ML22101A028
| ML22101A028 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 03/30/2022 |
| From: | Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2022-4354 | |
| Download: ML22101A028 (77) | |
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of 3
SSES MANUAL Manual Name:
TSBl Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL Table Of Contents Issue Date:
03/29/2022 Procedure Name Rev Issue Date Chang_e ID Chang_e Number TEXT LOES 134 01/03/2019
Title:
LIST OF EFFECTIVE SECTIONS TEXT TOC.
Title:
TABLE OF CONTENTS 25 03/05/2019 f.;,
TEXT 2.1.1 7
03/29/2022 Title : SAFETY LIMITS (SLS) REACTOR CORE SLS TEXT 2.1.2 1
10/04/200 7
Title:
SAFETY LIMITS (SLS) REACTOR COOLANT SYSTEM (RCS) PRESSURES TEXT 3. 0 5
03/18/2021
Title:
LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY TEXT 3.1. 1 2
03/29/2022
Title:
REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN (SDM)
TEXT 3.1.2 0
11/15/2002
Title:
REACTIVITY CONTROL SYSTEMS REACTIVITY ANOMALIES
...--...\\ "-J' 3
11/16/2016 TEXT 3.1. 3
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD OPERABILITY
/
TEXT 3.1.4 5
11/16/20 1 6
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 5
03/29/2022
Title:
REACTIVITY CONTROL SYSTEMS ROD PATTERN CONTROL Page 1 of 8
Report Date: 03/29/22
SSES MANUAL Manual Name:
TSBl Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.1.7 4
11/16/2016
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 5
03/29/2022
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 7
11/16/2016
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 5
11/16/2016
Title:
INSTRUMENTATION CONTROL ROD BLOCK INSTRUMENTATION TEXT 3.3.2. 2 3
11/16/2016
Title:
INSTRUMENTATION FEEDWATER MAIN TURBINE HIGH WATER LEVEL TRIP INSTRUMENTATION TEXT 3.3.3.1 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 3
11/16/2016
Title:
INSTRUMENTATION END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT) INSTRUMENTATION
- Page 2 of 8
Report Date: 03/29/22
SSES MANUAL Manual Name:
TSBl Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.3.4.2 1
11/16/2016
Title:
INSTRUMENTATION ANTICIPATED TRANSIENT WITHOUT SCRAM RECIRCULATION PUMP TRIP (ATWS-RPT) INSTRUMENTATION.
TEXT 3.3.5.1 5
03/05/2019
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 0
03/05/2019
Title:
INSTRUMENTATION REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM INSTRUMENTATION
- PREVIOUSLY TEXT 3.3.5.2 REV 1**
TEXT 3;3.6.1 9
03/05/2019
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 4
03/18/2021
Title:
INSTRUMENTATION LOSS OF POWER (LOP) INSTRUMENTATION TEXT 3.3.8.2 1
11/16/2016
Title:
INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS) ELECTRIC POWER MONITORING TEXT 3.4.1 6
03/29/2022
Title:
REACTOR COOLANT SYSTEM (RCS) RECIRCULATION LOOPS OPERATING TEXT 3.4.2 4
11/16/2016
Title:
REACTOR COOLANT SYSTEM (RCS) JET PUMPS TEXT 3.4.3 3
01/13/2012
Title:
REACTOR COOLANT SYSTEM RCS SAFETY RELIEF VALVES S/RVS Page 3 of 8
Report Date: 03/29/22
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 05/14/20i9
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 7
03/29/2022
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ECCS OPERATING TEXT 3.5.2 4
03/18/2021
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 6
03/05/2019 Page 4
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ECCS OPERATING of 8
Report Date: 03/29/22
SSES MANUAL Manual Name:
TSBl Manual Title*: TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.6.1.1 6
11/16/2016
Title:
PRIMARY CONTAINMENT TEXT 3.6.1:2 2
11/16/2016
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCK TEXT 3.6.1.3 17 03/18/2021
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT ISOLATION VALVES (PCIVS)
TEXT 3.6.1.4 2
11/16/2016
Title:
CONTAINMENT SYSTEMS CONTAINMENT PRESSURE TEXT 3.6.1.5 2
11/16/2016
Title:
CONTAINMENT SYSTEMS DRYWELL AIR TEMPERATURE TEXT 3.6.1.6 1
11/16/2016
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 Titl-e: CONTAINMENT SYSTEMS SUPPRESSION POOL WATER LEVEL TEXT 3.6.2.3 2
11/16/2016
Title:
CONTAINMENT SYSTEMS RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING TEXT 3.6.2.4 1
11/16/2016
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: 03/29/22
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 6
03/03/2020
Title:
PLANT SYSTEMS RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SYSTEM AND THE ULTIMATE HEAT SINK (UHS)
TEXT 3.7.2 4
03/03/2020
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 TEXT3.7.4 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
11/1.6/2016
Title:
PLANT SYSTEMS Page 6 of 8
Report Date: 03/29/22
SSES MANUAL Manual Name:
TSBl Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL*
TEXT 3.8.1 1.4 05/26/2021
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 4
11/16/2016
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 2
11/16/2016
Title:
ELECTRICAL POWER SYSTEMS BATTERY CELL PARAMETERS TEXT 3.8.7
.3 09/04/2019
Title:
ELECTRICAL POWER SYSTEMS DISTRIBUTION SYSTEMS -
OPERATING TEXT 3.8.8 2
03/05/2019
Title:
ELECTRICAL POWER SYSrEMS 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: 03/29/22
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.l 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. 1.0. 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 3
03/29/2022
Title:
SPECIAL OPERATIONS SHUTDOWN MARGIN (SDM) TEST -
REFUELING Page 8 of 8
Report Date: 03/29/22
Rev. 7 Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)
B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design 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. The MCPR fuel cladding integrity SL ensures that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core do not experience transition boiling.
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 in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in ari uncontrolled release of activity to the reactor coolant.
SUSQUEHANNA - UNIT 1 2.0-1
BASES APPLICABLE SAFETY ANALYSES Rev. 7 Reactor Core SLs B 2.1.1 The fuel cladding must not sustain damage as a result of normal operation and AOOs. The reactor core SLs are established to preclude violation of the fuel design criterion that an MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.
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-ff.
The use of the ACE/ATRIUM 11 (Reference 6) correlation is valid for critical power calculations with A TRI UM 11 fuel at pressures ~ 588.8 psi a (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 x106 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
BASES APPLICABLE SAFETY ANALYSES (continued) 2.1.1.2 MCPR Rev. 7 Reactor Core SLs B 2.1.1 The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an AOO from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e., MCPR = 1.00) and the MCPR SL is based on a detailed statistical procedure that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty in the critical power correlation. References 2, 4, 5, and 6 describe the methodology used in determining the MCPR SL.
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.
The effects of channel bow on MCPR are explicitly included in the 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.
SUSQUEHANNA - UNIT 1 2.0-3
BASES APPLICABLE SAFETY ANALYSES
( continued)
SAFETY LIMITS APPLICABILITY SAFETY LIMIT VIOLATIONS 2.1.1.3 Reactor Vessel Water Level Rev. 7 Reactor Core SLs B 2.1.1 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.
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.
SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.
Exceeding an SL may cause fuel damage and create a potential for 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
BASES REFERENCES
- 1.
10 CFR 50, Appendix A, GDC 10.
Rev. 7
- Reactor Core Sls B 2.1.1
- 2.
ANP-10307PA, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," (as identified in the COLR).
- 3.
Not used.
- 4.
EMF-2209(P)(A), "SPCB Critical Power Correlation," AREVA NP, (as identified in the COLR).
- 5.
EMF-2158(P)(A), "Siemens Power Corporation Methodology for.
Boiling Water Reactors: Evaluation and Validation of CASMO-4/Microburn-B2," (as identified in the COLR).
- 6.
ANP-10335P-A, "ACE/ATRIUM 11 Critical Power Correlation," (as identified in the COLR).
SUSQUEHANNA - UNIT 1 2.0-5
BASES THIS PAGE INTENTIONALLY LEFT BLANK
- SUSQUEHANNA - UNIT 1 2.0-6 Rev. 7 Reactor Core Sls B 2.1.1
B 3.1 B 3.1.1 BASES BACKGROUND APPLICABLE SAFETY ANALYSES REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN (SOM)
SOM requirements are specified to ensure:
Rev. 2 SOM B 3.1.1
- 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.
The control rod drop accident (CRDA) analysis (Refs. 2 and 3) assumes the core is subcritical with the highest worth control rod withdrawn.
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.
SUSQUEHANNA - UNIT 1 3.1-1
BASES
. APPLICABLE SAFETY ANALYSES (continued)
LCO APPLICABILITY ACTIONS Rev. 2 SOM B 3.1.1 Prevention or mitigation of reactivity insertion events is necessary to limit energy deposition in the fuel to prevent significant fuel damage, which could result in undue release of radioactivity. Adequate SOM ensures 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).
