BSEP 07-0026, Enclosure 1 - Design Calc. 21321-1267, Revision 0, Brunswick, Unit 2, Cycle 18 Core Operating Limits Report March 2007
| ML070990200 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/31/2007 |
| From: | Blom M, Dresser T BWR Fuel Engineering |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| BSEP 07-0026 21321-1267, Rev 0 | |
| Download: ML070990200 (70) | |
Text
BSEP 07-0026 Brunswick Unit 2, Cycle 18 Core Operating Limits Report March 2007
NGG Nuclear Fuels Mgmt. & Safety Analysis 132C18 Core Operating Limits Report Design Calc. No. 21321-1267 Page 1, Revision 0 BRUNSWICK UNIT 2, CYCLE 18 CORE OPERATING LIMITS REPORT March 2007 Blom, Michael 2007.03.28 1,3:54:48 -04'00O Prepared By:
Approved By:
Michael Blom BWR Fuel Engineering Digitally signed by Dresser. Thornas
- Dresser, ON:m=rDreassor.Thomas.
ou-NGG. Uses HNP. em~il=Ioo,.drosser@pgnmril.hoflr dc-com, progress-negy. zonel. oak T:wn ho fli as Tom Dresser Supervisor, Nuclear Fuel Supply
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C1 8 Core Operating Limits Report Design Cabc. No. 2B21-1267 Page 2, Revision 0 LIST OF EFFECTIVE PAGES Page(s) 1-30 Revision 0
NGG Nuclear Fuels Mgmt. & Safety Analysis Design Caic. No. 2B321-1267 B32C18 Core Operating Limits Report Page 3, Revision 0 TABLE OF CONTENTS Subject Pa2e Cover.........................................................................................................
1..
List of Effective Pages........................................................................................
2 Table of Contents..............................................................................................
3 List of Tables...................................................................................................
4 List of Figures..................................................................................................
4 Introduction and Summary....................................................................................
5 Single Loop Operation........................................................................................
5 Inoperable Main Turbine Bypass System....................................................................
6 Feedwater Temperature Reduction...........................................................................
6 APLHGR Limits...............................................................................................
7 MCPR Limits...................................................................................................
7 RBM Rod Block Instrumentation Setpoints.................................................................
7 Stability Option III.............................................................................................
7 Power/Flow Maps..............................................................................................
8 References......................................................................................................
9
NGG Nuclear Fuels Mgmt. & Safety Analysis Design Oalc. No. 21321-1267 B2C18 Core Operating Limits Report Page 4, Revision 0 CAUTION References to COLR Figures or Tables should be made using titles only; figure and table numbers may change from cycle to cycle.
LIST OF TABLES Table Title Page Table 1:
MCPR Limits.............................................................................................
10 Table 2:
RBM System Setpoints.................................................................................11.I Table 3:
PBDA Setpoints..........................................................................................
12 LIST OF FIGURES Figure Title or Description Page Figure 1:
APLHGR Limit Versus Average Planar Exposure - EDB # 2572..................................... 13 Figure 2:
APLHGR Limit Versus Average Planar Exposure - EDB # 2573....................................
14 Figure 3:
ALPLFIGR Limit Versus Average Planar Exposure - EDI3 # 2574...................................
15 Figure 4:
APLHGR Limit Versus Average Planar Exposure - EDB #I 2575....................................
16 Figure 5:
APLHGR Limit Versus Average Planar Exposure - EDB # 2660....................................
17 Figure 6:
APLHGR Limit Versus Average Planar Exposure - EDB # 2853....................................
18 Figure 7:
APLHGR Limit Versus Average Planar Exposure - EDB # 2854....................................
19 Figure 8:
[not used]..................................................................................................
20 Figure 9:
Flow-Dependent MAPLHGR Limit, MAPLHGR(F).................................................
21 Figure 10: Power-Dependent MAPLHGR Limit, MAPLHGR(P)................................................
22 Figure 11: Flow-Dependent MCPR Limit, MCPR(F).............................................................
23 Figure 12: Power-Dependent MCPR Limit, MCPR(P)............................................................
24 Figure 13: Stability Option III Power/Flow Map: OPRM Operable, Two Loop Operation, 2923 MWt.....25 Figure 14: Stability Option III Power/Flow Map: OPRM Inoperable, Two Loop Operation, 2923 MWt....26 Figure 15: Stability Option Ill Power/Flow Map: OPRM Operable, Single Loop Operation, 2923 MWt....27 Figure 16: Stability Option Ill Power/Flow Map: OPRM Inoperable, Single Loop Operation, 2923 MWt..28 Figure 17: Stability Option Ill Power/Flow Map: OPRM Operable, FWTR, 2923 MWt...................... 29 Figure 18: Stability Option III Power/Flow Map: OPRM Inoperable, FWTR, 2923 MWt....................
30
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C1 8 Core Operating Limits Report Design Cale. No. 21321-1267 Page 5, Revision 0 Introduction and Summary CAUTION References to COLR Figures or Tables should be made using titles only; figure and table numbers may change from cycle to cycle.
This report provides the values of the power distribution limits and control rod withdrawal block instrumentation setpoints for Brunswick Unit 2, Cycle 18 as required by TS 5.6.5. Revision 0 of this report supports operation at up to 2923 MWt. The main changes are those associated with the thermal limits and Power-Flow maps.
OPERATING LIMIT REQUIREMENT Average Planar Linear Heat Generation Rate (APLHGR) limits TS 5.6.5.a.1 (with associated core flow and core power adjustment factors)
Minimum Critical Power Ratio (MCPR) limits TS 5.6.5.a.2 (with associated core flow and core power adjustment factors)
Period Based Detection Algorithm (PBDA) Setpoint for Function 2.f of TS 3.3. 1. 1, TS 5.6.5.a.3 Oscillation Power Range Monitor (OPRM)
Allowable Values and power range setpoints for Rod Block Monitor Upscale TS 5.6.5.a.4 Functions of TS 3.3.2.1 Per TS 5.6.5.b and 5.6.5.c, these values have been determined using NRC approved methodology and are established such that all applicable limits of the plant safety analysis are met. The limits specified in this report support single recirculation loop operation (SLO) as required by TS LCO 3.4.1 and inoperable Main Turbine Bypass System as required by TS 3.7.6.
In order to support the Stability Option III with an inoperable OPR.M scram function, the following is also included in this report:
OPERATING LIMIT REQUIREMENT BWROG Interim Corrective Action Stability Regions TS LCO 3.3. 1. 1, Condition I This report conforms to Quality Assurance requirements as specified in Reference 1.
Sin2le Loop Operation Brunswick Unit 2, Cycle 18 may operate over the entire MEOD range with SLO over the entire cycle as permitted by TS 3.4.1 with applicable limits specified in the COLR for TS LCO's 3.2. 1, and 3.2.2.
The applicable limits are:
LCO 3.2. 1, Average Planar Linear Heat Generation Rate (APLHGR) Limits: per Reference 1, the Figures 9 and 10 described in the APLHGR Limits section below include a SLO limitation of 0. 8 on the MAPLHGR(F) and MAPLHGR(P) multipliers.
NGG Nuclear Fuels Mgmt. & Safety Analysis Design Caic. No. 21321-1267 B2C18 Core Operating Limits Report Page 6, Revision 0 LCO 3.2.2, Minimum Critical Power Ratio (MCPR) Limits: per Reference 1, Table 1 and Figures 11I and 12, the MCPR limits presented apply to SLO without modification.
Various indicators on the Power/Flow maps are provided not as operating limits but rather as a convenience for the operators: a single ioop operation (SLO) Entry Rod Line is shown on the two ioop operation maps to avoid regions of instability in the event of a pump trip; a maximum core flow line is shown on the single loop operation maps to avoid vibration problems; a maximum core power of 50%
RTP in SLO mitigates a spurious trip signal which could result from APRM noise;, and Average Power Range Monitors (APRM) Simulated Thermal Power (STP) Scram and Rod Block nominal trip setpoint limits are shown to illustrate where these setpoints occur. Note that the APRM STP setpoints are only approximations, shown at the estimated core flow corresponding to the actual drive flow-based setpoints. The approximations serve to indicate where the operator may encounter the APRMSTP -
High Allowable Value setpoints (LCO 3.3. 1. 1, Table 3.3. 1. 1-1: Reactor Protection System Instrumentation, Function 2. b).
Inoperable Main Turbine Bypass System Brunswick Unit 2, Cycle 18 may operate with an inoperable Main Turbine Bypass System in accordance with TS 3.7.6 with applicable limits specified in the COLR for TS LCO 3.2.1 and LCO 3.2.2. The applicable limits are as follows:
LCO 3.2. 1, Average Planar Linear Heat Generation Rate (APLHGR) Limits: in accordance with Reference 1 as shown in Figure 10, TBPOOS does not require an additional reduction in the MAPLGHR(P) limits, as the Turbine Bypass Operable and Inoperable limits are identical.
LCO 3.2.2, Minimum Critical Power Ratio (MCPR) Limits: in accordance with Reference 1, TBPOOS does not require an additional increase in the MCPR(P) multiplier as shown in Figure 12, as the Turbine Bypass Operable and Inoperable limits are identical. TBPOOS requires increased MCPR limits, included in Table 1.
The system response time assumed by the safety analyses from event initiation to start of bypass valve opening is 0.10 seconds, with 80% of the bypass flow capacity achieved in 0.30 seconds Although the Turbine Bypass Out-of-Service (TBPOOS) analysis supports operation with all bypass valves inoperable for the entire maximum extended operating domain (MEOD) range and up to 1 I100F rated equivalent feedwater temperature reduction, three or more turbine bypass valves inoperable renders the Main Turbine Bypass System inoperable.
Feedwater Temperature Reduction A variation within 1 0O F of nominal feedwater temperature or a power level less than 3 0% RTP has been evaluated as being in compliance with normal feedwater temperature operating limits. AI feedwater temperature reduction in excess of 1 0O F together with the reactor power at a minimum of 30% rated requires the use of the Reduced FWTR MCPR limits (Table 1) and Stability Option III Power/Flow limits (Figures 17 and 18).
NGG Nuclear Fuels Mgmt. & Safety Analysis Design Caic. No. 21321-1 267 B2C18 Core Operating Limits Report Page 7, Revision 0 APLHGR Limits The limiting APLHGR value for the most limiting lattice (excluding natural uranium) of each fuel type as a function of planar average exposure is given in Figures 1 through 7. These values were determined with the SAFER/GESTR LOCA methodology described in GESTAR-11 (Reference 2).
Figures 1 through 7 are to be used only when hand calculations are required as specified in the bases for TS 3.2. 1. Hand calculated results may not match a POWERPLEX calculation since normal monitoring of the APLHGR limits with POWERPLEX uses the complete set of lattices for each fuel type provided in Reference 3.
The core flow and core power adjustment factors for use in TS 3.2.1 are presented in Figures 9 and 10.
For any given flow/power state, the minimum of MAPLHGR(F) determined from Figure 9 and MAPLHGR(P) determined from Figure 10 is used to determine the governing limit.
MCPR Limits The Scram Speed MCPR OPTION A, OPTION B, and non-pressurization transient MCPR limits for use in TS 3.2.2 for each fuel type as a function of cycle average exposure are given in Table 1. These values were determnined with the GEMINI(TRACG) methodology and GEXL14 critical power correlation described in GESTAR-11 (Reference 2) without assuming EOC-RPT, and are consistent with a Safety Limit MCPR of 1. 11 specified by TS 2.1.1.2.
The core flow and core power adjustment factors for use in TS 3.2.2 are presented in Figures I11 and
- 12. For any given power/flow state, the maximum of MCPR(F) determined from Figure 11I and MCPR(P) determined from Figure 12 is used to determine the governing limit. All MCPR limits presented in Table 1, Figure 11I and Figure 12 apply to two recirculation pump operation (TLO) and SLO without modification.
