ML030450307

License Amendment Modification of the Technical Specification Surveillance Requirements for the Safety/Relief Valves

ML030450307
Person / Time
Site:	River Bend
Issue date:	02/13/2003
From:	Michael Webb NRC/NRR/DLPM/LPD4
To:	Hinnenkamp P Entergy Operations
Webb M, NRR/DLPM, 415-1347
References
TAC MB5090
Download: ML030450307 (15)

Text

February 13, 2003 Mr. Paul D. Hinnenkamp Vice President - Operations Entergy Operations, Inc.

River Bend Station P. O. Box 220 St. Francisville, LA 70775

SUBJECT:

RIVER BEND STATION, UNIT 1 - ISSUANCE OF AMENDMENT RE: MODIFICATION OF THE TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS FOR THE SAFETY/RELIEF VALVES (TAC NO. MB5090)

Dear Mr. Hinnenkamp:

The Commission has issued the enclosed Amendment No. 130 to Facility Operating License No. NPF-47 for the River Bend Station, Unit 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated May 14, 2002, as supplemented by letter dated December 6, 2002.

The amendment revises the TS safety function lift setpoint tolerances for the Safety/Relief valves (S/RVs). The changes also allow surveillance of the relief mode of operation of the S/RVs to be performed without physically lifting the disk of a valve off the seat at power.

A copy of our related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commissions next biweekly Federal Register notice.

Sincerely,

/RA/

Michael Webb, Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-458

Enclosures:

1. Amendment No. 130 to NPF-47

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

February 13, 2003 Mr. Paul D. Hinnenkamp Vice President - Operations Entergy Operations, Inc.

River Bend Station P. O. Box 220 St. Francisville, LA 70775

SUBJECT:

RIVER BEND STATION, UNIT 1 - ISSUANCE OF AMENDMENT RE: MODIFICATION OF THE TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS FOR THE SAFETY/RELIEF VALVES (TAC NO. MB5090)

Dear Mr. Hinnenkamp:

The Commission has issued the enclosed Amendment No. 130 to Facility Operating License No. NPF-47 for the River Bend Station, Unit 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated May 14, 2002, as supplemented by letter dated December 6, 2002.

The amendment revises the TS safety function lift setpoint tolerances for the Safety/Relief valves (S/RVs). The changes also allow surveillance of the relief mode of operation of the S/RVs to be performed without physically lifting the disk of a valve off the seat at power.

A copy of our related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commissions next biweekly Federal Register notice.

Sincerely,

/RA/

Michael Webb, Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-458

Enclosures:

1. Amendment No. 130 to NPF-47

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

PUBLIC RidsNrrDlpmPdiv (WRuland)

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Accession No.: ML030450307

• No major changes to SE.

• • No legal objection OFFICE PDIV-1/PM PDIV-1/LA DE/EMEB/SC OGC**

PDIV-1/SC NAME MWebb DJohnson DTerao RHoefling RGramm DATE 1/31/03 1/24/03 12/20/02 2/6/03 2/10/03 OFFICIAL RECORD COPY

March 2002 River Bend Station cc:

Winston & Strawn 1400 L Street, N.W.

Washington, DC 20005-3502 Manager - Licensing Entergy Operations, Inc.

River Bend Station P. O. Box 220 St. Francisville, LA 70775 Senior Resident Inspector P. O. Box 1050 St. Francisville, LA 70775 President of West Feliciana Police Jury P. O. Box 1921 St. Francisville, LA 70775 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 Ms. H. Anne Plettinger 3456 Villa Rose Drive Baton Rouge, LA 70806 Mr. Michael E. Henry, Administrator and State Liaison Officer Department of Environmental Quality P. O. Box 82135 Baton Rouge, LA 70884-2135 Wise, Carter, Child & Caraway P. O. Box 651 Jackson, MS 39205 Executive Vice President and Chief Operating Officer Entergy Operations, Inc.

P. O. Box 31995 Jackson, MS 39286-1995 General Manager - Plant Operations Entergy Operations, Inc.

River Bend Station P. O. Box 220 St. Francisville, LA 70775 Director - Nuclear Safety Entergy Operations, Inc.

River Bend Station P. O. Box 220 St. Francisville, LA 70775 Vice President - Operations Support Entergy Operations, Inc.

