ML19254E978
| ML19254E978 | |
| Person / Time | |
|---|---|
| Issue date: | 10/17/1979 |
| From: | Salah S Office of Nuclear Reactor Regulation |
| To: | Phillips L Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7911050206 | |
| Download: ML19254E978 (25) | |
Text
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UNITED STATES 3,
NUCLEAR REGULATCRY COMMissICN
[ ] * ;Y $
.vA>lNGTON, 3. C. 20555
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\\ S i l 5,,],J OCT 17 '979 MEMORANDUM FOR:
L.E. Phillips, Acting Chief, Analysis Branch, OSS FROM:
S. Salah, Analysis Branch, DSS
SUBJECT:
PROGRESS ON THE REVISED B&W STEAM LINE BREAK ANALYSIS WITH BWKIN A meeting was held in Bethesda, Maryland on Septeder 6,1979 with the members of Babcock & Wilcox (G&W) and Consumer Pcwer Comoany (CPCO) to discuss the progress made on steam line break analysis methods being developed by B&W. Meeting attendees are listed in enclosure 1.
Copies of viewgrapns used during the presentations are provided as enclosure 2.
The BWKIN code and the associated calculational methods being developed by B&W will be applied to the Midland Plant, Units 1 and 2.
BWKIN is a modified MEKIN code. Final description and the results of this analysis will be submitted by B&W as an addendum to a future topical report on the rod ejection accident. The BWKIN code is being developed as a result of the staff position regarding power distribution prior to and during the steam line break accident with a stuck control rod Q. 211.166, 211.168, 211.169).
At the present development stage, B&W has performed steady-state calcu-lations with BWKIN. The results of these calculations were compared with two-dimensional PDQ calculations. Three-dimensional BWKIN calculations were performed for half core geometry with a 2x2 mesh representing each fuel box radially and 8 axial nodes. A full power beginning-of-life case with stuck rod confhuration following a reactor trip was chosen
~
as the worst case. Core inlet coolant temperature, system pressure, and core coolant flow rates were obtained from the system transient calculations utilizing the TRAP computer code.
The essential thermal-hydraulic and neutronic input parameters for the BWKIN code are listed in enclosure 2. also lists the capa-bilities of the BWKIN code. Core geometry, enriched fuel distribution in the core, control rod geometry, and poison rod gecmetries for 2WKIN calculations are shown in the figures (Subject No.'s 5, 6, 7, 8, 9 and
- 10) of enclosure 2.
The results of the steady-state axially averaged BkKIN pcwer distributions are comoared with PDQ calculations, and 3WKIN axial powe distributions are compared with FLAME calculations. The results incicate good agreement.
}2hb b0 7911050 1
L.E. Phillips OCT I ng As of this meeting, B&W has completed its s',1idy-state phase of code development which is prerequisit for development of the transient phase necessary for analysis of the steam line break accident. The transient phase of the development effort will be followed by a DNBR analysis.
B&W plans to complete the report on the steam line break analysis and submit it for staff review around February 1980.
The staff stated that a further meeting on the transient development effort appears appropriate at the con,clusion of '.his phase and prior to the documentation effort.
f
-/
- .1 S. Salah Analysis Branch Division of Systems Safety
.nclosures:
1.
Attendance !;st 2.
'liewgraphs
}2Ib bh
Attendance List Sectember 6,1979 NRC H. Richings L. Kopp S. Salah
- 0. Hood M. Dunenfeld L. Phillips M. McCoy 3&W J. Maxley R.D. Vosbueih J. Howard S. Sian R.L. Rees J. Rodes A. Gharakhani CPC0 J.J. Zabritski
<r
^?0 M.J. Salerno G.E. Clyde
y,y nVd '10'b3.1.N33 WC E
_Loarens
.00 00 JN 001V143 DJ N
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AGENDA
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INTRCDJCTIQ1 l
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i SHJimti 52SITIVIT(
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SLB STUCX F.0D PROGRN1 OBJECTIVES i
- STUDY TRANSIENT PEAK ASSOCIATED WITH MAXIMLM WORTH STUCK R0D DURING TRIP.
i 1
-- CONFIRM NO PRCPAGATioN IN THE LIMITED NLMSER OF P!NS I
IN DIl3 DURING TRIP.
1
- PROVIDE ANSWER To !!RC QUESTIONS 211.166, 211.163, 211.169, i
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AND 211.171.
