ML063330085

Response to Request for Additional Information Re Integrated Leakage Rate Test Interval

ML063330085
Person / Time
Site:	Millstone
Issue date:	11/27/2006
From:	Gerald Bichof Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
06-315A
Download: ML063330085 (13)

Text

Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, Virgini.3 2.3060 U'ch Address: www.dom.com November 27, 2006 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 1 1555 Rockville Pike Rockville, MD 20852-2738 Serial No.

06-31 5A NSS&UDF RO Docket No.

50-423 License No.

N PF-49 DOMINION NUCLEAR CONNECTICUT. INC.

MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION INTEGRATED LEAKAGE RATE TEST INTERVAL In a letter dated June 14, 2006, Dominion Nuclear Connecticut, Inc. (DNC) submitted a request to extend the test interval for the integrated leakage rate test for Millstone Power Station Unit 3 (MPS3). In a facsimile dated September 29, 2006, the NRC staff forwarded a request for additional information (RAI) in order to complete its review of DNC's request. The response to the RAI is provided in the attachment to this letter.

The additional information provided in this letter does not affect the conclusions of the significant hazards consideration discussion in the DNC letter dated March 28, 2006.

If you have any questions in regard to the responses provided, or require additional information, please contact Mr. Paul R. Willoughby at (804) 273-3572.

Very truly yours, Vice President - Nuclear Engineering Commitments in this letter: None Attachments: (1 )

Serial No. 06-31 5A Docket Nos. 50-423 Response to Request for Additional Information Page 2 of 3 cc:

U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1 41 5 Mr. V. Nerses Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North 1 1555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852-2738 Mr. 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 061 06-51 27

Serial No. 06-31 5A Docket Nos. 50-423 Response to Request for Additional Information Page 3 of 3 COMMONWEALTH OF VIRGINIA

)

COUNTY OF HENRICO

)

1 The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Gerald T. Bischof, who is Vice President

- Nuclear Engineering 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 27 day of n-*

,2006.

My Commission Expires:

3 1. d( ood.

a&

Notary Public (SEAL)

Serial No. 06-31 5A Docket No.05-423 ATTACHMENT 1 INTEGRATED LEAKAGE RATE TEST INTERVAL LICENSE AMENDMENT REQUEST RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No. 06-31 5A Docket No. 50-423 Response to Request for Additional Information Page 1 of 9 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION INTEGRATED LEAKAGE RATE TEST INTERVAL LICENSE AMENDMENT REQUEST In a letter dated June 14, 2006, Dominion Nuclear Connecticut, Inc. (DNC) submitted a request to extend the test interval for the integrated leakage rate test for Millstone Power Station Unit 3 (MPS3). In a facsimile dated September 29, 2006, the NRC staff forwarded a request for additional information (RAI) in order to complete its review of DNC's request. The response to the RAI for MPS3 is provided in the balance of this attachment.

NRC Question No. 1.

The population dose for Class 1 accidents (no containment failure) at MPS3 (1.65E-t4 person roentgen equivalent man (rem) per event) is at the high end of the range of values reported for intact containment release classes in other studies, including Level 3 analyses submitted as part of recent license renewal applications.

Population doses for accidents with an intact containment are typically on the order of 1000 person-rem per event. The impact of the proposed integrated leakage rate test (ILRT) extension on population dose could be substantially overstated as a result of this apparent, albeit conservative, bias in the dose estimate.

Provide additional justification for the population dose estimate used for Class 1 accidents in the ILRT analysis in view of the aforementioned disparity.

DNC Res~onse The population dose for Class 1 accidents (intact containment) at Millstone Power Station Unit 3 (MPS3) was calculated to be 1.65E+04 person-rem. The MPS3 technical specification maximum allowable leakage rate is 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 />. The source term release fractions for the population dose evaluation were calculated using the MAAP4 computer code based on the technical specification maximum allowable leak rate. The MAAP4 model used a conservative containment leakage methodology for an intact containment.

This conservative leak rate methodology was also used in the MPS3 license renewal severe accident mitigation alternatives analysis (SAMA).

The use of a conservative containment leakage methodology in this application for an intact containment does not impact the overall conclusions.

NRC Question No. 2.

The total large early release frequency (LERF) for MPS3 prior to the requested change is stated to be 3.17E-7 per year (page 20 of Attachment 2 to the June 14, 2006, request). However, this value does not include the contribution to LERF

Serial No. 06-31 5A Docket No. 50-423 Response to Request for Additional Information Page 2 of 9 from external events. As stated in Section 2.2.4 of Regulatory Guide 1.I 74, the risk-acceptance guidelines (in this case, for LERF) are intended for comparison with a full-scope risk assessment, including internal and external events.

