ML20040H478
| ML20040H478 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 02/10/1982 |
| From: | Caruso R Office of Nuclear Reactor Regulation |
| To: | Kay J YANKEE ATOMIC ELECTRIC CO. |
| References | |
| TASK-15-05, TASK-15-5, TASK-RR LSO5-82-02-059, LSO5-82-2-59, NUDOCS 8202180293 | |
| Download: ML20040H478 (8) | |
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Mr. James A. Kay 7d n 5lE;ljj y [ " M " [S Senior Engineer - Licensing Ic Yankee Atomic Electric Company N
1671 Worcester Road
/?g Framingham, Massachusetts 01701
Dear Mr. Kay:
SUBJECT:
YANKEE - 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,
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which completes the review of this topic for Yankee.
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, S6od 5;
Ralph Caruso. Project Manager Operating Reactors Branch No. 5 tsu 4 {")
Division of I.fcensing Ann:
Enclosure:
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As stated cc w/ enclosure:
See next page 8202180293 820210 PDR ADOCK 05000029 P
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1 Mr. James A. Kay cc Mr. James E. Tribble, President Yankee Atomic Electric Company 25 Research Drive Westborough, Massachusetts 01581 Greenfield Community College 1 College Drive Greenfield, Massachusetts 01301 Chairman Board of Selectmen Town of Rowe Rowe, Massachusetts 01367 Energy Facilities Siting Council 14th Floor One Ashburton Place Boston, Massachusetts 02108 U. S. Environmental Protection Agency Region 1 Office ATTN: EIS COORDINATOR JFK Federal Building Boston, Massachusetts 02203 Resident Inspect;r Yankee Rowe Nuclear Power Station c/o U.S. NRC Post Office Box 28 Monroe Bridge, Massachusetts 01350 1
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SYSTEMATIC EVALUATION PROGRAM TOPIC XV-5
' YANKEE TOPIC:
XV-5, Loss of Normal Feedwater Flow I.
INTRODUCTION A loss of normal feedwater flow event at the Yankee plant could occur from a regulating system failure, an operator error, or several concurrent valve and pump failures. A large and rapid decrease in feedwater flow when operating at powel without a corresponding reduction in steam flow results in a reduction of the water inventory in the steam generators and thus in a loss of heat removal.
This results in an increase in the reactor coolant temperature and pres.sure which eventually requires a reactor trip to prevent fuel damage. The reactor protection system was designed to trip the reactor on a low water level signal from a steam generator and thus protect the reactor, s
The Yankee Atomic Electric Company (YAEC) submitted the results of an analysis of this loss of feedwater event in its FSAR (Reference 1). After the TMI-2 accident YAEC submitted a revised and more detailed analysis (Reference 2) and agreed to enhance Yankee-Rowe's protection for this event by providing two additional auxiliary feedwater pumps, which could be controlled from the control room, and a reactor trip on high primary system pressure. On June 30, 1981, YAEC submitted the results of an analysis of the loss of flow event with the latest nuclear parameters (Reference 3).
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 performance of structures, systems, and components of the facility with the e
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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 normal 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 occurrences.
GDC 15 " Reactor Coolant System Design" requires that the reactor coolant and associated protection systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation, including the effects of anticipated operational j
oCCurences.
l GDC 26 " Reactivity Control System Redundancy and Capability" requires that the l
reactivity control systems be capable of reliably controlling reactivity changes l
to assure that-under conditions of norm,al o,peration, including anticipated operational occurrences, and with appropriate margin for malfunctions such as i
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. As stated in Ref. 4 the NRC (Auxiliary Systems Branch) is reviewing the ne'ed for automatic initiation of the two additional auxiliary feedwater pumps.
I V.
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.
V.
E VAL UATION As stated in Reference 2, YAEC's revised analysis of the transient due to a loss of feedwater flow at Yankee'
) included consideration of the total avail-able energy and the conservation of the total masses of H O in the primary and 2
secondary systems, and a more detailed consideration of the heat capacity of the steam generator structure.
These considerations were added to calculations which were initially done with the GEMINI-II computer program.
These calculations assumed a step drop in the normal feedwater flow rate to zero at a time when the plant was operating at full power plus 3% to account for uncertainties.
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. For this situation the water level in the steam generators drops to the low level set point in 18 seconds at which time the reactor is tripped.
The primary pressure increases and peaks at 2178 psia at 22 seconds and then decreases to 1860 psia at 60 seconds. Since this-peak pressure is below the 2400 psia set point on the power operated relief valve, it does not open. The secondary system pressure increases significantly after the turbine throttle valve is tripped closed at 20 seconds.
It peaks about 890 psia and then slowly decreases to the steam bypass set point of 760 psia.
For this sequences of events, if no feedwater is added, it will take more than 40 minutes for all of the water to boil out of the steam generators.
The departure from nucleate boiling (DNB) ratio remains above its limit during the transient.
This analysis was approved by the NRC on July 22,1981 (Reference 5).
Non-safety grade component and system failures during the above chain of events could result in the pressure rising to the set points of some or all of t$e code safety relief valves. As stated in Reference 6 the total > capacity of the 12 valves in the secondary system is in excess of the steam generating capability of the reactor coolant system, thus ensuring that the steam generators and associated piping cannot be overpressurized if the turbine throttle valves were to close and the reactor failed to scram. As stated in Reference 7 the total capactiy of the 2 safety valves in the primary system is sufficient to protect against overpressure during a full load rejection accident which causes a high pressurizer level scram.
The required relieving capacities of the safety valves will be higher for these events than for 1
b failures. in non-safety grade components and systems during a loss of normal feedwater
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l flow event.
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. Reference 3 includes a summary of YAEC's latest analysis of the loss of normal feedwater flow event.
This was made with the nuclear parameters of Core XV.
The results are the same as those given above, which are from Reference 2.
VI.
CONCLUSION As part of the SEP review for Yankee
, the analysis for a loss of normal feedwater event has been evaluated. We have concluded that the consequences of this event are in conformance with the criteria of SRP Section 15.2.7 and therefore, the analysis is acceptable.
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RE FERENCES 1.
Yankee Nuclear Power Station Final Safety Analysis Report; Yankee Atomic Electric Company, Westboro, Massachusetts; January 3,1974; Volume III, Section 15.2.6 2.
Letter; Vandenburgh, D. E. of YAEC to USNRC; " Yankee Rowe Loss of Feedwater Analyses"; September 12, 1979.
3.
YAEC letter to NRC on Systematic Evaluation Program Topic Assessments; June 30,1981; pps 22-29.
4.
U.S. NRC; Safety Evaluation By The Office of Nuclear Reactor Regulation Supporting Amendment No. 69 to Facility Operating License No. DPR-3, Yankee Nuclear Power Station (Yankee-Rowe); Washington, D.C.; July 22, 1981; page 55.
5.
ibid; page 9.
6.
Yankee Nucl~ ear Power Station Final Safety Analysis Report; Yankee Atomic Electric Company, Westboro, Massachuset ts; January 3,1974, Volume II, page 10.3-1.
7.
ibid; Volume 1, page 5.2-5.
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