ML20055A327
| ML20055A327 | |
| 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-15-19, TASK-RR LSO5-82-07-027, LSO5-82-7-27, NUDOCS 8207160153 | |
| Download: ML20055A327 (9) | |
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q; July 13,1982 i
Docket No.50-029 LS05-82-07-027 Hr. James A. Kay Senior Engineer - Licensing Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701
Dear Mr. Kay:
StBJECT:
YANKEE - SEP TOPIC XV-19, LOSS-OF-COOLANT ACCIDENTS REStA. TING FROM SPECTRUM 0F POSTULATED PIPING BREAKS WITHIN TE REACTOR COOLANT PRESSURE BOUNDARY By letter dated November 18, 1981, 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 the systems review of this topic for the Yankee Nuclear Power Station. Radiological consequences of this event will be addressed in a separate evaluation.
This evaluation will be a basic input to the integrated safety assessnent 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 is modified before the integrated assessment is completed.)
g804 Sincerely, u.sE( 88)
A D D'.
i+. &ylE Ralph Caruso Project Manager Operating Reactors Branch No. 5 Division of Licensing
Enclosure:
As stated cc w/ enclosure:
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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 Rowe Rowe,Mahsachusetts 01367 Energy Facilities Siting Council 14th Floor One Ashburton Place Boston, Massachusetts 02108 l
U. S. Environmental Protection Agancy 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 23 Monroe Bridge," Massachusetts 01350 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I 631 Park Avenue King of Prussia, Pennsylvania 19406
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i TOPI C XV-19 (SYSTEMS)
LOSS OF COOLANT ACCID ENTS RESULTING FROM SPECTR UM O F POSTUL ATED PIPING BREAKS WITHIN THE R E ACTOR COOLANT PR ESSUR E BOUND AR Y Y ANKEE NUCLE AR POWER STATION I.
I NER OD U CT ION 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 10 C FR 50.46 and Appendix K t o 10 C FR 50 are met.
Loss-of-coolant accidents are postulated accidents that would result from the loss of reactor coolants at a rate in excess of the capability of the reactor coolant make-up system, from piping breaks in the reactor coolant pressure boundary.
The review consists of evaluating the licensee's analysis of the spectrum of loss-of-coolant accidents including break location, break size, initial conditions assumed, the evaluation model used, failure modes and the acceptability of auxil.iary systems used.
II. R EVIEW CR I T ER I A 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 performance of systems provided for the prevention of accidents and the mitigation of the consequences of accidents.
. Section 50.46 of 10 C FR Part 50 requires that all light water reactors with rircatoy cladding shall be provided with an emergency core cooling system designed so that its performance following a LOCA satisfies the criteria set forth in that section.
Performance is calculated with an evaluation model satisfying the requirements of Appendix K t o 10 C FR 50.
I i
The General D esign Criteria (Appendix A to 10 CFR Part 50) set forth the criteria for the design of water-cooled reactors.
GD C 35 "Emergenc y Core Cooling" requires that a system be provided to provide abundant emergency core cooling whose function is to transfer heat from the core following 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 negligible 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.
l III.R EL ATED SAFETY TOPICS Topi c III-5. A.
" Effects of Pipe Breaks on Structures, Systems, and Components Inside Containment" ensures that the ability to achieve safe shutdown or mitigate the consequences of an accident are maintained.
The adequacy of the features provided for Switchover from Injection t o R eci rculat ion modes is addressed in Topic VI-78.
. 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 Engineered Sjfety Feature systems are considered as part of 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 overpressurized.
IV. R EVIEW GUID ELINES The review of ECCS performance during a LOCA is conducted in accordance with Standard Review Plan Section 15.6.5 and 6.3.
The plant is considered to be adequately designed against a LOCA if the criteria of 10 CFR 50.46 are met.
The radiological consequences are addressed in a separate evaluation.
V.
EVALUATION Assuming the most pessimistic combination of circumstances which could lead to core uncovery and excessive heatup following a loss-of-coolant accidente fuel cladding integrity is ultimately maintained by successful operation of the Emergency Core Cooling System (ECCS).
The ECCS in the Yankee Nuclear Power Station provides the necessary protection to mitigate the consequences of a loss of-coolant accident.
The ECCS consists of three high pressure safety injection pumpsi three low pressure injection pumps, and one accumulator.
