ML20205L874

Safety Evaluation Accepting Util 831110,840508 & 0607 Responses to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1,3.2.2 & 4.5.1

ML20205L874
Person / Time
Site:	Limerick
Issue date:	03/26/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20205L822	List: ML20205L818 ML20205L874
References
GL-83-28, NUDOCS 8704020212
Download: ML20205L874 (5)

Text

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1. o g UNITED STATES

-  ! g NUCLEAR REGULATORY COMMISSION 7f  : E WASHINGTON, D. C. 20555

...../ SAFETY EVALUATION RELATED TO GENERIC LETTER 83-28, ITEMS 3.1.1, 3.1.2, 3.2.1. 3.2.2, AND 4.5.1 PPILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION, UNIT 1 DOCKET NO. 50-352

1. Introduction On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal from the reactor protection system. This incident occurred during the plant startup, and the reactor was tripped manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers has been determined to be related to the sticking of the undervoltage trip attachment. Prior to this incident, on February 22, 1983, at Unit 1 of the Salem Nuclear Power Plant, an automatic trip signal was generated due to a steam generator low-low level during plant startup. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip.

Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EDO), directed the staff to investigate and report on the generic implications of these occurrences at Unit .1 of the Salem Nuclear Power Plant. The results of the staff's inquiry into the generic implications of the Salem I incidents are reported in NUREG-1000,

" Generic Implications of ATWS Events at the Salem Nuclear Power Plant."

As a result of this investigation, the Director, Division of Licensing, Office of Nuclear Reactor Regulation requested (by Generic Letter 83-28 dated July 8,1983) all licensees of operating reactors, applicants for an operating license, and holders of construction permits to respond to certain generic concerns. These concerns are categorized into four areas:

(1) Post-Trip Review, (2) Equipment Classification and Vendor Interface, (3) Post-Maintenance Testing, and (4) Reactor Trip System (RTS) Reliabil-ity Improvements. Within each of these areas, various specific actions were delineated.

This safety evaluation (SE) addresses the following actions of Generic Letter 83-28:

3.1.1 and 3.1.2, Post-Maintenance Testing (Reactor Trip System Components) 3.2.1 and 3.2.2, Post-Maintenance Testing (All Other Safety-Related Components) 4.5.1, Reactor Trip System Reliability (System Functional Testing) 8704020212 DR 870326 ADOCK 05000352 PDR

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By letters dated November 10, 1983, Ma 1984 and June 7, 1984, Philadelphia Electric Compa ny (Licensee)y 8, described their planned and completed actions regarding the above items for the Limerick Generating .

Station, Unit 1.

2. Evaluation 2.1 General Generic Letter 83-28 included various NRC staff positions regarding the specific actions to be taken by operating reactor licensees and operating license applicants. The Generic Letter 83-28 positions and discussions of licensee compliance regarding Actions 3.1.1, 3.1.2, 3.2.1, 3.2.2 and 4.5.1 for Limerick are presented in the sections that follow.

2.2 Actions 3.1.1 and 3.1.2, Post-Maintenance Testina (Reactor Trip System Components)

Position Licensees and applicants shall submit the results of their review of test and maintenance procedures and Technical Specifications to assure that in components post-maintenance the reactor trip operability system (testing RTS) isofrequired safety-related to be i conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.

Licensees and applicants shall submit the results of their check of vendor and engineerin recommendations (regarding safety-related components in the RTS) gto ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications, where required.

Actions 3.2.1 and 3.2.2, Post-Maintenance Testino (All Other Safety-Related Components) i Position Licensees and applicants shall submit a report documenting the extending of test and maintenance procedures and Technical Specif-ications review to assure that post-maintenance operability testing of all safety-related equipment is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.

4 Licensees and applicants shall submit the results of their check of vendor and engineering recommendations (all other safety-related components) to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifica-tions, where required.

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, Discussicn i By a letter dated November 10. 1983, in response to Item 3.1.1, the licensee

described the use of an Operation Verification Fonn that specifies all testing requirements along with acceptance criteria and actions necessary

, to assure; that equipment has been returned to fully operable status fol-lowing preventive and corrective maintenance. Also documented or referenced on the form are all testing performed, verification methodology and test results. Approved administrative procedures govern these actions and requirements. The licensee also stated that preventive and corrective maintenance activities would be incorporated into a Computerized History

>j and Paintenance Planning System (CFAMPS) by January 1, 1985. The'adminis-trative controls and procedures assures that post maintenance operability testing is required to be conducted for all safety related components in the reactor trip system.

