ML20065M997

TS Change Request 213 to License DPR-16,updating & Clarifying TS 3.4.B.1 to Be Consistent W/Existing TS 1.39 & 4.3.D Re Five Electromatic Relief Valves Pressure Relief Function Inoperable During Pressure Testing

ML20065M997
Person / Time
Site:	Oyster Creek
Issue date:	04/19/1994
From:	GENERAL PUBLIC UTILITIES CORP.
To:
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ML20065M989	List: ML20065M986 ML20065M997 ML20065M999
References
NUDOCS 9404260309
Download: ML20065M997 (5)

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1 GPU NUCLEAR CORPORATION OYSTER CREEK NUCLEAR GENERATING STATION facility Operating License No. DPR-16 Technical Specification Change Request Request No. 213 Docket No. 50-219 y 1

Applicant submits, by this Technical Specification Change Request No.

213 to the Oyster Creek Nuclear Generating Station Operating License, a change to page 3.4 4.

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By Y J fJ. J ll@ ton iVic P /sident and Director er Creek Sworn and Subscribed to before me this ' lh ay of 994.

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'tary Public of NJ

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JUDITH M.CROWE Neery Public of N Je MyCommissionExpires U 8 #

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0YSTER CREEK NUCLEAR GENERATING STATION FACILITY OPERATING LICENSE NO. OPR-16

-DOCKET NO. 50-219 TECHNICAL SPECIFICATION CHANGE REQUEST NO. 213 Reactor Pressure Vessel Testina - EMRV Bvoass Applicant hereby requests the Commission to change Appendix A to the above captioned license as indicated below, and pursuant to 10 CFR 50.92, an analysis concerning the determination of no significant hazards considerations is also presented:

1.0 .SffTLQN TO BE CHANGJJ Section 3.4.0.1.

2.0 EXTENT OF CHANGE Update and clarify Technical Specification 3.4.B.1 to be consistent with existing Specifications 1.39 and 4.3.0 (ASME Code Section XI, Article 5000 requirements).

The requested change deletes reference to the ASME Code Section XI, IS-500 ten year hydrotest inspection interval and replaces this with references to (1) the Technical Specification 1.39 definition for Reactor Vessel Pressure Testing and (2) the Technical ?9ecification 3.3. A.(i) Reactor Vessel Pressure Testing limits (P/T and 250 F maximum ,

test temperature). l The requested change will clarify that the five electromatic relief l valves' (EMRV) pressure relief function may be inoperable or bypassed I 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.

3.0 [J1ANGES REQUESIfQ The requested change is shown on attached Technical Specification page 3.4-4.

9 4.0 DISCUSSION l During the 14R Cutage, a question arose regarding the interpretation of Technical Specification 3.4.B.l. The current Technical Specification is as follows, "B. Automatic Depressurization System 1

1. Five electromatic relief valves of the L automatic depressurization system shall be l 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 R these valves (but not the automatic  !

depressurization function) may be inoperable -)

or bypassed during the system hydrostatic q pressure test required by ASME Code Section .l XI,1S-500 at or near the end of each ten year inspection interval ."

Upon completion of outage work, the reactor pressure boundary must be-pressure leak tested per ASME Code Section XI lWA-5000 (Technical i Specification 4.3.0) and requires pressurizing up to 1055 +10/-0 psig. .

This is consistent with Technical Specification Amendments 82 and 90, References 6.2 and 6.3. Since the Electromatic Relief Valve (EMRV) setpoint is 1060 psig, the EMRV setpoint must be increased to prevent i possible inadvertent actuation. Bypass of EMRV actuation causes the u automatic pressure relief function of the valve to be inoperable.

Technical Specification 3.4.B.1 requires five EMRVs.to be operable when reactor water temperature is greater than 212 F'and pressurized above 110 psig except that "...during the system hydrostatic pressure test required by ASME Code Section XI, IS-500 at or near the end of each ten l year inspection interval" the automatic pressure relief (not automatic depressurization) function may be inoperable.

The NRC Safety Evaluation for Technical Specification Amendment 44, Reference 6.1, approved EMRV pressure relief bypass for the ten year hydrotest. based on the observation that

...even though the pressure relief function of the j EMRVs is bypassed, over pressure protection would '

continue to be provided by...(the code)... safety valves. Elimination of this relief function does not effect the reactor safety analysis, since credit was i C not taken for the relief function."

