ML073511682

Proposed License Amendment Request, One-Time Five-Year Extension to Type a Test Interval

ML073511682
Person / Time
Site:	Surry
Issue date:	12/17/2007
From:	Matthews W Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
07-0802, RG-1.163
Download: ML073511682 (59)

Text

10 CFR 50.90 VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 December 17, 2007 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Serial No.

NL&OS/ETS Docket No.

License No.

07-0802 RO 50-281 DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

SURRY POWER STATION UNIT 2 PROPOSED LICENSE AMENDMENT REQUEST ONE-TIME FIVE-YEAR EXTENSION TO TYPE A TEST INTERVAL Pursuant to 10 CFR 50.90, Dominion requests an amendment, in the form of a change to the Technical Specifications to Facility Operating License Number DPR-37 for Surry Power Station Unit 2. The proposed change will permit a one-time five-year exception to the ten-year frequency of the performance-based leakage rate testing program for Type A tests as required by Regulatory Guide (RG) 1.163. This one-time exception to the requirement of RG 1.163 will allow the next Type A test to be performed no later than October 26,2015.

Attachments 1 and 2 provide an evaluation of and the risk assessment for the proposed

change, respectively.

The marked-up and proposed Technical Specifications pages are provided in Attachments 3 and 4, respectively.

The proposed change has been reviewed and approved by the Facility Safety Review Committee.

To permit effective outage planning, Dominion requests approval of the proposed Technical Specification change by March 31, 2009. Upon issuance, the amendment will be implemented within 60 days.

Serial No. 07-0802 Docket No. 50-281 One-Time Five-Year Extension to Type A Test Interval Page 2 of 3 If you have any questions or require additional information, please contact Mr. Gary Miller at (804) 273-2771.

Very truly yours,

~~u/~

William R. Matthews Senior Vice President - Nuclear Operations Attachments

1. Evaluation of Proposed License Amendment

2. Probabilistic Risk Assessment

3. Marked-up Technical Specifications Page

4. Proposed Technical Specifications Page Commitments made in this letter: None.

COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by William R. Matthews, who is Senior Vice President -

Nuclear Operations, of Virginia Electric and Power Company.

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 /7,1;,,-- day of &~

,2007.

My Commission Expires:

~i J~ Jot)8.

(SEAL)

MMCaARET *. IENNITT Notary PublIc s~42>O" Commonwealth of Virginia My Commtlllon Explrel Aug 31. 2001

~~.,~

NOtary Public

Serial No. 07-0802 Docket No. 50-281 One-Time Five-Year Extension to Type A Test Interval Page 3 of 3 cc:

U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 State Health Commissioner Virginia Department of Health James Madison Building - 7th floor 109 Governor Street Suite 730 Richmond, Virginia 23219 Mr. C. R. Welch NRC Senior Resident Inspector Surry Power Station Mr. S. P. Lingam NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 0-8 G9A 11555 Rockville Pike Rockville, Maryland 20852 Mr. R. A. Jervey NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 0-8 G9A 11555 Rockville Pike Rockville, Maryland 20852

Serial No. 07-0802 Docket No. 50-281 One-Time Five-Year Extension to Type A Test Interval Evaluation of Proposed License Amendment Surry Power Station Unit2 Virginia Electric and Power Company (Dominion)

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 1 of 17 EVALUATION OF PROPOSED LICENSE AMENDMENT

3.0 BACKGROUND

3.1 10 CFR 50, Appendix J, Option B Requirements 3.2 Reason for Proposed Amendment

4.0 TECHNICAL ANALYSIS

4.1 Implementing 10 CFR 50, Appendix J, Option B 4.2 Surry Integrated Leak Rate Test History 4.3 Description of Containment 4.4 Containment Leakage Consideration for Operability 4.5 Containment Operational Performance 4.6 IWE/IWL Inservice Inspection (lSI) activities to support TYPE A 5.0 PLANT SPECIFIC RISK ASSESSMENT FOR THE EXTENDED TYPE A TEST INTERVAL 5.1 Method of Analysis 5.2 Conclusions 6.0 EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION 7.0 IMPLEMENTATION OF THE PROPOSED CHANGE

9.0 REFERENCES

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 2 of 17

1.0 INTRODUCTION

Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion) requests a change to the Surveillance Requirements in Section 4.4 of the Technical Specifications for the containment structure.

The proposed change will permit a one-time five-year exception for Surry Unit 2 from the requirement of Regulatory Guide (RG) 1.163 (Reference 3) and NEI 94-01 (Reference 1), which specify performance of a Type A test at a frequency of up to ten years with allowance for a fifteen-month extension.

2.0 PROPOSED CHANGE

This application for amendment to the Surry Technical Specifications proposes to revise the Technical Specification Surveillance Requirement in Section 4.4.8.1, Containment Leakage Rate Requirements.

The exception is to allow Type A testing within fifteen years from the last Type A, performed on October 26,2000. This application represents a cost beneficial licensing change.

From a differential safety benefit perspective, the improvement by performing the integrated leak rate test within ten years rather than fifteen years is not commensurate with the significant additional cost associated with the test frequency. The specific change to TS Section 4.4.8.1 (page 4.4.1) is as follows:

Revise the current Unit 1 exception from NEI 94-01 from Unit 1 to Unit 2:

"NEI 94 1995, Section 9.2.3: The first Unit 2 Type A test performed after the October 26, 2000 Type A test shall be performed no later than October 26, 2015."

3.0 BACKGROUND

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.

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, Band C tests.

Frequency is based upon an evaluation which looks at the "as 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

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 3 of 17 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 RG 1.163 and 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 testing frequency to one per twenty years was found to lead to an imperceptible increase in risk. The estimated increase in risk is small because Type A tests identify only a few potential leakage paths that cannot be identified by Type Band 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 Type A testing had minimal impact on public risk."

"While Type Band 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 smaiL" Exceptions to the requirements of RG 1.163 and NEI 94-01, 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 8.

The Surry Power Station Unit 2 current ten-year Type A test interval ends on October 26, 2010.

In order to meet the interval requirements of NEI 94-01, this test must be performed during refueling outage (S2-R-22), scheduled to commence in October of 2009.

By granting the proposed one-time exception, Surry would benefit by not having to perform the Type A test for an additional five years.

Direct cost savings are estimated at $200,000 for equipment and materials and $122,500 in expended man hours for containment setup, valve lineups, etc. as a result of elimination of the actual performance of the test.

In addition, approximately 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> of critical path outage time can be eliminated by not performing the Type A test. The critical path time is estimated at a savings of $1.3 million.

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 4 of 17

4.0 TECHNICAL ANALYSIS

4.1 Implementing 10 CFR 50, Appendix J, Option B 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 50, 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."

Amendments 208/208 (Reference 2) were issued to Surry Power Station Units 1 and 2 to permit implementation of 10 CFR 50, Appendix J,

Option B.

Amendments 208/208 modified Technical Specification Section 4.4 to require testing in accordance with the Containment Leakage Rate Testing Program and RG 1.163 (Reference 3), respectively.

RG 1.163 specifies a method acceptable to the NRC for complying with Option B by approving the use of NEI 94-01 and ANSI/ANS 56.8-1994 (Reference 4), subject to several regulatory positions in the guide.

Exceptions to 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."

Therefore, this application does not require an exemption to Option B.

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 performance leakage rate was less than 1.0La and performance factors were considered consistent with NEI 94-01).

Based on the May 1991 and October 2000 Type A tests, the current interval for Surry Unit 2 is once every ten years.

The extension requested for Surry Unit 2 only applies to the 10 CFR 50, Appendix J, Type A test.

Appendix J, Type B and Type C tests are performed at the intervals required by Appendix J, Option B.

The current rule for Type B requires completion of electrical penetrations within 120 months. Some portion of other required Type B tests are conducted each refueling, and are completed in approximately eO-month intervals consistent with the Type C testing requirements.

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 5 of 17 4.2 Surry Type A Test History 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 Band C tests.

Previous Type A tests confirmed that the Surry Unit 2 reactor containment structure has low leakage and represents insignificant potential risk contributor to increased containment leakage. The increased leakage is minimized by continued Type Band Type C

testing for penetrations with direct communication with containment atmosphere.

Also, the In-Service Inspection (lSI) 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 three Type A tests are reported in the following table for Surry Unit 2.

Test Date As-Found Leakage November 1986 Measured Leakage 0.635 of La with upper confidence Limit (UCL) Margin Type C Penalty 0.003 of La TOTAL 0.638 of La Acceptance Limit (*)

1.0La May 1991 October 2000 Measured Leakage with UCL Margin Type C Penalty TOTAL Measured Leakage with UCL Margin Type C Penalty TOTAL 0.414 of La 0.004 of La 0.418 of La 0.050 of La 0.010 of La 0.060 of La 1.0La 1.0La The total allowable "as-left" leakage is 0.75 La, (La, 0.1 % of primary containment air by weight per day, is the leakage assumed in dose consequences) with 0.6 La, the maximum leakage from Type Band C components.

4.3 Description of Containment The containment is a steel-lined, heavily reinforced concrete structure with vertical cylindrical wall and hemispherical dome supported on a flat base mat. Below grade, the containment structures are constructed inside a cofferdam of steel sheet piling.

The

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 6 of 17 structures are soil-supported. The base of the foundation mat is located approximately 66 feet below finished ground grade.

Each containment structure has an inside diameter of 126 ft. 0 in. The spring line of the dome is 122 ft. 1 in. above the top of the foundation mat. The inside radius of the dome is 63 ft. 0 in. The interior vertical height is 185 ft. 1 in., and the base mat is 10 ft. 0 in.

thick. The steel liner for the wall is 3/8-inch thick, except over the base mat, where 0.25-inch and 0.75-inch plate is used. The steel liner for the dome is 0.50-inch thick. A waterproof membrane is placed below the containment structural mat and carried up the containment wall to ground level.

Attached to and entirely enveloping the part of the structure below grade, the membrane protects the structure from the effects of ground water and the steel liner from external hydrostatic pressure. Ground water immediately adjacent to the containment structure is kept below the top surface of the foundation mat by pumping as required.

Access to the containment structure is provided by a 7 ft. inside diameter personnel hatch penetration, and a 14 ft. 6 in. inside diameter equipment hatch penetration. Other smaller containment structure penetrations include hot and cold pipes, main steam and feedwater pipes, fuel transfer tube, and electrical conductors.

The reinforced concrete structure has been designed to withstand all loadings and stresses anticipated during the operation and life of the unit. The steel lining is attached to and supported by the concrete. The liner functions primarily as a gastight membrane and transmits incident loads to the concrete. The containment structure does not require the participation of the liner as a structural component. No credit has been taken for the presence of the steel liner in designing the containment structure to resist seismic force or other design loads.

The steel wall and dome liners are protected from potential interior missiles by interior concrete shield walls.

CRDM missile protection is provided by a concrete shield on Unit 1 and a steel shield on Unit 2.

The base mat liner is protected by a 1.5 to 2-foot thick concrete cover, except where a O.75-inch-thick liner plate was used beneath the reactor vessel incore instrumentation, and at a drainage trench where floor grating provides additional protection.