Consistent with this guidance, and to the extent supportable by the available risk models for MPS3, provide an estimate of the total LERF when external events and the impact of the requested change are included within the assessment.

DNC R ~ S D O ~ S ~

The submittal included a brief discussion of the impact of the contribution of external events on the LERF.

An external events model is not available for MPS3. Consistent with the MPS3 license renewal SAMA, a conservative factor of 1.6 was used to account for the potential impact of external events based on review of the Individual Plant Examination (IPE) and Individual Plant Examination of External Events (IPEEE) results. EPRl class 3b frequency represents LERF in this evaluation. This factor is applied to the EPRl class 3b frequency to estimate the change in LERF.

Baseline Class 3b frequency = 7.78E-08 x 1.6 = 1.24E-07 Iyr 15 year Class 3b frequency = 3.89E-07 x 1.6 = 6.22E-07 Iyr ALERF = 6.22E 1.24E-07 = 4.98E-07Iyr The ALERF from the baseline to 15 year frequency accounting for external events yields a value of 4.98E-071yr. This ALERF is on the same order of magnitude as that reported in the submittal.

Reg. Guide 1.I 74 defines both these changes in LERF as "small" and acceptable.

NRC Question No. 3.

The discussion of conservatisms in the ILRT analysis (page 28 of Attachment 2 to the June 14, 2006, request) implies the existence of a more recent version of the MPS3 probabilistic risk assessment (PRA) than the October 2002 version on which the ILRT analysis is based. Describe the major differences between the October 2002 version and the updated version of the PRA, including changes to models and assumptions, and estimated results for core damage frequency and LERF. Provide an expanded discussion of the impact on ALERF and total LERF if the ILRT analysis was based on the more recent PRA.

DNC R ~ S D O ~ S ~

The 2002 PRA model was used for the MPS3 ILRT evaluation. The 2004 PRA model was not used in the ILRT evaluation because the LERF portion of the 2004 PRA model had not been updated at the time of the submittal to support changes made to the 2004 PRA Level 1 internal events model.

Serial No. 06-315A Docket No. 50-423 Response to Request for Additional Information Page 3 of 9 A list of the major changes made to the 2002 PRA model included in the 2004 PRA model is provided below.

Addition of a main feedwaterlcondensate model for steam generator cooling Modification of the anticipated transient without scram (ATWS) coincident with steamline break sequence Modification of turbine plant component cooling water (TPCCW) model for isolation on a containment depressurization actuation (CDA) signal rather than safety injection signal (SIS)

The CDF calculated for the 2004 PRA model is lower than the CDF calculated for the 2002 PRA model.

Model CDF (lvr) 2002 2.88E-05 2004 1.49E-05 A comparison of the 2002 and 2004 PRA model LERF values is shown below.

Model Base Case 10vr ILRT Freq 15vr ILRT Freq 2002 3.1 7E-07 4.98E-07 6.28E-07 2004 3.1 1 E-07 4.05E-07 4.72E-07 The comparison above shows that the 2004 PRA model yields slightly lower 10 and 15 year LERF values.

NRC Question No. 4 Section 4.4 of the submittal includes a brief description of the containment inservice inspection (ISI) program being implemented at MPS3. Please provide a schedule and description of the IS1 methods used to provide assurance that, in the absence of a containment ILRT for 15 years, the containment structural and leak-tight integrity will be maintained.

DNC Response The Containment In-service Inspection Program implementation meets the requirements of 10 CFR 50.55a and Subsections IWE & IWL of the ASME Code,Section XI, to uncover any evidence of structural damage or degradation which could affect containment structural integrity or leak-tightness.

IWE visual examinations include visual examinations of the accessible portions of the metallic liner, associated integral attachments, pressure retaining welds, moisture

Serial No. 06-31 5A Docket No. 50-423 Response to Request for Additional Information Page 4 of 9 barrier materials, and penetration liners.

IWL visual examinations include accessible Class CC surfaces and components.

The IWE general visual examination method is employed for the following items:

a. accessible surface areas of the metal liner and integral attachments and mechanical and electrical penetrations, and pressure-retaining portions of the personnel and equipment hatches in the scope of Subsection IWE.

b. moisture barrier materials in areas depicted in Figure IWE-2500-1.

General visual examinations for IWE Category E-A items may be performed directly or remotely, depending on accessibility of the item.

