ML20073G100
| ML20073G100 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 09/27/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20073G093 | List: |
| References | |
| NUDOCS 9410040007 | |
| Download: ML20073G100 (5) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION Ia 4E WASHINGTON, D.C. 3088H001
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i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 170 i
TO FACILITY OPERATING LICENSE NO. DPR-16 GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY i
OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219 t
1.0 INTRODUCTION
The GPU Nuclear Corporation (the licensee) letter of April 19, 1994 request for a Technical Specification (TS) change (Ref.1) is to update and clarify TS l
3.4.B.1 to be consistent with TSs 1.39 and 4.3.D.
The Technical Specification Request (TSCR) deletes reference to the ASME Code Section XI,1S-500 ten-year hydrotest inspection interval and replaces this with references to (1) the Technical Specification 1.39 definition for Reactor Vessel Pressure Testing.
i and (2) the Technical Specification 3.3.a(i) Reactor Vessel Pressure Testing limits (P/T 250*F maximum test temperature).
The TSCR clarifies that the five electromatic relief valves (EMRV) pressure relief functicn may be inoperable or bypassed during system pressure testing required by ASME Code Section XI, ARTICLE IWA-5000, including system leakage and hydrostatic test, with reactor vessel completely solid, core not critical and Technical Specification 3.2.A (core reactivity limits) satisfied.
2.0 DISCUSSION Technical Specification 4.3.D addresses ASME Code Section XI Article 5000 i
requirements for pressure leak testing of the reactor pressure boundary follcaing refueling outages or major repairs to the reactor coolant system (RCS). The licensee states that pressure leak testing requires a pressure of 1055 +10/-0 psig, a requirement inconsistent with the EMRV setpoint of 1060 psig (Ref. 1). The inconsistency is similar to one addressed by the staff in i
Amendment 44 (Ref. 2) in which it stated:
The licensee has also proposed to allow the pressure relief function of the electromatic relief valves to be inoperable or bypassed (the ADS [ automatic depressurization system] function of the valves would be maintained) during the system hydrosta?.ic pressure test required by AMSE[ sic]CodeSectionXI,15-500atorneartheendofeachtenyear inspection interval. This allowance is necessary since the hydrostatic test pressure is above the setpoint of the relief valves.
UDR P
. Even though the pressure relief function of the electromatic relief valves is bypassed, over pressure protection would continue to be provided by the 16 [ code] safety valves.
Elimination of this relief function does not affect the reactor safety analyses, since credit was not taken for the relief function. Therefore, we find the modification acceptable.
The staff has been informed that the statement regarding the safety valves was incorrect for the 10 year hydro test since the valves are gagged to prevent weeping (Refs. 3 end 4). Therefore in an SE issued August 29, 1994, the staff concluded that it is acceptable to gag the safety valves during a hydrostatic system test. However, the EMRVs are configured to provide overpressure protection for that test.
In the leak testing being addressed herein, the safety valves are not gagged and the mistake has no impact.
The staff later, in Amendment 150 (Ref. 5). determined that the 16 code safety valves could be reduced to nine on the bai, of analyses that the staff judges to bound shutdown-relued conditions appr0priate to the Reference I request.
(The Standard Review Plan does not address operation of boiling water reactors under the shutdown conditions of concern here.) Although the licensee concludes in its Reference I request that only non-critical conditions are appropriate, the staff considered inadvertent criticality in addition to non-critical conditions. The staff concludes that an inadvertent criticality is unlikely. The licensee assures that a withdrawn rod will not result in criticality prior to the pressurization test (Ref. 6). The moderator temperature coefficient will be negative under the test conditions (Ref. 6) and heatup will not induce criticality.
Even if a criticality were to occur, the doppler coefficient will inhibit the power increase and any void formation would reduce reactivity. The staff is satisfied that the combination of unlikely criticality, limiting characteristics, and design basis analyses means that criticality need not be considered further.
3.0 EVALUATION Bypassing the EMRV actuation (setpoint of 1060 psig) will prevent operation of the automatic pressure relief function, but will have no effect on automatic or operator-actuated depressurization capability provided by the EMRVs.
Reactor coolant system overpressure protection will continue to be provided by the remaining nine code safety valves, four of which have a setpoint of 1212 psig, rather than the EMRV setpoint of 1060 psig (Refs. I and 6).
The test process is to (Ref. 7):
Maintain temperature between 215 *F and 225 *F by operation of reactor recirculation pumps (Sections 4.1.1,6.22.1).
