ML20247N772
| ML20247N772 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/30/1998 |
| From: | Hetzel W, Watford G GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20013J582 | List: |
| References | |
| J1103244SRLR, J1103244SRLR-R, J1103244SRLR-R00, NUDOCS 9805270222 | |
| Download: ML20247N772 (28) | |
Text
[
l ENCLOSURE 2 -
- BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO.1 DOCKET NO. 50-325 LICENSE NO. DPR-71 TRANSMITTAL OF CORE OPERATING LIMITS REPORT,'.
l SUPPLEMENTAL RELOAD LICENSING REPORT, AND LOSS-OF-COOLANT-ACCIDENT ANALYSIS REPORT
" SUPPLEMENTAL RELOAD LICENSING REPORT FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 1 RELOAD 11 CYCLE 12,"
J1103244SRLR, REVISION 0, CLASS 1, APRIL 1998 EEM E5
- P-r*DR i
[.
O f
l GENuclearEnergy J1103244SRLR Revision 0 l'
ClassI April 1998 l
Supplemental Reload Licensing Report for BRUNSWICK STEAM ELECTRIC PLANT UNIT 1 Reload 11 Cycle 12 f
&)
GE Nuclear Energy J1103244SRLR Revision 0 ClassI April 1998 J1103244SRLR, Rev. 0
[
Supplemental Reload Licensing Report for Brunswick Steam Electric Plant Unit 1 Reload 11 Cycle 12 Approved Approved [k,
. A. A9dford, Manager W. H. Hetzel l
Nuclear Fuel Engineering Fuel Project Manager
[.
l l
. BRUNSWICK 1 J1103244SRLR Reload 11 Rev. O Important Notice Regarding I
Contents of This Report Please Read Carefully This report was prepared by General Electric Company (GE) solely for Carolina Power and Light Company (CP&L) for CP&L's use in defining operating limits for the i
Brunswick Steam Electric Plant Unit 1. The information contained in this report is believed by GE to be an accurate and true representation of the facts known or obtained or provided to GE at the time this report was prepared.
The only undertakings of GE respecting information in this document are contained in the contract between CP&L and GE for nuclear fuel and related services for the nuclear system for Brunswick Steam Electric Plant Unit 1 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 unauthorized use, neither GE nor any of the contributors to this document makes any representation or warranty (expressed 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 I
such information.
1 4
i Page 2
I l
I BRUNSWICK 1 J1103244SRLR Reload 11 Rev. O l
Acknowledgement The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload l
Licensing Report, were performed by S. B. Shelton, H.M. Schrum, and P. Wei. The Supplemental Reload Licensing Report was prepared by S. B. Shelton. This document has been verified by J. Su.
l i
Page 3
.f BRUNSWICK 1L J1103244SRLR Reload 11 Rev. O The basis for this report is General Electric Standard Applicationfor Reactor Fuel, NEDE-2401l-P-A-
- 13. August 1996; and the U.S. Supplement, NEDE-24011-P-A-13-US, August 1996.
1.
' Plant-unique Items
. Appendix A: Analysis Conditions Appendix B: Main Steamline Isolation Valve Out of Service (MSIVOOS)
Appendix C: Decrease in Core Coolant Temperature Events Appendix D: Feedwater Temperature Reduction (FWTR)
Appendix E: Maximum Extended Operating Domain (MEOD)
Appendix F: Turbine Bypass Out of Service (TBPOOS)
L p
?.
Reload Fuel Bundles l
Cycle Fuel Type Loaded Number Irradiated:
GE l o-P811XB322-1 I GZ-70M-150-T (G E8 x 8NB-3) 9 8
GElo-P81IXB346-12GZ-100M-150 T (GE8x8NB-3) 10 156
- GE13-P9DTB380-10G5.0A 100T-146-T (gel 3) -
11 132 GE13 P9DTB380-IlG5.0A-100T-146-T(GE13) 11 68 NCE G E 13 -P9 DTB403-7G 6.0/7G 5.0. ! 00T-146-T (G E 13) 12 160' G E 13 -P9 DTB403-5G 6.0/7G 5.0- 100T-146-T (G E l 3) 12 36 Total 560 I
l 3.
