ML23107A129
| ML23107A129 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 03/31/2023 |
| From: | Susquehanna |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 2023-5391 | |
| Download: ML23107A129 (1) | |
Text
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Page 1 of 2
- MANUAL HARD COPY DISTR I BUTION DOCUMENT TRANSMITTAL 2023-5391 USER INFORMATION:
GERLACH*ROSEY M EMPL#:028401 CA#: 0363 Address: NUCSA2 Phone#: 542-3194 TRANSMITTAL INFORMATION:
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SSES MANUAL Manual Name: TSB2
. n u a l Ti tle : TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL Table Of Contents Issue Date: 03/30/2023 Procedure Name Rev Issue Date Change ID Change Number TEXT LOES 138 01/03/2019
Title:
LIST OF EFFECTIVE SECTIONS TEXT TOC 25 03 / 05/2019 Title : TABLE OF CONTENTS TEXT 2.1.l 7 03 / 30/2023
Title:
SAFETY LIMITS (SLS) REACTOR CORE SLS TEXT 2.1.2 1 10 / 04/2007
Title:
SAFETY LIMITS (SLS) REACTOR COOLANT SYSTEM (RCS) PRESSURE SL TEXT 3.0 5 03 / 18/2021
Title:
LIMITING CONDITION FOR OPERATI ON (LCO) APPLI CABILITY TEXT 3 .1.1 2 03 / 31/2021 Title : REACTIVI TY CONTROL SYSTEMS SHUTDOWN MARGIN (SDM)
TEXT 3.1.2 0 11 / 18 / 2002 Title : REACTIVI TY CONTROL SYSTEMS REACTIVITY ANOMALIES TEXT 3.1.3 4 01/15/2023 Title : REACTIVI TY CONTROL SYSTEMS CONTROL ROD OPERABILITY TEXT 3.1.4 5 11/16/2016 Title : REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM TIMES TEXT 3.1.5 2 11/16/2016 Ti tle: REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS TEXT 3.1.6 6 01/15/2023 Title : REACTIVITY CONTROL SYSTEMS ROD PATTERN CONTROL Pagel of 8 Report Date: 03/31/23
SSES MANUAL Manual Name: TSB2 ianual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.1.7 5 01/05/2023
Title:
REACTIVITY CONTROL SYSTEMS STANDBY LIQUID CONTROL (SLC) SYSTEM TEXT 3.1.8 4 11/16/2016
Title:
REACTIVITY CONTROL SYSTEMS SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVES TEXT 3.2.1 6 03/31/2021
Title:
POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
TEXT 3.2 . 2 6 03/30/2023
Title:
POWER DISTRIBUTION LIMITS MINIMUM CRITICAL POWER RATIO (MCPR)
TEXT 3.2.3 4 03/31/2021
Title:
POWER DISTRIBUTION LIMITS LINEAR HEAT GENERATION RATE LHGR EXT 3.3.1.1 7 01/05/2023
Title:
INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS) INSTRUMENTATION TEXT 3.3 . 1.2 4 01/23/2018
Title:
INSTRUMENTATION SOURCE RANGE MONITOR (SRM) INSTRUMENTATION TEXT 3.3.2.1 6 0 1 /15/2023 Title : INSTRUMENTATION CONTROL ROD BLOCK INSTRUMENTATION TEXT 3.3.2.2 4 01/05/2023 Title : INSTRUMENTATION FEEDWATER - MAIN TURBINE HIGH WATER LEVEL TRIP INSTRUMENTATION TEXT 3.3.3.1 9 11/16/2016 Title : INSTRUMENTATION POST ACCIDENT MONITORING (PAM) INSTRUMENTATION TEXT 3.3.3.2 2 11/16/2016 Ti tle : INSTRUMENTATION REMOTE SHUTDOWN SYSTEM TEXT 3.3.4.1 3 01/05/2023
Title:
INSTRUMENTATION END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT) INSTRUMENTATION ,.
Page 2 of 8 Report Date: 03/31/23
SSES MANUAL Manual Name: TSB2
- !anual
Title:
TECElUCAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.3.4.2 2 01/05/2023
Title:
INSTRUMENTATION ANTICIPATED TRANSIENT WITHOUT SCRAM RECIRCULATION PUMP TRIP (ATWS-RPT) INSTRUMENTATION TEXT 3.3.5.1 8 01/05/2023
Title:
INSTRUMENTATION EMERGENCY CORE COOLING SYSTEM (ECCS) INSTRUMENTATION TEXT 3.3.5.2 3 03/18/2021
Title:
REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL INSTRUMENTATION TEXT 3.3.5.3 1 01/05/2023
Title:
UNIT 1 REACTOR PRESSURE VESSEL WIC TS CHANGES TEXT 3.3.6.1 10 01/05/2023
Title:
INSTRUMENTATION PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION
- 'EXT 3. 3. 6. 2 6 03/05/2019
Title:
INSTRUMENTATION SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION TEXT 3.3.7.1 4 03/05/2019
Title:
INSTRUMENTATION CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS) SYSTEM INSTRUMENTATION TEXT 3.3.8.1 7 01/05/2023
Title:
INSTRUMENTATION LOSS OF POWER (LOP) INSTRUMENTATION TEXT 3.3.8.2 1 11/16/2016
Title:
INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS) ELECTRIC POWER MONITORING TEXT 3.4.1 6 03/31/2021
Title:
REACTOR COOLANT SYSTEM (RCS) RECIRCULATION LOOPS OPERATING TEXT 3.4.2 4 11/16/2016
Title:
REACTOR COOLANT SYSTEM (RCS) JET PUMPS TEXT 3.4.3 3 01/13/2012
Title:
REACTOR COOLANT SYSTEM (RCS) SAFETY/RELIEF VALVES (S/RVS)
Page 3 of 8 Report Date: 03/31/23
SSES MANUAL Manual Name: TSB2
~ual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL 1 11/16/2016 TEXT 3.4.4
Title:
REACTOR COOLANT SYSTEM (RCS) RCS OPERATIONAL LEAKAGE TEXT 3.4.5 3 03/10/2010
Title:
REACTOR COOLANT SYSTEM (RCS) RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE TEXT 3.4.6 5 11/16/2016
Title:
REACTOR COOLANT SYSTEM (RCS) RCS LEAKAGE DETECTION INSTRUMENTATION TEXT 3.4.7 3 11/16/2016
Title:
REACTOR COOLANT SYSTEM (RCS) RCS SPECIFIC ACTIVITY TEXT 3.4.8 3 11/16/2016
Title:
REACTOR COOLANT SYSTEM (RCS) RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM
- HOT SHUTDOWN
'EXT 3 .4. 9 2 11/16/2016 *
Title:
REACTOR COOLANT SYSTEM (RCS) RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM
- COLD SHUTDOWN TEXT 3.4.10 6 05/14/2019
Title:
REACTOR COOLANT SYSTEM (RCS) RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TEXT 3.4.11 l 11/16/2016
Title:
REACTOR COOLANT SYSTEM (RCS) REACTOR STEAM DOME PRESSURE TEXT 3.5.1 9 01/05/2023
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) REACTOR PRESSURE VESSEL {RPV) WATER INVENTORY CONTROL AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ECCS OPERATING TEXT 3.5.2 7 06/09/2022
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ECCS OPERATING TEXT 3.5.3 7 01/05/2023
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL AND REACTOR CORE ISOLATION COOLING {RCIC) SYSTEM ECCS OPERATING Page 4 of 8 Report Date: 03/31/23
SSES MANUAL Manual Name: TSB2
.tanual
Title:
TECHNICAL SPECIFICATIONS BASES DNIT 2 MANUAL TEXT 3.6.1.1 6 11/16/2016
Title:
PRIMARY CONTAINMENT TEXT 3.6.1.2 3 01/05/2023
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCK TEXT 3.6.1.3 20 01/05/2023
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT ISOLATION VALVES (PCIVS)
TEXT 3.6.1.4 2 11/16/2016
Title:
CONTAINMENT SYSTEMS CONTAINMENT PRESSURE TEXT 3.6.1.5 2 11/16/2016
Title:
CONTAINMENT SYSTEMS DRYWELL AIR TEMPERATURE
- .'EXT 3. 6 .1. 6 2 01/05/2023
Title:
CONTAINMENT SYSTEMS SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS TEXT 3.6.2.1 3 11/16/2016
Title:
CONTAINMENT SYSTEMS SUPPRESSION POOL AVERAGE TEMPERATORE TEXT 3.6.2.2 2 03/05/2019
Title:
CONTAINMENT SYSTEMS SUPPRESSION POOL WATER LEVEL TEXT 3.6.2.3 3 01/05/2023
Title:
CONTAINMENT SYSTEMS RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING TEXT 3.6.2.4 2 01/05/2023
