ML20203K035
| ML20203K035 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 02/26/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20203K030 | List: |
| References | |
| NUDOCS 9803040435 | |
| Download: ML20203K035 (7) | |
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1- UNITED STATES
/ s* NUCLEAR REQULATORY COMMISSION WAsHINGToh D.C. ReseHoot SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE THIPD TEN-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN REQUEST FOR RELIEF NO. PR-10 E.OB NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DCCKET NUMBERE 50 298
1.0 INTRODUCTION
The Technical Specifications for Cooper Nudear Station, state that the inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 components shall
- be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and l ._ applicable Mdsnda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR 50.55a(s)(3) states that altamatives to the requirements of parograph (g) may be used, when authorized by the NRC, if (i) the proposed attematives would provide en accepteble level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual d;fficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code CIm 1,2, and 3 components (including supports) shall meet the requirements, except the, Gesign and access provisions and the preservice examination requiremertis, set forth in the ASME Code,Section XI, " Rules for Inservice inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first ten year interval and subsequent intervals comply with the requirements in the latest edition ar,J addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) on the date twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for the Cooper Niiclear Station third 10-year inservice inspection (ISI) interval is the 1989 Edition. The components (including supports) may meet the require.r. ems set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission approval.
ENCLOSURE 1 9903040435 980226 PDR ADOCK 05000298 P PDR , _ . -
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. 2-Persuant to 10 CFR 50.55a(g)(5)(iii), if the licensee detennines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the Commission in support of that determination and a request made for relief from the ASME Code requirement. After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(G)(i), the Commission may grant relief and r.my impose attemative requirements that are determined to be authorized by law, will not endanger life, property, or the com non defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed, in letters dated Jar uary 9,1998, and Janussy 30,1998, the Nebraska Public Power District submrtted to the NRC its Third Ten-Year Interval inservice Inspection Program plan, Request for '
Relief No. PR-to for Cooper Nuclear Station.
2.0 EVALUATION / CONCLUSIONS The staff, with technical assistance from its contractor, the Idaho National Engineering and Environmental Laboratory (INEEL), has evaluated the information provided by the licensee ir, support of its Third Ten-Year interval inservice inspection (ISI) Program Plan, Request for Relief No. PR 10 fer Cooper Nuclear Station. Based on the information submitted, the steff adopts the
! contractor's conclusions and recommendations presented in the Technical Letter Report l attached.
L Pursuant to 10 CFR 50.55a(a)(3)(ii), the hcensee proposed that, in lieu cf performing a Class 1 Inservice system pressure test on the Main Steam scfety/ relief valve pilot cartridge assemblies at 1005 psig, it will perform a pressure test on these components at a minimum of 900 psig during the normal startup following their eplacement. In addition, the licensee proposed that if there is ,
an unplanned shutdown with a drywell entry before the next refueling outage, another inspection of these bolted connections wi!! be performed to look for any evidence of leakage.
The staff concludes that requiring the licensee to perfctm a system leakage test at a trvninal operating pressure associated with 100% rated reactor power would result 10 a hardship without
- a compensating increase in quality and safety, because of the following: 1) the safety / relief valves are not isolable from the reactor vessel arid the entire primary system would need to be pressurized in order to petform this test; 2) a laakage test at a pressure not less than the nominal operating pressure associated with 100% rated reactor power (1005 psig) cannot be performed during a rummi plant startup, due to the excessive temperature and radiological exposure conditions to whM::h licensee personnel would be exposed in the primary containment; and 3) the hcensee would have to perform extensive valve manipulations, system lineups and procedural controls, ihat would be rquired in order to heat up and pressurize the primary system to establish the necessary test prea ture, during plant outage conditions, without withdrawal of control rods.
