ML20040E204
| ML20040E204 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/12/1982 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TAC-43320, TAC-47355, TAC-47389, TAC-47473, TAC-54199, NUDOCS 8202030278 | |
| Download: ML20040E204 (12) | |
Text
_ _ _ ~
f g
y
.+
e
.+
tladb JAN 121982 Docket No. 50-336 q
d
/
.it Hr. O. G. Counsil. Vice President
'g f* $ h tiuclear Engineering & Operations i
'i Northeast Nuclear Energy Company M-t H
Hartford, Connecticut 06101
\\p y,]p /)k 4"
P. O. Box 270 h
Y
Dear Mr. Counsil:
ITQ We have completed our review of Transient Analysis covered in Sections 5.3.2 through 5.3.9, 5.3.13 and 5.3.15 through 5.3.17 of the Basic Safety Report (DSR) submitted by your letter dated March 6,1980. This BSR is intended to serve as a reference fuel assembly and safety analysis report for use of Westinghouse fuel assemblies in NSSS designed by Combustion Engineering, specifically in Millstone Nuclear Power Station, Unit No. 2.
~
Based on our review, we conclude that the transient analysis for startup i
of an inactive reactor coolant pump, excess load, loss of electrical load and/or turbine trip, loss of normal feedwater, excess heat removal due to feedwater malfunction, reactor coolant system depressurization, loss of coolant flou, transients resulting from malfunctions of one steam generator.
steam line rupture, steam generator tube rupture, and reactor coolant pump seized rotor events are acceptable. The uncontrolled boron dilution and pump shaft break events remain unresolved and will be addressed in more detail in the Cycle 5 reload safety evaluation. A copy of our safety evaluation for the above transients is enclosed, t
i Please note that the 1981 version of the Westinghouse Appendix K Evalua-tion Model, as described in Westinghouse letter of May 15, 1981, has been approved as documented by the December 1, 1981 letter from our 2
i J. R. !! iller to E. P. Rahe of Westinghouse.
i We currently expect to complete our BSR review in about one month. At t
that time we will issue a final Safety Evaluation covering fuel design and thermal hydraulic characteristics including the remaining transient and accident analyses addressed in the BSR.
4 Sincerely, Original sip.ned by.
Robert A. Clark Robert A. Clark, Chief 8202030278 820112 Operating Reactors Branch #3 PDR ADOCK 05000336 Division of Licensing P
Enclosure:
As stated j
- See previous page for concurrence and distribution i
cc-See next page omce>.....:..........
.9.RB.#.3 ;,01.*,,.
.QRB#3,;D(*,
.R S B.*.,,,,
08.BD.;D...
PMKreutzer EConner/pn TSpeis 4
sunme >
.s 1/sza2,,,,,,
11.8/s2,,,,,,1/,9/82,,,,
Jy82 7,,
om>
j.
nac ronu ais oo-so> uncu eno OFFICIAL RECORD COPY usm u-swa
m DISTRIBUTION:
Docket File NRC PDR L PDR NSIC-TERA Docket No. 50-336 ORB #3 Rdg DEisenhut OELD I&E-3
!!r. W. G. Counsil, Vice President ACRS-10 Nuclear Engineering & Operations JHeltemes Northeast Nuclear Energy Company PMKreutzer-3 P. O. Box 270 RAClark llartford, Connecticut 06101 EConner Gray File
Dear Mr. Counsil:
MMcCoy TSpeis We have completed our review of Transient Analysis covered in Sections 5.3.2 through 5.3.9, 5.3.13 and 5.3.15 through 5.3.17 of the Basic Safety Report (BSR) submitted by your letter dated March 6, 1930. This BSR is intended to serve as a reference fuel assembly and safety analysis report for use of Westinghouse fuel assemblies in Millstone Nuclear Power Station, Unit No. 2.
Based on w r review, we conclude that the transient analysis for startup of an inactive reactor coolant pump, excess load, loss of electrical load and/or turbine trip, loss of normal feedwater, excess heat removal due to feedwater malfunction, reactor coolant system depressurization, loss of coolant flow, transients resulting from malfunctions of one steam generator, steam line rupture, steam generator tube rupture, reactor coolant pump seized rotor and pump shaft break events are acceptable.
The uncontrolled boron dilution event remains unresolved and will be addressed in more detail in the Cycle 5 reload safety evaluation. A copy of our transient section safety evaluation is enclosed.
Please note that the 1981 version of the Westinghouse Appendix K Evalua-tion Model, as described in Westinghouse letter of May 15, 1981, has been approved aa locumented by the December 1, 1981 letter from our J. R. Miller to E. P. Rahe of Westinghouse.
