ML24103A215
| ML24103A215 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 04/05/2024 |
| From: | Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2024-3270 | |
| Download: ML24103A215 (1) | |
Text
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TSBl - TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL
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SSES MANUAL r
Manual Name: TSBl
Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL
Table Of Contents Issue Date: 04/04/2024 Procedure Name Rev Issue Date Change ID Change Number TEXT LOES 134 01/03/2019
Title:
LIST OF EFFECTIVE SECTIONS
TEXT TOC 25 03/05/2019 0:-,
Title:
TABLE OF CONTENTS
TEXT 2.1.1 8 04/04/2024 <0~)
Title:
SAFETY LIMITS (SLS) REACTOR CORE SLS "-,,/> "- ',-./ /
TEXT 2.1.2 1 10/04/2007 "-./ // ~' ;;
Title:
SAFETY LIMITS (SLS) REACTOR COOLANT SYSTEM-(RCS)"/RESSURE S /"*.,, /
f 1---..,,, \\ ',,,// ~ "'- *v,
TEXT 3.0 5 03/18/20'21 \\ ) \\ ', )
Title:
LIMITING CONDITION FOR OPERATION-(LCO) ':A:PPLICABILITY 1~() ~ -__/'
TEXT 3.1.1 3.> 0*4/04/Z0-2'4 /,,,/,.,., /
/ ( '"' '\\
Title:
REACTIVITY CONTROL SYSTEMS SHUTDGWN MARGIN (SDM)
~/ ',, '* V TEXT 3.1.2 ~'\\, _',.,<""- -~> "-'> 11/15/2002
Title:
REACTIVITY co~~~Z:,SMS-REACTIVITY ANOMALIES
TEXT 3.1. 3,,,--,.,_\\CJ) 4 04/04/2024
Title:
REACTIV~TY-CONTR0-1!-,SYSTEMS CONTROL ROD OPERABILITY
TEXT 3. 1. 4 ', \\~JJ _ ___...- 5 11/16/2016
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM TIMES
TEXT 3.1.5 2 11/16/2016
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS
TEXT 3.1. 6 6 04/04/2024
Title:
REACTIVITY CONTROL SYSTEMS ROD PATTERN CONTROL
Page 1 of 8 Report Date: 04/04/24 SSES MANUAL
Manual Name: TSBl
Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 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 4 03/29/2022
Title:
POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
TEXT 3.2.2 6 04/04/2024
Title:
POWER DISTRIBUTION LIMITS MINIMUM CRITICAL POWER RATIO (MCPR)
TEXT 3.2.3 4 03/29/2022
Title:
POWER DISTRIBUTION LIMITS LINEAR HEAT GENERATION RATE (LHGR)
TEXT 3. 3. 1. 1 8 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 7 04/04/2024
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.l 10 11/16/2016
Title:
INSTRUMENTATION POST ACCIDENT MONITORING (PAM) INSTRUMENTATION
TEXT 3.3.3.2 2 11/16/2016
Title:
INSTRUMENTATION REMOTE SHUTDOWN SYSTEM
TEXT 3.3.4.1 4 01/05/2023
Title:
INSTRUMENTATION END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT) INSTRUMENTATION
Page 2 of B Report Date: 04/04/24
,' SSES MANUAL '
Manual Name: TSBl
Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 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 6 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
TEXT 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 6 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 7 05/13/2022
Title:
REACTOR COOLANT SYSTEM (RCS) RECIRCULATION LOOPS OPERATING
TEXT 3.4.2 5 05/13/2022
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: 04/04/24 SSES MANUAL
Manual Name: TSBl
Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL
TEXT 3.4.4 1 11/16/2016
Title:
REACTOR COOLANT SYSTEM (RCS) RCS OPERATIONAL LEAKAGE
TEXT 3.4.5 2 04/13/2016
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
TEXT 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 1 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 5 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: 04/04/24 SSES MANUAL
Manual Name: TSBl
Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL
TEXT 3.6.1.1 6 11/16/2016
Title:
TEXT 3.6.1.2 3 01/05/2023
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCK
TEXT 3.6.1.3 19 01/05/2023
Title:
C<;)NTAINMENT SYSTEMS PRIMARY CONTAINMENT ISOLATION VALVES (PCIVS)
TEXT 3. 6. 1. 4 3 11/27/2023
Title:
CONTAINMENT SYSTEMS CONTAINMENT PRESSURE
TEXT 3.6.1.5 2 11/16/2016
Title:
CONTAINMENT SYSTEMS DRYWELL AIR TEMPERATURE
TEXT 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 TEMPERATURE
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 PRIMARY CONTAINMENT HYDROGEN RECOMBINERS
TEXT 3.6.3.2 4 04/22/2020
Title:
CONTAINMENT SYSTEMS DRYWELL AIR FLOW SYSTEM
Page 5 of 8 Report Date: 04/04/24 SSES MANUAL
Manual Name: TSBl
Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL
TEXT 3.6.3.3 3 09/29/2017
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT OXYGEN CONCENTRATION
TEXT 3.6.4.1 16 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 8 01/05/2023
Title:
PLANT SYSTEMS RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SYSTEM AND THE ULTIMATE HEAT SINK (UHS)
TEXT 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 CONTROL 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 3 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
TEXT 3.7.8 1 11/16/2016
Title:
PLANT SYSTEMS
Page 6 of 8 Report Date: 04/04/24 SSES MANUAL
Manual Name: TSBl
Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL
TEXT 3.8.1 15 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 FUEL 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
TEXT 3.8.6 3 11/27/2023
Title:
ELECTRICAL POWER SYSTEMS BATTERY CELL PARAMETERS
TEXT 3.8.7 4 01/05/2023
Title:
ELECTRICAL POWER SYSTEMS DISTRIBUTION SYSTEMS - OPERATING
TEXT 3.8.8 2 03/05/2019
Title:
ELECTRICAL POWER SYSTEMS DISTRIBUTION SYSTEMS - SHUTDOWN
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-OUT INTERLOCK
TEXT 3.9.3 1 11/16/2016
Title:
REFUELING OPERATIONS CONTROL ROD POSITION
TEXT 3.9.4 0 11/15/2002
Title:
REFUELING OPERATIONS CONTROL ROD POSITION INDICATION
Page 7 of 8 Report Date: 04/04/24 SSES MANUAL
Manual Name: TSBl
Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 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
TEXT 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 03/29/2022
Title:
SPECIAL OPERATIONS CONTROL ROD TESTING - OPERATING
TEXT 3.10.8 4 04/04/2024
Title:
SPECIAL OPERATIONS SHUTDOWN MARGIN (SDM) TEST - REFUELING
Page 8 of 8 Report Date: 04/04/24 Rev. 8 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 limits 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 barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier 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 SLMCPRss/9s, 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 1 2.0-1 Rev. 8 Reactor Core SLs B 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 SLMCPR9519s.
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 for critical power calculations with ATRIUM 11 fuel at pressures~ 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, with the following basis:
Provided that the water level 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 106 lb/hr-ft2. 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 x10 6 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 1 2.0-2 Rev.8 Reactor Core S~s 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 fuel cladding integrity 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 MCPRss1ss.
The SL is based on ATRIUM 11 fuel. For cores with a single fuel product line, the SLMCPRss1ss is the MCPRss1ss for the fuel type. For cores loaded with a mix of applicable fuel types, the SLMCPRss,ss is based on the largest (i.e., most limiting) of the MCPRss1es 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 SLMCPRes1ss.
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 correlations (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 1 2.0-3 Rev.8 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 NRC Bulletin No. 90-02 entitled "Loss of Thermal 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 1 2.0-4 Rev. 8 Reactor Core SLs 8 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.
- 6. ANP-10335P-A, "ACE/ATRIUM 11 Critical Power Correlation," (as identified in the COLR).
- 7. ANP-3857P; "Design Limits for Framatome Critical Power Correlations," Revision 2.