The design basis accident was selected as the design basis for the containment structure because all other bases would result in lower temperatures and pressures.

The containment structure is also designed for the normal subatmospheric operating conditions.

Further, the containment structure is designated for a leakage rate not to exceed 0.1 % of the contained volume per day at 45 psig.

The operating pressure for the containment is greater than 10.1 psia and less than 11.3 psia partial air pressure for Unit 2.

The temperature of the containment air fluctuates between a maximum temperature of 125°F and a minimum of 75°F during normal operation and 60°F during shutdown, depending upon the ambient temperature of available service water.

The normal operating pressure allows accessibility for

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 7 of 17 inspection and minor maintenance during operation without requiring containment pressurization.

The containment structure is designed by ultimate strength methods conforming to ACI 318-63, Part IV-B.

4.4 Containment Leakage Consideration for Operability Containment operability is maintained by limiting leakage to.:: 1.0 La, except prior to entering a Mode where operability is required for the first time following the performance of a periodic test performed in accordance with 10 CFR 50, Appendix J, Option B (see Surry TS 4.4.B.2). At that time the combined Type Band C leakage must be < 0.6 La on a maximum pathway leakage rate basis and the overall Type A leakage must be

< 0.75 La. At all other times prior to performing as found testing, the acceptance criteria for Type Band C leakage testing is < 0.60 La on a minimum pathway leakage rate basis.

In addition, to leakage considerations following a design basis LOCA, containment operability also requires structural integrity following a design basis accident.

Compliance with the Technical Specification discussed above will ensure a containment configuration, including personnel and equipment hatches, that is structurally sound and that will limit leakage to the rates assumed in the safety analysis.

4.5 Containment Operational Performance During power operation, Surry Unit 2 is maintained at a subatmospheric condition.

Instrumentation constantly monitors containment pressure.

If pressure rises, an alarm annunciates conditions approaching the limits specified in the Technical Specifications.

Although not as significant as the differential pressure resulting from a design basis accident, the fact that the containment can be maintained subatmospheric provides a degree of assurance of containment structural integrity (i.e., no large leak paths in the containment structure). This feature is a complement to visual inspection of the interior and exterior of the containment structure for those areas that may be inaccessible for visual examination.

4.6 IWE/IWL Inservice Inspection (lSI) Activities to Support Type A Test Surry Unit 2 has completed the requirements of their first ten-year concrete containment inservice inspection program (lWL).

Concrete containment examinations were completed by August 31, 2007 in accordance with the requirements of the 1992 Edition with the 1992 Addenda of ASME Section XI completing the first ten-year interval.

These examinations on the concrete exterior were conducted by the Responsible Engineer using the visual method (VT-3C and VT-1C).

The second ten-year concrete containment examinations (lWL) will be completed in accordance with the requirements of the 2001 Edition through the 2003 Addenda of ASME Section XI and have specified dates of August 31,2011 and August 31, 2016.

General and detailed visual examinations shall be completed in accordance with Category L-A of the Code no earlier than or later than one year of the specified date. The Surry Unit 2 containment

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 8 of 17 does not have an unbonded post-tensioning system.

As such, examinations required by Category L-B do not apply.

The first ten-year interval metallic shell and penetration liner examination (IWE) requirements are still being completed.

These examinations are being in accordance with the 1992 Edition with the 1992 Addenda of ASME Section XI with the one exception for Item E1.12 (wetted surfaces of submerged areas). For this item, the NRC recently approved the request to use the 2001 Edition through the 2003 Addenda of ASME Section XI as modified by 10 CFR 50.55a(b) limitations.

The interval date for Surry Unit 2 is October 20, 1997 to May 21,2008.

Examinations performed for Surry Unit 2 include the general visual of all accessible areas, and a visual (VT-1) examination of pressure retaining bolting. Additionally, visual (VT-3) examinations are planned for item E1.12 (wetted surfaces of submerged areas).

The examination requirements are detailed in Categories E-A and E-G of the applicable Code. No other examination categories apply.

The second ten-year interval IWE examination requirements will be performed in accordance with the 2001 Edition through the 2003 Addenda of ASME Section XI as modified by the 10 CFR 50.55a (b) limitations. At this time no augmented Category E-C examinations are planned.

The remaining examinations are based on Category E-A, and are visual (General, VT-3, and VT-1) examinations based on Code or 10 CFR requirements.

The following relief requests were reviewed to assess the effect, if any, resulting from the proposed Type A test frequency extension:

Relief Request RR-IWE2 obtained relief from Section XI of the ASME Code, 1992 Edition, 1992 Addenda, Code Items E5.10 and E5.20 which require a visual examination of metal containment seals and gaskets.

The relief permits continued acceptance of containment seals and gaskets through the performance of 10 CFR Appendix J testing rather than by individual visual inspection.

NRC letter dated April 14, 1999 granted this relief to Surry Units 1 and 2. The proposed Type A test frequency extension only affects Type A testing.

The Type B testing program remains unaffected and, therefore, the relief request remains valid and unaffected by the proposed change.

Relief Request RR-IWE5 obtained relief from Section XI of the ASME Code, 1992 Edition, 1992 Addenda, Code Item E8.20 which requires a bolt torque or tension test for bolted connections that have not been disassembled and reassembled during the inspection interval.

The relief request permits the leak tightness of bolted connections to be verified through the performance of 10 CFR 50 Appendix J testing. NRC letter dated April 21, 1999 granted the relief request. The proposed frequency extension affects Type A testing only.

The Type B testing program is not affected. As a result, the relief request remains valid and unaffected

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 9 of 17 by the proposed change.

Relief Request RR-IWE8 obtained relief from Section XI of the ASME Code 1992

Edition, 1992

Addenda, Table IWE-2500-1, Category E-P, which contains examination requirements in conjunction with post

repair, replacement, and 10 CFR 50 Appendix J requirements. NRC letter dated March 8, 2000 granted the relief request for Surry Units 1 and 2. The relief request is administrative in nature, removing redundant Code requirements addressed by Appendix J and eliminating unnecessary Authorized Nuclear Inservice Inspector (ANI!) involvement. As a result, the relief request remains valid and unaffected by the proposed change.

The second ten-year interval IWE program for Surry meets the requirements of the 2001 Edition through the 2003 Addenda of ASME Section XI. Categories E-D and E-G are no longer part of the Code.

The relief requests above are not needed for the second ten-year interval since examination of seals and gaskets and bolt torque or tension tests are no longer addressed by ASME Section XI.

As such, the extension request will no longer impact the ASME Section XI program upon second interval start for each unit. Given the short time period remaining in the first ten-year IWE lSI interval for Surry Unit 2, and the Type Band C tests performed during the first ten-year IWE lSI interval, the Appendix J, Type A extension is seen as having a negligible impact.

Surry Engineering performs IWE/IWL lSI inspection activities in support of the required Type A test. There will be no change to the schedule for these inspections due to the extension of the Type A test interval.

The activities performed that assure continued containment integrity include:

During refueling outages, IWE general visual examinations are performed on the Containment Metal Liner (IWE - MC component).

All or parts of accessible areas are examined.

Although localized rust and surface anomalies were detected, no repairs have been required to meet Code requirements.

The 2002 IWL containment lSI Program inspections of the Surry Unit 2 containment structure identified embedded material in the containment dome area.

In the Unit 2 containment dome the material appears to be wood, approximately 2 in. x 12 in. with the side grain exposed. The findings represent a direct inspection of approximately one-third of the containment dome and a remote inspection of the remainder of the containment dome.

The embedded material, as described above, was inadvertently left in the containment structure during original plant construction. The slight depression in the location of the wood and below the adjacent concrete indicates that the wood was likely concealed below a thin layer of cement paste immediately following removal of the concrete form-work. Over time this thin layer of concrete has spalied off, leaving the wood exposed.

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page10of17 Engineering performed an assessment of the significance of the embedded material. The assessment concluded that the containment structures remain fully capable of meeting the functional design requirements as described in Technical Specification 5.2 and UFSAR Section 15.5. This assessment assumed that the piece of wood extends from the concrete surface through the concrete placement and also assumed that similar embedded pieces of wood could exist in the two-thirds of the containment dome that was not directly inspected at that time.

An Engineering evaluation of the inspection findings was performed and concluded that:

the leak-tight integrity of the liner has not been jeopardized, any degradation of the underlying reinforcing steel as a result of the embedded wood is insignificant to the structure, the loss of concrete displaced by the wood will have an insignificant effect upon the structure, and no significant loss of radiological shielding or missile protection has occurred.

Direct inspection of the remainder of the containment dome was completed in April 2002. Any identified embedded debris was removed and the containment repaired, as necessary.

The ASME Section XI, Article IWL interval 2 (2nd 5-year) inspection of the Surry Unit 2 containment structure was conducted between May 2006 and December 2006. The following was observed during the inspection:

Embedded materials were identified in the containment dome area. The materials included small scraps of dimensional lumber, a rubber gasket, metal rod, and wire.

This debris was all located within the concrete cover above the first row of steel reinforcing bar. No reinforcing bar was exposed as a result of debris removal and no concrete deterioration was identified. All identified debris was extracted and all void areas deeper than one half inch were repaired using cement based mortar.

Numerous pop-outs and abandoned anchors were identified on the containment exterior concrete and uncoated areas of the containment structure adjacent to other building interiors.

The pop-outs and abandoned anchors identified on the exterior concrete were repaired using cement based mortar.

Rock pockets, sand pockets, a crack, a spall, two small holes, and a number of loose patches were identified.

All identified defects were excavated to sound material and repaired using cement based mortar or concrete.

Based upon these inspection findings, the Unit 2 containment structure is generally found to be in good material condition.

No significant defects or concerns were observed on the exterior concrete and all observed defects were due to original

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 11 of 17 construction flaws.

Taken together or individually, the defects identified did not represent a significant structural concern. The containment structure continues to retain its ability to perform as designed under all load cases including the design basis earthquake and a postulated strike from a tornado generated missile.

During the fall 2003 (S2-R-18) refueling outage Surry Unit 2 replaced the reactor pressure ves:)el head. Completion of the replacement head required an opening in the containment larger than the equipment hatch.

Therefore, an opening ranging from a size of approximately 18 ft. 6 in. high by 17 ft. 0 in. wide in the steel liner plate and 26 ft. 0 in.

hi~lh by 27 ft. 4 in. wide at the outside face of the concrete. The opening was repaired, examined, and tested in accordance with the appropriate ASME Code requirements for the metal liner and concrete structure.

The cut steHI liner plate was welded back to its original configuration using full penetration welds. After the cut steel liner plate was welded back to its original configuration Non-Destructive Examination (NDE) was performed on the completed full penetration weld. A surface examination was performed on the weld as well as a spot radiographic examination.

In addition, although not an original design requirement, vacuum box testing was performed as required in accordance with Relief Request IWE9 (Reference 8).

After vacuum box testing of the liner seam weld and installation of the channel, the channel to liner weld was tested by a static pressure test (decay test) and the weld wa) soap bubble tested for leakage with an acceptance criterion of zero leakage.

In addition, following the containment building pressure test, the channel was pressurized and an "as-found" local leak rate test was performed at 50 psi in accordance with NRC approved relief request RR-IWE9.