P. O. Box 31995 Jackson, MS 39286-1995 Attorney General State of Louisiana P. O. Box 94095 Baton Rouge, LA 70804-9095 Brian Almon Public Utility Commission William B. Travis Building P.O. Box 13326 1701 North Congress Avenue Austin, Texas 78701-3326

ENTERGY GULF STATES, INC. **

AND ENTERGY OPERATIONS, INC.

DOCKET NO. 50-458 RIVER BEND STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 130 License No. NPF-47 1.

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

A.

The application for amendment by Entergy Gulf States, Inc.* (the licensee) dated May 14, 2002, as supplemented by letter dated December 6, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and

• Entergy Operations, Inc. is authorized to act as agent for Entergy Gulf States, Inc., and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.

• • Entergy Gulf States, Inc., has merged with a wholly owned subsidiary of Entergy Corporation.

Entergy Gulf States, Inc., was the surviving company in the merger.

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions 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 2.C.(2) of Facility Operating License No. NPF-47 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 130 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Robert A. Gramm, Chief, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: February 13, 2003

ATTACHMENT TO LICENSE AMENDMENT NO. 130 FACILITY OPERATING LICENSE NO. NPF-47 DOCKET NO. 50-458 Replace the following pages of the Appendix A 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 Insert 3.4-10 3.4-10 3.4-11 3.4-11 3.5-5 3.5-5 3.6-24 3.6-24

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 130 TO FACILITY OPERATING LICENSE NO. NPF-47 ENTERGY OPERATIONS, INC.

RIVER BEND STATION, UNIT 1 DOCKET NO. 50-458

1.0 INTRODUCTION

By application dated May 14, 2002, as supplemented by letter dated December 6, 2002, Entergy Operations, Inc. (the licensee), requested changes to the Technical Specifications (TSs) for the River Bend Station, Unit 1 (RBS). The supplement dated December 6, 2002, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on June 25, 2002 (67 FR 42822).

The proposed changes would modify the safety/relief valve (S/RV) TS setpoint tolerance and manual actuation surveillance requirements (SRs). Specifically, the proposed changes would revise TS SR 3.4.4.1 for the S/RV setpoint tolerance; SRs 3.4.4.3, 3.5.1.7, and 3.6.1.6.1 for the S/RV manual actuation; and the associated Bases for these requirements.

The proposed amendment would change the safety function lift setpoint tolerances for the S/RVs that are listed in SR 3.4.4.1. The tolerances for these valves would be changed from

+/-3% to +3%, -5% of the safety lift setpoint. This change only applies to the as-found tolerance and not to the as-left tolerance, which will remain unchanged at +/-1% of the safety lift setpoint.

The proposed amendment would also change TS SRs 3.4.4.3 for S/RVs, 3.5.1.7 for Automatic Depressurization System (ADS) valves, and 3.6.1.6.1 for Low-Low Set (LLS) valves. The proposed changes would verify that the relief-mode actuators stroke when the valves are manually actuated, without physically lifting the disks off their seats at power. In addition, the proposed amendment would change the frequency of these SRs from 18 months to in accordance with the inservice testing program, consistent with SR 3.4.4.1.

Each S/RV is a Crosby, 8 X 10, direct-acting, spring loaded, safety valve with attached pneumatic actuator for relief-mode operation. The valve is direct acting; the single safety valve stem penetrates the valve bonnet and attaches directly to the valve disk. The S/RV bonnet is directly mounted to the top of the S/RV body. The S/RV body and cap assembly provide mounting facilities for the functional parts of the relief components. RBS has a total of 16 S/RVs installed on the four main steam lines (MSLs). All S/RVs are of the same design and can perform their intended function through two modes of operation.

The safety mode of operation consists of direct action of the reactor vessel steam pressure against a single spring-loaded disk. This disk will lift off of its seat when the reactor vessel pressure exceeds the spring force, thereby allowing vessel steam to flow directly through the seat-to-disk opening to the discharge piping and suppression pool. The safety function set pressure of each S/RV is established by changing the compressed spring force.

