ML20010E088
| ML20010E088 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/26/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TASK-15-08, TASK-15-19, TASK-15-8, TASK-RR LSO5-81-08-047, LSO5-81-8-47, NUDOCS 8109030059 | |
| Download: ML20010E088 (8) | |
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August 26, 1981 Docket l'o. 50-245
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Mr. W. G. Counsil, Vice President tiuclear Engineering and Operations E/., u.5. @ 7 8#h./
tiortheatt fluclear Energy Company
(,A Post Office Box 270
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Dear "r.
.ounsil:
SUBJECT:
MILLST0!;E 1 - SEP TOPICS XV-S, C0llTROL R0D f tISOPERATIO:1 A!!D XV-19, LOSS OF C00LAfiT.',CCIDEllTS RESULTIrlG FROM SPECTRUit 0F POSTULATED PIPI!1G BREAKS WITHI!! THE REACTOR COOLA!!T PRESSURE BOUT;DARY By letter dated June 30, 19S1, you submitted safety assesstent reports for the above topics.
The staff has reviewed these assessnents and our conclusions are presented in the enclosed safety evaluation reports, which complete these topic evaluations for I!illstone 1.
These evaluations will be basic input to the integrated safety assessnent for your facility.
The evaluation may be revised in the fu'.are if your facility design is changed or if !;RC criteria relating to these topics are rodified before the integrated assessment is conpleted.
Sincerely, Dennis !!. Crutchfield, Chief Operating Reactors Branch I!o. 5 Division of Licensing
Enclosure:
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OFFICIAL RECORD COPY usa m ni-m ua nac rosu ais o, Sn Nacu cm
111LLSTONE 1 Docket No. 50-245 Mr. W. G. Counsil i
CC William H. Cuddy, Esquire Connecticut Energy Agency Day, Berry & Howard ATTN: Assistant Direr. tor Counselors at Law Research and rolicy One Constitution Plaza De vel opmen*.
Hartford, Connecticut 06103 Department of Planning and Energy Policy Natural Resources Defense Council 20 Grand Street 917 15th Street, N. W.
Hartford, Connecticut 06106 Washington, D. C.
20005 Northeast Nuclear Energy Company ATTN: Superintendent Millstone Plant P. O. Box 128 Waterford, Connecticut 06385 i
Mr. Richard T. Laudenat Manager, Generation Facilities Licensing Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 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 j
200 Boston Post Road Waterford, Connecticut 06385 4
John F. Opeka Systems Superintendent i
Northeast Utilities Service Company i
P. O. Box 270 Hartford, Connecticut 06101 U. S. Environmental Protection Agency Region 1 Office ATTN: EIS COORDINATOR JFK /ederal Building Bsston, Massachusetts 02203 i
i 1
4
EVALUATION OF CONTROL R0D MISOPERATION EVENTS IN MILLSTONE UNIT 1 (XV-8)
In boiling water reactors up to BWR/65 control rods are moved one at a time.
During startup and up to a preset power level (15 to 25 percent of full power) a rod withdrawal sequence is specified and enforced by orocedures, by computer software, or by hard-wired circuitry.
Above the preset power level rod movements are per formed in such a way as to keep the power distribution within *he requirements of the limiting conditions for operation and to achieve a desired power shape.
A control rod misoperation event occurs in boiling water reactors when a control rod is moved out of sequence or is moved too far after having been properly selected for motion.
Two such events - one during startup or low power operation and one during operation at power - are usually analyzed.
Fod Miscoeration Durina Startug DJring startup and low power c peration, the rod withdrawal sequence at.
Millstone Unit 1 is enforced by the Rod Worth Minimizer (RWM) a computer based system that provides motion blocks to the controi rod drive system when out-of-seqt ence rod motions are attempted.
In the event that the RWM is not operable second operator is required to approve the rod selection and motion before the rod is moved.
Correctly following the withdrawal sequence constitutes normal operation and reactivity additions are designed to permit untroubled increases in power through the startup range (i.e., no periods 50 short that a reactor trip occurs before the operator can take action to prevent it).
- The probability that a misoperation event will occur in Millstone Unit i during startup is very small.
Nevertheless, a generic analysis of the consequences of such an event has been perforned and the results presented in the LaSalle County Station Final Safety Analysis Report (Docket 50 J41, Section 15.4.1).
.The calculation is performed in two steps - first a detailed analysis, including multidimensional effects, is performed for a rod having a worth slightly higher than would be anticipated (1.6 percent reactivity change) after which point kinetics calculations are used to extrapolate the results to rod worths to be expected for out -of-sequence rods.
Calculations were done with an initial reactor power of one percent of rated power because a sensitivity study had shown that the consequences were maximum at this l evel.
Transient terrrination is assumed to occur by means of the APRM scram at low (15 percent) power or by the degra.ded (worst bypass condition) 1RM Scram.
The withdrawal speed is assumed to be the maximum value attainable and rod worths up to 2.5 percent reactivity change were analyzed.
In no case did a peak enthalpy greater than 60 calcries per gram result.
Our acceptance criterion for fuel damage is 170 calories per gram for this event.