Reference Core Loading Pattern j
Nominal previous cycle core average exposure at end of cycle:
27730 mwd /MT
( 25156 mwd /ST)
- 1 Minimum previous cycle core average exposure at end of cycle 27450 mwd /MT' from cold shutdown considerations:
-( 24902 mwd /ST)
Assumed reload cycle core average exposure at beginning of 14253 mwd /MT cycle:
( 12930 mwd /ST) 1 Assumed reload cycle core average exposure at end of cycle:
- 29865 mwd /MT
( 27093 mwd /ST)
Reference core loading pattern:
Figure I l
' ' This value corresponds to 492.2 EFPD -
Page 4
__--_________L
e 4
. BRUNSWICK l!
J1103244SRLR Reload 11 Rev. 0
/4.
Calculated Core Effective Multiplication and Control System Worth - No Voids,20 C -
Beginning of Cycle, k genjve e
Uncontrolled 1.112 Fully controlled
' O.958 Strongest control rod out 0.986 R, Maximum increase in cold core reactivity with exposure into cycle, Ak 0.000 5.
Standby Liquid Control System Shutdown Capability Boron (ppm)
Shutdown Margin (Ak)
(at 20 C)
- o 660-0.032 6.
Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parameters Exposure: BOCl2 to EOCl2-2205 mwd /MT (2000 mwd /ST) with ICF Peaking Factors L
Fuel Bundle Bundle Initial Design Local Radial Asial R-Factor Power Flow MCPR
( M W t)
(1000 lb/hr) gel 3 1.45 1.45 1.40 1.020 6.466 109.5 1.35 l
l Exposure: EOCl2-2205 mwd /MT (2000 mwd /ST) to EOCl2 with ICF Peaking Factors Fuel Bundle Bundle Initial
- Design -
Local Radial Axial R Factor Power Flow MCPR.
(MWt)
(1000 lb/hr)
GE13 1.45 1.40 1.47.
1.020 6.217 112.7 1.36 i
l Page 5
r; BRUNSWICK 1.
Jl103244SRLR Reload 11 Rev. 0 '
Exposure: BOCl2 to EOCl2 with TBPOOS -
I Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor
' Power Flow MCPR -
( M W t)
(1000 lb/hr)
I gel 3 1.45 1.48 1.19 1.020 6.569 108.1 1.39 Exposure: BOC12 to EOCl2 with TBPOOS and FWTR l-Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor-Power Flow MCPR (MWt)
(1000 lb/hr)
GEi3 1.45 1.51 1.24 1.020 6.676 106.9 1.39 7.
Selected Margin Improvement Options Recirculation pump trip:
No Rod withdrawal limiter:
No Thennal power monitor:
Yes Improved scram time:
Yes (ODYN Option B)
,n Measured scram time:
No Exposure dependent limits:
Yes Exposure points analyzed:
2 (EOCl2-2205 mwd /MT and EOCl2)
Page 6
- [
BRUNSWICK 1 J1103244SRLR Reload 11 Rev.- 0 l
8.
Operating Flexibility Options l
Single-loop operation:
Yes Load line limit:
Yes Extended load line limit:
Yes Maximum extended load line limit:
Yes Increased core flow throughout cycle:
Yes Flow point analyzed:
104 3 %
Increased core flow at EOC:
Yes Feedwater temperature reduction throughout cycle:
Yes Temperature reduction:
110 3F Final feedwater temperature reduction:
Yes ARTS Program:
Yes Maximum extended operating domain:
Yes Moisture separator reheater OOS:
No Turbine bypass system OOS:
Yes Safety / relief valves 00S:
Yes (credit taken for 9 of 11 valves, however, ATWS evaluations require 10 in-service for power uprate)
ADS OOS:
Yes (2 valves OOS)
No Main steam isolation valves 005:
Yes Page 7
BRUNSWICK 1 J1103244SRLR Refoed 11 Rev 0 9.
Core-wide AOO Analysis Results Methods used: GEMINI; GEXL-PLUS Exposure range: BOC12 to EOCl2-2205 mwd /MT (2000 mwd /ST) with ICF Uncorrected ACPR Event Flux Q/A GE13 Fig.
(%NBR)
(%NBR)
Load Reject w/o Bypass 502 121 0.26 2
Exposure range: EOCl2-2205 mwd /MT (2000 mwd /ST) to EOCl2 with ICF Uncorrected ACPR Event Flux Q/A GE13 Fig.
(%NilR)
(*/.NIIR)
Tuthine Trip w/o Bypass 569 125 0.27 3
Exposure range: 110C12 to EOC12 with TBPOOS Uncorrected ACPR Event Flux Q/A
- gel 3 Fig.
(%NilR)
(%NilR)
FW Controller Failure 369 122 0.30 l4 s
Exposure range: IlOCl2 to EOCl2 with TBPOOS and FWTR Uncorrected ACPR Event Flux QIA GE13 Fig.
(%NBR)
(*/.NBR)
FW Controller Failure 403 125 0.30 5
- 10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary The rod withdrawal error event in the maximum extended operating domain was originally analyzed in the GE BWR Licensing Report, Afaximum Erlended Operatmg Domain Analysts for Brunswick Steam E/cctric Plant, NEDC-31654P, February 1989. The MCPR limit for rod withdrawal error is bounded by the operating limit MCPRs presental in section 11 of this report for RBM setpoints shown in Tables 10-5(a) or 10 5(b) of NEDC-31654P, Additionally, the RBM operability requirements specified in Section L
10.5 of NEDC-31654P have been evaluated and shown to be sufHcient to ensure that the Safety Limit L
MCPR and cladding 1*/o plastic strain criteria will not be exceeded in the event of an un-blocked RWE event.
Page 8
[:..
' BRUNSWICK 1
= 31103244SRLR Reload 11 Rev. O II. Cycle MCPR Values
In agreement with commitments to the NRC (letter from M. A. Smith to the Document Control Desk, 10CFR Part 21, Reportable Condition, Safety Limit MCPR Evaluation, May 24,1996) a cycle-specific l
. Safety Limit MCPR calculation was performed, and has been reported in both the Safety Limit MCPR and the Operating Limit MCPR shown below. This cycle specific SLMCPR was determined using the analysis basis documented in GESTAR with the following exceptions:
1; The reference cere loading was analyzed.
- 2. The actual bundle parameters (e.g., local peaking) were used.
- 3. The full cycle exposure range was analyzed.
Safety limit:
1.09 l
Single loop operation safety limit: 1.10 l
Non-pressurization events:
Exposure range: BOCl2 to EOCl2 GE13 Fuel Loading Error (disoriented) 1.26 I
Pressurization events:
d Exposure range: BOCl2 to EOCl2-2205 mwd /MT (2000 mwd /ST) with ICF Exposure point: EOC12-2205 mwd /MT (2000 mwd /ST) l Option A Option B GE13 GE13 lead Reject w/o Bypass 1.40 1.35 l
l l
2 The Operating Limit MCPR for two loop operation (TLO) bounds the Operating Limit MCPR for single loop operation (SLO); therefore, the Operating Limit MCPR need not be changed for SLO.
i I
The ICF Operating Limits for the exposure range of BOCl2 to EOCl2-2205 mwd /MT (2000 mwd /ST) bound the Operating Limits for the following domains: MELLL, ICF and FWTR, MSIVOOS and ICF, I
l Page 9 -
BRUNSWICK 1 Jl103244SRLR Reload 11 -
Rev. 0 8
Exposure range: EOCl2-2205 mwd /MT (2000 mwd /ST) to EOC12 with ICF Exposure point: EOC12 l
Option A Option B GE13 GE13 Turbine Trip w/o Bypass.
1.46 1.38 l
l Exposure range: DOCl2 to EOCl2 with TBPOOS' L
Exposure point: EOC12 Option A Option B GE13 GE13 FW Controller Failure 1.49 1.41 Exposure range: HOCl2 to EOCl2 with TBPOOS and FWTR Exposure point: EOCl2 Option A Option B GE13 GE13 FW Controller Failure 1.50 1,42
- 12. Overpressurization Analysis Summary Psi Pv Plant Event (psig)
(psig)
Response
MSIV Closure (Flux Scram) 1285 13:6 Figure 6
- 13. Loading Error Results Variable water gap disoriented bundle analysis: Yes '
Disoriented Fuel Bundle ACPR G E I 3-P9DTB403-5G6.0/7GS.0-100T-146-T (GE 13) 0.06 GE13-P9DTB403-7G6.0/7G5.0-100T-146-T (GE13) 0.10 GE13 P9DTB380-10G5.0A-100T-146-T(GE13) 0.16
)
. GE13-P9DTB380-I lG5.0A-100T-146-T (GE13) 0.17 5 The ICF Operating Limits for the exposurt range of EOC12-2205 mwd /MT (2000 mwd /ST) to EOCl2 bound the Operating Limits for the following domains: ME11L, ICF and FWTR, MSIVOOS and ICF.
' ' includes a 0.02 penalty due to variable water gap R-factor uncertainty.
Page 10 c__=_u________
L.
BRUNSWICK 1 Jl103244SRLR Rev. O Beload 11
- 14. Control Rod Drop Analysis Results This is a banked position withdrawal sequence plant, therefore, the control rod drop accident analysis is not required. NRC approval is documented in NEDE-24011-P-A-US.
l l
- 15. Stability Analysis Results GE SIL-380 reconunendations and GE interim corrective actions have been included in the Brunswick Steam Electric Plant Unit 1 operating procedures. Regions of restricted operation defined in Attachment I to NRC Bulletin No. 88-07 Supplement 1, Power Oscillations in Bolling Water Reactors Reactors (BWRs), are applicable to Brunswick 1.
Brunswick Unit I Cycle 11. Reactor Stability Long-Term Solution Enhanced Option I-A, Stability Region Boundary Generation and Validation, GENE-A13-00367-47 documents the Enhanced Option I-A (EI A) stability region boundaries for Brunswick Unit I Cycle i1 and the analysis associated with their generation and validation. Reload validation of these stability region boundaries has been performed in accordance with NEDO-32339 A, Revision I, Reactor Stability, Long Term Solution: Enhanced Option I-A. The Reload Validation Matrix confirms best-estimate code boundary validation stability ct' ria. The results are shown in Figure 7.
'Ihe GE8x8EB LOCA analysis results presented in Sections 5 and 6 of Brunswick Steam Electric Plant Units 1 and 2 SAFElVGE57'R-LOCA Loss-of-Coolant Accident Analysis, NEDC-31624P, Revision 2, July 1990, conservatively bound the LOCA analysis of the GE8x8NB-3 fuel types. This analysis yielded a licensing basis peak clad temperature of 1533 'F, a peak local oxidation fraction of <0.30%, and a core-wide metal-water reaction of 0.046%.
An additional LOCA analysis was performed for the GE13 fuel type. The results, presented in Brunswick Steam Electric Plant Units 1 and 2 SAFEIUGESTR-LOCA Loss-of-Coolant Accident Analysis:
Application to GE13 Fuel, NEDC-31624P, Supplement 3, Revision 0, January 1996, show a licensing basis peak clad temperature of 1535 F. The peak local oxidation fraction and core-wide metal-water reaction were shown to be bounded by the results from GE8x8EB LOCA analysis.
A single loop operation MAPLHGR multiplier of 0.80 is applicable to both GE8x8NB-3 and gel 3 fuel types. Therefore, the power-and flow-dependent MAPLHGR adjustment factors identified in Figures 4-2 and 4-4 of Maximum Extended Operating Domain Analysis for the Brunswick Steam Electric Plant, NEDC-31654P, Class III (GE Nuc! car Energy Proprietary), February 1989, should be used with the limitation that no multiplier greater than 0.80 is used during SLO.
The most and least limiting MAPLHGRs for the new gel 3 fuel designs are as follows:
Page11
l:
I BRUNSWICK 1 J1103244SRLR Jteload 11 Rev.- 0
.16.* Loss-of-Coolant Accident Results (cont.)
Bundle Type: GE13-P9DTB403-5G6.0/7GS.0-100T-146-T Average Planar Exposure -
MAPLHGR (kw/ft)
- (GWd/ST)
(GWd/MT).
Most Limiting Least Limiting 0.00 0.00 10.65 10.73 0.20 0.22 10.72 10.79 l.00' l.10 10.85 10.88 2.00 2.20.
I1.00 11.03 3.00 3.31 11.12 11.21
- 4.00 -
4.41 11.25 11.35 5.00 5.51 11.38 11.50 6.00-6.61 11.52 11.66 7.00 7.72 11.66 11.82 8.00 8.82 11.81 11.99 9.00 9.92 11.95 l' ?2 10.00 11.02 12.05-12.26 12.50 13.78 12.04 12.36 15.00 16.53 11.97 12.29 17.50 19.29 11.79 12.07 20.00 22.05 11.54 11.79 25.00 27.56 11.02 11.23
_)
30.00 33.07 10.44 10.49 35.00 38.58 9.69 9.85 j
40.00 44.09 8.98 9,14 i
45.00 49.60 8.30 F.44 50.00 55.12 7.64 7.74
'55.00 60.63 7.00 7.05 58.49 64.48 6.54 6.55 59.19-65.25 6.45 l
l l
i i
Page 12 l
w_--______:__-_
_ _ = _.
[
BRUNSWICK 1 J1103244SRLR l
Reload 11 Rev. 0 l
l l
l 16, Loss-of-Coolant Accident Results (cont.)
Bur.dle Type: GE13-P9DTB403-7G6.0/7G5.0 100T-146-T Average Planar Exposure MAPLIIGR (kw/ft)
(GWd/ST)
(GWd/MT)
Most Limiting Least Limiting 0.00 0.00 10.44 10.44 0.20 0.22 10.51 10.51 1.00 1.10 10.61 10.63 2.00 2.20 10.74 10.77 3.00 3.31 10.88 10.93 4.00 4.41 11.02 11.09 5.00 5.51 11.17 11.26 6.00 6.61 11.32 11.43 7.00 7.72 11.48 11.59 8.00 8.82 11.62 11.74 9.00 9.92-11.73 11.89 10.00 11.02 11.85 12.04 12.50 13.78 11.86 12.16 15.00 16.53 11.86 12.21 17.50 19.29 11.76 12.06 20.00 22.05 11.54 11.80 25.00 27.56 11.02 11.25 30.00 33.07 10.49 10.70 35.00 38.58 9.85 10.01 40.00 44.09 9.13 9.26 45.00 49.60 8.43 8.52 50.00 55.12 7.73 7.81 55.00 60.63 7.03 7.14 58.33 64.29 6.56 6.61 59.06 65.I1 6.49 Page 13
I 3
BRUNSWICK 1 J1103:44SRLR Reload I r Rev 0 gggggggggg 52 50
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B=GE13-P9DTB403-7G6.0nG5.0-100T-146-T (Cycle 12)
E= GE 13-P9DTB 380- 10G 5.0 A-100T-146-T (Cycle 11)
C=GElo-P81DW322 1IGZ-70M-150-T (Cycle 9)
F= G E 13-P9DTB380- 1 I G5.0 A-100T-146-T (Cycle 11)
~
Figure 1 Reference Core Loading Pattern Page 14
4 BRUNSWICK 1-J1103244SRLR
' Reload 11 Rev. O
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1 mwd /ST) ICF_IIBB )
4 i
Page 15 i
i
r BRUNSWICK 1 11103244SRLR
'*eload 11 Rev. 0 l
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~
i Page 16 j
l J
i BRUNSWICK 1 11103244SRLR Reload 11 Rev. 0 t
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Figure 4 Plant Response to FW Controller Failure (BOC12 to EOCl2 TBPOOS_ NOM ).
Page 17
t J1103244SRLR g..
BRUNSWICK 1 Reload I1 Rev. 0
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' ' BRUNSWICK 1 Jl103244SRLR Reloa'd i1:
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A Page 19 "
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BRUNSWICK 1 Jl103244SRLR Reload i1 Rev. O Appendix A Analysis Conditions To reflect actual plant parameters accurately, the values shown in Table A-1 were used this cycle.
Table A-1 Analysis Value Parameter ICF_IIBB FWTR_IIBB Thermal power, MWt 2558.0 2558.0 Core flow, Mib/hr 80.3 80.3 Reactor pressure, psia 1060.8 1041.2 Inlet enthalpy. BTU /lb 530.7 517.0 Non-fuel power fraction 0.036 0.036 Steam flow analysis, Mlb/hr 11.09 9.67 Dome pressure, psig 1030.0 1011.7 Turbine pressure, psig 984.8 976.9 No. of Safety / Relief Valves 9
9 Relief mode lowest setpoint, psig 1164.0 1164.0 Recirculation pump power source on-site '
on-site '
Turbine control valve mode of operation Partial are Partial arc I
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Bounds operation with oft-site power source for reload licensing events for Cycle 12.