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).
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 criticality 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).
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
BASES ACTIONS (continued)
B.1 Rev. 2 SOM B 3.1.1 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, D.3, and 0.4 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 (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 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.
SUSQUEHANNA - UNIT 1 3.1-3
BASES ACTIONS (continued)
SURVEILLANCE REQUIREMENTS E.1, E.2, E.3, E.4, and E.5 (continued)
Rev. 2 SOM B 3.1.1 Action must also be immediately initiated to fully insert all insertable 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.
SR 3.1.1.1 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 burnup, 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 SUSQUEHANNA - UNIT 1 3.1-4
BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.1.1.1 (continued)
Rev. 2 SOM H 3.1.1 demonstrations that rely solely on calculation of the highest worth control rod, additional margin (0.10% t.k/k) must be added to the SDM limit of 0.28% t.k/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 BPWS. 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 BPWS, 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 SDM 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.
SUSQUEHANNA - UNIT 1 3.1-5
BASES REFERENCES
- 2. FSAR, Section 15.
Rev.2 SOM B 3.1.1
- 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.
SUSQUEHANNA - UNIT 1 3.1-6
BASES THIS PAGE INTENTIONALLY LEFT BLANK SUSQUEHANNA - UNIT 1 3.1-7 Rev. 2 SOM B 3.1.1
Rev. 5 Rod Pattern Control B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Rod Pattern Control BASES BACKGROUND APPLICABLE SAFETY ANALYSES 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.
The analytical methods and assumptions used in evaluating the CRDA are summarized in References 1 and 2. CRDA analyses assume that the 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 banked position withdrawal sequence (BPWS). The BPWS is applicable from the condition of all control rods fully inserted to 10% RTP (Ref. 2). For the BPWS, 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 (e.g., between notches 08 and 12). 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 demonstrate SUSQUEHANNA - UNIT 1 3.1-34
BASES APPLICABLE SAFETY ANALYSES
( continued)
Rev. 5 Rod Pattern Control B 3.1.6 that the 280 cal/gm fuel damage limit will not be violated during a CRDA while following the BPWS mode of operation for control rod patterns.
These analyses consider the effects of fully inserted inoperable and OPERABLE control rods not withdrawn in the normal sequence of BPWS, but are still in compliance with the BPWS requirements regarding out of sequence control rods. These requirements allow a limited number (i.e.,
eight) and distribution of fully inserted inoperable control rods.
When performing a shutdown of the plant, an optional BPWS 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 :c:;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 BPWS control rod insertion, or may be bypassed and the improved BPWS shutdown sequence implemented under LCO 3.3.2.1, Condition D controls.
In order to use the Reference 9 BPWS 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 :c:;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 :c:;10% 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
- c:;10% 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 BPWS control rod insertion process contained in Reference 9 are followed, the plant is considered to be in compliance with the BPWS requirements, as required by LOC 3.1.6.
Rod pattern control satisfies Criterion 3 of the NRC Policy Statement (Ref. 8).
SUSQUEHANNA - UNIT 1 3.1-35
BASES LCO APPLICABILITY ACTIONS Rev. 5 Rod Pattern Control B 3.1.6 Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with the BPWS. 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 BPWS.
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.
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
BASES ACTIONS (continued)
SURVEILLANCE REQUIREMENTS A.1 and A.2 (continued)
Rev. 5 Rod Pattern Control B 3.1.6 "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 BPWS, 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.
SR 3.1.6.1 The control rod pattern is periodically verified to be in compliance with the BPWS to ensure the assumptions of the CRDA analyses are met. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The RWM which 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
BASES REFERENCES Rev. 5 Rod Pattern Control B 3.1.6
- 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).
- 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.
- 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. 4 APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
BASES BACKGROUND APPLICABLE SAFETY ANALYSES The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that limits specified in 1 O CFR 50.46 are not exceeded during the postulated design basis loss of coolant accident (LOCA).
LOCA calculations for the ATRIUM 10 and ATRIUM 11 fuel designs were performed. The analytical methods and assumptions used in evaluating Design Basis Accidents (DBAs) that determine the APLHGR Umits are presented in References 7, 8 and 9.
LOCA analyses are performed to ensure that the APLHGR limits are adequate to meet the Peak Cladding Temperature (PCT), maximum cladding oxidation, and maximum hydrogen generation limits of 10 CFR 50.46. The analyses are performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within the assembly.