SUSQUEHANNA - UNIT 1 2.0-5 Rev. 8 Reactor Core SLs B 2.1.1
BASES
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SUSQUEHANNA - UNIT 1 2.0-6 Rev.3 SOM B 3.1.1
B 3.1 REACTIVITY CONTROL SYSTEMS
B 3.1.1 SHUTDOWN MARGIN (SOM)
BASES
BACKGROUND SOM requirements are specified to ensure:
- a. The reactor can be made subcritical from all operating conditions and transients and Design Basis Events;
- b. The reactivity transients associated with postulated _accident conditions are controllable within acceptable limits; and
- c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
- These requirements are satisfied by the control rods, as described in GDC 26 (Ref. 1 ), which can compensate for the reactivity effects of the fuel and water temperature changes experienced during all operating conditions.
APPLICABLE The control rod drop accident (CRDA) analysis (Refs. 2 and 3) assumes SAFETY the core is subcritical with the highest worth control rod withdrawn.
ANALYSES Typically, the first control rod withdrawn has a very high reactivity worth and, should the core be critical during the withdrawal of the first control rod, the consequences of a CRDA could exceed the fuel damage limits for a CRDA (see Bases for LCO 3.1.6, "Rod Pattern Control"). Also, SOM is assumed as an initial condition for the control rod removal error during refueling and fuel assembly insertion error during refueling accidents (Ref. 4). The analysis of these reactivity insertion events assumes the refueling interlocks are OPERABLE when the reactor is in the refueling mode of operation. These interlocks prevent the withdrawal of more than one control rod from the core during refueling.
(Special consideration and requirements for multiple control rod withdrawal during refueling are covered in Special Operations LCO 3.10.6, "Multiple Control Rod Withdrawal-Refueling.") The analysis assumes this condition is acceptable since the core will be shut down with the highest worth control rod withdrawn, if adequate SOM has been demonstrated.
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APPLICABLE Prevention or mitigation of reactivity insertion events is necessary to limit SAFETY energy deposition in the fuel to prevent significant fuel damage, which ANALYSES could result in undue release of radioactivity. Adequate SOM ensures
( continued) inadvertent criticalities and potential CRDAs involving high worth control rods (namely the first control rod withdrawn) will not cause significant fuel damage.
SOM satisfies Criterion 2 of the NRC Policy Statement (Ref. 5).
LCO The specified SOM limit accounts for the uncertainty in the demonstration of SOM by testing. Separate SOM limits are provided for testing where the highest worth control rod is determined analytically or by measurement.
This is due to the reduced uncertainty in the SOM test when the highest worth control rod is determined by measurement. When SOM is demonstrated by calculations not associated with a test (e.g., to confirm SOM during the fuel loading sequence), additional margin is included to account for uncertainties in the calculation. To ensure adequate SOM during the design process, a design margin is included to account for uncertainties in the design calculations (Ref. 6).
APPLICABILITY In MODES 1 and 2, SOM must be provided because subcriticality with the highest worth control rod withdrawn is assumed in the CRDA analysis (Ref. 2). In MODES 3 and 4, SOM is required to ensure the reactor will be held subcritical with margin for a single withdrawn control rod. SOM is required in MODE 5 to prevent an open vessel, inadvertent ~riticality during the withdrawal of a single control rod from a core cell containing one or more fuel assemblies or a fuel assembly insertion error (Ref. 4).
ACTIONS
With SOM not within the limits of the LCO in MODE 1 or 2, SOM must be restored within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Failure to meet the specified SOM may be caused by a control rod that cannot be inserted. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is acceptable, considering that the reactor can still be shut down, assuming no failures of additional control rods to insert, and the low probability of an event occurring during this interval.
SUSQUEHANNA - UNIT 1 3.1-2 Rev. 3 SOM 8 3.1.1
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ACTIONS B.1
( continued)
If the SOM cannot be restored, the plant must be brought to MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, to prevent the potential for further reductions in available SOM (e.g., additional stuck control rods). The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
C.1
With SOM not within limits in MODE 3, the operator must immediately initiate action to fully insert all insertable control rods. Action must continue until all insertable control rods are fully inserted. This action results in the least reactive condition for the core.
0.1, D.2. 0.3. and D.4
With SOM not within limits in MODE 4, the operator must immediately initiate action to fully insert all insertable control rods. Action must continue until all insertable control rods are fully inserted. This action results in the least reactive condition for the core. Action must also be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to provide means for control of potential radioactive releases. This includes ensuring secondary containment is OPERABLE; at least one Standby Gas Treatment (SGD subsystem is OPERABLE; and secondary containment isolation capability (i.e., at least one secondary containment isolation valve and associated instrumentation are OPERABLE, or other acceptable administrative controls to assure isolation capability) in each secondary containment penetration flow path not isolated and required to be isolated to mitigate radioactivity releases. This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.
E.1. E.2, E.3, E.4, and E.5
With SOM not within limits in MODE 5, the operator must immediately suspend CORE ALTERATIONS that could reduce SOM (e.g., insertion of fuel in the core or the withdrawal of control rods). Suspension of these activities shall not preclude inserting control rods or removing fuel from the core to reduce the total reactivity.
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BASES
ACTIONS E.1, E.2, E.3, E.4, and E.5 (continued)
(continued)
Action must also be immediately initiated to fully insert all inser:table control rods in core cells containing one or more fuel assemblies. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies have been fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and therefore do not have to be inserted.
Action must also be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to provide means for control of potential radioactive releases. This includes ensuring secondary containment is OPERABLE; at least one SGT subsystem is OPERABLE; and secondary containment isolation capability (i.e., at least one secondary containment isolation valve and associated instrumentation are OPERABLE, or other acceptable administrative controls to assure isolation capability) in each associated penetration flow path not isolated that is assumed to be isolated to mitigate radioactivity releases. This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the Surveillances as needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to restore the component to OPERABLE status. Action must continue until all required components are OPERABLE.
SURVEILLANCE SR 3.1.1.1 REQUIREMENTS SOM must be verified to be within limits to ensure that the reactor can be made subcritical from any initial operating condition. Adequate SOM is demonstrated by testing before or during the first startup after fuel movement, control rod replacement, or shuffling within the reactor pressure vessel. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another core location. Since core reactivity will vary during the cycle as a function of fuel depletion and poison burn up, the beginning of cycle (BOC) test must also account for changes in core reactivity during the cycle. Therefore, to obtain the SOM, the initial measured value must be increased by an adder, "R", which is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated BOC core reactivity. If the value of "R" is zero (that is, BOC is the most reactive point in the cycle), no correction to the BOC measured value is required (Ref. 6). For the SOM
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SURVEILLANCE SR 3.1.1.1 (continued)
REQUIREMENTS (continued) demonstrations that rely solely on calculation of the highest worth control rod, additional margin (0.10% Llk/k) must be added to the SOM limit of 0.28% Llk/k to account for uncertainties in the calculation.
The SOM may be demonstrated during an in sequence control rod withdrawal, in which the highest worth control rod is analytically.
determined, or during local criticals, where the highest worth control rod is determined by analysis or testing.
Local critical tests require the withdrawal of control rods in a sequence that is not in conformance with the analyzed rod position sequence. This testing would therefore require re-programming or bypassing of the rod worth minimizer to allow the withdrawal of control rods not in conformance with the analyzed rod position sequence, and therefore additional requirements must be met (see LCO 3.10. 7, "Control Rod Testing -
Operating").
The Frequency of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification.
During MODE 5, adequate SOM is required to ensure that the reactor does not reach criticality during control rod withdrawals. An evaluation of each planned in-vessel fuel movement during fuel loading (including shuffling fuel within the core) is required to ensure adequate SOM is maintained during refueling. This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern. For example, bounding analyses that demonstrate adequate SOM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence. These bounding analyses include additional margins to the associated uncertainties. Spiral offload/reload sequences inherently satisfy the SR, provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle. Removing fuel from the core will always result in an increase in SOM.
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REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.
- 2. FSAR, Section 15.
- 3. ANP-10333P-A, "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA)," (as identified in the COLR).
- 4. FSAR, Section 15.4.1.1.
- 5. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
- 6. FSAR, Section 4.3.
