ML20064B984
| ML20064B984 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 12/31/1982 |
| From: | Browning R, Charnley J, Hill R GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19262G943 | List: |
| References | |
| Y1003J01A54, Y1003J1A54, NUDOCS 8301040266 | |
| Download: ML20064B984 (28) | |
Text
Y1003J01A54 Revision 0 Class I December 1982 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PEACH BOTTOM ATOMIC POWER STATION UNIT 3, RELOAD 5 Prepared: - ~ - - ~/
R. A. Browning Licensing Engineer
/
Verified:
R. T. Hill Senior Licensing Engineer
- dd Approved: _J.g. Charnley [/
Fuel Licensing Manager V l
l l
l NUCLEAR POWER SYSTEMS DIVISION
- GENERAL ELECTRIC COMPANY I
SAN JOSE, CAltFORNIA 95125 GENERAL $ ELECTRIC 8301040266 821230 PDR ADOCK 05000278 P I PDR
Y1003J01A54 Rev. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Philadelphia Electric Company (PECo) for PECO's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending PECo's operating license of the Peach Bottom Atomic Power Station Unit 3. The information contained in this report is believed by General Electric 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 " Contract between Philadelphia Electric Company and General Electric Company for Fuel Bundles and Services for Reload Fuel Supply for Peach Bottom Atomic Power Station Units 2 and 3, dated November 16, 1979," and nothing contained in this document shall be construed as changing said contract. The use of this information except as I defined by said contr.ict, or for any purpose other than that for which it is
- intended, is not authorized; and with respect to any such unauthorized use, l neither General Electric nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.
I 11
Y1003J01AS4 Riv. 0
- 1. PLANT UN1QUE ITEMS (1.0)*
Appendix A - Description of Peach Bottom Lead Test Assemblies (PBLTA's)
Appendix B - Rotated Bundle Error Results for PBLTAs Appendix C - Pressurized Test Assembly (PTA) Fuel Rod Replacement Appendix D - Single Loop Operation
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 and 4.0)
Fuel Type Cycle Loaded Number Number Drilled Irradiated P8DRB284H 4 263 263 P8DRB299 5 216 216 PTA 2 1 1 New P8DRB284H 6 56 56 P8DRB299 6 224 224 PBLTA 1 6 2 2 6 2 2 PBLTA 2 Total 764 764
+3. REFERENCE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at end of cycle: 19.0 GWd/ST Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 18.8 GWd/ST Assumed reload cycle core average exposure at end of cycle: 18.2 GWd/ST Core loading pattern: Figure 1
- ( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A (latest approved revision); a letter "S" preceding the number refers to the appropriate country-specific supplenent.
1
Y1003J01A54 Riv. 0
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTDi WORTH - NO VOIDS, 20*C (3.3.2.1.1 and 3.3.2.1.2)
Beginning of Cycle, K-effective ,
Uncontrolled 1.113 Fully Controlled 0.966 Strongest Control Rod out 0.989 R, Maximum Increase in Cold Core Reactivity with Exposure into Cycle, Delta K 0.001
- 5. STANDBY LIQUID CONTROL. SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (ak) ppm (20"C, Xenon Free) 660 0.029
- 6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 and S.2.2)
(REDY EVENTS ONLY)
EOC EOC - 2.0 GWd/ST Void fraction (%) 39.8 39.8 Average Fuel Temperature (*F) 1278 1278 Void Coefficient N/Aa (c/% Rg) -8.14/-10.17 -9.05/-11.32 Doppler Coefficient N/Aa (c/ F) -0.233/-0.221 -0.224/-0.213 Scram Worth N/A" (S) l "N = Nuclear Input Data A = Used in Transient Analysis b
Generic, exposure independent values are used as given in " General Electric Boiling Water Reactor Generic Reload Fuel Application," NEDE-240ll-P-A-2, July 1981.
