ML082210549

Issuance of Amendment Regarding Changes to the Technical Specifications Related to the Isolation Valve Seal Water System

ML082210549
Person / Time
Site:	Robinson
Issue date:	10/03/2008
From:	Vaaler M Plant Licensing Branch II
To:	Walt T Carolina Power & Light Co
Vaaler, Marlayna, NRO/DNRL 415-3178
Shared Package
ML082210560	List: ML082210549
References
TAC MD7469
Download: ML082210549 (17)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 3, 2008 Mr. Thomas D. Walt, Vice President H. B. Robinson Steam Electric Plant, Unit NO.2 Carolina Power & Light Company 3581 West Entrance Road Hartsville, South Carolina 29550-0790

SUBJECT:

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.2 - ISSUANCE OF AMENDMENT REGARDING CHANGES TO THE TECHNICAL SPECIFICATIONS RELATED TO THE ISOLATION VALVE SEAL WATER SYSTEM (TAC NO. MD7469)

Dear Mr. Walt:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 220 to Renewed Facility Operating License No. DPR-23 for the H.B. Robinson Steam Electric Plant, Unit No.2, in response to your application dated November 29, 2007. The amendment consists of changes to Technical Specification Section 3.6.8, "Isolation Valve Seal Water (IVSW) System."

The amendment revises Surveillance Requirements (SR) 3.6.8.2 and 3.6.8.6 related to IVSW tank volume and header flow rates. Specifically, the change clarifies the wording of SR 3.6.8.2, and revises SR 3.6.8.6 to provide a total flow rate limit from all four headers in place of the individual header limits.

A copy of the related safety evaluation is also enclosed. The Notice of Issuance will be included in the NRC's next biweekly Federal Register notice.

Sincerely, r-"~7 'cJ, h r . ~).;; l ("

ft.* Marlayna Vaaler, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-261

Enclosures:

1. Amendment No. 220 to DPR-23

2. Safety Evaluation cc w/enclosures: See next page

Mr. T. D. Walt H. B. Robinson Steam Electric Plant, Carolina Power & Light Company Unit No.2 cc:

Mr. Ernest J. Kapopoulos, Jr. Associate General Counsel II - Legal Plant General Manager Department H. B. Robinson Steam Electric Plant, Progress Energy Service Company, LLC Unit NO.2 Post Office Box 1551 Progress Energy Carolinas, Inc. Raleigh, North Carolina 27602-1551 3581 West Entrance Road Hartsville, South Carolina 29550 Public Service Comm ission State of South Carolina Mr. Eric McCartney Post Office Drawer 11649 Director of Site Operations Columbia, South Carolina 29211 H. B. Robinson Steam Electric Plant, Unit No.2 Ms. Beverly Hall, Section Chief Progress Energy Carolinas, Inc. N.C. Department of Environment 3581 West Entrance Road and Natural Resources Hartsville, South Carolina 29550 Division of Radiation Protection 3825 Barrett Dr.

Mr. Curt Castell Raleigh, North Carolina 27609-7721 Supervisor, Licensing/Regulatory Programs H. B. Robinson Steam Electric Plant, Mr. Robert P. Gruber Unit NO.2 Executive Director Progress Energy Carolinas, Inc. Public Staff - NCUC 3581 West Entrance Road 4326 Mail Service Center Hartsville, South Carolina 29550 Raleigh, North Carolina 27699-4326 Mr. C. T. Baucom Ms. Susan Jenkins Manager - Support Services* Nuclear South Carolina Department of Health H. B. Robinson Steam Electric Plant, Bureau of Land & Waste Management Unit NO.2 2600 Bull Street Progress Energy Carolinas, Inc. Columbia, South Carolina 29201 3581 West Entrance Road Hartsville, South Carolina 29550 Ms. Margaret A. Force Assistant Attorney General Mr. Scott D. West State of North Carolina Superintendent - Security Post Office Box 629 H. B. Robinson Steam Electric Plant, Raleigh, North Carolina 27602 Unit NO.2 Progress Energy Carolinas, Inc. Mr. John H. O'Neill, Jr.