Two out of three high pressure and low pressure injection pumps are
[
, required to be operable; these are actuated on low main coolant pressure or high containment pressure signals.
The licensee has analyzed the performance of the emergency core cooling system (ECCS) in accordance with 10 CFR 50.46 and 10 CFR 50 Appendix K.
The limiting failure is the loss of qpe Safety Injection Subsystem (one HPSI pump and one LPSI pump).
The break spectrum analysis performed with t he WR EM-Based PWR ECCS Evaluation Model (R e f e renc e 1) identified the most limiting break as a double ended guillotine break at the pump discharge with a discharge coefficient of 1.0.
The highest peak clad temperature (1875 F) is reached for this break; therefore, small breaks are bounded by the large break analysis (R e f s.
2 and 3).
The ECCS performance has been found acceptable by the staff in connection with its evaluation of the Cycle XV reload for Yankee Nuclear Power Station (R e f e renc e 4).
The ECCS design for the Yankee Nuclear Power Station incorporates separate injection lines to the f our R CS cold legs.
Each line is provided with separate high pressure safety injection to accomodate small breaks.
A HPSI revision was part of a 1977 j
backfit proposed in change No. 145 to the facility license.
In the revised design, a check valve is installed in each 4-inch safety injection branch line upstream of the junction from the high pressure line.
The check valve prevents the high pressure injection from backflowing toward the broken' branch line.
This modification accomodates a single failure by separating I
. the high pressure and Low pressure injection headers.
In Reference 3r the Licensee provided a small break analysis for a break postulated to occur in a small Length of ECCS piping (one to two feet) immediately downstream of the check valve which is nearest one of the reactor coolant system (R CS) cold leg injection points.
This break location results in R CS blowdown through a 2.25 in. ID thermal sleeve and ECCS spillage through one 3.438 in. ID ECCS Line to containment.
With the modified system the PCT was calculated t o be' 1124 F for this break.
The most limiting smatt break was determined in a break spectrum analysis to be a 4.0 inch break for which a 0
PCT of 1793 F was calculated.
The 2.25 inch break near the HPSI injection point is thus bounded by both the large and smaLL break spectrum analyses.
The ECCS design modification and the accompanying ECCS analysis were both approved by the staff in R eference 5.
The staff concluded that the modified design (1) provides conformance t o the NR C's rules and regulations relating to the ECCS performance requirements; (2) satisfies the single active failure criterion; (3) allows testing of individual ECCS/
accumulator components; and (4) does not require operator action.
The accumulator sub-system was found acceptable on similar bases and by including Technical Specifications for periodic testing and for Limiting conditions'fpr operation.
. VI. CONCLUSIONS As part of the SEP review of the Yankee Nuclear Power Station, the loss-of-coolant analysis was reviewed against the acceptance criteria of SRP s e c t i ons 15.6.5 86.3.
The initial conditions relative to single failure, break size and location, power level, and operating conditions have been reviewed and found to conform to the requirements of the SR P.
The analysis was performed with an approved evaluation model and the results were found to be acceptable.
One portion of the evaluation related to the seismic design of essential ECCS equipment was not addressed. The SRP requires that the piping and instrumentation for essential ECCS components be designated seismic Category I and Safety Class II. This aspect will be addressed under SEP Topics III-l and under III-6 when the licensee completes its overall evaluation of the Yankee Nuclear Power Station seismic design.
VII.R E FER EN CES 1.
YAEC-1160 "Apptication of Yan kee WR EM-Ba sed Gener ic PWR ECCS Evaluation Model to Maine Yankee" July 1978.
2.
Proposed Change No. 145, Supplement No. 5, " Yank ee R owe Core XIII ECCS Performance Evaluation", August 1, 1977.
3.
Proposed Change No. 145, Supptement No. 7, WYR 77-90,
" Additional Yankee R owe Core XIII Small Break Analysis",
September 21, 1977.
4.
Safety Evaluation by the Of fice of Nuclear R eactor R egula t ion Supporting Amendment No. 69 to Facility Ope ra t ing L icen se No. D PR-3, Yankee Nuc tear Power Station, July 22, 1981.
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1 1 i 5.
Safety Evaluation by the Of fice of Nuclear Reactor Regulation Supporting Amendment No. 43, facility operating License No. DPR-3, Yankee Atomic Electric Company, Yankee I
Nuclear Power Stat ion, D ocket No.50-029, August 25, 1977.
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