ML20202F098
| ML20202F098 | |
| Person / Time | |
|---|---|
| Issue date: | 07/03/1986 |
| From: | Paulson W Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8607150092 | |
| Download: ML20202F098 (27) | |
Text
. _
July 3, 1986 LICENSEE: B&W Owners Group
SUBJECT:
SUMMARY
OF JUNE 19, 1986 MEETING WITH B&W 0WNERS GROUP On June 19, 1986, the NRC staff met with representatives of the B&W Owners Group (BWOG) in the offices of MPR Associates,1050 Connecticut Avenue, N.W.,
Washington, D.C.
The purpose of the meeting was to discuss the sensitivity study being conducted by MPR for the owners group. Enclosure 1 is a list of attendees. Enclosure 2 are copies of the viewgraphs provided by the BWOG and MPR.
By letter dated January 24, 1986, the EDO informed the BWOG that the NRC staff would initiate a reassessment of B&W reactor designs. The BWOG was encouraged to assume a leadership role in accomplishing key aspects of the overall effort required. By letter dated May 15, 1986, the BWOG submitted their program plan. One element of this program is an independent comparison of different PWR designs with regard to operational transients. MPR Associates has been contracted by the BWOG to perform this study. MPR has developed and is verifying models to perform this sensitivity study. The results of several verification runs are included in Enclosure 2.
j Staff coments on the MPR study included:
1.
The sensitivity study should either address a range of parameters or if significant sensitivity is found, MPR should include appropriate recommendations for future work.
2.
Concern was expressed regarding the extrapolation of the results from a
" generic" plant.
3.
Concern was expressed that MPR dropped the steam generator tube rupture from the study. The staff asked the owners group to think somemore about this decision.
i 4.
The staff stated that they could not formally participate in the peer review of the MPR study; however, the staff stated that they would provide comments on the program during working meetings such as this meeting.
[
The BWOG indicated that the sensitivity effort is scheduled to be completed in the third or fourth quarter (calendar) of 1986.
t
/s/WPaulson Walter A. Paulson, Technical Assistant Division of PWR Licensing-B
Enclosures:
Asstatyd DPh hTA p
WPaulson/pws Qy 07h)/86 8607150092 860703 PDR TOPRP ENVBW C
1 se ENCLOSURE 1 MPR - BWOG - NRC MEETING JUNE 19,-1986 NAME ORGANIZATION W. Paulson NRC/NRR H. Estrada MPR D. Strawson MPR G. R. Skillman GPUNC S. T. Rose Duke Power Larry Reed Duke Power N. M. Cole MPR R. W. Ganthner B&W P. L. Ferguson NRC M. P. Rubin NRC W. Beckner NRC A. J. Szukiewicz NRC/DSR0/EIB H. C. Garg NRC/NRR/PEICSB L. B. Marsh NRC/NRR/RSB R. C. Jones NRC/NRR/RSB T. L. Dragon 1 MPR Nancy Richardson MPR N. Ehrich MPR Jose A. Calvo NRC Jim Carlton B&W Ed.Branagan NRC/NRR/PWR-B/RSB Bharat Agrawal NRC/RES/DAE D. L. Basdekas NRC/RES Charles Turk AP&L G. B. Swindlehurst Duke Robert J. Schomaker B&W Lou Lanese GPUNC Henry Bailey NRC/IE/DEPER
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MEETING
SUMMARY
DISTRIBURTION cket File NRC PDR L PDR DELD E. Jordan B. Grimes l
ACRS 10 i
B. Borsum i
AD Rdg F. Schroeder D. Crutchfield PEICSB RSB s
J. Stolz G. Vissing F. Miraglia W. Paulson P. L. Ferguson i
M. P. Rubin W. Beckner A. J. Szukiewicz H. C. Garg t
L. B. Marsh R. C. Jones Jose A. Calvo Ed Branagan 1
Bharat Agrawal D. L. Basdekas Henry Bailey
}
l i
4
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I i_
i i
1 ENCLOSURE 2 AGENDA 1
OPENING COMMENTS G.
R.