RBM Rod Block Instrumentation Setpoints The nominal trip setpoints and allowable values of the control rod withdrawal block instrumentation for use in TS 3.3.2.1 (Table 3.3.2. 1-1) are presented in Table 2. These values were determined to be consistent with the bases of the ARTS program, and the determination of MCPR limits with the GEMINI(TRACG) methodology and the GEXL-PLUS critical power correlation described in GESTAR-11 (Reference 2) as well as the NUMAC PRNM system as discussed in Reference 8. The table also includes the cycle-specific MCPR limits regarding the required operability of the RBM, consistent with Technical Specification Table 3.3.2. 1 -1.
Stability Option III Brunswick Unit 2 has implemented BWROG Long Termn Stability Solution Option III using Oscillation Power Range Monitors (OPRMs) with the methodology described in Reference 4. Plant specific analysis incorporating the Option III hardware is described in Reference 5. Reload validation has been performed in accordance with Reference 6. The resulting stability based MCPR Operating Limit is provided for two conditions as a function of OPRM amplitude setpoint in Table 3. If desirable, Table 3 would support higher stability limits for various MCPR operating limits greater than the least limiting AQO OLMCPR values provided in Table 1 or Figures I11 or 12, but the suggested stability setpoints ensure no OLMCPR restrictions from Option III. Table 3 shows that for an OPRM setpoint
NGG Nuclear Fuels Mgmt. & Safety Analysis Design Caic. No. 21321-1267 B2C18 Core Operating Limits Report Page 8, Revision 0 (Amplitude Setpoint Sp) of 1. 13, OLMCPR(SS) is less restrictive than Figure 11I at 45% maximum flow or Figure 12 at 60.6% maximum power. Table 3 also shows that OLMCPR(2PT) is less restrictive than any limiting Table I OLMCPR for Sp equal to 1. 13. Therefore the OPRM Period Based Detection Algorithm (PBDA) setpoint limit for Amplitude Sp referenced by fuinction 2.f of Table 3.3. 1. 1-1 of Technical Specification 3.3. 1.1 is 1. 13 for Cycle 18. Per Table 3-2 of Reference 6, an Sp value of 1. 13 supports selection of a Confirmation Count Setpoint NP of 15 or less.
Stability Option III also affects the power/Flow maps as described below.
Power/Flow Maps
'Six Power/Flow maps for use at up to 2923 MWt (Figures 13-18) were developed based on References 1 and 7 to facilitate operation under Stability Option III as implemented by LCO 3.3.1..1, Condition I and function 2.f of Table 3.3.1.1-1 of the Technical Specifications. All six maps illustrate the region of the power/flow map above 25% power and below 60% drive flow (which is approximately 63-64%
core flow) where the system is required to be enabled. [Note that the power/flow maps can only approximate the licensed Enabled Region, because the maps display core flow, while the Enabled Region is based on drive flow]
Figures 13, 15 and 17 support an operable OPRM function 2.f and Figures 14, 16 and 18 support an inoperable OPRM function 2.f for TLO, SLO or FWTR, respectively. Each figure for an operable OPRM shows a Scram Avoidance Region where the OPRM system may generate a scram to avoid an instability event. The figures for an inoperable OPRM event show the additional stability regions which address BWROG-94078 Interim Corrective Actions required to support LCO 3.3. 1. 1, Condition I. Figures 14, 16 and 18 for OPRIV inoperable also includes a 5% Buffer Region around the Immediate Exit Region as an operator aid [Note for Figure 16 (SLO), the 5% Buffer Region does not fully enclose the Immediate Exit Region as on Figures 14 and 18].
Figures 15 and 16 for SLO show the maximum allowable core flow at 45 Mlbs/hr and has the STP scram and rod block limits appropriately reduced. Note that the STP scram and rod block limits are defined in Technical Specifications, the Technical Requirements Manual, and Plant procedures.
Figures 15 and 16 also implement the corrective action for A/R 217345217345which restricts reactor operation to no more than 50% RTP in SLO with OPRM operable or inoperable. This operator aid is intended to mitigate a spurious trip signal which could result from APRM noise while operating at high power levels.
Figures 17 and 18 for FWTR differ from ILO and SLO by including extended scram regions to provide additional stability protection. Although the scram region for SLO above 50% RTP is more restrictive, intentional operation with SLO and FWTR is prohibited.
NGG Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 2B21-1 267 B2C18 Core Operating Limits Report Page 9, Revision 0 References
- 1)
BNP Design Calculation 2B321-1267, "Preparation of the B32C 18 Core Operating Limits Report," Revision 0, March 2007.
- 2)
NEDE-2401 1 -P-A (including U.S. supplement), "General Electric Standard Application for Reactor Fuel," Revision 15.
- 3)
NEDC-31624P, "Loss-of-Coolant Accident Analysis Report for Brunswick Steam Electric Plant Unit 2 Reload 17 Cycle 18," Supplement 2, Revision 10, January 2007.
- 4)
NEDO-3 1960-A, "BWR Owners Group Long-Term Stability Solutions Licensing Methodology," November 1995.
- 5)
GE-NE-CS 1-00251-00-01, Revision 0, "Licensing Basis Hot Bundle Oscillation Magnitude for Brunswick 1 and 2," March 2001.
- 6)
NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.
- 7)
BNP Design Calculation 0B321-1015, Revision 6, "Power/Flow Maps," March 2007.
- 8)
BNP Design Calculation 2C5 1-0001 Revision 3, "Power Range Neutron Monitoring System Setpoint Uncertainty and Scaling Calculation (2-C5 1 -APRM 1 through 4 Loops and 2-C5 I1-RBM-A and B Loops," May 26, 2004.
- 9)
NEDE-32906P-A, Revision 1, and Supplements, "TRACG Application for Anticipated Operational Occurrences (AOO) Transient Analyses," April 2003.
- 10)
GE-NE-0000-0022-8 180-RO, "Brunswick Nuclear Station TRACG Implementation for Reload Licensing Transient Analysis," February 2004.
- 11)
GE-NE-0000-0036-9469-R1, Revision 1, "Brunswick 1 and 2 Off-Rated Analyses Below the PLU Power Level," April 2006.
- 12)
Global Nuclear Fuel Document 0000-0052-0675-SRLR, Rev. 0; "Supplemental Reload Licensing Report for Brunswick Steam Electric Plant Unit 2, Reload 17 Cycle 18," February 2007.
NGG Nuclear Fuels Mgmt. & Safety Analysis 132C18 Core Operating Limits Report Design CaIc. No. 21321-1267 Page 10, Revision 0 Table 1 MCPR Limits (EOC RPT not required)
Non-pressurization Transient MCPR Limits Pressurization Transient MCPR Limits 100% Power OLMCPR Turbine Feedwater Scram Exposure Range:
Exposure Range:,
Bypass Temperature Fuel Type Speed OtoER50Md/To System Normal?
MCPR BOO590 to/M EO-5 MdMTt Operable?
Option ER90MdM O
Operable Normal GE14 A
1.53 1.57 B
1.35 1.39 Operable Reduced GE14 A
1.53 1.57 B
1.35 1.39 Inoperable Normal GE14 A
1.61 1.61 B
1.43 1.43 Inoperable Reduced GE14 A
1.61 1.61 B
1.43 1.43 This Table is referred to by Technical Specifications 3.2.2, 3.4.1 and 3.7.6.
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C18 Core Operating Limits Report Design CaIc. No. 21321-11267 Page 11, Revision 0 Table 2 RBM System Setpoints ta Setpoint Trip Setpoint Allowable Value Lower Power Setpoint (LPSP b 27.7
<- 29.0 Intermediate Power Setpoint (PSP b) 62.7
< 64.0 High Power Setpoint (HPSP b 82.7
- - 84.0 Low Trip Setpoint (LTSPC)
- 114.1
- 114.6 Intermediate Trip Setpoint (ITSPc)
< 108.3
- 108.8 High Trip Setpoint (HTSPc)
- ý 104.5
- 105.0 RBM Time Delay NOd) 2.0 seconds
- 2.0 seconds a RBM Operability requirements are not applicable:
(1) if MCPR> 1.76; or (2) if MCPR Ž 1.45 and thermal power Žý 90% Rated Thermal Power.
b Setpoints in percent of Rated Thermal Power.
C Setpoints relative to a full scale reading of 125.
For example, *ý114.1 means! *114.1/125.0 of full scale.
This Table is referred to by Technical Specification 3.3.2.1 (Table 3.3.2.1-1).
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C18 Core Operating Limits Report Design Cabc. No. 21321-1267 Page 12, Revision 0 Table 3 PBDA Setpoints OPRM Setipoint OLMCPR(SS)
OLMCPR(2PT')
1.05 1.224 1.097 1.06 1.248 1.119 1.07 1.273 1.142 1.08 1.300 1.165 1.09 1.327 1.190 1.10 1.356 1.215 1.11 1.384 1.241 1.12 1.414 1.268 1.13 1.445 1.295 1.14 1.477 1.324 1.15 1.511 1.355 Acceptance Criteria Off-rated OLMCPR @
Rated Power OLMCPR 45% Flow, 60.6% power PBDA Setpoint Setpoint Value Amplitude SP 1.13 Confirmation Count NP 15 This Table is referred to by Technical Specification 3.3.1.1 (Table 3.3.1.1-1).
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C18 Core Operating Limits Report Design Calc. No. 2B21-1 267 Page 13, Revision 0 Figure 1 Fuel Type GEI4-PI ODNAB42O-1 8GZ-1 OOT-1 50-T-2572 Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure - EDB# 2572 12.0 11.0 10.0 Tis Figure is Referred To By
___I.
111Technical Specification 3.2.1 a-9.0-8.0 7.0-Exposure Limit (GWd/Mt) (kW/ft) 0.00 9.29 0.22 9.37 1.10 9.51 2.20 9.70 3.31 9.91 4.41 10.111 5.51 10.26 6.61 10.40 7.72 10.53 8.82 10.66 9.92 10.78 11.02 10.89 12.13 11.00 13.23 10.99 14.33 10.97 15.43 10.96 16.53 10.93 18.74 10.86 22.05 10.67 27.56 10.25 33.07 9.80 38.58 9.29 44.09 8.81 49.60 8.34 55.12 7.86 60.63 5.62 62.23 4.90 I N.
I j
I
- t
____ +/-
____ t t ____
-I + I 1'
____ I I ____
____________ I
____ I Permissible Region of Operation I __
I I
I IT I
IT 1
____ T I
-I
____ T I it _____-1~
____ I
____ 1-I _____ ~i
____ I I ____
6.0 5.0-4.0-0 5
10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWdIMT)
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C18 Core Operating Limits Report Design Calc. No. 2B21-1267 Page 14, Revision 0 Figure 2 12.0 11.0 10.0 I-0=
-J 0~
9.0 8.0 7.0 Fuel Type GEl 4-PI ODNAB4I 9-6G7.0/7G6.013G2.O-1 OOT-1 50-T-2573 Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure - EDB# 2573
___ {
'-I is Figure is Referred To By Technical Specification 3.2.1 Exposure Limit
-Ij b
(GWd/Mt)
(kW/ft)
N___I______
0.00 9.24 0.22 9.31 1.10 9.43
-I__
2.20 9.60
[
I___
3.31 9.77 I_________
4.41 9.96 F__
~5.51 10.113 6.61 10.27 1
1 1
1 7.72 10.40 8.82 10.52
[__
_I_
9.92 10.64 F__
_I_
_ T_
11.02 10.77 F
12.13 10.89 Permissible 13.23 10.88 Region of 14.33 10.88 jOperation 15.43 10.88 1
1 16.53 10.87 ~ I___F I__
18.74 10.83 ~
22.05 10.67 27.56 10.23
-I 33.07 9.78 ~
38.58 9.33
-I___
44.09 8.87
-IA 49.60 8.37 55.12 7.84 60.63 5.55 62.07 4.90 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWdIMT) 6.0 5.0 4.0
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C18 Core Operating Limits Report Design Caic. No. 2B21-1 267 Page 15, Revision 0 Figure 3 12.0 11.0 10.0
(-
0 9.0 8.0 7.0 Fuel Type GEl 4-PI ODNAB425-3G7.O/14G6.O/1 G2.O-l OOT-l 50-T-2574 Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure - EDB# 2574
___T
..... Th is Figure is Referred To By I
I 11 1 Technical Specification 3.2.1 Exposure Limit
_II (GWd/Mt) (kW/ft)
__I I_
2.20 9.70
__1__
3.31 9.85
__I___
___I 4.41 10.01 1_
5.51 10.16
{_
_ I___
6.61 10.30 1__
7.72 10.45 8.82 10.60
_1_
_ 1 _
_1_
9.92 10.70 11.02 10.79 j_
12.13 10.87
~
13.23 10.87 14.33 10.85 1emisTl 15.43 10.84 Rermissible 16.00 10.84 OpRegaion o 16.53 10.84i Oprto 18.74 10.82 1_1_
21.09 10.72 1
__1__
22.05 10.68
~
27.56 10.26 33.07 9.82 I__________
38.58 9.33 44.09 8.83 49.60 8.35 55.12 7.85 6.0 5.0 4.0 0
5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWd/MT)
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C1 8 Core Operating Limits Report Design Calc. No. 2B21-1 267 Page 16, Revision 0 Figure 4 12.0 11.0 10.0 I-
-j 9.0 8.0 7.0 Fuel Type GEl 4-PI ODNAB439-1 2G6.O-1 OOT-1 50-T-2575 Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure - EDB# 2575 zJThis Figure is Referr ed To By
-~-Technical Specification 3.2.1 Exposure Limit (GWd/Mt)
(kW/ft) 0.00 9.68 0.22 9.72 1.10 9.79 2.20 9.89 3.31 9.99 4.41 10.09 5.51 10.20 6.61 10.31 7.72 10.43 8.82 10.55 1
1 9.92 10.67 11.02 10.79 12.13 10.92 13.23 10.93 14.33 10.92 Permissible 15.43 10.90 Region of 16.00 10.89 Operation 16.53 10.88 18.74 10.81 21.09 10.66 22.05 10.60 27.56 10.18 33.07 9.76 38.58 9.32 44.09 8.87 49.60 8.37 55.12 7.83 60.63 5.54 62.06 4.88 0
5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWdIMT) 6.0 5.0 4.0
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C18 Core Operating Limits Report Design Calc. No. 2B21-1267 Page 17, Revision 0 Figure 5 12.0 11.0 10.0 F-a.