In accordance with IWL Article 5000 of ASME Code, a containment structure pressure test was performed at 45 psi. The surface of the replacement concrete was examined in accordance with IWL-5250 prior to pressurization, at test pressure and following completion of the pressurization test.

The extensive testing ensures that the Unit 2 containment:;tructure has been restored to its original design condition.

5.0 PLANT SPECIFIC RISK ASSESSMENT FOR THE EXTENDED TYPE A TEST INTERVAL 5.1 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 ma33 over time.

Likewi3e, thi3 approach is used in the analyses presented in EPRI TR-104285, NUREG-1493, and the NEI Interim Guidance. The

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 12 of 17 analysis performed examines plant specific accident sequences in which the containment integrity remains intact or the containment is impaired.

Specifically, the following wer,e 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 te!st 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 rates across containment isolation valves.

Core damage sequences in which containment integrity is impaired due to containmEmt 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*1 04285 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 accounted for in this evaluation.

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

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., valve failing to close following a valve stroke test) where no1 accounted for in this evaluation (EPRI TR-104285 Class 6 sequences).

5.2 Conclusions Based on the above sequences considered, the following conclusions are made regarding the plant risk associated with extending the Type A test frequency from ten years to fiftee!n years:

RG 1.174 (Reference 8) provides guidance for determining the risk impact of plant-speGific changes to the licensing basis. RG 1.174 defines small changes in risk as resulting in increases of core damage frequency (CDF) below 1E-5/yr and

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 13 of 17 increases in large early release frequency (LERF) below 1E-6/yr. Since the Type A test does not impact CDF, the relevant criterion is LERF.

The increase in LERF resulting from a change in the Type A test interval from once-per-ten-YEiars to once-per-fifteen-years is 1.3E-7/yr based on internal events.

RG 1.174 states that when the calculated increase in LERF is in the range of 1E-7/yr to 1E-6/yr, applications will be considered if it can be shown that the total LERF is less than 1E-~I/Yr.

Since the total LERF for the 15-year metric is 9.8E-7/yr, then the proposed change is considered acceptable.

The increase in the total dose rate is defined here by person-rem/year increases for those accdent sequences influenced by Type A testing. The one-time change to the Type A test interval from ten years to fifteen years increases the Type A test dose rate by 0.024%.

This change in dose rate is due to the conservative assumption made in the calculation of the Class 3 frequencies.

The change in conditional containment failure probability (CCFP) is calculated to demonstrate the impact on 'defense-in-depth.'

The changes in the contribution of sequences involving containment failure are negligible from the ten-year interval to the propo8ed fifteen-year interval. Therefore, the Surry Unit 2 containment defense-in-depth i~; maintained.

The risk assessment calculation performed for the 5-year Type A test extension for Unit 2 from 1I) years to 15 years is included in Attachment 2.

6.0 EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION The proposed revision to 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, will be extended on a one-time basis to fifteen years from the last Type A test.

In accordance with the requirements of 10 CFR 50.9:2, the enclosed application is judged to involve no significant hazards based upon the following information:

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

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 an activity or modification by itself that could lead to equipment failure or accident initiation.

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 14 of 17 The proposed 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 reducing the Type A testing frequency to once per twenty years leads to an imperceptible increase in risk.

Surry provides a high degree of assurance through 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 Band C testing, required by Technical Specifications, identifies any containment opening from design penetrations, such as valves, that would otherwise be detected by a Type A test. These factors establish that an extension to the Surry Type A test interval will not represent a significant increase in the consequences of an accident.

2.

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

The proposed revision to Surry 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. The proposed extension to Type A testing does 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.

3.

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

The proposed revision to Surry 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, will be extended on a one-time basis to fifteen years from the last Type A test.

Regulatory Guide (RG) 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis.

RG 1.174 defines small changes in risk as resulting in increases of CDF below 1E-5/yr and increases in LERF below 1E-6/yr.

Since the Type A test does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A test interval from a once-per-ten-years to a once-per-fifteen-years is 1.3E-7/yr, based on internal events.

RG 1.174 states that when the calculated increase in LERF is in the range of 1E-7/yr to 1E-6/yr, applications will be considered if it can be shown that the total LERF is less than 1E-6/yr. Since the total LERF is 9.8E-7/yr, the change is considered small and not a significant reduction in margin.

Increasing the Type A test interval from ten to fifteen years is, therefore, considered non-risk significant and will not significantly reduce the margin of safety.

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 15 of 17 The NUREG-1493 generic study of the effects of extending containment leakage testing found that a 20-year extension in Type A leakage testing resulted in an imperceptible increase in risk to the public.

NUREG-1493 found that, generically, the design containment leakage rate contributes about 0.1 percent of the overall risk and that 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 be detected by Type Band C testing.

Furthermore, for Surry, maintaining the containment subatmospheric during plant operations further reduces the risk of any containment leakage path going undetected.

7.0 IMPLEMENTATION OF THE PROPOSED CHANGE This amendment request meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) as follows:

(i) The amendment involves no significant hazards consideration.

As described in Section 6 of this evaluation, the proposed change involves no significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed change does not involve the installation of any new equipment, or the modification of any equipment that may affect the types or amounts of effluents that may be released offsite. Therefore, there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupation radiation exposure.

The proposed change does not involve plant physical changes, or introduce any new mode of plant operation. The extended interval will not include any activities that will increase individual or cumulative occupation radiation exposure. Therefore, there is no significant increase in individual or cumulative occupational radiation exposure.

Based on the above, Dominion concludes that the proposed changes meet the criteria specified in 10 CFR 51.22 for a categorical exclusion from the requirements of 10 CFR 51.22 relative to requiring a specific environmental assessment by the Commission.

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page 16of17

8.0 CONCLUSION

The proposed one-time change will not alter assumptions relative to the mitigation of an accident or transient event and will not adversely affect normal plant operation and testing. The proposed change is consistent with the current safety analysis assumptions and with the Technical Specifications. As such, riO question of safety exists.

RG 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines small changes in risk as resulting in increases of CDF below 1E-5/yr and increases in LERF below 1E-6/yr.

Since the Type A does not impact CDF, the relevant criterion is LERF.

The increase in LERF resulting from a change in the Type A test interval from once-pElr-ten-years to once-per-fifteen-years is 1.3E-7/yr, based on internal events. RG 1.174 states that when the calculated increase in LERF is in the range of 1E-7/yr to 1E-6/yr, applications will be considered if it can be shown that the total LERF is less than 1E-5/yr. Since the total LERF is 9.8E-7/yr then the change is considered acceptable.

The Facility Safety Review Committee has reviewed this proposed change to the Technical Specifications and have !Concluded that it does not involve a significant hazards consideration and will not endanger the health and safety of the public.

9.0 REFERENCES

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

2. NRC letter to Surry Units 1 and 2, Technical Specification Amendments 208/208, dated, April 18, 1996 that permitted implemE~ntation of the requirements of 10 CFR 50, Appendix J, Option B for performance-based primary reactor containment leakage testing.

3.

Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program,"

September 1995.

4. American National Standard ANSVANS - 56.8 - 1994, "Containment System Leakage Testing Requirements."

5.

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

6.

EPRI TR-104285, "Risk Assessment of Revised Containment Leak Rate Testing Intervals," dated August 1994.

Serial No. 07-0802 Docket No. 50-281 Unit 2 Type A Test Extension Page17of17 7.

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

8. NRC letter to Surry Power Station approving RR-IWE9, "Surry Power Station, Unit 2 - ASME Section XI, Inservice Inspelction Program, Relief Request (RR)

RR IWE9 for Containment Testing," dated May1, 2003.

9.

Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated July 1998.

Serial No. 07-0802 Docket No. 50-281 One-Time Five-Year Extension to Type A Test Interval Probabilistic Risk Assessment Five Year Type Extension for Surry Unit 2 Surry Power Station Unit 2 Virginia Electric and Power Company (Dominion)

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK Part V, Volume RA.LIA, REVISION 0 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 Surry Power Station Probabilistic Risk Assessment Notebook Part VPRA Risk Analysis Volume RA.LI.4 Calculation of 5 year ILRT Extension fol' Unit 2 from 10 years to 15 years Revision No.2 Effective: Date: December 2007

Purpose:

To provide a risk impact assessment on extending the Integrated Leak Rate Test (ILRT) interval for Surry Unit 2 from 10 years to 15 years.

==

Conclusion:==

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 1.33E-07 / yr.

Therefore, the risk impact when compared to other severe accident risks is small.

Prepared By:

Signature Date Reviewed By:

Signature Date Approved By:

Signature Date

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.i Part V, Volume RA.LI.4, REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 Table of Contents 1.0 2.0 3.0 3.1 3.2 3.3 4.0 5.0 6.0 PURPOSE -------------------------------------------..------------------------------------------------ 1 INTRODUCTION------------------------------..------------------------------------------------ 1 ANALYSIS ----------------------------..------------..------------------------------------------------ 2 InJluts--------------------------------*---------------------------------------------------------------- 2 AssulllJltions -------------------------------------------------------------------------------------- 2 Methodology--------------------------------------------------------------------------------------- 3 BODY OF CALCULATION------.--------------.-------------------------------------------- 5 RESULTS AND CONCLUSIONS-------------..---------------------------------------------24 REFERENCES ----------------------.---------------------------------------------------------------26 ATTACHMENT A - SPS FREQUENCY AND DOSE DATA ----------------------------------- Al ATTACHMENT B - JUSTIFICATION OF VOLUNlE CHANGE------------------------------- Bl ATTACHMENT C - REVIEWER COMMENTSIRl~SOLUTIONS--------------------------- Cl List ofTables TABLE 1 - Baseline Containlllent Frequencies - Giv<<m Accident Class--------------------------- 9 TABLE 2 - Person-Relll Measures - Given Accident Class-----------------------------------------l0 TABLE 3 - Baseline Mean Person-Relll Measures - Given Accident Class ----------------------10 TABLE 4 - Mean Consequence Measures for 10-Year Test Interval-Given Accident Class ------------------------------------------..--------------..----------------------------------------------- 12 TABLE 5 - Mean Consequence Measures for 15-Year Test Interval-Given' Accident Class------------------------------*-------------**----------------------------------------------14 TABLE 6 - Evaluated IIIlJlact of Containment Leak Size on Containment Leak Rate---------16 TABLE 7 - Liner Corrosion Base Case -----------------..------------------------------------------- 21 TABLE 8 - Summary ofResults----------------------..---------------------------------------------- 25 TABLE A-l-SPS Frequency and Dose Data ---------..------------------------------------------ Al TABLE A Surry Total EPRI Class Frequency and Dose Data--------------------------------- Al List ofFigurE!s FIGURE*1 - Fractional Impact on Risk Associated with Containment Leak Rates-------------!7

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.iii Part V, Volume RA.LI.4, IU:VISION 0 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2

SUMMARY

OF CHANGES

~-

Revision Author Summary 0

Initial issuance.

f----- 1 Editorial clarification of comments 2

Clarification of the Conditional Containment Failure Probability for the 15 year test int~~rval f-----

~-

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.I Part V, Volume RA.LI.4, REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 1.0 PURPOSE The purpose ofthis document is to provide a risk impact assessment on extending the Integrated Leak Rate Test (ILRT) interval for Surry Unit 2 from once ill 10 years to once in 15 years.

