ML20217P922
| ML20217P922 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/28/1998 |
| From: | Eric Simpson Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20013H969 | List: |
| References | |
| LR-N97117, NUDOCS 9805070139 | |
| Download: ML20217P922 (9) | |
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Puthe Servce Dectrc arsi Gas Company E. C. simpson Public Servico Electnc and Gas Company F.o. Box 230. Hancocks Bndge. NJ 08038 609 339 4 700 ScNv V<e Pms @nt %cacar Er g+eenng APR 2 81998 LR-N97117 LCR H97-10 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
l REQUEST FOR CHANGE TO TECHNIC /.L SPECIFICATIONS SAFETY RELIEF VALVE SETPOINT TOLERANCES HOPE CREEK GENERATING STATION l
FACILITY OPERATING LICENSE NPF-57 l
DOCKET NO. 50-354 10 accordance with 10CFR50.90, Public Service Electric & Gas (PSE&G) Company j
hereby requests a revision to the Technical Specifications (TS) for the Hope Creek 4
Generating Station (HC). In accordance with 10CFR50.91(b)(1), a copy of this l
submittal has been sent to the State of New Jersey.
l The proposed revision represents changes tc Technical Specification section 3/4.4.2,
" Safety / Relief Valves" and is being made to implement more appropriate safety / relief valve (SRV) setpoint tolerances. Justification for these proposed changes was developed from: 1) an NRC Safety Evaluation Report, dated 3/8/93, for the "BWR Owners Group lnservice Pressure Relief Technical Specification Licensing Topical j
Report", NEDC-31753P; and 2) General Electric report NEDC-32511P, " Safety Review l
for Hope Creek Safety / Relief Valve Tolerance Analyses" The proposed changes have been evaluated in accordance with 10CFR50.91(a)(1),
using the criteria in 10CFR50.92(c), and a determination has been made that this request involves no significant hazards considerations. The basis for the requested change is provided in Attachment 1 to this letter. A 10CFR50.92 evaluation, with a determination of no significant hazards consideration, is provided in Attachment 2. The marked up Technical Specification pages affected by the proposed changes are provided in Attachment 3. Pursuant to Attachment 4 of this letter, Attachment 5 of this submittal, General Electric report NEDC-32511P, " Safety Review for Hope Creek Safety /Reiief Valve Tolerance Analyses", dated April 1996, contains proprietary information and therefore should be withheld from public disclosure.
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l Document Control Desk l LR-N97117 l-Upon NRC approval of this proposed change, PSE&G requests that the amendment be made effective on the date of issuance, but allow an implementation period of sixty l
days to provide sufficient time for associated administrative activities.
ShotJd you have any questions regarding this request, we will be pleased to discuss them with you.
Sincerely, l
Affidavit Attachments (5)
C Mr. H. Miller, Administrator - Region l U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. R. Ennis, Licensing Project Manager - HC U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. S. Pindale (X24)
-l USNRC Senior Resident inspector - HC Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering i
P. O. Box 415 l
Trenton, NJ 08625 95 4933
i 8
REF: LR-N97117 LCR H97-10 STATE OF NEW JERSEY )
) SS.
COUNTY OF SALEM
)
i E. C, Simpson, being duly sworn according to law deposes and says:
I am Senior Vice President - Nuclear Engineering of Public Service Electric and Gas i
Company, and as such, I find the matters set forth in the above referenced letter, concerning Hope Creek Generating Station, Unit 1, are true to the best of my knowledge, information and belief.
OYX146>1 u-Subscribgcpand Swor this d J, day of [n to befor rne'.1998 i
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/)tA Notary Publi(of New Jersey My Commission expires on /# /38*
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4 Document Control Dask LR-N97117 Attachmsnt 1 LCR H97-10 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 SAFETY RELIEF VALVE SETPOINT TOLERANCES BASIS FOR REQUESTED CHANGE:
The changes proposed in this request establish more appropriate TS setpoint tolerance limits for Hope Creek's safety / relief valves (SRVs).
Implementation of these proposed changes can reduce:
- 1) the number of SRVs tested each outage; and 2) the number of Licensee Event Reports (LERs) transmitted by Hope Creek to document occurrences of SRV setpoint drift.
