ML20199G561
| ML20199G561 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 01/30/1998 |
| From: | Graham P NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NLS980024, NUDOCS 9802040297 | |
| Download: ML20199G561 (7) | |
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.n NLS980024 January 30,1998 '
U.S. Nuclear Regulatory Commission Attention: Document Control Desk
- Washington, D.C. 20555-0001 Gentlemen:
Subject:
Third 10-Year inteival inservi,:e inspection Program; Relief Request PR-10, Revision 1 Cooper Nuclear Station, NRC Docket 50-298, DPR-46
Reference:
Letter (No. NLS980003) to U.S. NRC Document Control Desk from Phillip D.
Graham (NPPD) dated January 9,1998,"Tnird 10 Year Interval inser'fice Inspection Program; Relief Request PR-10" The purpose of this letter is to submit Relief Request PR-10, Rethion 1, for Nuclear Regulatory Commission (NRC) review and approval; Revision 0 of this relief request had been previously submitted by the Nebraska Public Power District (District) per the above reference.
- The revised relief request contains additional information in support of the basis for relief, including impact on personnel hazards, valve lineups and system configuration. The revised y
relief request (contingent upon NRC approval) also contains an additional commitment to inspect the bolted connections for evidence ofleakage if there is an unplanned. hutdown with drywell s
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entry before the next refueling outage. A component drawing is included with this submittal to aid in the NRC's review and approval of the subject relief request. The District is requesting NRC approval by February 20,1998, in order to support the mid-cycle outage that is planned to commence February 27,1998, but could occur sooner, l
Should you have any questions concerning this matter, please contact me.
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- NLS980024--
- -l January 30,1998 -
Page 2 'of 2; cc: Regional Administrator USNRC - Region IV-Senior Project Manager USNRC - NRR Project Directorate -
Senior Resident inspector
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Cooper Nuclear Station Third Interv:1 to NLS980024 Inservice Inspection Program Pag *e1of7 RELIEF REQUEST NUMBER: PR 10, REVIEION 1 COMPONENT IDENTIFICATION Code Classes:
1
References:
IWB-5221(a)
Examination Categories:
B-P ltem Numbers:
B15.10
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Description:==
System leakage test pressure
- Component Numbers:
MS-RV-71 ARV, MS-RV-71BRV, MS-RV-71CRV, MS-RV-71DRV, MS-RV-71ERV, MS-RV-71FRV, MS-RV-71GRV, MS-RV-7111RV 3
CODE REOUIREMEN_T IWB-5221(a) requires the cystem leakage tests to be conducted at a pressme not less than the nominal operating pressure associated with 100% rated reactor power.
DASIS FOR RELIEF During 'he 1998 mid-cycle outage, Cooper Nuclear Station (CNS) will be replacing the pilot cartridge assemblies on the eight Main Steam safety relief valves. The pilot cartridge assemblies will have been set point tested and certified to OM-1 requirements prior to installation. ASME
. X1 requires an i 1 service leak test and n VT-2 visual examination for these replacements.
Because the pilot cartridge assemblies will have been set point tested to pressures greater than that required by IWB-5221(a), this system leakage test focuses on the mechanical connection between the valve body and the pilot cartridge assembly. Since the safety relief valves are not 6
isolable from the reactor vessel, the entire primary system will need to be pressurized in order to perform this test.
A leakage test at a pressure not less than the nominal operating pressure associated with 100%
rated reactor power (1005 psig) cannot be performed during a normal plant startup, due to the excessive temperature and radiological exposure conditions which the inspectors would be exposed to in the primary containment.
Extensive valve manipulations, system lineups and procedural controls are required in order to heat up and pressurize the primary system to establish the necusary test pressure, during plant outage conditions, without the withdrawal of control rods. This is done in order to perform the necessary leakage test without exposing tLe inspectors to the excessive temperature and i
radiological exposure conditions as described in the paragraph abow. This special te::t usually takes one full day of plant outage time, and the additional valve lineups and system reconfigurations necessary to support thic special test, impose an additional challenge to the affected systems.- A normal plant startup then occurs after completion and subsequent recovery from the test procedure.