2.0 INTRODUCTION

On 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 Surry Nuclear Power Station (SPS) selected the requirements under Option B as its testing program [PROCDROl].

The surveillance testing requirements as proposed in NEI 94~OI [REPORTOl] for Type A testing is at least once every 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 Surry Unit 2 current IO-year Type A test interval e::nds in October 2010. The proposed amendment to the Technical Specification (TS) is for a one~time exception to the lO~yearfrequency ofthe performance-based leakage rate testing program for Type A test as documented in [REPORTOl]. The exception will allow ILRT testing within 15 years from the last ILRT which was performed in October 2000.

This calculation will provide a risk impact assessment on extending the plant's ILRT interval by five years. The risk assessment will be performed in accordance with the guidelines set forth by NEI

[REPORTOl] and [REPORT02], the methodology used by EPRI [REPORT03] and [REPORT04], and the NRC regulatory guidance:: on the use ofProbabilistic Risk Assessment (PRA) findings and risk insights in support ofa licensee request for changes to a plant's licensing basis, Regulatory Guide 1.174

[RGOl].

In addition, the results and findings from the Surry Individual Plant Examination (IPE) [REPORT09],

the revised model [CALC02], [CALC04], [CALC11], and previous calculations [CALC06], [CALC07],

and [CALC09] were used for this risk assessment calculation.

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.2 Part V. Volume RA.LL4, REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 3.0 ANALYSIS 3.1 Inputs This calculation will use Surry 50 mile population data for calculating the population dose, which was also used for license extension Severe Accident Mitigation Alternatives (SAMA) analysis as discussed in [CALCOl]. The Source Tenn Category (STC) release fractions and corresponding frequencies were taken from the [CALC11] and revised data in [CALC02]. Source tenn category is defined here as a grouping oflike releases from the Containment Event Tree (CET) endpoints such that the offsite consequences are expected to be similar. There are enough STCs to cover the spectrum ofreleases.

3.2 Assumptions As stated in the Surry Technical Specifications, the leakage rate (La) acceptance criterion is defined as:

La = 0.1 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)

The NEI interim guidance [REPORT02] was instrumental in making all ofthe following assumptions:

1.

Containment leak rates greater than 1La, but less than 35La, indicate an impaired containment.

Leak rates within this range are considered 'small'.

2.

Containment leak rates greater than 35La, indicate a containment breach. These leak rates are considered to be 'large'.

3.

Containment leak rates less than 1La indicate an intact containment. These leak rates are considered to be 'negligible'.

4.

The maximum containment leakage for Class 3A sequences is lOLa.

5.

The maximum containment leakage for Class 3B sequences is 35La*

6.

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.

7.

Containment leakage related to Classes 4, 5, and 6 are not affected by changes in ILRT test frequency. Therefore, these classes are not considered in this assessment methodology.

8.

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

9.

The containment releases are not impacted with time.

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.3 Part V, Volume RA.LI.4, REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year.Integrated Leak Rate Test Extension for Unit 2 3.3 Methodology A simplified bounding analysis approach for evaluating the change in risk associated with increasing the interval from 10 years to 15 years for the Type A test was used. Type A tests measure the containment air mass and calculates the leakage from the change in mass over time. This approach is similar to that presented in the EPRI [REPORT03] and NEI [REPORT02] reports, as well as NUREG-1493

[NUREGOl]. Namely, the analysis perfonned examined the SPS IPE [REPORT09] 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 TRlO4285 Class 1 sequences).

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

For example, liner breach or steam generator man way leakage. (EPRI TR-I04285 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.

Accident sequences involving containment failure induced by severe accident phenomena (EPRI TR-I04285 Class 7 sequences), containment bypassed (EPRI TR-I04285 Class 8 sequences) and large containment isolation failures (EPRI TR-I04285 Class 2 sequences).

Small containment isolation 'failure-to-seal' events (EPRI TR-I04285 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.

Core damage sequences in which containment integrity is impaired due to containment isolation failures ofpathways left 'opened' following a plant post-maintenance test (e.g. a valve failing to close following a valve stroke test) where not accounted for in this evaluation (EPRI TR-I04285 Class 6 sequences).

Jhe steps taken to perform this risk assessment evaluation are as follows:

Step 1 - Quantify the baseline risk in terms ofcore damage frequency per reactor year for each ofthe eight accident classes presented in Table I. Map the Level 3 release categories into 8 release classes defined by EPRI TR-I04285 [REPORT03]. See Table A-I of Attachment A.

Step 2 - Develop baseline plant specific population dose in person-rem per reactor year for each ofthe eight accident classes evaluated in the EPR! Report [REPORT03].

Step 3 - Evaluate risk impact ofextending the Type A test interval from 10-to-15 years.

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.4 Part V. Volume RA.LI.4. REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 Step 4 - Determine the change in risk in terms ofLarge Early Release Frequency (LERF) in ac~ordancewith RG 1.174 [RGOl]

Step 5 - Evaluate the Risk Impact in Terms of ~LERF Step 6 - Determine Impact on Conditional Containment Failure Probability

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.5 Part V. Volume RA.LI.4. REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 4.0 BODY OF CALCULATION Step 1-Quantify the baseline risk in terms of core damage frequency per reactor year.

This step involves the review ofthe SPS IPE [REPORT09] containment event tree (CET). The CET

<~ha..11lcterizesthe response ofthe containment to important severe accident sequences. The CET used in this evaluation is based on important phenomena and systems-related events identified in NUREG-1335

[NUREG02] and NSAC-159, Volume 2 [REPORT05] and on plant features that influence the phenomena.

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

A review ofthe containment isolation model was perfonned [CALC03] and [CALC04]. The five issues associated with containment isolation in NUREG-1335 [NUREG02] were examined, and these issues are:

(1) The identity ofpathways 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 oftesting and maintenance procedures.

(5) The quantification ofeach containment isolation mode.

The containment isolation model in [CALC03] and [CALC04] screened out lines less than 5.5 inches in diameter which was the minimum cutofffor the LERF definition. This evaluation considers lines that are sized between 0.1 inches and 5.5 inches as potential candidates for significant containment leakage.

The Level 3 release categories were mapped into 8 release classes (See Table A-I in Attachment A) as defined in the EPRI Report [REPORT03]. These EPRI containment failure classifications are listed below.

EPRI Containment Failure Classifications Class 1 Containment remains intact including accident sequences that do not lead to containment failure in the long tenn. The release offission products (and attendant consequences) is detennined 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 ofcontainment volume per day for PWRs (e.g. SPS measured at Pa, calculated peak containment pressure related to the design basis accident). Changes to leak rate testing frequencies do not affect this classification.

SPS PROBABn..ISTIC RISK ASSESSMENT NOTEBOOK P.6 Part V, Volume RA.LI.4, REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 Class 2 Containment isolation failures (as reported in the IPEs) include those accidents in which the pre-existing leakage is due to failure to isolate the containment. These include those that are dependent on the core damage accident in progress (e. g., initiated by common cause failure or support system failure ofpower) and random failures to close a containment path. Changes in Appendix J testing requirements do not impact these accidents.

Class 3 Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal (i.e., provide a leakMtight 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.

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

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

Class 6 Containment isolation failures include those leak paths not identified by the LLRTs. The type ofpenetration failures considered under this class includes those covered in the plant test and maintenance requirement or verified by in service 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 ofaccidents.

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

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 frequencies are needed to detennine the Class 1 frequencies and will be calculated first.

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

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.7 Part V, Volume RA.LI.4, REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 To calculate the probability that a liner leak will be large (Event CLASS-3B), the data presented in NUREG-1493 [NUREGOl] was used. 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 (La). Since 21La does not constitute a large release (please refer to the write-up in Step 4 and Table 6 for Large Release information), no large releases have occurred based on the 144 ILRTs reported in NUREG-1493 [NUREG01].

An improvement in the methodology used to determine the frequencies ofleakages detectable only by ILRTs, classes 3A and 3B was made using the methods documented in [REPORT02]. The method utilized in the aforementioned utility submittals (discussed in [REPORT02]) involved using a 95%

confidence ofa i2 distribution ofthe noted ILRT failures (4 of 144 reported in [NUREGO1]). 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 the statistical information should be based on 5/182. Rather than using the 1.,2 distribution used previously, it has been considered more appropriate to utilize the mean (5/182 = 0.027) for the class 3A (small leak) distribution. From the NEI document [REPORT02], the Jeffrey's non-informative prior distribution was used to calculate the class 3B (large leak) distribution as follows:

L"1 P

b b'l' (NumberojFailures) + (0.5) raZ ure r 0 a z tty =....:----::...--~-....:

(Numberofl'ests) +1 Failure Probability = (0) + (0.5)

(182) + 1 The number oflarge failures is zero, so the class 3B probability is 0.5 / 183 =0.0027 The respective frequencies per year are determined as follows:

CLASS-3A-FREQUENCY =PROBclass-3A

• CDF CLASS-3B-FREQUENCY = PROBclass-3B

• CDP where:

PROBclass-3A = probability ofsmall pre-existing containment liner leakage

= 0.027 PROBclass-3B =probability oflarge pre-existing containment liner leakage

= 0.0027 CDP = 2.93E-05 / year

[Table A-I, Attachment A]

CLASS-3A-Base-Frequency = 0.027 *2.93E-05/ year =7.91E-07 / year CLASS-3B-Base-Frequency = 0.0027

• 2.93E-05 / year = 7.9lE-08 / year

SPS PROBABll,ISTIC RISK ASSESSMENT NOTEBOOK P.8 Part V. Volume RA.LI.4. REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 For this analysis the associated maximum containment leakage for class 3A is lOLa and for class 3B is 35La Class 1 Sequences: This group consists ofall core damage accident progression bins for which the containment remains intact. The frequency per year for these sequences is 3.17E-06 / year (see Attachment A, Table A-I). For this analysis the associated maximum containment leakage for this group is 1La. The SPS IPE did not model Class 3 type failures; therefore they need to be accounted for in the Class 1 accident class. Using NEI interim guidance methodology [REPORT02], the frequency for Class 1 should be reduced by the new estimated frequencies in Class 3A and Class 3B in order to preserve the total CDF. The revised Class 1 frequency is therefore:

CLASS-I-FREQ =FREQClass (FREQClass3A + FREQClass3B)

CLASS-I-FREQ = 3.17E (7.91E-07 + 7.91E-08)

CLASS-I-Base-Frequency = 2.30E-06/ year Class 2 Sequences: This group consists ofall 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-I, Attachment A as Class 2.

CLASS-2-FREQUENCY = 6.52E-IO / year Class 4 Sequences: This group consists ofall core damage accident progression bins for which a failure-to-seal containment isolation failure ofType 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 ofall core damage accident progression bins for which a failure-to-seal containment isolation failure ofType 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 Sequences: This group is similar to Class 2 and addresses additional failure modes not typically modeled in PRAs due to the low probability ofoccurrence. These are sequences that involve core damage accident progression bins for which a failure-to-seal containment leakage due to failure to isolate the containment occurs. These sequences are dominated by misalignment of containment isolation valves following a test/maintenance evolution.