The SRV setpoint tolerance is being modified to reflect valve capability and will not challenge the safety function of these valves.
REQUESTED CHANGE AND PURPOSE:
1 As shown in Attachment 3 of this letter, LCO 3.4.2.1 is being modified to replace the il% setpoint tolerance limit for SRVs with a i3% setpoint tolerance limit.
Implementation of this j
change enables Hope Creek:
- 1) to potentially reduce the amount of SRV testing performed to comply with Surveillance Requirement 4.4.2.2; and 2) to potentially reduce the number of LERs required to document occurrences of setpoint drift.
In addition, Surveillance Requirement 4.4.2.2 is being revised to state that all SRVs must be certified to be within il% of the TS setpoint prior to returning the valves to service after setpoint testing.
The purpose of this change is to ensure that the SRVs are within established limits contained in plant analyses.
BACKGROUND:
On March 8, 1993, the NRC issued a Safety Evaluation Report (SER) for a BWR Owners Group Licensing Topical Report NEDC-31753P, "BWROG In-service Pressure Relief Technical Specification Revision Licensing Topical Report".
In the SER, the NRC stated that: 1) a generic change of [SRV) setpoint tolerance to i3% is acceptable; 2) the relaxation of the [SRV) tolerance from 1% to 3% will in itself result in a reduction in the number of LERs required; 3) plant specific TS changes to raise the allowable setpoint tolerance to 3% must include the requirement that additionel testing be conducted if failures are experienced (as currently defined in ASME Boiler and Pressure Vessel Code, Section lWV 3513); and 4) in all cases the valve setpoint will be restored to within 1% prior to plant startup (this testing is consistent with current requirementsi.
Page 1 of 4
1 Documant Control Desk LR-N97117 LCR H97-10 In the SER, the NRC stated that a licensee choosing to implement these TS changes must provide certain plant specific analysis.
The analysis provided should include the following:
1.
Transient analysis of all abnormal operational occurrences as described in NEDC-31753P, utilizing a 36 setpoint tolerance for the safety mode of SRVs.
2.
Analysis of the design basis overpressurization event using the 3% tolerance Jimit for the SRV setpoint to confirm that the vessel pressure does not exceed the ASME pressure vessel code upset limit.
3.
The analyses conducted must assure that the number of SRVs correspond to the number of valves required to be operable in the TS.
4.
Re-evaluation of the performance of high pressure l
systems (pump capacity, discharge pressure, etc.), motor-operated valves, and vessel instrumentation and associated piping considering the 13% tolerance limit.
5.
Evaluation of the 3% tolerance on any plant specific alternate operating modes (e.g.,
increased core flow, extended operating domain, etc.).
6.
Evaluation of the effect of the 3% tolerance limit on i
the containment response during loss of coolant accidents and the hydrodynamic loads on the SRV discharge lines and containment.
The NRC also stated that these plant specific analyses should be used for the initial justification for the TS modifications and be addressed in all future reload submittals.
In April, 1996, General Electric completed Hope Creek specific evaluations for plant operation with the proposed 3% setpoint tolerance limit for the SRVs.
These results were documented in General Electric report NEDC-32511P, " Safety Review 'or Hope Creek Generating Station Safety / Relief Valve Tolerance Analyses."
(
JUSTIFICATION OF REQUESTED CHANGES:
I PSE&G has reviewed General Electric analyses for Hope Cteek, j
performed with the proposed i3% setpoint tolerances for the safety mode of the SRVs, and concurs with the results.
The l
analysis performed for Hope Creek assumed that:
- 1) the plant operating parameters and the core design are consistent with the l
Page 2 of 4
l Document Control Desk LR-N97117 i
Attachmnnt 1 LCR H97-10 i
l Hope Creek Reload 6 Cycle 7 licensing calculations (24A5173, Supplemental Reload Licensing Report for Hope Creek Generating l
Station Unit 1 Reload 6 Cycle 7, Rev.
O, July, 1995); and 2) one SRV was out of service (consistent with the TS).
The results l
concluded that there was no impact on fuel therma] limits.
1 l
Specific results from these analyses are as follows:
1.
