ML20239A063

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SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Sequoyah Nuclear Plant,Units 1 & 2
ML20239A063
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 08/27/1998
From:
NRC (Affiliation Not Assigned)
To:
Shared Package
ML20239A062 List:
References
GL-95-07, GL-95-7, NUDOCS 9809080118
Download: ML20239A063 (4)


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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 1

LICENSEE RESPONSE TO GENERIC LETTER 95-07. " PRESSURE LOCKING AND THERMAL BINDING OF SAFETY-RELATED POWER-OPERATED GATE VALVES" I

SEQUOYAH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NUMBERS 50-327 AND 50-328

1.0 INTRODUCTION

l Pressure locking and thermal binding represent potential common-cause failure mechanisms that can render redundant safety systems incapable of performing their safety functions. The identification of susceptible valves and the determination of when the phenomena might occur requires a thorough knowledge of components, systems, and plant operations. Pressure locking occurs in flexible-wedge and double-disk gate valves when fluid becomes pressurized inside the valve bonnet and the actuator is not capable of overcoming the additional thrust requirements resulting from the differential pressure created across both valve disks by the pressurized fluid in the valve bonnet. Thermal binding is generally associated with a wedge gate valve that is closed while the system is hot and then is allowed to cool before an attempt is made to open the valve.

Pressure locking or thermal binding occurs as a result of the valve design characteristics (wedge and valve body configuration, flexibility, and material thermal coefficients) when the valve is subjected to specific pressures and temperatures during various modes of plant operation. Operating experience indicates that these situations were not always considered in many plants as part of the design basis for valves.

2.0 REGULATORY REQUIREMENTS 10 CFR Part 50 (Appendix A, General Design Criteria 1 and 4) and plant licensing safety analyses require or commit (or both) that licensees design and test safety-related components and systems to provide adequate assurance that those systems can perform their safety functions. Other individual criteria in Appendix A to 10 CFR Part 50 apply to specific systems.

In accordance with those regulations and licensing commitments, and under the additional provisions of 10 CFR Part 50 (Appendix B, Criterion XVI), licensees are expected to act to ensure that safetydelated power-operated gate valves susceptible to pressure locking or thermal binding are capable of performing their required safety functions.

On August 17,1995, the U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 95-07, " Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves," to request that licensees take certain actions to ensure that safety-related power-l operated gate valves that are susceptible to pressure locking or thermal binding are capable of performing their safety functions within the current licensing bases of the facility. GL 95-07 9809080118 980827 PDR ADOCK 05000327 G

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requested that each licensee, within 180 days of the date of issuance of the generic letter (1) evaluate the operational configurations of safety-related power-operated gate valves in its plant to identify valves that are susceptible to pressure locking or thermal binding, and (2) perform further analyses and take needed corrective actions (or justify longer schedules) to f

ensure that the susceptible valves, identified in (1) above, are capable of performing their intended safety functions under all modes of plant operation, including test configuration. In addition, GL 95-07 requested that licensees, within 180 days of the date of issuance of the generic letter, provide to the NRC a summary description of (1) the susceptibility evaluation used to determine that valves are or are not susceptible to pressure locking or thermal binding, (2) the results of the susceptibility evaluation, including a listing of the susceptible valves identified, and (3) the corrective actions, or other dispositioning, for the valves identified as susceptible to pressure locking or thermal binding. The NRC issued GL 95-07 as a

" compliance backfit" pursuant to 10 CFR 50.109(a)(4)(i) because modification may be necessary to bring facilities into compliance with the rules of the Commission referenced above.

In a letter dated February 13,1996, the Tennessee Valley Authority (TVA) submitted its 180-day response to GL 95-07 for Sequoyah Nuclear Plant, Units 1 and 2. In a letter dated

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March 15,1996, the licensee supplemented its 180-day response to GL 95-07. The NRC staff reviewed the licensee's submittals and requested additional information in a letter dated July 9,1996. In a letter dated August 6,1996, the licensee provided the additional information.

During the period of May 25 through July 5,1997, the NRC staff performed an inspection to review specific aspects of information summarized in the licensee's responses to GL 95-07.

This inspection is documented in NRC Inspection Report 50-327, 328/97-06. In a letter dated December 1,1997, the licensee further supplemented its 180-day response to GL 95-07.

3.0 STAFF EVALUATION 3.1 Scope of Licensee's Review GL 95-07 requested that licensees evaluate the operational configurations of safety-related power-operated gate valves in their plants to identify valves tnat are susceptible to pressure locking or thermal binding. The TVA letters of February 13, March 15, and August 6,1996, and December 1,1997, described the scope of valves evaluated in response to GL 95-07. The NRC staff has reviewed the scope of the licensee's susceptibility evaluation performed in response to GL 95-07 and found it complete and acceptable.

3.2 Corrective Actions 4

GL 95-07 requested that licensees, within 180 days, perform further analyses as appropriate, and take appropriate corrective actions (or justify longer schedules), to ensure that the susceptible valves identified are capable of performing their intended safety function under all l

modes of plant operation, including test configuration. The licensee's submittals discussed proposed corrective actions to address potential pressure-locking and thermal-binding problems.

