ML19209C234
| ML19209C234 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 08/31/1979 |
| From: | Brugge R, Ervin A, Rash J GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19209C231 | List: |
| References | |
| NEDO-24182A, NEDO-24182A-R1, NUDOCS 7910120233 | |
| Download: ML19209C234 (28) | |
Text
{{#Wiki_filter:.* . NEDO-24182A 79NED300 Class I August 1979 Revision 1 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 RELOAD 2 REVISION 1: REVISED JRANSIENT AND GETAB ANALYSES Pre pared : !/ WA P pa e A. M. Ervin, Engineer . Operating Licenses II nior Engineer Fuel and Services Licensing
,m n Approved: k b' ,b o ' c)(/ Approved: y R. O. Brugge, Mahsger R. E. Engel, Manager Operating Licenses II Reload Fuel Licensing 1144 324 NUCLE AR E NE aGY P A C.ECTS Div:SiCN OENE A A L E LE CT A.C CCYpa'.Y SAN .CSE. CAL FC ANI A 951:5 GENER Alh ELECTRIC 3910120
NED0-24182 IMPORTANT NOTICE REGARDING CONTENTS OF TlilS REPORT
?lcace Read Carefully This report was prepared by General Electric solely for Carolina Power and Light Company (CP&L) for CPSL's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending CP&I's operating license of the Brunawick Steam Electric Plant Unit 2. The information contained in this report is believed by General Electric to be an accurate and true representatie of the facts known, obtained or provided to General Electric at the time this re, et 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 nuclear 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-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 privatelv owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information. M [3
NEDO-24182A
- 1. PLANT-UNIQUE ITEMS (1.0)*
Rotated Bundle Analysis Procedure: Appendix A Total Number and Capacity of Safety / Relief Valves: Reference 2 Fuel Loading Error LHGR: Appendix B
- 2. RELOAD FUEL BUNDLES (1.0, 3.3.1 and 4.0)
Fuel Type Number Numter Drilled Irradiated Initial Core Type 1 108 .' 0 8 Initial Core Type 3 176 176 7DB230 4 4 8DP274L 100 100 SOB 274H 40 40 New 8DRB265H 64 64 8DRB283 68 68 Total 560 560
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle exposure: 11,570 mwd /t Assumed reload cycle exposure: 13,080 mwd /t Core loading pattern: Figure 1
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM
'a' ORTH - 50 VOIDS , 200C (3.3.2.1.1 and 3.3.2.1.2)
BOC kel_f Uncontrolled 1.120 Fully Controlled 0.958 Strongest Control Rod Out 0.989 R, Maxi =um increase in Cold Core Reactivity with Exposure Into Cycle, 2k 0.000
- 5. STANDBY LIOUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (ak) Eem (20 C, Xenon Free) 600 0.032
*( ) refers to areas of discussion in Reference 1.