The specific analytical methods and assumptions used in evaluating the fuel design limits from 10 CFR 50.46 for the ATRIUM 1 O fuel design are presented in Reference 3 and 4. The specific analytical methods and assumptions used in evaluating the fuel design limits from 1 O CFR 50.46 for the A TRI UM 11 fuel design are presented in Reference 11.
APLHGR limits are developed as a function of fuel type and exposure.
LOCA analyses were performed for the regions of the power/flow map bounded by the rod line that runs through 100% RTP and maximum core flow and the upper boundary of the MELLLA region. The MELLLA region is analyzed to determine whether an APLHGR multiplier as a function of core flow is required. The results of the analysis demonstrate the PCTs
. are within the 10 CFR 50.46 limit, and that APLHGR multipliers as a function of core flow are not required.
SUSQUEHANNA - UNIT 1 3.2-1
BASES APPLICABLE SAFETY ANALYSES (continued)
LCO APPLICABILITY ACTIONS Rev. 4 APLHGR B 3.2.1 Finally, the LOCA analyses were performed for Single-Loop Operation.
The results of the analysis for ATRIUM-10 fuel shows that an APLHGR limit which is 0.8 times the two-loop APLHGR limit meets the 10 CFR 50.46 acceptance criteria, and that the PCT is less than the limiting two-loop PCT. The results of the analyses for ATRIUM 11 fuel show that an APLHGR limit which is 0.8 times the two-loop APLHGR limits meets the 1 O CFR 50.46 acceptance criteria, and that the PCT is less than the limiting two-loop PCT.
The APLHGR satisfies Criterion 2 of the NRC Policy Statement (Ref. 10).
The APLHGR limits specified in the COLR are the result of the OBA analyses.
The APLHGR limits are primarily derived from LOCA analyses that are assumed to occur at high power levels. Design calculations and operating experience have shown that as power is reduced, the margin to the required APLHGR limits increases. At THERMAL POWER levels
< 23%RTP, the reactor is operating with substantial margin to the APLHGR limits; thus, this LCO is not required.
A.1 If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the OBA may not be met. Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates 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 sufficient to restore the APLHGR(s) to within its limits and is acceptable based on the low probability of a OBA occurring simultaneously with the APLHGR out of specification.
B.1 If the APLHGR 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.
SUSQUEHANNA-UNIT 1 3.2-2
BASES SURVEILLANCE
. REQUIREMENTS REFERENCES SR 3.2.1.1 Rev. 4 APLHGR B 3.2.1 APLHGRs are required to be initially calculated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is ~ 23% RTP and periodically thereafter. Additionally, APLHGRs 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 />. APLHGRs are 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 APLHGRs must be calculated prior to exceeding 44% RTP. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
- 1.
Not used.
- 2.
Not used.
- 3.
EMF-2361 (P)(A), "EXEM BWR-2000 ECCS Evaluation Model,"
Framatome ANP, (as identified in the COLR).
- 4.
EMF-2292(P)(A), "ATRIUM'-10: Appendix K Spray Heat Transfer Coefficients," (as identified in the COLR).
- 5.
Not used.
- 6.
Not used.
- 7.
FSAR, Chapter 4.
- 8.
FSAR, Chapter 6.
- 9.
FSAR, Chapter 15.
- 10. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
11 ANP-10332P-A, "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Loss of Coolant Accident Scenarios," (as identified in the COLR).
SUSQUEHANNA - UNIT 1 3.2-3
BASES THIS PAGE INTENTIONALLY LEFT BLANK SUSQUEHANNA - UNIT 1 3.2-4 Rev.4 APLHGR B 3.2.1
Rev. 5 MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)
BASES BACKGROUND APPLICABLE SAFETY ANALYSES 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 MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs). 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.
The analytical methods and assumptions used in evaluating the AOOs to establish the operating limit MCPR are presented in References 2, 3, 5, 7, and 10 for ATRIUM 10 fuel design analysis and references 2, 3, 5, 7, 10, arid 12 through 15 for ATRIUM 11 fuel designs. To ensure that the MCPR 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 (~CPR). When the largest
~CPR is added to the MCPR SL, the required operating limit MCPR is obtained.
The MCPR operating limits derived from the transient analysis 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 SUSQUEHANNA - UNIT 1 3.2-5
BASES Rev.5 MCPR B 3.2.2 APPLICABLE combinations of plant conditions (i.e., control rod scram speed, bypass SAFETY ANALYSES valve performance, EOC-RPT, cycle exposure; etc.). Flow dependent (continued)
MCPR limits are determined by analysis of slow flow runout transients.
