Liner Corrosion Analysis The approach documented in the Calvert Cliffs Nuclear Power Plant submittal in Reference 19 was used to determine the change in likelihood, due to extending the ILRT, of detecting

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 23 of 30 liner corrosion. This likelihood was then used to determine the resulting change in risk.

The fcdlowing 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 half failure is assumed for basemat concealed liner corrosion due to the lack of identified failures. (See Table 7, Step 1.)

The success data was limited to 5.5 years to reflect the years since September 1996 wlhen 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 7, Step 1.)

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 increased 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 7, Steps 2 and 3.)

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 of Reference 19 for sensitivity studies.)

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 7, Step 4.)

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 7, Step 5.) Sensitivity studies are inlcluded that evaluate total detection failure likelihoods of 5%. (See Table 4 of Reference 19 for Calvert Cliffs sensitivity studies.)

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 24 of 30 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.

Table 7

- Liner COI ~ s i o nBase Case Step Description Containment Containment

- Cylinder and Dome Basemat Historical Liner Flaw Likelihood Events: 2 Events: 0 Failure Data: Containment location Assume half a failure specific. (Brunswick 2 and Success Data: Based on 70 steel-lined Containments and 9 years North Anna 2) since the 10 CFR 50.55a requirement for periodic visual inspections of containment

-surfaces.

Aged Adjusted Liner Flaw Year Failure -

Year Failure L.ikelihood -

Rate -

Rate During 15-year interval, assumed 1 2.1 E-3 1 5.OE-4 failure rate doubles every five avg 5-10 5.2E-3 avg 5-10 1.3E-3 years (14.9% increase per year). 15 1.4E-2 15 3.5E-3 The average for 5thto 10th year was set to the historical failure 15-year avg = 6.27E-3 15-year avg = 1.57E-3 risk. (See Table 5 from Ref. 19 for

- example.)

a.n Increase in Flaw Likelihood Eletween 3 and 15 years Uses aged adjusted liner flaw likelihood (Step 2), assuming failure rate doubles every five years. (See Tables 5 and 6 in Ref.

19.)

L.ikelihood of Breach in Pressure Likelihood Pressure Likelihood Containment given Liner Flaw (PW of Breach (psi@ of Breach The upper end pressure is consistent with the Calvert Cliffs 20 0.1%

F'robabilistic Risk Assessment 64.7 (ILRT) 1.1%

(13RA) Level 2 analysis. 0.1% is 100 7.02%

assumed for the lower end. 120 20.3%

Intermediate failure likelihoods are 150 100%

determined through logarithmic interpolation. The basemat is assumed to be 1110 of the cylinderldome analysis.

-The same value will be used for

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 25 of 30 Step Description Containment Containment Cvlinder and Dome Basemat MPS3 as was used for CCNP, since the containment design is somewhat similar. The design pressure of MPS3 is 45 psig "ersus 50 psig for CCNPP.-

visual Inspection Detection F:ailure~ikelihood 5% failure to identity Cannot be visually visual flaws plus 5% inspected.

likelihood that 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.

L,ikelihood of Non-Detected 0.0096%

Containment Leakage

$Steps 3*4*5)

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% =

0.0123h.

The non-large early release frequency (LERF) containment over-pressurization failures for MPS3 is estimated at 1.31E-5 per year. This is based on the total CDF minus the Class 1, 38 and 8 frequencies from Table 1 (1.31 E-5 = 2.88E (1.37E-05 + 7.78E-08 + 1.22E-06)). The total CDF for MPS3 is 2.88E-5. 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) = 0.00012

• 1.31E-5 = 1.57E-9 per year.

Thus, it has been independently shown that the increase in LERF, due to a liner corrosion failure, is 1.57E-9 per year which meets the screening criterion of less than E-7 in Reg.

Guide 1.174.

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 26 of 30 Results Summary The results for the Baseline, 10-year and 15-year ILRT evaluation are summarized in Table 8 below.

Serial No. 06-31 5 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 27 of 30 Table 8: Summarv of Results EPRl Base Base Base re^

10yr 15yr II 15yr II 15yr Class 1

2 Freq IRx-yr 1.46E-05 0.00E+00 Dose Person-Rem 1.65E+04 (3110)

Freq

/RX-yr 1.37E-05

(

D~;e

t Person-2.27E-01

/Rx-yr 1.18E-05 I~ose~;te Person-Reml r 1.94E-01 Metrics I

Freq

/RX-yr Dose Rate Person-Metrics 0.00E+00 0.00E+00 0.00E+00 O.OOE+O 0.00E+00 3b N/A 5.78E+05 7.78E-08 4 N/A 0.00E+00 0.00E+00 5 N/A 0.00E+00 0.00E+00 6 N/A 0.00E+00 0.00E+00 7 1.30E-05 5.83E+05 1.30E-05 8 1.22E-06 4.1 0E+06 1.22E-06 ~ ~ - -

2.88E-05 2.88E-05 2.88E-05 1 13.61 (CDF) (CDF) (CDF)

ADR Change from Base I ILRT DR I

% of total I I I dose A%

Change in ILRT DR from Base I

I LERF 1 A LERF I from base CCFP, %

A CCFP, %

CDF 2.88E-05

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 28 of 30 Identified Conservatisms in the ILRT Analysis The results in Table 8 have shown that the 10 and 15 year metrics for some parameters are in line with expectations and others that do not meet expectations. As a result, it is important to analyze the results and to identify some conservatisms that exist in the methodology adopted for this ILRT extension analysis. These conservatisms include:

A first conservatism exists in the conservatively high CDF that was used in this analysis. If the updated PRA model CDF (which is less) was used in this analysis, the resulting Class 3a and 3b frequencies would decrease the 10 and 15 year metrics proportionately.