1144 Jikfi 1
NEDO-24182A 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 and 5.2) EOC Voie Coefficient N/A* (c/% Rg) 7.89/9.86 *** Void Fraction (%) 41.76 Doppler Coefficient N/A (c/: F) 0.1937/0.1840 Average Fuel Temperature ( F) 1538 Scram I? orth N/A ($) 38.75/31.00 Scram Reactivity Figure 2 7. RELOAD-UNIQUE CETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS Exposure
- 0*b
- EOC EOC EOC Peaking factors (local, radial and axial) 1.24/1.260/1.40 1.22/1.349/1.40 1.20/1.493/1.40 ***
R-Factor 1.100 1.098 1.051 Bundle Power (>r.;t ) 5.379 5.752 6.362 *** Bundle Flow (103 lb/hr) 125.26 115.54 115.19 *** Initial MCPR 1.22 1.29 1.28 *** 8. SELECTED StARGIN IMPROVEMENT OPTIONS (5.2.2) None
*N = Nuclear Input Data A = Used in Transient Analysis 4
3-Q'
- Denotes change from Rev. 0 }kk 2
NEDO-24182A
- 9. CORE-WIDE TRANSIENT' ANALYSIS RESULTS (5.2.1)
Power Flow : Q/A Psi Py aCPR Plant Transient Exposure (%) () (%) (%) (psig) (ps12) 7x7 8x8/8x8R Response Generator Load Rejection v/o Bypass BOC-EOC3 104 100 269.9 109.5 1169 1216 0.15 0.22 Figure 3 *** Inadvertent HPCI Pump Start --- 104 100 122.4 113.1 1018 1067 0.11 0.14 Figure 4 Feedwater Controller Failure LOC-ECC3 104 100 109.0 105.1 1028 1076 0.05 0.06 Figure 5 ***
- 10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)
TRANSIENT
SUMMARY
(5.2.1) Rod Position Rod Block (Feet aCPR MLHCR (kW/ft) Limiting Reading Withdrawn) 7x7 8x8 8x8R 7x7 8x8 8x8R Rod Pattern 104 4.0 0.13 0.10 0.19 18.0 15.3 12.5 Figure 6 105* 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 6 107 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 6 110 9.0 0.24 0.24 0.36 18.2 16.5 14.3 Figure 6
- 11. OPERATING MCPR LIMIT (5.2)
BOC3 - EOC3 1.29 (8x8/8x8R fuel) *** 1.22+ (7x7 fuel) ***
- 12. OVERPRESSL".tIZATION ANALYSIS
SUMMARY
(5. 3) Power Core Flow Psi Py Plant Transient (%) (%) (psig) (psig) Response MSIV Closure (Flux Scram) 104 100 1214 1259 Figure 7 *** 1144 328
- Indicates setpoint selected.
*** Denotes change from Rev. O.
+If scoop tube block is set at 102.5% flow and 112% flow K curve g is used.
3
NED0-24182 . 13. S.' ABILITY ANALYSIS RESULTS (5.4) Decay Ratio: Figure 8 Reacter Core Stability: Decay Ra t io , 0.62 x2!*O (105: Rod Line - Natural Circulation Power) Channel Hydrodynamic Performance Decay Ratio, x /x 3 0 (105% Rod Line - Natural Circulatier. Power) 8x3/8x3R channel 0.28 7x7 channel 0.13
- 14. . LOSS-OF-COOLANT ACCIDENT RESULTS. (5.5.2) 8DRB265 Exposure MAPLHCR PCT Local 0xidation (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 '.2.0 2200 0.033 20,000 11.8 2197 0.033 25,000 11.3 2138 0.027 30,000 10.7 2056 0.021 8DRB283 Exposure MAPLEGR PCT Local Oxidation (mwd /t) (kW/ft) ( F) Fraction 200 11.2 2122 0.027 1,000 11.2 2117 0.026 5,000 11.8 2184 0.032 10,000 12.0 2197 0.033 15,000 11.9 2194 0.032 20,000 11.3 2197 0.033 25,000 11.3 2132 0.027 3C,000 11.1 2106 0.025 4
))hh )
NEDO-2 4182A
- 15. LOALING ERROR RESULTS* (5.5.4, Appendix A)
Limiting Event: Rotated bundle 8DRB283 or 8DRB265H MCPR: 1.07**
- 16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and 13
*Using New Rotated Bundle Analysis Procedures described in Appendix A.
