ML20010H328
| ML20010H328 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/18/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TASK-15-03, TASK-15-04, TASK-15-14, TASK-15-3, TASK-15-4, TASK-RR LSO5-81-09-049, LSO5-81-9-49, NUDOCS 8109240348 | |
| Download: ML20010H328 (13) | |
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September 18, 1981 Docket No. 50-245 LS05-81 09-049 c4 j T m%
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5 Mr. W. G. Counsil, Vice President it' fN %
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Northeast Nuclear Energy Company
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[6 Post Office Box 270 gmi s
Hartford, Connecticut 06101
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Dear Mr. Counsil:
SUBJECT:
MILLSTONE 1 - SEP TOPICS XV-3, XV-4, XV-14 By letter dated June 30, 1981, you submitted safety assessment reports for the above topics. The staff has reviewed these assessments and our conclusior.s are presented in the enclosed safety evaluation reports, which complete these topic evaluations for Millstone 1.
These evaluations will be basic input to the integrated safety assessment for your facility. The evaluation 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.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing SEo/
Enclosure:
As stated 5
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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 Coun:11 20 Grand Street 917 15th Sti aet, N. W.
Hartford, Connecticuo 06106 Washington, D. C.
20005 Northeast Nuclear Energy Company ATTN: Superintendent Millstone Plant 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 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 l
of Waterford Hall of Records 200 Boston Post Road l
Waterford, Connecticut 06385 l
l John F. Opeka Systems Superintendent Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 U. S. Environmental Protection Agency Region 1 Office ATTN: EIS C0ORDINATOR JFK Federal Building Boston, Massachusetts 02203
MILLSTFd.. SEr TOPIC XV-3 EVALUATION LOSS OF EXTERNAL LOAD I.
ItiTRODUCTION
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Upon loss of e'xternal load the main generator breaker opens causing the turbine generator to overspeed.
This is sensed by an acceleration relay, which initiates a rapid closure of the turbine control valves and opens
, the bypass valves.
Insertion of selected rods is also started. Activa-tion of fast closure of the turbine control valves gives a delayed scram signal which is suppressed by position switches df the bypass valves if these valves open in less than 260 milliseconds.
Proper operation of the 100% capacity bypass system would make the pressure transient relatively mild.
The licensee has performed an analysis on the loss of external load assuming the turbine bypass valves are unavailable.
Because of the delay in cenerating a reac. tor trip signal, tht transient can be limiting.
respect to MCPR and therefore is separately analyzed for each reload. The most recent analysis is presen+ed in reference 1.
I:.
EVALUATION The licensee has submitted an analysis which was performed usin; the nethods and the assumptions presented in the General Electric generic reload reoorts, (reference 2).
These reports have been previously evaluated by NRC and found acceotable.
The rethods and the assumptions used in the analysis are in conformance with the acceptance criteria of SRP Section 15.2.1 except for the value of the initial power, which is assumed to be 100% instead of 102%.
. The results of the analysis show that the maximu cressure reached in the vessel is 1226 psig.
A sensitivity analysis performed by General Electric for a typical BWR (ref. 2) shows that an initial power level of 102% would result in about 3 psi higher peak pressure. This would still be well be-low the 1375 psig (110%) maximum allowable pressure.
Of all the transients described in FSAR, the loss of external load is the most limiting, with regard to MCPR. The MCPR operating limit for each cycle is obtained by addition of the maximum AMPCR value for this transient to the fuel cladding integri$y safety limit MCPR of 1.07.
Because the maximum AMCPR is calculated using an initial power level of 100%, a transient starting from MCPR operating limit and an actual calorimetric power of 102% would lead to MCPR less than the safety limit.
To account for the power measurement uncertainty, a higher operating limit L* ?
- uld be needed.
General Electric nas performed a sensitivity study to de: ermine the relative dependency of CPR upon changes in bundle power.
The res;lts of the study (ref 2, table 5-10) show that to assure no violation of tr.e 'CPR safety 1imit, during the transient, an increase of tuo percent in the "T ~ :rerating. limit wuuld be neeced.
This increase in MCPR would restrict c. c;erational flexibility but would not re:uire any plant modifications or s!.;ification of the operating power.
The licensee has stated that the analyses for tne next fuel cycle are performed using the recently approved new code ODYN which takes the power measurement uncertainty into account.
Thus the criteria of 3RD section 15.2.1 for MCPR will be met for the next reload.
III. CONCLUSIONS As cart of the SEP review for Millstone-1, we have' evaluated the licensee's analysis of the loss of external load event (ref.1), against the criteria
. of SRP section 15.2.1.
Base: on this evaluation, we have concluded that excluding the MCPR the results are in conf.ormanc.e.with the crite[r_ia,[__Jhe_.
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MCPRwill.be_injonformancewiththe[criteriahftef_thenext,reloidjen___
the operating limi_t MCPR is calculated using a new recently approved code.
