ML20054C130
| ML20054C130 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 04/16/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W CONNECTICUT YANKEE ATOMIC POWER CO. |
| References | |
| TASK-15-05, TASK-15-15, TASK-RR LSO5-82-04-051, NUDOCS 8204200097 | |
| Download: ML20054C130 (12) | |
Text
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April 16,1982 IN Docket flo. 50-213 g
LS05-82-04-051 p
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r.o93 Mr. H. G. Counsil, Vice President F
Nuclear Engineering and Operations c
Connecticut Yankee Atomic Power Company 3
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N Post Office Box 270 g
Hartford, Connecticut 06101
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Dear Mr. Counsil:
SUBJECT:
HADDAM FLECK - SEP TOPICS XV-5, LOSS OF NORiiAL FEEDWATER FLOW AND XV-15, It!ADVERTEllT OPENIllG OF A PWR PRESSURIZER SAFETY / RELIEF VALVE By letter dated September 30, 1981, you submitted safety assessment reports for the above topics. The staff has reviewed these assess-ments and our conclusions are presented in the enclosed safety evalua-tion reports, which complete the review of these topics for Haddam fleck.
The evaluations will be a basic input to the integrated assessment for your facility. The evaluations may be revised in the future if your facility design is changed or if NRC criteria relating to these topics are modified before the integrated assessment is completed.
360I
.s Sincerely, oss M G bO A DO!
Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 5 Slo w t/
Division of Licensing
Enclosures:
As stated cc w/ enclosures:
See next page 8204200097 820416 PDR ADOCK 05000213
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Mr. W. G. Counsil CC Day, Berry & Howard Counselors at Law One Constitution Plaza Hartford, Connecticut 06103 Superintendent Haddam Neck Plant RFD #1' Post Office Box 127E East Hampton, Connecticut 06424
. Mr. Richard R. Laudenat Manager, Generation Facilities Licensing Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 Board of Selectmen Town Hall Haddam, Connecticut 06103 State of Connecticut 0Ffice of Policy and Management ATTN: Under Secretary Energy Division 80 Washington Street Hartford, Connecticut 06115 U. S. Environmental Protection Agency Region I Office ATTN:
Regional Radiation Representative 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 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 E
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HADDAM NECK PLANT SEP TOPIC XV-5 LOSS OF NORMAL FEEDWATER FLOW 1.
Introduction A loss of normal feedwater flow could be caused by main feed pump failure, feed control valve malfunction or a loss of offsite power.
The loss of feedwater flow would result in the reduction of feedwater flow to the steam generator when operating at power without an equivalent reduction in steam generator steam flow, thus reducing the water inventory in the steam generator and the heat removal capability of the secondary system.
The consequences of the transient would be an increase in intet temperature, increase in pressurizer level and inc rease in primary oressure.
Reactor protection is provided by trips on steam generator low level with steam-feed flow mismatch, and high pressurizer pressure.
The reactor trip woutd cause a turbine trip and steam would be dumped to either the condenser or atmosphere depending on whether offsite power is available.
2.
Review Criteria Section 50.34 of 10 CFR 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 objective of assessing the risk to public health and safety resulting from operation of the facility, including determinations of the margins of safety during normat operations and transient conditions anticipated durine 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 radio-activity.
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 occurrences.
GDC 25 " Reactivity Control System Redundancy and Capability" requires that the reactivity control systems be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and'with appropriate margin for malfuctions such as stuck rods, specified i
l acceptable fuel design limits are not exceeded.
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3.
Related Safety Topics Various other SEP topics evaluate such items as the auxiliary feedwater system and reactor protection system.
The effects of single failures on safe shutdown capability are considered under I
3-Topic VII-3.
4.
Review Guidelines The review is conducted in accordance with SRP 15.2-6.
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 are required.
The extent to which operator action is required is also evaluated.
5.
Evaluation The loss of feedwater flow event has been analyzed in Ref erences is 2, and 3.
A detailed digital simulation was used to represent the primary and secondary systems.
The transient was analyzed using the RETRAN1/ MOD 2 computer code.
Beginning of life reactor kinetics data were used, reactor power was 102% and a decay heat uncertainty of 20% was assumed.
The most reactive controt rod was assumed stuck in the fully withdrawn position.
The feed-water flow to ali steam generators was assumed to stop at time zero without flow coastdown.
The steam generator (SG) water level would drop from the normat water level to the low level setpoint.
The level in all steam generators was assumed to reach the Low level setpoint simultaneously.
The reactor would trip on coincidence of low SG water level and feedwater flow and steam flow mismatch.
The auxiliary feedwater system would be initiated automatically at 3 minutes after reactor scram.
The results of the analysis indicate that a minimum of 320 gpm of feedwater is sufficient to remove the decay heat.
The minimum DNBR predicted was 1.70 and the peak pressure was 2254 psia which was much less than the design pressure of 2500 psia.
. 6.
