ML20034G517
| ML20034G517 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 02/28/1993 |
| From: | Klapproth J, Noble L GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20034G512 | List: |
| References | |
| 23A7147, 23A7147-R02, 23A7147-R2, NUDOCS 9303100012 | |
| Download: ML20034G517 (25) | |
Text
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I GENuclearEnergy 175Curmet A.enue San Jcse. CA 951;S 23A7147 Revision 2 Class I February 1993 23A7147, Rev. 2 Supplemental Reload Licensing Report for Perry Nuclear Power Plant Unit 1 Reload 3 Cycle 4 Approved Y
h2M Approved Av J. F. Klapproth, *s'ilanger L D. Noble, Manager Fuel Licensing Reload Nuclear Engineering
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Perry I nem Reload 3 an. 2
)
Important Notice Regarding Contents of This Report j
Please Read Carefully l
l The only undertakings of the General Electric Company (GE) respecting information in this l
document are contained in the contract between Cleveland Electric Illuminating Company (CEI) and GE, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than CEI for any purpose other than that for which it is l
intended, is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, and assumes no liability as to the completeness, accuracy or usefulness of the information contained in this document.
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Perry I num Reload 3 nn 2 Acknowledgment j
The engineering and reload licensing analyses which form the technical basis of this Supple-mental Reload Licensing Report, were performed by P. A. Hahn and J. L Casillas of the Fuel En-gineering Section. The Supplemental Reload Licensing Report was prepared by P. A. Lambert of Fuel Licensing. This report has been verified by Fuel Engineering and J. L. Embley of Fuel l
Licensing.
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j Page 3
Perry 1 2347247 Reload 3 Rev 2 The basis for this report is General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-10, February 1991; and the U. S. Supplement. NEDE-24011-P-A-10-US, March 1991.
1.
Plant-unique items Appendix A:
Analysis Conditions Appendix B:
Basis For Analysis of Loss-of-feedwater Heating Event Appendix C:
Analyzed Operating Domain Appendix D:
Transient Analyses Appendix E:
Rotated Bundle Analysis 2.
Reload Fuel Bundles Fuel Tree Cycle leaded Number Irradiated BPSSRB176 (BP8x8R) 1 10 G E8B-P8SQB301-7GZ-120M-150-T ( BS301 E) (G E8x8EB) 2 128 GESB-P8 SOB 301-5GZ-120M-150-T (BS301F) (G E8x8EB) 2 134 G E8 B-P8SQ B320-9G Z-120M-150-T (G E8x8E B) 3 104 G E8 B-P8SO B322-7G Z-120 M-150-T (G E8x8 EB )
3 168 New G E 10-P8S X B306-1 1 G Z3-120M - 150-T (G E8x8N B-3 )
4 68 G E 10-P8SX B306-10GZ2-120M-150-T (G E8x8 NB-3 )
4 JM l
Total 748 3.
Refer ecce Core leading Pattern
- mwd /ST mwd /MT Nc minal previous cycle core average exposure at end of cycle:
16,110 17.758 Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations:
16,110 17,758 Assumed reload cycle core average exposure at beginning of cycle:
15,946 17,577 Assumed reload cycle core average exposure at end of cycle:
21,881 24,120 Core loading pattern:
Figure 1
- The information in Sections 3,4 and 5 considers only the remainder of Cycle 4; all other information applies to the entire Cycle 4.
Page 4
= --
i Perry 1 2 min Reksad 3 Rev 2 4.
Cgculated Core EITective Multiplication and Control System Worth - No Voids.
20 C*
Beginning of Cycle, K nen>w e
Uncontrolled 1.121
{
Fully conti ded 0.956 l
l Strongest control rod out 0.984 R, Maximum increase in cold core reactivity with exposure l
into cycle. AK 0.005 l
l S.
Standby Liquid Control System Shutdown Capability
- l t
i Boron Shutdown Margin (AK) l (CDm)
(20*C. Xenon Free) i l
660 0.029 6.
Reload Unique GETAB AOO Analysis Initial Condition Parameters j
)
Fuel Peakine Factors Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (MWt)
(1.000 lb/hr)
MCPR l
Exposure: BOC4 to EOC4 Increased core flow /Feedwater temperature 420*F GE8x8NB-3 1.20 1.58 1.40 1.000 7.369 116.5 1.21
)
GE8x8EB/
1.20 1.48 1.40 1.051 6.925 120.3 1.17 BP8x8R Exposure: BOC4 to EOC4 Increased core flow /Feedwater temperature reduction to 250*F GE8x8NB-3 1.20 1.64 1.40 1.000 7.620 114.6 1.20 GE8x8EB/
1.20 1.53 1.40 1.051 7.14 0 118.6 1.18 l
B.'3x8R j
Exposure: BOC4 to EOC4 Increased core flow /Feedwater temperature reduction to 320*F GE8x8NB-3 1.20 1.61 1.40 1.000 7.523 115.3 1.20 GE8x8EB/
1.20 1.51 1.40 1.051 7.048 119.3 1.18 BP8x8R Exposure: BOC4 to EOC4 Increased core flow /Feedwater temperature reduction to 370*F
)
GE8x8NB-3 1.20 1.61 1.40 1.000 7.517 115.5 1.19 GE8x8EB/
1.20 1.50 1.40 1.051 7.024 119.5 1.17 BP8x8R
)
- The information in Sections 3. 4 and 5 considers only the remainder of Cycle 4; all other information applies to the entire Cycle 4.
)
Page5
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Perry 1 2mm Reload 3 nev :
7.
Selected Margin Improvement Options Recirculation pump trip:
Yes Rod withdrawal limiter:
Yes Thermal power monitor:
Yes Measured scram time:
No Exposure dependent limits:
No Exposure points analyzed:
1 (EOC) 8.
Operating Flexibility Options (S.S.2)
Single-loop operation:
Yes Load line limit:
No Extended load line limit:
No Maximum extended load line limit:
No Increased core flow at end af cycle:
Yes Increased core flow throughout the cycle:
Yes Flow point analyzed:
105Fe Feedwater temperature reduction throughout the cycle:
Yes Final feedwater temperature reduction:
Yes Temperature reduction:
50
- F,100
- F,170
- F ARTS Program:
No Maximum extended operating domain:
Yes Main steam isolation valve out of service:
No Recirculation pump trip out of service:
No Turbine bypass out of service:
No
)
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y Page 6
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Perry I neic Reload 3 Rn :
9.
Core-wide AOO Analysis Results Methods used: GEMINI and GEXL-PLUS Uncorrected ACPR Flux Q/A GE8x8 Ell /
Event
('"c NilR)
('~r N H R ) GE8s8 Nil-3 IIP 8x8R Ficure Exposure range: HOC 4 to EOC4 increased core flow /Feedwater temperature 420* F Load rejection 401 113 0.14 0.10 2
without bypass Feedwater controller 289 112 0.11 0.09 3
failure (143F )
c Pressure regulator 146 105 0.07 0.05 4
failure downscale Loss of 100* F 0.12 0.12 feedwater heating Exposure: BOC4 to EOC4 Increased core flow /Feedwater temperature reduction to 250*F Feedwater controller 280 117 0.13 0.12 5
failure (143r )
c
)
Pressure regulator 148 106 0.07 0 06 6
failure downscale Exposure: BOC4 to EOC4 increased core flow /Feedwater temperature reduction to 320*F Feedwater controller 292 116 0.12 0.11 7
-)
failure (143rc)
Exposure: HOC 4 to EOC4 Increased core flow /Feedwater temperature reduction to 370*F Feedwater controller 293 114 0.12 0.10 8
failure (143rc)
)
- 10. IAcal Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary The generic bounding BWR/6 rod withdrawal error (RWE) is analyzed in NEDE-24011-P-A-9-US and GESSAR-Il Appendix 15B is applied; the resulting ACPR is 0.11. The generic RWE ACPR was verified to be applicable to the new fuel design. The original generic analysis in GESSAR-II was not applicable for control cell core operation; however, it was subsequently shown to be applicable for control cell core operation and
)
GESSAR-il was revised to reflect this application in Revision 21.
- See Appendix B.
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Perry 1
- mc Reload 3 Rev 2
- 11. Cycle MCPR Values *
)
Safety limit: 1.07 l
Single loop operation safety limit: 1.08 Exposure range: BOC4 to EOC4
)
Non-pressurizadon events GE8x8EB/
GE8x8NH-3 HP8x8R
)
Rod withdrawal error 1.18 1.18
)
Fuel loading error *
- 1.23 1.21 Loss of 100* F feedwater heating 1.19 1.19 (Bounding from 420*F feedwater temperature condition)
Pressurization events
)
]
Option A GE8x8EB/
i GE8x8NH-3 BP8x8R j
)
Exposure range: BOC4 to EOC4 Increased core flow /Feedwater temperature 420*F Load rejection without bypass 1.21 1.18 Feedwater controller failure 1.19 1.16 Pressure regulator failure downscale 1.15 1.13
)
Exposure: BOC4 to EOC4 Increased core flow /Feedwater temperature reduction to 250* F Feedwater controller failure 1.21 1.20 Pressure regulator failure downscale 1.15 1.14 J'
Exposure: BOC4 to EOC4 Increased core flow /Feedwater temperature reduction to 320*F Feedwater controller failure 1.21 1.19 Exposure: BOC4 to EOC4 increased core flow /Feedwater temperature reduction to 370*F
)
j Feedwater controller failure 1.20 1.18 I
- GEMINI ODYN adjustment factors are provided in the letter from J. S. Charnley (GE) to M. W.
Hodges (NRC), GEMINI ODYNAdjustment Factorsfor BlVR/6, dated July 6,1987. The MCPR limit does not change because of channel bow. Channel bow is reflected in the monitoring of the
)
core.
- See Appendix E.
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1 Page 8
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Perry I num l
Reload 3 Rev :
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- 12. Overpressurization Analysis Summary 1
)
Pu Pv Event (psic)
(psic)
Plant Resnonse MSIV closure (flux scram)*
1241 1272 Figure 9
)
- 13. IAading Error Results Variable water gap misoriented bundle analysis: Yes l
l ACPR i
)
Event GE8x8NB-3 GE8xREB/BP8x8R t
Miscriented fuel bundle 0.16 "
0.14 "
)
- 14. Control Rod Drop Analysis Results l
Banked Position Withdrawal Sequence is utilized at the Perry Nuclear Power Plant Unit 1:
I therefore, the bounding control rod drop analysis (CRDA) described in
)
NEDE-240ll-P-A-10-US is applied. NRC approval of the bounding analysis is given in the l
letter to 3. S. Charnley (GE), Acceptance for Referencing of Licensing Topical Report l
l NEDE-24011, Revision 6, Amendment 9 "GESTAR-Il General Electric Standard Application for Reactor Fuel," January 25,1985.
- 15. Stability Analysis Results GE SIL-380 recommendations have been included in the Perry Nuclear Power Plant Unit 1
)
l operating precedures and/or Technical Specifications and, therefore, the stability analysis is i
not required. NRC approval for deletion of a cycle-specific stability analysis is documented l
in Amendment 8 to NEDE-24011-P-A-US. In addition, the Perry Nuclear Power Plant Unit I recognizes the issuance of NRC Bulletin No. 88-07, Supplement 1, Power Oscillations in
)
Boiling Water Reactors (BWRs), and will continue to comply with the recommendations contained herein.
)
- The MSIV closure (flux scram) analysis is perfo. med using GEMINI methods at the 1027c power level to account for the power level uncertainties specified in Regulatory Guide 1.49. The analysis was performed with 13 highest setpoint s afety valves operational.
l "ACPR penalty of 0.02 for the tilted misoriented bundle has been applied. See Appendix E.
)
Page 9
I 23A7147 u*
'O Rev 2
.cf
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ee the Perry Nuclear Power Plant Unit I
.alysis Report, as amended) u o..
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. 0.M-150-T (GE8s8NB-3)
M APLHG R (kw/ft) l
~
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,,G J.
_,/ N n Most Limiting Least Limiti g 0.0 0.0 11.55 12.43 0.2 0.2 11.61 12.47 1.0 1.1 11.71 12.58 2.0 2.2 11.92 12.72 3.0 3.3 12.17 12.88 4.0 4.4 12.41 13.04 5.0 5.5 12.61 13.20 6.0 6.6 12.81 13.33 7.0 7.7 12.99 13.41
)
8.0 8.8 13.16 13.50 9.0 9.9 13.31 13.56 10.0 11.0 13.34 13.43 12.5 13.8 13.23 13.40 f
15.0 16.5 12.92 13.07
)
i 20.0 22.0 12.16 12.40 25.0 27.6 11.44 11.76 35.0 38.6 10.14 10.40 1
45.0 49.6 8.90 9.15 51.7 57.0 5.87 6.03 51.9 57.2 5.95
)
The peak clad temperature (PCT) is $2149'F at all exposures; the local oxidation (fraction) is s0.061 at all exposures. The MAPLHGR multiplier for single-loop operation (SLO) is 0.80.
)
)
Page 10
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Perry I n 47:47 Egload 3 ne 2
- 16. Less-of-coolant Accident Results (continued)
Hundle Type: GE10-P8SXB306-10GZ2-120M-150-T (GE8x8NB.3)
Averace Planar Exoosure M APLHGR (kw/ft)
)
(GWd/ST)
(GWd/MT)
Most Limiting Least Limiting 0.0 0.0 11.21 12.35 0.2 0.2 11.26 12.45
)
1.0 1.1 1136 12.62 2.0 2.2 11.56 12 75 3.0 33 11.81 i.'.i5 4.0 4.4 12.08 12.95
)
5.0 5.5 12.35 13.06 6.0 6.6 12.57 13.17 7.0 7.7 12.80 13.28 8.0 8.8 13.00 13.38
)
9.0 9.9 13.20 13.4 6 10.0 11.0 13.37 13.52 12.5 13.8 13.45 13.52 15.0 16.5 13.14 13.20
)
20.0 22.0 12.40 12.57 25.0 27.6 11.61 11.94 35.0 38.6 10.12 10.57 45.0 49.6 8.83 9.29
)
52.1 57.4 5.87 5.96 52.3 57.6 5.89 The Peak Clad Temperature (PCT) is $2129'F at all exposures; the Local Oxidation (Fraction) is
)
50.058 at all exposures. The MAPLHGR multiplier for single-loop operation (SLO) is 0.80.
1 l
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1 Page 11 l
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.BEBEME8MMME+888s sMMBEMMMMMMMM.
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- -sEMMHEEBEBEEEMMBiHBEBEEe
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i:M M M M M M EGBE M M M EE N M M i HBEMMBEMBEMMMMBEMBEM 21:EEMBEEEBEMMBEBEBEBEBEMBEEE
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- "MMMEEMBEMMMEEMME" "BEEEEE E M M M M M BEM" "MEMMMMMMM"
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IIIIIIIIIIIi 1 3 5 7 911131517192123252729313335373941434547495153555759
)
FUEL TYPE A = BP85RB176 E = GE88-P850B320-9GZ-120M-150-T B = GE8B-P8SQB301-5GZ-120M-150-T F = GE10-P8SXB306-10GZ2-120M-150-T
)
C = GE88-P85Q8301-7GZ-120M-150-T G = GE10-P8SXB306-llGZ3-120M-150-T D = GE8B-P850B322-7GZ-120M-150-T Figure 1 Reference Core Imading Pattern
)
{
l Perry I m,c Reload 3 Rn 2 i turaw rtex i vtsstt Pa ss sist(as a l
2 ave SJWAct {At FLUX 2 SATETY v 'vE rt; 3 CORE 14.E7 rLCW 3, K d' { g } pid m.
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Tra (somes) ist (Ecoes) l Figure 2.
Plant Response to Load Rejection without Bypass (ICF/FWT 420'F)
Page 13
Perry 1
,],
Reload 3
$ [ N 2 (ATrwx jJ4]ltGlt8:dyb 25 $$N Mi5'E W
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Figure 3.
Plant Response to Feedwater Controller Failure (ICF/FWT 420*F)
Obs Be
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{AT RUX h
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Figure 4.
Plant Response to Pressure Regulator Failure Downscale (ICF/FMT 420'F)
Page 15
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Figure 5.
Plant Response to Feedwater Controller Failure (ICF/F%TR to 250'F) i Page 16
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i Figure 6.
Plant Response to Pressure Regulator Failure Downscale
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l (ICF/FWTR to 250'F) l t
h Page 17
Perry I m ic Reload 3 Rn 2 s
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Figure 7.
Plant Response to Feedwater Controller Failure (ICF/FWTR to 320'F) l
)
l Page 18
Perry I nar Reload 3 nes :
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Plant Response to Feedwater Controller Failure (ICF/FWTR to 370*F)
Page 19
Perg i ucic Reload 3 nn 2
,,mo.
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Plant Response to MSIV Closure, Flux Scram l
Page 20
1 Perry I nunc Reload 3 nev 2 Appendix A i
Analysis Conditions i
To reflect actual plant parameters accurately, the values shown in Table A-1 were used this cycle to reflect the bounding conditions.
Table A-1 i
Analvsis Value (FW Temo.)
