ML20136H634

Safety Evaluation Accepting Licensee 851210 Response to Generic Ltr 83-28,Item 4.1 Re Reactor Trip Sys Reliability

ML20136H634
Person / Time
Site:	Fort Calhoun
Issue date:	01/06/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20136H579	List: ML20136H574 ML20136H634
References
GL-83-28, NUDOCS 8601100293
Download: ML20136H634 (2)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. I DOCKET NO. 50-285 GENERICLETTER83-28, ITEM 4.1,REACIORTRIPSYSTEMRELIABILIH I. INTRODUCTION:

On February 24, 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 breaker; nas been detennined to be related to the sticking of the under voltage trip at achment. Prior to this incident, on February 22, 1983, at Unit 1 of the Sale a Nuclear Plant, an automatic trip signal was generated based on 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 (ED0), 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 unit incidents are reported in NUREG-1000,

" Generic Implications of ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Comission (NRC) 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)PostmaintenanceTesting,and(4)ReactorSystemReliability Improvements.

The fourth action item, Reactor Trip System Reliability Improvements, consists of Action Item 4.1, " Reactor Trip System Reliability (Vendor-Related Modifications)"; Action Item 4.2,," Reactor Trip System Reliability (Preventative Maintenance and Surveillance Program for Reactor Trip Breakers)";

Action Item 4.3, " Reactor Trip System Reliability (Automatic Actuation of Shunt Trip Attachments for Westinghouse and B&W Plants)"; Action Item 4.4,

" Reactor Trip System Reliability.(Improvements in Maintenance and Test Procedures for B&W Plants)"; and Action Item 4.5, " Reactor Trip System Reliability (System Functional Testing)." This safety evaluation (SE) addresses Action Item 4.1 only.

II. REVIEW GUIDELINES -

The following review guidelines were used to evaluate the response from the licensee to item 4.1 of Generic Letter 83-28:

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The licensee or applicant shall submit a statement that he has reviewed all vendor-recommended reactor trip breaker modifications and determined that (1) each modification has, in fact, been l implemented; or (2) a written evaluation of the technical reasons for not imp'lementing a modification exists.

l III. EVALUATION AND CONCLUSION By letter dated December 10, 1985, the licensee of the Ft. Calhoun Station i provided information regarding vendor recommended modifications to the reactor trip system. We have reviewed the licensee's response against the review guidelines as described in Section II. A brief description of the licensee's response and the staff's evaluation of the response against the review guidelines is provided below:

The licensee stated that vendors were requested to investigate the reactor trip breaker history and indicate any recommended modifications. The vendors indicated that no modifications are recommended for the reactor trip breakers. Based on this vendor ,

information the licensee does not plan to modify any reactor. trip '

breakers.

Based on our review, we conclude that the licensee's response to vendor recommended modifications to the reactor trip system for the Ft. Calhoun Station is acceptable.

Principal NRC Contributor: J. Bess l

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