SKILLMAN 2
GENERAL OBJECTIVES OF MPR THE SENSITIVITY STUDY 3
DESCRIPTION OF MODELS MPR 4
COMPARATIVE ANALYSIS MPR MATRIX 5
VERIFICATION PROCESS MPR 6
SCHEDULE FOR PEER CHARLES TURK REVIEW MEETINGS 7
CONCLUDING REMARKS G.
R.
SKILLMAN l
J
GENERAL OBJECTIVES OF THE SENSITIVITY STUDY 1
FOR B&W PLANTS VS.
OTHER PWRs i
CHARACTERIZE RESPONSE NORMAL CONDITIONS UPSETS AND ACCIDENTS INCLUDING COMPOUND UPSETS NOT NORMALLY INCLUDED IN FSARs EVALUATE SAFETY MARGINS:
OVERPRESSURE DNBR KW/ FOOT
- OTHER, IF APPROPRIATE RECOMMEND CORRECTIVE MEASURES FOR B&W UNITS, IF UPSETS OF A SPECIFIC KIND ARE MORE LIKELY SAFETY MARGINS IN SPECIFIC UPSETS ARE SIGNIFICANTLY SMALLER l
'E G :
2500 PSI DESIGN SAFETY LIMIT 4 1.3 DNBR
,KW/FTA
=
a DESIGN SAFETY MRGIN OPERATING SAFETY MARGIN PLANT PROTECTION SYSTEM (REACTOR PROTECTION SYSTEM /
ENGINEERED SAFETY FEATURE SYSTEMN MERGENCY FEEDWATER SYSTEM LIMITS)
NORMAL OPERATION ENVELOPE O
MARGINS 0
TIME TO REACH LIMITS 0
FREQUENCY OF OCCURRENCE
DESCRIPTION OF MODELS UNITS ANALYZED TWO B&W UNITS
[MAY ADD A THIRD]
ONE CE UNIT, PRE-75 WILL EXAMINE DIFFERENCES WITH A POST-75 UNIT ONE )_ UNIT, RECENT, D-TYPE RSGs WILL EXAMINE DIFFERENCES WITH A PRE-75 UNIT
.,_._wm
ANALYSIS NATRIX:
STUDY OF THE SENSITIVITY OF BsW PWRS VS. OTHER PWR$
PRELlHINARY TRANSIENTS ANALYZED H0 DEL FSAR PURPOSE 1
SMALL DISTURBANCES. wITHOUT CONTROLS A.
REACTIVITY STEPS 3
(1)
HIGH POWER X
(2)
Low power X
B.
STEAM FLOW STEPS QUANTIFY TIME TO REACH DEFINED LIMITS FOR KEY (1)
HIGH POWER X
VARIABLES WITH THE " BARE" (2)
Low POWER X f PLANT (NO CONTROLS) AS A QUANTITATIVE MEASURE OF SENSITIVITY FOR 8 8 W, j[
AND CE PLANTS.
C.
FEED flow STEPS (1)
HIGH POWER X
(2)
Low P0wER X)
D.
REACTIVITY IMPULSES (1)
HIGH power X
(2)
Low POWER X
E.
STEAM Flow IMPULSES DETERMINE TRANSFER FUNCTIONS l
(INPUT-0UTPUT HAGNITUDE AND Xh (1)
HIGH POWER PHASE VS.
FREQUENCY) FOR (2)
LOW POWER XI PURPOSES OF CHARACTERIZING DIFFICULTY OF CONTROL.
F.
FEED Flow IMPULSES (1)
HIGH POWER X
l (2)
Low POWER
{
ANALYSIS MATRIX:
STUDY OF THE' SENSITIVITY OF B&W PWR$ VS. OTHER PWRS PRELIMINARY PAGE 3 TRANSIENTS ANALYZED MODEL FSAR PURPOSE 3
SIGNIFICANT DISTURBANCES. WITH CORRECTIVE ACTION B.
IURBINE IRIP (CONTINUED)
(3)
STEAM RELIEF SYSTEM x
EVALUATE SUSCEPTABILITY OF PWRS TO OVERC00 LING ON STEAM MALFUNCTIONS RELIEF CONTROL MALFUNCTION.
