ML20062H073
| ML20062H073 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/31/1979 |
| From: | Brugge R, Ervin A GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20062H070 | List: |
| References | |
| 79NED-262, NEDO-24182, NUDOCS 7904170226 | |
| Download: ML20062H073 (28) | |
Text
.. .
NEDO-24182 I 79NED262 I, Class I March 1979 i
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} SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR
$ BRUNSWICK STEAM ELECTRIC PLANT l UNIT 2 RELOAD 2 I 1 ,
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Prepared: .[ MA A. M. Ervin, Engineer Operating Licenses II ,
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Approvedi k. , bg R. O. Brugge, Man ger Operating Licenses II s
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NuC:.s AA (NgaGy SGC.EC*3 Olyt$;CN
G E N E R A l. h. E l.E C T R I C '
. - - _ _ _ _ ._.___....__u- .__---------------.___ ------_.________._ . _ _ _ _ _ . _ . _ .
NED0-24182 -
i IMPORTANT NOTICE REGARDING i
CONTENTS OF THIS REPORT '
Please Read Carefully k l
l 1
l This report was prepared by General Electric solely for Carolina Power and Light
- Company (CP&L) for CP&L's ua n with the U.S. Nuclear Regulatory Commission (USNRC) for amending CP&L's operating license of the Brunswick Steam Electric Plant i Unit 2. The information contained in this report is believed by General Electric l to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting information in ,
this document are contained in the contract between Carolina Power and Light (
Company and General Electric Company for nuclear fuel and related services for the nuclaar system for Brunswick Steam Electric Plant, dated January 28, 1974, and nothing contained in this document shall be construed as changing said contract.
The use of this information except as defined by said contract, or for any purpose '
other than that for which it is intended, is not authorized; and with respect to ,
any such unathorized use, neither General Electric Company nor any of the con- i tributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information con-tained in this document or that such use of such information may not infringe privately owned rights; not do they assume any responsibility for liability or .
damage of any kind which may result from such use of such information. ,
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I j 1. PLANT UNICUE ITEMS (1.0)*
Rotated Bundle Analysis Procedure: Appendix A l Total Number and Capacity of Safety / Relief Valves: Reference 2 i
Fuel Loading Error LHGR: Appendix B l
- 2. RELOAD FUEL SUNDLES (1.0, 3.3.1 and 4.0) j Fuel Type Number Number Drilled i Irradiated Initial Core Type 1 108 103
- Initial Core Type 3 176 176 7DB230 4 4 8DB274L 100 100 SDB274H 40 40 New 8DRB265H 64 64 s, SDRB283 68 68 j Total 560 560
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle exposure: 11,570 mwd /t Assumeu reload cycle exposure: 13,080 mwd /t Core loading pattern: Figure 1
- 4. CALCULATED CCRE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 200C (3.3.2.1.1 and 3.3.2.1.2)
BOC k ett Uncontrolled 1.120 Fully Controlled 0.958
- (-- Strongest Control Red Out 0.989 I '-
R, Max 1=um Increase in Cold Core Reactivity j with Exposure Into Cycle, ak 0.000 5.
STEDBY LIOUID CCNrROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3) i .
Shutdown Margin (ak)
( EEE (20 C, Xenon Free) s 600 0.032 i *
- ( ) reterst:o areas of discussion in Reference 1.
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- 6. RELOAD UNIQUE TRANSIENT ANALYSIS IMPUTS (3.3.2.1.5 and 5.2) '
EOC Void Coefficient N/A* (C/* Rg) 7.46/9.49 t
Void Fraction (*) 41.76 Doppler Coefficient N/A (c/* F) 0.1937/0.1840 Average Fuel Te=perature ( F) 1538 Scram Worth N/A (S) 38.75/31.00 !
Scram Reactivity Figure 2
- 7. RELOAD UNIQUE CETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) I i
Peaking factors l
(local. radial [
4 and axial) 1.24/1.269/1.40 1.22/1.371/1.40 1.20/1.512/1.40 R-Factor 1.100 1.098 1.051 I i :
j Bundle Power 4
(MWC) 5.417 5.847 6.442 Bundle Flov (103 lb/hr) 124.91 114.98 115.73 Initial MCPR 1.21 1.27 1.27
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
None '
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- N = Nuclear Input Data A = Used in Transient Analysis i t
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- 9. CORE-WIDE TRANSIENT' ANALYSIS RESULTS (5.2.1)
{
. Power Flev 4 Q/A Psi Py ACPR Plant ,
l Transient Exposure () () () (%) (psig) (psig) 7x7 8x8/8x8R Response J Generator Load i Rejection w/o 8 '
Bypass 30C-EOC3 104 100 261.4 108.9.1168 1215 0.14 0.20 Figure 3 !