3581 West Entrance Road Pillsbury Winthrop Shaw Pittman, LLP Hartsville, South Carolina 29550 2300 N Street NW.

Washington, DC 20037-1128 U. S. Nuclear Regulatory Commission Resident Inspector's Office Robert J. Duncan II H. B. Robinson Steam Electric Plant Vice President, Nuclear Operations 2112 Old Camden Road Progress Energy Carolinas, Inc.

Hartsville, South Carolina 29550 3581 West Entrance Road Hartsville, South Carolina 29550 Mr. J. Paul Fulford Manager, Performance Evaluation and Brian C. McCabe Regulatory Affairs PEB 5 Manager, Nuclear Regulatory Affairs Progress Energy Carolinas, Inc. Progress Energy Carolinas, Inc.

Post Office Box 1551 3581 West Entrance Road Raleigh, North Carolina 27602-1551 Hartsville, South Carolina 29550 Mr. David T. Conley

October 3, 2008 Mr. Thomas D. Walt, Vice President H. B. Robinson Steam Electric Plant, Unit No.2 Carolina Power & Light Company 3581 West Entrance Road Hartsville, South Carolina 29550-0790

SUBJECT:

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.2 - ISSUANCE OF AMENDMENT REGARDING CHANGES TO THE TECHNICAL SPECIFICATIONS RELATED TO THE ISOLATION VALVE SEAL WATER SYSTEM (TAC NO. IVID7469)

Dear Mr. Walt:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 220 to Renewed Facility Operating License No. DPR-23 for the H.B. Robinson Steam Electric Plant, Unit No.2, in response to your application dated November 29,2007. The amendment consists of changes to Technical Specification Section 3.6.8, "Isolation Valve Seal Water (IVSW) System."

The amendment revises Surveillance Requirements (SR) 3.6.8.2 and 3.6.8.6 related to IVSW tank volume and header flow rates. Specifically, the change clarifies the wording of SR 3.6.8.2, and revises SR 3.6.8.6 to provide a total flow rate limit from all four headers in place of the individual header limits.

A copy of the related safety evaluation is also enclosed. The Notice of Issuance will be included in the NRC's next biweekly Federal Register notice.

Sincerely, IRA by FSaba forI Marlayna Vaaler, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-261

Enclosures:

1. Amendment No. 220 to DPR-23

2. Safety Evaluation cc w/enclosures: See next page DISTRIBUTION:

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 CAROLINA POWER & LIGHT COMPANY DOCKET 1\10. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 220 Renewed License No. DPR-23

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Carolina Power & Light Company (the licensee), dated November 29,2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this*

amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 3.B. of Renewed Facility Operating License No. DPR-23 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 220 are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

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3. This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION

,./1 /-?

/' /. /. )?vJ~ .

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Tho;as'H. BOy;e, 'Chiet!

Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-23 and the Tec~lnical Specifications Date of Issuance: October 3, 2008

ATTACHMENT TO LICENSE AMENDMENT NO. 220 RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace page 3 of Renewed Operating License No. DPR-23 with the attached page 3.

Replace the following pages of Appendix A of the Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Page Insert Page 3.6-20 3.6-20 3.6-21 3.6-21

-3 neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; E. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A. Maximum Power Level The licensee is authorized to operate the facility at a steady state reactor core power level not in excess of 2339 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 220 are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

(1) For Surveillance Requirements (SRs) that are new in Amendment 176 to Final Operating License DPR-23, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 176. For SRs that existed prior to Amendment 176, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 176.

Renewed Facility Operating License No. DPR-23 Amendment No. 220

Isolation Valve seal Water System 3.6.8 3.6 CONTAINMENT SYSTEMS 3.6.8 Isolation Valve Seal Water (IVSW) System LCO 3.6.8 The IVSW System shall be OPERABLE.

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

ACTIOOS coomTIOO REQUIRED ACTIOO CCt1PLETHJ'J TIME A. IVSW system A.I Restore IVSW system 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. to OPERABLE status.

B. Required Action B.1 Be in MOOE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and associated C~l etion Tine

-AND not rret.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREOCIENCY SR 3.6.8.1 Verify IVSW tank pressure is ~ 46.2 pSig. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.6.8.2 Verify the IVSW tank water volume is 31 days

~ 85 ga 11 ons .

