ML20055A289
| ML20055A289 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 07/13/1982 |
| From: | Caruso R Office of Nuclear Reactor Regulation |
| To: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| References | |
| TASK-15-06, TASK-15-6, TASK-RR LSO5-82-07-026, LSO5-82-7-26, NUDOCS 8207160082 | |
| Download: ML20055A289 (7) | |
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a July 13,1982 Docket No.50-029 LS05-82 026 Mr. Janes A. Kay Senior Engineer - Licensing Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701
Dear Mr. Kay:
SUBJECT:
YANKEE - SEP TOPIC XV-6, FEEDWATER SYSTEM PIPE BREAKS INSIDE AND OUTSIDE CONTAINMENT (PWR)
By letter dated February 1,1982, you submitted a safety assessment report for the above topic. The staff has reviewed this assessment and our conclusions are presented in the enclosed safety evaluation report, which completes this topic for the Yankee Nuclear Power S*stion.
This evaluation will be a basic input to the integrated assessment for your facility. The evaluation may be revised in the future if your facility design is changed or if HRC criteria relating to this topic is modified before the integrated assessment is completed.
Sincerely, s GC>V Ralph Caruso, Project Manager D5" Operating Reactors Branch No. 5 Division of Licensing ADDI m.kyk
Enclosure:
As stated cc w/ enclosure:
See next page 9207160082 820713
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OFFICIAL RECORD COPY usoeoa ni-m.m NRC FORM 318 (10-80) NRCM 024o i
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e Mr. James A. Kay cc Mr. James E. Tribble, President Yankee Atomic Electric Company 25 Research Drive Westborough, Massachusetts 01581 Chairman Board of Selectmen Town of Rpwe Rowe, Massachusetts 01367 Energy Facilities Siting Council 14th Floor One Ashburton Place Boston, Massachusetts 02108 U. S. Environmental Protection Agency Region I Office ATTN:
Regional Radiation Representative JFK Federal Building Boston, Massachusetts 02203 Resident Inspector Yankee Rowe Nuclear Power Station c/o U.S. NRC Post Office Box 28 Monro Bridge,* Massachusetts 01350 Ronald C. Haynes, Regional Administrator Nuclear Rcgulatory Commission, Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 e
4 YANKEE NUCLEAR POWER STATION SEP 70PIC XV-6 FEEDWATER SYSTD4 PIPE BREAK l
I. Introduction A rain feedwater system piping break will result in reduced feedwater flow rates to the secondary plant. Eventually, inadequate feedwater supply to one or more steam generators will result in a reactor scran induced by several possible conditions e.g.,1m steam generator water levels, high reactor coolant loop pressure, high containment pressure, or low main steamline pressure. Following reactor scram, the auxiliary feedwater system is available for decay heat removal.
II.
Review Criteria section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and perfomance of structures, systens, and ccrponents of the facility with the objective of assessing the risk to public health and safety resulting from operation of the racility,, including deterrunation of the margins of safety during norral operations and transient conditions anticipated during the life of the facility.
Section 50.36 of 10 CFR PART 50 requires the Technical Specifications include safety limits which protect the integrity of the physical barriers which guard against the uncontrolled release of radioactivity.
The General Design Criteria (Appendix A to 10 CFR Part 50) gestablish mininum requirements for the principal design criteria for water cooled reactors.
GDC 27, "Cmbined Reactivity Control System Capability," requires that the reactivity control systms, in conjunction with poison addition by the energency core cooling system, have the capability to reliably control reactivity changes to assure that under postulated accident conditions, and with appropriate margin for stuck rods, the capability to cool the core is maintained.
GDC 28, " Reactivity Limits," requires that the reactivity control systems be designed with appropriate limits on the potential amount and rate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than lirited local yielding, nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to irpeir sianificantly the capability to cool the core.
GDC 31," Fracture Prevention of Reactor Coolant Pressure Boundary,"
requires that the boundary be designed with sufficient margin to assure that when stressed under operating, mainter.ance, testing and postulated accident conditions, (1) the boundary behaves in a nonbrittle manner, and (2) the probability of rapidly propagating fractures is minimized.
GDC 35, "D1ergency Core Cooling," requires that a systen be provided to provide abundant emergency core cooling whose function 6 is to provide abundant crergency core cooling whose function is to transfer heat frcm the core cooling a loss of coolant such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented, and (2) clad metal water reaction is limited to negligibic amounts. The system should have suitable redundancy and interconnections such that system function can be maintained assuming a single failure and assuming availability of only on-site or only off-site power supplies.
10 CFR Part 100.11 provides dose guidelines for reactor siting against which calculated accident dose consequences may be ccrpared.
III Related Safety Topics SEP 'Ibpics III-5A, " Effects of Pipe Break on Structures, Systems and Ccrponents Inside Containment," and III-5.B. "P.ioe Breaks outside Containment," consider the dynamic effects (pipe wttipi, jet impingement, adverse environment) on safety related equiprnent.
Other SEP topics address such items as ESF initiation, the auxiliar,f feedwater systs capacity, and contaiment isolation.
IV Review Guidelines The review is conducted in accordance with SRP 15.2.8.
The 6 evaluation includes review of the analysis of the event.
Identification of the features in the plant that mitigate the consequences of the event as well as the ability of these systes to function is required.
V Evaluation If the feedwater line break occurs upstream of the non-return valve the stean generator would not blowdown. The consequences of this event are less limiting than a caplete loss of feedwater event. A feedwater line break (FLB) between the rmn feedwater line non-return valve and the steam generator would result in the blev30an of that steam generator and a loss of all feedwater to all steam generators until the break is isolated, using safety grade cmponents in the auxiliary feedwater system. The blowdown results i
in an initial cooldown of the primary system. The cooldown transient is bounded by the steam line break event (Ref.2). For this event either the low steam generator water level trip, the low rain stean line pressure trip or the high RCS pressure trip is available to scram the reactor. One of the two safety related auxiliary feedwater trains would be sufficient to remove the decay
heat and prevent the prinny system frcxn exceeding 110% of design pressure (Ref.1). Because of the relatively long steam generator dryout time, approximately 40 minutes (Ref.1), the operator would have sufficient tire to divert auxiliary feedwater flow to the unaffected steam generators. The peak FCS pressure for this transient is 2178 psia significantly lower than 110% of the design apressure.
VI Conclusion As part of the SEP review for the Yankee Nuclear Power Station, the staff has evaluated the licensee's analysis of the feedwater pipe brea k event. The results indicate that the operator would have sufficient time to isolate the break and divert auxiliary feedwater to the unaffected steam generators (Ref.1). The maximum reactor coolant system pressure is always below 110% of the design pressure.
We therefore, fir.d the results of the analysis for the feedwater pipe break transient acceptable.
REFERENCES 1.
Letter from Yankee Atomic Electric Co. to NRC, (FYR 81-95),
Systematic Evaluation Program topic assessments, dated June 30, 1981.
2.
Letter from Yankee Atomic Electric Co. to NRC, (FYR 81-11),
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SEP Topic Assessment Completion, dated February 1,1982.
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