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SUSQUEHANNA - UNIT 1 3.1-7 Rev.4 Control Rod OPERABILITY B 3.1.3
B 3.1 REACTIVITY CONTROL SYSTEMS
B 3.1.3 Control Rod OPERABILITY
BASES
BACKGROUND Control rods are components of the control rod drive (CRD) System, which is the primary reactivity control system for the reactor. In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes to ensure under conditions of normal operation, including anticipated operational occurrences, that specified acceptable fuel design limits are not exceeded. In addition, the control rods provide the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System. The CRO System is designed to satisfy the requirements of GOC 26, GOC 27, GOC 28, and GDC 29 (Ref. 1).
The CRD System consists of 185 locking piston control rod drive mechanisms (CROMs) and a hydraulic control unit for each drive mechanism. The locking piston type CROM is a double *acting hydraulic piston, which uses condensate water as the operating fluid. Accumulators provide additional energy for scram. An index tube and pisto'n, coupled to the control rod, are locked at fixed increments by a collet mechanism. The collet fingers engage notches in the index tube to prevent unintentional withdrawal of the control rod, but without restricting insertion.
This Specification, along with LCO 3.1.4, "Control Rod Scram Times," and LCO 3.1.5, "Control Rod Scram Accumulators," ensure that the performance of the control rods in the event of a Design Basis Accident (OBA) or transient meets the assumptions used in the safety analyses of References 2, 3, and 4.
APPLICABLE The analytical methods and assumptions used in the evaluations involving SAFETY control rods are presented in References 2, 3, and 4. The control rods ANALYSES provide the primary means for rapid reactivity control (reactor scram), for maintaining the reactor subcritical and for limiting the potential effects of reactivity insertion events caused by malfunctions in the CRD System.
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BASES
APPLICABLE The capability to insert the control rods provides assurance that the SAFETY assumptions for scram reactivity in the OBA and transient analyses are not ANALYSES violated. Since the SOM ensures the reactor will be subcritical with the *
( continued) highest worth control rod withdrawn (assumed single failure), the additional failure of a second control rod to insert, if required, could invalidate the demonstrated SOM and potentially limit the ability of the CRO Systern to
- hold the reactor subcritical. If the control rod is stuck at an inserted position and becomes decoupled from the CRO, a control rod drop accident (CROA) can possibly occur. Therefore, the requirement that all control rods be OPERABLE ensures the CRO System can perform its intended function.
The control rods also protect the fuel from damage which could result in release of radioactivity. The limits protected are the MCPR Safety Limit (SL) (see Bases for SL 2.1.1, "Reactor Core SLs," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," and LCO 3.2.4, "Average Power Range_
Monitor (APRM) Gain and Setpoints"), and the fuel damage limit (see Bases for LCO 3.1.6, "Rod Pattern Control") during reactivity insertion events.
The negative reactivity insertion (scram) provided by the CRO System provides the analytical basis for determination of plant thermal limits and provides protection against fuel damage limits during a CROA. The Bases for LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6 discuss in more detail how the SLs are protected by the CRO System.
Control rod OPERABILITY satisfies Criterion 3 of the NRC Policy Statement (Ref. 5).
LCO The OPERABILITY of an individual control rod is based on a combination of factors, primarily, the scram insertion times, the control rod coupling integrity, and the ability to determine the control rod position. Accumulator OPERABILITY is addressed by LCO 3.1.5. The associated scram accumulator status for a control rod only affects the scram insertion times; therefore, an inoperable accumulator does not immediately require declaring a control rod inoperable. Although not all control rods are required to be OPERABLE to satisfy the intended reactivity control requirements, strict control over the number and distribution of inoperable control rods is required to satisfy the assumptions of the OBA and transient.
analyses.
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BASES
APPLICABILITY In MODES 1 and 2, the control rods are assumed to function during a OBA or transient and are therefore required to be OPERABLE in these MODES.
In MODES 3 and 4, control rods are not able to be withdrawn (except as permitted by LCO 3.10.3 and LCO 3.10.4) since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements for control rod OPERABILITY during these conditions.
Control rod requirements in MODE 5 are located in LCO 3.9.5, "Control Rod OPERABILITY-Refueling."
ACTIONS The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each control rod. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable control rod. Complying with the Required Actions may allow for continued operation, and subsequent inoperable control rods are governed by subsequent Condition entry and application of associated Required Actions.
A.1, A.2, A.3 and A.4
A control rod is considered stuck if it will not insert by either CRD drive water or scram pressure. With a fully inserted control rod stuck, no actions are required as long as the control rod remains fully inserted. The Required Actions are modified by a Note, which allows the rod worth minimizer (RWM) to be bypassed if required to allow continued operation.
LCO 3.3.2.1, "Control Rod Block Instrumentation," provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis. With one withdrawn control rod stuck, the local scram reactivity rate assumptions may not be met if the stuck control rod separation criteria are not met. This separation criteria stipulates that a stuck control rod is equivalent to a "slow" control rod for purposes of separation requirements between "slow" control rods.
Therefore, a verification that the separation criteria are met must be performed immediately. The separation criteria are not met if a) the stuck control rod occupies a position adjacent to two "slow" control rods, b) the stuck control rod occupies a position adjacent to one "slow" control rod and the one "slow" control rod is also adjacent to another "slow" control rod, or, c) if the stuck control rod occupies a location adjacent to one "slow" control rod when there is another pair of "slow" control rods adjacent to one another. Adjacent control rods include control rods that are either face or diagonally adjacent. The description of "slow" control rods is provided in LCO 3.1.4, "Control Rod Scram Times." In addition, the associated control rod drive must be disarmed in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The allowed Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is acceptable, considering the
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ACTIONS A.1, A.2, A.3 and A.4 (continued)
(continued) reactor can still be shut down, assuming no additional control rods fail to insert, and provides a reasonable time to perform the Required Action in an orderly manner. Isolating the control rod from scram prevents damage to the CROM. The control rod can be isolated from scram and normal insert and withdraw pressure, yet still maintain cooling water to the CRD.
Monitoring of the insertion capability of each withdrawn control rod must also be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM. SR 3.1.3.3 performs periodic tests of the control rod insertion capability of withdrawn control rods. Testing each withdrawn control rod ensures that a generic problem does not exist. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." The Required Action A.3 Completion Time only begins upon discovery of Condition A concurrent with THERMAL POWER greater than the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of rod pattern control (LCO 3.1.6) and the RWM (LCO 3.3.2.1 ). The allowed Completion Time provides a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests. To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SOM is also required within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Should a OBA or transient require. a shutdown, to preserve the single failure criterion, an additional control rod would have to be assumed to fail to insert when required.
Therefore, the original SOM demonstration may not be valid. The SOM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn.
The allowed Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to verify SOM is adequate, considering that with a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram. Even with the postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control remains to reach and maintain MODE 3 conditions.
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ACTIONS B.1
( continued)
With two or more withdrawn control rods stuck, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The occurrence of more than one control rod stuck at a withdrawn position increases the probability that the reactor cannot be shut down if required. Insertion of all insertable control rods eliminates the possibility of an additional failure of a control rod to insert.
The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
C.1 and C.2
With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, operation may continue, provided the control rods are fully inserted within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and disarmed (electrically or
- hydraulically within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations.
The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. The control rods can be electrically disarmed by disconnecting power from all four directional control valve solenoids. Required Action C.1 is modified by a Note, which allows the RWM to be bypassed if required to allow insertion of the inoperable control rods and continued operation. LCO 3.3.2.1 provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis.
The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging plant systems.
D.1
If any Required Action and associated Completion Time of Condition A or C are not met, or there are nine or more inoperable control rods, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i.e., scram) of the control rods. The number of control rods permitted to be inoperable when operating above 10% RTP (e.g., no CRDA considerations) could be more than the value specified, but the occurrence of a large number of inoperable control rods could be indicative of a generic problem, and
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ACTIONS D.1 (continued)
(continued) investigation and resolution of the potential problem should be undertaken.
The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.1.3.1 REQUIREMENTS The position of each control rod must be determined to ensure adequate information on control rod position is available to the operator for determining CRD OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
NOT USED
Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal. These Surveillances are not required when THERMAL POWER is less than or equal to the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the analyzed rod position sequence (LCO 3.1.6) and the RWM (LCO 3.3.2.1).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Verifying that the scram time for each control rod to notch position 05 is
~ 7 seconds provides reasonable assurance that the control rod will insert when required during a OBA or transient, thereby completing its shutdown function. This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1, "Reactor Protection System RPS) Instrumentation," and the functional testing of SDV vent and drain valves in LCO 3.1.8, "Scram Discharge Volume (SDV) Vent and Drain Valves," overlap this Surveillance to provide complete testing of the
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SURVEILLANCE SR 3.1.3.4 (continued)
REQUIREMENTS (continued) assumed safety function. The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.
Coupling verification is performed to ensure the control rod is connected to the CROM and will perform its intended function when necessary. The Surveillance requires verifying a control rod does not go to the withdrawn overtravel position. The overtravel position feature provides a positive check on the coupling integrity since only an uncoupled CRD can reach the overtravel position. The verification is required to be performed any time a control rod is withdrawn to the "full out" position (notch position 48) or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This includes control rods inserted one notch and then returned to the "full out" position during the performance of SR 3.1.3.3. This Fre;iquency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved and operating experience related to uncoupling events.
REFERENCES 1. 10 CFR 50, Appendix A GDC 26, GDC 27, GDC 28, and GDC 29.
- 2. FSAR, Section 4.3.2.
- 3. FSAR, Section 4.6.
- 4. FSAR, Section 15.
- 5. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
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SUSQUEHANNA - UNIT 1 3.1-21 Rev. 6 Rod Pattern Control B 3.1.6
B 3.1 REACTIVITY CONTROL SYSTEMS
B 3.1.6 Rod Pattern Control
BASES
BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO 3.3.2.1, "Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10% RTP. The sequences limit the potential amount of reactivity addition that could occur in the event of a Control Rod Drop Accident (CRDA).
This Specification assures that the control rod patterns are consistent with the assumptions of the CRDA analyses of References 1 and 2.
APPLICABLE The analytical methods and assumptions used in evaluating the CRDA are SAFETY summarized in References 1 and 2. CRDA analyses assume that the ANALYSES reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis.
The RWM (LCO 3.3.2.1) provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the CRDA analysis are not violated.
Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage which could result in the undue release of radioactivity. Since the failure consequences for UO2 have been shown to be insignificant below
- fuel energy depositions of 300 cal/gm (Ref. 3), the fuel damage limit of 280 cal/gm provides a margin of safety from significant core damage which would result in release of radioactivity (Refs. 4 and 5). Generic evaluations (Ref. 1 & 6) of a design basis CRDA have shown that the maximum reactor pressure will be less than the required ASME Code limits (Ref. 7). The offsite doses are calculated each cycle using the methodology in reference 1 to demonstrate that the calculated offsite doses will be well within the required limits (Ref. 5). Control rod patterns analyzed in Reference 1 follow the analyzed rod position sequence. The analyzed rod position sequence is applicable from the condition of all control rods fully inserted to 10% RTP (Ref. 2). For the analyzed rod position sequence, the control rods are required to be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. The banked positions are established to minimize the maximum incremental control rod worth without being overly restrictive during normal plant operation. For each reload cycle the CRDA is analyzed to
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BASES
APPLICABLE demonstrate that the 280 cal/gm fuel damage limit will not be violated SAFETY during a CRDA while following the cycle specific analyzed rod position ANALYSES sequence for control rod patterns.
(continued)
These analyses consider the effects of fully inserted inoperable and OPERABLE control rods not withdrawn in the normal sequence of the analyzed rod position sequence. These requirements allow a limited number (i.e., eight) of fully inserted inoperable control rods.
When performing a shutdown of the plant, an optional control rod sequence (Ref. 9) may be used provided that all withdrawn control rods have been confirmed to be coupled prior to reaching THERMAL POWER of :::;10% RTP. The rods may be inserted without the need to stop at intermediate positions since the possibility of a CRDA is eliminated by the confirmation that withdrawn control rods are coupled. When using the Reference 9 control rod sequence for shutdown, the RWM may be reprogrammed to enforce the requirements of the improved control rod insertion, or may be bypassed and the analyzed rod position sequence implemented under LCO 3.3.2.1, Condition D controls.
In order to use the Reference 9 shutdown process, an extra check is required in order to consider a control rod to be "confirmed" to be coupled. This extra check ensures that no Single Operator Error can result in an incorrect coupling check. For purposes of this shutdown process, the method for confirming that control rods are coupled varies depending on the position of the control rod in the core. Details on this coupling confirmation requirement are provided in Reference 9, which requires that any partially inserted control rods, which have not been confirmed to be coupled since their last withdrawal, be fully inserted prior to reaching THERMAL POWER of :::;10% RTP. If a control rod has been checked for coupling at notch 48 and the rod has since only been moved inward, this rod is in contact with it's drive and is not required to be fully inserted prior to reaching THERMAL POWER of s10% RTP. However, if it cannot be confirmed that the control rod has been moved inward, then that rod shall be fully inserted prior to reaching the THERMAL POWER of s10% RTP. This extra check may be performed as an administrative check, by examining logs, previous surveillance's or other information. If,
the requirements for use of the control rod insertion process contained in Reference 9 are followed, the plant is considered to be in compliance with the analyzed rod position sequence as required by LCO 3.1.6.
Rod pattern control satisfies Criterion 3 of the NRC Policy Statement (Ref. 8).
SUSQUEHANNA - UNIT 1 3.1-35 Rev. 6 Rod Pattern Control B 3.1.6
BASES
LCO Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with the analyzed rod position sequence. This LCO only applies to OPERABLE control rods. For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.3, "Control Rod OPERABILITY," consistent with the allowances for inoperable control rods*
in the analyzed rod position sequence.
APPLICABILITY In MODES 1 and 2, when THERMAL POWER is~ 10% RTP, the CRDA is a Design Basis Accident and, therefore, compliance with the assumptions of the safety analysis is required. When THERMAL POWER is
> 10% RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Ref. 2). In MODES 3, 4, and 5, since the reactor is shut down and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SOM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn.
ACTIONS A.1 and A.2
With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, actions may be taken to either correct the control rod pattern or declare the associated control rods inoperable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Noncompliance with the prescribed sequence may be the result of "double notching," drifting from a control rod drive cooling water transient, leaking scram valves, or a power reduction to ~ 10% RTP before establishing the correct control rod pattern. The number of OPERABLE control rods not in compliance with the prescribed sequence is limited to eight, to prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence. When the control rod pattern is not in compliance with the prescribed sequence, all control rod movement should be stopped except for moves needed to correct the rod pattern, or scram if warranted.
Required Action A.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3.3.2.1 requires verification of control rod movement by a qualified member of the technical staff. This ensures that the control rods will be moved to the correct position. A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required Action A.2. OPERABILITY of control rods is determined by compliance with LCO 3.1.3, "Control Rod OPERABILITY," LCO 3.1.4,
SUSQUEHANNA - UNIT 1 3.1-36 Rev. 6 Rod Pattern Control B 3.1.6
BASES
ACTIONS A.1 and A.2 (continued)
(continued)
"Control Rod Scram Times," and LCO 3.1.5, "Control Rod Scram Accumulators." The allowed Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is reasonable, considering the restrictions on the number of allowed out of sequence control rods and the low probability of a CRDA occurring during the time the control rods are out of sequence.
B.1 and B.2
If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has less impact on control rod worth than withdrawals have. Required Action B.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3.3.2.1 requires verification of control rod movement by a qualified member of the technical staff.
When nine or more OPERABLE control rods are not in compliance with the prescribed control rod sequence, the reactor mode switch must be placed in the shutdown position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. With the mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability_of a CRDA occurring with the control rods out of sequence.
