ML20235Z345

Safety Evaluation Supporting Util Compliance W/Atws Rule, 10CFR50.62 Re Power Testability Features of Alternate Rod Insertion Sys & Recirculating Pump Trip Design

ML20235Z345
Person / Time
Site:	Brunswick
Issue date:	03/08/1989
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20235Z334	List: ML20235Z331 ML20235Z345
References
NUDOCS 8903150350
Download: ML20235Z345 (4)

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L -SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO COMPLIANCE WITH ATWS RULE 10 CFR 50.62 CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM EL ECTRIC PLANT, UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324

1.0 INTRODUCTION

On September 18, 1987, the NRC issued a Safety Evaluation (SE) for the Brunswick Steam Electric' Plant, Units 1 and 2, concerning compliance with the Anticipated Transient Without Scram Rule (ATWS Rule), 10 CFR 50.62.

In that SE, the staff concluded that the at power testability features of the alternate rod insertion (ARI) system were not acceptable. It stated that Carolina Power & Light Company (the licensee) should modify the testability portion of the arf design so that at power surveillance testing does not prevent the ARI system from responding to an automatic ARI initiation signal. The staff also concluded that the recirculating pump trip (RPT) design was not acceptable because the Brunswick RPT design does not appear to be as reliable as the two previously approved designs (Monticello and the modified Hatch designs.) This position is discussed in more detail in the staff's generic SE which accepted the GE owners group proposed ATWS ARI and RPT designs.

In a letter dated December 4, 1987, the licensee provided a design modification to address the ARI system testability concern; and in a letter dated November 13, 1987, the licensee provided additional information pertinent to the reliability of the RPT design.

2.0 EVALUATION AND CONCLUSION

1. The ARI system Testability The ATWS Rule guidance states that the ARI system should be testable while the plant is at power operation. The boiling water reactor (BWR) owners group licensing topical report NEDE-31096-A specifies the method of compliance as follows:

The ARI system is designed such that periodic surveillance tests can be performed during normal plant operation to provide assurance that the ARI logic is capable of functioning as designed. This testing includes the relay logic to initiate ARI valves actuation. Trip units can be tested using existing test capabilities. Testing of final actuation devices (ARI valves), while the reactor is at power, is not required, since this could affect plant availability.

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Surveillance testing should not prevent the ARI system from responding to an automatic ARI initiation signal.

By letter dated December 4,1987, the licensee provided a modified ARI design which allows at power testing without preventing the automatic actuation signal to initiate the intended function. The modified ARI design provides additional contacts to the test isolation switches, which will allow the isolation of a single circuit, while the remaining circuits remain available for ARI initiation. The staff has reviewed the modified circuitry and finds that the modified design meets the ARI testability requirements. Therefore, the staff concludes that the ARI at power testability concern is resolved.

2. The RPT System Reliability By letter dated November 13, 1987, the licensee presented additional information pertinent to the reliability of the Brunswick ATWS/RPT sy stem. The licensee has separated the breaker failure rates into two parts (1) a coil / trip mechanism dominant failure rate, and (2) the balance of the breaker's component failure rate. The coil / trip mechanism failure contribution to the overall breaker failure rate is estimated to be 13 percent. The overall breaker failure per demand for Monticello is estimated to be 8.12 E-3 per demand, while the overall breaker failure per demand for Brunswick is estimated to be 2.2.0 E-4 per demand.

In addition, the NRC Region I staff performed a survey on Region I BWRs in 1987. This survey shows that out of the 14 Region I BWR plants, three used 480 V RPT breakers (low-voltage breakers) similar to those used at Monticello, and seven plants used four KV breakers (high-voltage breakers) similar to those used at Brunswick. Of the three plants with the low-voltage breaker design, two plants had a total of nine reported failures to trip. However, none of the seven plants with the high voltage breaker design had any reported failure to trip. Four plants (the remainder of Region I BWRs which trip the recirculating pump motor directly) also use high-voltage breakers.

None of the four plants had any reported failures to trip. While the reporting of breaker failures may not be complete, the above survey indicated that the high-voltage breakers are not less reliable than the low-voltage breakers.

Furthermore, the NRC Region I staff report also states that the MG set drive-motor breakers (the type used at Brunswick) and pump-motor breakers are not as mechanically complex as the AKF-2-25 breakers (the type ured at Monticello), and, therefore, might explain the better performance of the high-voltage breakers. On the basis of its study, the Region I staff found that the high-voltage breakers are more reliable than the low-voltage, Monticello-type breakers.

Based on the failure rate calculation presented by the licensee, and the NRC Region I independent survey, the NRC staff concludes that the relia-bility of the Brunswick RPT system is equivalent to the Monticello RPT system and, therefore, is acceptable.

Principal Contributor: H. Li Dated: March 8, 1989 L

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3. REFERENCE 1.. .LetterZimmerman,S.R.,CP&L,.toUSNRC,ATWS-ARIDrahings, dated December 4, 1987.

2. Letter Zimmerman S. R., CP&l., to USNRC, ATWS-ARI System, dated November 13, 1987.

3. Memorandum Murley, T. E., Region I, to Denton, H. R.-(and others),

NRR, Reliability of Nonsafety-Related Breakers During an ATWS, dated January 20, 1987.

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