~
C.
LOSS OF ONE FEED PUMP x
ASSESS DIFFERENCES AMONG PWRS IN SAFETY MARGINS FOR THIS DISTURBANCE.
EVALUATE CAPABILITY OF PWRS TO WITHSTAND THIS DIS-TURBANCE WITHOUT TRIP.
D.
LOSS OF ALL EEED PUMPS Rx TRIP (1)
NORMAL EMERGENCY FEED x
X "EALIBRATE" MODEL FOR SAFETY MARGIN ESTIMATION FOR DIS-TURBANCES OF THIS TYPE.
(2)
DELAYED EMERGENCY FEED x
ASSESS DIFFERENCES AMONG PWR$ IN SAFETY MARGINS FOR THIS DISTURBANCE.
(3)
NO EMERGENCY FEED, BLEED-AND-OTHER ANALYSES ASSESS DIFFERENCES AMONG FEED DECAY. HEAT. REMOVAL VF.RIFIED FOR PWRS IN SAFETY MARGINS FOR THE PURPOSE THIS DISTURBANCE.
[* ASSESS DIFFERENCES AMONG (4)
EXCESSIVE EMERGENCY FEED x
t PWRS IN SAFETY MARGINS FOR THIS DISTURBANCE.
EVALUATE SUSCEPT AB IL ITY OF PWRS TO OVERC00 LING ON THIS DISTURBANCE.
ANALYSIS NATRIX:
STUDY OF THE SENSITIVITY OF BtW PWRS VS. OTHER PWRS PREllNINARY PAGE 2 TRANSIENTS ANALYZED N0 DEL FSAR PURPOSE 2
StaNtriCANT DrsfuRBANCES. WITHOUT C_0RRECTIVE ACTION A.
ROD DROPS SMALL ax x
x ASSESS DIFFERENCES AMONG PWRS IN SAFETY MARGINS FOR LARGE dK x
x
'l THIS DISTURBANCE.
" CALIBRATE" HODEL FOR SAFETY MARGIN ESTIMATION FOR DIS-TURBANCES OF THIS TYPE.
VERIFY MODEL DYNAMICS FOR 4
REACTIVITY DISTURBANCES.
B.
TUR8INE TRIP x
ASSESS DIFFERENCES AMONG PWRS IN SAFETY MARGINS FOR THIS DISTURBANCE.
C.
LOSS OF ONE FEED PUhP x
ASSESS DIFFERENCES AMONG PWRS IN SAFETY MARGINS FOR THIS DISTURBANCE.
D.
LOSS OF ALL FEED PUMPS x
ASSESS DIFFERENCES AMONG PWRS IN SAFETY MARGINS FOR THl3 DISTURBANCE.
E.
LOSS OF ONE COOLANT PUMP x
3 SIGNIFICANT DISTURBANCES. WITH CORRECTIVE ACTION A.
ROD WITHDRAWAL ACCIDENT, HIGH POWER x
ASSESS DIFFERENCES AMONG PWRS IN SAFETY MARGINS FOR THIS DISTURBANCE.
B.
IURBINE TRIP (1)
NORMAL STEAM RELIEF W/Rx CUTBACx x
EVALUATE CAPABILITY OF PWRS TO WITHDSTAND LOAD REJECTION WITHOUT TRIP.
(2)
NORMAL STEAM RELIEF W/Rx TRIP x
x ASSESS DIFFERENCES AMONG I
PWRS IN SAFETY MARGINS FOR THIS DISTURBANCE.
1
ANALYSIS MATRIX:
STUDV 0F THE SENSITIVITY OF B&W PWR$ VS. OTHER PWRS PRELIMINARY PAGE 4 TRANSIENTS ANALYZED MODEL FSAR PURPOSE 3
SrCNIFICANT DISTURBANCES. WITH CORRECTIVE ACTION (CONTINUED)
E.