The relief mode of operation is accomplished when an automatic or manual control circuit signal provides electric power to the actuator solenoids. The actuator solenoids reposition, admitting control air to the pneumatic actuator cylinder. The pneumatic actuator piston strokes vertically, rotating an attached lever and dog/load plate assembly, which contacts the spindle nut threaded onto the S/RV stem. The S/RV stem, which is directly attached to the disk, strokes vertically and lifts the disk off of the seat. The relief-mode of operation can actuate each S/RV regardless of reactor vessel pressure. The pneumatic operator and linkage are designed so that a malfunction of this linkage will not inhibit the safety-mode of operation. The relief-mode of operation requires three major elements to perform its intended function: actuation logic and controls, electrical power, and control air.

Seven of the S/RVs use the relief mode to perform the ADS function. Also, five S/RVs, one of which is also an ADS S/RV, use the relief mode to perform the LLS function.

2.0 REGULATORY EVALUATION

General Design Criterion 15 of Appendix A to Part 50 to Title 10 of the Code of Federal Regulations (10 CFR Part 50) requires that The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences. Further, the NRC's regulatory requirements related to the content of TSs are set forth in 10 CFR 50.36.

Specifically, 10 CFR 50.36(c)(2)(ii) sets forth four criteria to be used in determining whether a limiting condition for operation (LCO) is required to be included in TSs. These criteria require an LCO for: (1) installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant system pressure boundary; (2) initial plant conditions that are assumed in a design-basis transient and accident analysis; (3) components or systems that are used for mitigating consequences of the design-basis transient and accident; and (4) components or systems which probabilistic risk assessment has shown to be significant to public health and safety. The Standard TSs (STSs) were developed based on the criteria in 10 CFR 50.36(c)(2)(ii). Existing LCOs and related SRs included as TS requirements which satisfy any of the criteria specified in 10 CFR 50.36(c)(2)(ii) must be retained in the TSs. The NRC encourages the licensee to upgrade their TSs consistent with those criteria and conforming, to the extent practical and consistent with the licensing basis for the plant, to the current STS.

The S/RVs are part of the primary success path and are assumed in the Updated Safety Analysis Report in the accident and safety analyses to mitigate the effects of the licensing-basis accidents. In accordance with Criterion 3 of 10 CFR 50.36(c)(2)(ii) discussed above, TS LCOs are required for both the S/RV safety and relief modes of operation. Since the setpoint tolerance and SRs for S/RVs are proposed to be changed, the licensee is required to provide acceptable analyses to support the adequacy of the TS changes.

The staff finds that the licensee in Section 5 of its May 14, 2002, submittal identified the applicable regulatory requirements. The regulatory requirements for which the staff based its acceptance include the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) and the ASME Operations and Maintenance (OM) Code.

There have been several other somewhat similar TS changes approved for other facilities regarding both setpoint tolerance and stroke testing. However, the -5% as-found lower setpoint tolerance has not been approved at any facility in prior changes. The staff has determined that the proposed changes are acceptable, as discussed below.

3.0 TECHNICAL EVALUATION

The staff has reviewed the licensees regulatory and technical analyses in support of its proposed license amendment which are described in Sections 3 and 4 of the licensees May 14, 2002, submittal. The detailed evaluation below will support the conclusion 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 Commissions 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.

Currently, all 16 S/RVs installed on the MSLs are removed during each refueling outage, bench tested for as-found setpoint, and replaced with valves certified to have zero seat-to-disk leakage and to have as-left setpoints within +/-1% of the setpoints specified in the TS. The licensee plans to initiate a sampling program for the S/RVs in accordance with the ASME Code requirements in lieu of replacing all 16 valves each refueling outage. The sampling program would require removal of a sample and testing for as-found conditions as required by the Code.

The licensee is requesting this change to reduce plant personnel exposure and reduce costs associated with unnecessary valve refurbishment, additional testing, and increased outage durations. The as-found setpoint tolerances would be changed only for the lower tolerance setting of the safety lift function. The relief mode of S/RV operation is dependent upon external power sources, and the setpoint for the relief mode would remain unchanged.

The licensee states that the current S/RV safety lift setpoint tolerances were established during the power uprate process and included in the associated license amendments (Amendment Numbers 109 and 114). The power uprate required the as-left safety valve function settings to be within +/-1% and allowed an as-found tolerance of +/-3% before additional testing would be required. The change was made pursuant to Licensing Topical Report (LTR) NEDC-31753P "BWROG [Boiling Water Reactor Owners Group] In-Service Pressure Relief Technical Specification Revision Licensing Topical Report" and the associated NRC Safety Evaluation Report (SER).