LCO APPLICABILITY ACTIONS The MCPR satisfies Criterion 2 of the NRC Policy Statement (Ref. 11 ).
The MCPR operating limits specified in the COLR 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.
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.
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.
SUSQUEHANNA - UNIT 1 3.2-6
BASES ACTIONS (continued)
SURVEILLANCE REQUIREMENTS B.1 Rev. 5 MCPR B 3.2.2 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.
SR 3.2.2.1 The MCPR is required to be initially calculated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is ~ 23% RTP and 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 then the MCPR operating limits corresponding to the Maximum Allowable Average Scram Insertion Time must be implemented. Determining MCPR 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.
SUSQUEHANNA - UNIT 1 3.2-7
BASES REFERENCES
- 1.
NUREG-0562, June 1979.
Rev. 5 MCPR B 3.2.2
- 2.
XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors," Exxon Nuclear Company, March 1983.
- 3.
XN-NF-80-19(P)(A) Volume 3 Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERM EX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company, January 1987.
- 4.
Not used.
- 5.
XN-NF-80-19 (P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, June 1986.
- 6.
7.
- 8.
- 9.
- 10.
Not used.
EMF-2209(P)(A), "SPCB Critical Power Correlation," AREVA NP, (as identified in the COLR).
Not used.
Not used ANF-1358(P)(A), "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Framatome ANP, (as identified in the
- COLR).
11.
Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
- 12.
ANP-10300P-A, "AURORA-B: An Evaluation Model for Boiling 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).
SUSQUEHANNA - UNIT 1 3.2-8
BASES REFERENCES (continued)
- 15.
SUSQUEHANNA - UNIT 1 Rev. 5 MCPR B 3.2.2 ANP-10335P-A, "ACE/ATRIUM-11 Critical Power Correlation," (as identified in the COLR).
3.2-9
Rev. 4 LHGR B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)
BASES BACKGROUND APPLICABLE SAFETY ANALYSES The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location. Limits on LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation. Exceeding the LHGR limit could potentially result in fuel damage and subsequent release of radioactive materials. Fuel design limits are specified to ensure that fuel system damage, fuel rod failure, or inability to cool the fuel does not occur during the normal operations identified in Reference 1.
The analytical methods and assumptions used in evaluating the fuel system design are presented in References 1, 2, 3, and 4. The fuel assembly is designed to ensure (in conjunction with the core nuclear and thermal hydraulic design, plant equipment, instrumentation, and protection system) that fuel damage will not result in the release of radioactive materials in excess of regulatory limits. The mechanisms that could cause fuel damage during operational transients and that are considered in fuel evaluations are:
- a. Rupture of the fuel rod cladding caused by strain from the relative expansion of the UO2 pellet; and
- b. Severe overheating of the fuel rod cladding caused by inadequate cooling.
A value of 1 % plastic strain of the fuel cladding has been defined as the limit below which fuel damage caused by overstraining of the fuel cladding is not expected to occur (Ref. 3).
Fuel design evaluations have been performed and demonstrate that the 1 % fuel cladding plastic strain design limit is not exceeded during continuous operation with LHGRs up to the operating limit specified in the COLR. Transient evaluations were also performed. Reference 4 establishes LHGR acceptance criteria on strain and fuel overheating (fuel centerline melt) for both normal operation and anticipated operational occurrences.
The LHGR satisfies Criterion 2 of the NRC Policy Statement (Ref. 7).
SUSQUEHANNA - UNIT 1 3.2-10
BASES LCO APPLICABILITY ACTIONS Rev.4 LHGR B 3.2.3 The LHGR is a basic assumption in the fuel design analysis. The fuel has been designed to operate at rated core power with sufficient desigh margin to the LHGR calculated to cause a 1 % fuel cladding plastic strain. The operating limit to accomplish this objective is specified in the COLR.
The LHGR limits are derived from fuel design analysis that is limiting at high power level conditions. At core thermal power levels < 23% RTP, the reactor is operating with a substantial margin to the LHGR limits and, therefore, the Specification is only required when the reactor is operating at
- 23% RTP.
A.1 If any LHGR exceeds its required limit, an assumption regarding an initial condition of the fuel design analysis is not met. Therefore, prompt action should be taken to restore the LHGR(s) to within its required limits such that the plant is 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 LHGR(s) to within its limits and is acceptable based on the low probability of a transient or Design Basis Accident occurring simultaneously with the LHGR out of specification.