TURBINE TRIP I.,JNTRODUCTION A turbine trip is actuated by fast closure of the turbine stop valves, which abruptly interrupt steam flow to the turbine.
Independent of the cause, a turbine trip is followed by a reactor trip initiated directly by turbine stop valve position switches.
The turbine bypass valves ope, to remove steam generated in tne core.
The licensee has cresented an analysis of the turbine trip event in the FSAR.
For the latest fuel cycles, the transient has not been reanalyzed.
- Rather, reference is made to th'e results of a loss of load event which is a more 1imiting transient (ref. 3) and considered to bound the results of a turbina trio event.
II.
EVALUATION The turbine trip transient is bounded by the loss of lead event.
In the turbine trip event the reactor is tripped before the steam f'.ow to the turbine is interrupted.
In the loss of load event, however, the reactor trip signal is delayed and the trip is initiated after the turbine control valves have fully closed.
III.
CONCLUSIONS As part of the SEP review for Millstone-1 we have evaluated the licensee's analysis (ref.1) of the turbine trip event and have concluded that it is bounded by the loss of external load event.
.4-LOSS OF COEEZ ER VACUUM I.
INTRODUCTION In the extreme case of a sudden loss of condens.c vacuum, the transient would be identical to the turbine trip transient with failure of bypass.
The most limitino sir.gle failure during 'that transient has been identified as a safety /
relief valve failure to open.
The licensee has presented an analysis of the loss of condenser vacuum in the FSAR.
Since then, the transient has not been reanalyzed.
Rather, reference is made to the results of a loss of load event (ref. 3) which is considered to bound the loss of condenser vacuum event.
II.
EVALUATION The worst' case loss of condenser vacuum transient is identical to the tc-bine trip transient with failure of bypass.
However, since loss of cc.ndenser vacuun results in a loss of bypass, an additional singie failure should be assuned to satisfy the SRP 15.2.1, section II, acceptance criteria 2d.
The most limiting single failure that could produce the highest peak pressure is a safety relief valve failure to open.
General Electric's sensitivity analysis (ref. 4) has shown that a 17% decrease in relief ca acity which corresponds to the capacity of one safety valve, would result in a pressure increase of less than 30 psi. This increase in pressure added to the bound-ing loss of load event would produce a peak pressure of less than 1200 psig, which is below the maximum allowable pressure of 1375 psig.
A relief valve failure to open would not influence the minimum MCHFR because the minimum is reached before any relief valve is opened.
. I '. I.
CQnCLUSIONS As part of the SEP review for Millstone-1 we have evaluated the licensee's analysis (ref. 3) of the loss of condenser vacuum and h' ave concluded that
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this transient is in conformance with the criteria of SRP section 15.2.1.
CLOSURE OF MAIN STEAM ISOLATION VALVES I.
INTRODUCTION Inadvertent closure of the main steam isolation valves results in the loss-of the steam removal path from the reactor to the turbine and may cause vessel everpressurization.
A direct reactor trip is initiated on 10% closure of the isolation valves.
The licensee has performed an analysis of tne closure of the main steam isola-tion valves for each reload.- The most recent analysis is presented in reference l.
In that analysis, a failure of the direct reactor trip sicnal is cisured and the reactor is tripped only on high neutron flux. This assumption cres bcycr.d the criteria of SRP section 15.2.1 and the purpose of the analysis is to demonstrat" that the relief capacity meets the ASME III criterit.
' PR is not calculated' for this event as described in the evaluation sect :-
II.
EVALUATION The licensee has submitted an analysis whicn has been performed using r.s method and the assumptions presented in General Electric generic reload repor:s, reference 2.
Those reports have been evaluated by NRC.
The methods and the assumptions used in the. analysis are in conformance with the acceptance criteria of SRP section 15.2.1 except that the value of the initial power is assumed to be 100% instead of 102%.
However, the assumption of direct reactor trip failure goes beyond the SRp criteria.
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The results of the analysis indicate that the maxima. ;ressure reached in the vessel is 1276 psig.
General Electric's generic sensitivity analyses show that a'ssujming an initial power of.102% would add about 3 psi to the
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maximum pressure and assuming a failure of'one safety / relief valve to open would add less than 30 psi.
Even after those additions the pressure would still be below the maximum allowable pressure of 1375 psig.
MCPR is not calculated.
However, from the isolation valve characteristics it can be seen that if a direct reactor trip from 100% valve closure functions properly, the reactor trip decreases the cere the~rmal power to about 50%
before isolation valve closure starts throttling the steam flow.
That trip signal is protected'acainst single failures and credit for its proper functioning is consistent with the criteria of SRP section 15.2.1.
Early decrease -in reactor power would prevent MCPR decrease below its normal full power value.
III.
CC"CLUSIONS As part of the SEP review for Millstone-1, we have evaluated the licensee's analysis of the closure of the main steam isolatic
<a'ves event (ref.1),
against the cri,teria of SR? section 15.2.1.