SURVEILLANCE SR 3.1.6.1 REQUIREMENTS The control rod pattern is periodically verified to be in compliance with the analyzed rod position sequence to ensure the assumptions of the CRDA analyses are met. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The RWM provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at~ 10% RTP.
SUSQUEHANNA - UNIT 1 3.1-37 Rev. 6 Rod Pattern Control 83.1.6
BASES
REFERENCES 1. ANP-10333P-A, "AURORA-8: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA)," (as identified in the COLR).
- 2. "Modifications to the Requirements for Control Rod Drop Accident Mitigating System," BWR Owners Group, July 1986.
- 3. NUREG-0979, Section 4.2.1.3.2, April 1983.
- 4. NUREG-0800, Section 15.4.9, Revision 2, July 1981.
- 5. 10 CFR 100.11.
- 6. NEDO-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors,"
December 1978.
- 7. ASME, Boiler and Pressure Vessel Code.
- 8. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
- 9. NEDO 33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.
SUSQUEHANNA - UNIT 1 3.1-38 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 ratio 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 (i.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 determined 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 1 O for ATRIUM 1 O 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 determine the largest reduction in critical power ratio (CPR). The types of transients evaluated 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 MCPRee.e%, the required operating limit MCPR is obtained.
MCPRee.e% is determined 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 MCPRee.e% calculation are given in
SUSQUEHANNA - UNIT 1 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 MCPR99.9% 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 (MCPR99.9% value, MCPRf 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 MCPR99.9% limit specified in the COLR.
APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed 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 1 3.2-6 Rev. 6 MCPR B 3.2.2
BASES
ACTIONS A.1
If 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 OBA occurring simultaneously with the MCPR out of specification.
8.1
If 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 SR 3.2.2.1 REQUIREMENTS 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 then 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 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. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
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 "Control 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
SUSQUEHANNA - UNIT 1 3.2-7 Rev. 6 MCPR B 3.2.2
BASES
SURVEILLANCE then the MCPR operating limits corresponding to the Maximum Allowable REQUIREMENTS Average Scram Insertion Time must be implemented. Determining MCPR (continued) operating limits based on interpolation between 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, THERMEX: 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 Nuclear 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 Loss 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).
SUSQUEHANNA - UNIT 1 3.2-8 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 Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," (as identified in the COLR).
- 14. BAW-10247P-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).
SUSQUEHANNA - UNIT 1 3.2-9 Rev. 7 Control Rod Block Instrumentation B 3.3.2.1
B 3.3 INSTRUMENTATION
B 3.3.2.1 Control Rod Block Instrumentation
BASES
BACKGROUND Control rods provide the primary means for control of reactivity changes.
Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events. During low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions, control rod blocks from the Reactor Mode Switch - Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities.
The Nominal Trip Setpoint (NTSP) is a predetermined setting for a protective device chosen to ensure automatic actuation prior to the process variable reaching the Analytical Limit and thus ensuring that the Safety Limit (SL) would not be exceeded. The NTSP accounts for various uncertainties. As such, the NTSP meets the definition of a Limiting Safety System Setting (LSSS) because the protective instrument channel actuates to protect a reactor core or RCS pressure boundary Safety Limit. Rod Block Monitor functions 1 a, 1 b and 1 care LSSSs.
Technical Specifications contain values related to the OPERABILITY of equipment required for safe operation of the facility. OPERABLE is defined in Technical Specifications as "... being capable of performing its specified safety function(s)." For automatic protective devices related to SLs, the required safety function is to ensure that a SL is not exceeded and therefore the NTSP is the LSSS, as defined by 10 CFR 50.36. However, use of the NTSP to define OPERABILITY in Technical Specifications would be an overly restrictive requirement if it were applied as an OPERABILITY limit for the "as-found" value during a Surveillance. This would result in Technical Specification compliance problems, as well as reports and corrective actions required by the rule which are not necessary to ensure safety.
Use of the NTSP to define "as-found" OPERABILITY under the expected circumstances described above would result in actions required by both the rule and Technical Specifications that are not warranted. However, there is also some point beyond which the device would have not been able to perform its function due, for example, to greater than expected drift. This value needs to be specified in the Technical Specifications in order to
SUSQUEHANNA - UNIT 1 3.3-44 Rev. 7 Control Rod Block Instrumentation 8 3.3.2.1
BASES
BACKGROUND define OPERABILITY of the devices and is designated as the Allowable (continued) Value which, is the least conservative value of the as-found setpoint that a channel can have during testing.
The Allowable Value specified in SR 3.3.2.1. 7 is the least conservative value of the as-found setpoint that a channel can have when tested, such that a channel is OPERABLE if the setpoint is found conservative with respect to the Allowable Value during the CHANNEL CALIBRATION.
The purpose of the RSM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude a MCPR Safety Limit violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above the low power range setpoint. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit. The RSM channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn. A simulated thermal power signal from one of the four redundant average power range monitor (APRM) channels supplies a reference signal for one of the RSM channels and a simulated thermal power signal from another of the APRM channels supplies the reference signal to the second RBM channel. This reference signal is used to determine which RSM range setpoint (low, intermediate, or high) is enabled. If the APRM simulated thermal power is indicating less than the low power range setpoint, the RBM is automatically bypassed. The RBM is also automatically bypassed if a peripheral control rod is selected (Ref. 2).
The purpose of the RWM is to control rod patterns during startup, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP. The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA. Prescribed control rod sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses steam flow signals to determine when the reactor power is above the preset power level at which the RWM is automatically bypassed (Ref. 1). The RWM is a single channel system that provides input into RMCS rod block channel 2.
The function of the individual rod sequence steps (banking steps) is to minimize the potential reactivity increase from postulated CRDA at low power levels. However, if the possibility for a control rod to drop can be
SUSQUEHANNA - UNIT 1 3.3-45 Rev. 7 Control Rod Block Instrumentation B 3.3.2.1
BASES
BACKGROUND eliminated, then banking steps at low power levels are not needed to (continued) ensure the applicable event limits can not be exceeded. The rods may be inserted without the need to stop at intermediate positions since the possibility of a CRDA is eliminated by the confirmation that withdrawn control rods are coupled.
To eliminate the possibility of a CRDA, administrative controls require that any partially inserted control rods, which have not been confirmed to be coupled since their last withdrawal, be fully inserted prior to reaching the.
THERMAL POWER of :;;;10% RTP. If a control rod has been checked for coupling at notch 48 and the rod has not been moved inward, this rod is in contact with it's drive and is not required to be fully inserted prior to reaching the THERMAL POWER of :;;;10% RTP. However, if it cannot be confirmed that the control rod has been moved inward, then that rod shal.I be fully inserted prior to reaching the THERMAL POWER of :;;;10% RTP..
The remaining control rods may then be inserted without the need to stop at intermediate positions since the possibility of a CRDA has been eliminated.
With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is required to be in the shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods.
APPLICABLE Allowable Values are specified for each applicable Rod Block Function SAFETY listed in Table 3.3.2.1-1. The NTSPs (actual trip setpoints) are selected to ANALYSES, ensure that the setpoints are conservative with respect to the Allowable LCO, and Value. A channel is inoperable if its actual trip setpoint is non-conservative APPLICABILITY with respect to its required Allowable Value.
NTSPs are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The Analytical Limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the Analytical Limits, corrected for calibration, process, and some of the instrument errors. The NTSPs are then determined, accounting for the remaining channel uncertainties. The trip setpoints derived in this manner provide adequate protection because instrumentation-uncertainties, process effects, calibration tolerances, and drift are accounted for.
SUSQUEHANNA - UNIT 1 3.3-46 Rev. 7 Control Rod Block Instrumentation B 3.3.2.1
BASES
APPLICABLE The specific Applicable Safety Analyses, LCO, and Applicability SAFETY discussions are listed below on a Function by Function basis.
- ANALYSES, LCO, and 1. Rod Block Monitor APPLICABILITY (continued) The RBM is designed to prevent violation of the MCPRSL and the cladding 1 % strain Fuel design limit that may result from a single control rod withdrawal (RWE) event.