At the time the Amendment 44 SER was issued, Oyster Creek had 16. safety valves for over pressure protection. Technical Specification. Amendment -

150, Reference 6.4, approved a reduction of the number of safety valves

atiOysterCreekfrom16to9. The NRC Safety Evaluation for Amendment 150, based on GPU Nuclear analyses, determined that over pressure protection was within the acceptable 1375 psig_ limit when only nine safety valves are assumed to operate with no credit given to (a) the five EMRVs, (b) recirculation pump trip (RPT), or (c) the isolation condensers. This determination was based on the limiting over pressure protection case of main steam isolation valve closure with high flux scram and no RPT. This full power limiting case with no RPT more than bounds the conditions for this TSCR since EMRV bypass for' reactor vessel pressure testing would occur by Technical Specification limitation only when the core is not critical. The other condition evaluated in the Amendment 150 SER was for ATWS response which is not applicable for core not critical conditions under which EMRV bypass for reactor vessel pressure testing wou1J be allowed.

The justification for concluding that the Technical Specification 3.4.B.1 exemption allowing inoperable EMRV pressure relief function during the ten year hydrotest also applies to reactor vessel pressure testing is that (1) reactor pressure vessel leak testing is conducted at a pressure lower than for the hydrotest,-(2) leak testing is similar to hydro testing in nature (i.e., reactor vessel completely water solid, core not critical), (3) the leakage testing is required by Technical Specification 4.3.D, and most importantly (4) two currently valid existing NRC Safety Evaluations (Technical Specification Amendment 44, Reference 6.1 and Technical Specification Amendment 150, Reference 6.4) documented that the code safety pressure relief valves, not the EMRV pressure relief function, is the licensing design basis for the reactor safety analysis.

Reactor Vessel Pressure Testing is currently defined in the Technical Specification 1.39 definition as follows,

" 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 section 3.2.A (Core Reactivity Limits) satisfied."

The proposed revised Technical Specification 3.4.B.1, which invokes this definition, is as follows (requested change shown in bold),

"B. Automatic Depressurization 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. Reactor Vessel Pressure Testing consistent with Specifications 1.39 and 3.3.A.(i)."

5.0, DETERMINATION We have determined that the proposed Technical Specification change involves no significant hazards considerations as discussed below.

1. The requested change will not involve a significant increase in the probability or consequence of any accident previously evaluated .

because this change (a) merely updates and clarifies Technical Specification 3.4.B.1 to be consistent with other existing Technical Specifications, (b) contains no adverse changes to any existing safety function necessary for the reactor vessel solid, core not critical condition, and (c) makes no modification or physical changes to plant equipment, performance or operation necessary to respond to accidents for the reactor vessel solid, core not critical condition.

2. The requested change does not create the possibility of a new or different accident from any accident previously evaluated because this change (a) merely updates and clarifies Technical Specification 3.4.B.1 to be consistent with other existing Technical Specifications, (b) contains no adverse changes to any existing safety function necessary for the reactor vessel solid, core not critical condition, and (c) over pressure protection would continue to be provided by the code safety valves when the EMRV pressure relief function is bypassed.

3. A significant reduction in margin of safety is not involved because even though the EMRV pressure relief function is bypassed, over pressure protection would continue to be provided by the code safety valves. Elimination of this relief function does not affect the reactor safety analysis, since credit was not taken for the EMRV pressure relief function (References 6.1 and 6.4).

6.0 REFERENCES

6.1 NRC Safety Evaluation, Technical Specification Amendment No. 44, " Bypass of EMRV Pressure Relief Function During System Hydrotest," January 4, 1980.

6.2 NRC Safety Evaluation, Technical Specification Amendment No. 82, " Inservice Inspection and Testing," May 22, 1985.

6.3 NRC Safety Evaluation, Technical Specification Amendment No. 90, " Inservice Inspection and Testing," October 18, 1985.

6.4 NRC Safety Evaluation, Technical Specification Amendn:ent l No. 150, " Main Steam Safety Valve Reduction," March 6, 1991.

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