ML20236K108

Safety Evaluation Supporting Amend 4 to License DPR-21

ML20236K108
Person / Time
Site:	Millstone
Issue date:	07/17/1987
From:	Office of Nuclear Reactor Regulation
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ML20236K088	List: ML20236K085 ML20236K099 ML20236K108
References
NUDOCS 8708060413
Download: ML20236K108 (3)

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1. 'o UNITED STATES

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 4 TO FACILITY OPERATING LICENSE NO. DPR-21 <

NORTHEAST NUCLEAR ENERGY COMPANY MILLSTONE NUCLEAR POWER STATION, UNIT N0. 1 DOCKET NO. 50-245

1.0 INTRODUCTION

By letter dated February 13, 1987, Northeast Nuclear Energy Company (NNECO 1 or the licensee) submitted an application to amend Facility Operating License No. DPR 21 for Millstone Nuclear Power Station, Unit No 1 (MNPSI).

The proposed amendment would lower the reactor water cleanup (RWCU) system l isolation set point in the Technical Specification (TS) from the existing l Group 3 Isolation Signal (reactor low water level) to the Group 5 Isolation Signal (reactor low-low water level).

2.0 DISCUSSION AND EVALUATION The licensee states that the proposed change to the TS involves the auto-matic action of the RWCU system isolation valves which close on a Group 3 Primary Containment Isolation System (PCIS) signal. The proposed transfer j of the RWCU system isolation valves from Group 3 to Group 5 will isolate RWCU at the low-low level and avoid RWCU isolation at low water level conditions. This will eliminate the unnecessary RWCU isolations that can occur following events such as main steam isolation valve (MSIV) closure with subsequent indicated reactor water level decrease due to the resulting  ;

pressure spike. The set point logic remains the same (one-out-of-two-taken-twice) and failure of any single sensor will not prevent system isolation.

The licensee has concluded that there is no impact on the probability of ,

the failure of the RWCU or PCIS systems.

The low water level signals isolate lines that penetrate the reactor vessel and the primary containment (drywell), and connect to primary systems which are not required during isolation conditions and are located outside of the primary containment. The TS set points for low level and low-low level l i

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are less than or equal to 127 inches above the top of the active fuel and 79 (+4, -0) inches above the top of the active fuel, respectively. RWCU 1' system isolation is achieved by closure of redundant valves in the RWCU.

8-inch return line to the reactor vessel via the feedwater piping (closure of a check valve on backflow and automatic closure of motor-operated iso- j lation valve 1-CU-28) coincident with closure of redundant motor-operated isolation valves in the 8-inch line from the reactor vessel via the recir-culation loop to the RWCU system (isolation valves 1-CU-2 and parallel l l

valves 1-CV-3 and 1-CU-5).

Automatic RWCU system isolation limits the amount of reactor coolant that .i can be released into the reactor building in the unlikely event of a gross 1 RWCU system failure outside the drywell. Lowering the reactor vessel water l 1evel trip setting will delay RWCU system isolation and increase the amount j of potential radioactivity release into the reactor building. However, the low-low level isolation of RWCU will still initiate isolation with 6.75' of water. level above the top of the active fuel.

At present, the RWCU system typically isolates on reactor trips and other upsets. Pressure transients resulting in coolant void collapse such as that following a reactor trip / turbine trip or main steam isolation typically .

cause the indicated water level to decrease enough to activate RWCU isola- I tion on low water level. This then results in loss of capability to remove excess reactor water through the RWCU system as level then increases while  ;

the feedwater control system throttles back to satisfy the reduced feedwater demand. Without this capability, there is increased risk of feedwater system isolation (automatic tripping of feedwater pumps and closure of feedwater control valves) on high reactor water level. The possibility of failure to restart the feedwater pumps increases the dependence on emergency safety features to provide. core cooling.

The balance to be reached is between a potential radioactivity release )

into the reactor building following failure of the RWCU system (set point i change from low level to low-low level) and the risk that continual chal-lenge of the emergency core cooling systems due to isolation of the feed-water system will result in more adverse accident consequences. The licensee provided an off-site dose calculation and a reactor building EEQ profile comparison of RWCU system rupture (RWCU isolation on low-low level) against the main steam line break accident. The results showed the RWCU system failure is bounded by the main steam line break accident. In both cases, the calculated thyroid dose is less than 11 percent of that allowed by 10 CFR Part 100 for the exclusion area bSundary and less than 4 percent of that allowed in the low population zone. The whole body calculated dose is less.than 20 percent of the 10 CFR Part 100 limit for both postulated l accidents. NNECO's offsite dose calculation method has been reviewed by i the NRC and found acceptable as documented in the radiological effluent TS I amendment issued on October 1, 1985.

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f Since the change in plant response due to lowering the RWCU system isola-tion set point is bounded by current accident analyses, the consequences of J the proposed change will not impact the margins of safety in that tha fuel  ;

cladding, the primary coolant boundary, and the containment will remain intact. Although the proposed change could allow a greater quantity of cooling water to escape into the reactor building following a RWCU system ,

pipe break outside the drywell, the consequences are bounded by the design  !

basis loss-of-coolant accident, the main steam line break.

i Based on the above, we find the proposed change to lower the RWCU .ystem s I set point from low level to low-low level in TS limiting conditions for  !

operation acceptable. ]

3.0 ENVIRONMENTAL CONSIDERATION

This amendment involves a change to a requirement with respect to the installation or use of facility components located within the restricted  ;

area as defined in 10 CFR Part 20 requirements. The staff has determined i that the amendment involves no significant increase in the amounts, and no significant change in the typci, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has'previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. j Accordingly, this amendment meets the eligibility criteria for categorical I exclusion set forth in 10 CFR Section 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmenatl assessment need be prepared in connection with the issuance of this amendment.

4.0 CONCLUSION

We have concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense or security or to the health and safety of the public.

5.0 ACKNOWLEDGEMENT Principal Contributor: E. Conner Dated: July 17, 1987

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