The licensees evaluation of the S/RV setpoint test results indicate that the majority of the S/RVs experience minor setpoint drift in the negative direction over time. Based on a review of the 33 previous test results, the average drift over one operating cycle equated to approximately -1.6% from the previous as-left test results. The highest drift experienced was at

-4.65% and was attributed to seat leakage. Excluding this valve, the average drift from the previous as-left test results was -1.4%. In the past, RBS has replaced all 16 S/RVs each refueling outage, and as such, there is no history for the plant valves to determine if drift continues over longer periods of time. However, test results provided by another boiling water reactor plant with similar Crosby S/RVs indicate that the amount of drift over multiple cycles was not significantly different than that experienced over one cycle. For 231 tests from single and multiple cycles, the average drift was -0.8%. A drift of -0.3% was the average for tests conducted after 30 months or more from the previous test. Therefore, the licensee states that excessive drift over multiple operating cycles is not anticipated for the RBS S/RVs. The licensee states that more significant drift occurs when S/RVs are leaking by their seats, but that leaking S/RVs will typically not be permitted to remain in place beyond the next refueling outage. Further, the licensee stated that a valve seat modification, which was made in 1995, significantly improved seat leakage.

The licensee performed an evaluation to demonstrate that changing the as-found tolerance as proposed does not compromise plant safety or result in an increase in the number of unnecessary S/RV actuations. The licensee determined that assuming an as-found setpoint of

-5%, the lowest set S/RV (1195 psig) would lift at 1135 psig, which would provide an operating margin of 80 psi above the nominal reactor operating pressure of 1055 psig. The licensee determined that this provides sufficient margin to prevent unwanted actuation of the S/RVs.

The licensee also evaluated the ADS function, the scram response, and the LLS logic functions and determined that there is no significant impact on these functions as discussed below.

The ADS valves open automatically as part of the emergency core cooling system for events involving small breaks. Since the purpose of the ADS is to reduce reactor pressure, the licensee determined that the proposed change to the as-found safety setpoint tolerance would have no impact on the ADS operation.

The reactor vessel dome pressure high scram setpoint is 1094.7 psig. The high pressure scram terminates a pressurization transient in conjunction with high neutron flux scram. Drift of the S/RV safety function to the proposed as-found tolerance of -5% remains above this limit.

The licensee determined that this provides assurance that there is a low probability of inadvertent scram on high reactor pressure.

In order to assure that no more than one relief valve reopens following a reactor isolation event, one ADS valve and four non-ADS valves are provided with lower opening and closing setpoints.

These setpoints override the normal relief setpoints following the initial opening of the relief valves and act to hold these valves open longer, thus preventing more than a single valve from subsequentially reopening. This system logic is referred to as the LLS logic and functions to ensure that the containment design basis of one S/RV opening on subsequent actuation is met.

The licensee determined that the overlap between the as-found safety mode setpoint tolerance with some of the LLS instrument setpoints will not significantly impact response of the SRVs during transient events.

3.1 Evaluation of Proposed S/RV Setpoint Tolerance The staff has reviewed the licensees proposed TS change to increase the S/RV as-found lower setpoint tolerance to -5% and finds that the change would reduce plant personnel radiation exposure and reduce costs associated with valve refurbishment, additional testing, and increased outage durations. This change would also reduce the potentially undesirable effects of opening the S/RVs at power.

The licensees proposed sample testing of the S/RVs during refueling outages meets the maximum five-year ASME Code test period, which has been determined to be acceptable for this type of valve. The licensee has demonstrated with test data that the S/RVs have a good setpoint performance history with a general trend for the valve setpoints to drift slightly low.

Further, a review of data from another facility having the same valve model indicates that with longer service times, the setpoint is not expected to further degrade significantly.