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- 3) CacSSsECTICNS l
- 4) DELAYED NEUIRON FRACTICNS
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NO TEGERATURE, PETN. TEGE2ATLRE, NO AXIN. CCNTRCL RCD I
FRACTICNS
- 7) FAST NO DWAL IfiVERSE t;EJTRCN W!CCITIES B) TWR'AL PDTL'LIC t
- 1) It.tET R.CW RATE
- 2) ItLET eawY j
- 3) InITIrSTkYSTATEPChER I
- 4) INITIAL STEADY STATE PRESSI.FE
- 5) K-FMTCRS
- 6) FLCH TRANS!E.7 FCRCING FUCICN j
- 7) EWWY TR#iSIENT FCRCING FCCICN l
- 8) ERESSLRE TRN4SIEfT FGCIPE PJETICN i
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SUSJECT
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BWKIN SHUTDOWN MARGIN CCNSIDERATICX
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l I.
Ft.ux RsmsTRIsurzon II.
STucx Roo Au.cwAncs s
LHCERTA!NTIES ON CONTROL ROD CROSS SECTICNS e MXIEM ALLOWABLE INSERTED RCD INDEX j
e POWER DEFICIT, HFP TC lEP l
t l
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NO.
MAC E IN 'J. '3. A.
I i
CONTROL RCD CROSS I
SECTICN UtiCERTAINTIES i
i 2 10% 4e l
ROD WORTH 'JNCERTAINTY i
y CONTROL R0D ASSORPTION CROSS SECTICN UNCERTAINTIES r
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l oi
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l 12C5 '86 i
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J SU BJECT 3@ maca.yo. n-NO.*-
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SENSITIVITY OF HOT ASSEMBLY P0' DER PEAK TO ROD CROSS SECTICII UlCERTAINTIES l
1 l
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i ABSORPTION HOT ASSEMBLY RELATIVE CROSS SECTION RADIAL PEAx!No FACTCR*
CHANGE NOMINAL VALUES 3.109 BASE MAXIMUM VALUES **
3.115
+0.2%
MINiMun VALUES *"
3.097
-0.4%
i i
IT MAY BE CONCLUCED THAT THE RADIAL POWER PEAX OF HOT ASSEMBLY IS NOT SENSITIVE TO THE CONTROL RCD ABSORPTION l
1 CROSS SECTIONS.
l 4
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CC:iTF.0L RCD ASSORPTIO1 CROSS SECTICiS (STEADY STATE, HFP, H-14 STUCX OUT)
NCM!nAL Caoss SECT!cNS 1.755 1.656l1.055 1.775 1.S79 1.072 max!Mun Caoss SECT!aNS -
Mt91 mum Caoss SECTrans l
1.737 '
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0.445 0.504 U.443 0.867 0.d34 1.9W 3.109 2.674 0.436 0.589 0.430 0.852 0.821 1.975 3.115 2.682 l
0.456 0.6 19 0.456 0.880 0.845 1.978 3.097 2.663
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NOTE:
1/4 CoaE GECMETaY DISPLAYED.
PRCBLEM !s 1/2 Ccas SYMMETRIC).t.
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$U 3d.C7 CATALCG NO. 00-3M CENTER. ST. 3
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MIDLAND 1&2-FSAR TABLE 15.1-12 PARAMETERS APPLICABLE TO FTIEL FAILURE CASE FOR STCA:1 LLC BRL'AK (hCRST D:;3) l l
i Paraceter Assumotion t
l-l Break s:. c 33.5" double ended rupture outside contal.nment Power level 102 l
Time in life EOL or SQL offsite power Not available Process steam plant Not connected Number of RC pumps operating at ti=c cf rupture 4
t 12 C
'.8 9 l
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/
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g 120 inuraai 80 Peter 5 of 40 2501 nt 3
+l.0 Total 00 }
1.0 Inactivity.
2.0
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[
Temp, F
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- Pressure, 1333 psia 1400 A
1203
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5 la 15 20 25 30 35 40 45 50 55 80 Tlee, see CCllSU.!ERS PC','.'ER COMP Ai~
MIDLA !D PLA li UI'!TS 1 & 2 FlHAL SAFETY A."ALYSIS REPC.li i
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Stesmline 3resk, **crst DN3 Case e
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- 4200 Max. Clad Temp 2.5 i
1 1C00 3 40C0 MC.ssa 3
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.5 4
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r q 800
._ 3600 l
1.5 700 3400
.5 z
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FINAL SAFETY AllALYSIS REP 03T Steamline 3reak, '='orst ON3 C.sse rsAn ris;ure 13.1-a 3/73 Revision 7
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SUSJECT
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ANALYSIS l
l ANALYSIS 3
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CPCo LICENSING SCHEDULE
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