The IWE detailed visual examination method is em~loved to:

a. assess the initial condition of a surface requiring augmented examination in accordance with IWE-1241,

b. determine the magnitude and extent of deterioration and distress of a suspect containment surface initially detected by general visual examination, and

c. assess the structural condition of an area affected by repairlreplacement activity.

The criteria applied to detailed visual examinations is specified in IWE-231 O(e).

The IWL aeneral visual examination method is em~loved for the accessible portions of the outer concrete for the Dureose of identifvina the followina:

a. areas of concrete deterioration and distress as defined in ACI-201.I (IWL-251 O(a)), and ACI-349.3R.

General visual examinations for IWL Category L-A concrete surfaces may be performed directly or remotely, depending on accessibility of the examination surface.

The IWL detailed visual examination method is emdoved to determine the maanitude and extent of deterioration and distress of susDect concrete surfaces, includina:

a. suspect exterior concrete surface area initially detected by general visual examination, and

b. concrete surfaces affected by repairlreplacement activity.

Serial No. 06-31 5A Docket No. 50-423 Response to Request for Additional Information Page 5 of 9 Both direct and remote visual examination techniaues are used, direct visual beina the referred method.

The examination is performed from floors, platforms, walkways, ladders, or other temporary or permanent vantage points. Lighting and distance conditions are acceptable when the conditions or indications for which the visual examination is performed can be detected at the chosen distance and illumination.

During the initial baseline examination(s), adequacy of lighting and distance was verified by demonstrating the ability to resolve a 1/32" black line on an 18%

neutral gray card for general visual examination, or a 1/64" black line on an 18%

neutral gray card for detailed visual examination. The program manual and implementing procedures have since been revised to require adequacy of lighting and distance to be verified by demonstrating the ability to resolve meeting the character height requirements of Table IWA-2010-1 (0.044" for VT-1 and 0.1 05" for VT-3).

Remote examinations use supplemental visual aids, such as binoculars, transits, telescopes, fiber optics, cameras, or other suitable instruments, provided such systems have a resolution capability at least equivalent to that attainable by direct visual examination. The adequacy of lighting and distance for remote examination is verified in the same manner as for direct visual examination.

Although the Containment In-Service Inspection Program Manual does allow visual examinations by the Responsible Engineer (IWE) and the Responsible Professional Engineer (IW L), all examinations following the initial baseline(s) have been performed by visual examiners certified IWA ANSTIASNT CP-189 supplemented with additional training specific to IWE and IWL.

Engineering evaluations may provide a basis for accepting a flaw or degradation in lieu of immediate repairlreplacement. The Responsible Engineer (IWE) or the Responsible Professional Engineer (IWL) performs these evaluations.

Surface areas that are likely to experience accelerated degradation and aging, for which detailed visual inspection and/or ultrasonic thickness measurements may be required, have augmented inspections performed in accordance with IWE-2500.

Per IWL-3300, items not meeting acceptance standards of IWL-3100 or IWL-3200 must be documented. The documentation must include the cause of the condition, the applicability of the condition to other plants at the same site, and the acceptability of the concrete containment without repair of the item, repair/replacement requirements, and the extent, nature, and frequency of additional examinations. Such areas are subject to additional examinations

Serial No. 06-31 5A Docket No. 50-423 Response to Request for Additional Information Page 6 of 9 when specified by the Responsible Professional Engineer. The engineering evaluations at MPS3 address the requirements in IWL-3300, including whether the indication is confined to a local area or exhibits potential to exist in inaccessible areas.

The interval for IWE-2000 examinations and inspections is September 9, 1998 to September 9, 2008. The effective start date of the first IWE period in the first interval has been determined based on the mandated finish date specified in 10 CFR 50.55a(g)(6)(ii)(B). The three periods within this interval start and end on the following dates:

First IWE Period September 9, 1998 to September 9,2001 Second IWE Period September 9,2001 to September 9,2005 Third IWE Period September 9, 2005 to September 9, 2008 The next three periods for the second 10 year interval will be as follows:

First IWE Period September 9, 2008 to September 9,201 1 Second IWE Period September 9,201 1 to September 9, 201 4 Third IWE Period September 9, 201 4 to September 9,201 8 As allowed by IWA-2430(d)(l), adjustments to each interval may be made, as long as requirements in IWA-2430(d)(2) through IWA-2430(d)(4) are met. As allowed by IWA-2430(e), adjustments to each interval may be made as a result of an extended outage.

Millstone Unit 3 has completed the First interval First and Second Period Examinations. The Third Period Examination is scheduled for 3R11 (Spring of 2007).