B.1 If the LHGR 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 is 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.
SUSQUEHANNA - UNIT 1 3.2-11
BASES SURVEILLANCE REQUIREMENTS REFERENCES SR 3.2.3.1 Rev. 4 LHGR B 3.2.3 The LHGR is required to be initially calculated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is ~ 23% RTP and periodically thereafter. Additionally, LHGRs 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 />. The LHGR 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 lower power levels and because the LHGRs must be calculated prior to exceeding 44% RTP. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
1, FSAR, Section 4.
- 2. FSAR, Section 5.
- 3. NUREG-0800,Section II.A.2(g), Revision 2, July 1981.
- 4. ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs," (as identified in the COLR).
- 5. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
SUSQUEHANNA - UNIT 1 3.2-12
BASES THIS PAGE INTENTIONALLY LEFT BLANK SUSQUEHANNA-UNIT 1 3.2-13 Rev. 4 LHGR B 3.2.3
Rev. 6 Recirculation Loops Operating B 3.4.1 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.1 Recirculation Loops Operating BASES BACKGROUND The Reactor Coolant Recirculation System is designed to provide a forced coolant flow through the core to remove heat from the fuel. The forced coolant flow removes more heat from the fuel than would be possible with just natural circulation. The forced flow, therefore, allows operation at significantly higher power than would otherwise be possible. The recirculation system also controls reactivity over a wide span of reactor power by varying the recirculation flow rate to control the void content of the moderator. The Reactor Coolant Recirculation System consists of two recirculation pump loops external to the reactor vessel. These loops provide the piping path for the driving flow of water to the reactor vessel jet pumps. Each external loop contains one variable speed motor driven recirculation pump, a motor generator (MG) set to control pump speed and associated piping, jet pumps, valves, and instrumentation. The recirculation pump, piping, and valves are part of the reactor coolant pressure boundary and are located inside the drywell structure. The jet pumps are reactor vessel internals.
The recirculated coolant consists of saturated water from the steam separators and dryers that has been subcooled by incoming feedwater.
This water passes down the annulus between the reactor vessel wall and the core shroud. A portion of the coolant flows from the vessel, through the two external recirculation loops, and becomes the driving flow for the jet pumps. Each of the two external recirculation loops discharges high pressure flow into an external manifold, from which individual recirculation inlet lines are routed to the jet pump risers within the reactor vessel. The remaining portion of the coolant mixture in the annulus becomes the suction flow for the jet pumps. This flow enters the jet pump at suction inlets and is accelerated by the driving flow. The drive flow and suction flow are mixed in the jet pump throat section. The total flow then passes through the jet pump diffuser section into the area below the core (lower plenum), gaining sufficient head in the process to drive the required flow upward through the core. The subcooled water enters the bottom of the fuel channels and contacts the fuel cladding, where heat is transferred to the coolant. As it rises, the coolant begins to boil, creating steam voids within the fuel channel that continue until the coolant exits the core.
Because of reduced moderation, the steam voiding introduces negative reactivity that must be compensated for to maintain or to increase reactor power. The recirculation flow control allows operators to increase recirculation flow and sweep some of the voids from the fuel channel, SUSQUEHANNA - UNIT 1 3.4-1
BASES BACKGROUND (continued)
APPLICABLE SAFETY AI\\IALYSES Rev. 6 Recirculation Loops Operating B 3.4.1 overcoming the negative reactivity void effect. Thus, the reason for having variable recirculation flow is to compensate for reactivity effects of boiling over a wide range of power generation without having to move control rods and disturb desirable flux patterns.
Each recirculation loop is manually started from the control room. The MG set provides regulation of individual recirculation loop drive flows. The flow in each loop is manually controlled.
The operation of the Reactor Coolant Recirculation System is an initial condition assumed in the design basis loss of coolant accident (LOCA)
(Ref. 1). During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the accident. The initial core flow decrease is rapid because the recirculation pump in the broken loop ceases to pump reactor coolant to the vessel almost immediately. The pump in the intact loop coasts down relatively slowly. This pump coastdown governs the core flow response for the next several seconds until the jet pump suction is uncovered (Ref. 1).
The analyses assume that both loops are operating at the same flow prior to the accident. However, the LOCA analysis was reviewed for the case with a flow mismatch between the two loops, with the pipe break assumed to be in the loop with the higher flow. While the flow coastdown and core response are potentially more severe in this assumed case (since the intact loop starts at a lower flow rate and the core response is the same as if both loops were operating at a lower flow rate), a small mismatch has been determined to be acceptable based on engineering judgment. The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2), which are analyzed in Chapter 15 of the FSAR.