(continued)

HBRSEP Unit No. 2 3.6-20 Amendnent No .220

Isolation Valve Seal Water System 3.6.8 SURVEILLANCE REQUIREMENTS (continued)

SURVEY LLANCE FREOOENCY SR 3.6.8.3 Verify the opening time of each air In accordance operated header injection valve is within With the limits. Inservice Testing Program SR 3.6.8.4 Verify each automatic valve in the IVSW 18 mnths System actuates to the correct position on an actual or simulated actuation signal.

SR 3.6.8.5 Verify the IVSW dedicated nitrogen 18 months bottles will pressurize the IVSW tank to

~ 46.2 psig.

SR 3.6.8.6 Verify total IVSW seal header flow rate 18 months is S 124 cc/m1nute.

HBRSEP Unit NO.2 3.6-21 Amendment No. 220

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 220 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-23 CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON STEAM ELECTRlC PLANT, UNIT NO.2 DOCKET NO. 50-261

1.0 INTRODUCTION

By application dated November 29, 2007 (Agencywide Documents Access Management System (ADAMS) Accession No. ML073381185), Carolina Power and Light Company, now doing business as Progress Energy Carolinas, Inc. (the licensee), requested changes to the technical specifications (TSs) for the H. B. Robinson Steam Electric Plant, Unit No.2 (HBRSEP). The proposed amendment would revise HBRSEP TS Section 3.6.8, "Isolation Valve Seal Water (IVSW) System."

The proposed change revises Surveillance Requirements (SRs) 3.6.8.2 and 3.6.8.6 related to IVSW tank volume and header flow rates. Specifically, the change would clarify the wording of SR 3.6.8.2, and revise SR 3.6.8.6 to provide a total flow rate limit from all four headers in place of the individual header limits.

SR 3.6.8.2 currently states, "Verify the IVSW tank volume is ?: [greater than or equal to]

85 gallons." The intent of this SR is to ensure that the water volume in the tank is greater than or equal to 85 gallons; however, the word water is missing from the statement.

SR 3.6.8.6 provides limits on the IVSW seal header flow rates, which thereby measures the leakage rate from the system boundary. The intent of SR 3.6.8.6, in combination with SR 3.6.8.2, is to ensure that there is sufficient water volume available in the IVSW system to compensate for leakage, such that makeup to the system is not required for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SR 3.6.8.6 currently provides individual flow rate limits for each of the four system headers.

Notice of consideration of this amendment and an opportunity to request a hearing was published in the Federal Register on January 15, 2008 (73 FR 2548).

Enclosure

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2.0 REGULATORY EVALUATION

Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The TSs ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the pUblic. The regulatory requirements of the U.S. Nuclear Regulatory Commission (NRC, the Commission) related to the content of the TSs are contained in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36, which requires that the TSs include items in the following categories:

(1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation (LCOs);

(3) surveillance requirements; (4) design features; and (5) administrative controls.

As stated in 10 CFR 50.36(d)(3), SRs "are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met."

The proposed change relates to SRs that ensure there is sufficient water volume available in the IVSW system to compensate for leakage, such that makeup to the system is not required for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Reliable operability of the IVSW System, as verified by the SRs, ensures that the system is available to limit the fission product release from the containment under accident conditions.

The relevant regulatory requirements for acceptance of the proposed change are contained in the HBRSEP Updated Final Safety Analysis Report (UFSAR) as it relates to containment design integrity and containment heat removal, and in 10 CFR Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors."

General Design Criteria (GDC) 10, 49, and 52, as discussed in Chapter 3.1, "Conformance with General Design Criteria," of HBRSEP's UFSAR relate to containment design integrity and containment heat removal. The GDC in existence at the time HBRSEP was licensed for operation in july 1970 were contained in Proposed AppendiX A to 10 CFR 50, "General Design Criteria for Nuclear Power Plants," published in the Federal Register on July 11, 1967.

