ML20154C873
| ML20154C873 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 09/29/1998 |
| From: | Eric Simpson Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LCR-H97-10, LR-N98423, NUDOCS 9810070050 | |
| Download: ML20154C873 (5) | |
Text
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Pubhe. Service Electric and Gas Company E. C. simpson Public Service Electric and Gas Company PO Box 236. Hancocks Bndge, NJ 08038 609-339-1700 Sono vice Bes do'n. Naclear Engeneering SEP 2 91998 LR-N98423 LCR H97-10 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
SUPPLEMENT TO A REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS SAFETY RELIEF VALVE SETPOINT TOLERANCES HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 On April 28,1998, via letter LR-N97117, Public Service Electric & Gas (PSE&G)
Company transmitted Li::ense Change Request (LCR) H97-10 to the NRC to request a revision to the Technical Specifications (TS) for the Hope Creek Generating Station.
Specifically, LCR H97-10 requested, in part, a revision to the Safety Relief Valve (SRV) tolerance acceptance limits contained in the TS. On August 6,1998, the NRC issued a Request for Additional Information (RAI) associated with that LCR. The NRC questions contained in the RAI, as well as PSE&G's responses, are contained in Attachment 1 of this letter.
PSEdG has determined that the information contained in the attachment to this letter does not alter the conclusions reached in the 10CFR50.92 No Significant Hazards analysis previously submitted with LCR H97-10. In accordance with 10CFR50.91(b)(1),
a copy of this submittal has been sent to the State of New Jersey.
Should you have any questions regarding this request, please contact James Priest at 609-339-5434.
Sincerely, 1
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Attachment
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9810070050 990929 PDR ADOCK 05000354 p
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SEP 2 91998 Document Control Desk LR.-N98423 h
C Mr. H. Miller, Administrator - Region i U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 4
Mr. R. Ennis.
Licensing Project Manager - Hope Creek c"
U. S. Nuclear Regulatory. Commission One White Flint North Mail Stop 14E21 11555 Rockville Pike Rockville, MD 20852 Mr. S. Pindale (X24)
USNRC Senior Resident Inspector - HC Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P. O. Box 415 Trenton, NJ 08625 i
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4 95 4933
REF: LR-N98423 LCR H97-10 STATE OF NEW JERSEY )
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- COUNTY OF SALEM
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l E. C. Simpson,'being duly sworn according to law deposes and says:
I am Senior Vice President - Nuclear Engineering of Public Service Electrb and Gas Company, and as such, I find the matters set forth in the above reference d letter, concerning Hope Creek Generating Station, Unit 1, are true to the best of my knowledge, information and belief.
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Subscribed and Swor to before me thisd9M ay o
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Attachmsnt 1 LR-N98423 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION l
1.
In Licensee Event Report (LER)97-024, the licensee reported the setpoint 3
performance for the plant Target Rock 2-Stage safety / relief valves (SRVs).
Several as-found setpoint values were much in excess of the 11%
tolerance in the current technical specifications (TS) and the i3% toler::nce proposed by the licensee. The LER states that platinum ion-beam implanted pilot disks were installed in 7 of the 14 SRVs and that the effectiveness of this modification would be evaluated following Cycle 8. If the evaluation of the 7 platinum ion-beam implanted disks indicates that 4
they are ineffective, what additional corrective actions would be necessary to ussure that the SRVs will operate within the i3% tolerance? If the j
evaluation of the disks indicates they an effective in reducing the setpoint drift to within i3%, is it planned that the new disks would be installed in all j
14 SRVs?
1 J
Evaluation of the effectiveness of the platinum ion-beam implanted discs is an ongoing process at Hope Creek and also within the industry. The lead plant utilizing platinum ion-beam ing... itation has performance data spanning three operating cycles, which has demonstrated significant reductions in corrosion induced bonding. PSE&G believes that the success established by this procedure at the lead o'mt provides a high level of assurance that as found pilot assembly setpoint test wits will not exhibit the classic corrosion bonding phenomena experie 1ced at Hope Creek. Since PSE&G believes that platinum ion-beam implantation will effectively reduce SRV setpoint drift resulting from corrosion bonding, Hope Creek is planning on installing the remaining complement of platinum ion-beam implanted disks during the next refueling outage (RF08), such that all 14 SRVs will have platinum ion-beam implanted disks. Reuse of the seven platinum ion-beam implanted disks currently installed is planned, provided that i
machining is not required to recondition the surface. If machining is necessary, then re-installation of stellite disks may be required since re-application of the platinum ion coating may not support the outage schedule. PSE&G believes that additiond corrective actions to address corrosion bonding induced SRV setpoint drift are not necessary.
In addition, the Hope Creek SRV overpressure relief design utilizes a Low-Low i
set function, which cycles two SRVs at reduced setpoints (1017 and 1047 psig) utilizing Automatic Depressurization System (ADS) channel logic and air actuators force to open the valves. The function of this design feature and logic Page 1 of 2
' Attachmint 1 LR-N98423 is similar in nature to the " pressure sw;tch" proposal approved in LTR NEDC-32121, " Pressure Switch / Transmitter for Two Stage Target Rock Safety / Relief Valve". The design of the SRV Low-Low set function provides additional assurance that the safety limits established in TS would not be exceeded.
Therefore, PSE&G believes that additional actions to address SRV setpoint drift are not necessary.
2.
Uncertainties of analysis parameters should be accounted for in the safety 4
analyses used to bound limits defined by the plant TS. Provide a
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discussion of the SRV setpoint testirig instrument accuracy and how this j
source of uncertainty is accounted for in the licensee's safety analysis associated with the proposed TS SRV setpoint tolerance.
i SRV as-found setpoint testing is performed in accordance with approved vendor procedures that meet the requirements of ASME/ ANSI OM-1987, " Operation and Maintenance of Nuclear Power Plants". This standard specifies, in section 1.4.1.2, a testing uncertainty for SRV set pressure of +2% to -1% at the pressure level of interest, with an overall combined eccuracy not to exceed +1% to -2% of the indicated (measured) set pressure. NWS Technologies and other vendors performing setpoint testing for SRVs typically use instruments with an accuracy of i 0.15% to ensure ASME/ ANSI criteria are met. This equates to a potential instrument inaccuracy of approximately 1.7 psi for any SRV, which is a small fraction of any of the SRV setpoints. For example,1.7 psiis approximately 0.15%
of the 1108 psig SRV setpoint and is well within the 13% (or approximately 33 psi) as-found testing tolerance associated with this TS change.
The safety analyses that were pedormed to justify the 13% TS acceptance limit did not specifically account for setpoint testing instrument inaccuracy. Bouding safety analyses were performed, which justified an as-found condition of 13%
around the nominal setpoints. The safety anWyses directly utilized an input assumption of 13% or an analytical input assumption of a more bounding absolute setpoint in order to justify the proposed 3% TS acceptance limit.
Since the proposed TS require the SRVs be adjusted to within 1% ( 11psig) tolerance prior to reinstallation in the plant, an approximate 2% (22 psig) margin to the TS acceptance limit is ensured. However, the setpoint testing instrument inaccuracy will be accounted for in the acceptance criteria of the surveillance procedures used to determine the as-found SRV setpoint. Therefore, PSE&G concludes that instrument accuracy is appropriately accounted for and that the safety analyses performed will remain bounding.
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