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

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.9 Part V. Volume RA.LI.4. REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for lJnit 2 Class 7 Seguences: This group consists ofall core damage accident progression bins in which containment failure is induced by severe accident phenomena (Early and Late Failures). The frequency of Class 7 is the sum ofthose release categories identified in Table A-I, Attachment A as Class 7.

CLASS-7-FREQUENCY = 2.361:-05/ year Class 8 Sequences: This group consists ofall core damage accident progression bins in which containment bypass occurs. The frequency ofClass 8 is the sum ofthose release categories identified in Table A-I as Class 8.

CLASS-8-FREQUENCY = 2.49]j~-06/ year Note: for this class the maximum release is not based on nonnal containment leakage, because the releases are released directly to the environment. Therefore, the containment structure will not impact the release magnitude.

A

'd tCI G'

tF T bl 1 B lin C ta" a

e ase e

on mmen reQuenCles -

lven CCI en ass Class Description Frequency (per Rx-year) 1 No Containment Failure 2.30E-06 2

Large Contaitnnent Isolation Failures (Failure-to-close) 6.52E-10 3A Small Isolation Failures (Type A test) 7.91E-07 3B Large Isolation Failures (rypeA test) 7.91E-08 4

Small isolation failure - failure-to-seal (Type B test)

Not Analyzed 5

Small isolation failure - failure-to-seal (Type C test)

Not Analyzed 6

Containment Isolation Failures (dependent failures, personnel errors)

Not Analyzed 7

Severe Accident Phenomena Induced Failure (Early and late Failures) 2.36E-05 8

Containment Bypassed (SGTR & V-Sequence) 2.49E-06 CDF Core Damage All CET End states 2.93E-05 Step 2 - Develop baseline plant specific person-rem dose (population dose) per reactor year.

Plant-specific MAAP/MACCS2 analysis was perfonned to evaluate the person-rem dose to the population, within a 50-mile radius from the Surry power plant. The dose for Class 1 and Class 2 accidents is the sum ofthe Class 1 and Class 2 dose values from Table A-I, Attachment A, respectively.

Using the total population dose for Class 1 accidents as the starting reference point, the Class 3, Class 7, and Class 8 accidents are calculated below. The population dose is converted to the corresponding Class value using the appropriate dose multiplier as was used in the NEI methodology [REPORT02] to predict the person-rem dose for Class 3 accidents.. Note that the multiplier (i.e. 10) for Class 3A is the maximum containment leakage multiplier assumed for small ieaks, and the Class 3B multiplier (Le. 35) is the maximum containment leakage multiplier assumed for large leaks. The dose for the Class 7 accidents was obtained by frequency weighting all the Class 7 dose values. This was done by dividing

SPS PROBABll..ISTIC RISK ASSESSMENT NOTEBOOK P.lO Part V, Volume RA.LI.4. REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 the sum ofthe products by the sum ofthe frequencies from Table A~l. Class 8 sequences include contaimnent bypass failures; as a result, the person-rem dose is not based on nonnal contaimnent leakage. The releases for this class are expected to be released directly to the enviromnent. The class 8 doses are frequency weighted as were done for Class 7. The frequency weighted Class 8 dose from Table A-I represent the sum ofthe dose for the Event-V and SGTR sequences. The baseline dose results are calculated below and are summarized in Tahle 2.

Class 1 = 5.98E+02 person-rem Class 2 =2.38E+04 person-rem Class 3A = 5.98E+02

• 10:= 5.98E+03 person~rem Class 3B = 5.98E+02

• 35 := 2.09E+04 person-rem Class 4 =Not analyzed Class 5 = Not analyzed Class 6 = Not analyzed Class 7 =ID(Preq x Dose) / IDPreq = 1.00E+05 person-rem Class 8 = ID(Freq x Dose) /IDFreq = 3.61E+06 person-rem Table 2: Person-Rem Measures - Given Accident Class Class Description Person-Rem (50-Miles) 1 No Contaimnent Failure 5.98E+02 2

Large Containment Isolation Failures (Failure~to~close) 2.38E+04 3A Small Isolation Failures (Type A test) 5.98E+03 3B Large Isolation Failures (Type A test) 2.09E+04 4

Small isolation failure - failure-to-seal (Type B test)

N/A 5

Small isolation failure ~ failure-to-seal (Type C test)

N/A 6

Other Isolation Failures (e.g., Dependent Failures)

N/A 7

Failure Induced by Phenomena (Early and Late Failures) 1.00E+05 8

Contaimnent Bypassed (SGTR & V-Sequence) 3.61E+06 The ab~ve dose results when combined with the frequency results presented in Table 1 yields the SPS 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-Rem/yr (per Rx-yr)

(50-Miles)

(50-Miles)

]

No Containment Failure 2.30E-06 5.98E+02 1.38E-03 2

Large Isolation Failures (Failure-to-close) 6.52E~10 2.38E+04 1.55E-05 3A Small Isolation Failures (Type A test) 7.91E-07 5.98E+03 4.73E-03 3B Large Isolation Failures (Type A test) 7.91E-08 2.09E+04 1.65E-03 4

Small isolation Failure-to-Seal (Type B test)

N/A N/A N/A

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.II Part V. Volume RA.LI.4. REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 5

Small isolation Failure-to-Seal (Type C test)

N/A N/A N/A 6

Other Isolation Failures (e.g., Dependent Failures)

N/A N/A N/A 7

Failure Induced by Phenomena (Early and Late 2.36E-05 1.00E+05 2.36 Failures) 8 Containment Bvpassed (SGTR & V-Sequence) 2.49E-06 3.61E+06 8.99 CDP All CET End States 2.93E-05 N/A 11.36 Based on the above values, using the same methodology as [CALClO], the baseline percent risk contribution ofthe Dose Rate (DR) related to Type A testing is as follows:

% ofTotal DRBASE = [( CLASS3ABASE + CLASS3BBAsE) / TotaIBASE] X 100 where:

CLASS3ABASE = class 3A person-rem/year =4.73E-03 person-rem/year

[Table 3]

CLASS3BBASE = class 3B person-rem/year = 1.65E-03 person-rem/year

[Table 3]

TotalBASE =total person-rem/year for baseline interval = 11.36 person-rem/year

[Table 3]

% ofTotal DRBASE = [(CLASS3ABASE + CLASS3BBASE ) / TotalBAsE ] x 100%

% ofTotal DRBASE = [(4.73E-03 + 1.65E-03) /11.36] x 100 %

% ofTotal DRBASE = 0.056%

Therefore, the baseline percent oftotal dose rate due to Type A testing is 0.056%.

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

The revised methodology in [REPORT02] suggests that a multiplier should be factored into the analysis to represent the change in probability ofleakage. As stated in [REPORT03] and [NUREGOl], relaxing the initial test interval from three ILRTs in a ten year period, to one ILRT in a ten year period increases the average time that a leak detectable only by an ILRT would go undetected from 18 months (3yrs / 2) to 60 months (10 yrs / 2). This is a factor of3.333 (i.e. 60/18). The baseline dose associated with the ten-year interval was previously calculated using the percentage increase (10%), 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, the average time that a leak detectable only by an ILRT would go undetected is 90 months (15 yrs / 2). Therefore, a fac~or of5.0 (i.e. 90/18) should be applied.

Risk Impact Related to to-year Test Interval As previously stated, Type A tests impact only Class 1 and Class 3 sequences. In addition, the increased probability ofnot detecting excessive leakage has no impact on the frequency ofoccurrence 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 remains the same, even though the probability ofnot detecting the liner

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.12 Part V. Volume RA.LI.4, REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 opening increases). Thus, only the frequency ofClass 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 ofleakage of3.33.

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:

FREQClass3AlO= 7.91E-07

• 3.33 = 2.63E-06/ year FREQClass3BIO= 7.9IE-08

• 3.33 = 2.63E-07/ year 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-I-FREQlO = FREQClass-l - (FREQClass3A + FREQClass3B)

CLASS-1-FREQlO = 3.17E (2.63E-06 + 2.63E-07)

CLASS-I-FREQI0 =< 2.77E-07 I year The results ofthese calculations are presented in Table 4 below.

A

'd tel I G' Ii 10 Y Ttl t M

T bl 4 M C

a e

ean onsequence easures or

- ear es n erva -

lven CCI en ass Class Description Frequency Person-Rem Person-Remlyr (per Rx-yr)

(50-Miles)

(50-Miles) 1 No Containment Failure 2.77E-07 5.98E+02 1.66E-04 2

Large Isolation Failures (Failure-to-c1ose) 6.52E-1O 2.38E+04 1.55E-05 3A Small Isolation Failures (Type A test) 2.63E-06 5.98E+03 1.57E-02 3B Large Isolation Failures (Type A test) 2.63E-07 2.09E+04 5.50E-03 4

Small isolation Failure-to-Seal (Type B test)

N/A N/A N/A 5

Small isolation Failure-to-Seal (Type C test)

N/A N/A N/A 6

Other Isolation Failures (e.g., Dependent N/A N/A N/A Failures) 7 Failure Induced by Phenomena (Early and 2.36E-05 1.00E+05 2.36 Late Failures) 8 Bypass (SGTR) 2.49E-06 3.61E+06 8.99 CDF All CET End States 2.93E-05 N/A 11.37 Based on the above values, the Type A 10-year test frequency percent oftotal dose rate for Class 3 is as follows:

% ofTotal DRIO = [(CLASS3AlO +CLASS3BIO) / TotallO] X 100 where:

CLASS3AlO = Class 3A person-rem/year = 1.57E-02 person-rem/year

[Table 4]

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.13 Part V, Volume RA.LI.4, REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 CLASS3BlO =Class 3B person-rem/year = 5.50E-03 person-rern/year TotallO =total person-rem year for to-year interval = 11.37 person-rem/year

% ofTotal DRIO = [(CLASS3AIO+ CLASS3BI0) / TotallO] x 100

% ofTotal DRlO = [(1.57E-02 + 5.50E-03) / 11.37] x 100

% of Total DR10 =0.186%

[Table 4]

[Table 4]

Therefore, the totall0-year test frequency ILRT interval percent oftotal dose rate related to Type A testing is 0.186%.

The Ll% change in the 10 year ILRT Dose Rate from the baseline value is 0.186% - 0.056% =0.130%.

The ten-year dose rate change (related to an ILRT) over the baseline case is as follows:

DR ChangelO =[z:n(Class 1, 3A, 3B) 10 - z:n(Class 1, 3A, 3B) Base]

where:

z:n (Class 1,3A,3B) Base = 1.38E-03 + 4.73E-03 -+- 1.65E-03 person-rem/year z:n (Class 1,3A,3B) Base = 7.76E-03 person-rem/year z:n(Class 1,3A,3B) 10 = 1.66E-04 + 1.57E-02 + 5.50E-03 person-rem/year z:n(Class 1,3A,3B) 10 =2.14E-02 person-rem/year DR Changel0 = [2.14E-02 -7.76E-03] person-rem/year DR Changelo =1.36E-02 person-rem/year

[Table 3]

[Table 4]

Therefore, the ten-year dose rate change from the baseline case is 1.36E-02 person-rem/year.