The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 14. The fuel thermal performance as a function of RBM Allowable Value is determined from the analysis. The NTSP and Allowable Values are chosen as a function of power level.
NTSP operating limits are established based on the specified Allowable Values.
The RBM function satisfies Criterion 3 of the NRC Policy Statement (Ref. 7).
Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Value for the associated power range, to ensure that no single instrument failure can preclude a rod block for this Function. The actual setpoints are calibrated consistent with applicable setpoint methodology.
Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS.
Nominal trip setpoints are those predetermined values of output at which an action should take place. The trip setpoints are compared to the actual process parameter, the calculated RBM flux (RBM channel signal). When the normalized RBM flux value exceeds the applicable trip setpoint, the RBM provides a trip output.
The analytic limits are derived from the limiting values of the process parameters. Using the GE setpoint methodology, based on ISA RP 67.04, Part II "Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation" setpoint calculation Method 2, the RBM Allowable Values are determined from the analytical limits using the statistical combination of the RBM input signal calibration error, process measurement error, primary element accuracy and instrument accuracy under trip conditions. Accounting for these errors assures that a setpoint found during calibration at the Allowable Value has adequate margin to protect the analytical limit thereby protecting the Safety Limit.
SUSQUEHANNA - UNIT 1 3.3-47 Rev. 7 Control Rod Block Instrumentation B 3.3.2.1
BASES
APPLICABLE 1. Rod Block Monitor (continued)
SAFETY ANALYSES, For the digital RBM, there is a normalization process initiated upon rod LCO, and selection, so that only RBM input signal drift over the interval from the rod APPLICABILITY selection to rod movement needs to be considered in determining the (continued) nominal trip setpoints. The RBM has no drift characteristic with no as-left or as-found tolerances since it only performs digital calculations on the digitized input signals provided by the APRMs.
The RBM Allowable Value demonstrates that the analytical limit would not be exceeded, thereby protecting the safety limit. The Nominal trip setpoirits and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and environment errors are accounted for and appropriately applied for the RBM. There are no margins applied to the RBM nominal trip setpoint calculations which could mask RBM degradation.
The RBM will function when operating greater than or equal to 28% RTP.
Below this power level, the RBM is not required to be OPERABLE.
The RSM-selects one of three different RBM flux trip setpoints to be applied based on the current value of THERMAL POWER. THERMAL POWER is indicated to each RBM channel by a simulated thermal power (STP) reference signal input from an associated reference APRM channel. The OPERABLE range is divided into three "power ranges," a "low power range," an "intermediate power range," and a "high power range." The RBM flux trip setpoint applied within each of these three power ranges is, respectively, the "low trip setpoint," the "intermediate trip setpoint," and the "high trip setpoint" (Allowable Values for which are defined in the COLR).
To determine the current power range, each RBM channel compares its current STP input value to three power setpoints, the "low power setpoint",
(28%), the "intermediate power setpoint" (current value defined in the COLR), and the "high power setpoint" (current value defined in the COLR),
which define, respectively, the lower limit of the low power range, the lower limit of the intermediate power range, and the lower limit of the high power range. The trip setpoint applicable for each power range is more restrictive than the corresponding setpoint for the lower power range(s). When STP is below the low power setpoint, the RBM flux trip.outputs are automatically bypassed but the low trip setpoint continues to be applied to indicate the RBM flux setpoint on the NUMAC RBM displays.
SUSQUEHANNA-UNIT 1 3.3-48 Rev. 7 Control Rod Block Instrumentation B 3.3.2.1
BASES
APPLICABLE 1. Rod Block Monitor (continued)
SAFETY ANALYSES, The calculated setpoints and applicable power ranges are bounding LCO, and values. In the equipment implementation, it is necessary to apply a APPLICABILITY "deadband" to each setpoint. The deadband is applied to the RBM trip (continued) setpoint selection logic and the RBM trip automatic bypass logic such that the setpoint being applied is always equal to or more conservative than the required setpoint. Since the RBM flux trip setpoint applicable to the higher power ranges are more conservative than the corresponding trip setpoints for lower power ranges, the trip setpoint applicable to the higher power range (high power range or intermediate power range) continues to be applied when STP decreases below the lower limit of that range until STP is below the power range setpoint by a value exceeding the deadband.
Similarly, when STP decreases below the low power setpoint, the automatic bypass of RBM flux trip outputs will not be applied until STP decreases below the trip setpoint a value exceeding the deadband.
The RBM channel uses THERMAL POWER, as represented by the STP input value from its reference APRM channel, to automatically enable RBM flux trip outputs (remove the automatic bypass) and to select the RBM flux trip setpoint to be applied. However, the RBM Upscale function is only required to be OPERABLE when the MCPR values are less than the values defined in the COLR, depending on the THERMAL POWER level.
Therefore, even though the RSM Upscale Function is implemented in each RBM channel as a single trip function with a selected trip setpoint, it is characterized in Table 3.3.2.1-1 as three Functions, the Low Power Range
- Upscale Function, the Intermediate Power Range - Upscale Function, and the High Power Range - Upscale Function, to facilitate correct definition of the OPERABILITY requirements for the Functions. Each Function corresponds to one of the RBM power ranges. Due to the deadband effects on the determination of the current power range, the transition between these three Functions will occur at slightly different THERMAL POWER levels for increasing power versus decreasing power.
The RWM enforces the analyzed rod position sequence to ensure that the initial conditions of the CRDA analysis are not violated.
The analytical methods and assumptions used in evaluating the CRDA are
- summarized in Reference 15. The analyzed rod position sequence requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the analyzed rod position sequence are specified in LCO 3.1.6, "Rod Pattern Control."
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APPLICABLE 2. Rod Worth Minimizer (continued)
SAFETY ANALYSES, When performing a shutdown of the plant, an optional control rod LCO, and sequence (Ref. 11) may be used if the coupling of each withdrawn control APPLICABILITY rod has been confirmed. The rods may be inserted without the need to (continued) stop at intermediate positions. When using the Reference 11 control rod insertion sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved control rod insertion process, or may be bypassed and the improved control rod shutdown sequence implemented under the controls in Condition D.
The RWM Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 7).
Since the RWM is designed to act as a backup to operator control of the rod sequences, only one channel of the RWM is available and required to be OPERABLE. Special circumstances provided for in the Required Action of LCO 3.1.3, "Control Rod OPERABILITY," and LCO 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the analyzed rod position sequence. The RWM may be bypassed as required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed.
Compliance with the analyzed rod position sequence, and therefore OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is< 10% RTP. When THERMAL POWER is
> 10% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Ref. 15). In MODES 3 and 4, all control rods are required to be inserted into the core (except as provided in 3.10 "Special Operations");
therefore, a CRDA cannot occur. In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.
- 3. Reactor Mode Switch - Shutdown Position
During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed.
The Reactor Mode Switch - Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.
The Reactor Mode Switch - Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 7).
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BASES
APPLICABLE 3. Reactor Mode Switch-Shutdown Position (continued)
SAFETY ANALYSES Two channels are required to be OPERABLE to ensure that no single LCO, and channel failure, will preclude a rod block when required. There is no APPLICABILITY Allowable Value for this Function since the channels are mechanically
.(continued) actuated based solely on reactor mode switch position.
During shutdown conditions (MODE 3, 4, or 5), no positive reactivity insertion events are analyzed because assumptions are that control rod withdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch is in the shutdown position, the control rod withdrawal block is required to be OPERABLE. During MODE 5 with the reactor mode switch in the refueling position, the refuel position one-rod-out interlock (LCO 3.9.2) provides the required control rod withdrawal blocks.
ACTIONS
With one RBM channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod block function; however, overall reliability is reduced because a single failure in the remaining OPERABLE channel can result in no control rod block capability for the RBM. For this reason, Required Action A.1 requires restoration of the inoperable channel to OPERABLE status. The Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on the low probability of an event occurring coincident with a failure in the,
remaining OPERABLE channel. Alternatively, a Completion Time can be*
determined in accordance with the Risk Informed Completion Time Program. However, because Function 3, Reactor Mode Switch -
Shutdown Position, is only applicable in MODES 3, 4, and 5, the Risk Informed Completion Time Program may not be entered for inoperable channel(s) of Function 3.