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4 BRUNSWICK 1 J1103244SRLR Edp3d iI Rev. Q Appendix B Main Steamline isolation Valve Out of Service (MSIVOOS)
Reference B-1 provided a basis for operation of Brunswick Steam Electric Plant (BSEP) with one Main Steamline Isolation Valve Out of Service (MSIVOOS) (three steamline operation) and all S/RVs in service.
i For this mode of operation in BSEP Unit I tLoughout Cyc!c 12, the MCPR limits presented in Section 11_
of this report are bounding and should be applied when operating in the MSIVOOS mode at any time during the cycle. *Re peak steamline and peak ves;ci pressures for the limiting overpressurization event (MSIV closure with flux scram) were not calculated for the MSIV0OS mode of operation. In this mode of
- operation it ir required that all S/RVs be operational versus the assumed 2 S/RVs OOS for the events evaluated during normal pleit operation. Previous cycles analyses have shown that MSIV closure with flux scram, evaluated in the MSIVOUS mode, has resulted in the peak vessel pressure being reduced by -
more than 25 psi, when compared to the same case evaluated with all (four) stearnlines operational.
.- Reference B-l. Main Steamline Isolation Valve Out ofService for the Brunswick Steam Electric Plant EAS-l17-0987, GE Nuclear Energy, April 1988.
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- BRUNSWICK 1 ;
Jl103244SRLR Reload 11 Rev. ' O Appendix C Decrease li1 Core Coolant Temperature Events The Loss of Feedwater Heater (LFWH) event and the HPCI inadvenent start-up event are the only cold water injection AOOs checked on a cycle-by-cycle basis. A Cycle 11 analysis showed a LFWH ACPR'of 0.12. The HPCI event was shown to be bounded by the LFWH event for Cycle 11. There is no reason why these events would be expected to be more severe for Cycle 12. The results of the AOOs presented in Section 11 of this report sufTsciently bound the expected results of the LWFH and HPCI inadvertent stan-up events, therefore these events were not analyzed for Cycle 12.
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BRUNSWICK 1 Jl103244SRLR ILcloadi1 Rev O Appendix D Feedwater Temperature Reduction (FWTR)
Reference D 1 provides the basis for operation of Brunswick Steam Electric Plant (BSEP) with FWTR.
He MCPR limits presented in Section ii of this report are bounding and should be applied when operating with IMR. Previous analysis has shown the FWCF event is most severe at ICF and FWTR. The analyses used to calculate FWTR limits were based on constant turbine pressare which bounds constant dome pressure.
Reference D-l. Feedwater Temperature Reduction with Manmum Extended Load line and Jrcreased Core Flowfor Brunswick Steam Electric Plant Units 1 and 2.NEDC-32457P, Revision l, December 1995.
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' BRUNSWICK 1 Jl103244SRLR Reload 1I Rev. 0
- Appendix E Maximum Extended Operating Domain (MEOD) i
- Reference E-l provided a basis for operation of Brunswick Steam Electric Plant (BSEP) in the Max mum Extended Operating Domain (MEOD). Previous cycles have shown that these low flow conditions are bounded by ICF, therefore this domain was not analyzed for Cycle 12.. Application of the GEXL-PLUS correlation to the reload fuel has been confirmed as required in reference E-1. The applicability of GE13 was addressed and found acceptable.
Reference E-1. Maximum Extended Operating Domain Analysis for the Brunswick Steam Electric Plant.NEDC-31654P, GE Nuc! car Energy (Proprietary), February 1989.
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- BRUNSWICK 1-31103244SRLR Reload 11 Rev. 0 Appendix F Turbine Bypass Out of Service (TBPOOS)
Reference F-1 provides a basis for operation of Brunswick Steam Electric Plant (BSEP) with all Turbine j_
Bypass Valves Out of Senice (TBPOOS) and 2 S/RVs Out of Senice (2 SRVOOS). ' Reference F-1 has been confirmed applicable to the operation of Brunswick-1, Cycle 12. This is the first cycle of operation with this operating domain. Section 11 of this report presents the MCPR limits for the modes of operation with TBPOOS.
Reference F-l. Turbine Bypass Out ofService Analysisfor Carolma Power & Light Company's Br mswick Nuclear Plant Units I and 2.NEDC-32813P, Revision 1, GE Nuclear Energy (Proprietary), March 1998.
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