ML20079J182
| ML20079J182 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 10/07/1991 |
| From: | J. J. Barton GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| C321-91-2259, NUDOCS 9110150297 | |
| Download: ML20079J182 (5) | |
Text
.
GPU lluclear Corporadon lliO Nuclear
- ees:r388 Forked River, New Jersey 08731-o389 Go9 971-400o Writer s Direct Dial Nurnber.
October 7,1991 C321-91-2259 U. S. Nuclear Regulatory Commission Att: Document Control Desk Washington, DC 20555
Dear Sir:
Subject:
Oyster Creek Nuclear Generating Station Facility License No. DPR-16, Docket No. 50-219 Response to Safety Evaluation Report -
Station Blackout Analysis By letters dated April 17, 1989 and March 30, 1990, GPU Nuclear (GPUN) submitted the Oyster Creek Nuclear Generating Station (OCNGS) response and supplemental response to the Station Blackout SBO) rule (10 CFR 50.63).
The NRC reviewed the submittals and formally respon(ded via the subject Safety Evaluation Report (SER) received on August 28,1991.
The SER finds the OCNGS submittals to be in conformance with 10 CFR 50.63 contingent upon GPUN implementing various recommendations identified therein.
The enclosed attachment provides our response to each recommendation.
With respect to plant modifications, GPUN had previously indicated that the connection of the combustion turbines (CT) to the 4.16KVAC Emergency Bus IC or ID would be accomplished by disconnecting the start-up transformer SA or SB via a new disconnect switch, and feeding 4.16KVAC Bus IA or IB from either of the two cts.
The proposed modification will no longer utilize a disconnect switch to isolate the start-up transformer.
Instead, the Alternate AC (AAC) power supply will be routed from the new step-down SB0 transformer to a normally open 4.16KV Bus IA circuit breaker. The
' art-up transformer will be isolated via its existing normally open 4.16KV circuit breaker.
This change provides the same electrical isolation capabil!ty as the original design, while eliminating unnecessary hardware.
9110150297 911007 FDR ADOCK 0500 9
p
.L A U U O : >
GPU Nuclear Corporation is a subsdary of General Pubhc Utihties Corporation
Page 2 The modifications, associated procedure changes, and supporting documentation for SB0 are expected to be completed by the end of refueling outage 14R, which is currently scheduled for the fall of 1992.
Sincerely, h
o 3.J.Barton Vice President and Director, Oyster Creek JJB/EP/ pip Attachment cc:
Administrator, Region 1 Senior NRC Resident inspector Oyster Creek NRC Project Manager e
ATTACHMF,NL1 N
PROPOSED ALTERNATE AC POWER SOURCE Staff Recommendation:
The licensee should take positive steps to improve the combustion turbine sy"em reliability from the present value of 0.93 to the minimum target value 01 0.95.
Response: The reliability value previously cited was based on early, short-term operational history of the combustion turbines, which were installed in 1989.
Positive steps taken with respect to additional data collecting and review indicate that we can achieve and maintain an aggregate alternate AC source reliability of 0.95.
EFFECTS OF LOSS OF VENTILATION Staff Recommendation:
The licensee should use an initial temperature for the SB0 control room heat-up calculation no lower than that allowed by the Technical Specifications or the adminirtrative procedures.
The licensee should verify and confirm that the equipment in the inverter rooms and other areas which have heat generation sources are qualified for the expected heat-up that would occur in these areas during the first hour before the AAC power source is available.
Further, the licensee should include all analyses and related information in the supporting documentation that is to be maintained by the licensee for staff review.
Response: As indicated in our previous supplemental res)onse to SB0 dated March 30, 1990, the effects on the control room due to tie loss of the ventilation system were the sub:ect of the control room heat up analysis.
This analysis demonstrated that the temperature inside the control room will remain below acceptable limits (120'F) following loss of the ventilation system.
In this analysis, the initial control room temperature was assumed to be 75*F and the heat generation rates based on plant operation at 100%
power with all attendant control room instrumentation functioning. At the end of four (4) hours with.ut ventilation the average control room temperature was calculated to be 106.3 F.
A four (4) hour period is the OCNGS proposed station blackout duration.
Subsequent to performing the heat up analysis, a control room loss of ventilation test was conducted unuer Test Procedure TP 254/13 MTX 26.12.2.6.
This test confirmed the conservatism of the analytical method used in the heat up analysis.
The test was conducted during November 1989 with the plant operating at full power.
The cabinet doors and the control room doors were closed during the test.