We currently expect to complete our BSR review in about one month. At that time we will issue a final Safety Evaluation covering fuel design and thermal hydraulic characteristics including the remaining transient and accident analyses addressed in the BSR.
Sincerely, Robert A. Clark, Chief ld Operating Reactors Branch #3 DMsion o]cy}7jng,yg c}gg *g
Enclosure:
As stated hbh83 hQ
.QR
- DL
.f.M5.N.
.E,,,,,,,,,[, h....
.e.0RB#3.:DL.....
OFFICE) z er....
,,I,S,p e,,,,,
,,R,A,C,],a,(k,,,,,,,,,
. - >....cc;.....see..ne x t..page.........
..../....../. 8 2 1
........./. 8 2 1
........./. 82 1/q 1../...r/. 82 omp Nac ronu ais oo-so nacu ano OFFICIAL RECORD COPY usam i-m.m
f*
g
' UNITED STATES.
3 g
NUCLEAR REGULATORY COMMISSION r.
j wAswinoTom. o. c.2 ossa t
4
/
JAN 121982 i
Docket No. 50-336 Mr. W. G. Counsil, Vice President Nuclear Engineering & Operations Northeast Nuclear Energy Company P. O. Box 270 Hartford, Connecticut 06101
Dear Mr. Counsil:
We have completed our review of Tran ient Analysis covered in Sections 5.3.2 through 5.3.9, 5.3.13 and 5.3.15 through 5.3.17 of.the Basic Safety Report (BSR) submitted by your letter dated March 6, 1980. This BSR is int'nded e
to serve as a reference fuel assembly and safety analysis report for use of Westinghouse fuel assemblies in NSSS designed by Combustion Engineering, specifically in Millstone Nuclear Power Station, Unit No. 2.
Based on our review, we conclude that the transient analysis for startup of an inactive reactor coolant pump, excess load, loss of electrical load and/or turbine trip, loss of normal feedwater, excess heat removal due to feedwater malfunction, reactor coolant system depressurization, loss of coolant flow, transients resulting from malfunctions of one steam generator, steam line rupture, steam generator tube rupture, and reactor coolant pump -
seized rotor events are acceptable. The uncontrolled baron dilution and pump shaft break events remain unresolved and will be addressed in more detail in the Cycle 5 reload safety evaluation. A copy of our safety evaluation for the above transients is e~nclosed.
Please note that the 1981 version of the Westinghouse Appendix K Evalua-tion Model, as described in Westinghouse letter of May 15, 1981, has been approved as documented by the December 1, 1981 letter from our J. R. Miller to E. P. Rahe of Westinghouse.
We currently expect to complete our BSR review in.about one month. At that time we will issue a final Safety Evaluation covering fuel design and thermal hydraulic characteristics including the remaining transient and acci' dent analyses addressed in the BSR.
~
Sincerely,
{f 7
debertA. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Enclosure:
As stated cc: See next page 9
r i
?
Northeast Nuclear Energy Company cc:
William H. Cuddy, Esquire Mr. John Shedlosky Day, Berry & Howard Resident Inspector / Millstone Counselors at Law c/o U.S.N.R.C.
One Constitution Plaza P. O. Drawer KK Hartford, Cormacticut 06103 Niantic, CT 06357 Mr. Charles Brinkman Manager - Washington Nuclear Operations C-E Power Systems Combustion Engineering, Inc.
4853 Cordell Aven., Suite A-1 Bethesda, MD 20014 Mr. Lawrence Bettencourt, First Selectman Town of Waterford Hall of Records - 200 Boston. Post Road Waterford, Connecticut 06385 Northeast Nuclear Energy Company ATTN-Superintendent Millstone Plant Office of Policy & Management Post Office Box 128 ATTN: Under Secretary Energy Waterford, Connecticut 06385 Division 80 Washington Street Waterford Public Library Hartford, Connecticut 06115 Rope Ferry Road, Route 156 Waterford, Connecticut 06385 U. S. Environmental Protection Agnecy Region I Office ATTN:
Regional Radiation Representative John F. Kennedy Federal Building Boston, Massachusetts 02203 Northeast Utilities Service Company ATTN:
Mr. Richard T. Laudenat, Manager Generation Faci.11 ties Licen' sing P. O. Box 270 Hartford, Connecticut 06101 D
e
.t SAFETY EVALUATION REPORT OF THE Wp TINGHOUSE BASIC SAFETY REPORT 5.0 EVALUATION OF TRANSIENTS AND ACCIDENTS
5.1 INTRODUCTION
The transients discussed in the Basic Safety Report are categorized based on the method used to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded. The first category includes those Anticipated Operational Occurrences for which Reactor Protection System trip setpoints assure that the SAFDLs are not exceeded.