Risk Impact Related to i5-year Test Interval The risk contribution for a 15 year interval is similar to the lO-year interval. The difference is in the increase in probability ofleakage value. This increase in containment leakage is accounted for by using the multiplier 5.0 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:

FREQClass3A15 = 7.91E-07

• 5.0 = 3.96E-06 / year FREQclass 3BI5 = 7.91E-08

• 5.0:;;; 3.96B-07 / year 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:

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.14 Part V, Volume RA.LI.4, REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 CLASS-I-FREQ15 = FREQClass-l - (FREQClass3A + FREQclass3B)

CLASS-I-FREQ15 = 3.17E (3.96E-06 + 3.96E-07)

CLASS-I-FREQ15 = -1.18E-06/ year.

Since the Class I frequency for the 15 year interval is less than zero (which is impossible), the Class I frequency will be set to zero and an over estimation ofthe total frequency will result. The results ofthis calculation are presented in Table 5 below.

ti 15 Y T

tInt al G'

A'd tCI M

T bl 5 M C

a e

ean onseQuence easures or

- ear. es erv -

Iven CCI en ass Class Description Frequency Person-Rem Person-Remlyr (per Rx-yr)

(50-Miles)

(50-Miles)

I No Containment Failure O.OOE-OO 5.98E+02 O.OOE+OO 2

Large Isolation Failures (Failure-to-close) 6.52E-IO 2.38E+04 I.55E-05 3A Small Isolation Failures (Type A test) 3.96E-06 5.98E+03 2.37E-02 3B Large Isolation Failures (Type A test) 3.96E-07 2.09E+04 8.28E-03 4

Small isolation Failure-to-Seal (Type B test)

N/A N/A N/A 5

Small isolation Failure-to-Seal (Type C test)

N/A N/A N/A 6

Other Isolation Failures (e.g., Dependent N/A N/A N/A Failures) 7 Failure Induced by Phenomena (Early and 2.36E-05 1.00E+05 2.36E+00 Late Failures) 8 Bypass (SGTR) 2.49E-06 3.6IE+06 8.99E+00 Total All CET End States 3.04E-05*

N/A 11.38

• The CDF remains at 2.93E-05 Based on the above values, the Type A 15 year test frequency percent oftota! dose rate (DR) for Class 3 is as follows:

% ofTota! DR15 = [(CLASS3AI5 + CLASS3B1S) / Tota!I5] x 100 where:

CLASS3A15 = Class 3A person-rem/year = 2.37E-02 person-rem/year

[Table 5]

CLASS3BIS =Class 3B person-rem/year = 8.28E-03 person-rem/year Totahs = total person-rem/year for I5-year interval = 11.38 person-rem/year

[Table 5]

[Table 5]

% ofTotal DR15= [(2.37E-02 + 8.28E-03) /11.38] x 100

% of Total DRlS = 0.281%

Therefore, the total 15 year test frequency ILRT interval percent oftotal dose rate related to Type A testing is 0.281 %.

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.15 Part V, Volume RA.LI.4.REVISION 2 RISK ANALYSIS -

¢alculation for a 5 Year Integrated Leak Rate Test Extension for

~Jnit 2 The i\\% change in the 15 year ILRT DR from the baseline value is 0.281% - 0.056% =0.225%.

The..1.% change in the total dose rate between the ten-to-fifteen year intervals related to Type A testing IS:

i\\% ChangelO-ls = % ofTotal ORIs - % ofTotal DRIO = 0.281% - 0.186% = 0.095%

The fifteen-year dose rate Mange (related to an ILRT) over the baseline case is as follows:

DR Changels = [2:

D (Class 1, 3A, 3B)15 - 2:

D (Class 1, 3A, 3B)Base]

where:

2:

D (Class 1, 3A, 3B)Base = 1.38B-03 + 4.73E-03 + 1.65B-03 person-rem/year 2:n (Class 1, 3A, 3B)Base = 7.76E-03 person-rem/year 2:

D (Class 1, 3A, 3Bhi5 = O.OOE+OO + 2.37E-02 + 8.28E-03 person-rem/year 2:

D (Class 1, 3A, 3B)115 = 3.20E-02 person-rem/year DR Change15 = [3.201E-02 -7.76B-03] person-rem/year DR Changel5 =2.42:E-02person-remlyear

[Table 3]

[Table 5]

Therefore, the fifteen-year d~)se rate change from the baseline case is 2.42E-02 person-rem/year.

Step 4 - Determine the cha,ge in risk in terms ofLarge Early Release Frequency (LERF)

The one time extension ofinpreasing the Type A test interval involves establishing the success criteria for a large release. This criterion is based on two prime issues:

1) The containment l~ak rate versus breach size, and

2) The impact on ris1<l: versus leak rate.

The containment leak size far the corresponding leak rate was calculated using the same methodology as in [CALC04]. The leak size',and the corresponding leak rate data for Surry were used from [CALC08].

The effect ofcontainment l~ size on the containment leak rate is shown in Table 6. In addition, Oak Ridge National Laboratory (ORNL) [REPORT06] completed a study evaluating the impact ofleak rates on public risk using information from WASH-1400 [REPORT07] as the basis for its risk sensitivity calculations (see Figure 1).

Based upon the information in Table 6 and ORNL, it is judged that small leaks resulting from a severe accident (that are deemed nQt to dominate public risk) can be defined as those that change risk by less than 5%. This definition wQuld include leaks ofless than 35%/day. Based on the Table 6 data, a 35%/day containment leak rate equates to a diameter leak ofslightly smaller than 0.7 inches. It is to be

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.16 Part V. Volume RA.LI.4. REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 noted that for Surry a containment leak with a diameter ofO.? inches was calculated as opposed to 2.0 inches for Indian Point 3. This difference in containment leak diameter is due to the difference in containment size between Surry and Indian Point 3. Therefore, this study defines small leakage as containment leakage resulting from an opening of0.012in2 or less, large leakage as greater than 0.012in2 and negligible leakage as O.OOli~ or less.

tL kR t C t' tL kS' t fC taO T bl 6 E ltd I a

e va ua e mpac 0

on mmen ea lZeon on ammen ea ae Containment Leak Size Approximate Containment Leak Rate at Design Pressure Diameter (inches)

Area La (in2)

(wt%/day) 0.036 0.001 0.1 (acceptable by Tech Specs) 0.115 0.010 1.0 (lOla) 0.126 0.012 3.5 (35La) 0.364 0.104 10.0 0.681 0.363 35.0 1.152 1.043 100.0 5.647 25.05 2400 The risk impact associated with extending the ILRT interval involves the potential ofa core damage event, normally resulting 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 ifa pre-existing leak were present.

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

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.I?

Part V, Volume RA.LI.4, REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 Figure 1: Fractional Impact on Risk Associated with Containment Leak Rates [REPORT06]

0.12 ~---r"""";-++-r+++f-.--,-".f**""**-+*~H+/-~H---~4*d*-4~~H Late releases are excluded regardless ofthe size ofthe leak because late releases are, by definition, not a LERF event. At the same time, sequences in the Surry IPE [REPORT09], which result in large releases (e.g., large isolation valve failures), are not impacted because a LERF will occur regardless ofthe presence ofa pre-existing leak. Therefore, the frequency ofClass 3B sequences (Table 4) is used as the LERF for Surry. This frequency, based on a ten-year test interval, is 2.63E-07/ yr.

Reg. Guide 1.174 [RG01] provides guidance for detemlining the risk impact ofplant-specific changes to the licensing basis. Since the ILRT does not impact CDF, the relevant metric is LERF. Regulatory Guide 1.174 [R001] states, when the calculated increase in LERF is in the range of 10-7 per reactor year to 10-6 per reactor year, applications will be considered only ifit can be reasonably shown that the total LERF is less than 10-5 per reactor year (Region II). Calculating the increase in LERF requires detemrining the impact ofthe ILRT interval on the leakage probability.

Step 5 - Evaluate the Risk Impact in Terms of LU-ERF The.6. LERF from Base to once-per-IO years (10 year metrics) is calculated to be the difference between the Class 3B frequencies in Tables 3 and 4.

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.18 Part V. Volume RA.LI.4. REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 A LERF = Class 3BlO - Class 3BBase A LERF = 2.63E M 07 -7.91E-08= 1.84E-07 The baseline total LERF for Surry has been calculated to be 6.62E-07 / yr in [NBO1].

This A LERF increases the baseline LERF to 6.62E-07 + 1.84E-07 = 8.46E-07 / yr.

The A LERF from Base to once-per-I5 years (15 yearmetrics) is calculated to be the difference between the Class 3B frequencies in Tables 3 and 5.

A LERF = Class 3B15 - Class 3BBase A LERF = 3.96E 7.91E-08 = 3.17E-07 This A LERF increases the baseline LERF to 6.62E-07 +3.17E-07 = 9.79E-07 / yr.

The ALERF from once-per-I0 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 = Class 3B15 - Class 3BlO A LERF = 3.96E 2.63E-07 = 1.33E-07 The guidance in [RGOl] states that when the calculated increase in LERF is in the range of 10.7 per reactor year to lO*6 per reactor year, applications will be considered only ifit can be shown that the total LERF is less than 10.5per reactor year. The total new LERF value for the 15 year change for SPS has been calculated to be 9.79E-07 / yr. Since guidance in [Rool] defines small changes in LERF, thus the ma~tudein the difference between the 10 year and 15 year LERF value (1.33E-07) is in the range of 10* /yr to 10.6 /yr, increasing the ILRT interval to 15 years is considered acceptable.

Step 6 - Determine Impact on Conditional Containment Failure Probability Another parameter that the NRC Guidance in [ROOI] states can provide input into the decision making process is the consideration ofchange in the Conditional Containment Failure Probability (CCFP). The change in CCFP is indicative ofthe effect ofthe ILRT on all radionuclide releases, not just LERF. The CCFP can be calculated from the risk calculations perfonned in this analysis.

In this assessment, based on the NEI Interim Guidance [REPORT02], 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 ofthe 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 [REPORT02]:

CCFP% = [l-<<Classl Frequency +Class 3AFrequency) / Total CDF)] x 100%

For the Base interval, the values are obtained from Table 3:

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.19 Part V, Volume RA.LI.4. REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 CCFPBase = [1 - <<2.30E-06 + 7.91E-07) / 2.93E-05)] x 100%

CCFPBase= 89.5%

For the 1O-year interval, the values are obtained from Table 4:

CCFPlO =[1- <<2.77E-07 + 2.63E-06) / 2.93E-05)] x 100%

CCFPlO = 90.1%

For the 15-year interval, the values are obtained from Table 5:

CCFPl5 = [1 - <<O.OOE-OO + 3.96E-06) / 2.93E-(5)] x 100%

CCFP15 = 86.5%

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

L\\CCFP% =CCFPl5 - CCFPlO =-3.6%

(see Note 1)

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

L\\CCFP% = CCFPIO - CCFPBase=0.6%

The 15 year change in the conditional containment failure probability is:

L\\CCFP% = CCFPl5 - CCFPBase = -3.0%

(see Note 1)

This 15 year calculated change in CCFP% is negative. Intuitively, this is not realistic and the expeCted 15 year change in the CCFP% is negligible.