9.0 8.0 7.0 Fuel Type GEl 4-PI ODNAB4I 3-1 6GZ-1 OOT-1 50-T-2660 Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure - EDB# 2660 This Figure is Referred To By
-Technical Specification 3.2.1 (GWd/Mt) (kW/ft)
__I______
4.41 10.08 7 5.51 10.26___
6.61 10.44 7.72 10.59 8.82 10.74 11.0 11.00.
12.13 11.12 13.23 11.15 14.33 11.16 Permissible 15.43 11.16 Region of 16.00 11.16 16.53 11.16 Operation 18.74 11.13 21.09 11.02 22.05 10.98
~
27.56 10.57 33.07 10.15 38.58 9.65 44.09 9.121 49.60 8.59 55.12 8.04--
60.63 6.'48 63.50 5.18 64.16 4.88 0
5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWd/MT) 6.0 5.0 4.0
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C18 Core Operating Limits Report Design Calc. No. 2B21-1 267 Page 18, Revision 0 Figure 6 12.0 11.0 10.0 x
9.0 8.0 7.0 Fuel Type GEl 4-P1 ODNAB407-1 6GZ-1 OOT-l 50-T-2853 Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure - EDB# 2853
_This Figure is Referred To By Technica Specification 3.2.1TY Exposure Limit (GWd/Mt)
(kW/ft) -
0.00 9.47 1.10 9.71
~
2.20 9.87
~
3.31 10.02 4.41 10.17 5.51 10.31 6.61 10.45 7.72 10.57 8.82 10.70 -I___1 9.92 10.82 j___
_I___
11.02 10.93 12.13 11.05
_ I_
13.23 11.06 14.33 11.05 Permissible 15.43 11.05 Region of 16.00 11.05 Operation 16.53 11.05 18.74 11.02
.1 21.09 10.91 t
22.05 10.87 27.56 10.49 33.07 10.08 38.58 9.58 44.09 9.07 49.60 8.54
~-
55.12 7.'99 60.63 6.34 jj~J 63.50 5.03 63.8 4.8-8 0
5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWd/MT) 6.0 5.0 4.0
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C18 Core Operating Limits Report Design Calc. No. 2B21-1 267 Page 19, Revision 0 Figure 7 Fuel Type GEl 4-P1 ODNAB425-1 8GZ-1 OOT-1 50-T-2854 Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure - EDB# 2854 12.0 11.0 10.0 9.0 M
8.0 0
7.0 6.0 5.0 4.0 0
5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWdIMT)
NGG Nuclear Fuels Mgmt. & Safety Analysis E32C18 Core Operating Limits Report Design Caic. No. 21321-1267 Page 20, Revision 0 Figure 8
[Not Used]
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C18 Core Operating Limits Report Design Calc. No. 2B21-1 267 Page 21, Revision 0 Figure 9 Flow-Dependent MAPLHGR Limit, MAPLHGR(F) 1.10 I
I 1.05
- 0. 95)
-L 0.90 0.85
<0.85 0.75 0.6 S0.80 0.5 0.5 This Figure is Referred To By[
Technical Specifications 3.2.1, 3.4.1 and 3.7.6 ITwo Loop Operation LimitlI
/
ax lw= 102.5% -
107%
112%
-IfSingle Loop Operation Limitf
__~-j~-
/
I,/ / /
0/or MAPLHGR(F) = MAPFACF
- MAPLHGRSTD MAPLHGRSTD = Standard MAPLHGR Limits MAPFACF(F) = Minimum (1.0, AFWC/IOO0+BF)
Wc= % Rated Core Flow AF And BF Are Fuel Type Dependent Constants Given Below:
Max Core Flow
(% Rated) 102.5 107.0 112.0 117.0 AF 0.6784 0.6758 0.6807 0.6886 BF 0.4861 0.4574 0.4214 0.3828 0.45 I
I I
I I
I I
I I
I 0.40 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 110 Core Flaw (% Rated)
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C18 Core Operating Limits Report Design CaIc. No. 2B21-1 267 Page 22, Revision 0 Figure 10 Power-Dependent MAPLHGR Limit, MAPLHGR (P) 1.05' 1.00-0.95-0.90' c.0.85-0.-0 2 0.75-U.)
0.-0 o 0.70 0.60
<0.655 0
- a. 0.55' I
rjcor Flo I
<65% Turbin
.-i Bypass Operable or InoperableL...
MAPLHGR(P) = MAPFACp
- MAPLHGRSTD MAPLHGRSTD = Standard MAPLHGR Limits
___4_ý
\\'
f Core Flow > 65%
Turbine Bypass Operable or Inoperable I
/
I I
I 1
1 1
0.45 0.40 0.35 Core Flow < 50% - Turbine Bypass Operable or InoperableF
__1 Core Flow < 65% - Turbine
~Bypass Operable or Inoperable For P < 23%:
No Thermal Limits Monitoring Required
-or 23% <_ P < 26% & Turbine Bypass Operable or Inoperable:
For Core Flows~ 50%:
MAPFACp = 0.567 + 0.0157 (P - 26%)
For Core Flow:5 65%:
MAPFACp = 0.495 + 0.0130 (P - 26%)
For Core Flow > 65%:
MAPFACp = 0.433 + 0.0063 (P - 26%)
For 26% :5 P < 40% & Turbine Bypass Operable or Inoperable:
For Core Flow *ý 65%:
MAPFACp = 1.0 + 0.005224 (P - 100%)
For Core Flow > 65%:
MAPFACp = 0.634 + 0.0035 (P - 40%)
For P Ž 40%:
MAPFACp = 1.0 + 0.005224 (P - 100%)
-0ý
~>Core Flow > 65% - Turbine Bypass Operable or Inoperable
.1-4-I~
~~~~~~~~~~~
I 030 20 3¶0~
Pbypass (26%)
35 40 45 50 55 60 65 70 75 80 85 90 95 Power (% Rated) 100
NGG Nuclear Fuels Mgmt. & Safety Analysis B2C18 Core Operating Limits Report Design Caic. No. 21321-1267 Page 23, Revision 0 Figure 11 Flow-Dependent MCPR Limit, MCPR(F) 1.80
~ m I -
I i
i i
i Ii i
I
+
I-
-I I~
-~
MCPR(F) = Max (1.24, AFWC/IOO+BF)
Max Core Flow
(% Rated)
Aý B
102.5
-0.592 1.717 107.0
- 0.608 1.760 112.0
-0.625 1.812 117.0
-0.656 1.877
-I 1.70 1.60 o~1.50 1.40 1.30 1.20 1
-Max Flow =117%
_-This Figure is Referred To By Technical Specification 3.2.2,
___3.4.1 and 3.7.6 112%0/,_
107%
_102.5%_
_ I 20 30 40 50 60 70 80 90 100 110 120 Core Flow (% Rated)
NGG Nuclear Fuels Mgmt. & Safety Analysis 132C1 8 Core Operating Limits Report Design Calc. No. 2B21-1267 Page 24, Revision 0 Figure 12 Power-Dependent MCPR Limit, MCPR (P) a.
VI 0-It Al 0
V-3.80 3.70 3.60 3.50 3.40 3.30 3.20 3.10 3.00 2.90 2.80 2.70 2.60 2.50 2.40 2.30 2.20 2.10 2.00 1.90 1.80 1.70 1.60 1.50 1.40 1.30 1.20 1.10 1.00 OLIMCPR Rated MCPR Multiplier (Kp)
-Core Flow > 65% - Turbine
- Bypass Operable or Inoperable OprtnLitMCR OLMCPR(P) = Kp *OLMCPR(100)
____________For P < 23% :
No Thermal Limits Monitoring Required Core Flow < 65% - Turbine Bypass Operable or Inoperable For 23%:5 P < 26% & Turbine Bypass I -
I I
r 1
1 perable or Inoperable:
1For Core Flow:550%:
OLMCPR(P) = [ 2.34 + 0.0700(26% - P)]
Core Flow < 50% - Turbine For Core Flow 5 65%:
Bypass Operable or Inoperable OLMCPR(P) = [2.74 + 0.0967(26% - P)1 F E" ;r I
I IFor Core Flow > 65%:
L I
I OLMCPR(P) = [ 3.13 + 0.0900(26% - P)]
- Core Flow > 65% - Turbine For 26%
- 5 P < 40% & Turbine Bypass Bypass Operable or Inoperable Operable or Inoperable:
For Core Flow < 65%:
OLMCPR(P) =-1.88 + 0.0129 (40% - P)
____________For Core Flow > 65%:
OLMCPR(P) = 2.10 + 0.0157 (40% - P)
For 40:5 P < 45%:
Kp = 1.28 + 0.0135 (45% - P)
___For 45%:5 P < 60% :
Core Flow < 65% - Turbine Kp = 1. 18 + 0.00667 (60% - P)
Bypass Operable or In operable For P >Ž 60%
7777777Kp
= 1.00 + 0.00450 (100% - P)
This Figure is Referred To By Technical Specification
___3.2.2, 3.4.1, 3.7.6 20 2
Pbypass (26 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Power (% Rated)
NGG Nudlear Fuels Management & Safety Analysis B2C1 8 Core Operating Limits Report Figure 13 Stability Option Ill Power/Flow Map Design Calc. No. 21321-1 267 Revision 0 Page 25 OPRM Operable, Two Loop Operation, 2923 MWt IThis Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 I
120.0 110.0 100.0 90.0 80.0 70.0 60.0 a-
~
iI~ K A1/4APRM STP Rod Block
-~~~
I
- t.
, 0 M in imumn M.. im Urn (M ELLI)
(IC F)
C 0re C 0r*
P.wa,
- FlInw, Flonw.