Area temperature readings were taken at 5'-6" above the control room floor.
Also, temperature readings were taken on top of some cabinets.
No auxiliary ventilation was provided. The initial average control room temperature was 73.5 F.
The control room design temperature (thermostatically controlled) is 75'F 5'F (FSAR 9.4.1.2).
The results of this test demonstrated that for a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period, following loss of Heating, Ventilating, and Air Conditioning (HVAC), the control room temperature remained below 84.2'F (average room temperature). The maximum temperature experienced in the control room was 87.5*F.
No abnormalities or instrumentation drifting were observed during the loss of ventilation test.
\\
'At,tachment page 2 Following an SB0 event, power is expected to be restored within a one (1) hour period. According to test results during the first one (1) hour period following loss of HVAC, the average temperature in the control room was less than 79.2*F, far below maximum allowable limits. The maximum temperature experienced during this one (1) hour period was 82*F.
This test was conducted with the plant at 100% power.
During an SB0 event, the control room heat generation rate is expected to be less than that during 100% power operation since less AC equipment would be able to operate.
The Oyster Creek Technical Specifications do not specify a control room temperature limit.
Station Procedure 331.1 (R.3) indicates a 75'F 5'F set point for the local thermostat.
Based on these test results and based on the conservatism associated with it, we do not conside* it necessary to incorporate any limits into the administrative procedures to maintain the Control Room temperature at or below the 75'F value used in the Control Room Heat Up Analysis or require opening of control room cabinet doors.
Additionally, temperature effects due to heat generated from the inverters in the A&B battery room will be confirmed. Other similar areas will also be reviewed.
CONTAINMENT ISOLATION Staff Recommendation: The licensee needs to list and address in an appropriate procedure (s) the containment isolation valves (CIVs) which are normally closed and the CIVs which are normally open and which fail as-is upon loss of AC power and cannot be excluded by the criteria given in Regulatory Guide (RG) 1.155. The procedure needs to identify the actions necessary to ensure that the valves are fully closed, if needed. The staff's position is that the valve closure needs to be confirmed by position indication (local, mechanical, remote process information, etc.). The licensee should include the abova information in the SB0 supporting documentation that is to be maintained by the licensee for the staff's review.
Resoonse: Containment isolation requirements were reviewed as part of the Oyster Creek coping assessment to ensure adequate containment integrity during an SB0 event.
This review is in accordance with the criteria in RG 1.155, Section 3.2.7.
The review identified several containment isolation valves that may be in the open position at the onset of an SBO.
For these valves, position indication confirmation and/or manual operation, as applicable, will be procedurally implemented, unless satisfactory engineering justification determines it is not necessary.
' Attachment
'Page 3 PROPOSED MODIFICATIONS Staff Recommendation: The licensee should iiclude sufficient technical information on the proposed modifications in the SB0 supporting documentation that is to be maintained by the licensee for staff review.
Response
Plant modifications in support of SB0 rule compliance will be documented.
Such documentation will contain sufficient technical information describing the modification and, will be available for staff review.
QUALITY ASSURANCE AND TECHNICAL SPECIFICATIONS Staff Recommendation: The licensee should verify that the SB0 equipment is covered by an appropriate QA program consistent with the guidance of RG 1.155.
This evaluation should be documented as part of the documentation supporting the SB0 rule response.
Response: SB0 equipment will be classified and, as applicable, included in an appropriately graded quality QA program. The quality program will be consistent with the applicable guidance in RG 1.155.
EMERGENCY DIESEL GENERATOR (EDG) RELIABILITY PROGRAM Staff Recommendation: The licensee should provide confirmation and include in the documentation supporting the SB0 package that is to be maintained by the licensee that an EDG reliability program meeting the guidance of RG 1.155, Position 1.2, is in place or will be implemented.
Response: The April 17, 1989 submittal indicated that a target EDG reliability of 0.975 was selected for the OCNGS, consistent with NUMARC 87-00, Section 3.2.4.
The existing, proceduralized surveillance and maintenance program including but not limited to deviation reporting, graded root-cause analysis, independent reviews (oversight),- etc., is believed to be consistent with the intent of RG 1.155, Section 1.2.
This enables maintaining a 0.975 EDG target reliability.
It is our understanding that the NRC staff is preparing a performance-based rule and regulatory guide which will provide licensees guidance for their own individual EDG reliability program.
Upon issuance, we will review the new rule and regulatory guide, and determine if any cleages to the existing program are necessary.