Inadvertent HPCI Pump -
Start ---
104 100 122.4 113.1 1018 1067 0.11 0.14 Figure 4 :
i Feedvater Controller Failure BOC-EOC3 104 100 109.2 105.1 1027 1076 0.06 0.07 Figure 5
( ~ /) 10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) t TRANSIENT
SUMMARY
(5.2.1) i Rod Position ,
Rod Block (Feet ACPR MLHGR (kW/ft) Limiting ,
Reading Withdrawn) 7x7 8x8 8x8R 7x7 _8g '8x8R Rod Pattern !
104 4.0 0.13 0.10 0.19 18.0 15.3 12.5 Figure 6 l 105* 4.0 0.13 0.10 0.19 18.0 15.3 12.5 Figure 6 ,
106 4.5 0.15 0.11 0.22 18.8 16.3 13.1 Figure 6 f 107 5.0 0.17 0.13 0.25 19.4 16.8 13.6 Figure 6 108 5.5 0.20 0.14 0.27 19.8 17.3 14.1 Figure 6 ,
109 6.0 0.22 0.16 0.29 20.0 17.5 14.4 Figure 6 l
. 110 9.0 0.24 0.24 0.36 18.2 16.5 14.3 Figure 6 l t
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- 11. OPERATING MCPR LIMIT (5.2) l
) BOC3 - EOC3 L
) . J L 1.27 (8x8/8x8R fuel) l 1.21 (7x7 fuel)
- 12. OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3)
Power Core Flow Ps1 Py Plant ,
Transient (t) (%) (psig) (psig) Response i
MSIV Closure
, (Flux Scram) 104 100 1213 1258 Figure 7
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- Indicates setpoint selected {
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, .c.- . . . . .. .--- ---- . . - - - - - - - . . - - - . - . . . _ . . . - -
- 13. STABILITY ANALYSIS RESULTS (5.4)
Decay Ratio: Figure 8 .
i Reactor Core Stability:
Decay Ratio, x2 /*0 *
(1052 Rod Line - Natural t Circulation Power) !
a Channel Hydrodyna=ic Perfor=ance Decay Ratio, x2 /*0 (1052 Rod Line - Natural Circulation Power) i l 8x8/8x8R channel 0.28 7x7 channel 0.13 ,-
I 14. LOSS-OF-COOLANT ACCIDENT RESUI.TS, (5. 5. 2 ) ,
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8DRB265 ;
Exposure MAPLEGR PCT Local Oxidation ,
(mwd /t) (kW/ft) ( F) Fraction 200 11.5 2154 0.030 1,000 11.6 2156 0.029 5,000 11.9 2192 0.032 10,000 12.0 2196 0.032 15,000 12.0 2200 0.033 20,000 11.8 2197 0.033 25,000 11.3 2138 0.027 30,000 10.7 2056 0.021 i
8DRB283
()
e I Exposure MAPLMGR PCT Local Oxidation (mwd /t) (kW/ft) (OF) Fraction $-
200 11.2 2122 0.027 l 1,000 11.2 2117 0.026 {
t 5,000 11.8 2184 0.032 l 10,000 12.0 2197 0.033 ;
15,000 11.9 2194 0.032 -
20,000 11.8 2197 0.033 25,000 11.3 2132 0.027
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30,000 11.1 2106 0.025 t
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i 15. LOADING ERROR RESULTS* (5.5.4, Appendix A) 1 i
Limiting Event: Rotated bundle 8DRB283H or 8DRB265H ,
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- -e MCPR
- 1.07**
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- 16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1) l Doppler Reactivity Coefficient: Figure 9 )
Accident Reactivity Shape Functions: Figures 10 and 11 l Scram Reactivity Functions: Figures 12 and 13
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$Using New Rotated Sundle Analysis Procedures described in Appendix A. >
[ ** Includes added penalty of 0.02 imposed by NRC.
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,i Figure 1. Reference Core Loac!ing Pattern l
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j Figure 4. Plant Response to Inadvertent Startup of IIPCI Purep
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o NOTES: 1. ROD PATTERN IS II4 CORE MIRROR SD! METRIC. i UPPER LETT QUADRANT SHOWN ON MAP I- 2. NUMBERS INDICATE NDGER OF NOTCHES WITHDRAWN -
CUT OF 48. BLANK IS A WI'HDRN#J RCD 4
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- ( 3. ERROR ROD IS (22,27)
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O soVNOING VALUE FOR 280 cal /g COLD Q 80UNQiNG VALUE FCR 290 cal /g MS8 6 CALCULATED VALUE - COLD / ;
h CALCULATED VALUE - H$8 g/}
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Figure 11. RCA Reactivity Shape Function a: 236 0 [
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i REFERE' ICES .