Plant specific LOCA analyses have been performed assuming only one operating recirculation loop. These analyses have demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided that the APLHGR limits for ATRIUM 10 and ATRIUM 11 fuel are modified.
SUSQUEHANNA - UNIT 1 3.4-2
BASES APPLICABLE SAFETY ANALYSES
( continued)
LCO Rev. 6 Recirculation Loops Operating B 3.4.1 The transient analyses of Chapter 15 of the FSAR have also been performed for single recirculation loop operation and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR, LHGR, and MCPR limits for single loop operation are specified in the COLR. The APRM Simulated Thermal Power-High setpoint is in LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation." In addition, a restriction on recirculation pump speed is incorporated to address reactor vessel internals vibration concerns and assumptions in the event analysis.
Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 5).
Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. With the limits specified in SR 3.4.1.1 not met, the recirculation loop with the lower flow must be considered not in operation. With only one recirculation loop in operation, modifications to the required APLGHR limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE"), LHGR limits (LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGRr'), MCPR limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), and APRM Simulated Thermal Power-High setpoint (LCO 3.3.1.1) may be applied to allow continued operation consistent with the safety analysis assumptions.
Furthermore, restrictions are placed on recirculation pump speed to ensure the initial assumption of the event analysis are maintained.
The LCO is modified by a Note that allows up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to establish the required limits and setpoints after a change from two recirculation loops operation to single recirculation loop operation. If the limits and setpoints are not in compliance with the applicable requirements at the end of this period, the ACTIONS required by the applicable specifications must be implemented. This time is provided to stabilize operation with one recirculation loop by: limiting flow in the operating loop, limiting total THERMAL POWER, monitor APRM and local power range monitor (LPRM) neutron flux noise levels; and, fully implementing and confirming the required limit and setpoint modifications.
SUSQUEHANNA - UNIT 1 3.4-3
BASES APPLICABILITY ACTIONS Rev. 6 Recirculation Loops Operating B 3.4.1 In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.
In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.
A.1 When operating with no recirculation loops operating in MODE 1, the potential for thermal-hydraulic oscillations is greatly increased. Although this transient is protected for expected modes of oscillation by the OPRM system, when OPERABLE per LCO 3.3.1.1, Function 2.f (Reference 3, 4),
the prudent response to the natural circulation condition is to preclude potential thermal-hydraulic oscillations by immediately placing the mode switch in the shutdown position.
B.1 Recirculation loop flow must match within required limits when both recirculation loops are in operation. If flow mismatch is not within required limits, matched flow must be restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. If matched flows are not restored, the recirculation loop with lower flow must be declared "not in operation." Should a LOCA occur with recirculation loop flow not matched, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed prior to imposing restrictions associated with single loop operation. Operation with only one recirculation loop satisfies the requirements of the LCO and the initial conditions of the accident sequence.
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 based on the low probability of an accident occurring during this time period, providing a reasonable time to complete the Required Action, and considering that frequent core monitoring by operators allows abrupt changes in core flow conditions to be. quickly detected.
These Required Actions do not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing SUSQUEHANNA - UNIT 1 3.4-4
BASES ACTIONS (continued)
SURVEILLANCE REQUIREMENTS B.1 (continued)
Rev.6 Recirculation Loops Operating B 3.4.1.
recirculation pump speed to re-establish forward flow or by tripping the pump. Additional recirculation loop flow rate restrictions during single loop operation are provided in TRM Section 3.4.6.
C.1 With no recirculation loops in operation while in MODE 2 or if after going to single loop operations the required limits and setpoints cannot be established, the plant must be brought to MODE 3, where the LCO does not apply within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics.
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 to reach MODE 3 from full power conditions in an orderly manner without challenging plant systems.
SR 3.4.1.1 This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < 75 million lbm/hr), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is < 75 million lbm/hr. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop.
The mismatch is measured in terms of core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered inoperable. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.1.2 As noted, this SR is only applicable when in single loop operation. This SR ensures the recirculation pump limit is maintained. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SUSQUEHANNA - UNIT 1 3.4-5
BASES REFERENCES
- 1. FSAR, Section 6.3.3.7.
- 2. FSAR, Section 5.4.1.4.