The NRC SER for the above described LTR, NEDC-31753P, did not approve an as-found tolerance outside the +/-3% range. The licensees proposed -5% lower as-found setpoint tolerance is a precedent which has not been applied at other facilities. One of the issues of concern to the staff is that of possible effects of increased valve seat leakage due to lower simmer margins. That is, if the expanded lower as-found tolerance were to result in less testing and maintenance applied to the valves, this might result in an increased tendency for the S/RVs to drift to lower setpoints and leak. With increased leakage, there is an increased risk of a damaging water hammer occurrence following a loss-of-offsite power, as described in NRC Information Notice 87-10, "Potential for Water Hammer During Restart of Residual Heat Removal Pumps," and Supplement 1 thereto. However, since the licensee will continue to apply the +/-1% as-left tolerance for resetting the S/RVs, and since the valves are not expected to drift significantly greater amounts for longer operational service, the -5% as-found tolerance is not expected to result in increased leakage.

The staff has reviewed the effects of the change in S/RV setpoint tolerance from -3% to -5%

and finds that it is not significant to S/RV operation during overpressure transients. The nominal setpoints for the relief and safety modes are not changed. The nominal operating pressure of the reactor pressure vessel at power is 1055 psig. For the lowest as-found setpoint of -5%, the lowest nominal set S/RV (1195 psig) would lift at 1135 psig and provide an operating margin of 80 psi between the nominal reactor operating pressure and the S/RV lift pressure. This provides sufficient margin and is sufficient to prevent unwanted actuation of the S/RVs postulated to occur during pressurization transients. Further, since the high setpoint tolerance is unchanged, the capability of the S/RVs to ensure ASME overpressure protection is maintained.

The staff also finds that there is not a significant impact on the ADS function, the scram response, and the LLS logic function for the proposed change to the lower S/RV setpoint tolerance. Lower S/RV setpoints would not affect the ADS function, and the lowest set S/RV would remain above the scram setpoint. For the LLS logic function, the change to the lower S/RV setpoint tolerance could result in an overlap between the lowest S/RV setpoints and some of the higher LLS settings, considering instrument drift, but not for the lowest LLS setting. This provides reasonable assurance that only one of the S/RVs will open for subsequent LLS logic S/RV operation, in accordance with the plant licensing basis.

Therefore, the staff finds that the proposed TS SR with the +3%, -5% as-found setpoint tolerance to be acceptable.

3.2 Licensees Basis for Proposed S/RV Stroke Testing Currently, all S/RVs are removed during each refueling outage, tested on a test fixture, and certified for safety-mode set pressure, actuator stroke and leakage, and seat leak tightness. A certified S/RV is installed at the location of each removed S/RV, and the electrical and pneumatic controls are connected, after which a surveillance is performed to verify proper operation of the solenoid valves and a leakage test is performed to verify proper connection of the control air. The S/RV stroke test surveillance is performed during reactor startup while the plant is diverting steam flow to the main condenser via the turbine bypass/control valves. Each S/RV hand-switch is taken to the "OPEN" position, giving the S/RV an open signal in the relief mode. Valve operation is verified by indication that the S/RV acoustic monitor red status light is on, a change occurs for the main steam control valves or bypass valves indicated position, or there is indication of decreasing MSL flow. This test physically strokes the S/RV using the relief mode of operation.

The licensee states that, following the currently required stroke testing surveillance, a potential exists for the valves to begin leaking due to the actuations. Each leaking S/RV directs reactor vessel steam flow to the suppression pool where the steam is condensed and mixed with the suppression pool contents. The condensation transfers heat to the suppression pool, and as a consequence, one of the Residual Heat Removal (RHR) systems must be periodically re-aligned from its normal safety mode of vessel injection to provide suppression pool cooling.

This operation of the RHR system generates additional wear and cycling of system components, and the re-alignment places the system in a configuration that differs from its intended normal standby mode of operation. The transfer of heat to the suppression pool is also a source of thermal heat loss from the power generation steam cycle, thereby reducing electrical generating efficiency. The reactor vessel steam that leaks through each S/RV provides an additional source of radioactive nuclides, which become a potential source for personnel contamination. The containment at RBS is accessible during power operations, and containment contamination would increase if an S/RV has a leaking seat. This contamination is undesirable from two aspects: personnel must spend additional time to exit the radiologically controlled area, and staff efforts must be utilized to reduce or remove this contamination.