Based :: :r.is evaluation we have concluded that the ana'.jsis is in conformance the present SRP criteria.
STEAM PRES _URE. REGULATORY FAILURE S
I.
INTRODUCTION In case of a steam pressure regulator failure in the direction of decreasing flow, the turbine control valves start to close.
After a slight increase in pressure an independent backup regulator takes over the pressure control i
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and reopens the valves.
Steam pressure is stabilized at the setting of the
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backup regulator dich is somewhat higher than the normal operating pressure.
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,3 y c,.~.r The most limiting single failure is considered to be a failure of the backup regulator which is the only equipment needed to mitigate the initial event.
Failure of the backup regul'ator is equivalent to a loss of external load.
The licensee has presented only a qualitative discussion on the event (ref.1).
II.
EVALUATION The' event induces a very mild transient on the plant.
In the case of the most limiting single failure the transient is equivalent to a loss of ex-ternal load.
III.
CONCLUSIONS A stear cressure regulator failure event is considered to be bounded by the loss of ex:er.51 load transient whic
- -fer.s to the S?.P criteria.
Therefore, a quantitative analysis of the consecuences of a steam pressure regulator. failure is I
not needed.
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~~r/:E-1, SEP TODIC XV-4 EVALUATION LC'] 0F N0',-E".ED3ENCY A-C p0WER TO THE STATION AUXILI:~::5 7
I.
INTRODUCTION Loss of non-energency A-C power to the station auxiliaries initiates a direct reactor trip.
It also trips the condenser cooling water camps, feedwater pumps, and recirculation pumps:
Coastdown of the recirculation flow and sudden loss of feedwater make the loss of non-amergency A-C power transient different from all transients analyzed under topic XV-3.
The licensee has not presented an analysis of the loss of non-emergency A-C power transient but has stated that it is bounded by more severe transients reanalyzed for each cycle.
II.
EVALUATION The worst case pressure transient resulting from a loss of non-emergency A-C power is ecuica'.ent to the turbine trip without byoass, with exception that in the loss :" :n-emergency A-C power, reactor trip is initiated somewhat earlier and thus the pressure transient aspects of loss of A-C power are bounded by tr.E turbine trip analysis.
Coastdown of the recirculation flow and loss of #sei-ater decrease the reactor power and thus tend to decrease the peak or :3 out this influence is only cinor.
A trio of tne e:irculation pumps, following loss of A-C power, might cause a mismatch be: ween reactor thermal power and recirculation flow. This mismatch is not expected following a turbine trip.
However, our experience is that the severity of the turbine trip transient is reduced if power to the recirculation pump motors is cut off when turbine step valve closure occurs. This is because the rapid reduction in recirculation flow increases the core void l-
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-2 centent thereby reducing the peak transient power and the heat flux.
GE's analysis perfonned to supp6rt the recirculation pump trip (RPT) design indicated thN adding the RPT features will ' result in a reduction of 50 to 70% in ACPR.
Thus we conclude that (4CPR during loss of A-C
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power transientis bounded by the turbine trip analysis,also._,.
Sudden loss of feedwater flow which is also caused by loss of A-C power
' does not lead to significant loss of vessel water inventory.
Power to the feedwater pumps is restored within about one minu,te when the gas turbiner; is started.
III.
C_0NCLUSIONS As part of the SEP review of Millstone-1, the analysis of loss of non-emergency A-C power has been evaluated and we have concluded that this tran-sient is bounded by the turbine trio analysis which is evaluated under SEP topic XV-3.
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MILLI-
-1, SEP' TOPIC'XV-14 EVALUATION INADVERTENT-0PERA?p'. OF ECCS THAT INCREASES REACTOR COOLANT INVENTORY I.
INTRODUCTION - -
The high-pressure ECCS system for Millstone-1 is the feedwater coolant injection system. This utilizes components of the main feedwater system, which are normally in use during power operation.
The licensee has not presented a separate analysis under this topic.
II.
EVALUATION Millstone-1 does not have a separate, high-pressure ECCS. Thus'. the event is
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equivalent to the increase in feedwater flow event which is evaluated under topic XV-1.
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CONCLUSIONS Conclusions are cresented under topic XV-1.
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REFERENCES,
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1.
"illstone-1, Reload No. 7 Licensing Submittals, September 1980.
2.
General Electric Boiling Water Reactor Generic Reload Fuel Appli-cation, NEDE-24011-P, Original and Revisions, May 1977, March 1978, and July 1979.
i 3.
Letter fn
. G. Counsel, Northeast Utilities, to D. M. Crutchfield.
NRC,
Subject:
Millstone Nuclear Power Statics '!r.it No.1, SEP Section XV Topics, June 30, 1981.
4.
Letter from Ivan.'. Stuart, GE, to Vi'.Mr Stello, Jr., $sbject: Code Overpressure Protection Analysh, Sensitivity of Peak Vessel Pressures to Valve Operability, December 23,1P5.
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