In accordance with IWL-2410 (Concrete), the effective startlfinish date of the MPS-3 IWL interval shall be as follows, based on a March 28, 1985 completion date of the Structural Integrity Test:

First IWL Interval March 28, 1999 to March 28, 2001 Second IWL Interval March 28,2004 to March 28,2006 The First and Second lnterval Examinations have been completed.

Third IWL Interval March 28, 2009 to March 28, 201 1 Fourth IWL Interval March 28, 201 1 to March 28, 201 6

Serial No. 06-31 5A Docket No. 50-423 Response to Request for Additional Information Page 7 of 9 NRC Question No. 5 The third paragraph of Section 4.4, lists IWE and IWL activities conducted by DNC personal (sic), as stated in the submittal:

DNC Engineering performs IW EIIWL 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:

The subsequent discussion only discusses IWE activities. Describe IWL actions and inspections that have identified problematic areas (such as significant cracking, spawling of concrete) and the disposition required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)

Section XI.

MPS3 has a conventionally reinforced concrete containment surrounded by an enclosure building.

The first and second interval IWL outer concrete examinations for MPS3 are complete. They were implemented in August of 2001 and March of 2006. Some horizontal seam and vertical surface cracks with measurable widths but with little to no depth were identified. These cracks were limited in number and not unexpected as they were previously identified during the initial ILRT. The cracks were recorded and detailed by the visual examiners and were conservatively classified as suspect areas and accepted by engineering evaluation. The cracks are being tracked for future examination.

None of the indications required rework.

The first and second interval examinations resulted in no identified degradation or active mechanisms that would prevent the containment from performing its safety function.

Due to the lack of permanent vantage points from inside the enclosure building proper, additional rigging from the upper enclosure building structural beams is used to provide direct access to the outer portions of the concrete surface. The visual examiners use a bosun's chair and separate safety lines for visual examinations. Where the visual examinations cannot be performed by this method, the enclosure building girts are used to access surfaces for direct and remote visual examinations.

Outer concrete examinations at other building areas, such as the engineered safety features building, auxiliary building, main steam valve building, and hydrogen recombiner building, are accessible from vantage points and through the use of extension ladders and step ladder(s).

Serial No. 06-31 5A Docket No. 50-423 Response to Request for Additional Information Page 8 of 9 NRC Question No. 6 For the examination of penetration seals and gaskets, and examination and testing of bolted connections associated with the primary containment pressure boundary (Examination Categories E-D and E-G), the licensee requested relief from the requirements of Section XI of the ASME Code, 1992 Edition, 1992 Addenda. As an alternative, the licensee proposed to examine the above items under the 1998 Edition, during the leak-rate testing of the primary containment.

Option B of Appendix J for Type B testing and Type C testing (per Nuclear Energy Institute 94-01 and Regulatory Guide 1.1 63), and the ILRT extension requested in this amendment for Type A testing, provide flexibility in the scheduling of these inspections. Discuss your schedule for examination and testing of seals, gaskets, and bolted connections as modified by Title 10 of the Code of Federal Regulations Section 50.55a(b)(2)(ix) that provide assurance regarding the integrity of the containment pressure boundary.

MPS3 Technical Specifications have been amended to apply the performance based Option B of 10CFR50, Appendix J.

The industry guideline allows extension of the interval to once in 10 years (120 months) for Type B electrical penetrations.

It is assumed that a portion of the penetration(s) is tested periodically during the 120 months.

At MPS3, a sample of approximately 25% (four groups) of the electrical penetrations is tested at the refueling outage frequency (1 8 months) even though the testing is routinely performed online.

Conservatively, each electrical penetration is Type B tested approximately every 6 years.

Other Type B penetrations include the fuel transfer tube and bellows, equipment hatch, and personnel airlock. Since each of these penetrations is open each refueling outage, the Appendix J testing is performed prior to re-establishing containment integrity following the refueling outage. Additionally, the personnel airlock door seals are verified leak tight following online containment entries.

At the equipment hatch and personnel airlock, all bolted connections are accessible and inspected each inspection period (three times per 10-year interval) per the requirements of Category E-A of Table IWE-2500-1, whether assembled or disassembled. This exceeds the once per interval requirement of the 1992 edition.

Serial No. 06-31 5A Docket No. 50-423 Response to Request for Additional Information Page 9 of 9 In addition, indications of damage on assembled bolted connections require connection disassembly for detailed visual inspection.

Existing station maintenance procedures are relied upon to ensure that the integrity of re-assembled bolted connections is maintained.