2
Y1003J01A54 Rev. 0
- 7. RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)
Fuel Peaking Factors R- Bundle Bundle Flow Initial Design Local Radial Axial Factor Power (MWT) (1000 lb/hr) MCPR Exposure: EOC P8x8R/PTA 1.20 1.45 1.40 1.051 6.105 112 1.34 PBLTA2a 1.25 1.30 1.40 1.106 5.462 117 1.33 PBLTAla 1.25 1.32 1.40 1.106 5.537 126 1.34 Exposure: E0C-2.0 GWd/t P8x8R/PTA 1.20 1.52 1.40 1.051 6.385 110 1.27 PBLTA2" 1.25 1.36 1.40 1.106 5.728 116 1.27 PBLTAl" 1.25 1.38 1.40 1.106 5.801 124 1.27
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)
Transient Recategorization: No Recirculation Pump Trip : No Rod Withdrawal Limiter : No Thermal Power Monitor : No Improved Scram Time : No Exposure Dependent Limits : Yes Exposure Points Analyzed : EOC E0C-2.0 GWd/ST "See Appendix A for PBLTA description 3
Y1003J01A54 Rev. 0
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)
A CPR Flux, Q/A, PTA/
Transient % NBR % NBR P8x8R PBLTA2 PBLTAl Figore Exposure: BOC to EOC - 2.0 GWd/ST Load Rejection w/o Bypass 641 123 0.20 0.20 0.20 2a Loss of 100*F Feedwater Heating 125 124 0.16 0.15 0.15 3 Feedwater Controller Failure 268 119 0.12 0.12 0.12 4a Exposure: EOC - 2.0 GWd/ST to EOC Load Rejection w/o Bypass 695 128 0.26 0.26 0.27 2b Loss of 100*F Feedwater Heating 125 124 0.16 0.15 0.15 3 Feedwater Controller Failure 322 124 0.19 0.19 0.19 4b
- 10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)
SUMMARY
(S.2.2.1)
Limiting Rod Pattern: Figure 5 Includes 2.2% Power Spiking Penalty: Yes .
^
Rod Position Rod Block (feet PTA/ PTA/
Reading withdrawn) P8x8R PBLTA2 PBLTAl P8x8R PBLTA2 PBLTAl 104 4.0 0.15 0.15 0.15 18.71 18.71 18.71 105 4.0 0.15 0.15 0.15 18.71 18.71 18.71 106 4.5 0.17 0.17 0.17 19.07 19.07 19.07 107* 5.0 0.19 0.19 0.19 19.20 19.20 19.20 108 5.5 0.20 0.20 0.20 19.20 19.20 19.20 109 7.0 0.24 0.24 0.24 19.20 19.20 19.20 110 12.0 0.31 0.31 0.31 19.20 19.20 19.20 1
- Indicates setpoint selected j 4
Y1003J01A54 Rev. 0
- 11. CYCLE MCPR VALUES (S.2.2)
Non-Pressurization Events PTA/
Exposure Range: BOC to EOC P8x8R PBLTA2 PBLTAL Loss of 100*F Feedwater Heating 1.23 1.22 1.22 Fuel Loading Error 1.22 *
- Rod Withdrawal Error 1.26 1.26 1.26 Pressurization Events Option A Option B PTA/ PTA/
P8x8R PBLTA2 PBLTAl P8x8R PBLTA2 PBLTAl Exposure Range:
BOC to EOC - 2.0 GWd/ST Load Rejection w/o Bypass 1.33 1.33 1.33 1.12 1.12 1.12 Feedwater Controller Failure 1.24 1.24 1.24 1.18 1.18 1.18 Exposure Range:
EOC - 2.0 GWd/ST to EOC Load Rejection w/o Bypass 1.39 1.39 1.40 1.27 1.27 1.28 Feedwater Controller Failure 1.32 1.32 1.32 1.25 1.25 1.25
- 12. OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3) si v Plant Transient (psig) (psig) Response MSlV Closure (Flux Scram) 1245 1276 Figure 6
- See Appendix B for PBLTA Rotated Bundle Results 5
, 7 Y1003J01A54 2 .
Rev. 0
, , o-
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- 13. STABILITY ANALYSIS RESULT (S.2.4) ' ' '
i
?
,',s Rod Line Analyzed: 105% Rod Line -T r
Decay Ratio:
- Figure )
i Reactor Core Stability Decay Ratio.'x2/*0 '- *0.37 .
Channel ifydrodynamic Performance Decay Ratio, x2/ *0 -# '
Channel Type
./
P8x8R/PTA -
/
- ,'C.29 ,
PBLTA1 . O.12 '
PBLTA2 , , 0.27 ,
1 I
_l
' v*
- 14. IOADINC ERif0R RESULTS (S.2.5.4) ,/
i !