2 Also described in the subject letter in response to Item 3.1.2 were those

actions and controls implemented by the Architect-Engineer for the Reactor Trip System including a document control system that assured the latest vendor technical information had been included in preoperational testing procedures, engineering drawings and vendor manuals. The licensee stated that these actions and controls assure plant operations would begin with the most recent vendor and engineering reconnendations for testing.

The updating of appropriate documents during operations is discussed in

the licensee's response to Items 2.1 and 2.2 of the generic letter.

The licensee's letter dated November 10, 1983, in response to Items 3.2.1

, and 3.2.2, further stated that post-maintenance testing and vendor and engineering recommendations for all other safety related components are.

reviewed, evaluated, conducted and implemented in precisely the same way as for the reactor protection system discussed in paragraph 2.2.  ;

Further, License Condition (8) required the licensee to implement its

, connitments applicable to Generic Letter 83-28 on a schedule consistent i with that given in their letters dated November.10, 1983, May 8, 1984 and l June 7, 1984

, Durino a routine staff inspection (Report No. 50-352/85-47) it was verified l that the licensee met License Condition (8), that post maintenance testing was conducted as described.in the licensee's letter and that the preventive and corrective maintenance activities had been incorporated into CHAMPS.

Also verified was the implementation of Vendor Technical Information Controls for all- safety related items using the Nuclear Utilities Task Action Com-mittee (NUTAC) publication, " Vendor Equipment Technical Information Program,"

j February 1984, on Generic Letter 83-28.

1 l' Based on the above, the staff concludes that the licensee has complied with the NRC staff pocitions for Actions 3.1.1 and 3.1.2, 3.2.1 and 3.2.2 of Generic Letter 83-28.

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2.4 Action 4.5.1. Reactor Trip System Reliability (System Functional Testina)

Position l On-line functional testing of the reactor trip system, including independent testing of the diverse trip features shall be performed on all plants. The diverse trip features to be tested include the breaker undervoltage and shunt trip features on Westinghouse, B&W and CE plants; the circuitry used for_ power interruption with the silicon controlled rectifiers on B&W plants; and the scram pilot valve and backup scram valves (including all initiating circuitry) on GE plants.

Discussion By letters dated November 10, 1983, May 8, 1984 and June 7, 1984, the licensee described their surveillance testing of the reactor protection system includina the scram pilot valves and backup scram j valves. License Condition '(8) required the licensee to implement its commitments applicable to the generic letter on a schedule consistent with that given in the above referenced letters.

The licensee stated, in their May 8, 1984 letter that the Reactor

, Protection System can be tested during operation by an overlapping i series of tests. The scram pilot valves are tested as follows.

(1) The manual scram trip channel test verifies the ability to de-energize the channel's scram pilot valve solenoid without a reactor scram by using the manual scram push button switches.

Each channel is independently tested.

(2) The single rod scram test verifies the capability of each rod l to scram using the toggle switches on- the hydraulic control unit.

With respect to the backup scram valves the licensee stated that  !

, on-line testing will not be conducted as this would result in a  !

i 1 reactor scram. In lieu of on-line testing the licensee stated that the backup scram valves will be tested during each refueling outage i to avoid spurious full - scram. The NRC staff finds the licensee position not to test the backup scram valves on line acceptable with the existing design as on-line testing would result in unnecessary reactor scrams and challenges to the Reactor Protection System.

, Table 4.3.1.1-1 of the facility Technical Specifications (TS) requires that the manual scram be tested monthly. Additionally, TS 4.1.3.2 requires single rod scram testing once per 120 days of power operation.

The licensee's intent to conduct testing of reactor protection system components at a frequency indicated in the plant Technical Specif-ications and the above referenced letters was verified durina~ a December, 1985, staff inspection (Report No. 50-352/85-47).

Rased on the above the staff concludes that the licensee's actions are consistent with the NRC staff pcsition for Action 4.5.1 of Generic Letter 83-28.

3. Conclusion Based upon the foregofng discussions, the staff concludes that the licensee has complied with Actions 3.1.1, 3.1.2, 3.2.1, 3.2.2 and 4.5.1 of Generic Letter 83-28.

Dated:

Principal Contributor:

4 G. Napuda, Division of Reactor Safety, Region I I

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