Note 1: The negative ACCFP% is caused by the adjustment ofthe Class 1frequency to preserve the total CDF. In reality, the factor of5 increases ofthe Class 3 frequencies for the 15 year case cannot occur without an adjustment in other, more severe release categories.

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.20 Part V, Volume RA.LI.4, REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 External Event Sensitivity Analysis The SPS IPE [REPORT09] has limited discussion pertaining to external events, and it appears that external events would have the largest impact on the EPRI Class 7 event for the ILRT evaluation.

However, in the Severe Accident Mitigation Alternatives analysis (SAMA) for the SPS license renewal

[CALC02], a factor was used to account for the potential impact ofexternal events.

The benefits of each SAMA were multiplied by a factor of2.0 to account for the external events. This factor could be applied to the CDP used here to calculate the EPRI 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 = 7.91E-08 x 2.0 = 1.58E-07 /yr 15 year Class 3B frequency = 3.96E-07x 2.0 =7.92E-07 /yr The external events change in LERF from the Baseline to the 15 year test interval is 6.34E-07 /yr (7.92E 1.58E-07). This compares to the internal events Baseline to 15 year change in LERF as 3.17E-07 /yr (3.96E 7.91 E-08).

Thus it has been independently shown that with external events included, the change in LERF due to a 15 year ILRT interval still meets the screening criterion in [ROOl]. Since guidance in [ROOl] defines small changes in LERF in the range of 10*7/yr, increasing the ILRT interval to 15 years is considered acceptable.

Linear Corrosion Analysis The approach documented in the Calvert Cliffs Nuclear Power Plant submittal in [REPORT08] was used to determine the change in likelihood, due to extending the ILRT, ofdetecting liner corrosion. This likelihood was then used to determine the resulting change in risk. The following issues are addressed:

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

Assumptions A.

A halffailure is assumed for basemat concealed liner corrosion due to the lack ofidentified failures. (See Table 7, Step 1.)

B.

The success data was limited to 5.5 years to reflect the years since September 1996 when 10 CFR 50.55a started requiring visual inspection. Additional success data was not used to limit the

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.21 Part V. Volume RA.LI.4. REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 aging impact ofthis 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).

C.

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

Sensitivity studies are included that address the doubling ofthis rate every 10 years and every two years (see Table 7, Steps 2 and 3).

D.

The likelihood ofthe containment atmosphere reaching the outside atmosphere given a liner flaw exists, is a function ofthe pressure inside the Containment. Even without the liner, the Containment is an excellent barrier. But as the pressure in Containment increases, cracks will form. Ifa 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 ofcontainment 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 detennined through logaritlunic interpolation. Sensitivity studies are included that decrease and increasethe 20 psia anchor point by a factor of10 (see Table 4 of [REPORT08] for sensitivity studies).

E.

The likelihood of l~age 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).

F.

A 5% visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 10% is used. To date, all liner corrosion events have been detected through visual inspection. (See Table 7, Step 5.) Sensitivity studies are included that evaluate total detection failure likelihoods of5% (see Table 4 [REPORT08] for sensitivity studies).

G.

All non-detectable containment over-pressurization failures are assumed to be large early releases. This approaoh avoids a detailed analysis ofcontainment failure timing and operator recovery actions.

Table 7: Liner Corrosion Base Case Step Description Containment Cylinder Containment Basemat and Dome 1

Historical Liner Flaw Likelihood Events: 2 Events: 0 Failure Data: Containment location Assume half a failure specific.

(Brunswick 2 and North Success Data: Based on 70 steel-lined Anna 2) 0.5/(70*5.5) = 1.3E-3 Containments and 9 years since the 2/(70*5.5) = 5.2E-3 10 CPR 50.55a requirement for periodic visual inspections of containment surfaces.

2 Aged Adjusted Liner Flaw Flaw Flaw Likelihood Year Likelihood Year Likelihood During 15-year interval, assume 0

1.79E-03 0

4.47E-04 1

2.0SE-03 1

5.13E-04

SPS PROBABILISTIC !USK ASSESSMENT NOTEBOOK P.22 Part V, Volume RA.LI.4, !mVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 Step Description Containment Cylinder Containment Basemat and Dome failure rate doubles every five years 2

2.36E-03 2

5.89E-04 (i.e. a 14.9% increase per year). The 3

2.71E-03 3

6.77E-04 average over the 5th through lOth year 4

3.11E-03 4

7.77E-04 period was set to the historical failure 5

3.57E-03

~

8.93E-04 oJ rate of Step 1 (See Table-5 from 6

4.10E-03 6

1.03E-03

[REPORT08] for an ~ample). These 7

4.71E-03 7

1.18E-03 assumptions are used to calculate the 8

5.41E-03 8

1.35E-03 flaw likelihood for each year (for a 9

6.22E-03 9

1.55E-03 15 year period).

10 7.14E-03 10 1.79E-03 11 8.21E-03 11 2.05E-03 12 9.43E-03 12 2.36E-03 13 1.08E-02 13 2.71E-03 14 1.24E-02 14 3.11E-03 15 1.43E-02 15 3.57E-03 3

Increase in Flaw LiU.elihood 0.71% (1 to 3 years) 0.18% (1 to 3 years)

Between 3, 10, and 15 years 4.14% (1 to 10 years) 1.03% (1 to 10 years) 9.65% (1 to 15 years) 2.41% (1 to 15 years)

This cumulative prob~i1ityuses the age adjusted liner flaW likelihood of (Note that the Calvert (Note that the Calvert Step 2 (see Tables 5 and 6 in Cliffs analysis presents the Cliffs analysis presents the

[REPORT08]). For elK-ample, the delta between 3 and 15 delta between 3 and 15 7.12E-03 (at 3 years) cumulative flaw years of8.7% to utilize in years of2.2% to utilize in likelihood is the sum ofthe year 1, the estimation ofthe the estimation ofthe year 2, and year 3 lik¢lihoods ofstep ALERF value. For this ALERF value. For this 2.

analysis, however, the analysis, however, the values are calculated based values are calculated based on the 3,10, and 15 year on the 3, 10, and 15 year intervals consistent with intervals consistent with the desired presentation of the desired presentation of the results.)

the results.)

4 Likelihood of Breacl~in Pressure Likelihood Pressure Likelihood Containment given J..iner Flaw (psia) ofBreach (psia) ofBreach The upper end pressute is consistent 20 0.1%

20 0.01%

with the Calvert Cliffs Probabilistic 64.7 (ILRT) 1.1%

64.7 (lLRT) 0.11%

Risk Assessment (pItA.) Level 2 100 7.02%

100 0.7%

analysis. 0.1% is asSlilIlled for the 120 20.3%

120 2.0%

lower end. Intermediate failure 150 100%

150 10%

likelihoods are detem1lined through logarithmic interpolation. The basemat is assumed to be 1/10 ofthe cylinder/dome analysis.

SPS PROBABn..ISTIC RISK ASSESSMENT NOTEBOOK P.23 Part V, Volume RA.LI.4. REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 Step Description Containment Cylinder Containment Basemat and Dome The same value will be used for SPS as was used for CCNP, since the containment design is somewhat similar. The design pl'lessure of SPS is 45 psig versus 50 psig for CCNPP.

5 Visual Inspection Detection Failure 10%

100%

Likelihood 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.

6 Likelihood ofNon-Detected 0.0106%

0.0027%

Containment Leakage (Steps 3*4*5) 9.65%*1.1%*10%

2.41%*0.11%*100%

The total likelihood ofthe 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 = 0.0106% + 0.0027% = 0.0133%

The non-large early release :&equency (LERF) containment over-pressurization failure for SPS is estimated at 2.44E-05 per year. This is based on the total CDF minus the Class 1, 3B and 8 frequencies from Table 1 (2.44E-05 =2.93E (2.30E-06 + 7.91E-08 + 2.49E-06>>. The total CDF for SPS is 2.93E-05. Ifall non-detectabile containment leakage events are considered to be LERF, then the increase in LERF associated with the liner corrosion issue is:

Increase in LERF (lLRT 3 to 15 years) = 0.000133

• 2.44E-05 =3.25E-09 per year Thus it has been independently shown that the increase in LERF due to a liner corrosion failure is 3.25E-09 per year which meets the screening criterion ofless than 10-1 in [R001].

SPS PROBABll..ISTIC RISK ASSESSMENT NOTEBOOK P.24 Part V, Volume lllA.LI.4. REVISION 2 RISK ANALY8IS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 5.0 RESULTS AND CONCLUSIONS 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 1.33E-07 / yr.

Therefore, the risk impact when compared to other severe accident risks is small. The results for the Baseline, 10 year and 15 year ILRT evaluation are summarized in Table 8below.

ISPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.25 Part V, Volume M.LI.4, REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 Table 8: Summary of Results EFRI Balit:Frt:q Balit: (311'1)

DaSevUse

'W'. ___

'II"'lI._..L_

.. 1\\_-- ",_. __

lln_.~ n ___ 'T'1I_L_

1n___

1l!!___ lr.'-___

... ~_~ 'n..- ___ n

_<4-_

15ji' Dlilit:.uu:",~ niilll:

~UJI.l'.l:lJ.

~UJ'*

.lJ'U~l:.R.iilLl:

I.vy.

1..:11. 1.'1 "'l{

A..,}3J.

.A.lU~1;.L,-at.lI; Class IRx-yr FreQ I Rx-yr Person-Rem Person-Rem I yr I Rx-yr Person-Rem I yr Metrics I Rx-yr Person-Remlyr Metrics 1

3.17E-06 2.30E-06*

5.98E+02 1.3SE-03 2.77E-07 1.66E-04 O.OOE-OO O.OOE+OO 2

6.52E-IO 6.52E-IO 2.38E+04 I.S5E-OS 6.52E-I0 1.5SE-05 6.52E-I0 I.SSE-05 3A N/A 7.9IE-07 5.98E+03 4.73E-03 2.63E-06 1.57E-02 3.96E-06 2.37E-02 3B N/A 7.9IE-OS 2.09E+04 1.65E-03 2.63E-07 5.50E-03 3.96E-07 8.28E-03 4

N/A N/A N/A N/A N/A N/A N/A N/A 5

N/A N/A N/A N/A N/A N/A N/A N/A 6

N/A N/A N/A N/A N/A N/A N/A N/A 7

2.36E-05 2.36E-OS 1.0OE+05 2.36 2.36E-05 2.36 2.36E-05 2.36E+OO 8

2.49E-06 2,49E-06 3.61E+06 8.99 2.49&06 8.99 2,49E-06 8.99E+OO TOTAL 2.93E-05 2.93E-05 N/A 11.36 2.93E-05 11.37 3.04E-05 11.38 (CDF)

(CDF)

(CDF)

(CDF)

.6DRChange l.36E-02 2,42E-02 from Base ILRTDR 0.056%

0.186%

0.281%

% oftotal dose

.6.% Change 0.130%

.0225%

in ILRT DR from Base LERF 6.62E-07 8,46E-07 9.79E-07

.6.LERF 1.84E-07 3.17E-07 From base CCFP, %

89.5%

90.1%

87.0%

.6.CCFP %

0.60%

-2.50%

CDP 2.93E-05

SPS PROBABll.ISTIC RISK ASSESSMENT NOTEBOOK P.26 Part V. Volume ]RA.LI.4. REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2

6.0 REFERENCES

[CALC01]

SM-1241 Revision 0, "MACCS2 model for Surry Level 3 Application", 2-28-2000

[CALC02]

SM-1256 Revision 1, "Surry Severe Accident Mitigation Alternative (SAMA)", 09 2000.