, /
Variable Water Cap Misoriented Juadle' Analysis: \es *~-
Includes 2.2% Power Spiking Penalty: Yes ' '
? -s* ,
~
Event Initial CPR Re s ul t in g - G.' R Resulting LHGR 1.22 ')7.5 Rotated Bundle Error l.07 1S. CONTROL ROD DROP ANALYS1S itESULTS (S . L 5.1) ,
_~
n
~
i ,
Bounding Analysis Results: ,
Doppler Reactivity Coefficient: Figure 8 '
- Accident Reactivity Shape Functions: Figures 9 and 10 Scram Reactivity Functions: Figures 11 and 12 Plant Specific Analysis Results: /
Parameter (s) not Bouw!ed , Cold : Scram Reactivity Resultant Peak Enthilpy,' Cold: 164.0 ' -
I Parameter (s) not Bc unde J , -IISB : None ,, ,
~
- 16. LOSS-OF-COOLANT ACCIDENT'RE.SU11 ( S .' . 5. 2 )
See " Loss-of-Coolant Acciden; Analysis for Peacl. Bot tom Atomii[iawer Station '
l'n i t 3," General Electris- C rapany, December 1977 (hEDO-24082, Is am[nded) .
[ +-
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FUEL TYPE A= P80R8284H E = P80RB294H 8 = P8DRE299 F = P90R8299 C= P800833BLTA G= PSOR8200 D= P80R8294H H= PTA Figure 1. Reference Core LoadAng Pattern 7
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Y1003J01A54 Rev. 0 02 06 10 14 18 22 26 30 59 55 24 14 30 51 24 6 47 24 6 8 6 43 24 4 24 39 14 8 32 35 6 24 0 31 30 6 32 22 NOTES: 1. R0D PATTERN IS 1/4 CORE SYFDETRIC.
UPPER LEFT QUADRANT SHOWN ON MAP.
- 2. NUliBER INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48. BLANK IS A WITHDRAWN ROD.
- 3. ERROR ROD IS (26,35).
Figure 5. Limiting RWE Rod Pattern 13
Y1003J01A54 Rev. 0
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Y1003J01AS4 Rev. O A NA1 URAL CIRCU .ATION 8 1 05 PERCENT R )D LINE C ULl . PERFORMA 4CE LIMIT 1.00 E A
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- 0. 0 20.0 40.0 60.0 80.0 100.0 120.0 PERCENT POWER Figure 7. Reactor Core Decay Ratio versus Power 15
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0 CALCULATED VALUE - HS8 C BOUND VAL 280 CAUG COLD D BOUND VAL 280 CAUG HSB 0 500 1000 1500 2m 2500 3000 FUEL TEMPERATURE (OC)
Figure 8. Doppler Reactivity Coefficient Comparison for RDA 16
Y1003J01A54 Rsv. O 0 o20 0015 - - - -
gCC 6 A A A t 0010 2
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A ACCIDENT FUNCTION I
B BOUNDING V ALUE 280 CAL /G 0,i O 5 10 15 20 ROu POSITION (ft withdrecen)
Figure 9. RDA Reactivity Shape Function at 20*C 17
Y1003J01A54 Rev. 0 0.020 b
0 015 ,
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A ACCIDENT FUNCTION 8 BOUNDING V ALUE 280 CAUG O .i O 5 10 15 20 ROD POSITION (f t withdrawn)
Figure 10. RDA Reactivity Shape Function at 286 C 18 l 9
Y1003J01A54 Rev. 0 0.030 A SCRAM FUNCTION B SOUNDING VALUE 200 CAL /G 0.025 0.020 m 9 *
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Figure 11. RDA Scram Reactivity Function at 20*C 19
Y1003J01AS4 Rev. O o.os , ,
B BOUNDING VALUE 280 CAUG A SCRAM FUNCTION O 04 0 03 r 2
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Y1003J01A34 Rev. O APPENDIX A DESCRIPTION OF PEACH BOTTOM LEAD TEST ASSEMBLIES (PBLTAs)
Two of the LTAs utilize an improved pressure drop spacer (low-aP spacer),
while the remaining two LTAs have the normal spacer found on 8x8R fuel.