The vessel overpressurization analyses for the most limiting pressurization event, the main steam isolation
{
valve (MSIV) closure with flux scram, indicates the vessel l
l pressure remains within the ASME Upset Code limit of 1375 psig (with a conservative margin of more than 40 psig) at I
the increased setpoint tolerance.
Instrumentation utilized for normal and accident operation will not be affected based on the fact that the system was designed to meet a transient of 1375 psig as described in the Technical Specifications.
2.
The Anticipated Transient Without Scram (ATWS) analysis for the MSIV closure event determined that vessel pressure remains well within the ATWS design criteria of 1500 psig for the maximum vessel overpressure.
3.
A loss of coolant accident (LOCA) evaluation determined that the increase in setpoint tolerances would have negligible or no impact on the accident analyses conducted.
4<
Analyses performed on the high pressure emergency core cooling systems indicate that these systems can perform their design functions under the new setpoint tolerances due to the very small increase in operating pressure and/or temperature.
In addition, PSE&G has evaluated the impact on motor-operated valve operability with SRV setpoints at +3%.
3 This evaluation concluded that the high pressure systems at Hope Creek will perform their required functions satisfactorily with the potential increase in differential pressure.
5.
There is no impact on the design basis accident LOCA containment pressure and temperature and on the peak l
suppression pool temperature.
In addition, an increase in l
the initial SRV opening setpoint pressure due to SRV drift will have negligible effect on the peak drywell temperature for small line breaks.
6.
The increase in the SRV opening setpoint tolerances will not-result in the exceedance of allowable stresses in the containment structures.
Page 3 of 4 l
4 8
Documnnt Control Dask LR-N97117 Attachmnnt 1 LCR H97-10 The results for emergency core cooling system /LOCA performance, l
high pressure system performance, containment response, and ATWS l
mitigation are cycle independent.
The vessel overpressurization analysis and fuel thermal limits evaluation are cycle dependent.
As stated previously, the above analyses were performed for Cycle 7 (for normal and abnormal operating modes in the current operating cycle).
The vessel overpressurization and thermal margin determinations will be re performed for succeeding cycles as part of the required reload licensing evaluations.
These roload licensing evaluations will use the TS SRV setpoint tolerance limits proposed in this submittal.
The proposed changes to Surveillance Requirement 4.4.2.2 are being made to ensure that the SRVs are returned to service within the i1% limits assumed by the analyses justifying the change to i3% SRV setpoint tolerance limits.
This requirement is consistent with the current TS requirements for SRV operability.
PSE&G believes that the above conclusions Justify the TS changes being proposed in this submittal.
The proposed changes and analyses conducted have been performed in accordance with the recommendations contained in the NRC's 3/8/93 SER and will not challenge the systems, structures and components from performing their safety functions.
CONCLUSIONS:
The changes proposed in this request are being made to implement more appropriate SRV setpoint tolerance limits.
PSE&G has concluded that these proposed changes are adequately justified and result in No Significant Hazards Considerations as described in Attachment 2 of this letter.
Page 4 of 4
Document Control Desk LR-N97117 LCR H97-10 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 SAFETY RELIEF VALVE SETPOINT TOLERANCES 10CFR50.32 EVALUATION Public Service Electric & Gas (PSE&G)
Technical Specifications do not involveproposed changes to the the ation (HC) consideration.
In support of this determination,a significant hazards of each of the three standards set forth in 10CFR50 an evaluation provided below.
.92 is REQUESTED CHANGE The proposed revisions to Technical Specifi replace the setpoint tolerance limit.1% setpoint tolerance limit cation LCO 3.4.2.1 In addition, for SRVs with a 3%
4.4.2.2 is being revised to state that all SRVSurveillance Requirement to be within 1% of the TS setpoint prior to returning ths must be certified to service.
e valves BASIS 1.
the probability or consequences of an aThe proposed changes d cant increase in eval ua t ed.
ccident previously The proposed TS revisions 2) involve:
1) changes; systems or components in normal or accident opno significant c and 3) no changes to existing structures, erating conditions; Therefore these changes will not systems or components.
accident previously evaluated.
