ML20033D003
| ML20033D003 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 12/02/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W CONNECTICUT YANKEE ATOMIC POWER CO. |
| References | |
| TASK-15-19, TASK-RR LSO5-81-12-002, LSO5-81-12-2, NUDOCS 8112040581 | |
| Download: ML20033D003 (7) | |
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December 2,1981 g.Tr Ch Dc.cket No. 50-213 LS05-81 12-002 E
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CI Mr. W. G. Counsil. Vice President 9
Nuclear Engineering and Operations y
Connecticut Yankee Atomic Power Co.
Post Office Box 270 4
Hartford, Connecticut 061 01 A
Dear Mr. Counsil:
SUBJECT:
HADDAM NECK - SEP TOPIC XV-19 (SYSTEMS), LOSS OF COOLANT ACCIDENTS RESULTING FROM SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY By letter dated September 30, 1981, you sulnitted a safety assessment re-port for the above topic. The staff has reviewed this assessment and our conclusions are presented in the enclosed safety evaluation report, which completes the systems review of this topic for Haddam Neck.
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 NRC criteria relating to this topic are modified before the integrated assessment is completed.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing 5
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Enclosures:
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HADDAM NECK Docket No. 50-213 Mr. W. G. Counsil
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cc Day, Berry & Howard Counselors at Law One Constitution Plaza Hartford, Connecticut 06103 Superintendent Haddam Neck Plant RFD el Post Office Box 127E East Hampton, Connecticut 06424 Mr. Richard R. Laudenat Manager, Generation Facilities Licensing 40 theast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 Russell Library 119 Broad Street Middletown, Connecticut 06457 Eoard of Selectmen Town Hall Haddam, Connecticut 06103 Connecticut Energy Agency ATTN: Assistant Director Research and Policy Development Department of Planning and Energy Policy 20 Grand Street Hartford, Connecticut 06106 U. S. Environmental Protection Agency Region 1 Office ATTN: EIS COORDINATOR JFK Federal Building Boston, Massachusetts 02203 Resident Inspector Haddam Neck Nuclear Power Station c/o U. S. NRC East Haddam Post Office East Haddam, Connecticut 06423
TOPIC XV-19 LOSS OF COOLANT ACCIDENTS RESULTING FROM SPECTRU4 0F POSTULATED PIPING BREAV.S WITHIN THE REACTOR COOLANT PRESSURE 80lDQ',hY HADDAM NECK PLANT I.
INTRODLCTION The objective of this review is to assure that the consequences of a Loss of Coolant Accident (LOCA) are acceptable, i.e., that the requirements of the AEC Interim Policy Statement and Appendix K to 10 CFR 50 are met.
Loss-of coolant accidents are postulated accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makc-up system, from piping breaks in the reactor coolant pressure bounda ry.
The review consists of evaluating the licensee's analysis of the spectrum of loss-of-coolant accidents including break locations, break size, and initial conditions assumed, the evaluation model used, failure modes and the acceptability of auxiliary systems used.
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 evaluatien of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility, including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility.
Light water reactors with stainless steel clad must be provided with an emergency core cooling system designed so that its performance following a LOCA satisfies the Interim Acceptance Criteria.
The system must be designed so that it can fulfill its safety function with a single active failure, and loss of offsite power.
The evaluation model must be shown to be suitably conservative as de-termined by the staff.
The General Design Criteria (Appendix A to 10 CFR Part 50) establish mini-num requirements for the principal design criteria for watcr-cooled reactors.
G9C 35 " Emergency Core Cooling" requires that a system be provided to' provide abundant errercency core cooling whose function is to transfer heat from the core following a inss 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 negligible amounts.
The system should have suitable redundancy and interconnections such that system function can be i
maintained assuming a single failure and assuming available of only on-site or only off-site power supplies.
i 10 CFR Part 100.11 provides dose guidelines for reactor siting against which I
calculated accident dose consequences may be compared.
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. III.
RELATED SAFETY TOPICS Topic III-5. A. " Effects of Pipe Breaks on Structures, Systems and Components Inside Containment" ensures that the ability to achieve safe shutdown and mitigate the consequences of a pipe break accident is not impaired by the dynamic effects of the break.
The adequacy of the features provided for Switchover from Injection to Recirculation nodes is addressed in Topic VI-7.B.
Other SEP topics consider the emergency power supplies, effects of flooding of safety-related equipment (VI-7.D), prevention of boron precipitation (IX-4) as well as failure modes of the ECCS (VI-7.C).
In addition such areas as containment integrity and isolation, post accident _ chemistry and ESF systems are considered as part of other SEP topics. Topics VI-2.D and VI-3 address the capability of'the containment heat removal systems to alleviate the pressure / temperature transient so that the containment is not overpressurizcd.
IV.
REVIEW GUIDELINES The review of ECCS performance during a LOCA is conducted in accordance with Standard Review Plan 15.6.5 and 6.3.
The plant is considered to be adequately designed against a LOCA if the interim acceptance criteria (IAC) are met.
The radiological consequences are addressed in a separate evaluation.
_4 V.
EVALUkTION Assuming the most pessimistic combination of circumstance which could lead to core uncovery and excessive heatup following a loss-of-coolant accident, fuel cladding integrity is ultimately maintained by successful operation of the Emergency Core Cooling System.
The Safety Injection System in the Haddam Neck Plant provides the necessary protection to mitigate the consequences of a loss-of-coolant accident.
The Safety Injection pumps are in redundant t rains each capable of injecting 7250 gpm.
One high pressure and one low pres-sure pump are required to be available; these are automatically initiated on low
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pressurizer pressure and level or on Ligh containment pressure signals.
Charging pumps are available to accontmodate small breaks and are included as part of the Safety Injection trains, but are not credited in the ECCS analyses.
The licensee has analyzed the performance of the emergency core cooling system (ECCS) in accordance with the Interim Acceptance Criteria (IAC) for Emergency Core Cooling Systems (effective June 29, 1971).
Since the Haddam Neck Plant uses stainless steel clad fuel, the ECCS performance analysis under the guide-lines of the IAC is acceptable.
The limiting failure is the loss of one Safety Injection Train.
The break
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l spectrum analys!s performed with the Westinghouse evaluation model identified e
I the worst break as an 8.24 square foot cold leg split (Ref.1).
The highest peak clad temperature (2182 F) is reached for this break; therefore, small breaks are bounded by the large break analysis (Ref. 2).
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VI.
CONCLUSIONS As part of the SEP review of the Haddam Neck Plant, the loss-of-coolant analysis was reviewed against the acceptance criteria of SRP Section 15.6.5 and Section 6.3..
The initial conditions relative to single failure, break size and location, power level, and operating conditions have been reviewed and found to con. form to the requirements of the SRP.
The analysis was performed with an approved evaluation rodel and the results were found to be acceptable.
VII. REFERENCES 1.
D. C. Switzer (CYAPCO) letter to A. Schwencer (NRC), " Revised ECCS Evaluation Upper Head Ruid (UHF) Temperature Assumption", dated liay 2,1977.
- 2. W. G. Counsil (CYAPCO) letter to D. L. Ziemann, dated August 28, 1979, "Small Break LOCA Analyses."
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