ML20211Q029
| ML20211Q029 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 10/16/1997 |
| From: | Huffman W NRC (Affiliation Not Assigned) |
| To: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 9710220064 | |
| Download: ML20211Q029 (5) | |
Text
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,j NUCLEAR REGULATORY COMMISSION W ASHINGT ON, D.C. soub0001 October 16, 1997 Mr. Nicholas J. Liparulu, Manager Nuclear Safety and Regulatory Analysis Nuclear and Advanced Technology Division Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, PA 15230
SUBJECT:
OPEN ITEM ASSOCIATED WITH AP600 PRESSURE ISOLATION VALVE LEAK TESTING
Dear Mr. Liparulo:
The Plant Systems Branch of the Nuclear Regulatory Commission (NRC) has provided a Safety Evaluation Report (SER) to Standardization Project Director-ate which evaluates the detection of leakage through the reactor coolant pressure boundary (RCPB) for the Ar500 design, included in the SER is an open item associated with insufficient leak testing of reactor coolant boundary pressure isolation valves (PIVs).
PlVs have been defined in the Westinghouse standard technical specifications (TSs) and in various NRC generic communica-tions as any two valves in series within the RCPB which separates the high pressure reactor coolant system (RCS) from an attached low pressure system.
However, there has been significant variation within the nuclear industry on the interpretation of which PlVs are leak tested as ASME,Section XI, Category A valves (seat leakage important) as specified in the 1988 version of OMa, Part 6, Sections 4.2.2.2 and 4.2.2.3.
The NRC staff previously raised issues on the need for Westinghouse to ioentify the AP600 PlVs which need to be leak tested and included in TSs in letters to Westinghouse dated December 24, 1996, and May 28, 1997. Westing-house letter dated August 29, 1997, stated that based on evolutionary design certification precedence, the AP600 has no PlVs, however, Westinghouse developed AP600 specific criteria for identifying PIVs.
Based on the devel-oped criteria, the only talves Westinghouse has identified for the AP600 design as PlVs are the accumulator discharge check valves. Westinghouse also proposed a technical specification requirement for operability and testing of these Plv valves, it aapears that Westinghouse is only requiring a full stroke test of the chec( valves and does not specify any need for leak testing.
The staff does not find the Westinghouse response sufficient or acceptable.
The need to identify and leak test PIVs separating certain systems and subsys-tems connected to the RCS is implied in the review and evaluation associated J)p[y w
with an applicant's compliance with Intersystem Loss-of-Coolant Accident (ISLOCA) evaluations identified in SECY-93-087, " Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Retetor (ALWR) Designs." One of the criteria specified in the ISLOCA SECY is that for l
systems not designed to withstand full RCS pressure (ultimate rupture strength I equivalent to RCS pressure), capability for leak testing of pressure isolation
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f Mr. Nicholas J. Liparulo October 16, 1997 valves should be provided. Westinghouse claims the AP600 conforms to the SECY intersystem loss-of-coolant accident criteria. However, the staff notes that where the RCPB interfaces with systems or subsystems which penetrate contain-ment and have an ultimate rupture strength less than RCS pressure, Westing-house does not require leak testing beyond containment isolation valve leak testing capability.
The staff believes that the intent of the SECY PlV criteria was that the PlVs be leak tested as-ASME Section XI, Category A valves in addition to 10 CFR Part 50, Appendix J.
The specific information the~ staff requires to resolve this issue are provided in the enclosure to this letter.
If you have any questions regarding this matter, you may contact me at=
(301) 415-1141.
Sincerely, original signed by:
William C. Huffman, Project Manager Standardization Project Directorate Division of Reactor Program Managertent Office of Nuclear Reactor Regulation Docket No.52-003
Enclosure:
As stated cc w/ enc 1:
See next page DISTRIBUTION:
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DATE 10/15/97 W D6 10/g/9T OFFICIAL RECORD COPY
Mr. Nicholas J. Liparulo Docket No.52-003 Westinghouse Electric Corporation AP600 cc: Mr. B. A. McIntyre Mr. Russ Bell Advanced Plant Safety & Licensing Senior Project Manager, Programs Westinghouse Electric Corporation Nuclear Energy Institute Energy Systems Business Unit 1776 I Street, NW P.O. Box 355 Suite 300 Pittsburgh, PA 15230 Washington, DC 20006-3706 Mr. Cindy L. Hang Ms. Lynn Connor Advanced Plant Safety & Licensing Doc-Search Associates Westinghouse Electric Corporation Post Office Box 34 Energy Systems Business Unit Cabin John, MD 20818 Box 355 Pittsburgh, PA 15230 Dr. Craig D. Sawyer, Manager Advanced Reactor Programs Mr. Sterling Franks GE Nuclear Energy U.S. Department of Energy 175 Curtner Avenue, MC-754 NE-50 San Jote, CA 95125 19901 Germantown Road Idaho Falls, ID 83415 Germantown, MD 20874 Mr. Robert H. Buchholz Mr. Frank A. Ross GE Nuclear Energy U.S. Department of Energy, NE-42 175 Curtner Avenue, MC-781 Office of LWR Safety and Technology San Jose, CA 95125 19901 Germantown Road Germantown, MD 20874 Barton Z. Cowan, Esq.