A second conservatism exists in the conservatively high baseline LERF that was used in this analysis. If the updated PRA model LERF (which is less) was used in this analysis, the baseline LERF would decrease the 10 and 15 year metrics proportionately.

7. Design Review The most current drawings and procedures were used. This calculation does not perform a design verification or affect the design of the plant.

8. Enclosure -A: MPS3 Frequency and Dose Data

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 29 of 30 ENCLOSURE 1-A MPS3 FREQUENCY AND DOSE DATA

Serial No.06-315 Docket No. 50-423 Five Year Extension of Type A Test Interval Attachment 2 Page 30 of 30 Table A-1 I Dose Data Release Frequency* EPRl 1 Description Category Per year Class 1 M-1.A 2.21 E-07 Cont Bypass, V-7 Early Cont Failure Early Melt, No Sprays 7 Early Cont Failure Late Melt, No Sprays 2 Cont Iso Failure I lntermediate Cont Fail Late Melt. No S ~ r a v s 7 Intermediate Cont Fail Early Melt, No Sprays 7 Late Cont Fail No S ~ r a v s 7 Intermediate Cont Fail With Sprays 7 Late Cont Fail With Sprays 7 Basemat Failure No Sprays 7 Basemat Failure With Sprays 1 No Cont Failure

• • Ref. 6

Serial No.06-315 Docket No. 50-423 ATTACHMENT 3 PROPOSED RISK-INFORMED TECHNICAL SPECIFICATIONS CHANGE FIVE-YEAR EXTENSION OF TYPE A TEST INTERVAL MARK-UP OF TECHNICAL SPECIFICATIONS DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

BATTVF CnNTRnl S PROCEDURES AND PROGRAMS (Continued)

2) P r e - p l anned o p e r a t i n g procedures and backup i n s t r u m e n t a t i o n t o be used i f one o r more m o n i t o r i n g instruments become inoperable, and

3) A d m i n i s t r a t i v e procedures f o r r e t u r n i n g Jnoperabt e instruments t o OPERABLE s t a t u s as soon as practicable.

f. Containment Leakaqe Rate Testing Proqram A program s h a l l be e s t a b l i s h e d t o implement the leakage r a t e t e s t i n g of t h e containment as r e q u i r e d by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions'. This program s h a l l . be i n accordance w i t h t h e guide1 i n e s contained i n Regul a t o r y Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1 9 9 3 T h e peak c a l c u l ated containment i n t e r n a l pressure f o r the design b a s i s l o s s o f c o o l a n t accident, Pa, i s 38.57 psig.

The maximum allowable containment leakage r a t e L,, a t Pa, s h a l l be 0.3 percent by weight' o f the containment a i r per 2 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Leakage r a t e acceptance c r i t e r i a are:

1) Containment o v e r a l l leakage r a t e acceptance c r i t e r i o n is

-< 1.0 La. During t h e f i r s t u n i t s t a r t u p f o l l o w i n g t e s t i n g i n accordance w i t h t h i s program, t h e leakage r a t e acceptance c r i t e r i a are < 0.60 1, f o r t h e combined Type B and Type C t e s t s , and 5 0.042 La f o r a l l p e n e t r a t i o n s t h a t a r e Secondary Containment bypass leakage- paths, and t0.75 La f o r Type A tests;

2) A i r lock t e s t i n g acceptance c r i t e r i a are:

a. Overall a i r l o c k leakage r a t e i s 0.05 La when t e s t e d a t 2 Pa.

b. For each door, seal leakage r a t e i s < 0.01 La when p r e s s u r i z e d t o 2 Pa.

The p r o v i s i o n s o f S p e c i f i c a t i o n 4.0.2 do n o t a p p l y t o the t e s t frequencies s p e c i f i e d i n the Containment Leakage Rate T e s t i n g Program.

ification o n t a i nment Leakage Rate T e s t i n g -

CJ M j approved by t h e NRC on December 6, 1985.

UTI I C T ~ M C .- IIMTT 2 6-17 Amendment Wn. l g . 186

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/LBDCR 06-MP3-010)

RETYPED TECHNICAL SPECIFICATIONS PAGE DOMINION NUCLEAR CONNECTICUT, INC MILLSTONE POWER STATION UNIT 3

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

2) Pre-planned operating procedures and backup instrumentation to be used if one or more monitoring instruments become inoperable, and

3) Administrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable.

f. Containment Leakage Rate Testing:Promam A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions*. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, as modified by the following exception to NEI 94-01, Rev. 0, "Industry Performance Based Option of 10 CFR Part 50 Appendix J": The first Type A test performed after the January 6, 1998 Type A test shall be performed no later than January 6, 2013.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 38.57 psig.

The maximum allowable containment leakage rate La, at Pa, shall be 0.3 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Leakage rate acceptance criteria are:

1) Containment overall leakage rate acceptance criterion is I 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 I 0.042 La for all penetrations that are Secondary Containment bypass leakage paths, and < 0.75 La for Type A tests;

2) Air lock testing acceptance criteria are:

a. Overall air lock leakage rate is I0.05 La when tested at 2 P,.

b. For each door, seal leakage rate is < 0.01 La when pressurized to 2 Pa.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

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

• An exemption to Appendix J, Option A, paragraph III.D.2(b)(ii), of 10 CFR Part 50, as approved by the NRC on December 6, 1985.

MILLSTONE - UNIT 3 6-17 Amendment No. fi9,4-86,