Protection is provided by a reactor trip.
The second category includes those anticipated operating occurrences for which initial steady-state overpower margins are maintained by Limiting Conditions of Operation to assure that the S.AFDLs are not exceeded. Protection is provided by the overpower margin, and a trip.
by the RPS, although available, may not be required.
Events in this category are Loss of Coolant Flow, Full Length CEA Drop, and transients resulting from the malfunction of one steam generator.
A third ' category used in the Basic Safety Report covers postulated accidents other than. LOCA. These accidents are analyzed to ensure that the extent of fuel failure and subsequent radioactive release will be acceptable.
The following sections 5.2.6 and 5.3 discuss the review of the analytical techniques used and the individual events analyzed.
5.2.6 ANALYTICAL TECHNIQUES The analytical techniques used for the Westinghouse Basic Safety Report postulated transients and accidents are reviewed on a generic basis.
The status of the staff review of these methods is discussed below.
5.2.6.1 FACTRAN The FACTRAN code is described in WCAP-7908, "A FACTRAN IV Code for ' Thermal Transients in a 00 Fuel Rod".
This code is used 'to perfonn fuel rod 2
heatup calculations through a cross section of a metal clad U0 fuel rod and 2
the transient heat flux at the surface of the clad. The staff preliminary
review of this computer program was completed during the final stage of the review of RESAR-414 for preliminary design approval and the code was found to be acceptable. However, our evaluation of WCAP-7908 specified certain requirements to be used for various applications of the code. Our review indicates that there is reasonable assurance that the conclusions based on the anal,yses presented in the Basic Safety Report will not be appreciably altered by the completion of the analytical methods review.
If the final
~
evaluation of the methods indicates.that revisions to the analyses are necessary, the ' licensee will be required to implement the results of such revisions.
5.2.6.2 LOFTRAN This computer program is used to evaluate various system transients and accidents, and is still under review by the staff. LOFTRAN is described in WCAP-7907, "LOFTRAN Code Description". Revisions to the code dealing with the addition of multi-loop capability and an improved' pump model to allow calculation of punp coastdown transients have been submitted. Our review has progressed to the point that there is reasonable assurance that the conclusions based on these analyses will not be appreciably altered by completion of the analytical review.
If the final evaluation of the method indicates that revisions to the analyses are necessary, the licensee will be required to implement the results of such revisions.
5.2.6.5 THINC For Westinghouse designed reactors, the THINC computer code has been used to calculate core thermal-hydraulic perfonnance characteristics.
The code considers cross-flow between adjacent assemblies in t,he core and cross-flow and thermal diffusion between adjacent subchannels in the assemblies.
^
The THINC-I code is described in WCAP-7838, " Application of the THINC
~
Program to PWR Design".
Departure from nucleate boiling analysis is performed using the THINC-I code in conjunction with the W-3 correlation both of which have been found acceptable for Westinghouse reloads.
5.3 TRANSIENT ANALYSIS e
,-1 n--
- w
5.3.2 UNCONTROLLED BORON DILUTION Present NRC criteria require that at least 15 minutes be available from the time the operator is made aware of an unplanned boron dilution event to the time a loss of shutdown margin occurs during power operation, startup, hot standby, hot shutdown and cold shutdown. Thirty minutes warning is required during refueling. The staff has determined that control room alanns should be available to alert the operating staff to boron dilution events in all modes of operation.
If a r'edundant alann is not provided, the licensee should perform an analysis to show that the consequences of the most limiting unmitigated boren dilution event meet the following criteria:
(a) the DNBR does not fall below the minimum acceptable DNBR, (b) the priury system pressure does not exceed-110% of the design pressure, and, (c) the pressure-temperature limits of Appendix G are not violated for all postulated unmitigated boron dilution events.
This issue is considered open and will be resolved on.a. plant specific basis.
5.3.3 STARTUP 0F AN INACTIVE REACTOR COOLANT PUMP This event was not reanalyzed because the Millstone Unit 2 Technical Specifications do not pennit operation with less than.four reactor coolant pumps operating.
5.3.4 EXCESS LOAD EVENT This event, characterized as a moderate frequency event, was not considered in the Basic Safety Report since it is not a limiting cooldown event.
The staff concludes that the steam line break is a more' limiting cooldown event and that in an excess load transient the thermal margin /
low pressure trip' will provide assurance that the departure from nucleate boiling ratio will not go below its limiting value. Therefore, no. fuel is expec'ted to fail as a result of the less severe cooldown anticipated operational occurrences.