3 The staff's evaluation of the licensee's actions is discussed in the following paragraphs:

The licensee stated that the following valves were modified to eliminate the potential for a.

pressure locking:

1,2 FCV-63-008 Residual Heat Removal (RHR) Pump To Safety injection (SI) 1,2 FCV-63-011 RHR Pump To Sl Pump Suction 1,2 FCV-63-022 Sl Pump To Cold Leg injection 1,2 FCV-63-039/040 Charging Pump injection 1,2 FCV-63-072/073 Containment Sump To RHR Pump Suction 1,2 FCV-63-172 RHR Hot Leg injection 1,2 FCV-74-001/002 RHR Suction From Hot Legs 1,2 FCV-74-003 RHR Pump A Suction 1,2 FCV-74-021 RHR Pump B Suction 1,2 FCV-74-033/034 RHR Crosstie Valves 1,2 FCV-63-172, RHR Hot Leg injection, were modified by the installation of a bypass line that connected the bonnet cavity to the high pressure side of each valve.

The valve in each bypass line is normally closed and is required to be opened prior to opening valves 1,2 FCV-63-172 to initiate hot leg recirculation flow. The position of the valve in each bypass line is controlled by procedures. The staff finds that physical modification to valves susceptible to pressure locking is an appropriate corrective action to ensure operability of the valves and is thus acceptable.

b.

The licensee stated that the following valves would be modified by the end of the Unit i refueling outage scheduled for October 1998, and the Unit 2 refueling outage scheduled for April 1999, to eliminate the potential for pressure locking:

1,2 FCV-63-025/026 Boron Injection Tank Outlet Isolation 1,2 FCV-72-002/039 Containment Spray Pump Discharge 1,2 FCV-72-040/041 Containment Spray Header Isolation i

As short-term corrective action, the licensee is relying on bonnet leakage or actuator capability calculations during locked rotor conditions to demonstrate that the valves can operate under pressure-locking conditions. The staff finds that the short-term corrective action is acceptable until the modifications to eliminate the potential for pressure locking are complete.

c.

The licensee stated that it used a thrust-prediction methodology developed by Commonwealth Edison Company (Comed) to demonstrate that the following valves could open under pressure-locking conditions:

1,2 FCV-01-016 Steam Supply to Turbine Driven Auxiliary Feedwater Pump 1,2 FCV-62-138 Emergency Boration isolation 1,2 FCV-63-001 RHR Pump Suction From Refueling Water Storage Tank 1,2 FCV-63-006/007 S1 Pump Suction From RHR 1,2 FCV-63-156/157 Sl Pump A to Hot Legs 1,2 FCV-68-332/333 Pressurizer Power Operated Relief Valves (PORV) Block Valves

On April 9,1997, the staff held a public meeting to discuss the technical adequacy of the Comed pressure-locking thrust prediction methodology and its generic use by licensees in their submittals responding to GL 95-07. The minutes of the public meeting were issued on April 25,1997. At the public meeting, Comed recommended that, when using j

its methodology, minimum margins should be applied between calculated pressure-locking thrust and actuator capability. These margins along with diagnostic equipment

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accuracy and methodology limitations are defined in a letter from Comed to the NRC dated May 29,1998 (Accession Number 9806040184). The NRC considers the use of l

the Comed pressure locking methodology acceptable provided these margins, diagnostic equipment accuracy requirements and methodology limitations are incorporated into the pressure-locking calculations. Comed indicated that its f

methodology may be revised. The staff considers that calculations that are used to l

l demonstrate that valves can overcome pressure locking are required to meet the requirements of 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants, and therefore, controls are required to be in place to ensure that any industry pressure-locking thrust prediction methodology requirements and revisions are properly implemented. Under this condition, the staff finds that the Comed methodology provides a technically sound basis for assuring that valves susceptible to pressure locking are capable of performing their intended safety-related function.

d.

The licensee stated that valves within the scope of GL 95-07 were evaluated for thermal binding. When evaluating whether valves were susceptible to thermal binding, the licensee assumed that thermal binding would not occur below specific ternperature thresholds. These assumptions were based on industry experience. The licensee did not consider that gate valves in systems with a normal operating temperature less than i

approximately 200*F were susceptible to thermal binding. Further, flexible wedge gate valves that are shut and experience a cooldown differential temperature of less than 100*F and solid wedge gate valves that are shut and experience a cooldown differential l

temperature of less than 50 F prior to opening were not considered by the licensee to l

be susceptible to thermal binding. The pressurizer PORV block valves, 1,2 FCV-68-332/333, exceeded these temperature limitations; however, operational history demonstrated that the valves are not susceptible to thermal binding.

The screening criteria and operational history results used by the licensee appear to provide a reasonable approach to identify those valves that might be susceptible to thermal binding. Until more definitive industry criteria are developed, the staff concludes that the licensee's actions to address thermal binding of gate valves are acceptable.

4.0 CONCLUSION

On the basis of this evaluation, the NRC staff finds that the licensee has performed appropriate evaluations of the operational configurations of safety-related power-operated gate valves to identify valves at the Sequoyah Nuclear Plant, Units 1 and 2, that are susceptible to pressure locking or thermal binding. In addition, the NRC staff finds that the licensee has taken, or is l

scheduled to take, appropriate corrective actions to ensure that these valves are capable of j

performing their intended safety functions. Therefore, the staff concludes that the licensee has adequately addressed the requested actions discussed in GL 95-07.

Principal Contributor: S. Tingen, NRR Date: