ML20093E222
| ML20093E222 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 03/31/1984 |
| From: | Verbryke P GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20093E217 | List: |
| References | |
| 23A1694, 23A1694-R, 23A1694-R00, NUDOCS 8407170425 | |
| Download: ML20093E222 (20) | |
Text
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I 23A1694 MAR 19 4 l
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GENERAL ELECTRIC BOILING WATER REACTOR SUPPLEMENTAL RELOAD
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LICENSING SUBMITTAL FOR PILGRIM NUCLEAR POWER STATION UNIT 1, RELOAD 6 l i
a 7
23A1694 l
Revision 0 Class I March 1984 GENERAL ELECTRIC BOILING WATER REACTOR SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PILGRIM NUCLEAR POWER STATION, UNIT 1 RELOAD 6 Prepared: - (d f P. A. Verbryke g Verified:
C. L. lillf/
/
Approved- ,
/F . S. Charnley, Ma uel Licensing NUCLEAR ENERGY BUSINESS OPERATIONS + GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENER AL h ELECTRIC 1/2
23A1696 Revo 0 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Boston Edison Company (BECo) for BECo's use with the U.S. Nucicar Regulatory Commission (USNRC) for amending BECO's operating license of the Pilgrim Nuclear Power Station. 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 the contract between Boston Edison Company and General Electric Company for reload fuel fabrication for the nuclear system for Pilgrim Nuclear Power Station, dated July 14, 1972, 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 au*horized; and with respect to any such unauthorized 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 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.
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23A1694 Rev. 0
- 1. PLANT UNIQUE ITEMS (1.0)*
Increased Core Flow Throughout Cycle: Appendix A Bounding Control Rod Drop Accident Analysis: Appendix B Safety / Relief Valve Low Setpoint: Appendix C
- 2. RELOAD FUEL BUNDLES (1.0, 2.7, 3.3.1 AND 4.0)
Fuel Cycle Number Designation Loaded Number Drilled Irradiated 8DB219L 4 24 24 8DB219H 4 8 8 P8DRB265L 5 120 120 P8DRB282 5 64 64 P8DRB265H 6 60 60 P8DRB282 6 112 112 New P8DRB282 7 160 160 BP8DRB282 7 32 32 Total 580 580
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at end 16474 mwd /ST of cycle:
Minimum previous cycle core average exposure at end 16074 mwd /ST of cycle from cold shutdown considerations:
Assumed reload cycle core average exposure at end 18070 mwd /ST of cycle:
Core loading pattern: Figure 1
- ( ) refers to areas of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A-6 and NEDD-24011-P-A-6-US.
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I 23A1694 Rev. 0
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTil - I NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2) 1 Minimum Shutdown Margin, BOC, k,gg f Uncontrolled 1.113 Fully Controlled 0.955 Strongest Control Rod Out 0.983 j R, Maximum Increase in Cold Core Reactivity with 0.007 Exposure into Cycle, Ak
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (ak) ppm (20*C, Xenon Free) 700 0.048
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)
(REDY Events Only)
BOC7 +
6000 mwd /ST EOC7 35.7 35.7 Void Fraction - (%)
1150 1150 Average Fuel Temperature ('F)
Void Coefficient N/A* (c/% Rg) -6.43/-8.04 -5.65/-7.07 Doppler Coefficient N/A (c/*F) -0.216/-0.205 -0.228/-0.217 Scram Worth N/A ($)** ,
L 1
- N = Nuclear Input Data, A = Used in Transient Analysis
- Generic, exposure independent values are used as given in " General Electric Application for Reactor Fuel," NEDE-24011-P-A-6-US.
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23A1696 Rev. 0
- 7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETER (S.2.2)
Fuel Peaking Factors Bundle Power Bundle Flow Initial Design (Local, Rrd'al, Axial) R-Factor (MWt) (1000 lb/hr) MCPR BOC7 + 6000 mwd /ST 4 BP8x8R/ 1.20 1.73 1.40 1.051 5.836 106.2 1.34 P8x8R ,
8x8 1.22 1.63 1.40 1.098 5.480 105.5 1.30 j EOC7
-BP8x8R/ 1.20 1.67 1.40 1.051 5.634 107.4 1.39 P8x8R 8x8 1.22 1.56 1.40 1.098 5.246 107.1 1.37
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2) 1 I
Transient Recategorization: No Recirculation Pump Trip: No I Rod Withdrawal Limiter: No Thermal Power Monitor: No Measured Scram Time: No Number of Exposure Points: 2
- 9. OPERATING FLEXIBILITY OPTIONS 4
Single Loop Operation: No Load Line Limit: No Extended Load Line Limit: Yes Increased Core Flow: Yes Flow Point Analyzed: 107.5%
Feedwater Temperature Reduction: No j 7
23A1696 Rev. 0
- 10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)
ACPR Flux Q/A BP8x8R/
Transient (% NBR) (% NBR) P3x8R 8x8 Figure Exposure: BOC7 to BOC7 +
6000 mwd /ST Load Rejection Without Bypass 527 121 0.27 0.23 2a Feedwater Controller Failure 318 121 0.22 0.20 3a l Exposure: BOC7 + 6000 mwd /ST to EOC7 Load Rejection Without Bypass 580 124 0.32 0.30 2b Feedwater Controller Failure 336 124 0.28 0.26 3b Exposure: BOC7 to EOC7 Inadvertent Startup of HPCI 121 114 0.13 0.13 4
- 11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)
SUMMARY
(S.2.2.1)
(Generic Bounding Analysis Results)
ACPR Rod Block Reading (%) (All Fuel Types) 104. 0.13 105. 0.16 106. 0.19 107. 0.22 108. 0.28 109. 0.32 110. 0.36 Setpoint Selected is: 107.
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23A1694 Rev. 0
- 12. CYCLE MCPR VALUES (S.2.2)
Nonpressurization Events:
Exposure Range: BOC7 to EOC7 BP8x8R/
P8x8R 8x8 Inadvertent Startup of HPCI 1.20* 1.20*
Fuel Loading Error 1.24 ----
Rod Withdrawal Error 1.29 1.29 Pressurization Events: l l
Option A Option B l BP8x8R/ BP8x8R/ l P8x8R 8x8 P8x8R 8x8 Exposure Range: BOC7 to BOC7 +
6000 mwd /ST Load Rejection Without Bypass 1.40 1.36 Feedwater Controller Failure 1.35 1.33 Exposure Range: BOC7 + 6000 mwd /ST to EOC7 Load Rejection Without Bypass 1.45 1.43 1.40 1.38 Feedwater Controller Failure 1.41 1.39 1.32 1.30
- 13. OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3)
P,1 P v
Transient (psig) (psig) Plant Response MSIV Closure 1315 1330 Figure 5 (Flux Scram)
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23A1694 Rev. O j l
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- 14. STABILITY ANALYSIS RESULTS (S.2.4) l l
Rod Line Analyzed: Extrapolated Rod Block l
Decay Ratio: Figure 6 Reactor Core Stability Decay Ratio, x2 /*0: 0.65 Channel Hydrodynamic Performance Decay Ratio, x2 /*0 8x8 Channel: 0.23 BP8x8R/P8x8R Channel: 0.18 ,
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- 15. LOADING ERROR RESULTS (S.2.5.4) l l
Variable Water Cap Misoriented Bundle Analysis: Yes Event In!tial MCPR Resulting MCPR Misoriented 1.22 1.07
- 16. CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)
See Appendix B.
- 17. LOSS-OF-COOLANT ACCIDENT RESULTS, NEW FUEL (S.2.5.2)
See " Loss-of-Coolant Accident Analysis Report for Pilgrim Nuclear Power Station," August 1977, NEDD-21696, as amended.
10
23A1694 Rev. 0
- ' MBEMMBsBEM
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- BEM M BsBsBEBEBiBs l BEB2BsBEBEBEBE IIIIIIIIIIIII1 1 3 5 7 9111315171921232527293133353739414345474951
. FUEL TYPE A = 8DB219L E = P8DRB265H B = 8DB219H F = P8DRB282 C = P8DRB265L G = P8DRB282 D = P8DRB282 11 = BP8DRB282 Figure 1. Reference Core Loading Pattern 11
23A1694 Rev. 0 l
1 NEUTRON FLU ( 1 VESSEL PRESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 SAFETY V AL%E FLOW i 3 CORE IM.ET TLOW 3 RELIEF VALVE FLOW l 150.0 300.0 i nvoace vartr ri m 8 '
O 100.0 I 200.0 b
8 W \
k S 0. 0
-w 100.0 a : : :-
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- 0. 0 2.0 4.0 6. 0 0.0 2. 0 4. 0 6.0 TIME (WCONOS) ilME (SECONOS) 1 LEVEL (INCH-7EF.SEP-SKRT) 1 V010 R CT ITV 2 VESSEL STcA1 FLOW 2 00PPLE E TIV!TV 200.0 ' M_ _ EE_P5%'o'
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Figure 2a. Plant Response to Load Rejectior. Without Bypass, 100% Power, 107.5% Flow at BOC + 6000 mwd /ST 12
23A1694 Rev. 0 1 hEUTR0t4 FLU ( 1 VESSEL PRESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 SAFETY V ALVE FLOW 3 CORE INLET rLOW 3 RELIEF VALVE FL0d 139.0 300.0 m evo_m.?? u m_LuE ettw k
7 5 300.0 J 2 0.0 /
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Figure 2b. Plant Response to Load Rejection Without Bypass, 100% Power 107.5% Flow at EOC 13
23A1694 Rev. 0 150.0 1 NEUTRON FLUX f I VES SEL PRESS RISECPSI) 2 AVE SURFACE HEJ [ FLU 2 SAF ETY VALVE FLOV 3 COR EINLETFLOq 3 REL lEF VALVE FLO' 150.0 ' C0= r mLET SL? 4 BYPASS VALVE FLO )
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ILEV EL(!NCH-REF.SEP.SKRT) I VOI ) REACTIVITY 2VES EL STEAMFLOW 2 00P 'LER REACTI TY 3 TUR )!NE STEAMFLOW 3, SCR 1.0 rnt LM it er REA.CTIVI ceru,!