CONTROL SYSTEM UPSETS ICS POWER FAILURES
- WITH (1)
VARYING FEED SYSTEM INITIAL
- ASSESS DIFFERENCES AMONG CONDITIONS x
PWR$ IN SAFETY MARGINS FOR THIS DISTURBANCE.
EVALUATE SUSCEPTABILITY OF PWRS TO OVERCOOLING ON THIS DISTURBANCE.
g (2)
VARYING FEED SYSTEM x
CONFIGURATIONS F.
LOSSES OF COOLANT FLOW
[l (1)
LOSS OF ONE PUMP x
ASSESS DIFFERENCES AMONG PWRS IN SAFETY MARGINS FOR THIS DISTURBANCE.
~
EVALUATE CAPABILITY OF PWRS TO WITHSTAND THIS DIS TURBANCE WITHOUT TRIP.
[* ASSESS (2)
LOSS OF ALL PUMPS x
x DIFFERENCES AMONG PWRS IN SAFETY' MARGINS FOR THis DISTURBANCE.
- CALIBRATE" MODEL DYNAMICS FOR DISTURBANCES OF THIS TYPE.
ASSESS DIFFERENCES AMONG
( PWRS IN SAFETY MARGINS FOR THIS DISTURBANCE.
)
FOR PLANTS OTHER THAN B&W, PLAUSIBLE POWER SUPPLY FAILURES WILL BE EVALUATED ON AN AD HOC BASIS.
w-
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ANALYSIS NATRIX:
STUDY OF Tile SENSITIVITY OF B&W PWRS VS. OTHER PWRS PREllMINARY PAGE 5 TRANSIENTS ANALYZED MODEL FSAR PURPOSE G.
LOSSES OF COOLANT SPECTRUM OF BREAK SIZES x**
ASSESS DIFFERENCES AMONG PWRS IN SAFETY MARGINS FOR THIS DISTURBANCE.
4.
CONTROL SYSTEM ANALYSES AND OPERATIONAL BURDEN A.
Rx TRIP WITH AND WITHOUT (1)
SECURING STEAM LOADS x
- ASSESS DIFFERENCES AMONG PWRS IN OPERATOR BURDEN FOR A RELATIVELY FREQUENT UPSET.
(2)
INCREASING MAKEUP x
B.
OPERATOR MANUAL CONTROL REQUIRE-ASSESS DIFFERENCES AMONG MENTS IN RESPONDING TO SELECTED PWRS IN BACKING UP AUTO-FAILURES OF AUTOHATic CONTROL LOOPS x
MATIC CONTROLS.
5 VERrricArioN ANALYSES COMPARE MODEL PREDICTIONS A.
B&W UNIT RESPONSES AGAINST ACTUAL PLANT RESPONSES FOR SELECTED TRANSIENTS.
(1)
TURBINE TRIP WITH Rx TRIP x
(2)
LOSS OF ONE FEED PUMP x
(3)
TURBINE TRIP WITHOUT Rx TRIP x
(4)
LOSS OF ALL MAIN FEED PUMPS x
DEtAYED EMERGENCY FEED x
OTHER ANALYSES AND TEST DATA WILL BE USED IN THIS EVALUATION, AS APPROPRIATE.
ANALYSIS MATRIX:
STUDY OF THE SENSITIVITY OF BtW PWRS VS. OTHER PWRS PRELIMINARY PAGE 6 TRANSIENTS ANALYZED MODEL FSAR PURPOSE s.
CE UNIT RESPONSFS (1)
LOSS OF ALL MAIN FEED PUMPS X
a c.