Title:
CONTAINMENT SYSTEMS RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL SPRAY TEXT 3.6.3.1 2 06/13/2006
Title:
CONTAINMENT SYSTEMS INTENTIONALLY LEFT BLANK TEXT 3.6.3.2 4 08/02/2021
Title:
CONTAINMENT SYSTEMS DRYWELL AIR FLOW SYSTEM Page 5 of 8 Report Date: 03/31/23
SSES MANUAL Manual Name: TSB2 tanual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.6.3.3 3 09/29/2017 Title1 CONTAINMENT SYSTEMS PRIMARY CONTAINMENT OXYGEN CONCENTRATION TEXT 3.6.4.1 17 12/16/2020
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT TEXT 3.6.4.2 14 03/05/2019
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT ISOLATION VALVES (SCIVS)
TEXT 3.6.4.3 7 03/05/2019
Title:
CONTAINMENT SYSTEMS STANDBY GAS TREATMENT (SGT) SYSTEM TEXT 3.7.1 10 01/05/2023
Title:
PLANT SYSTEMS RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SYSTEM AND THE ULTIMATE HEAT SINK (UHS)
'EXT 3. 7. 2 6 01/05/2023
Title:
PLANT SYSTEMS EMERGENCY SERVICE WATER (ESW) SYSTEM TEXT 3.7.3 4 03/05/2019
Title:
PLANT SYSTEMS GONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS) SYSTEM TEXT 3.7.4 2 03/05/2019
Title:
PLANT SYSTEMS CONTROL ROOM FLOOR COOLING SYSTEM TEXT 3.7.5 2 11/16/2016
Title:
PLANT SYSTEMS MAIN CONDENSER OFFGAS TEXT 3.7.6 4 11/16/2016
Title:
PLANT SYSTEMS MAIN TURBINE BYPASS SYSTEM TEXT 3.7.7 2 11/16/2016
Title:
PLANT SYSTEMS SPENT FUEL STORAGE POOL WATER LEVEL 11/16/2016 TEXT 3.7.8 1
Title:
MAINE TORBINE PRESSURE REGULATION SYSTEM Page 6 of 8 Report Date: 03/31/23
SSES MA.NU.AL Manual Name: TSB2
.ianual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.8.1 17 01/05/2023
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES - OPERATING TEXT 3.8.2 2 03/18/2021
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES - SHUTDOWN TEXT 3.8.3 7 08/07/2019
Title:
ELECTRICAL POWER SYSTEMS DIESEL FOEL OIL LUBE OIL AND STARTING AIR TEXT 3.8.4 5 01/05/2023
Title:
ELECTRICAL POWER SYSTEMS DC SOURCES - OPERATING TEXT 3.8.5 2 03/05/2019
Title:
ELECTRICAL POWER SYSTEMS DC SOURCES - SHUTDOWN
.'EXT 3.8.6 2 11/16/2016
Title:
ELECTRICAL POWER SYSTEMS BATTERY CELL PARAMETERS TEXT 3.8.7 9 01/05/2023
Title:
ELECTRICAL POWER SYSTEMS DISTRIBUTION SYSTEMS OPERATING TEXT 3.8.8 2 03/05/2019
Title:
ELECTRICAL POWER SYSTEMS DISTRIBUTION SYSTEMS - SHTITDOWN TEXT 3.9.1 1 11/16/2016
Title:
REFUELING OPERATIONS REFUELING EQUIPMENT INTERLOCKS TEXT 3.9.2 2 11/16/2016
Title:
REFUELING OPERATIONS REFUEL POSITION ONE-ROD-Our INTERLOCK TEXT 3.9.3 1 11/16/2016
Title:
REFUELING OPERATIONS CONTROL ROD POSITION TEXT 3.9.4 0 11/18/2002
Title:
REFUELING OPERATIONS CONTROL ROD POSITION INDICATION Page 7 of 8 Report Date: 03/31/23
SSES MANUAL Manual Name: TSB2
~ual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.9.5 1 11/16/2016
Title:
REFUELING OPERATIONS CONTROL ROD OPERABILITY - REFUELING TEXT 3.9.6 2 11/16/2016
Title:
REFUELING OPERATIONS REACTOR PRESSURE VESSEL (RPV) WATER LEVEL TEXT 3.9.7 1 11/16/2016
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL TEXT 3.9.8 1 11/16/2016
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL TEXT 3.10.1 2 03/05/2019
Title:
SPECIAL OPERATIONS INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION
'EXT 3.10.2 1 11/16/2016
Title:
SPECIAL OPERATIONS REACTOR MODE SWITCH INTERLOCK TESTING TEXT 3.10.3 1 11/16/2016
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD WITHDRAWAL - HOT SHUTDOWN TEXT 3 .10 .4 1 11/16/2016
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD WITHDRAWAL - COLD SHUTDOWN TEXT 3.10.5 1 11/16/2016
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD DRIVE (CRD) REMOVAL - REFUELING TEXT 3.10.6 1 11/16/2016
Title:
SPECIAL OPERATIONS MULTIPLE CONTROL ROD WITHDRAWAL - REFUELING TEXT 3.10.7 2 03/31/2021
Title:
SPECIAL OPERATIONS CONTROL ROD TESTING - OPERATING 03/31/2021 TEXT 3.10.8 4
Title:
SPECIAL OPERATIONS SHUTDOWN MARGIN (SDM) TEST - REFUELING Page 8 of 8 Report Date: 03/31/23
Rev. 7 Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)
B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design 1imits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).
The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2 for ATRIUM 10 and ATRIUM 11 fuel. MCPR greater than the specified limit represents a conservative margin' relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical baniers that separate the radioactive materials from the environs. The integrity of this cladding
- banier is related to its relative freedom from perforations or cracking .
Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.
While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration.
Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e., MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. This is accomplished by having a Safety Limit Minimum Critical Power Ratio (SLMCPR) design basis, referred to as the SLMCPR95195, which corresponds to a 95% probability at a 95%
confidence level (the 95/95 MCPR criterion) that transition boiling will not occur.
Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results SUSQUEHANNA- UNIT 2 2.0-1
Rev. 7 Reactor Core SLs 8 2.1.1 BASES BACKGROUND in oxidation of the fuel cladding to a structurally weaker form. This weaker (continued form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.
APPLICABLE The fuel cladding must not sustain damage as a result of normal operation SAFETY and AOOs. The Technical Specification SL is set generically on a fuel ANALYSES product MCPR correlation basis as the MCPR which corresponds to a 95% probability at a 95% confidence level that transition boiling will not occur, referred to as SLMCPR95195.
The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in combination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure,. and THERMAL POWER level that would result in reaching the MCPR limit.
2.1.1.1 Fuel Cladding Integrity The use of the SPCB (Reference 4) correlation is valid for critical power calculations with ATRIUM 10 fuel at pressures.!: 571.4 psia (conservatively bounded by 575 psig) and bundle mass fluxes > 0.087 x 106 lb/hr-ft2
- The use of the ACE/ATRIUM 11 (Reference 6) correlation is valid fo(
critical power calculation with ATRIUM 11 fuel at pressures ;a:: 588.8 psia (conservatively bounded by 575 psig) with no minimum bundle mass flux.
For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERMAL POWER, wtth the following basis:
Provided that the water leVE?I in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to ensure a minimum bundle flow for all fuel assemblies that have a relatively high power and potentially can approach a critical heat flux condition. For ATRIUM 10 and ATRIUM 11 fuel, the minimum bundle flow is> 28 x 103 lb/hr and the coolant minimum bundle flow and maximum area are such that the mass flux is always> 0.24 x 108 lb/hr-ft.2. Full scale critical power test data taken from various fuel designs at pressures from 14.7 psia to 1400 psia indicate that the fuel assembly critical power at 0.24 x 106 lb/hr-ft2 is approximately 3.35 MWt. At 23% RTP, a bundle power of approximately 3.35 MWt corresponds to a bundle radial peaking factor of approximately 2.8, which is significantly higher than the expected peaking factor. Thus, a THERMAL POWER limit of 23% RTP for reactor pressures< 575 psig is conservative and for conditions of lesser power would remain conservative.
SUSQUEHANNA UNIT 2 2.0-2
Rev. 7 Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2 MCPR SAFETY ANALYSES The fuel cladding integrity SL is set such that no significant fuel damage (continued) is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. The Te,chnical Specification SL value is dependent on the fuel product line and the corresponding MCPR correlation, which is cycle independent. The value is based on the Critical Power Ratio (CPR) data statistics and a 95% probability with 95% confidence that rods are not susceptible to boiling transition, referred to as MCPR95195.
The SL is based on ATRIUM 11 fuel. For cores with a single fuel product line, the SLMCPR95195 is the MCPR95/95 for the fuel type. For cores loaded with a mix of applicable fuel types, the SLMCPR95195 is based on the largest (i.e., most limiting) of the MCPR95195 values for the fuel product lines that are fresh or once-burnt at the start of the cycle. References 4, 6, and 7 described the methodology used in determining the SLMCPR95195.
The SPCB and ACE/ATRIUM 11 critical power correlations are based on a significant body of practical test data. As long as the core pressure and flow are within the range of validity of the correlation (refer to Section B 2.1.1.1 ), the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. These conservatisms and the inherent accuracy of the SPCB and ACE/ATRIUM 11 correlations provide a reasonable degree of assurance that during sustained operation at the MCPR SL there would be no transition boiling in the core. If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not be compromised.
Significant test data accumulated by the NRC and private organizations Indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach. Much of the data indicate that BWR fuel can survive for an extended period of time in an environment of boiling transition .
- SUSQUEHANNA - UNIT 2 2.0-3
Rev. 7 Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2 MCPR (continued)
SAFETY ANALYSES The effects of channel bow on MCPR are explicitly included in the (continued) calculation of the MCPR SL Explicit treatment of channel bow in the MCPR SL addresses the concerns of the NRC Bulletin No. 90-02 entitled "Loss of Thennal Margin Caused by Channel Box Bow."
Monitoring required for compliance with the MCPR SL is specified in LCO 3.2.2, Minimum Critical Power Ratio.
2.1.1.3 Reactor Vessel Water Level During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor.vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height.
- The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.
SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria.
SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.
APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.
SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential for VIOLATIONS radioactive releases in excess of regulatory limits. Therefore, it is required to insert all insertable control rods and restore compliance with the Sls within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal. '
- SUSQUEHANNA UNIT 2 2.0-4
Rev. 7 Reactor Core SLs B 2.1.1 BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.
- 2. Not used.
- 3. Not used.
- 4. EMF-2209(P)(A), "SPCB Critical Power Correlation," (as identified in the COLR).
- 5. Not used.
0
- 6. ANP-10335P-A, "ACE/ATRIUM 11 Critical Power Correlation, (as identified in the COLR).
- 7. ANP-3857P, HDesign Limits for Framatome Crjtical Power Correlations," Revision 2 .
- SUSQUEHANNA- UNIT 2 2.0-5
Rev. 6 MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)
BASES BACKGROUND MCPR is a ratlo of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs), and that 99.9% of the fuel rods are not susceptible to boiling transition if the limit is not violated.
Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref. 1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.
The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power 0.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and detemiined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.
APPLICABLE The analytical methods and assumptions used in evaluating the AOOs to SAFETY establish the operating limit MCPR are presented in References 2, 3, 5, 7, ANALYSES and 10 for ATRIUM 10 fuel design analysis and References 2, 3, 5, 10, and 12 through 15 for ATRIUM 11 fuel designs. To ensure that the MCPR Safety Limit (SL) is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to detem,ine the largest reduction in critical power ratio (CPR). The types of transients ~valuated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (LlCPR). When the largest LlCPR is combined with the MCPRgg_s%, the required operating limit MCPR is obtained.
MCPRoo 9% is detem,ined to ensure more than 99.9% of the fuel rods in - .
the core are not susceptible to boiling transition using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved Critical Power correlations. Details of the MCPRoo 9% calculation are given in SUSQUEHANNA - UNIT 2 3.2-5
Rev. 6 MCPR B 3.2.2 BASES APPLICABLE References 7, 15, and 16. References 7 and 15 also include a tabulation SAFETY of the uncertainties and the nominal values of the parameters used in the ANALYSES MCPR99 9% statistical analysis.
(continued)
The MCPR operating limits are derived from the MCPRgg 9% value and the transient analysis, and are dependent on the operating core flow and power state to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency. These analyses may also consider other combinations of plant conditions (i.e., control rod scram speed, bypass valve performance, EOC-RPT, cycle exposure, etc.). Flow dependent MCPR limits are determined by analysis of slow flow runout transients.
The MCPR satisfies Criterion 2 of the NRC Policy Statement (Ref. 11 ).
LCO The MCPR operating limits specified in the COLR (MCPRgg 9% value, MCPRt values and MCPRp values) are the result of the Design Basis Accident (OBA) and transient analysis. The operating limit MCPR is determined by the larger of the flow dependent MCPR and power dependent MCPR limits, which are based on the MCPRgg 9% limit specified in the COLR.
APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assum~ to occur at high power levels. Below 23% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 23% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.
Studies of the variation of limiting transient behavior have been _performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 23% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs.
When in MODE 2, the intermediate range monitor provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels< 23% RTP, the reactor is operating with substantial margin to the MCPR .limits and this LCO is not required .
- SUSQUEHANNA - UNIT 2 3.2-6
Rev. 6 MCPR B 3.2.2 BASES ACTIONS tf any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met.
Therefore, prompt action should be taken to restore the MCPR(s) to within
.the required limits such that the plant remains operating within analyzed conditions. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is normally sufficient to restore the MCPR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.
B.1 tf the MCPR cannot be restored to within its required limits Within the associated Completion Time, the plant must be brought to.a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 23% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 23% RTP in an orderly manner and without challenging plant systems .
- SURVEILLANCE REQUIREMENTS SR 3.2.2.1 The MCPR is required to be initially calculated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is ;::: 23% RTP and periodically thereafter. Additionally, MCPR must be calculated prior to exceeding 44% RTP unless performed in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. MCPR is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance after THERMAL POWER;::: 23% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels and because the MCPR must be calculated prior to exceeding 44% RTP.
SR 3.2.2.2 Because the transient analysis takes credit for conservatism in the scram time performance, it must be demonstrated that the specific scram time is consistent with those used in the transient analysis. SR 3.2.2.2 compares the average measured scram times to the assumed scram times documented in the COLR. The COLR contains a table of scram times based on the LCO 3.1.4, UControl Rod Scram Times" and the realistic scram times, both of which are used in the transient analysis. If the average measured scram times are greater than the realistic scram times then the MCPR operating limits corresponding to the Maximum Allowable Average Scram Insertion Time must be implemented. Determining MCPR SUSQUEHANNA - UNIT 2 3.2-7
Rev.6 MCPR B 3.2.2 BASES SURVEILU\NCE SR 3.2.2.2 (continued)
REQUIREMENTS (continued) operating limits based on interpolation betw_een scram insertion times is not permitted. The average measured scram times and corresponding MCPR operating limit must be determined once within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each set of scram time tests required by SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3 and SR 3.1.4.4 because the effective scram times may change during the cycle. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is acceptable due to the relatively minor changes in average measured scram times expected during the fuel cycle.
REFERENCES 1. NUREG-0562, June 1979.
- 2. XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors," Exxon Nuclear Company, March 1983.
- 3. XN-NF-80-19(P)(A) Volume 3, Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERM EX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company, January 1987.
- 4. Not used.
- 5. XN-NF-80-19 (P)(A), Volume 4, Revision 1, *Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nudear Company, June 1986.
- 6. Not used.
- 7. EMF-2209(P)(A), "SPCB Critical Power Correlation," (as identified in the COLR).
- 8. Not used.
- 9. Not used.
- 10. ANF-1358(P)(A), "The L.:oss of Feedwater Heating Transient in Boiling Water Reactors," (as identified in the COLR).
- 11. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132) .
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Rev.6 MCPR B 3.2.2 BASES REFERENCES 12. ANP-10300P-A, "AURORA-B: An Evaluation Model for Boiling (continued)* Water Reactors; Application to Transient and Accident Scenarios,"
(as identified in the COLR). *
- 13. BAW-10247PA, "Realistic Themial-Mechanical Fuel Rod Methodology for Boiling Water Reactors," (as identified in the COLR).
- 14. BAW-10247P-A Supplement 2P-A, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors Supplement 2:
Mechanical Methods," (as identified in the COLR).
- 15. ANP-10335P-A, "ACE/ATRIUM-11 Critical Power Correlation,"(as identified in the COLR).
- 16. ANP-10307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," (as identified in the COLR) .
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