Eckert Seamans Cherin & Mellott Mr. Charles Thompson, Nuclear Engineer 600 Grant Street 42nd Floor AP600 Certification Pittsburgh, PA 15219 NE-50 19901 Germantown Road Mr. Ed Rodwell, Manager Germantown, MD 20874 PWR Design Certification Electric Power Research liistitute 3412 Hillview Avenue Palo Alto, CA 94303 l
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o SER ASSOCIATED WITH AP600 SER SECTION 5.2.5 250.30F To be consistent with SECY-93-087 intersystem LOCA concerns, West mghouse should apply the following criteria for identifying AP600 PlVs which should be leak tested and included in technical specifications (TSs):
(1) any portion of the connected low pres-sure systems outside containment not meeting the ISLOCA acceptance criteria, i.e., a component having the ultimate rupture strength less than the RCS operating pressure, or (2) the connected low pressure system is an engineered safety feature for mitigation of DBAs. Westinghouse should list all PlVs meeting the inclusion criteria and require technical specification (TS) operability conditions.
250.31F Westinghouse states'in the AP600 TS bases (pg. B 3.4-69) that the normal residual heat removal system (RNS) PIVs should be excluded from TSs because the RNS ultimate rupture strength is designed to be not less than the RCS operating pressure.
The NRC staff does not agree that the RNS design fully complies with the SECY 93-087 criteria that low ~>ressure systems interfacing with the RCS have an URS exceeding tie RCS operating pressure. Specifically, the RNS pump seal is not designed to meet this URS criterion.
Because the RNS pump seal is not designed to the RCS URS, failure of the RNS PlVs could result in the failure of RNS seals, which could re; ult in a LOCA outside containment. Therefore, the RNS PlVs should be included in the set of PlVs identified in TSs.
250.32F Unlike conventional operating reactor charging systems, the AP600 CVS charging system operates only intermittently and is normally not running. The staff notes that there is only one check valve (V064) which is normally closed and separates the reactor coolant pressure boundary from the CVS charging system outside of contain-ment. Although there are both check valves and discharge isola-tion valves on the CVS charging line outside of containment, this is not a conventional arrangement for PlVs (since PlVs are nor-mallv within the reactor coolant pressure boundary). Westinghouse should consider the need for another check valve within contain-ment on the charging system. The staff believes that Westinghouse will need to designate PIVs for the CVS charging line since the charging pump suction does not meet the ISLOCA criteria of SECY-93-087.
250.33F AP600 TS 3.4.16, "RCS Pressure Isolation Valve (PlV) Integrity,"
appears to be deficient to the staff. The TS should be improved to address the following concerns:
(a)
Condition A in TS 3.4.16 is defined as one or more RCS PIVs being inoperable.
How is the inoperability of PIV defined here? TS Bases page B3.4.71 states that the purpose of this LCO is to verify the PlVs have not suffered gross failures and that the operability tests specified in the inservice Enclosure
e test program (IST) implies that an acceptable method of determining valve integrity is the ebility of the PlVs to transition from open to closed. How does this operability condition ensure PIV leak tightness at expected differential pressure across the valves? This ap) ears to be inconsistent with the leakage LCO of Standard Tec1nical Specifications which specifies a leakage limit at the expected differential pressure across the PIV.
SR 3.4.16.1 (note typo in surveillance numbering) requires verification of operability of each PIV in accordance with the IST program. How is the PIV operability defined in the IST program? Section 3.9.6 of SSAR indicates that the valve leakage inservice testing is performed only for the contain-ment isolation valves in accordance with Appendix J.
How is the PIV leakage tested under the IST program with the RCS pressure?
(b)
The LC0 3.4.16 applicability MODES are limited to MODES I and 2, and MODES 3 and 4 with pressurizer pressure greater than 1000 asig.
Is this limitation of P > 1000 psig based on Westing 1ouse's conclusion that the accumulator discharge check valves are the only PIVs required to be included in the TS? The applicability MODES should not be limited to Pressurizer P > 1000 psig. When Required Action A.1 is not met, B.2 requires the pressurizer pressure to be reduced to less than 1000 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This is not accept-able.
B.2 should require the plant to be in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
(c)
The staff believes that the AP600 PIV TS should be modeled after the standard TS for PIVs.
However, the staff is willing to forego the need for PIV leak testing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of any automatic or manual actuation of a PIV.
The staff would be satisfied if the PIVs are Icak tested in accordance with the IST program schedule and prior to enter-ing MODE 2 if the unit has been in MODE 5 for more than 7 days and the PIVs have not been tested in the previous 9 months.
(d)
The TS 3.4.16 BASES Background section addresses the rela-tionship between reactor coolant pressure boundary valves (RCPB) and PIVs.
The staff does not believe that Westing-house has defined the RCPB consistent with the definition of 10 CFR 50.2.
Westinghouse should revise the Bases appropri-ately.
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