AppendiX A to 10 CFR 50, which became effective in 1971 and as subsequently amended, is somewhat different from the proposed 1967 criteria. Discussions in this safety evaluation refer to the proposed GDC as contained in HBRSEP's UFSAR. The applicable proposed GDCs are summarized below:

• Proposed GDC 10, in part, refers to design of the containment to withstand a large reactor coolant system pipe break without loss of integrity, and, together with other engineered safety features, retain its functional capability.

• Proposed GDC 49, in part, refers to design of the containment structure to limit the leakage of radioactive materials from openings, penetrations, and any necessary heat removal systems, so as not to result in undue risk to the health and safety of the public.

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• Proposed GDC 52, in part, refers to the design of active heat removal systems such that they prevent exceeding containment design pressure under accident conditions and shall perform their required function assuming a single failure of an active component.

The NRC staff reviewed the licensee's license amendment request to verify that the above regulatory requirements, as well as the licensing basis criteria stated in the UFSAR, continue to be met with the proposed change.

3.0 TECHNICAL EVALUATION

3.1 Description of Affected System A description of the IVSW System is provided in the Bases documents for TS Section 3.6.8, as well as the HBRSEP UFSAR. A summary of the IVSW System, as provided in these licensee documents, follows:

The IVSW assures the effectiveness of certain containment isolation valves during any condition which requires containment isolation, by providing a water seal at the valves. These valves are located in lines that are connected to the Reactor Coolant System (RCS), or that could be exposed to the containment atmosphere in the event of a loss of coolant accident (LOCA). The system provides a reliable means for injecting seal water between the seats and stem packing of the globe and double disc types of isolation valves, and into the piping between closed diaphragm type isolation valves.

The system provides assurance that, should an accident occur, the containment leak rate is no greater than that assumed in the accident analyses by providing seal water at a pressure greater than or equal to 1.1 times the calculated peak containment internal pressure related to the design basis accident. The system is designed to maintain this seal for at least 30 days. The possibility of leakage from the containment or RCS past the first isolation point is thereby prevented by assuring that if leakage does exist, it will be from the IVSW System into containment.

The IVSW System operates to limit potential fission product releases from the containment. Although no credit was taken for operation of this system in the calculation of offsite accident doses, it does provide assurance that, should an accident occur, the containment leak rate is lower than that assumed in the accident analysis.

The system includes one 175 gallon seal water tank capable of supplying the total requirements of the system. The IVSW tank's minimum required volume of 85 gallons is maintained, with the normal supply of makeup water to the IVSW tank being the Primary Water System. In the event Primary Water is not available, emergency makeup can be supplied from the Service Water System.

-4 The tank is pressurized with a nitrogen blanket slJpplied from two independent sources. Primary supply is from the plant nitrogen supply header through a pressure regulating control valve. Automatic backup supply is provided from two high pressure nitrogen bottles through separate high and low pressure regulating valves.

The system is normally in a static condition with the seal water injection tank filled and pressurized. Indication of IVSW tank level and pressure, along with corresponding low level and low pressure alarms, are provided in the Control Room.

The IVSW tank supplies pressurized water to four distribution headers. Header "A" is the manual header, meaning an isolation valve on this header must be pressurized by opening a manual valve supplying the individual isolation valve.

Header "A" serves lines that are normally filled with fluid following a LOCA, and lines that must remain in service for a period of time following the accident.

Headers "B," "C," and "0" are automatic headers that are pressurized through one or both of two redundant, fail-open, air-operated valves in parallel. A loss of power will cause the automatic valves to open, since automatic initiation is a deenergized signal to vent air from the valve operators. System operation is initiated by a Phase A containment isolation signal which accompanies any Safety Injection (SI) signal.

3.2 Proposed TS Changes The proposed change adds the word water to SR 3.6.8.2. This is a clarification change and is consistent with the intent of the SR and the manner in which the surveillance is performed, and is therefore acceptable to the NRC staff. The proposed change also revises SR 3.6.8.6 to provide a total 'flow rate limit from all four headers in place of the current individual header limits.

The proposed total flow rate limit of less than or equal to 124 cubic centimeter (cc) per minute is equal to the sum of the existing header limits currently specified in SR 3.6.8.6.