** Includes added penalty of 0.02 imposed by NRC.
i144 330
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C1 C3 05 07 09 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 Fuel Type A=Init Core Type 1 E=d D B 2 7411 B-Init Core Type 3 F=8DRB26Sil C-Gen B (7DB230) G=8DRB283 D=8DB274L Figure 1. Reference Carc Laading Pattern e 1144 331
NEDO-24182 iOO 45 C C - 678 CRD IN PERCENT 1 - NOMINAL SCRAM CURVE IN 1-5)
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2 - SCRAM CURVE USED IN ANALYSIS _ 4o 1 80 -
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Scram Reactivity and Centrol Red Drive Specifications 1144 332
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TIME (SEC) 20. 30. 40. TIME (SEC) O 2 1ll N 1 LEVEL (INdH-REF-SEP-SKIRT
- 2 VESSEL SlEAMFLCW 1 VOID RE4d! 7 w 150-3 TureINE 5 TEAMFL CW--
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NEDO-24182 02 06 10 14 la 22 26 30 51 47 43 34 6 39 32 1; 34 4 2 31 28 27 6 2 0 23 NOTES: 1. ROD PATTERN IS 1/4 CORE MIRROR SYMMETRIC. LTPER LEFT QUADRANT SHOWN ON }dR
- 2. NUMBERS INDICATE NUBBER OF NOTCHES WITFDRAW'i OUT OF 48. BLANK IS A WITHDRAWN ROD
- 3. ERROR ROD IS (22,27)
Figure 6. Limiting RWE Red ?attern 1144 336
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2 VESSEL SlEAMFLOH 1 VOID RER(J dITT 200* 3 TURBINE FTEAMFlOW 20 lEACTIVITY GEE 5iATD TE0W )* 3S RFr2TIVITT S r cIIV1IY 100.
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Plant Response to MSIV Closure J N
NEDO-24182 12 ULTIMATE STABILtTY LIMIT 10 ------------- --o- -- 08 -
=
D N 5 0
? < 0.6 -
U
; N ATUR AL CIRCU L ATIO*.
04 - 105% ROD LINE O2 - 1 I I I OO-0 20 40 60 80 100 POWER 1%I Figure 8. Decay Ratio r o 1144 338
NEDO-24182 0
-5 -
-10 -
s g -15 - 52 i w 8 5 g -20 - C
- ~
O SOUNDING VALUE FOR 280 al/g COLO C BOUNDING VALUE FOR 280 cal /g HS8 d CALCULATED VALUE - COLO h CALCULATED VALUE - HS8
-30 V
_3 I l i l ! O 400 800 1200 1600 2000 2400 FUEL TEMPER ATURE ( Cl Figure 9. Doppler Reactivity Coefficient Comparison for RDA a)
NEDO-24182 24 O CALCULATED VALUE O BOUNDING VALUE FOR 280 cai/g 20 - 16 - G I R O 5 0 12 - g 5 P-~ v 4 w 8 - 4 - g i l l 1 0 4 8 12 16 20 RCD POSITION ift OUT) Figure 10. RDA Reactivity Shape Function 1144 340 15
NEDO-24182 28 24 - 20 - Ci O 2 16 s O N t o D 12 - N e 8 I O BOUNDING VALUE FOR 280 cal /g 4 - O CALCULATED VALUE I I I oc I O 4 8 12 16 20 ACO POSITION (f t CUT) Figure 11. RDA Reactivity Shape Function at 286 C
,p<
./
16
NEDO-24182 90 0 BOUNDING VALUE FOR 280 cal /g 80 - O CALCULATED VALUE 70 - 60 6 ) z
.j 50 -
3 9 C A
*3 I
Y C 4 2 0 3 30 - 20 - 10 -
, I I I O 2 4 6 8 10 ELAPSED TIME (sec)
Figure 12. RDA Scram 3eactivity Function at 20 C 1144 342
s . NEDO-24182 140 O BOUNOING VALUE FOR 280 cal /g O CALCULATED VALUE 120 - 100 - G 1 c j 80 - E 9 .G. w
- 1 1
t > 60 - E; 5 = 40 - OC I l I O O 2 4 6 8 to ELAPSED TIME !sec) Figure 13. RDA Scram Reactivity Function at 286 C 13 1144 343
NEDO-24182 REFERENCES
- 1. " General Electric Boiling Water Generic Fuel Application," NEDE-240ll-P, Revision 3, March 1978.
- 2. Letter No. NG-77-1060 from E. E. Utley (CP&L) to A. Schwencer (NRC),
September 20, 1977.