8.1
If Required Action A.1 is not met and the associated Completion Time has expired, the inoperable channel must be placed in trip within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. If both*
RBM channels are inoperable, the RBM is not capable of performing its intended function; thus, one channel must also be placed in trip. This initiates a control rod withdrawal block, thereby ensuring that the RBM function is met.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities and is acceptable because it minimizes risk while allowing time for restoration or tripping of inoperable channels.
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ACTIONS C.1, C.2.1.1, C.2.1.2, and C.2.2
( continued)
With the RWM inoperable during a reactor startup, the operator is still capable of enforcing the prescribed control rod sequence. However, the overall reliability is reduced because a single operator error can result in violating the control rod sequence. Therefore, control rod mo'vement must be immediately suspended except by scram. Alternatively, startup may continue if at least 12 control rods have already been withdrawn, or a reactor startup with an inoperable RWM was not performed in the last calendar year, i.e., the last 12 months. Required Actions C.2.1.1 and C.2.1.2 require verification of these conditions by review of plant logs and control room indications. A reactor startup with an inoperable RWM is defined as rod withdrawal during startup when the RWM is required to be OPERABLE. Once Required Action C.2.1.1 or C.2.1.2 is satisfactorily completed, control rod withdrawal may proceed in accordance with the restrictions imposed by Required Action C.2.2. Required Action C.2.2 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other*
qualified member of the technical staff. The RWM may be bypassed under these conditions to allow continued operations. In addition, Required Actions of LCO 3.1.3 and LCO 3.1.6 may require bypassing the RWM, during which time the RWM must be considered inoperable with Condition Centered and its Required Actions taken.*
D.1
With the RWM inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence. Required Action D.1 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff. The RWM may be bypassed under these conditions to allow the reactor shutdown to continue.
E.1 and E.2
With one Reactor Mode Switch - Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod withdrawal block function. However, since the Required Actions are consistent with the normal action of an OPERABLE Reactor Mode Switch - Shutdown Position Function (i.e., maintaining all control rods inserted), there is no distinction between having one or two channels inoperable.
SUSQUEHANNA - UNIT 1 3.3-52 Rev. 7 Control Rod Block Instrumentation.
B 3.3.2.1
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ACTIONS E.1 and E.2 (continued)
(continued)
In both cases (one or both channels inoperable), suspending all control rod withdrawal and initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies will ensure that the core is subcritical with adequate SOM ensured by LCO 3.1.1. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are therefore not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.
SURVEILLANCE As noted at the beginning of the SRs, the SRs for each Control Rod Block REQUIREMENTS instrumentation Function are found in the SRs column of Table 3.3.2.1-1.
The Surveillances are modified by a Note to indicate that when an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains control rod block capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Refs. 9, 12 and 13) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary.
A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. It includes the Reactor Manual Control Multiplexing System input. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs and by verifying proper indication of the selection error of at least one out-of-sequence control rod. As noted in the SRs, SR 3.3.2.1.2 is not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after any control rod is withdrawn in MODE 2. As noted, SR 3.3.2.1.3 is not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after THERMAL POWER is ~ 10% RTP in MODE 1.
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SURVEILLANCE SR 3.3.2.1.2 and SR 3.3.2.1.3 (continued)
REQUIREMENTS (continued) This allows entry into MODE 2 for SR 3.3.2.1.2, and entry into MODE 1 when THERMAL POWER is :s; 10% RTP for SR 3.3.2.1.3, to perform the required Surveillance if the Frequency is not met per SR 3.0.2. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
The RBM setpoints are automatically varied as a function of Simulated Thermal Power. Three control rod block Allowable Values are specified in Table 3.3.2.1-1, each within a specific power range. The power at which the control rod block Allowable Values automatically change are based on the APRM signal's input to each RBM channel. Below the minimum power setpoint, the RBM is automatically bypassed. These control rod block NTSPs must be verified periodically to be less than or
- equal to the specified Allowable Values. If any power range setpoint is non-conservative, then the affected RBM channel is considered inoperable. As noted, neutron detectors are excluded from the Surveillance because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.3 and SR 3.3.1.1.8. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
The RWM is automatically bypassed when power is above a specified value. The power level is determined from steam flow signals. The automatic pypass setpoint must be verified periodically to be not bypassed
- s; 10% RTP. This is performed by a Functional check. If the RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power setpoint channel can be placed in the conservative condition (nonbypass). If placed in the nonbypassed condition, the SR is met and the RWM is not considered inoperable. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch - Shutdown Position Function to ensure that the entire channel will perform the intended function. The CHANNEL FUNCTIONAL TEST for the
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SURVEILLANCE SR 3.3.2.1.6 (continued)
REQUIREMENTS (continued) Reactor Mode Switch - Shutdown Position Function is performed by attempting to withdraw any control rod with the reactor mode switch in the shutdown position and verifying a control rod block occurs.
As noted in the SR, the Surveillance is not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links.
This allows entry into MODES 3 and 4 if the Frequency is not met per SR 3.0.2. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
CHANNEL CALIBRATION is a test that verifies the channel responds to the measured parameter with the necessary range and accuracy.
CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibration consistent with the plant specific setpoint methodology.
As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.2.1.7 for the RBM Functions is modified by two Notes as identified in Table 3.3.2.1-1. The RBM Functions are Functions that are LSSSs for reactor core Safety Limits. The first Note requires evaluation of channel performance for the condition where the as-found setting for the channel setpoint is not the NTSP but is conservative with respect to the Allowable Value. For digital channel components, no as-found tolerance or as-left tolerance can be specified. Evaluation of instrument performance will verify that the instrument will continue to behave in accordance with design-basis assumptions. The purpose of the assessment is to ensure confidence in the instrument performance prior to returning the instrument to service. These channels will also be identified in the Corrective Action Program. Entry into the Corrective Action Program will ensure required review and documentation of the condition for continued OPERABILITY.
The second Note requires that the as-left setting for the instrument be
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SURVEILLANCE SR 3.3.2.1.7 (continued)
REQUIREMENTS (continued) returned to the NTSP. If the as-left instrument setting cannot be returned to th~ NTSP, then the instrument channel shall be declared inoperable.
The second Note also requires that the NTSP and NTSP methodology are to be contained in a document controlled by 1 O CFR 50.59.
The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors are possible.
REFERENCES 1. FSAR, Section 7.7.1.2.8.
- 2. FSAR, Section 7.6.1.a.5.7.
- 3. Deleted.
- 4. Deleted.
- 5. Deleted.
- 6. Deleted.
- 7. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 32193).
- 8. NEDC-30851-P-A, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation," October 1988.
- 9. GENE-770-06-1, "Addendum to Bases for changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation, Technical Specifications," February 1991.
- 10. FSAR, Section 15.4.2.
- 11. NEDO 33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.
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REFERENCES 12. NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power (continued) Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function," October 1995.
- 13. NEDC-3241 OP-A Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function," November 1997.
- 14. 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 Nuclear Company, June 1986.
- 15. ANP-10333P-A, "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA)," (as identified in the COLR).
SUSQUEHANNA - UNIT 1 3.3-54c Rev.4 SOM Test-Refueling B 3.10.8
B 3.10 SPECIAL OPERATIONS
B 3.10.8 SHUTDOWN MARGIN (SOM) Test-Refueling
BASES
BACKGROUND The purpose of this MODE 5 Special Operations LCO is to permit SOM testing to be performed for those plant configurations in which the reactor pressure vessel (RPV) head is either not in place or the head bolts are not fully tensioned.