Conclusion As part of the SEP review for Haddam Neck Plant, we have reviewed the licensee's analysis of loss of normal feedwater flow event according to the criteria of SRP Section 15.2.-6.
For the loss of normal feedwater flow event, the calculated DNBR of 1.70 is well above the minimum requirements of 1.30.
The predicted peak pressure of 2254 psia is well below the design pressure of 2500 psia.
We, therefore, conclude that the results are in conformance with SRP Section 15.2.6 and are acceptable.
7.
Reference:
(1)
Facility Description and Safety Analysis, for Haddam Neck Plant.
(2)
Plant design change #21 - October, l967, monthly report to AEC (3)
W.
G. Cbun sil l et t e r t o D.
M. Crutchfield, dated May 19, 1980 (4)
W.
G.
Counsil le t t e r t o D.
M.
Crutchfield, dated September 30, 1981.
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HADDAM NECK PLANT SEP TOPIC XV-15 INADVERTENT OPENING OF A PWR PRESSURIZER SAFETY / RELIEF VALVE I.
Introduction The inadvertent opening of a pressurizer safety or relief valve results in a reactor coolant inventory decrease and a decrease in reactor coolant syster pressure.
If the valve is not closed, the loss of coolant inventory causes depressurization of the reactor coolant system, draining of the vessel upper head, and fitting of the pressurizer with a two-phase mix.
ture. Safety injection is initiated which refills the vessel upper 1
head and pressurizer and a stable condition is established with safe-ty injection flow matching PORV discharge flow.
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 structuress 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 normat 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 coolanti control and protection systems be designed with appropriate margin to assura that specified acceptable fuel design limits are not exceeded during normal operations including the effects of anticipated opera-tional 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 occurrences.
GDC 26 " Reactivity Control System Redundancy and Capability" requires that the reactivity control systems be capable of reliably controlling reactivity changes to assure that under conditions of normal operations including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded.
. III. Related Safety Topics Various other SEP topics discuss such items as Engineered Safety Features (ESF). Topic XV-19 reviews the spectrum of loss of coolant accidents.
IV.
Review Guidelines The review is conducted in accordance with SRP 15.6.1.
The evaluation includes review of tha analysis for the event and identification of the features in the plant that mitigate the con-sequences 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.
Evaluation Overpressure protection for the reactor coolant pressure boundary is provided by means of pressurizer code safety valves end Power Operated Relief Valves. The design pressure of the Haddam Neck main coolant system is 2500 psia. Three pressurizer code safety valves and two power operated relief valves (PORV's) Limit the primary system pressure tran-sient below 110% of the design pressure. Each PORV is designed to relieve 210,000 lbs per hour of saturated steam at their set pressure of 2285 psia. The three code safety valves are each designed to relieve 240,000 lbs/hr of saturated steam at their set pressures. The safety valves are set to open at reactor pressures of 2500 psia, 2550 psia, and 2600 psia. Since the normal system pressure is 2015 psia there is
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. considerable margin between normal pressure and the set pressures of the PORV's and the code safety valves.
The f ailure of a PORV to reclose following an overpressure transient was a key factor during the Three Mile Island Unit 2 (TMI-2) accident.
Prior to the Three Mile Island accidents inadvertent opening of a PORV or safety valve was considered only as a small break LOCA, and no specific analyses of PORV opening and its unique opening characteristics were done. The re f ore r this event was not analyzed in Facility Description j
and Safety Analysis Report (Reference 1).
Generic analyses have been performed by Westinghouse in WCAP-9600 (Reference 2) in response to post-TMI requirements for the inadvertent opening of PORV's. Two separate pressurizer vapor space breaks corresponding to one PORV and 3 PORV's stuck open were analyzed in Reference 2.
The vapor space break corresponding to 2 PORV's is bounded by the analyses of these two cases.
The analyses showed no core uncovery and very slight voiding during the transient.
VI.
Conclusions As part of the SEP review of Haddam Neck Plant, the information provided by the licensee has been evaluated against the criteria of SRP Section 15.6.1 andfound to be in general conformance with the requirements of the SRP.
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The acceptance criteria require either that the plant responds such that the criteria for fuel damage and pressure are met or that the consequences of the transient are less severe than the consequences of another transient that results in a decrease of reactor coolant inventory and has the same anticipated frequency T
classification.
The inadvertent opening of PORV is a moderate frequency event whereas the small break LOCA falls under the
]
I frequency group of limiting faults.
No plant specific calculation of MDNBR was performed for this event.
However, since our generic evaluation of topical WCAP-9600 concluded that Westinghouse has 1
demonstrated that DNBR is not expected to decrease below the accept-4 able value of 1.3, we have concluded that the plant is adequately protected for this event.
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REFERENCES s
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Haddam Neck Facility Description and Safety Analysis (FDSA)
Report, July 19, 1976.
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" Report on Small Break Accidents for Westinghouse NSSS", WCAP-9600, 5
I June, 1979.
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