M N1b sth r Mlb./h r 00 76.19 890.4 7 99 75.04 80.4 7 99a 7 3.869 99.4 7 97 72.7 5 90.47 96 7.1 6 0.4 7 9
70.49 90.47 94 69.3 6 904 9 3 69.25 90.47 92 67.13 90.4 7 9 1 66.03 90.4 7 90 64.93 90.4 7 9
6.63 9047 89 62.74 00.4 7 867 61.66 890.51 66 66.59 90.60 65 595 90.7 0 84 56-.4 3 90.79 863 57.37 900 62 56.31 910 61 505.25 01.2 1 g0 04.260 91.36 79 53.16 91.52 78 52.12 81.67 77 5 1.09 681.8 3 76 500.0 5 a91.98 75 49.02 82.1 3 74 48.00 82.29 7 3 46.98 82.44 72 4 5.9 6 92.60 7 1 44.95 82.7 5 70 43.94 82.9 1 69 42.94 93.06 68 41.94 93.2 1 67 40.95 93.3 7 66 3 9.96 93.5 2 65 398.97 9 3.668 64 3 7.99 893.893 8 3 3 7.01 893.9 9 62 36.04 94.1 4 6 1 3 5.0 6 94.2 9 60 34.10 84.45 5 9 3 3.13 94.6 0 59a 32317 94-7 0 50.0 40.0 30.0 20.0 10.0 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbslhr Core Flow 0
10 20 30 40 50 60 70 80 90 100 110 120 % Core Flow
NGG Nuclear Fuels Management & Safety Analysis Figure 14 Design Caic. No. 2B21-1 267 B2C1 8 Core Operating Limits Report Stability Option Ill Power/Flow Map Revision 0 OPRMV Inoperable, Two Loop Operation, 2923 MWtPae2 This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 MinImuMM M axim Um APM TPScam(ME LL L (IC F 110.0 Co.
Coare Pow.r Flo*w,
- Flow, HAPRM STP Rod Block
~
Mlbslhr MUlbs/hr 70.08
.4
- 90.
85 59.50 85.470 84 5 8.43 80.479 8 0.907 500 79 3.16 813 7
52.1 810.67 7E~-
a-77 a10 01.8 70.L Entr RoLn-65.5 81a qu~ ReionA-anul tC84 48.03 82.28 73 467.98 82.44 Rgin B Imedat Ext 1
44.25 81217 7A a
43.94 82'.91 rn89 42.14 83.58 5008 8
42128 83.21
- 20.
187.8 39.8
.83.5 64 376136 MEL.ieEtyR n
63 5 7051 8.198a 40.0-bedego 62io A36aua CAM7
.04 0 84.14 60 34.10 84.45 52 33.13 84.60 30.0 70 432.17 84.701 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbslhr Core Flow 0
10 20 30 40 50 0 0 2 0
4 0
60 70 80 90 100 110 120 % Core Flow
NGG Nuclear Fuels Management & Safety Analysis Figure 15 Design Cabc. No. 2B21-1 267 B2C1 8 Core Operating Limits Report StabIiit Option III Power/Flow Map Revision 0 OPRM Operable, Single Loop Operation, 2923 MWt Page 27 This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 1
Minim um M aximuam JAPM TPScam(ME LLLI (IC F) 110.0 Core Cuare Power
- Flow, F low,
100
.7 9 8 0.47 100.0
-99 A7R 190G 7.4 804 STP Rod a9 739 0.4 7 90.0 95 70.49 80.47
- 80.
-93 80.47 104 93 8.43 80.47 03 57.137 00.40
- 0.
.0 82 5.031 91.05 a50.0 8.4 40.03
.4 2.0.004 20.0 8
16 05 10.058 806 4M~b hrU Cou:5 5
34.510 84.4 660.0-63 0
50.0-7.7 15.46
- 23.
308 3..62 539 6.593 770 82 24 M sh oeFo 0
10 20 30 40 50 0
0 2 0
4 0
60 70 80 90 100 110 120 % Core Flow
NGG Nudear Fuels Management & Safety Analysis Figure 16 Design Oakc. No. 2B21-1267 B2C1 8 Care Operating Limits Report StbltOto l
oe/lwMpRevision 0
Stablit Opton II PwerFlowMapPage 28 OPRM Inoperable, Single Loop Operation, 2923 MWt This Figure supports Improved Technical Specification 3.3. 1.1 and the Technical Requirements Manual Specification 3.3 120.0 1100 Core Care 100Power
- Flow, Flow,
.u~
- 80.
-91 04 93 84
.25.4 80.479 82059.31
- 81.
0.405 90
.4.20 80.37
- 50.
79 53.17 81.52 77 0.058 16 81.98 75 49.502 82.13 400 4
5849.0 92.29 7o6
.0 3
4 87. 9 9 9 2. 4 40 7
0' 00112 458.91 92.60 7FZR1 445.95 8 2.75
/YM99 42.94 9.506 85 41.94 83.213 80074 480.9 83.37
- 20.
83 39.98 93.52 5
38 4.97 03.60 74 37.99 93.83 8 3 3 7.
1 9 3. 9 61 354.094 94.29 Naturn A 38-bu 7~O 30 34.1 845 e 59 33,.13 84.80 0.0 58e:ý I--
32.17
- 84. 7 0 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbslhr Core Flow 0
10 20 30 40 50 60 70 80 9
0 1
2 Cr~o 90 100 110 120 % Core Flow
NGG Nudear Fuels Management & Safety Analysis B2C1 8 Core Operating Limits Report Figure 17 Stability Option Ill Power/Flow Map OPRM Operable, FWTR, 2923 MWt Design Caic. No. 2B21-1 267 Revision 0 Page 29 IThis Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 a.
120.0 110.0 100.0 90.0 80.0 70.0 60.0 50.0 40.0 30.0 20.0 10.0 0.0 Minimum M aximurn (M ELLL)
(IC F )
Po.W r
- Fow, Flow, M Ibalt, r lbs/h r I100 70.19 80.47 99 75.04 8.7 99 73.89 900.447 97 7 2. 70 5 90.47 96 1.9 80.447 95 70.49 90.4 7 94 9936 80.4 7 93 69.25 90.4 92 67.1 3 9047 9 1 69.0 3 90.4 7 90 64.9 3 80j. 47 89 63.93 90.4 7 89 62.74 90.4 7 97 6 1.96 90.5 1 96 60.0 90a.60 95 959.0 900.700 64 59a.43 90.7 9 03 5 7.37 90.9 0 9 2 06.31 a91.00 9 1
- 09. 25
.91.2 90 54.2 0 9 1.39f 79 5 3.16 91.5 2 79 512 9 a1.67 7 7 5 1.09 6.93 7 6 500.05 91.99a 7 5 4 9.02 892.13 74 49.00 92.2 9 7 3 46.9 9a 82.4 4 7 2 4 5.96 9 2.60 7 1 44.90 92.'75 70 4 3.94 92.91 89 4 2.:94 93.*09 4 41.94 93.2 1 47 40.95 893.'37 66 39.5 93.52 35 39.97 893.99 64 37.0989 30 83 3 7.0 93.99 62 36.04 94.1 4 61 35.06 94.2 9 9 0 34.10 9a4.405 59 3 3.13 9 4.60 59 32.17 94.7 0 0.0 7.7 15.4 23.1 30.8 38.5 46.2 53.9 61.6 69.3 77.0 84.7 92.4 Mlbslhr Core Flow 0
10 20 30 40 50 60 70 80 90 100 110 120 %Core Flow
NGG Nudlear Fuels Management & Safety Analysis Figure 18 Design Caic. No. 2B21-1267 B2C1 8 Core Operating Limits Report Stability Option IIl Power/Flow Map Raevso 30 OPRMV Inoperable, FWTR, 2923 MWt This Figure supports Improved Technical Specification 3.3.1.1 and the Technical Requirements Manual Specification 3.3 120.0 Minimum M axim um (ME LLI)
(ICF )
110.0-___
Core APRM STP Rod BI1dI M lb z/hr M lb sth r 90.0 000-8o7.4.9 80.47 79 3.83 80.47 88 82.74 60.4 7 880 19 80.581 8 0.80 80 5 58.50 80.70 84 58.43 8
- 0.478 83 51.3 80.90 0..
82 5.31 3 810.05 40.0-91~o A60 so 4m7SCA 89 0 42.84 83.08:
27 40 95 83.37 20.0-78 39 98a 83.52 85 3.8 97 8.88 8R4 5.37 99 8.83 83 6737 01 8.89 82 5638 04 84.1 81 3 50.25 41.29 5 9 533.13 84.80 50.0-79 3
7 47 0.0LL 7.7 15.4 23.1
- 30.
38.5
- 46.
53.92 616 6.770
- 8.
- 2.
oeFo 0
10 20 30 40 50 60 70 80 90 100 110 120 %Core Flow
BSEP 07-0026 0000-0052-0675-SRLR, Revision 0, Supplemental Reload Licensing Report for Brunswick Steam Electric Plant Unit 2 Reload 17 Cycle 18, February 2007
GNFr Global Nuclear Fuel A Joint Venture of GE, Toshiba, & Hitachi 0000-0052-0675-SRLR Revision 0 Class I February 2007 Supplemental Reload Licensing Report for Brunswick Steam Electric Plant Unit 2 Reload 17 Cycle 18
Brunswick Unit 2 0000-0052-0675-SRLR Reload 17 Revision 0 Important Notice Regarding Contents of This Report Please Read Carefully This report was prepared by Global Nuclear Fuel - Americas, LLC (GNF-A) solely for use by Progress Energy Carolinas, Inc. ("Recipient") in support of the operating license for Brunswick Unit 2 (the "Nuclear Plant"). The information contained in this report (the "Information") is believed by GNF-A to be an accurate and true representation of the facts known by, obtained by or provided to GNF-A at the time this report was prepared.
The only undertakings of GNF-A respecting the Information are contained in the contract between Recipient and GNF-A for nuclear fuel and related services for the Nuclear Plant (the "Fuel Contract") and nothing contained in this document shall be construed as amending or modifying the Fuel Contract. The use of the Information for any purpose other than that for which it was intended under the Fuel Contract, is not authorized by GNF-A. In the event of any such unauthorized use, GNF-A neither (a) makes any representation or warranty (either expressed or implied) as to the completeness, accuracy or usefulness of the Informnation or that such unauthorized use may not infringe privately owned rights, nor (b) assumes any responsibility for liability or damage of any kind which may result from such use of such information.
Page 2
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 Acknowledgement The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload Licensing Report, were performed by GNT-A/GEE-NE Nuclear Analysis personnel. The Supplemental Reload Licensing Report was prepared by G. M. Baka. This document has been verified by J. Su.
Page 3
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 The basis for this report is General Electric Standard Application for Reactor Fuel, NEDE-2401 I1-P-A-i15, September 2005; and the U.S. Supplement, NEDE-2401 Il-P-A-1I5-US, September 2005.
- 1.
Plant-unique Items Appendix A: Analysis Conditions Appendix B: Decrease in Core Coolant Temperature Events Appendix C: Operating Flexibility Options Appendix D: TRACG AOO Methodology Appendix E: MELLLA+ Implementation Appendix F: List of Acronyms
- 2.
Reload Fuel Bundles FeTyp Cycle Nme Fuel ype Loaded Nme Irradiated:
GE 14-PI1 DNAB42O-18GZ-l OOT-1 5-T-2572 (GE 14C) 16 14 GE 14-PI1ODNAB4 19-6G7.0/7G6.0/3G2.0- 1 OT-i 50-T-2573 (GE 14C) 16 8
GEI4-P1ODNAB425-3G7.O/14G6.0/1G2.0-IGOT-150-T-2574 (GE14C) 16 5
GEI4-PIODNAB439-12G6.0-lOOT-150-T-2575 (GE14C) 16 38 GE14-PlODNAB413-l6GZ-100T-150-T-2660 (GEl4C) 17 144 GE14-P1ODNAB425-3G7.0/14G6.0/1G2.0-IOOT-1 50-T-2574 (GEI4C) 17 64 GE 14-PI1ODNAB439-12G6.0- 1 OT-I 5G-T-2575 (GE 14C) 17 40 New:
GE14-PlGDNAB439-12G6.G-lGGT-1 50-T-2575 (GE14C) 18 39 GE 14-PI1 DNAB425-1 8GZ-I1 OT-i 5G-T-2 854 (GE 14C) 18 64 GE14-PlGDNAB4G7-16GZ-100T-15G-T-2853 (GE 14C) 18 144 Total:
560 Page 4
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0
- 3.