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Cooper Nuclear Station Third Interval to NLS980024 Inservice inspection Program Pag'e 2 of d
- RELIEF REQUEST NUMBER: PR-10, REVISION 1 During a normal plant startup, operators are required to perform walkdown inspections of the primary containment at approximateiy 500 and 900 psig. The bolted connection is not insulated,.
and is readily accessible for VT-2 examination during these primary containment walkdown inspections. During the next refueling outage, scheduled for the Fall of 1998, these connec+ ions will be reexamined for signs ofleakage, during the system leakage test, which will be performed at the end of the outage as required by ASME IX.
Research by the ASME in supnort of Code Cases N-4161 and N-4981 has demonstrated that leakage rates are proportional to the test pressure. Thus a pressure test at 900 psig, during a normal plant startup, will identify any leakage through the mechanical connection between our safety-relief valves and their pilot assemblies. As such, the alternate valve lineups required for i
- the inservice leakage test is an cdditional challenge to plant systems, which does not serve to add any notable value to the inspection process and/or results. Tnis is especially true for the upcoming mid-cycle outage, which is oflimited scope, and where the only components on the
- primary system which are being replaccd, and which result in the need for this leakage test, are the pilot assemblies on the eight safety relief valves.
Leakage at the bolted connection between the pilot assembly and the valve body would not preveot the safety relief valve (SRV) from performing its safety function. Leak ge from ihis connection would be detected by the drywell monitoring system. Plant Technical Specifications limit leakage to less than a two gaUon per minute increase in unidentified leakage in the previous 3
24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; a total of five gallons per minute unidentified; or 25 gallons per minute identified. Therefore, based on the set pomt tests, the drywell leakage system, the ability of the SRV to perform its desip. unction in the event of a leak at this connection, and our plant operating limitations, as established in the site Technical Specifications, reliefis requested in accordance with 10 CFR 50.55a(a)(3)(ii). The proposed leakage testing attemative provides en equivalent level of quality and safety.
PROPOSED ALTERNATE PROVISIONS In lieu of perfonning a Chtss I inservice system pressure test on the Main Steam safety relief valve pilot cartridge assemblics at 1005 psig, CNS shall perform a pressure test on these components at a minimum of 900 psig during the normal startup following their replacement. If there is an unplanned shutdown with a drywell entry before the next refueling outage, another inspection of these bolted connections will be performed to look for any evidence ofleakage.
APPLICABLE TIME PERIOD
.Reliefis requested for the 1998 mid-cycle outage ut CNS, which is currently scheduled to begin on Friday,2-27-98.
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l ATTACitMENT 3 LIST OF NRC COMMITMENTS l
Correspondenc3 No: NLS980024 The following table identifies those actions committed to by the District in this document.
Any other actions discussed in the submittal repreuent intended or planned actions by the District. They are described to the NRC for the NRC's information and are not regulatory commitments.
Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
COMMITTED DATE COMMITMENT OR OUTAGE CNS will perform a pressure test on MS-RV-71ARV, MS-RV-1998 Mid-Cycle Outage 71BRV, Pl-RV-71CRV, MS-RV-71DRV, MS-RV-71ERV, MS-RV-Contingent upon NRC 71FRV, MS-RV-71GRV, MS-RV-71HRV at a minimum of 900 peig approval.
This during the normal startup following their replacement, commitment replaces Disposition of observed leakags will consider the the same commitment marginal increase in leakage rates that would occur at made in Letter No.
the nominal operating pressure associated with 100% rated NLS980003, power.
Next unplanned If there is an unplanned shutdown with a drywell entry
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before the next refueling outage, another inspection of Y'
the bolts associated with MS-RV-71ARV, MS-RV-71BRV, MS-RV-71CRV, MS-RV-71DRV, MS-RV-71ERV, MS-RV-71FRV, MS-RV-71GRV, MS-RV-71HRV will be performed to look for any
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PROCEDURE 14 UMBER 0.42 REVISION NUMBER 5 PAGE 8 OF 13