ML20151T937
| ML20151T937 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/08/1988 |
| From: | NRC OFFICE OF SPECIAL PROJECTS |
| To: | |
| Shared Package | |
| ML20151T942 | List: |
| References | |
| TAC-00074, TAC-00075, TAC-00076, TAC-74, TAC-75, TAC-76, NUDOCS 8808180336 | |
| Download: ML20151T937 (3) | |
Text
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Q Wshe UNITED STATES 4Y NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATION BY THE OFFICE OF SPECIAL PROJECTS SUPPORTING AMENDMENT NO.153 TO FACILITY OPERATING LICENSE NO. DPR-33.
AMEN 0 MENT NO.148 TO FACILITY OPERATING LICENSE N0. OPR-52 AMENDMENT NO.123 TO FACILITY OPERATING LICENSE N0. OPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2 AND 3 00CKETS NOS. 50-259, 50-260 AND 50-296
1.0 INTRODUCTION
By letter dated June 13, 1988, the Tennessee Valley Authority (TVA or the 4
licensee) requested an amendment to Appendix A of the Browns Ferry Nuclear Plant (BFN) Units 1, 2 and 3 Technical Specifications (TS) for Facility Operating Licenses DPR-33, DPR-52, and OPR-68.
The amendment would modify TS Table 3.7.A to increase the maximum operating time for the inboard low pressure coolant injection (LPCI) valves, FCV 74-53 and FCV 74-67, from 30 seconds to 40 seconds.
Environmental qualification modifications required to meet 10 CFR Part 50.49 criteria resulted in longer stroke times for selected valves in the Emergency Core Cooling Systems.
The motor brakes for LPCI injection valves FCV 74-53 and FCV 74-67 could not be oualified for a harsh postaccident environment nor could qualified brakes be procured. The valve operator brakes were then removed and the valves were regeared which increased the valve stroke times.
2.0 EVALUATION LPCI is an operating mode of the Residual Heat Removal (RHR) system.
LPCI operation utilizes two ider.tical pump loops, with each loop.containing two pumps in parallel. The two loops are arranged to discharge water into different reactor recirculation loops. The inboard LPCI injection valves (FCV 74-53, FCV 74-67) are normally closed. The LPCI mode of the RHR system is initiated by:
(1) high drywell pressure (2.45 psig); or (2) low reactor vessel water level (378 inches). When the raactor vessel pressure has dropped to 450 psig, the LPCI injection valves to both recirculation loops open automatically to allow the LPCI pumps to inject water into the vessel as reactor pressure drops below the pump shutoff head.
Historically, the recirculation discharge line break with an assumed failure of the LPCI injection yalve has been the most limiting loss of coolant accident (LOCA) event for Browns Ferry.
Wi'h the increased valve stroke time, t
the limiting break event for BFN is still the recirculation discharge line 8808180336 880808 PDR ADOCK 05000259 P
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break with an assumed failure of a LPCI injection valve. A comprehensive LOCA analysis was performed by General Electric with the new valve stroke time.
This evaluation also examined the impact of the extended valve stroke time on non-LOCA events, such as high energy line breaks (HELB) and Appendix R fire events, the containment isolation function of the valve, and offsite dose calculations.
In addition to providing water to flood the reactor during a LOCA, the injection valves are part of the return path for the cooling water to the
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reactor vessel during operation of the shutdown cooling mode of the RHR system.
The shutdown cooling mode of the RHR involves long periods of manual operation such that the 10-second increase in the valve stroke time will not adversely affect the function of the LPCI valves in this mode.
The LPCI valves involved in the proposed change are also containment isolation valves. The containment isolation function of each LPCI line is provided by two valves in series:
the testable check valve inside the dr>vell and the
.normally closed injection valve.
The LPCI injection valves have an automatic isolation signal during shutdown cooling.
- The injection valves are nortnally closed, and open only during shutdown cooling, surveillance testing, and when required by a LOCA. During shutdown cooling, the reactor pressure is low enough that rapid reactor isolation is not necessary.
For a postulated break.
dlong the LPCI line, the testable check valve will provide isolation until the redundant isolaticn valves are closed.
The LPCI system is also used to protect core integrity for HELB event., and for certain Appendix R fire ewnts.
Tne licensee's analysis indicated the HELB event is not the most limiting for BFN. The Appendix R fire event is similar to the HELB event in that the reactor will be isolated for a long time after event initiation.
Reactor depressurization is accomplished with the main steam relief valves.
Thus, the core cooling capability is more dependent on the pump shutoff head than the valve stroke time.
A comprehensive LOCA analysis was performed with the new valve stroke times.
This evaluation also examined the impact of the extended valve stroke time on non-LOCA events, other safety functions of the valves, and offsite dose calculations. This safety evaluation demonstrated that the extendad valve stroke times will have insignificant impact on all the analyses above.
In i
addition, the increase in valve stroke time will not result in any changes in the Maximum Average Planar Linear Heat Generation Ratio for all fuel types at BFN. Based on the above, the staff concludes that the increase in the operating time of the LPCI injection valves does not significantly impact the safe operation of the plant and is therefore acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
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The amendments involve a change to a recuirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of
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3 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 these amendments involve no significant hazards consideration and there has been no public criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9)gibility comment on such finding. Accordingly, the amendments meet the eli Pursuant to 10 CFR 51.22(b), no environmental impact statement nor environmental assess-ment need be prepared in connection with the issuance of these amendments.
4.0 CONCLUSION
The staff has 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, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security nor to the health and safety of the public.
Principal Contributor:
J. Kelly Dated: August 8,1988 4
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