increase the probability of an Safety Evaluation Report, developed in accordance with the provisionT dated 3/8/93, s contained in an NRC Topical Report", Inservice Pressure Relief Technical Spe ififor the "BWR Owners c
NEDC-31753P as described in General Electcation Licensing report NEDC-32511P, "Safet Valve Tolerance Analyses".y Review for Hope Creek Safety / Relief ric all applicable design basis requirementswith these proposed Since the plant systems associated pable of: 1) meeting capability to mitigate the consequences o;f accid and 2) retain the the HC UFSAR, Therefore,the proposed changes were determined to bents described in justified.
these changes will not involve a e Page 1 of 2
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l Documsnt Control Desk LR-N97117 LCR H97-10 HOPE CREEK GENERATING STATION EACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 SAFETY RELIEF VALVE SETPOINT TOLERANCES l
10CFR; 92 EVALUATION
[
Public Service Electric & Gas (PSE&G) has concluded that the l
proposed changes to the Hope Creek Generating 3tation (HC)
Technical Specifications do not involve a significant hazards consideration.
In support of this determination, an evaluation of each of the three standards set forth in 10CFR50.92 is provided below, i
REQUESTED CHANGE i
The proposed revisions to Technical Specification LCO 3.4.2.1 replace the 1% setpoint tolerance limit for SRVs with a 3%
setpoint tolerance limit.
In addi tion, Surveillance Requirement 4.4.2.2 is being revised to state that all SRVs must be certified to be within 1% of the TS setpoint prior to returning the valves to service.
l BASIS 1.
The proposed changes do not involve a significant increase in the probability or consequences of an accident previously eval ua ted.
l The proposed TS revisions involve:
- 1) no significant hardware l
changes; 2) no significant changes to the operation of any systems or components in normal or accident operating conditions; and 3) no changes to existing structures, systems or components.
l Therefore these changes will not increase the probability of an accident previously evaluated.
These proposed changes were developed in accordance with the provisions contained in an NRC Safety Evaluation Report, dated 3/8/93, for the "BWR owners Group Inservice Pressure Relief Technical Specification Licensing j
Topical Report", NEDC-31753P as described in General Electric report NEDC-32511P, " Safety Review for Hope Creek Safety / Relief Valve Tolerance Analyses".
Since the plant systems associated with these proposed changes will still be capable of: 1) meeting all applicable design basis requirements; and 2) retain the capability to mitigate the consequences of accidents described in the HC UFSAR, the proposed changes were determined to be justified.
Therefore, these changes will not involve a Page 1 of 2
Document Control Desk LR-N07117 LCR H97-10 significant increase in the consequences of an accident previously evaluated.
2.
The proposed change does not create the possibility of a new or different kind of accident from any accident previously e val ua t ed.
Establishment of the 3% SRV setpoint tolerance limit will not adversely impact the operation of any safety related component or equipment.
Since the proposed changes involve:
- 1) no significant hardware changes; 2) no significant changes to the operation of any systems or components; and 3) no changes to existing structures, systems or components, there can be no impact on the occurrence of any accident.
These proposed changes were developed in accordance with the provisions contained in an NRC Safety Evaluation Report, dated 3/8/93, for the "BWR Ownere Group Inservice Pressure Relief Technical Specification Licensing Topical Report", NEDC-31753P as described in General Electric report NEDC-32511P, " Safety / Relief Valve Tolerance Analyses".
Furthermore, there is no change in plant testing proposed in this change reques which could initiate an event.
Therefore, these changes wi.
'ot create the possibility of a new or different kind of accluent from any accident previously evaluated.
3.
The proposed change does not involve a significant reduction in a margin of safety.
i 4
Establishment of the i3% SRV setpoint tolerance limit will not l
3 adversely impact the operation of any safety related component or equipment.
General Electric analyses performed for Hope Creek j
and contained in General Electric report NEDC-32511P,
" Safety / Relief Valve Tolerance Analyses," concluded that there is no significant impact on fuel thermal 1.imits, no significant impact on safety related systems, structures or components, and i
no significant impact on the accident analyses associated with j
the proposed changes.
Therefore, the changes contained in this request do not result in a significant reduction in a margin of j
safety.
L 1
CONCLUSION Based on the above, PSE&G has determined that the proposed changes do not involve a significant hazards consideration.
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