LCO 3.1.1, "SHUTDOWN MARGIN (SOM)," requires that adequate SOM be demonstrated following fuel movements or control rod replacement within the RPV. The demonstration must be performed prior to or within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after criticality is reached. This SOM test may be performed prior to or during the first startup following the refueling. Performing the SOM test prior to startup requires the test to be performed while in MODE 5, with the vessel head bolts less than fully tensioned (and possibly with the vessel head removed). While in MODE 5, the reactor mode switch is required to be in the shutdown or refuel position, where the applicable control rod blocks ensure that the reactor will not become critical. The SOM test requires the reactor* mode switch to be in the startup/hot standby position, since more than one control rod will be withdrawn for the purpose of demonstrating adequate SOM. This Special Operations LCO provides the appropriate additional controls to allow withdrawing more than one control rod from a core cell containing one or more fuel assemblies when the reactor vessel head bolts are less than fully tensioned.
APPLICABLE Prevention and mitigation of unacceptable reactivity excursions during SAFETY control rod withdrawal, with the reactor mode switch in the startup/hot A°NALYSES standby position while in MODE 5, is provided by the intermediate range monitor (IRM) neutron flux scram (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), and control rod block instrumentation (LCO 3.3.2.1, "Control Rod Block Instrumentation"). The limiting reactivity excursion during startup conditions while in MODE 5 is the control rod drop accident (CRDA).
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APPLICABLE CRDA analyses assume that the reactor operator follows prescribed SAFETY withdrawal sequences. For SOM tests performed within these defined ANALYSES sequences, the analyses of Reference 1 is applicable. However, for (continued) some sequences developed for the SOM testing, the control rod patterns assumed in the safety analyses of Reference 1 may not be met.
Therefore, special CRDA analyses, performed in accordance with an NRC approved methodology, are required to demonstrate the SOM test sequence will not result in unacceptable consequences should a CRDA occur during the testing. For the purpose of this test, the protection provided by the normally required MODE 5 applicable LCOs, in addition to the requirements of this LCO, will maintain normal test operations as well as postulated accidents within the bounds of the appropriate safety analyses (Ref. 1). In addition to the added requirements for the RWM, APRM, and_ control rod coupling, the notch out mode is specified for control rod withdrawals that are not in conformance with the analyzed rod position sequence. Requiring the notch out mode limits withdrawal steps to a single notch, which limits inserted reactivity, and allows adequate monitoring of changes in neutron flux, which may occur during the test.
As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply.
Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.
LCO As described in LCO 3.0.7, compliance with this Special Operations LCO' is optional. SOM tests may be performed while in MODE 2, in accordance with Table 1.1-1, without meeting this Special Operations LCO or its ACTIONS. For SOM tests performed while in MODE 5, additional requirements must be met to ensure that adequate protection against potential reactivity excursions is available. To provide additional scram protection, beyond the normally required IRMs, the APRMs are also required to be OPERABLE (LCO 3.3.1.1, Functions 2.a, 2.d, and 2.e) as though the reactor were in MODE 2. Because multiple control rods will be withdrawn and the reactor will potentially become critical, RPS MODE 2 requirements for Functions 2.a, 2.d, and 2.e of Table 3.3.1.1-1 must be enforced and the approved control rod withdrawal sequence must be enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2), or must be verified by a second licensed operator or other qualified member of the technical staff. The SOM may be demonstrated during an in sequence control rod withdrawal, in which the highest worth control rod is analytically determined, or during local criticals, where the highest worth control rod is determined by analysis or testing.
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LCO Local critical tests require the withdrawal of control rods in a sequence (continued) that is not in conformance with the analyzed rod position sequence. This testing would therefore require bypassing or reprogramming of the rod worth minimizer to allow the withdrawal of rods not in conformance with the analyzed rod position sequence, and therefore additional requirements must be met (see LCO 3.10.7, "Control Rod Testing Operating").
Control rod withdrawals that do not conform to the analyzed rod position sequence specified in LCO 3.1.6, "Rod Pattern Control," (i.e., out of sequence control rod withdrawals) must be made in the individual notched withdrawal mode to minimize the potential reactivity insertion associated with each movement.
Coupling integrity of withdrawn control rods is required to minimize the probability of a CRDA and ensure proper functioning of the withdrawn control rods, if they are required to scram. Because the reactor vessel head may be removed during these tests, no other CORE ALTERATIONS may be in progress. Furthermore, since the control rod, scram function with the RCS at atmospheric pressure relies solely on the CRD accumulator, it is essential that the CRD charging water header remain pressurized. This Special Operations LCO then allows changing the Table 1.1-1 reactor mode switch position requirements to include the startup/hot standby position, such that the SOM tests may be performed while in MODE 5.
APPLICABILITY These SOM test Special Operations requirements are only applicable if the SOM tests performed in accordance with LCO 3.1.1, "SOM" are to be*
performed while in MODE 5 with the reactor vessel head removed or the head bolts not fully tensioned. Additional requirements during these tests to enforce control rod withdrawal sequences and restrict other CORE ALTERATIONS provide protection against potential reactivity excursions.
Operations in all other MODES are unaffected by this LCO.
SUSQUEHANNA - UNIT 1 3.10-35 Rev.4 SOM Test-Refueling B 3.10.8
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ACTIONS
With one or more control rods discovered uncoupled during this Special Operation, a controlled insertion of each uncoupled control rod is required; either to attempt recoupling, or to preclude a control rod drop.
This controlled insertion is preferred since, if the control rod fails to follow the drive as it is withdrawn (i.e., is "stuck" in an inserted position), placing the reactor mode switch in the shutdown position per Required Action B.1 could cause substantial secondary damage. If recoupling is not accomplished, operation may continue, provided the control rods are fully inserted within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and disarmed (electrically or hydraulically) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. Electrically the control rods. can be disarmed by disconnecting power from all four directional control valve solenoids. Required Action A.1 is modified by a Note that allows the RWM to be bypassed if required to allow insertion of the inoperable control rods and continued operation. LCO 3.3.2.1, "Control Rod Block Instrumentation," Actions provide additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis.
The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging
- plant systems.
Condition A is modified by a Note allowing separate Condition entry for each uncoupled control rod. This is acceptable since the Required Actions for this Condition provide appropriate compensatory actions for each uncoupled control rod. Complying with the Required Actions may allow for continued operation. Subsequent uncoupled control rods are governed by subsequent entry into the Condition and application of the Required Actions.
8.1
With one or more of the requirements of this LCO not met for reasons other than an uncoupled control rod, the testing should be immediately stopped by placing the reactor mode switch in the shutdown or refuel position. This results in a condition that is consistent with the requirements for MODE 5 where the provisions of this Special Operations LCO are no longer required.
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SURVEILLANCE SR 3.10.8.1 REQUIREMENTS Performance of the applicable SRs for LCO 3.3.1.1, Functions 2.a and 2.d will ensure that the reactor is operated within the bounds of the safety analysis.
SR 3.10.8.1, SR 3.10.8.2, and SR 3.10.8.3
LCO 3.3.1.1, Functions 2.a, 2.d and 2.e, made applicable in this Special Operations LCO, are required to have applicable Surveillances met to establish that this Special Operations LCO is being met. However, the control rod withdrawal sequences during the SOM tests may be enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2 requirements) or by a second licensed operator or other qualified member of the technical staff.
As noted, either the applicable SRs for the RWM (LCO 3.3.2.1) must be satisfied according to the applicable Frequencies (SR 3.10.8.2), or the proper movement of control rods must be verified (SR 3.10.8.3). This latter verification (i.e., SR 3.10.8.3) must be performed during control rod movement to prevent deviations from the specified sequence. These surveillances provide adequate assurance that the specified test sequence is being followed.
Periodic verification of the administrative controls established by this LCO will ensure that the reactor is operated within the bounds of the safety analysis. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control rod is withdrawn to the "full out" notch position, or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved as well as operating experience related to uncoupling events.
SUSQUEHANNA - UNIT 1 3.10-37 Rev. 4 SOM Test-Refueling B 3.10.8
BASES
SURVEILLANCE SR 3.10.8.6 REQUIREMENTS (continued) CRD charging water header pressure verification is performed to ensure the motive force is available to scram the control rods in the event of a scram signal. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 940 psig is well below the expected pressure of 1100 psig. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCE 1. ANP-10333P-A, "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA)," (as identified in the COLR).
SUSQUEHANNA - UNIT 1 3.10-38