ML20237D955

TS Change Request 255 to License DPR-16,removing Requirement for ADS Function of EMRV to Be Operable During Rv Pressure Testing & Correcting Note H of Table 3.1.1 as Noted

ML20237D955
Person / Time
Site:	Oyster Creek
Issue date:	08/21/1998
From:	Roche M GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20237D953	List: ML20237D949 ML20237D955 ML20237D959
References
NUDOCS 9808280049
Download: ML20237D955 (6)

Text

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GPU NUCLEAR CORPORATION OYS13R CREEK NUCLEAR GENERATING STATION Facility Operating License No. DPR-16 .

l Technical Specification Change Request No. 255 Docket No. 50-219 Applicant submits, by this Technical Specification Change Request No. 255, to the Oyster Creek Nuclear Generating Station Operating License, a change to pages 3.1-16,3.4-3 and 3.4-4.

BY ) f Michael B. Roche Vice President and Director Oyster Creek Sworn and Subscribed to before me this hef day of MHgA8 .1998. i ht4LAau f /t< n l l A Notary Public of NJ '

GERALDINE E. LEVIN NOTARyMBLEOFMN g w p,. . p.. po.m u,- ~r++ ,

l C 9908290049 990821

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISS 0N In the Matter of. )

Docket No. 50-219 GPU Nuclear, Inc. )

CERTIFICATE OF SERVICE This is to certify that a copy of Technical Specification Change Request No. 255, for Oyster Creek Nuclear Generating Station Operating License, filed with the U.S. Nuclear Regulatory Commission on August 21, 1998 has this day of August 21, 1998, been served on the Mayor of Lacey Township, Ocean County, New Jersey by deposit in the -

United States mail, addressed as follows:

The Honorable Louis A. Amato I

Mayor of Lacey Township 818 West Lacey Road Forked River, NJ 08731 BY Michael B. Roche Vice President and Director Oyster Creek l

4 I. TECHNICAL SPECIFICATION CHANGE REQUEST (TSCR) No. 255 i .

l GPU Nuclear request the following replacement pages be inserted into existing Technical Specifications:

Replace existing pages 3.1-16,3.4-3 and 3.4-4 with the attached revised replacement pages 3.1-16,'3.4-3 and 3.4-4.

II. REASON FOR CHANGE This change to the Oyster Creek Technical Specifications will remove the requirement for the Automatic Depressurization (ADS) function of the Electromatic Relief Valves (EMRV) to be operable during Reactor Vessel Pressure Testing. Currently section 3.4.B.1 states "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 Specification 1.39 and 3.3.A.(i)."

l Additionally, note h of Table 3.1.1 will be corrected. Note h, was inadvertently changed in Amendment 75 to state " Relief valve controllers shall not be bypassed for

, . more than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (total time for all controllers) in any 30 day period ..." Amendment 44 of the FDSAR, original Technical Specifications, gives the correct bypass time as 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in a 30 day period. The wording of the note is being changed to clarify'what is meant by bypassing a relief valve controller.

III. SAFETY EVALUATION JUSTIFYING CHANGE' i

l The function of ADS as is to depressurize the reactor vessel in the event of a small break Loss of Coolant Accident (LOCA) to allow the~ Core Spray System to inject water l

to cool the core. Technical Specification definition 1.39 defines Reactor Vessel l ,

Pressure Testing as " System pressure testing ... with the reactor vessel pressure completely water solid, core not critical and section 3.2.A satisfied." Additionally, l Technical Specification section 3.3.A.(i) limits temperature to 250*F. Therefore,

! should a small break LOCA occur as reactor level began to lower some flashing would be expected to occur but the resulting pressure would be limited to the saturation pressure at 250 F, which is 29.8 psia (~ 15.1 psig). This would result in vessel pressure being rapidly reduced to well below 110 psig without ADS. Therefore, bypassing ADS during reactor vessel pressure testing would not change accident mitigation functions.

Amendment 120 of the OC Technical Specifications changed the Reactor Vessel Pressure / Temperature Limits and allowed temperatures to exceed 212*F during Reactor Vessel Pressure Testing. As part of this amendment a statement was added to the Primary Containment section of Technical Specifications stating that Primary Containment is not required during vessel pressure testing. As Primary Containment is not required, it would not be possible to receive an ADS actuation signal. High L

drywell pressure is a required signal for ADS and with the drywell hatch open this

. signal could not be generated. This condition was recognized and found acceptable with the issuance of Technical Specification Amendment 120. '

Correcting note h of Table 3.1.1 would not affect the function of ADS as described above. The original OC Technical Specifications included a requirement to limit bypassing the relief function of the EMRVs to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in a 30-day period.

' Subsequently, amendment 75 inadvertently changed the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> due to a typographical error. The licensing basis for the plant is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> versus the 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> stated in the existing note in the Technical Specifications. The statement " Relief valve controllers shall not be bypassed..." has led to some confusion because the EMRV control switches do not have a ." Bypass" position. The proposed change replaces terminology not used in reference to the EMRVs with the appropriate terms. The wording change is only a clarification and does not change the intent of the Technical Specification.

IV. NO SIGNIFICANT HAZARDS CONSIDERATION GPU Nuclear has determined that this TSCR poses no significant hazard as defined by 10 CFR 50.92.

1. State the basis for the determination that the proposed activity will or will not increase the probability of occurrence or consequences of an accident.

As the ADS is not required to mitigate a LOCA during reactor vessel pressure

. testing and this change will not affect the integrity of the reactor pressure vessel, bypassing the ADS during vessel pressure testing will not affect the probability of occurrence or the consequences of an accident previously evaluated in the SAR Correcting the allowed out of service time for the relief function of the EMRVs does not impact any of the accidents previously evaluated by the SAR.

2. State the basis for the determination that the activity does or does not create the possibility of an accident or malfunction of a different type than any previously identified in the SAR.

This change does not change the ADS system or affect its function; therefore, it does not create the possibility for an accident or malfunction of a different type than previously identified in the SAR.

3. State the basis for the determination that the margin of safety is not reduced.

The effect of the unavailability of Primary Containment has been previously analyzed for Amendment 120 to the Technical Specifications. This analysis may be applied to bypassing ADS since Primary Containment is required for ADS to initiate. Therefore, the Margin of Safety is not reduced by this change.

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. I This Technical Specification change reestablishes the out of service time to the value originally established in Amendment 44.

I V. IMPLEMENTATION GPU Nuclear requests that the amendment authorizing this change be effective upon issuance.

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Revised Technical Specification Pages

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