Reference Core Loading Pattern Core Average Cycle Exposure Exposure Nomnalprvios ed-f-ccleexosue:32737 MWdIMT 18233 MWd/MT Nomnalpreiou en-ofcyce eposre.(29699 MWd/ST)
(16541 MWd/ST)
Minimum previous end-of-cycle exposure 32737 MWdIMT 18233 MWdIMT (for cold shutdown considerations):
(29699 MWd!ST)
(16541 MWd/ST)
Assmedreoadbegnnng-f-ccl exosue:13456 MWdJMT 0 MWd/MT Assmedrelad eginin-ofcyce eposre:(12207 MWdIST)
(0 MWd/ST)
Assumed reload end-of-cycle exposure 32831 MWd/MT 119375 MWd/MT (rated conditions):
(29784 MWd/ST)
(17577 MWdIST)
Reference core loading pattern:
Figure 1
- 4.
Calculated Core Effective Multiplication and Control System Worth - No Voids, 201C Beginning of Cycle, keffective Uncontrolled 1.125 Fully controlled 0.955 Strongest control rod out 0.985 R, Maximum increase in strongest rod out reactivity during the cycle (Ak) 0.00 1 Cycl expsureat hichR ocurs17000 MWdIMT Cycl expsureat hichR ocurs(15422 MWd/ST)
- 5.
Standby Liquid Control System Shutdown Capability Boron(ppm)Shutdown Margin (Ak)
Boro (ppm)
(at 160'C, Xenon Free)
(at20C)Analytical Requirement Achieved 720
>-0.01 1 0.017 Page 5
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0
- 6.
Reload Unique TRACG Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parameters' Operating domain: WIT (HBB)
Exposure range
Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MP (MWt)
(1000 lb/hr)
MP GE14C 1.0 1.43 1.23 1.01 7.450 122.6 1.47 Operating domain: ICF (HBB)
Exposure range
Peaking Factors Fuel JBundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR 1.3 100 (MWt)
(1000 lb/br)
GE14C 1.0 1.37 131.07.148 125.1 1.47 Operating domain: ICF with TBVOOS (HBB)
Exposure range
Peaking Factors Fuel IBundle Bundle Initial Dein Local Radial Axial R-Factor Power Flow MP Design{
1.37(MWt)
(1000 lb/br)
MP GE14C 1.0 1.3 1.39 1.00 7.148 125.1 1.47
'Exposure range designation is defined in Table 7-1. Application condition number is defined in Section 11.
Page 6
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0
- 7.
Selected Margin Improvement Options 2 Recirculation pump trip:
Rod withdrawal limiter:
Thermal power monitor:
Improved scram time:
Measured scram time:
Exposure dependent limits:
Exposure points analyzed:
No No Yes Yes (Option B)
No Yes 2
Table 7-1 Cycle Exposure Range Designation Name Exposure Range 3 BOG toMOC BOCl8 to EOR18-5900 MWd/MT (5352 MWdIST)
MOC to EOC EOR18-5900 MWdIMT (5352 MWd/ST) to EOC18 BOG to FOG BOC18 to EOC18 2 Refer to the GESTAR basis document identified at the beginning of this report for the margin improvement options currently supported therein.
'End of Rated (EOR) is defined as the cycle average exposure corresponding to all rods out, 100% power/100% flow, and normnal feedwater temperature.
Page 7
Brunswick Unit 2 I? ýlnýA 1 7 0000-0052-0675-SRLR Revision 0
- 8.
Operating Flexibility Options 4 The following inform-ation presents the operational domains and flexibility options which are supported by the reload licensing analysis.
Extended Operating Domain (EOD):
Yes EOD type: Maximum Extended Load Line Limit (MELLLA)
Minimum core flow at rated power:
99.0 %
BOD type: Maximum Extended Load Line Limit Plus (MELLLA+)5 Minimum core flow at rated power:
85.0 %
Increased Core Flow:
Yes Flow point analyzed throughout cycle:
104.5 %
Feedwater Temperature Reduction:
Yes (MELLLA)
Feedwater temperature reduction during cycle:
Final feedwater temperature reduction:
ARTS Program:
Single Loop Operation:
Equipment Out of Service:
One Safety/relief valve Out of Service:
ADS Out of Service:
TBVOOS (w/ one SRV OOS) 6 No (MELLLA+)
11l0.3 0F I1l0.3 0F Yes Yes (MELLLA)
No (MELLLA+)
Yes (MELLLA)
No (MELLLA+)
Yes (1 Valve OOS)
Yes (MELLLA)
No (MELLLA+)
Yes 4~ Refer to the GESTAR basis document identified at the beginning of this report for the operating flexibility options currently supported therein.
'MELLLA+ operation is not allowed until approved by the U.S. Nuclear Regulatory Commission. See Appendix E.
At this time it is expected that the NRC review of MELLLA+ will include penalties/restrictions/commitments not addressed in the MELLLA+ evaluations presented in this document.
6 When the Turbine Bypass System is credited, 8 of 10 valves are assumed operable in the analysis.
Page 8
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0
- 9.
Core-wide AOO Analysis Results 7 Methods used: GEMINI (TRACG), GEXL-PLUS Operating domain: ICF (HBB)
Exposure range
Uncorrected ACPRJICPR 8 Event Flux Q/A 9 G1CFg
_________________________(%rated)
(%rated)
G1CFg Load Rejection w/o Bypass 327
-0.165 2
FW Controller Failure 147
-0.072 3
Operating domain: ICF (HBB)
Exposure range
Uncorrected ACPRIICPR Event Flux Q/A 9 G1CFg
_________________________(%rated)
(%rated)
G1CFg Load Rejection w/o Bypass 424
-0.185 4
FW Controller Failure 207
-0.103 5
Operating domain: ICF with TBVOOS (HBB)
Exposure range
Uncorrected ACPRIICPR Event Flux Q/A 9 G1CFg
(%rated) (%rated)
G1CFg FW Controller Failure 451
-0.202 6
7Exposure range designation is defined in Table 7-1. Application condition number is defined in Section 11.
8 Uncorrected ACPR/ICPR is being reported since this is the term used in developing the operating limit for TRACG-based analyses.
9 o available from the TRACG transient output.
Page 9
Brunswick Unit 2 11?.1 - ýA 1 7 0000-0052-0675-SRLR Revision 0
- 10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary The rod withdrawal error (RWE) event in the maximum extended operating domain was originally analyzed in the GE BWR Licensing Report, Maximum Extended Operating Domain Analysis for Brunswick Steam Electric Plant, NEDC-3 1654P, February 1989. The Brunswick Unit 2 Cycle 18 RWE MCPRs for the 108%, 111 %, and 114% RBM system setpoints are not generally bounded by the safety limit adjusted operating limit MCPRs in Tables 10-5(a), 10-5(b), and 10-5(c) of NEDC-3 1654P. However, the RWE MCPR for the 117% RBM system setpoint is bounded by the safety limit adjusted operating limit MCPR in Table 10-5(c) of NEDC-3 1654P. The limiting results are shown in the table below for the 108%, 111%,
114%, and 117% RBM system setpoints. The RBM operability requirements specified in Section 10.5 of NEDC-3 1654P (for RBM inoperable: OLMCPR Ž: 1.45 for power Ž: 90% and OLMCPR Ž! 1.76 for power
< 90% for a SLMCPR of 1. 11) have been evaluated and shown to be sufficient to ensure that the Safety Limit MCPR will not be exceeded in the event of an unblocked RWE event. In addition, the cladding 1%
plastic strain criteria have been met.
RBM Setpoint Cycle 18 All HTSP Results Without RBM Filter ACPR 108.0 0.19 111.0 0.23 114.0 0.27 117.0 0.29 Page 10
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0
- 11. Cycle MCPR Values 10 11 Two ioop operation safety limit:
Single loop operation safety limit:
1.11 1.13
-Stability MCPR Design Basis:
See Section 15 See Section 16 (Initial MCPR)
Non-pressurization events:
Exposure range: BOC to EOC GE14C Control Rod Withdrawal Error (RBM setpoint at 108%)
1.30 Loss of Feedwater Heating 12 1.24 Fuel Loading Error (mislocated)
Not limiting 13 Fuel Loading Error (misoriented) 1.27 Limnitinp_ Pressurization Events OLMCPR Summary Table: 14 Cond.
Exposure Range Option A Option B GE14C GE14C 1
Normal Operation (w/ Equipment-in-Service)
BOG to MOC1.3.5 MOC to EOC 1.57 1.3 2
BOG to EOC 1.61 1.43 10 Exposure range designation is defined in Table 7-1.
"'The Operating Limit MCPRs for Two Loop Operation (TLO) bound the Operating Limit MCPRs for Single Loop Operation (SLO); therefore, the Operating Limit MCPRs need not be changed for SLO.
12 See Appendix B.
13 The mislocated bundle fuel loading error OLMCPR is bounded by the pressurization event OLMCPR.
14 Each application condition (Appi. Cond.) covers the entire range of licensed flow and feedwater temperature unless specified otherwise. The OLMCPR values presented apply to rated power operation based on the two loop operation safety limit MCPR.
Page 1 11
Brunswick Unit 2 0000-0052-0675-SRLR PpcinC I -
Pressurization events: "~
Operating domain: ICF (HBB)
Exposure range BOC to MOC (Application Condition: 1)
Option A Option B GE14C GE14C Load Rejection w/o Bypass 1.53 1.35 FW Controller Failure Not Limiting 16Not Limiting16 Operating domain: ICF (HBB)
Exposure range
- MOC to EOC (Application Condition: 1)I______
Option A Option B GE14C GE14C Load Rejection w/o Bypass 1.57 1.39 FW Controller Failure Not Limiting 16 Not Limiting 16 Operating domain: ICF with TBVOOS (HBB)
Exposure range
- BOC to EOC (Application Condition: 2)
{Option A Option B GE14C GE14C FW Controller Failure j
1.61 1.43 1-2. Overpressurization Analysis Summary EetPSI Pdome Pv Plant Eet(psig)
(psig)
(psig)
Response
MSIV Closure (Flux Scram) - ICF (HBB) 1277 1286 1328 Figure 7 MSIV Closure (Flux Scram) - MELLLA+ (HBB) 1278 1288 1322 Figure 8 15 Application condition numbers shown for each of the following pressurization events represent the application conditions for which this event contributed in the determination of the limiting OLMCPR value.
16 The FW Controller Failure OLMCPR is bounded by the Load Rejection w/o Bypass event OLMCPR.
Page 12
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0
- 13. Loading Error Results Variable water gap misoriented bundle analysis: Yes 17 Misoriented Fuel Bundle ACPR GE 14-PI1ODNAB4O7-16GZ-I1OOT-150-T-2853 (GEl 4C) 0.06 GE 14-PI1ODNAB425-18GZ-I1 OT-150-T-2854 (GE 14C) 0.06 GE14-PI1 DNAB439-12G6.0- 1OOT-1 50-T-2575 (GE 14C) 0.16 GE 14-PI1 DNAB4 13-1 6GZ-lOOT-i 50-T-2660 (GEl 4C) 0.06 GE14-PlODNAB425-3G7.0/14G6.0/1G2.O-100T-150-T-2574 (GE 14C) 0.16
- 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-2401 1-P-A-US.
- 15. Stability Analysis Results 15.1 Introduction The BWROG Regional Mode DIVOM Guideline recommends that a plant specific DIVOM slope be used for Option III OPRM setpoint determination (Reference 1 in Section 15.4). However, since Brunswick Unit 2 will be implementing the Detect and Suppress Solution - Confirmration Density (DSS-CD) solution in the near future, the interim DIVOM approach as a function of Figure of Merit will be applied until DSS-CD is implemented (Reference 2 in Section 15.4).
Should the Option III OPRM system be declared inoperable, the BWROG Interim Corrective Action will constitute the stability licensing basis for Brunswick Unit 2 Cycle 18.
15.2 Stability Option III Brunswick Unit 2 has implemented BWROG Long Term Stability Solution Option III (Oscillation Power Range Monitor-OPRM) as described in Reference 3 in Section 15.4. Plant specific analysis incorporating the Option III hardware is described in Reference 4 in Section 15.4. Reload validation has been performed in accordance with the licensing basis methodology described in Reference 5 in Section 15.4. The stability based MCPR Operating Limit is provided for two conditions as a function of OPRM amplitude setpoint in the following table. The two conditions evaluated are for a postulated oscillation at 45% rated core flow steady state operation (SS) and following a two recirculation pump trip (2PT) from the limiting full power operation state point.