3.3 Background for Proposed Change to SR 3.6.8.6 The current flow rate values for each header were determined based on a design analysis assumption that each valve supplied by that header of the IVSW System leaks at a rate of 50 cc per hour per inch of nominal pipe diameter. The assumed leakage from each valve connected to a header was summed to obtain the acceptance limit for that individual header flow. Based on the number and size of the valves that were connected to the headers at the time SR 3.6.8.6 was last revised. this resulted in the values currently shown in SR 3.6.8.6, and a total flow from all four headers of 124 cc per minute (7440 cc per hour).

The licensee stated that during the fall 2008 refueling outage, it is planned that three Safety Injection valves (SI-870A, SI-870B, and SI-869) will be removed from IVSW Header "A" leakage testing, as it has been determined that these three valves should not be subject to 10 CFR 50 Appendix J leakage criteria. The valves will be removed from the testing program Via the 10 CFR 50.59, "Changes, Tests, and Experiments," process, based on American National

-5 Standards Institute (ANSI) Standard 56.8, Section 3.3.1. This section provides that Type B or Type C leakage testing is not required for containment boundaries that do not constitute potential primary containment atmospheric pathways during and following a design basis accident (DBA). The licensee concludes in the 50.59 evaluation that the subject flow paths are in service and pressurized during/following a DBA, and are therefore eligible to be removed from the requirements of the 10 CFR 50 Appendix J leakage criteria.

This change will result in the need to lower the currently allowed Header "A" flow rate value of 52 cc per minute in SR 3.6.8.6 in order to maintain consistency with the 50 cc per hour per inch design analysis assumption. According to HBRSEP, such a reduction would result in an unnecessary limitation, as 52 cc per minute still ensures compliance with the design requirement for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> water supply.

The design analysis also assumed there was a failure of the largest valve to close completely, resulting in an additional leakage of 1000 cc per hour per inch of nominal pipe diameter for that valve. The largest valve was determined to be a 6-inch valve, which resulted in an additional assumed leakage flow of 6000 cc per hour. Therefore, the total combined assumed leakage is 13,440 cc per hour, or 3.5 gallons per hour. These design assumptions resulted in the need for an inventory of 85 gallons of water in the tank to provide for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of leakage before makeup is required. Makeup can then be provided 'from either of two sources (Primary Water or Service Water) to ensure the system will function as designed for 30 days.

For demonstration that the 24-hour water supply is available, it becomes unimportant whether each individual header meets the current specified flow rate; it is the total flow from all four headers that is the determining factor. Therefore, the proposed change provides one combined seal header flow rate limit in order to assure that an adequate water supply is available. This represents a change from the original methodology used to determine the necessary tank volume. Individual header limits based on an assumed leak rate per valve could become inconsistent with this revised methodology if valves are added or removed from the header supply. In addition, the current surveillance procedures do not contain an acceptance criterion for individual valve leakage, only for the total header leakage limit based on the TSs.

The proposed license amendment therefore revises the analytical basis for SR 3.6.8.6. The revised methodology for establishing the header flow rate acceptance criterion for SR 3.6.8.6 would be based on the assessment that a total leak rate from all four headers of 124 cc per minute continues to ensure a 24-hour water supply based on the required tank volume of 85 gallons (the proposed change still assumes the failure of one valve to close, resulting in an additional 6000 cc per hour flow). This would replace the basis methodology that assumed a specific leakage rate per valve, an assumption that is not required to be verified or met on a per valve basis. The 50 cc per hour per inch leakage was a reasonable initial design input assumption used to estimate total IVSW leakage and hence establish a reasonable required tank volume. Under the new proposed methodology, surveillance flexibility would be gained, as individual header allowed leakage would increase. However, the design requirement - that is the underlying basis of the SR - to ensure a 24-hour water supply would not be impacted as a result of the proposed change.

-6 3.4 Technical Evaluation The DBA that results in a release of radioactive material within containment is a LOCA. The analyses for the LOCA assume the isolation of containment is completed and leakage from containment is at a rate equivalent to the design leakage rate. As part of the containment boundary, containment isolation valves function to support the leak tightness of containment.

By maintaining this barrier, offsite dose calculations will be less than the 10 CFR 100 guidelines during a DBA. The IVSW System thus functions as a post-accident dose mitigating system.

