ML20039E858
| ML20039E858 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/07/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W CONNECTICUT YANKEE ATOMIC POWER CO. |
| References | |
| TASK-15-05, TASK-15-5, TASK-RR LSO5-82-01-011, LSO5-82-1-11, NUDOCS 8201110551 | |
| Download: ML20039E858 (7) | |
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January 7,1982 Docket No. 50-245 L505-82-01-011
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Mr. W. G. Counsil, Vice President N
Nuclear Engineering and Operations 0.7 dSl/ [jgp' ' q; Connecticut Yankee Atomic Power Company Z
Post Office Box 270 C '.""
Hartford, Connecticut 06101
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Dear Mr. Counsil:
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SUBJECT:
MILLSTONE 1 - SEP TOPIC XV-5, LOSS OF NORMAL FEEDWATER FLOW By letter dated June 30, 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 review of this topic for Millstone 1.
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.
SW Dennis M. Crutchfield, Chief
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!,'erating Reactors Branch No. 5 j(oh Division of Licensing D54
Enclosure:
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111LLSTONE 1 Docket No. 50-245 Mr. W. G. Counsil cc William H. Cuddy, Esquire Connecticut Energy Agency Day, Berry & Howard ATTN: Assistant Director Counselors at Law Research and Policy One Constitution Plaza Development Hartford, Connecticut 06103 Department of Planning and Energy Policy Natural Resources Defense Council 20 Grand Street 91715th Street, N. W.
Hartford, Connecticut 06106 Washington, D. C.
20005 Northeast Nuclear Energy Company ATTN: Superintendent Millstone Plant i
P. O. Box 128 Waterford, Connecticut 06385 Mr. Richard T. Laudenat Manager, Generation Facilities Licensing Northeast Utilities Service Company.
P. O. Box 270 Hartford, Connecticut 06101 i
Resident Inspector c/o U. S. NRC P. O. Box Drawer KK Niantic, Connecticut 06357 Waterford Public Library Rope Ferry Road, Route 156 Waterford, Connecticut 06385 First Selectman of the Town of Waterford Hall of Records 200 Boston Post Road Waterford, Connecticut 06385
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John F. Opeka Systems Superintendent Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 U. S. Environmental Protection Agency Region I Office ATTN: EIS C0ORDINATOR JFK Federal Building Boston, Massachusetts 02203 1
MILLSIONE 1 SEP TOPIC XV-5: LOSS OF NORMAL FEEDWATER FLOW I.
INTRODUCTION
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Loss of feedwater flow could occur as a result of the simultaneous tripping of all feedwater pumps, or a feedwater controller failure that closes the feedwater control valves. When the feedwater flow drops to 20% of full flow, the recirculation loop control system is designed to reduce the speed of the motor-generators (M-G's) to a minimum to protect the recirculation and
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jet pumps from cavitation.
Loss of feedwater causes the water level in the reactor vessel to drop. The reactor protection system is designed to trip the reactor when the water level drops to the low-level set point.
Due to the decay heat and the release of steam, the water level will continue to drop until the low-low water level set point is reached at which time the protection system is designed to close the main steam isolation valves (MSIV's) 9 and to trip the recirculation pumps. An isolation condenser is designed to take away the decay heat after the MSIV's close and thereby independently maintain the water level above the top of the active fuel.
The tbrtheast Nuclear Energy Company (NNECO) submitted an analysis of the loss of feedwater event at Millstone 1 in March,1968 (Reference 1).
In August, 1972 (Reference 2) Millstone I transients were reexamined for end-of-cycle core char-acteristics, and the events were either reanalyzed or determined to not require reanalysis. This event was also assessed in a gencric study for a topical report (Reference 3), which was published in May,1977, and in another generic study l
after the Three Mile Island-2 accident (References '5 and 6).
i 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 m
and performance of structures, systems, and components of the facility with the
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objective of assessing the risk to public health and safety resulting from operation of the facility, including determination of the margins of safety during nonnal operations and transient conditions anticipated during the life of the facility.
Section.50.36 of 10 CFR Part 50 requires the Technical Specifications to 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) establish minimum requirements for the principal design criteria for water-cooled reactors.
GDC 10 " Reactor Design" requires that the core and associated coolant, control and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during normal operation, including the effects of anticipated operational occurrence.