.)L UNIT RESPONSES (1)
PARTIAL LOAD REJECTION X
(2)
TURBINE TRIP WITH Rx TRIP X
(3)
PARTIAL LOSS OF'FEEDWATER X
f I
f 4
l m
APPROACH TO VERIFICATION OF MODELS ADD COMPLEXITY AS REQUIRED BOUNDARY CONDITIONS CONTROLS AND PROTECTION DETAILS MODELING SOPHISTICATION e
e
_n
OVERALL PLANT PZR ADIABATIC STEAM / WATER INTERFACE PORY
- INSURGE HOMOGENOUS STEAM / WATER MI.bRE
- OUTSURGE gy (p
SPRAY /PORV
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Ex p
x r:
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MISV TSV l
TCV O
x V
FW RX LOOPS SG FP 1 GROUP KINETICS, HOT LEG ENERGY 1 EQUIVALENT SG
- 1, 2 PUMPS L' = 0 STORAGE VARIABLE SP
- 1 " DUMMY" FOR DOPPLER COEFFICIENT COLD LEG ENERGY ASYMMETRIC FLOW MODERATOR COEFFICIENT STORAGE 1 LUMP FUEL ENEi3GY
" PROGRAMMABLE STORAGE Flow SINGLE NODE FUEL
/C00 LATE HEAT TRAr.iSFER MODELING APPROACH
UTSG
" Ws THE MAL EQUILIBRIUM BETWEEN VAPOR AND
- NEGLIGIBLE CHANGE LIQUID PHASES IN VAPOR PHASE MASS STORAGE PROGRAMMED RISER FLOW WITH STEAMING RATE 100% EFFECTIVE i
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SEPARATOR l
DYNAMIC MOMENTUM BALANCE MFW EFW a c a
w DYNAMICALLY CONSTRAINED DOWNCOMER FLOW CONSTANT HEAT 1 LUMP DYNAMIC ENERGY ADDITION RATE IN BALANCE AROUND EACH OF TUBE BUNDLE REGION FOLLOWING:
DRun EQUAL STEAM AND
[-
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-- 1 DOWNCOMER UPPER VELOCITIES l
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OTSG TH I
T 1 LUMP UYNAMIC ENERGY EVAPORATIVE HEAT I
BALANCE AROUND SH REGION EFW,
TRANSFER UVER ENTIRE o
3 BUNDLE RATE OF CHANGE OF VAPOR hASS NEGLIGIBLE IN IlASS OR 6ALANCE AROUND SH REGION EVAPORATIVE HEAT TRANSFER OVER
" SUBMERGED" SECTION bb b
UNLY INSTANTANEOUS HEAT BALANCE WITH ASPIRATING FLOW CONSTANT RATE OF HEAT ADDITION IN BOILING SECTION OF IUBE BUNDLE DOWNCOMER FLOW DYNAMICALLY CONSTRAINED STEAM AND Lloulo BY fl0 MENTUM BALANCE PHASE VELCCITIES EOUAL CONSIDERING
- HEAD OF FLUID 1 LUMP UYNAMIC ENERGY IN IUBE BUNDLE 8ALANCE AROUND DOWNCOMER AND BOILING SECTION
- HEE OF F m IN DOWNCOMER
- 1 LUMP E0ILING REGIDH IRESSURE LOSS
- 1 LUMP UOWNCOMER PRESSURE LOSS (ORIFICE PLATE)
MODELING APPROACH
120 lieu tron 10 0 b(/
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Power, %
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120 110 Th eraal ico Power, %
90 80 70 60 f
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20 40 60 80 100 120 Time (seconds)
CE PWR MODEL VERIFICATION RUN FEED PUMP / REACTOR / TURBINE TRIP - 73% POWER
MPQ AC3CCIATES FLC2-27-2 5/27/86 Test Data
- - - Model Simulation j--
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' NOTE:
Test Data for TSAT unavailable.
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20 40 60 80 100 120 Time (seconds)
CE PWR MODEL VERIFICATION RUN FEED PUMP / REACTOR / TURBINE TRIP - 73% POWER
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MPR ASSCCIATES F162-27-4 5/27/86 Test Data
- - - Mocel Simulation Note:
Maxeup to reactor coolant system not simulated, leading to discrepancies in reactor coolant volume and pressure response in the long term.
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VERIFICATION ANALYSIS
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MPR DESCRIPTION
- RESULTS I
1 CONCLUSIONS l-I RECOMMENDATIONS o
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BWOGl BWOG BWOG TEAM WORKING TEAM l TEAM DESCRIPTION i
REPORT
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REVIEW l* CONCLUSIONS t-J t
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- OPERATOR BURDEN
- LOCA ASSESSMENT i
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