Maintain pressure between 1055 psig and 1065 psig by varying letdown through the reactor water cleanup system and by controlling flow to the control rod drives (CRDs) (Section 4.1.1).
l c Use one CRD pump (Section 4.1.1).
Precautions and limitations include (Ref. 7):
Limit pressurization rate to 150 psig/t.in. (Sections 4.2.1,6.23)
Maintain temperature it 215 *F while at 1055 - 1065 psig and < 240 'F as read at the suction of an operating recirculation pump (Sections 4.2.3, 4.2.5)
Limit reactor coolant temperature increases so that the temperature difference between reactor head metal and recirculation pump suction does not exceed 50 'F (Sections 4.2.6 and 6.22.1).
Other significant Reference 7 Sections include:
6.3.2 "Run at least one recire, pump from the time heat-up to test temperature is commenced until pressure is reduced to atmospheric."
6.22.2
" Soak reactor vessel at test temperature (215-225'F) for one hour prior to initiating pressurization."
The test procedure is consistent with Oyster Creek pressure / temperature limits
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specified in TS Figures 3.3.1.
The staff notes that the test procedure also addresses such potential concerns 1
as:
The only valid measurement of RCS water temperature is obtained via thermocouples located at the recirculation pump inlet. A meaningful 1
indication of RCS temperature is obtained only if there is significant flow in the recirculation pump piping.
Thermal stratification is a concern unless there is significant flow and RV water mixing. This is particularly true for the RV lower plenum and head, where cold water from the CRD could accumulate, with obvious implications to the RV wall temperature requirements of Figures 3.3.1 and to the temperature that would be measured in the pump piping.
There is a time lag between heating of RV water and the response of RV metal temperature.
Failure of the RV metal temperature to be adequately raised would result in a violation of TS Figures 3.3.1.
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4.0 TICHNICAL SPECIFICATION CHANGE j
The change is as follows where the new wording is in bold type:
I 3.4.B. Automatic Deoressurization System 1.
Five electromatic relief valves of the automatic depressurization system shall be operable when the reactor water temperature is greater than 212 "F and pressurized above 110 psig, except as specified in 3.4.B.2.
The automatic pressure relief function of these valves (but not the automatic depressurization function) may be inoperable or bypassed during l
Reactor Vessel Pressure Testing consistent with Specifications 1.39 and 3.3.A.(1).
Based on the staff evaluation above, the staff concludes that the proposed Technical Specification change is acceptable.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendment. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (59 FR 27056). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of i
the amendment.
7.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
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8.0 REFERENCES
1.
Barton, J.
J., "0yster Creek Nuclear Generating Station, Docket No. 50-219, Facility Operating License No. DPR-16, Technical Specification Change Request No. 213, Reactor Pressure Vessel Testing - EMRV (electromatic relief valve) Bypass," Letter to NRC, C321-93-2376, April 19, 1994, i
2.
" Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting j
Amendment No. 44 to Provisional Operating License No. DPR-16, Jersey l
Central Power & Light Company, Oyster Creek Nuclear Generating Station, Docket No. 50-219," NRC, January 4, 1980.
3.
Barton, J.
J., "0yster Creek Nuclear Generating Station, Docket No. 50-219, Facility Operating License No. DPR-16, ASME Code System Hydrostatic Pressure Testing," Letter to NRC from Vice President and Director, Oyster Creek, C321-94-2081, June 3, 1994.
4.
Barton, J.
J., "0yster Creek Nuclear Generating Station, Docket No. 50-219, Ten Year Hydrostatic Test," Letter to NRC from Vice President and Director, Oyster Creek, C321-94-2121, August 1, 1994.
5.
" Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No.150 to Provisional Operating License No. DPR-16, GPU Nuclear Corporation and Jersey Central Power & Light Company, Oyster Creek Nuclear Generating Station, Docket No. 50-219," NRC, March 6, 1991.
6.
Zak, Ron (GPU Nuclear), Telephone call with Warren Lyon and Ronald Hernan (NRC) to confirm staff understanding of licensee operations, August 16, 1994.
7.
" Nuclear Steam Supply System (NSSS) Leak Test," Oyster Creek Nuclear Generating Station Procedure Number 602.4.001 Rev. 19, pp 1 - 25, February 19, 1993. Received from Ron Zak (GPU Nuclear) as the latest version of the applicable procedure, August 16, 1994.
Principal Contributor: Narren Lyons Date:
September 27, 1994
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