ML20041F423
| ML20041F423 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 03/08/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20041F411 | List: |
| References | |
| NUDOCS 8203160486 | |
| Download: ML20041F423 (4) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 51 TO LICENSE NO. DPR-6 CONSUMERS POWER COMPANY BIG ROCK POINT PLANT DOCKET NO. 50-155
1.0 INTRODUCTION
l By letter dated October C7,1981, Consumers Power Company (CPC) (the licensee) requested changes to the Technical Specifications uppended
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to Facility Operating License No. DPR-6 for Big Rock Point Plant.
i The proposed changes institute performance and testing requirements on the under-voltage (degraded grid voltage) protection system for class lE power equipment.
By letter dated October 3,1980, CPC requested changes to the Technical Specifications appended to Facility Operating License No. DPR-6 for Big Rock Point Plant.
The proposed changes institute surveillance recuirements on the scram discharoe system to ensure proper operation on 4'
the system in the event of a scram. We have discussed these changes with CPC and made certain revisions to the proposed changes.
These j
revisions were agreed to by CPC.
Fir. ally, we have agreed with CPC to rectify a wording error which occurred curing the preparation of a correction to Amendment No. 49.
2.0 EVALUATION T.i Deoraded Grid Protection i
Under contract to ths NRC, EG&G performed a detailed technical review and eyaluation of the proposed modifications for degraded gr'id protection and the associated changes to the Technical Specifications. The enclosed technical evaluation report, EGG-EA-5374, Rev.1, " Degraded Grid Protection for Class IE Power Systems," ( Attachment 1) was prepared by EG&G to describe their review'and evaluation. We have reviewed this technical evaluation report and concur with it.
This technical evaluation provides the principle basis for our evaluation as described below.
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2 The criteria used by EG&G in its technical evaluation of the proposed changes include GDC-17 (" Electric Power Systems") of Appendix A to 10 CFR 50; IEEE Standard 279-1971 (" Criteria for Protection Systems for Nuclear Power Generating Stations"); IEEE Standard 308-1977 (" Class lE Power Systems for Nuclear Power Generating Stations"); ANSI Standard C84.lE (" Voltage Ratings for Electrical Power Systems and Equipment -
60 Hz"); and staff positions defined in NRC Generic Letter to CPC dated June 3,1977.
The following electrical system design modifications and Technical Specification changes were proposed by CPC.
cond level of undervoltage relays on Installation of a seThe second level undervoltage relaying A.
the 2400 volt bus.
will pr, ovide a three-out-of-three coincident 'Ogic scheme.
When the voltage is below 89.25% (of 2400 V) for approxi-mately 0.6 second the second level undervoltage relays will trip. The coincident signal is then fed through a single.
time delay relay set at 10 (+ 0.5) sec.
This logic will trip the supply breaker to tee 2400 volt bus.. De-energizing the 2400 volt bus will in turn trip the existing first level undervoltage relays (set at 50%) on the 480 volt class 1E bus. The first level undervoltage relays will provide the start signal to the diesel generator. When the voltage output of the diesel generator exceeds 91%, the diesel gene-rator breaker will close and the Class lE load will-be block loaded on the Class lE 4S0 volt bus.
Big Rock Point does not load shed or sequence safety related loads on the class.lE bus.
The licensee has proposed to install the second level of undervoltage relays on the 2400 volt bus which is not a Class lE system. The justification for selecting this vol-tage level is to provide protection for plant equipment l
al all voltage levels. This plant was built prior to the l
advent of Class lE and has only one emergercy 480V bus l
(2B) which supplies an emergency fire pump and supporting equipment for core cooling.
The second level undervoltage relays meet or exceed the requirements of IEEE 279-1971 and the breaker it trips (offsite power supply to 2400 volt l
switchgear) is on the plant Q list and is periodically l
tested.
We therefore find the licensee's proposed design acceptable.
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t B.. 'The staff requirement to block load shedding while the diesel
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generator is supplying the emergency load and reinstatement of this feature on a diesel generator trip is not applicable to Big Rock Point since diesel generator loads are block loaded on the Class lE bus and load shedding feature is not incorporated into the design.
C.
Additions and changes to plant Technical Specifications including the surveillance requirements, allowable limits for the setpoint and time delay, and limiting conditions for operation have been provided by the licensee. These additions and changes have been reviewed and found acceptable.
2.2 Scram Discharce System Under contract to the NRC, Franklin Research Center (FRC) performed a detailed technical review and evaluation of the proposed changes to the Technical Specifications. The enclosed technical evaluation report, TER-C-5506-56," BWR Scram Discharge Volume Long-Term Modifications," (Attachment 2) was prepared by FRC to describe their review and evaluation. We have reviewed this technical evaluation report and concur with it.
This technical evaluation report provides i
the principal basis for our evaluation as described below.
The licensee's response does not meet the explicit requirements of Paragraph 3.3-6 and Table 3.3.6-1 of the NRC's Model Technical Specifications.
However, Big Rock Point has a unique scram discharge.
system for which the NRC model Technical Specifications are not directly applicable.
In discussions with the NRC, CPC'did agree to i
institute a quarterly surveillance requirement to fully cycle the scram discharge volume vent and drain valves.
Inis surveillance requirement was not in the changes originally proposed by CPC.
Therefore, we conclude that the proposed changes as revised meet the NRC criteria on which the model Technical Specifications were based and are therefore, acceptable.
2.3 Wordine Correction In Amen'dment 49, as corrected November 24, 1981 a change to Section 7.6 was improperly ' worded. The proper frequency of routine tests for the High Radiation trip-closure of the Containment Ventilation Isolation Valves should be "at each major refueling shutdown." 'The improper words were "At each Major Frequency Shutdown."
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3.0 ENVIRONMENTAL CONSIDERATION
We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not. result in any significant environmental impact. Having made this determination, we.have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
4.0 C0!!CLUSIONS We have concluded, based on the considerations discussed above, that:
(1) because the amendment d'oes not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, thi amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be con-ducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Attachments:
As stated (2)
Date:
March 8, 1982
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