Parameter 420*F 250*F 320*F 370*F Thermal power, MWt 3579 3579 3579 3579 Dome pressure, psig 1028 1008 1015 1019 Steam flow, Mlb/hr 15.70 12.58 13.58 14.42 Turbine pressure, psig 976 974 975 975 Core flow, Mlb/hr 109.2 109.2 109.2 109.2 Reactor pressure, psia 1056 1056 1056 1056 Inlet enthalpy, Bru/lb 528.8 512.4 518.2 523.1
{
Non-fuel power fraction 0.038 0.038 0.038 0.038 i
No. of dual mode Safety / Relief Valves 17*
17*
17*
17*
Relief mode lowest setpoint, psig 1143*
1143*
1143*
1143' l
Safety mode lowest setpoint, psig 1177 1177 1177 1177 i
l 1
(
l i
t f
'There are a total of 19 valves; the 2 lowest setpoint safety / relief valves are assumed to be i
out-of-service in the transient analyses.
Pagt 23 I
\\
Perry 1 2 min Reload 3 nev 2 Appendix B Basis for Analysis of Loss-of-feedwater Heating Event The loss-of-feedwater heating event was analyzed using the BWR Simulator Code (Reference B-1). The use of this code is permitted in GESTAR II (Reference B-2). The transient plots, neutron flux and heat flux values normally reported in Section 9 are not an output of the BWR Simulator code; therefore, these item; are not included in this document.
The transient analysis inputs normally reported in Section 6 of the licensing submittal are internally calculated in the BWR Simulator Code and in ODYN.
References B-1 Steadt-State Nuclear Methods, NEDE-30130-P-A and NEDO-30130-A, April 1985.
B-2 General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-9, September 1988.
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Page 22 l-
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t
)
Perry 1 2mm l
Reload 3 Rev 2
)
Appendix C Analyzed Operating Domain l
l
)
The core-wide abnormal operational occurrence (AOO) analysis results reported in Section 9 are the most limiting values over the entire allowable operating range. This range covers the i
following operating options:
)
1.
Standard 1007c power / flow map; 2.
End-of-cycle power coastdown;
)
l 3.
MEOD with 100?c power flow range from 75Fc to 105cc of rated; and l
l l
4.
Partial feedwater heati.ig to 320*F during the cycle with final feedwater temperature reduction to 250* F after AllRods Out at end of cycle.
l Limiting events and conditi. s analyzed are based on Reference C-1 and the USAR analytical results. The Reload 3/ Cycle 4 analyses were performed assuming all four turbine control valves in a full arc mode of operation. This is conservative for partial are configuration.
The single-loop operation (SLO) analysis was reverified for the standard power / flow map with normal feedwater temperature.
References C-1 General Electric Standard Applicationfor Reactor Fuel, NEDE-24011-P-A-10-US, April 1991.
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Page 23
Perry 1 23A7147 Reload 3 Rev 2 Appendix D
)
Transient Analyses
)
The turbine trip without bypass (TTNBP) analysis AOO is a pressure increase event normally checked on a cycle-by-cycle basis to determine if this AOO could potentially establish the cycle MCPR operating limit.
)
The Perry turbine control valves will be operated in the full are mode throughout Cycle 4.
The load rejection without bypass (LRNBP) is always more limiting in this mode of operation; therefore, the TTNBP will not be limiting for Cycle 4 and was not analyzed.
)
The load rejection without bypass (LRNBP) AOO was run for the standard case only since it has been shown to be more limiting than the feedwater temperature reduction cases in previous reload analyses.
I
)
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)
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rw u
Perry I mm7 Reload 3 Rev 2 Appendix E Rotated Bundle Analysis The results for each fuel type are listed in Table E-1.
i i
Table E-1 ACPR BP8 SRB 176 (BP8x8R)
N/A G E8 B-P8SO B301-7G Z-120M-150-T ( BS301 E) (G E8x8 E B )
0.13 G E8B-P8SQ B301 5GZ-120 M-150-T (BS301 F) (G E8x8EB) 0.12 G E8 B-P8SQ B320-9G Z-120 M - 150-T (G E8x8 E B) 0.14 G E8 B-PSS OB322-7G Z-120M-150-T (G E8x8 E B) 0.14 G E 10- P8S XB306-l ! G Z3-120M-150-T (G E8x8N B-3 )
0.16 GE10-P8SXB306-10GZ2-120M-150-T(GE8x8NB 3) 0.09 l
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