The low- AP spacer LTAs are designated PBLTAl in this document and the normal spacer LTAs are designated PBLTA2.
For a detailed description of the 4 LTA bundles see Reference A-1.
REFERENCE A-1. Letter from S. L. Daltroff (Philadelphia Electric Company) to John F. Stolz (U.S. Nuclear Regulatory Commission), " Peach Bottom Unit 3 Lead Test Assemblies," December 1982.
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Y1003J01AS4 Rev. O Appendix B ROTATED BUNDLE ERROR RESULTS FOR PBLTAs*
Variable Water Gap Misoriented Bundle Analysis: Yes Includes 2.2% Power Spiking Penalty: Yes Bundle Initial CPR Resulting CPR Resulting LHCR PBLTA1 1.25 1.06 17.9 PBLTA2 1.27 1.07 17.9
- The rotated bundle error for the PBLTAs does not affect the limits due to special loading surveillance requirements to preclude a misoriented bundle.
Following core verification conducted by PECO, a second independent verifica-tion will be conducted to assure that the PBLTAs are not mislocated or misoriented.
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Y1003J01A54 Rev. O Appendix C PRESSURIZED TEST ASSEMBLY (PTA) FUEL ROD REPLACEMENT C.1 INTRODUCTION The Pressurized Test Assembly (PTA) was described in NED0-21363-4, Supple-ment 4. January 1977.
As a continuation of this program, during the Reload 5 refueling outage, up to 20 fuel rods will be removed from the PTA and replaced with irradiated rods from an 8DRB283 bundle (initially inserted as part of Reload 2) due to be dis-charged at EOC-5. The rods removed from the PTA will be examined and punctured for fission gas pressure measurement. These rods will not be used during future operation. The enrichment of the replacement rods was selected to assure that the reactivity of the reconstituted PTA will not exceed that of a nonreconstituted PTA. Therefore, the information contained in NEDO-21363-4 Supplement 4 is unaf fected by the fuel rod replacements of the PTA since the nuclear characteristics of the reconstituted bundle are essentially identical to a nonreconstituted bundle. The purpose of this appendix is to report the results of the analyses and safety evaluation for operation of the reconsti-tuted PTA during Cycle 6.
C.2 EVALUATIONS AND ANALYSES Standard lattice physics calculations were made for the reconstitutcJ PTA including simulation of the replacement rods. Cycle 6 operation was simulated by " burning" the reconstituted PTA in 8 exposure steps.
Over the exposure range of interest, the computed lattice reactivity of the reconstituted PTA is within 0.4% AK of the previously reconstituted PTA reactivity. The fuel rod power peaking for the reconstituted PTA is lower than that of the bundle prior to this reconstitution at the same exposures.
Since the local peaking is decreased as a result of this reconstitution, the reconstituted PTA will be no more limiting than the PTA at the same exposure prior to this reconstitution.
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Reconstitution of the PTA may result in a slight increase in PCT due to stored energy and local power distribution effects. The resulting increase ir Peak Cladding Temperature (PCT) on account of these effects is insignificant; com-pared to the margin to the PCT limit of 2200*F. Since the PCT of the non-reconstituted PTA is considerably less than 2200*F (i.e., the maximum PCT equals 1923*F), the PCT of the reconstituted PTA will also remain well below 2200*F. During Cycle 6 the exposure of the PTA is expected to exceed the maxi-mum value analyzed in the previous LOCA analysis. Thus, MAPLHGR values were calculated for the extended exposure points for Cycle 6 and are reported in Reference C-1. Since the peak linear hest generation rate of the reconstituted PTA is within the operating limit of 13.4 kW/f t, which was used in evaluating the mechanical performance of the maximum duty fuel rod, the results of the fuel rod thermal and mechanical design evaluations in NEDO-21363-4, Supple-ment 4, are conservatively applicable to the reconstituted PTA.
REFERENCE C-1. " Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 3," Ceneral Electric Company, December 1977 (NEDO-24082, as amended).
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Y1003J01A54 Rev. O Appendix D SINGLE LDOP OPERATION The single loop operation analysis, Reference D-1, has been verified for applicability to Cycle 6.
REFERENCE D-1 " Peach Bottom Atomic Power Ststion Units 2 and 3 Single Loop Operation,"
General Electric Company, May 1980 (NED0-24229-1) .
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