5.3.5 LOSS OF ELECTRICAL LOAD AND/0R TURBINE TRIP This event, characterized as a moderate frequency event, can result in an unplanned decrease in heat removal by' the secondary system.
This transient was evaluated by Westinghouse using LOFTRAN. The status of the staff review
. e
. of LOFTRAN is discussed in Section 5.2.6 of this Safety Evaluation Report.
The parameters used as input to this model were reviewed and found to be suitably conservative. The reactor is assumed to be initially at 102% of the rated core thermal power. Conservative values for.the moderator and Doppler coefficients of reactivity were assumed. No credit is taken for the operation of the steam dump system, steam generator power-operated relief valves, or pressurizer spray and pressurizer power-operated relief valves.
Main feedwater flow is assumed to be lost at the time of loss of load. The results of the analysis of this transient showed that cladding integrity was maintained; the minimum departure from nucleate boiling ratio did not decrease below its initial value. The maximum pressures within-the reactor coolant and main steam systems did not exceed 110% of their respective design pressures.
The staff concludes that appropriate analysis has been provided for this event for reload applications.
5.3.6 LO'SS OF NORMAL FEEDWATER A number of plant transients can result in an unplanned decrease in heat removal by the secondary system. For the loss of feedwater event analysis, no credit was taken'for the steam dumo and bypass system.
The pressurizer power-operated relief valves are assumed to function normally since this results in greater expansion of the RCS water and less margin to the point where water relief would occur. This event was evaluated by Westinghouse using the LOFTRAN code. The staff reviewed the parameters used as input to this model 'and found them to be suitably conservative.
The results of the analysis showed that cladding integrity was maintained by assuring that the-minimum departure from nucleate boiling ratio did not decrease below 1.30 and that the maximum pressure within the reactor coolant and. main steam system did not exceed 110 percent of the design pressures. The analysis was performed assuming manual initiation of the auxiliary feedwater system. As a result of recent TMI-related requirements, all ~ licensees have tiee~n required to install a system to automatically initiate.
auxilfary feedwater system flow. These systems have been or soon will be evaluated by the staff. The SSR analysis took credit for opening of the pressurizer PORVs. The licensees shoulo verify that the reactor coolant system does not overpressurize if the pressurizer PORVs fail to open.
This will be addressed in the plant specific evaluation.
l-
5.3.7 EXCESS HEAT REMOVAL DUE TO FEEDWATER MALFUNCTION This event was not censidere1 in the Basic Safety Report since it-is not a limiting cooldown event. This is acceptaole for reasons similar to those for the excess load event, for which the 5. team line break was established as the limiting cooldown event.
5.3.8 REACTOR COOLANT SYSTEM DEPRESSURIZATION The most severe core conditions detennined for this event result from t..e inadvertent opening of two pressurizer relief valves. This. transient was evaluated by Westinghouse using LOFTRAN. The status of the staff review of LOFTRAN is discussed in Section 5.2.6 of this Safety Evaluation
~
Report.
The parameters used as input to this model were reviewed and found to be suitably conservative. This transient is terminated by a thermal margin / low pntssure trip. The results of the analysis showed that cladding integrity was maintained by assuring that the departure from nucleate boiling ratio did not decrease below 1.50 and that the maximum pressure within the reactor coolant and main steam system did not exceed 110 percent of the design pressures. The staff finds this analysis and its results acceptable.
5.3.9 LOSS OF COOLANT FLOW Several types of occurrences in the facility can result in an unplanned decrease in reactor coolant flow rate. The ones,that inight be expected to occur with moderate frequency during the life of the facility are a partial loss of coolant flow caused by reactor coolant pump trip (s) or a complete loss of forced reactor coolant flow that may result f' rom the simultaneous loss of electrical power to all pumps. Operating with less than four loops in operation is not pennitted, therefore the most limiting decrease in flo'w event with regard to core thermal margins and pressure within the reactor coolant system is the complete loss of reactor coolant flow transient. This event was evaluated by Westinghouse using the LOFTRAN, FACTRAN and THINC codes. The analytical techniques requiring review completion by the NRC' staff are identified in Section 5.2.6 of this Safety Evaluation Report. The values of the parameters used for input to-this model were reviewed and found to be suitably conservative.
The results of the analysis of the complete loss of reactor coolant flow transient showed that cladding integrity was maintained by assuring that the minimum departure from nucleate boiling ratio did not decrease below 1.30 and that the maximum pressure within the reactor coolant.and rain 5
t steam system did not exceed 110 percent of the design pressures. The staff finds this analysis and its results acceptable.