The staff further concludes that for Cooper Nuclear Station, a system leakage test at a minimum pressure of 900 psig will provide reasonable assurance of continued operational readiness of the subject mechanical connections. The proposed test pressure will be adequate to cause leakage from the mechanical connection between the valve body and the pilot cartridge assembly if a leak-tight connection has not been established. Therefore, the staff concludes that the licensee's
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proposed aliemative contained in Request for Relief No. PR -10 to perform the system leakage
. test at 900 psig minimum following the reassembly of the mechanical connection between the valve bodies and the pilot cartridge assemblies of the 8 main steam safety / relief valves, is acceptable. Accordingly, pursuant to 10 CFR 50.55a(a)(3)(ii), the sta# authorizes Relief Request No. PR 10 for the Cooper Nuclear Station.
PrincipalContributor. T. McLellan Date: February 26, 1998 f $$
TECHNICAL LETTER RdPORT 1
[ QM THE THIRD 10-YEAR INTERVAL INSERVICC INSPECTION REQUEST FOR RELIEF PR 10 f.QB
- NEBRA8K4 PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NUMBER
- 50-298 i
- 1. INTRODUCTION e
! By letters dated January 9, and Jariuary 20,1998, the licensee, Nebraska Public Power
! District, submitted Requst for Relief PR 10 Revision 0 and Revision 1, respectively, containing a proposed altomative to the requirements of the ASME Code,Section XI, for
, the Cooper Nuclear Station. The proposed attemative examination is for use during the 1998 mid cycle outspe, which is scheduled to begin February 27,1998. TheIdaho National Engineering and Environmental Laboratory (INEEL) staff has evaluated the j licensee's proposed altamative in the following section.
- 2. EVALUATION The information provided by Nebraska Public Power District in support of the proposed
. altemative to he Code requirements has been evaluated and the basis for dbposition is documented below. The Code of record for the Cooper Nuclear Station, third 10-year ISI
! interval, which began March 1 1996 and ends February 28,2006, is the 1989 Edition of Section XI of the ASME Boiler and Pressure Vessel Code.
Request For Relief PR-10 Revision 1, Examir ation Category B-P, item Number B15.10, Reactor Vessel Pressure Retaining Boundary.
Code Reauirement: Section XI, Table IWB-2500, Examination Category B-P, item B15.10, requires a system leakage test of the Reactor Vessel Pressure Retaining Boundary, lWB-5221(a), requires that the system leakage test shall be conducted at a test pressure not less than the nominal operating pressure associated with 100% rated reactor power.
Ucensee's Proposed Altemative (as stated):
"In lieu of performing a Class 1 inservice system pressure test on the Main Steam safety relief valve pilot cartridge assemblies at 1005 psig, CNS shall perform a pressure test on these components at a minimum of 900 psig duiing the norma! :tartup following their replacement. If there is an unplanned shutdown with a drywell entry before the next ENCLOSURE 2
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2 refueling outage, another inspection of these bolted connectio.)s will be performed to look for any evidence ofleakage."
Lloensee's Basis for the Proposed Ahemative (as stated):
"During the 1998 mid-cycle outage, Cooper Nuclear Sation (CNS) will be replacing the pilot caitridge assemb!ies 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 XI requires an inservice leak test and a VT.2 visual examinatior, 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 .he valve body and the pilot cartridge assembly. Since the safety relief valves are not isolable from the reactor vessel, the entire primary system will need to be pressurized in order to perform this to.t.
"A leakage test at a pressure not less than the nominal operating pressure associated with 100% rated reacbr 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 conta'nment.
" Extensive valve manipulations, system lineups and procedural controls are required in order to heat up and pressurize the primary system to establish the necessary test pressure, during plant outage conditions, without withdrawal of control rods. This is done in order to perform the necessary leakage test without exposing the inspectors to the excessive temperature and radiological exposure conditions as desuibed in the paragraph ,
above. This special test usually takes one full day of plant outage time, and the additional valve lineups and system reconfigurations necessary to suppo.1 this special test, impose an additional challenge 10 the affected systems. A normal plant startup then occurs after completion and subsequent recovery from the test procedure.
"During a normal plant startup, operators are required to perform walkdown inspections of the primary containment at approximately 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 connections will be reexamined for signs of leakage, during the sptem leakage test, which will be performed at the end of the outage as required by ASME [XI).