Conclusion On the basis that the fuel loading and control rod designs of the Millstone Unit I reactor are essentially the same as that of the boiling water reactors for which the generic analysis of this event was performed we conclude that the analysis is applicable to the Millstone Unit I reactor.
Thus the analysis of this event meets the current requirements and is acceptable.
. Rod Misoperation at power The rod misoperation event at power occurs when the operator selects and withdraws an improper rod or withdraws a proper rod beyond proper limits.
In order to analyze a bounding event several conservative assumptions are made.
The core is assumed to be operating at full power with an assembly or assemblies in the vicinity of the rod to be. withdrawn operating at the limiting condition for operation for linear heat generation rate or critical power ratio.
The rod to be withdrawn is assumed to be fully inserted.
The existence of a rod pattern which would produce these initial conditions implies earlier mistakes by the operator.
The core is assumed to be free of xenon which tends to maximize rod worths.
The LPRM detectors which would yield the greatest response to the event are assumed to be inoperable or bypassed so that protective action is delayed.
The calculation is performed with a three-dimensional reactor simulator code and the assumption is made that the neutron and thermal fluxes have the same time response.
The currently used analysis methodology is described in NED0-240ll, " Generic Reload Fuel Application," and is the same technique that is used for all operating boiling water reactors.
The results of the calculation are used to establish the rod block setting for the rod block monitor.
The most recent plant-specific results were submitted in the Reload No. 7 licensing submittal on September 9, 1980.
Conclusion Since the analysis of this event meets the requirements that are currently used for other plants, we conclude that it is acceptable.
~
TOPIC XV-19 LOSS OF CODLANT ACCIDENTS RESULTING FROM SPECTRUP OF POSTULATED FIPING EREAKS s-s, WITH THE REACTOR C00LANT Y : c e_::.__no,_-. e.. v.
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. MILLSTONE UNIT 1 I.
INTE:2L'CTION The ctjective of this review is to assure that the consequences of Loss of Ccolant Accidents (LOCA) are acceptable, i.e., that the recuirerents of 10 CFR 50.a6 and Appendix K to 10 CFR 50 are met.
Less-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 make-up system, 1
fror pipir g breaks in the reactor coolant pressure boar :.ry.
The review con-l sists of evaluating the licensee's analysis of the spectrum of loss-of-coolant accidents includinc break locations, break sizes. and initial conditions assured, the evaluaticn model used, failure modes and the acceptability of auxiliary systems used.
II.
EVALUATION l
Assuming the most pessimistic combination of circumstances which could lead to core uncovery and excessive teatup following a loss-of-coolant accident, fuel cladding integrity is ultimately naintained by successful l
operation of the Emergency Core Cooling System.
The following systems in the Millstone 1 plant provide the necessary protection to mitigate the consequences of a less-of-coolant accident:
1)
Isolation Condenser System (ICS) initiation on reactor low-low water
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retctor high pressure.
inis svs-cr it ireiinec in :ne his icrci r-s..'s:er.s to insure that cladding integrity is na;r.:ained for costulated s~all treak LOCA conditions in the retirculaticr. cischarge piping with a gas turbine failure and LPCI injection into the daraged loop.
2)
Autor,atic Depressurization System ( ADS) initiation on coincident low-low level and high drywell pressure.
The ADS and ICS serve as backups to the s-Feed..a:er Ecolant injec ion (F!!:I)' System in the event the Fi.'CI coes not crera:e.
3)
Lo. Fressure Coolant Injection (LPCI) system initiation on reactor lcu-low water level and low reactor pressure.
4)
Ccre Spray System initiatico on reactor low-lor; water level and low reactor pressure.
5)
Feecwater Ccciant injection (F;'CI) system initiation on reactor low-low water level and high drywell pressure.
Emergency service water and other support systems must be operable to provide cooling.,ater for ECCS ccr.conents.
The licensee has cralyzed the perftrnance of the emergency core cooling system with General Electric nethcdology '.hich conforms to Section 50.46 of 10 CFR Part 50.
The evaluaticn nodel ccnforns to Appendix K to 10 CFR Part 50 with I
additional analytical rodels; one addition determines the isolation condenser heat removal rate and the second takes credit for LPCI flow past the broken loop.
Both of these model additions were approved by the staf.f in connection with the Safety Evaluation Supporting Amendment No. 67 to Provisional Operating License No. DPR-21 (Millstone Unit No.1), dated May 8,1950.
The failure of a gas turbine emergency power supply is a single failure unique to Mill stone 1.
The break spectrum analysis performed with the apprcved r4odel 2
l showed that this was the limiting failure for small breaks less than 0.1 ft in For larger breaks the single failure of a LPCI injection valve is limiting.
area.
N5 storst large L 'eik '..as detern ined to be the it".ccit suction lir.e break (Design Easis Accident).
ine s.5ii Dreak portion of the spectrum is bounded by the DBA.
III. CO*;CLUSIONS As part of the SEP review of Millstone Unit 1, the loss-of-coolant accident was revic.ved 2:ainst the acceptance criteria of SRP Section 15.6.5 and Section 6.3.
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e The initial conditions related to lir.iting sincle failure, breat size and locaticr.,
- ..er level and c; Era tir.; ccnditions L.5 ee teen rcvic..ed and fcur.
- to ccnfora to the recuire. ents of the SRP:
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