5.3.13 TRANSIENTS RESULTING FROM MALFUNCTIONS OF ONE STEAM GENERATOR Based on analyses provided for Millstone Nuclear Power Station, Unit No. 2, the loss of load to one steam generator has been found to be the most limiting asymmetric ' transient (Millstone Unit 2 Proposed License Amendment, Power Uprating, February 12, 1979). The analysis provided in the Basic Safety Report predicts a minimum departure from nucleate boiling ratio of 1.50 which is above the staff criterion of 1.30. The event was analyzed by Westinghouse using the LOFTRAN code. The status of the staff review of LOFTRAN is discussed in Section 5.2.6 of this Safety Evaluation Report. The values of the parameters used for input to the models were reviewed and found to be suitably conservative. The staff finds this analysis and its results to be acceptable.
5.3.15 STEAM LINE RUPTURE EVENT The analysis of a steam line break inside the containment, upstream of the flow measuring nozzle at initial no-load condition with offsite power available was made. The hot zero-power end-of-cycle case with the most reactive assembly fully withdrawn represents the most pessimistic initial condition. A single failure in the boric acid injection system.was assumed.
The licensee has stated (Reference 2) that for the case of a steam linee rupture with a loss of offsite power, the minimum approach to criticality would occur later in the transient and the core power increase would be slower than in the similar case with offsite power available. The peak power would remain below the. nominal full power value and the DNB design basis would not be violated. The conservatism of the methods used is being reviewed by.the staff as part of its review' of WCAP-9226 (Reference 3),
which incorporates the LOFTRAN and THINC codes. Our review of WCAP-9226 has progressed to the point that the conclusions based on the licensee's submittal will not be appreciably altered by the completion of the analytical methods review.
The steam line break is assumed to-occur during the no-load condition in order to maximize steam generator, water inventory and yield the most adverse
7 potential for radiological release. This is consistent with the approach used by Combustion Engineering in previous analyses of their Nuclear Steam Supply Systems.
The minimum departure from nucleate boiling ratio did not go below 1.30 using the W-3 correlation. The maximum pressure within the reactor coolant and main steam systems did not exceed 110 percent of the design pressures. The staff finds that' appropriate analysis has been provided "for this event for reload applications.
5.3.16 STEAM GENERATOR TUBE RUPTURE.
This event was not reanalyzed since the previous calculations are not affected by a core reload.
5.3.17 REACTOR COOLANT PUMP SEIZED ROTOR The analysis and effects of an instantaneous seizure of a reactor coolant pump motor.during any aliowed mode of operation have been reviewed.
The mathematical models used by Westinghouse for the evaluation are LOFTRAN, FACTRAN, and THINC. The review status of these codes is presented ~in Section 5.2.6 of'this Safety Evaluation Report. The parameters used as input to the analysis were reviewed and found to be suitably conservative. The analysis was performed with four pumps initially operating and one seized rotor. This is acceptable since operation with less than four pumps operating is not pennitted.
The results of the analysis show d that less than 2 percent of the fuel rods were predicted to experience departure from nucleate boiling and that the peak clad temperatur.e reach'ed was 1956 degrees Fahrenheit. Fuel damage'is minimal and no loss of core cooling capability will' result.
The maximum reactor coolant system pressure predicted for the analyzed event was 2488. pounds per square inch which is well below 110 percent of the design pressure of 2750 psia. Also the maximum pressure within the main steam system did not exceed 110 percent of the design pressure. The analysis
~
did not include a loss c# offside power nor the worst single failure. This requirement will be verified in the plant specific review.
9 9
6
=
=
D r.
The reactor coolant pump shaft break accident was also considered in the staff review of the Basic Safety Report. The acceptance criteria are the same as the pump seizure event.
Both events, rely on the same. reactor protection system trip function (low reactor coolant flow). The flow
~
coastdown for the pump seizure is faster than the coastdown for a broken pump shaft, resulting in a larger reduction in DNBR margin for the pump seizure event.
G 9
6 4
4 4
e e
s
REFERENCES 1.
Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 61 to Facility Operating License No. DPR-65, Northeast Nuclear Energy Company, et al. Millstone Nuclear Power Station, Unit 2, October 6,1980.
2.
Letter, W. G. Counsii (NNECO) to R. A. Clark (NRC), " Additional Information
.on Basic Safety Report", October 27, 1981.
3.
WCAP-9226, Rev. l',
" Reactor Core Response to Excessive Steam Releases",
January,.1978. '
4.
Letter, W. G. Counsil (NNECO) to R. A. Clark (NRC), June 16,1980.
=
O Y
w.
.,