[CALC03]

IP3-CALC-VC-03357 Revision 0, "Indian Point 3 Risk Impact Assessment ofExtending Conuinment Type A Test Interval", 1-4-2001

[CALC04]

SM-1237 Revision 0, "Surry and North Anna Containment Isolation Modeling",4 2000

[CALC05]

SM-1237 Revision 0, Addendum A "Surry and North Anna Containment Isolation Modding",4-24-2001

[CALC06]

SM-1325 Revision 0, Addendum A "Risk Impact Assessment ofExtending Containment Type A Test Interval at North Anna Power Station", 10-10-2001

[CALC07]

PRA06NQA-04178S3 Revision 2, "Risk Impact Assessment ofExtending Containment Type A Test Interval at Millstone Unit 3", 11-21-2003

[CALC08]

SM-1321 Revision 0, "Risk Impact Assessment ofExtending Containment Type A Test Interval at Surry Power Station", 8-14-2001

[CALC09]

C467060036-6915 Revision 0, "LaSalle ILRT Interval Extension Risk Assessment",

March 2006

[CALC10]

Florida Power Calculation, F-01-0001, Revision 2, "Evaluation ofRisk Significance of ILRT Extension", 6-19-01

[CALC11]

SM-1345 Revision 0, "Revised LERF Fractions Using Updated Surry Level 2 Model",

November 2003

[NB01]

PRA Model Notebook QU.2 Rev.3, "Model Quantification Results," Surry Power Station Units I&2, June 2006

[NUREG01]

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

[NUREG02]

NUREG-1335, "Individual Plant Examination: Submittal Guidance," August 1989.

[PROCDR01] Proc(x!ure, Engineering Periodic Test 2-NPT-CT-101, "Reactor Containment Building Integrated Leak Rate Test (Type A Containment Testing)", Revision 5

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.27 Part V, Volume RA.LI.4. REVISION 2 RISK ANALYS,IS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2

[REPORTOl] NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J", July 26, 1995, Revision 0

[REPORT02] Interim Guidance for Performing Risk Impact Assessments In Support ofOne-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Developed for NEI by EPRI, November 2001

[REPORT03] EPRI TR-I04285, "Risk Assessment ofRevised Containment Leak Rate Testing Intervals", August 1994

[REPORT04] EPRI TR~109325, "Risk Impact Assessment ofExtended Leak Rate Testing Intervals" December 2003

[REPORT05] Z. T. \\1endoza, et aI., "Generic Framework for Individual Plant Examination (IPE)

Back<::nd (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

[REPORT06] Burns, T.J., "Impact ofContainment Building Leakage on LWR Accident Risk'" Oak Ridge National Laboratory, NUREG/CR-3539, April 1984

[REPORT07] United States Nuclear Regulatory Commission, Reactor Safety Study, WASH-1400, October 1975

[REPORT08] Calvert Cliffs Nuclear Power Plant, Letter from Mr. Charles H. Cruse to NRC Document Control Desk, "Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leak Rate Test Extension", dated March 27,2002

[REPORT09] Probabilistic Risk Assessment Final Report "Surry Power Station Units 1 and 2 Individual Plant Examination", August 1991

[RG01]

Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk~lnfonnedDecisions On Plant-Specific Changes to the Licensing Basis" November 2002, Revision 1

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P.Al Part V, Volume UA.LL4, REVISION 2 RISK ANALYS,IS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2

~\\TTACHMENT A: SPS FREQUENCY AND DOSE DATA dD D t T bl A 1 SPS F a

e -.

requencyan ose aa Release Frequency*

Person-Rem**

EPRI Description Category Per year Class 1

O.OOE+OO 1

NoCF 2

2.5SE-06 5.98E+02 1

NoCF 3

O.OOE+OO 5.15E+06 8

7 EarlyCF 4

1.41E-07 2.S0E+04 8

7 LateCF S

9.16E-09 8.23E+OS a 7

LateCF 6

7.79E-09 2.50E+04 a 7

Late CF 7

8.lOE-07 8.23E+OS a 7

LateCF 8

4.43E-07 2.89E+OS 8

7 LateCF 9

1.3SE-OS 7.1OE+04a 7

LateCF 10 8.45E-06 7.1OE+04 a 7

LateCF 11 2.89E-07 2.50E+04 7

Melthru 12 O.OOE+OO 5.98E+02 R

2 No Cont. Iso 13 6.52E-10 4.71E+OS 8

2 No Cont. Iso 14 6.23E-07 1

Debris Cool IV 15 O.OOE+OO 1.19E+04 0 2

Debris Cool IV 16 O.OOE+OO 1.19E+04 0 2

Debris Cool IV 17 1.37E-06 2.7SE+06 a 8

Event V (attenuation) 18 2.42E-07 6.81E+06 8

8 Event V (no attenuation) 19 5.33E-07 5.07E+06 a 8

SGTR 20 3.42E-07 2.54E+06 c 8

SGTR (non-LERF)

CDFFreq 2.93E-05

• Frequency data taken from [CALCll].

• • Person-Rem data taken from [CALC01].

8 Recommended Altl~matevalues were used consistent with the IPE, SAMA analysis, and [CALC01].

bUsed IPE STC 21 JOT STC 15 and 16 based on review ofMAAP runs.

e Dose for STC 20 was assumed to be halfthat ofSTC 19 because it is a non-LERF event.

dD D ta T t IEPRI Cl F

. urry oa ass re ~uencyan ose a

J Frequency Dose Class 1 3.17E-06 5.98E+02 Class 2 6.52E-10 2.38E+04 Class 7 2.36E-05 7.30E+06 Class 8 2.49E-06 1.72E+08 T\\BLEA 2 S

SPS PROBABll..ISTIC RISK ASSESSMENT NOTEBOOK P. Bl Part V, Volume RA.LI.4, REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 ATTACHMENT B: JUSTIFICATION OF VOLUME CHANGE Revision 0 This is the original revision.

Revision 1 Editorial revision for enhancement and clarity prior to NRC LAR submittal.

Revision 2 Revised to clarify why the Conditional Containment Failure Probability was negative.

SPS PROBABILISTIC RISK ASSESSMENT NOTEBOOK P. Cl Part V, Volume lllA.LI.4. REVISION 2 RISK ANALYSIS -

Calculation for a 5 Year Integrated Leak Rate Test Extension for Unit 2 ATTACHMENT C: REVIEWER COMMENTSIRESOLUTIONS Comment Section!

Review Response to Review Comment Number Pa~e Comment 1

All Minor editorial comments Corrected 2

All Editorial revisions for Done enhancement and clarity 3

14 & 19 Clarification ofnegative CDF Corrected and CCFP calculations

Serial No. 07-0802 Docket No. 50-281 One-Time Five-Year Extension to Type A Test Interval Marked-up Technical Specifications Page Surry Power Station Unit2 Virginia Electric and Power Company (Dominion)

TS 4.4-1 4.4 CONTAINMENT TESTS Applicability Applies to containment leakage testing.

Objective To assure that leakage of the primary reactor containment and associated systems is held within allowable leakage rate limits; and to assure that periodic surveillance is performed to assure proper maintenance and leak repair during the service life of the containment.

Specification A. Periodic and post-operational integrated leakage rate tests of the containment shall be performed in accordance with the requirements of 10 CFR 50, Appendix J, "Reactor Containment Leakage Testing for Water Cooled Power Reactors."

B. Containment Leakage Rate Testing Requirements 1.

The containment and containment penetrations leakage rate shall be demonstrated by performing leakage rate testing as required by 10 CFR 50 Appendix J, Option B, as modified by approved exemptions, and in accordance with the guidelines contained in Regulatory Guide 1.163, dated September, 1995 as modified by the following exception:

NEI 94-01-1995, Section 9.2.3: The first Unit 24-Type A test performed after the October 26, 2000 April 23, 1992 Type A test shall be performed no later than October 26,2015 April 22, 2007.

2.

Leakage rate acceptance criteria are as follows:

a.

An overall integrated leakage rate of less than or equal to La, 0.1 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).

b.

A combined leakage rate ofless than or equal to 0.60 La for all penetrations and valves subject to Type Band C testing when pressurized to Pa.

Prior to entering an operating condition where containment integrity is required the as-left Type A leakage rate shall not exceed 0.75 La and the combined leakage rate of all penetrations subject to Type Band C testing shall not exceed 0.6 La.

3. The provisions of Specification 4.0.2 are not applicable.

Basis The leak tightness testing of all liner welds was performed during construction by welding a structural steel test channel over each weld seam and performing soap bubble and halogen leak tests.

Amendment Nos.

Serial No. 07-0802 Docket No. 50-281 One-Time Five-Year Extension to Type A Test Interval Proposed Technical Specifications Page Surry Power Station Unit2 Virginia Electric and Power Company (Dominion)

TS 4.4-1 4.4 CONTAINMENTlESTS Applicability Applies to containment leakage testing.

Objective To assure that leakage of the primary reactor containment and associated systems is held within allowable leakage rate limits; and to assure that periodic surveillance is performed to assure proper maintenance and leak repair during the service life of the containment.

Specification A. Periodic and post-operational integrated leakage rate tests of the containment shall be performed in accordance with the requirements of 10 CFR 50, Appendix J, "Reactor Containment Leakage Testing for Water Cooled Power Reactors."

B. Containment Leakage Rate Testing Requirements 1.

The containment and containment penetrations leakage rate shall be demonstrated by performing leakage rate testing as required by 10 CFR 50 Appendix J, Option B, as modified by approved exemptions, and in accordance with the guidelines contained in Regulatory Guide 1.163, dated September, 1995 as modified by the following exception:

NEI 94-01-1995, Section 9.2.3: The first Unit 2 Type A test performed after the October 26, 2000 Type A test shall be performed no later than October 26, 2015.

2.

Leakage rate acceptance criteria are as follows:

a.

An overall integrated leakage rate of less than or equal to La, 0.1 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).

b.

A combined leakage rate ofless than or equal to 0.60 La for all penetrations and valves subject to Type Band C testing when pressurized to Pa.

Prior to entering an operating condition where containment integrity is required the as-left Type A leakage rate shall not exceed 0.75 La and the combined leakage rate of all penetrations subject to Type Band C testing shall not exceed 0.6 La.

3. The provisions of Specification 4.0.2 are not applicable.

Basis The leak tightness testing of all liner welds was performed during construction by welding a structural steel test channel over each weld seam and performing soap bubble and halogen leak tests.

Amendment Nos.