GDC 15 " Reactor Coolant System Design" requires that the reactor coolant and associated protection systems be designed with sufficient margin to assure that 1
the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation, includi.ng the effects of anticipated operational occurrences.
GDC 26 " Reactivity Control System Redundance and Capability" requires tnat the reactivity control systems be capable of reliably controlling reactivity changes i
to assure that under conditions of normal operation, including anticipated opera-tional occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded.
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III.
RELATED SAFETY TOPICS Various other SEP topics evaluate such items as the reactor protection system.
The effects of single failures on safe shutdown capability are considered under Topic VII-3.
IV.
REVIEW GUIDELINES The review is conducted in accordance with SRP 15.2.7 The evaluation includes review of the analysis for the event and identification of the features in the plant that mitigate the consequences of the event as well as the ability of these systems to function as required.
The extent to which operator action is required is~also evaluated.
Deviations from the criteria specified in the Standard Review Plan are identified.
s V.
EVALUATION Since it is assumed that the reactor trips on a low-water-level signal and that the power then rapidly decreases, the critical power ratio does not decrease below its initial, steady-state value.
The main concern in this event is the loss of too much water so that the fuel elements are uncovered and inadequately cooled.
From the analysis in Reference 1, it can be concluded that:
1.
There is sufficient water in the reactor vessel so that a complete loss of flow event will not uncover the core before the main steam isolation valves (MSIV's) close.
1 2.
After the MSIV's close it is highly likely that the reactor will be generating
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more steam than the isolation condenser can condense.
This excess steam will cause the pressure in the reactor vessel to rise to the reli.ef valve setting L
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. resulting in blowdown through the relief valves to the torus suppression water. At approximately five minutes after a scram from sustained, full-power operation the condenser will be able to condense all of the steam that is produced by the decay heat. The blowdown will then stop; and the isolation condenser will thereafter maintain the water level in the reactor vessel.
In the first generic study (Reference 3) it was found that this loss of feedwater event is not as limiting as a turbine trip.
The NRC in Reference 4 agrees with this conclusion.
Following the Three Mile Island Unit 2 accident, the loss of feedwater event was generically reanalyzed for BWR's (Reference 5) using best estimate calcula-tions in response to information requested by the Bulletins & Orders Task Force.
f In its evaluation (Reference 6) of this report (Reference 5) the NRC stated that the basic requirements for this event, which are given in General Design Criteria 10 and 15, appear to have been met, even in the light of the TMI-2 experience.
VI.
CONCLUSION As part of the SEP review of Millstone 1, the analysis for loss of feedwater has been evaluated and we have concluded that the consecuences of this event are bounded by those of turbine trip.
This has been reviewed under SEP Topic XV-3, where it was found that a turbine trip is bounded by a loss of load event, for which it was concluded that, excluding the Minimum Critical Power Ratio (MCPR),
the results of the analysis are in conformance with the criteria of SRP section 15.2.1.
It was also found that the MCPR will be in conformance with this criteria after the next reload when the operating limit MCPR for the loss of load event is calculated using a'new recently approved code.
The staff I
evaluation of Topic XV-3 was issued in Reference 7.
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REFERENCES 1.
Millstone Point Nuclear Power Station Unit 1, Final Safety Analysis Report, Amendment 5, Chapter XI, Section 3.10; March 14,1968; Northeast Nuclear Energy Company, Hartford, Connecticut.
' ransient Reanalysis for End-of-Cycle Core Characteristics; August 14, 1972; 2.
T Northeast Nuclear Energy Company, Hartford, Connecticut.
3.
General Electric Company; Licensing Topical Report General Electric Boiling Water Reactor, Generic Re' loa'd Fuel Application; NEDE-24011-P-A; May,1977; pps 5-11, 5-68, 5-69.
4.
U. S. Nuclear Regulatory Commission; Safety Evaluation for the General Elec-tric Topical Report, Generic Reload Fuel Application, (NEDE-24011-P); April, 1978; page C-62.
5.
General Electric Company; Additional Information Required for NRC staff Generic Report on Boiling Water Reactors; Revision 1; NED0-2408A; August,1979.
6.
U. S. iiuclear Regulatory Cc.nission; Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in GE-Designed Generating Plants and Near-Term Operatina License Apolications; NUREG-0626; January,1980; page G-3.
7.
- Letter, D.' Crutchfield (NRC) to W. Counsil (NNECO),datedSeptember 18, 1981.
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