150.0 ' rEE 3witEn rLew _
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P 50.0 W .l.0 ll l Il e0 . 2. 0
- 0. 0 20.0 40.0 0. 0 20.0 40.0 TIME (SECONOS) f!ME (SECONOS)
Figure 3a. Plant Response to Feedwater Controller Failure, 100% Power 107.5% Flow at BOC + 6000 mwd /ST 14
23A1694 Rev. 0 150.0 1 NEurRON FLUX k i VESSEL PRESS RISEt SI5\
2 AVE SURFACE HEAT FLUX 2 SAF ETY VALVE FLOW 3 REL IEF VALVE FLOW
}
3 CDRE INLET FLOW 150.0 m con
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Figure 3b. Plant Response to Feedwater Controller Failure, 100% Power, 107.5% Flow at EOC 15 i
1
\
23A1694 Rev. O I
1 NEUTRON FLUX 1 VESSEL PRESS RISE (P$1) 2 AVE SURFACE HEAT FLUX 2 REl lEF VALVE FLOW 3 COR INLET FLOW 3 BYPLSS VALVE FLOW 150.0 ' e n= r 'u rT Ei = ' "C' rLC" " Ce rws 150.0 4
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Figure 4. Plant Response to Inadvertent Start Up of IIPCI Pump, 100% Power, 107.5% Flow at EOC 16
23A1694 Rev. 0 1 NEUTRON F UX 1 VESSEL PRESS RISE (PSI) 2 AVE SURF ACE HEAT FLUX 3 SAFETY AND RELIEF VALVE FLOW
. 3 CORE INLET FLOW 4 BYPASS VALVE FLOW a
.- 100.0 :I _-
N \ ,
280.0 b
1 30.0 N , ,0. 0 0.0 . . 0.0 ,_ ,,
- 0. 0 5.0 0.0 5. 0 TIME (SECONOSI TIME (SECONDS)
ILEVEL(INC4-REF.SEP-SKRT) i v010 REAcilv1TY 2 VESSEL STEAMFLOW PLER REACT!vtTY 3 TURBINE $TEAMFLOW r 20.0
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Figure 5. Plant Response to MSIV Closure, 100% Power, 107.5% Flow 17
23A1696 Rey, o Ab ATURAL C :RCULAT10 1 B1 00 PERCENT ROD LI 1E CL LTIMATEkTABILITY LINE 1.00 (: t:
$ .75 N,
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- 0. 0 20.0 40.0 60.0 80.0 100.0 120.0 PERCENT POWER Figure 6. Reactor Core Decay Ratio 18
23A1696 Rev. O APPENDIX,A INCREASE CORE FLOW TilROUCll0UT CYCLE l
The analyses performed for Cycle 7 inciuded increased core flow through-out the cycle. There are no concerns regarding reactor internals pressure drop or flow-induced vibration as discussed in the increased core flow analysis document for the EOC-6 (NEDO-30242).
The flow-biased instrumentation for the rod block monitor should be sig-nal clipped for a setpoint of 107%, since flow rates higher than rated would otherwise result in a ACPR higher than reported for the red withdrawal error.
l 19/20
t 23A1694 Rev. O I l
APPENDIX B CONTROL ROD DROP ANALYSIS l
l The cycle-specific control rod drop accident analysis has been discon-tinued for banked position withdrawat sequence (BPWS) plants based on the fact that in all cases the peak fuel enthalpy from a control rod drop accident would be much icas than the 280 cal /gm limit. This change in procedures was reported and justified in Reference B-1. Reference B-2 indicates this change is acceptabic to the NRC.
1 REFERENCES [
f B-1. Letter, R. E. Engel (GE) to D. B. Vassallo (NRC), " Control Rod Drop i
Accident," February 24, 1982.
B-2. NRC Memo, L. S. Rubenstein to C. C. Lainas, " Changes in GE Analysis of the Control Rod Drop Accident for Plant Reloads (TACS-48058),"
l February 15, 1983.
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APPENDIX C ,
I SAFETY / RELIEF VALVE LOW !1TPOINT l
The value used in the transient analyses for the safety / relief valve l
( low netpoint in 1126 paig. This is not consistent with the value of 1106 poig l t
I reported in NED& 24011-P-A-6-US. f t
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