"Research by the ASME in support of Code Cases N-416-1 and N 498-1 has demonstrated that leakage rates are proportional to the test pressure. Thus a pressure test at 900 psig, during a norm:.i plant startup, will identify any leakage through the mechanical connection betwun our safety-rohef valves and their pilot assemblies. As such, the altemate valve kneups requirect for the inse vice leakage test is an additional challenge to plant systems, which does not serve to add any notable value to the inspection process and/or results.
This is especially true for the upcoming mid-cycle outage, which is of limited scope, and where the only components on the primary system which are being replaced, and which
- result in the need for this leakage test, are the pilot assemblies on the eight safety relief valves.
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" Leakage at the bolted conaection between the pilot assembly and the valve body would not prevent the safety relief valve (SRV) from performing its safety function. Leakage from this connection would be detected by the drywell monitoring system. Plant Technical Specmcabr,s limit leakage to less than a two gallon per minute increase in unidentified '
leakage in the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; a totsi of five gallons per minuto unidentified; or 25 gallons per minute identified. Therefore, based on the set point tests, the drywell leakage system, the ability of the SRV to perform its design function in the event of a leak at this connection, and our plant operating limitations, as established in the site Technical Specifications, relief is requested in accordance viith 10 CFR 50.55a(a)(3)(ii). The
- proposed leakage testing alterriative provides sa equivalent level of quality and safety."
Evaluation: Paragraph IWB-5221(a) requires that system leakage tests be conducted at a test pressure not less than the nominal operatinq pressure associated with 100% rated reactor power. Releef is being requested from the Coda required system pressure test at nominal operating pressure following the replacement of the pilot cartridge assemblies on the eieht Main Steam safety relief valves. The Main Steam safety relief vahms are nonisolable and require the entire primary system to be pressurized in order to perfo m the required system leakage test.
To perform a system leakage test at the nominal operating pressure associated with 100%
rated reactor power, the reactor must achieve 100% power or system lineups with associated valve alignment must be performed to provide the required pressure.
Performance of the system leakaf,e examination at full reactor power would result in excessive temperature and radiological conditions. These conditions may compromise the quality of the examination as well as pose potential safety concems and hazardeus 9 conditions for the inspection personnel. To obtain nominal operating pressure without the reactor at full power, it would be necessary to perform extensive vabe manipulations, system lineups and procedural controls.- By performing these tr ' e excessive temperature and radiological conditions would be reduced. Hower, due to the additional valve lineups ano system reconfigurations necessary, an additional challenge to the affected systems will result.- Therefore, imposition of thn Code requirements would result in a considerable burden on the licensee.
As an attemative, the licensee proposed to perform the system leakage test on the Main Steam safety relief valve pilot cartridge assemblies at a minimum pressure of g00 psig during the normal startup. Also, if an unplanned shutdown occurs prior to the next refueling outage, and entry to the drywell occurs, the licensee committed to perform another inspection of the components to look for evidence of leakage.
The INEEL staff believes that the proposed test pressure (approximately g0% of the nominal operating pressure) will be adequate to cause leakage from the mechanical connection between the valve body and the pilot cartridge assembly if a leak-tight connection has not been established. Furtheimore, the INEEL staff believes that requiring the licensee to perform the system leakage test at the nominal operating pressure will result in a hardship without a compensating increase in quality and safety. Based on the proposed attemative test pressure of at least 000 psig combined with the drywell leakage monitoring system and Plant Technical Specification requirements for leakage allowance, it
is reasonable to conclude that leakage, N lt should occur, will be detected, providing reasoncble assurance of operational readiness.
- 3. CONCLUSION For Cooper Nuclear Station, a system leanage test at 900 psig minimum should provide reasonable assurance of continued operational readiness of the subject mechanical connections. Requiring the licensee to perform a system leakage test at a nominal operating pressure associated with 100% rated reactor power would result in a hartiship 3
without a compensating increase in quality and safety. Therefore, it is recommended that 1 the proposed altemative to perform the system leakage test at 900 psig minimum following the reassembly of the mechanical connection between tne valve bod / and the pilot cartridge assembly, be authorized pursuant to 10 CFR 50.55a(s)(3)(li).
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