The proposed change to SR 3.6.8.6 revises the methodology used to establish the system flow limits, but maintains the same total flow limitation and consistency with the system design requirements. There is no impact on the system's dose mitigation capability.

The IVSW System actuates on a containment isolation signal and functions to assure that the actual leakage is no greater than the design value. The system ensures the effectiveness of certain isolation valves to limit containment leakage by pressurizing the affected containment penetration flow paths at a pressure greater than the maximum calculated peak containment pressure during a LOCA. The IVSW System is designed to maintain this seal for at least 30 days. A single failure analysis shows that the failure of any active component will not prevent fUlfilling the design function of the system. By meeting these requirements, the IVSW System is considered a qualified seal system in accordance with 10 CFR 50, Appendix J.

The use of the IVSW System during a LOCA, while not considered in the analysis of the consequences of the accident, provides an additional means of ensuring that leakage is minimized. The operation of the system can be monitored after the accident, and provisions are included for manually replenishing the seal water if required. These measures further minimize the potential leakage to the environment. No detrimental effect on any other system will occur should the seal water system fail to operate.

As stated by the licensee, the acceptable leakage criteria are based on a total allowable leakage value calculation for each of the 4 IVSW System headers, using the IVSW System design leakage. The proposed change would modify SR 3.6.8.6 to allow verification of the total combined header flow, rather than the current individual header limits. The purpose of the .

IVSW leakage specification (SR 3.6.8.6) is to ensure that the 85 gallon water inventory in the IVSW tank (as required by SR 3.6.8.2) provides sufficient inventory for a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of leakage from the IVSW system. This provides sufficient time for the operators to make-up water to the IVSW tank. To demonstrate that capability, only the total leakage from the system needs to meet the acceptance criteria of 124 cc per min.

This change will not challenge the TS requirement for maintaining a 24-hour water supply based on the required tank volume of 85 gallons, as the maximum allowed header flow rate will not be increased, no matter how many valves or what proportion of the total leakage is associated with each header. To meet the overall intent of the TS, it becomes unimportant whether the leakage is all from one header, equally split between the four headers, or split between the headers based on the number of valves supported by each header. The establishment of leakage limits on a per-header or per-valve basis would impose unnecessary limitations in meeting the intent of the specification. The calculations performed that estimated the leakage from each header based on the number and sizes of the valves supported by the header, provided validation that

-7 the per-header leakage limits incorporated into the Technical Specifications were reasonable.

However, the calculations were unnecessary to meet the purpose of the specification.

To ensure that none of the IVSW headers becomes blocked, each penetration served by IVSW is tested individually, and then all leakages combined to determine the total header leakage.

During each penetration test, the leakage rate for the test boundary is determined, and then one valve within the test boundary is opened to verify that the flow meter goes off scale. This verifies that there is a clear flow path to the penetration. In addition, because of the nature of the IVSW system, a situation in which all of the allowed leakage was the result of a single penetration (the failure of a valve to close) would not challenge the function of the system as the maintained pressure assures that if leakage does exist, it is from the IVSW into containment.

Incorporating the proposed changes will continue to maintain the overall operability of the IVSW System and continue to ensure an adequate volume of water at sufficient pressure is available to provide the motive force necessary to move the system fluid into the applicable containment penetrations during accident conditions.

3.5 Conclusion Based on the evaluation described above, the NRC staff concludes that the proposed change, which allows SR 3.6.8.6 to verify one combined seal header flow rate limit in order to assure the availability of an adequate water supply for the IVSW System, (1) accounts for the impact of the proposed change on applicable accident analyses, design basis containment leakage limits, and other TS requirements, (2) maintains the IVSW System design bases and does not result in the possibility of an increased containment leakage rate during accident conditions, and (3) continues to meet the intent of all applicable proposed GDCs, and is, therefore, acceptable.

Based on this, the NRC staff also concludes that the proposed change continues to meet the TS requirements of 10 CFR 50.36, and the proposed amendment is, therefore, acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on the finding as published in the Federal Register on January 15, 2008 (73 FR 2548).

Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with issuance of the amendment.

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6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Marlayna Vaaler Date: October 3, 2008