The licensee states that the proposed TS change would provide verification of proper control connections by requiring the pneumatic and electrical controls to cycle the actuator on each S/RV following installation, without stroking the S/RV itself. In addition, the test population of S/RVs removed each outage for safety setpoint testing would also be tested in the relief mode during bench testing. The licensee states that this testing would provide reasonable assurance that the installed S/RVs will function properly in the relief mode. The licensee also states that the remaining installed S/RVs would continue to be tested for proper system function. As presently required by the TS and administrative procedures, proper operation of the solenoid valve will be demonstrated by providing an open signal to each S/RV, with a check to verify that each solenoid valve repositions. The licensee states that verification of proper solenoid valve operation, in addition to the proper relief mode operation of the test population, provides reasonable assurance that the S/RV would perform as expected when control air pressure is applied to the actuator assembly.

The licensee states that after the test population of S/RVs is replaced and S/RV controls have been connected, the clevis pin of each newly-installed S/RV would be removed from the lever arm and clevis, and the actuator assembly would be rotated away from the lever arm to allow an uncoupled actuation of the relief-mode actuator. Control air pressure to each actuator would be reduced from normal system pressure to prevent damaging the pneumatic relief-mode actuator. The actuator would be remotely operated from the control room, as required by current test methods, and visual verification would be performed for proper actuator response.

After proper actuator operation had been verified, the actuator would be re-coupled to the lever arm by re-installing the clevis pin and retaining rings. The licensee states that this test would demonstrate that no damage had occurred to the relief-mode actuator during reinstallation.

Further, the licensee would state that the S/RVs had been both removed and installed without incident.

The licensee states that the plant TS bases indicate that one purpose of mechanically stroking the S/RV is to verify that a change in measured steam flow indicates that no blockage exists in the S/RV discharge line. The licensee states that Foreign Material Exclusion (FME) controls are placed on all system openings when each S/RV is removed, and that these controls, as well as the horizontal orientation of the S/RV discharge lines, provide reasonable assurance that no obstruction will be admitted into the S/RV discharge tailpipe. The licensee states that there has not been any surveillance failure related to line blockage.

3.3 Evaluation of Proposed S/RV Stroke Testing The staff has reviewed the licensees proposed TS change to the S/RV stroke testing surveillance and finds that the current requirements can result in additional seat leakage of the S/RVs during power operation. Such leakage would be directed to the primary containment suppression chamber, causing a need to increase cooling to the suppression pool water which could potentially increase radiation exposure to plant personnel. The proposed testing provides for actual stroking of the S/RVs after performing the ASME Code setpoint testing on a sample of valves, combined with stroking only the S/RV actuators after the S/RVs have been installed.

A significant difference between the current TS testing and that proposed is that the proposed sample stroke testing of the S/RV population each outage, when the S/RVs are setpoint tested, is less than the current testing of all S/RVs each refueling. However, the 1987 Edition of the ASME OM Code, which is currently applicable to the licensee, requires only that a sample of S/RVs be setpoint tested in any test period. Further, the ASME OM Code requires the stroking of S/RV actuators only when setpoint tests, or maintenance or repair activities, are performed.

Therefore, the licensees proposed testing frequency meets the ASME OM Code-required frequency, which the staff has agreed is adequate for testing valves of this type.

A second difference between the current TS testing and that proposed is that the proposed testing of air actuators does not actually stroke the S/RVs themselves. However, the licensee has demonstrated that the bench testing, in combination with other tests and controls, provides adequate assurance of the capability of all of the necessary S/RV components to function properly in the relief mode of operation.

A third difference between the current TS testing and that proposed is that, since the S/RVs would no longer be stroke tested to discharge steam flow, the S/RV discharge lines would not be verified to be unblocked by steam flow. However, the licensee has demonstrated that the FME controls and the horizontal orientation of the lines provide reasonable assurance that the discharge lines would remain unblocked.

Therefore, the staff finds that the proposed TS SR for S/RV stroke testing is acceptable.

As described above, the licensee has proposed changes to the plant TS which would provide for an increased lower as-found tolerance and for testing of the S/RVs to demonstrate proper relief mode function, without the need for online stroking of the S/RVs. Based on the above evaluation, the staff finds that the licensee has adequately justified the proposed changes to the TS for RBS. Therefore, the proposed TS changes to SRs 3.4.4.1, 3.4.4.3, 3.5.1.7, and 3.6.1.6.1, and the associated Bases for these requirements, are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Louisiana State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 such finding (67 FR 42822 dated June 25, 2002). 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 the issuance of the amendment.

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 Commissions 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 Contributors: G. Hammer G. Thomas Date: February 13, 2003