" 1144 344 19/20
APPENDIX A NEk' BUNDLE LOADING ERz'OR EVENT ANALYSES PROCEDURES The bundle loading error analyses results presented in Section 15 in this supplement are based on new analyses procedures for the rotated bundle loadinf error events. The use of this new analysis procedure is discussed below. A.1 NEh ANALYSE 5 PROCED'JRE FOR THE RCTATED BUNDLE LOADINC ERROR EVENT Ine rotate; bandle loading error event analyses results presented in this sup-plement are based on the new analyses procedure described in References A-1 and A-2. This new method of performing the analyses is based on a more detailed analysis model, which reflects more accurate analyses than that used in previous analyses of this event. Tne principle difference between the previous analyses procedure and the new analyses procedure is the modeling of the water gap along the axial length of the bundle. The previous analyses used a uniform water gap, whereas the new analyses utilize a variable water gap which is representative of the actual condition. The effe:: of the variable water gap is to redu.e the power peaking and the R-factor in the upper regions of the limiting fuel rod. This results in the cal-culation of a reduced aCPR for the rotated bundle. The calculation was performed. using the same analytical models as were previously used. The only change is in the simulation of the water gap, which more accurately represents the actual ge: etry. In ::le new analyses, the axial alignment of a 180* rotated bundle conservatively 2;ncres the presence of the channel fastener. The more limiting condition of a s s u.- i n g that the spacer buttons are in contact with the top guide is assumed. There is nt known loading that could bend or break the channel spacer button during the insertion of a 180* rotated bundle, since both the top guide anJ spacer button are chamfered to provide lead-in. For a properly assembled bundle, no mechanisr. ex'.sts which could invalidate the assumption that a 180* rotated bundle leans to .ne side. A-1 1144
*
- 515
NEDO-24182 It should be noted that proper orientation of bundles in the reactor core is readily verified by visual observation and assured by verification procedures during core loading. Five separate visual indications of proper bundle orientation exist: (1) The channel fastener assemblies, including the spring and guard used to maintain clearances between channels, are located at one corner of each fuel asse=bly adjacent to the center of the control rod. (2) The iden ification boss on the fuel assembly handle points toward the adjacent control rod. (3) The chanrel spacing buttons are adjacent to the control rod passage area. (') The assembly identification numbers which are located on the fuel assembly handles are all readable from the direction of the center of the cell. (5) There is cell-to-cell replication. Experience has demonstrated that these design features are clearly visible se that any misloaded bundle would be readily identifiable during core loading verification. Figures A-1, A-2 and A-3 denote a normally loaded bundle, a 180* rotated bundle, and a 90' rotated bundle, respectively. Actual experience (References A-1 and A-2) has demonstrated that the probability of a rotated bundle is 1cw. The new analyses procedure results show that the minimum CPR for the most limit-ing rotated bundle in the core is greater than the safety limit. A-2 1144 345
NEDO-24182 REFERENCES A-1 Letter, R. E. Engel (CE) to D. Eisenhut (NRC), " Fuel Assembly Loading Error," M7N-219-77, June 1, 1977. A-2 Letter, R. E. Engel (GE) to D. Eisenhut (NRC), " Fuel Assembly Leadia-
~ ""*
Error, MFN-l.57-77, November 30, 1977 1144 347 A-3
NEDO-24182 II O'gjo s /N f <*4, O 'O O' O
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O,/N.O NOTE BUNDLE NUMBERS ARE FOR ILLUSTRATIVE PURPOSES ONLY Figure A-1. Normal Loading
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A-4
{. NEDO-24182
\f 180 ROTATeON - 8 iNDLE LJ 2348
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O[lO NOTE BUNOLE NUMBERS ARE FOR ILLUSTRATIVE PURPOSES ONLY Figure A-2. Rotated Bundle, 180 Degree Rotation 1144 349 A-5
NEDO-24182 O' O'\ 90' ROTATION - BUNOLE LJ 2348 f = f 0S s# O 'O O 'O
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O[lO NOTE. 8UNOLE NUM8ER$ ARE FOR IL LUSTRATIVE PURPOSES ONLY Figure A-3. Rotated Bundle, 90 Degree Rotation A-6 })kk JU
.'- 'A ' NEDO-24182 APPENDIX B Fuel Loading Error LHGR: 15.5 kW/ft
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