Current power and flow dependent limits provide adequate protection against violation of the Safety Limit MCPR for postulated reactor instability as long as the operating limit is greater than or equal to the specified value for the selected OPRM setpoint.
17 Includes a 0.02 penalty due to variable water gap R-factor uncertainty.
Page 13
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 The stability-based OLMCPR was calculated for Cycle 18. The reload validation calculation demonstrated that reactor stability does not produce the limiting OLMCPR for Cycle 18 as long as the selected OPRM setpoint produces values for OLMCPR(SS) and OLMCPR(2PT) that are less than the corresponding acceptance criteria. The OPRM setpoints for Two Loop Operation (TLO) are conservative relative to Single Loop Operation (SLO) and are therefore bounding.
Table 15.1 OLMCPR Results as a Function of OPRM Setpoint OPRM Setpoint OLMCPR(SS)
OLMCPR(2PT) 1.05 1.224 1.097 1.06 1.248 1.119 1.07 1.273 1.142 1.08 1.300 1.165 1.09 1.327 1.190 1.10 1.356 1.215 1.11 1.384 1.241 1.12 1.414 1.268 1.13 1.445 1.295 1.14 1.477 1.324 1.15 1.511 1.355 Rated Power Acceptance Off-rated OLMCPR OLMCPR as Criteria
@ 45% Flow 18 described in SRLR Section 11 15.3 Interim Corrective Action Stability GE SIL-380 recommendations and the BWROG Interim Corrective Actions in Reference 6 in Section 15.4 have been included in the Brunswick Unit 2 Cycle 18 operating procedures. Regions of restricted operation defined in Attachment I of Reference 7 in Section 15.4 and expanded in Reference 6 in Section 15.4, are applicable to Brunswick Unit 2. The standard ICA stability regions are expanded as appropriate to offer stability protection for Brunswick Unit 2 Cycle 18 in accordance with Reference 8 in Section 15.4.
18 The off-rated OLMCPR is the maximum of the Kp adjusted MCPR or the MCPRf at 45% core flow.
Page 14
Brunswick Unit 2 0000-0052-0675-SRLR Reload 17 Revision 0 15.4 References
- 1. Plant-Specific Regional Mode DlVOMProcedure Guideline, GE-NE-0000-0028-97 14-Ri, June 2, 2005.
- 2. Determination of Figure of Merit for Stability DIVOM Curve Applicability, OGO1-0228-O0l, July 16, 2001.
- 3. BWR Owners' Group Long-Term Stability Solutions Licensing Methodology, NEDO-3 1960-A, November 1995.
- 4. Licensing Basis Hot Bundle Oscillation Magn itude for Brunswick 1 and 2, GE-NE-C51-0025 1-00-0 1, Revision 0, March 200 1.
- 5. Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications, NEDO-32465 -A, August 1996.
- 6. BWR Owners' Group Guideline for Stability Interim Corrective Action, BWROG-94079, June 6, 1994.
- 7. Power Oscillations in Boiling Water Reactors, NRC Bulletin 88-07, Supplement 1, December 30, 1988.
- 8. Review of B WR Owners' Group Guidelines for Stability Interim Corrective Action, BWROG-02072, November 20, 2002.
- 16. Loss-of-Coolant Accident Results 16.1 10CFR5O.46 Licensing Results The ECCS-LOCA analysis is based on the SAFER'GESTR methodology. The licensing results applicable to each fuel type in the new cycle are summarized in the following table.
Table 16.1-1 Licensing Results Licensing Local Moeta-Water Fuel Type Basis PCT Oxidation Reta-action (OF)
Reaction GE14C 1557
< 1.00
< 0.10 The SAFERJGESTR ECCS-LOCA analysis results for the GE 14C fuel type are documented in Reference 1 for GE14C in Section 16.4.
Page 15
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 16.2 10CFR5O.46 Error Evaluation The I1OCFR5O.46 errors applicable to the Licensing Basis PCT are shown in the following table.
Table 16.2-1 Impact on Licensing Basis Peak Cladding Temperature for GE14C 10CFR5O.46 Error Notifications Number Subject PCT Impact
___(OF) 2002-01 Error in core spray injection elevation
+5 2002-02 Error in SAFER initial bulk water level
+10 2002-05 Error in WEVOL calculation of downcomer free volume 0
2003-01 Impact of SAFER level/volume table error on PCT
-5 2003-05 Impact of Postulated Hydrogen-Oxygen Recombination 0
2006-01 Impact of Top Peaked Power Shape for Small Break 0
LOCA Analysis Total PCT Adder (IF)
+10 The GE14C Licensing Basis PCT remains below the IlOCFR5O.46 limit of 2200 '17.
16.3 ECCS-LOCA Operating Limits The ECCS MAPLHGR operating limits have been merged with the thermal-mechanical MAPLHGR operating limits to produce a set of fuel type dependent composite MA-PLHGR limits representing the most restrictive values of both. The most and least limiting values of these composite MAPLHGRs for each of the new fuel bundles in this cycle are shown in the tables below.
Page 16
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 Table 16.3-1 MAPLHGR Limits Bundle Type: GE 14-PI1ODNAB4O7-16GZ-lOOT-i 50-T-2853 Average Planar Exposure MAPLHGR (kW/ft)
(GWd/ST)
(GWd/MT)
Most Limiting Least Limiting 0.00 0.00 9.47 9.91 0.20 0.22 9.57 9.98 1.00 1.10 9.71 10.10 2.00 2.20 9.87 10.25 3.00 3.31 10.02 10.41 4.00 4.41 10.17 10.58 5.00 5.51 10.31 10.74 6.00 6.61 10.45 10.88 7.00 7.72 10.57 11.02 8.00 8.82 10.70 11.17 9.00 9.92 10.82 11.31 10.00 11.02 10.93 11.46 11.00 12.13 11.05 11.60 12.00 13.23 11.06 11.70 13.00 14.33 11.05 11.68 14.00 15.43 11.05 11.66 14.51 16.00 11.05 11.63 15.00 16.53 11.05 11.61 17.00 18.74 11.02 11.48 19.13 21.09 10.91 11.27 20.00 22.05 10.87 11.19 25.00 27.56 10.49 10.65 30.00 33.07 10.08 10.12 35.00 38.58 9.58 9.61 40.00 44.09 9.07 9.10 45.00 49.60 8.54 8.59 50.00 55.12 7.99 8.06 55.00 60.63 6.34 6.49 57.61 63.50 5.03 5.18 57.91 63.84 4.88 5.03 58.17 64.12 4.91 58.19 64.14 4.90 Page 17
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 Table 16.3-2 MAPLHGR Limits Bundle Type: GE 14-PI1ODNAB425-18GZ-lOOT-150-T-2 854 Average Planar Exposure MAPLHGR (kWlft)
(GWd/ST)
(GWdIMT)
Most Limiting Least Limiting 0.00 0.00 8.85 9.20 0.20 0.22 8.91 9.25 1.00 1.10 9.01 9.34 2.00 2.20 9.14 9.46 3.00 3.31 9.27 9.59 4.00 4.41 9.41 9.73 5.00 5.51 9.54 9.86 6.00 6.61 9.68 9.99 7.00 7.72 9.82 10.12 8.00 8.82 9.95 10.25 9.00 9.92 10.09 10.38 10.00 11.02 10.23 10.51 11.00 12.13 10.27 10.60 12.00 13.23 10.29 10.62 13.00 14.33 10.31 10.65 14.00 15.43 10.34 10.70 14.51 16.00 10.35 10.74 15.00 16.53 10.37 10.77 17.00 18.74 10.44 10.88 19.13 21.09 10.47 10.90 20.00 22.05 10.48 10.90 25.00 27.56 10.28 10.52 30.00 33.07 9.84 10.06 35.00 38.58 9.38 9.60 40.00 44.09 8.91 9.14 45.00 49.60 8.41 8.65 50.00 55.12 7.88 8.12 55.00 60.63 5.70 6.47 56.45 62.23 4.96 5.74 56.61 62.40 5.66 57.61 63.50 5.16 57.77 63.68 5.08 58.04 63.98 4.95 Page 18
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 Table 16.3-3 MAPLHGR Limits Bundle Type: GE 14-PI1ODNAB439-12G6.0- 1OOT-150-T-2575 Average Planar Exposure MAPLHGR (kW/ft)
(GWdIST)
(GWdIMT)
Most Limiting Least Limiting 0.00 0.00 9.68 9.82 0.20 0.22 9.72 9.87 1.00 1.10 9.79 9.95 2.00 2.20 9.89 10.06 3.00 3.31 9.99 10.17 4.00 4.41 10.09 10.29 5.00 5.51 10.20 10.42 6.00 6.61 10.31 10.55 7.00 7.72 10.43 10.68 8.00 8.82 10.55 10.82 9.00 9.92 10.67 10.96 10.00 11.02 10.79 11.10 11.00 12.13 10.92 11.23 12.00 13.23 10.93 11.28 13.00 14.33 10.92 11.30 14.00 15.43 10.90 11.30 14.51 16.00 10.89 11.30 15.00 16.53 10.88 11.29 17.00 18.74 10.81 11.19 19.13 21.09 10.66 11.02 20.00 22.05 10.60 10.95 25.00 27.56 10.18 10.51 30.00 33.07 9.76 10.08 35.00 38.58 9.32 9.62 40.00 44.09 8.87 9.17 45.00 49.60 8.37 8.63 50.00 55.12 7.83 8.09 55.00 60.63 5.54 6.37 56.30 62.06 4.88 5.71 57.61 63.50 5.05 57.74 63.65 4.98 57.79 63.70 4.95 Page 19
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 The MAPLHGR operating limits for the remaining fuel bundles are documented in References 2 and 3 for GE14C in Section 16.4.
The single ioop operation multiplier on LHGR and MAPLHGR, and the ECCS analytical initial MCPR values applicable to each fuel type in the new cycle core, are shown in the following table.
Table 16.3-4 Initial MCPR and Single Loop Operation LHGR and MAPLHGR Multiplier Fuel Type Initial MCPR Single Loop Operation LHGR and MAPLHGR Multiplier GE14C 1.275 0.80 16.4 References The ECCS-LOCA analysis base reports applicable to the new cycle core are listed below.
References for GE14C I1. Brunswick Nuclear Plant Unit]I and 2 Extended Power Uprate Task 0407 ECCS-LOCA SAFER/GESTR Project Task Report, GE-NE-A22-0O1 13 0 1, Rev. 0, June 200 1.
- 2. Supplemental ReloadLicensing Report for Brunswick Steam Electric Plant Unit 2 Reload 16 Cycle 17, 0000-0028-0680-SRLR, Rev. 0, January 2005.
- 3. Supplemental Reload Licensing Report for Brunswick Steam Electric Plant Unit 2 Reload 15 Cycle 16, 0000-0005-1282-SRLR, Rev. 0, January 2003.
Page 20
Brunswick Unit 2 Reload 17 OOOO-0052-0675-SRLRR Revision 0 FOE H
nJ~
OFrm 0E1 E 'mE HO B EL ~
J~j
]IJI EH]D [EIOE AII
[F MEE [E JIM H L]
0 92HHH=HW L E E OHTf-HHV B D NN N [d 0l F9 Nl F
EW+I 1J2J I El E J Lý"j MHL F2ý N
LIP 2 N HHHý L
I-ILa]N 1
3l 5l 7 9l 11 13 15 E
17 192 32 2
93 3353 94 3
54 95 Fuel Type A=GE1 4-P 1 ODNAB4 13-1 6GZ-1 I OT-I 50-T-2660 (Cycle 17)
B=GE1 4-PI1ODNAB4 13-1 6GZ-I GOT-I 50-T-2660 (Cycle 17)
C=GE14-PI ODNAB425-3G7.0/14G6.0/I G20- lOOT-I 50-T-2574 (Cycle 17)
D=GEI4-PI ODNAB439-1 2G6.O-1 GOT-I 50-T-2575 (Cycle 17)
E=GE1 4-PI1ODNAB4O7-16GZ-I1 OT-I 50-T-2853 (Cycle 18)
F=GE1 4-PI GDNAB425-1I8GZ-I OT-I 50-T-2854 (Cycle 18)
G=GE14-P1 ODNAB439-1 2G6.O-1 GOT-I 50-T -2575 (Cycle 18)
H=GEI 4-P 1ODNAB4 13-1 6GZ-I1 OT-I 5O-T-266O (Cycle 17)
I=GE 14-PI1 DNAB42G-1I8GZ-I GOT-i 5G-T-2572 (Cycle 16)
J=GE 14-PI1 DNAB4 19-6G7.G/7G6.O/3G2.G-I OT-I 50-T-2573 (Cycle 16)
K=GE14-P1 GDNAB425-3G7.O/14G6.O/1 62.0-lOOT-i 50-T-2574 (Cycle 16)
L=GE14-P1 GDNAB439-I 2G6.- I GOT-I 5G-T-2575 (Cycle 16)
M=GE 14-PI1ODNAB4 13-I 6GZ-I GOT-I 5G-T-2660 (Cycle 17)
N=GE14-P1 GDNAB425-3G7.O/14G6.O/1 62.0-1 GOT-I 50-T-2574 (Cycle 17) 0=GE14-P1 GDNAB439-1I266.0-I GOT-i 50-T-2575 (Cycle 17)
Figure I Reference Core Loading Pattern Page 21
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 250.0 200.0 150.0 375.0 325.0+
275.0f 225.0 1 100.0, 00.0 0.0 2M 175.0 125.0 75.0 Vessel Press Rise (psi)
--a-Total SRV Flow
-*--Bypass Valve Flow
-25.0 I 0.0 1.0 2.0 3.0 Time (sec) 4.0 5,0 6.0 0.0 1.0 2.0 3.0 4.0 0.0 6.0 Time (sec) 0.0 1.0 2.0 3.0 Time (sec) 4.0 5.0 6.0 0.0 1.0 2.0 3.0 Time (sec) 4.0 0.0 6.0 Figure 2 Plant Response to Load Rejection w/o Bypass (MOC ICF (HBB))
Page 22
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 250.0 200.0 150.0 325.0 275.0 i.
225.0 I-
--+-Vessel Press Rise (psi)
--&- Total SRV Flow
-- Bypass Valve Flow 100.0 50.0.
0.0 125.0 1 75.0 25.0 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 10.0 20.0 Time (sec)
Time (sec) 1.5 200.0
--e-Level - Inch above Sep Skirt
-84(-Vessel Steam Flow -Turbine Steam Flow
-4+- Feedwater Flow 175.0 -
1.04 150.0-0.54 125.0 100.04 75.
50.0 25.0 0.0
.25.0 A
A 0.0
-0.5
-1.0 I
-1.5 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 18.0 20.0 Time (sec) 0.0 2.0 0.0 6.0 8.0 10.0 12.0 Time (sec) 14.0 15.0 18.0 20.0 Figure 3 Plant Response to FW Controller Failure (MOC ICF (HBB))
Page 23
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 375.0 325.0 275.0 225.0 125.0 75.0 25.0
-25.0
- Vessel Press Rise (psi)
--S-Total SRV Flow
--x-Bypass Valve Flow 0.0 1.0 2.0 3.0 4.0 Time (sec) 5.0 6.0 3.0 4.0 0.0 6.0 0.0 1.0 2.0 3.0 Time (sec)
Time (sec) 4.0 5.0 6.0 Figure 4 Plant Response to Load Rejection w/o Bypass (EOC ICF (HBB))
Page 24
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 200.0 150.0 100.0 50.0
-,0- Neutron Flux
--x-- Core Inlet Flow
- Core Inlet Subcooling 375.0 325.0 275.0+-
220.0 4 0: +-Vessel Press Rise (psi) Total SRV Flow
-- Bypass Valve Flow 125.0 1.
75.0 1 25.0 0.0.L
-25.
0.0 2-0 4.0 6.0 0.0 10.0 12.0 14.0 16.0 10.0 20.0 Time (sec) 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 10.0 20.0 Time (sec) 1.0, 2W0.01
-- Level - Inch above Sep Skirt
--X-Vessel Steam Flow
-A--Turbine Steam Flow
-- *- Feedwater Flow 170.0 150.0.
125.0 150.0 25.5 0.0
-25.0
-500.
F ------------------
1.0 0.5 0.0
-0.5
-1.0
-1.5 2.0 0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 10.0 20.0 Time (sec) 0.0 2.0 4.0 6.0 0.0 10.0 12.0 14.0 16.0 10.0 20.0 Time (sec)
Figure 5 Plant Response to FW Controller Failure (EOC ICF (HBB))
Page 25
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 375.0 250.0 H*
200.0 -1 100.0 1I 100.0
-0>-Neutron Flux
--- Core Inlet Flow
-- Core Inlet Subcooling 225.0 1 325.0 275.0 +
175.0 5-
--#-Vessel Press Rise (psi)
-a*-Total SRV Flow
--- Bypass Valve Flow 125.0 50.0 1 75.0 25.0
-25.0 0.0 0.0 2.0 4.0 6.0 8.0 10.0 12.0 Time (sec) 14.0 16.0 18.0 20.0 0.0 2.0 4.0 6.0 0.0 10.0 12.0 14.0 10.0 Time (sec) 10.0 2.0 1.5 1.0 0.5 0.0
-0.5 0
-1.0
-1.5 0
00
-2.0 0.0 2.0 4.0 8.0 0.0 10.0 12.0 14.0 16.0 18.0 20.0 Time (sec) 0.0 2.0 4.0 0.0 8.0 10.0 12.0 14.0 16.0 18.0 20.0 Time (sec)
Figure 6 Plant Response to FW Controller Failure (EOC ICF (HBB) with TBVOOS)
Page 26
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 25D.0 200.0 150.0 325.0 275.0 +
225.0f 00 0:
- 4-Vessel Press Rise (psi)
- Total SRV Flow
--X-Bypass Valve Flow 1Dw.0, 50.0 0.0 125.0 1 75.0 20.0 I
X M
0.0 1.0 2.0 3.0 4.0 5.0 6ý0 7.0
&0 9.0 Time (sec) 0 00 1.5 1.0 0.
0.0
.1.0
.1.5
-2.0 0.0 1.0 2.0 2.0 4.0 5.0 6,0 7.0 6.0 Time (sec) 0.0 1.0 2.0 3.0 4.0 5.0 6.0 70O 0.0 9.0 Time (sec)
Figure 7 Plant Response to MSIV Closure (Flux Scram)
Page 27
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 375.0 325.0 275.0 225.0 125.0 75.0 25.0
Vessel Press Rise (psi)
--Total SRV Flow
-a-Bypass Valve Flow 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 0.0 9.0 Time (sec) 00 2
1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 Time (sec)
Time (sec)
Figure 8 Plant Response to MSIV Closure (Flux Scram)
Page 28
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 Appendix A Analysis Conditions The reactor operating conditions and the pressure relief and safety valve configuration used in the reload licensing analysis for this plant and cycle are presented in Tables A-I and A-2 below.
Table A-i Reactor Operating Conditions Analysis Value Parameter NFWT RFWT Thermal power, MWt 2923.0 2923.0 Core flow, Mlb/hr 80.5 80.5 Reactor pressure (core mid-plane), psia 1059.6 1035.9 Inlet enthalpy, B~tu./lb 529.3 512.7 Sta lw l/r12.79 11.11 Doepesue sg1030.3 1007.6 Turbine pressure, psig 963.9 956.6 Table A-2 Pressure Relief and Safety Valve Configuration Number of Lowest Setpoint Valve Type Valves (psig)
Safety/Relief Valve 10 1163.9
'19 Not available in the TRACG output files.
Page 29
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 Appendix B Decrease in Core Coolant Temperature Events The Loss of Feedwater Heating (LFWH) event was analyzed for Brunswick Unit 2 Cycle 18 using the BWR Simulator Code. The use of this code is permitted in GESTAR 11. The transient plots, neutron flux and heat flux values normally reported in Section 9 are not an output of the BWR Simulator Code; therefore, those items are not included in this document. The OLMCPR result is shown in Section 11.
In addition, the Inadvertent HPCI start-up event was shown to be bounded by the LFWH event in Brunswick Unit 2 Cycle 18 in accordance with Reference B-i1.
Reference B-1. Determination ofLimiting Cold Water Event, NEDC-3253 8P-A, February 1996.
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Brunswick Unit 2 0000-0052-0675-SRLR Reload 17 Revision 0 Appendix C Operating Flexibility Options Reference C-I provides a basis for operation of the Brunswick Steam Electric Plant (BSEP) with one Main Steamline Isolation Valve Out of Service (MSIVOOS) (three steamline operation) and all S/RVs in service.
For MSIVOOS, the OLMCPRs presented in Section I11 and peak overpressure results in Section 12 are bounding. MSIVOOS will not be allowed while operating in the MELLLA+ domain (See Appendix E).
Reference C-2 provides a basis for operation of the Brunswick Steam Electric Plant (BSEP) with Feedwater Temperature Reduction (FWTR).
The required OLMCPRs are provided in Section 11.
FWTR and FWHOOS will not be allowed while operating in the MELLLA+ domain (See Appendix E).
Reference C-3 provides a basis for operation of the Brunswick Steam Electric Plant (BSEP) with Maximum Extended Operating Domain (MEOD). The required OLMCPRs are provided in Section 11.
Reference C-4 provides a basis for operation of the Brunswick Steam Electric Plant (BSEP) with all Turbine Bypass Valves Out of Service (TBVOOS). The required OLMCPRs are provided in Section 11.
The impact of GE14 fuel on the operating flexibility options is addressed in Reference C-5.
The ARTS power and flow dependent operating limits for all operating flexibility options are provided in References C-9, C-3, and C-6.
The ARTS power and flow dependent MCPR limits apply to two recirculation pump system operation and SLO without modification for Brunswick Unit 2 Cycle 18.
The ARTS power dependent limits in Reference C-9 include an adjustment to the limits provided in References C-3 and C-5 to address the 10 CFR Part 21 Communication in Reference C-8. The power dependent MCPR limits in Reference C-9 also include an adjustment to the Kp above P-bypass due to the implementation of TRACG (Reference C-7). The MCPRp limits provided in Reference C-9 are based on a Safety Limit of 1. 11. The Reference C-9 power dependent limits are not altered for Brunswick Unit 2 Cycle 18.
The ARTS flow dependent limits provided in Reference C-3 are based on a Safety Limit of 1.07. Due to the cycle-specific safety limit adjustment for Brunswick Unit 2 Cycle 18, there will be a required adjustment to the MCPRf limits and a validation of the required minimum GE14 OLMCPR for the recirculation pump seizure event. The Reference C-3 MCPRf limits are increased for a Safety Limit of 1.11 by the ratio of (1.11/1.07). The following coefficients apply:
Maximum Core Flow A(1)
B(f)
Flow Intercept MCPR
(% of Rated)
(% of Rated) 102.5
-0.592 1.717 80.51 1.24 107.0
-0.608 1.760 85.61 1.24 112.0
-0.625 1.812 91.64 1.24 117.0
-0.656 1.877 97.10 1.24 Page 31
Brunswick Unit 2 0000-0052-0675-SRLR Reload 17 Revision 0 The flow dependent MAPLHGR limit multiplier, MAPFACf, is not altered for Brunswick Unit 2 Cycle 18.
From Reference C-5, the SLO OLMCPR was determined to be 1.40 with a SLO SLMCPR of 1. 12. The initial power for the licensing basis pump seizure event is 2143 MWth, which was about 83.8% of the plant licensed thermal power at the time of the analysis.
Consequently, the Kp for this power level from Reference C-5 was approximately 1.06, resulting in a minimum full power OLMCPR of 1.32 to cover the pump seizure event for GEl14 fuel for a SLO SLMCPR of 1. 12. After EPU, the licensing basis GE 14 pump seizure initial power in MWth is unchanged, but represents about 73.3% of licensed thermal power, resulting in a Kp of about 1. 12. Since the SLO SLMCPR is 1. 13 for Brunswick Unit 2 Cycle 18, the SLO OLMCPR must be greater than or equal to 1.41 for the pump seizure event. Therefore, the minimum GE14 OLMCPR of 1.35 (See Section 11) is conservative for the SLO pump seizure event (l.35*K(73.3)=
1.35*1.12 =1.51 >1.41).
References C-i.
Main Steam line Isolation Valve Out of Service for the Brunswick Steam Electric Plant, EAS-l 117-0987, GE Nuclear Energy (Proprietary), April 1988.
C-2. Feedwater Temperature Reduction with Maximum Extended Load Line Limit and Increased Core Flow for Brunswick Steam Electric Plants Units 1 and 2, NEDC-32457P, Revision 1, GE Nuclear Energy (Proprietary), December 1995.
C-3.
Maximum Extended Operating Domain Analysis for Brunswick Steam Electric Plant, NEDC-3 1654P, GE Nuclear Energy (Proprietary), February 1989.
C-4. Turbine Bypass Out of Service Analysis for Carolina Power & Light Company's Brunswick Nuclear Plants Units I and 2, NEDC-328 13, Revision 3, GE Nuclear Energy (Proprietary), June 1998.
C-S.
GE] 4 Fuel Design Cycle-Independent Analyses For Brunswick Steam Electric Plant Units]I and 2, GE-NE-L 12-00876-00-01 P, Revision 1, GE Nuclear Energy (Proprietary), February 200 1.
C-6.
Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2 Extended Power Uprate, NEDC-33039P, GE Nuclear Energy (Proprietary), August 2001.
C-7. Brunswick Nuclear Station TRA CG Implementation for Reload Licensing Transient Analysis, GE-NE-0000-0022-8 180-RO, GE Nuclear Energy (Proprietary), February 2004.
C-8.
SC04-15, "Turbine Control System Impact in Transient Analyses," 10 CFR Part 21 Communication, October 3 1, 2004.
C-9. Brunswick]I and 2 Off-Rated Analyses Below the PLUPower Level, GE-NE-0000-0036-9469-Rl, Revision 1, GE Nuclear Energy (Proprietary), April 2006.
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Brunswick Unit 2 0000-0052-0675-SRLR Reload 17 Revision 0 Appendix D TRACG AQO Methodology Reference D-1 provides the results of the analyses and evaluations supporting the application of TRACG for AOO analyses for the Brunswick Steam Electric Plant Units 1 and 2. The report requires the increase of Kp from 1. 15 to 1. 18 at 60% power (See Appendix C). Additionally, the report specifies a scram speed adjustment factor (SSAF) of 0. 18 for GE 14 fuel. Section 11I of this report presents the MCPR limits based on the methodology of References D-2 and D-3.
References D-1. Brunswick Nuclear Station TRA CG Implementation for Reload Licensing Transient Analysis, GE-NE-O000-0022-8 180-RO, GE Nuclear Energy (Proprietary), February 2004.
D-2.
TRA CG Application for Anticipated Operational Occurrences Transient Analysis, NEDE-32906P-A, Revision 1, April 2003.
D-3.
TRA CG Application for Anticipated Operational Occurrences Transient Analysis, NEDE-32906P Supplement 2-A, March 2006.
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Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0
.Appendix E MELLLA+ Implementation Brunswick is seeking approval to operate in the MELLLA+ domain (Reference E-1), which would provide greater core flow flexibility, particularly as power approaches 120% of the original licensed thermal power (OLTP).
Approval is not expected prior to startup of Brunswick Unit 2 Cycle 18.
However, the cycle-specific reload licensing analyses were performed to support operation with or without the MELLLA+ domain. Special consideration of MELLLA+ was given during performance of the transient analyses, stability analyses, and LOCA analyses.
The pressurization transients are generally limiting at high flow conditions. However, the transients were performed at both the minimum MELLLA+ flow (85%) and the maximum ICF flow (104.5%). This ensures that the pressurization transient results bound both MELLLA and MELLLA+. Additionally, the loss of feedwater heating (LFWH) transient, which is more limiting at low flow, was performed at the minimum MELLLA+ flow.
Therefore, the limiting subcooling transient bounds both MELLLA and MELLLA+. Reduced feedwater temperature (FFWTR and FWHOOS) and single loop operation (SLO) will not be allowed in conjunction with operation in the MELLLA+ domain. Further, no SRVs will be allowed to be out of service in the MELLLA+ domain.
However, the pressurization transients were performned at MELLLA+ with 1 SRV out of service to ensure that both MELLLA and MELLLA+ are bounded. Finally, TBVOOS will be allowed in the MELLLA+ domain, but MSIVOOS will not.
For EPU/MELLLA+ operation, the Detect and Suppress Solution - Confirmation Density (DSS-CD) is the licensing basis for Brunswick Unit 2 Cycle 18. The reload checklist as outlined in Reference E-2 and shown in Table E-1 is used to confirm that the DSS-CD is applicable to Brunswick Unit 2 Cycle 18 EPU/M\\ELLLA+ operation.
Table E-1 DSS-CD Plant Specific Applicability Checklist Parameter:i Criterion Acpac BWR Product Line BWRl3-6 design as of July 2002 Yes (BWRJ4)
Fuel Product Line GE14 and earlier GE designs Yes (GE 14)
Operating Domain
<_ EPU/MELLLA+ including BWRI3-6 Yes licensed operational flexibility features as of July 2002 Rated TFW Reduction
- _ 1 20OF (EPU/MELLLA)
Yes (11 I00F)
(TFw Reduction not allowed for EPU/MELLLA+ operation))
MCPR Margin OLMCPR Rated - SLMCPR >067Yes (0. 178)
OLMCPR Rated____
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Brunswick Unit 2 0000-0052-0675-SRLR Reload 17 Revision 0 Should the DSS-CD OPRM system be declared inoperable, either the BSP Option 1 or the BSP Option 2 may be used as the backup stability solution for Brunswick Unit 2 Cycle 18 operation in accordance with Plant Technical Specification requirements. The appropriate BSP Option may be implemented based on the inform-ation provided in Reference E-3.
The application of the ECCS-LOCA analysis to operation in the MELLLA+ power/flow region was addressed in Reference E-4. Elimination of the 1 600'F Upper Bound peak cladding temperature limit has been incorporated as defined in Reference E-5.
References E-1. Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2 Maximum Extended Load Line Limit Analysis Plus, NEDC-3 3063P, GE Nuclear Energy (Proprietary), November 2002.
E-2.
General Electric Boiling Water Reactor Detect and Suppress Solution - Confirmation Density Licensing Topical Report, NEDC-3 3075P, Revision 5, November 2005.
E-3. DSS-CD Backup Stability Protection Evaluation for Brunswick Unit 2 Cycle 18, GE-NE-0000-0057-4507-RO, October 2006.
E-4. Brunswick Nuclear Plant Unit]I and 2 MELLLA + Task 040 7 ECCS-L OCA SAFER/GESTR Project Task Report, GE-NE-A22-001 13-83-01, Revision 0, September 2002.
E-5.
GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume INI, Supplement 1, Additional Information for Upper Bound PCT Calculation, NEDE-23785P-A, Volume III, Supplement 1, Revision 1, March 2002.
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Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 Appendix F List of Acronyms Acronym Description ACPR Delta Critical Power Ratio Ak Delta k-effective 2PT Two Recirculation Pump Trip ADS Automatic Depressurization System ADSOOS Automatic Depressurization System Out of Service AOO Anticipated Operational Occurrence APRM Average Power Range Monitor ARTS APRM, Rod Block and Technical Specification Improvement Program BOC Beginning of Cycle BSP Backup Stability Protection Btu British thermal unit BWROG Boiling Water Reactor Owners Group COLR Core Operating Limits Report CPR Critical Power Ratio DIVOM Delta CPR over Initial MCPR vs. Oscillation Magnitude DR Decay Ratio DS/RV Dual Mode Safety/Relief Valve ECCS Emergency Core Cooling System ELLLA Extended Load Line Limit Analysis EOC End of Cycle (including all planned cycle extensions)
EOR End of Rated (All Rods Out I100%Power / IlO0%Flow /NEWT)
ER Exclusion Region FFWTR Final Feedwater Temperature Reduction FMCPR Final MCPR FOM Figure of Merit FWCF Feedwater Controller Failure FWHOOS Feedwater Heaters Out of Service FWTR Feedwater Temperature Reduction GDC General Design Criterion GESTAR General Electric Standard Application for Reactor Fuel GETAB General Electric Thermal Analysis Basis GSF Generic Shape Function HAL Haling Bum HBB Hard Bottom Bum HBOM Hot Bundle Oscillation Magnitude HCOM Hot Channel Oscillation Magnitude HFCL High Flow Control Line HPCI High Pressure Coolant Injection ICA Interim Corrective Action Page 36
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 Acronym Description ICF Increased Core Flow IMCPR Initial MCPR IVM Initial Validation Matrix Kf Off-rated flow dependent OLMCPR multiplier KpOff-rated power dependent OLMCPR multiplier L8 Turbine Trip on high water level (Level 8)
LCF Low Core Flow LHGR Linear Heat Generation Rate LHGRFACf Off-rated flow dependent LHGR multiplier LHGRFA~p Off-rated power dependent LHGR multiplier LOCA Loss of Coolant Accident LPRM Local Power Range Monitor LRHBP Load Rejection with Half Bypass LRNBP Load Rejection without Bypass LTR Licensing Topical Report MAPFACf Off-rated flow dependent MAPLHGR multiplier MAPFACp Off-rated power dependent MAPLHGR multiplier MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MCPRf Off-rated flow dependent OLMCPR MCPRp Off-rated power dependent OLMCPR MELLLA Maximum Extended Load Line Limit Analysis MELLLA+
MELLLA Plus MOC Middle of Cycle MRB Maximal Region Boundaries MSIV Main Steam Isolation Valve MISIVOOS Main Steam Isolation Valve Out of Service MSR Moisture Separator Reheater MSROOS Moisture Separator Reheater Out of Service MTU Metric Ton Uranium MWd Megawatt day MWd/ST Megawatt days per Standard Ton MWd/MT Megawatt days per Metric Ton MWt Megawatt Thermal NBP No Bypass NCL Natural Circulation Line NFWT Normal Feedwater Temperature NOM Nominal Bum NTR Normal Trip Reference OLMCPR Operating Limit MCPR OOS Out of Service OPRM Oscillation Power Range Monitor Pbypass Reactor power level below which the TSV position and the TCV fast closure scrams are bypassed Pdome Peak Dome Pressure Page 37
Brunswick Unit 2 Reload 17 0000-0052-0675-SRLR Revision 0 Acronym Description PSI Peak Steam Line Pressure Pv Peak Vessel Pressure PCT Peak Clad Temperature PHE Peak Hot Excess PLHGR
-Peak Linear Heat Generation Rate PLU Power Load Unbalance PLUOOS Power Load Unbalance Out of Service PRFDS Pressure Regulator Failure Downscale PROOS Pressure Regulator Out of Service Q/A Heat Flux RBM Rod Block Monitor RC Reference Cycle RCF Rated Core Flow RFWT Reduced Feedwater Temperature RPS Reactor Protection System RPT Recirculation Pump Trip
.RPTOOS Recirculation Pump Trip-Out of Service RV Relief Valve RVM Reload Validation Matrix RWE Rod Withdrawal Error SC Standard Cycle SL Safety Limit SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single Loop Operation SRLR Supplemental Reload Licensing Report S/RV Safety/Relief Valve SRVOOS Safety/Relief Valve(s) Out of Service SS Steady State SSV Spring Safety Valve STU Short Tons (or Standard Tons) of Uranium TBV Turbine Bypass Valve TBVOOS Turbine Bypass Valves Out of Service TCV Turbine Control Valve TCVOOS Turbine Control Valve Out of Service TCVSC Turbine Control Valve Slow Closure TLO Two Loop Operation TRY Trip Reference Function TSIP Technical Specifications Improvement Program TSV Turbine Stop Valve TSVOOS Turbine Stop Valve Out of Service TT Turbine Trip TTHBP Turbine Trip with Half Bypass TTNBP Turbine Trip without Bypass UB IUnder Burn Page 38