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- g 1. " General Electric Boiling Water Generic Fuel Application," NEDE-24011-P, j 3 Revision 3, March 1978. ;
] 2. Letter No. NG-77-1060 frem E. E. Utley (CP&L) to A. Schwencer (NRC),
September 20, 1977.
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i APPENDIX A NEW SUNDLE LOADING ERROR EVENT ANALYSES PROCEDURES
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i The bundle loading error analyses results presented in Section 15 in this Ija supple =ent are based on new analyses procedures for the rotated bundle loadine
- error events.
J The use of this new analysis procedure is discussed below.
A.1 .
.I NEW ANA*YSIS PROCEDURE FOR THE ROTATED SUNDLE LOADING ERROR EVENT t
The rr:ated bundle leading error event analyses resul:s presented in :his sup-t ple:ent are based on the new analyses procedure described in References A-1 and tl } A-2.
This new method of performing the analyses is based en a more detailed analysis model, which reflects more accurate analyses than cha: used in previous analyses of this event.
The principle dif ference be:veen the previous analyses procedure and the new analyses procedure is the modeling of the wa:er gap along the axial leng:h of the bundle.
The previous analyses used a uniform wa:er gap, whereas :he new analyses u:ilize a variable water gap which is representative of the actual condition. '
The effe::
of :he variable water gap is to reduce the power peaking and the 1
R-factor in the upper regions of the limiting fuel rod. .
i j
This results in the cal-culation of a reduced ACPR for the rotated bundle. The calculatien was perfor:ed j ,
h using :he sane analy:1 cal models as were previously used. The only change is in
{ ; }
3 -e the simulation of the water gap, which more accurately represents the actual ge:ne:ry.
I in the new analyses, the axial align =ent of a 180* rota:ed bundle conserva:ively ignerss the presence of the channel fas:ener. The more limi:ing condition of 5
assuming tha:
the spacer buttons are in contact with the top guide is assumed.
j There is no known loading that could bend or break the channel sp.ieer bu::en j l
\. during the inser: ion of a 180* rotated bundle, since both the top guide ind l spacer bu::on are chamfered to provide lead-in.
j For a properly assembled bundle, no mechanist.
' exis:s which could invalidste the assumption that a 130' rotated 1 bundle leans to one side.
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A-1
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JEDO-24182 !
I: should be noted :ha: proper orien:a:ica of bundles in the reactor core is readily verified by visual observa: ion and assured by verifica ion procedures during core loading. Five separa:e visual indications of proper bundle '
crien:atien exis:: '
t (1) The channel fas:ener asse=blies, including :he spring and guard used [
to main:ain clearances be: ween channels, are located at one corner of !
e each fuel asse=bly adjacent :o the cen:er of the control red. i i.
(2) The identification boss on the fuel asse=bly handle poin:s :evard :he I adjacen: con:rol red, i t
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(3) The channel spacing buttons are adjacen: to the control rod passage 6 l
area. %. */ <
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(:) The assembly iden:1 fica: ion numbers which are located on the fuel )
asse:bly handles are all readable from :he direc:ica of the cen:er of L the cell.
(5) There is cell-:o-cell replica: ion.
Experience has demonstrated that these design features are clearly visible se that any misloaded bundle would be readily identifiable during core loa;!ing verifica: ion. 4 Figures A-1, A-2 and A-3 denote a nor ally loaded bundle, a leC' ~7 rota:ed bundle, and a 90* ro:a:ed bundle, respec:ively. Actual experience N,) - e (References A-1 and A-2) has demonstrated tha: the probability of a rotated 5
{
bundle is lov.
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The new analyses procedure results show that the minimum CPR for the mos: limit- i
(
ing rotated bundle in the core is greater than the safety limit. j; r.:
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A-2
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. NED0-24182 REFERENCES 4
i A-1 Letter, R. E. Engel (CE) to D. Eisenhut (NRC), " Fuel Assembly Leading 1
Errer," ME;-219-77, June 1, 1977 i A-2 Letter, R. E. Engel (GE) to D. Eisenhut (NRC), " Fuel Assembly Leadi.;
Errer," Mni-i.57-77. Neve=ber 30, 1977 6'
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NOTE. BUNCLE NUMBERS ARE FOR f LLUSTR ATivE PumPOSES ONLY f
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!'t Figure A-1. Normal Leadi::g f i
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--s i. i Il NOTE SUNOLE NUMBERS AAE FCA ILLUSTRATIVE PURPOSES CNLY ll 9
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Figure A-2. Rotated Sundic, 180 Degree Rotation
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1 Figure A-3. Rotated Sundle, 90 Degree Rotation ,
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- it NED0-24182 I i
- 4
- APPENDIX 3 ii Fuel Loading Error LHGR
- 15.5 PUft f
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5-1/3-2
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