Rev. 6 Recirculation Loops Operating B 3.4.1
- 3. GE NEDO-31960-A "BWROG Long Term Stability Solutions Licensing Methodology," November 1995.
- 4. GE NEDO-31960-A "BWROG Long Term Stability Solutions Licensing Methodology, Supplement 1," November 1995.
- 5. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
SUSQUEHANNA - UNIT 1 3.4-6
BASES Rev.6 Recirculation Loops Operating B 3.4.1 THIS PAGE INTENTIONALLY LEFT BLANK SUSQUEHANNA - UNIT 1 3.4-7
BASES Rev.6 Recirculation Loops Operating B 3.4.1 THIS PAGE INTENTIONALLY LEFT BLANK SUSQUEHANNA - UNIT 1 3.4-8
BASES Rev. 6 Recirculation Loops Operating B 3.4.1 THIS PAGE INTENTIONALLY LEFT BLANK SUSQUEHANNA - UNIT 1 3.4-9
Rev. 7 ECCS-Operating B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC).SYSTEM B 3.5.1 ECCS-Operating BASES BACKGROUND The ECCS is designed, in conjunction with the primary and secondary containment, to limit the release of radioactive materials to the environment following a loss of coolant accident (LOCA). The ECCS uses two independent methods (flooding and spraying) to cool the core during a LOCA. The ECCS network consists of the High Pressure Coolant Injection (HPCI) System, the Core Spray (CS) System, the low pressure coolant injection (LPCI) mode of the Residual Heat Removal (RHR) System, and the Automatic Depressurization System (ADS). The suppression pool provides the required source of water for the ECCS. Although no credit is taken in the safety analyses for the condensate storage tank (CST), it is capable of providing a source of water for the HPCI and CS systems.
On receipt of an initiation signal, ECCS pumps automatically start; simultaneously, the system aligns and the pumps inject water, taken either from the CST or suppression pool, into the Reactor Coolant System (RCS) as RCS pressure is overcome by the discharge pressure of the ECCS pumps. Although the system is initiated, ADS action is delayed, allowing the operator to interrupt the timed sequence if the system is not needed.
The HPCI pump discharge pressure quickly exceeds that of the RCS, and the pump injects coolant into the vessel to cool the core. If the break is small, the HPCI System will maintain coolant inventory as well as vessel level while the RCS is still pressurized. If HPCI fails, it is backed up by ADS in combination with LPCI and CS. In this event absent operator action, the ADS timed sequence would time out and open the selected safety/relief valves (S/RVs) depressurizing the RCS, thus allowing the LPCI and CS to overcome RCS pressure and inject coolant into the vessel. If the break is large, RCS pressure initially drops rapidly and the LPCI and CS cool the core.
Water from the break returns to the suppression pool where it is used again and again. Water in the suppression pool is circulated through a heat exchanger cooled by the RHR Service Water System. Depending on the location and size of the break, portions of the ECCS may be ineffective; however the overall design is effective in cooling the core regardless of the size or location of the piping break. Although no credit is taken in the safety analysis for the RCIC System, it performs a similar function as HPCI, but SUSQUEHANNA - UNIT 1 3.5-1
BASES BACKGROUND (continued)
Rev. 7 EGGS-Operating B 3.5.1 has reduced makeup capability. Nevertheless, it will maintain inventory and cool the core while the RCS is still pressurized following a reactor pressure vessel (RPV) isolation.
All ECCS subsystems are designed to ensure that no single active component failure will prevent automatic initiation and successful operation of the minimum required ECCS equipment.
The CS System (Ref. 1) is composed of two independent subsystems.
Each subsystem consists of two motor driven pumps, a spray sparger above the core, and piping and valves to transfer water from the suppression pool to the sparger. The CS System is designed to provide cooling to the reactor core when reactor pressure is low. Upon receipt of an initiation signal, the CS pumps in both subsystems are automatically started when AC power is available. When the RPV pressure drops sufficiently, CS System flow to the RPV begins. A full flow test line is provided to route water from and to the suppression pool to allow testing of the CS System without spraying water in the RPV.
LPCI is an independent operating mode of the RHR System. There are two LPCI subsystems (Ref. 2), each consisting of two motor driven pumps and piping and valves to transfer water from the suppression pool to the RPV via the corresponding recirculation loop. The two LPCI subsystems can be interconnected via the RHR System cross tie valves; however, at least one of the two cross tie valves is maintained closed with its power removed to prevent loss of both LPCI subsystems during a LOCA. The LPCI subsystems are designed to provide core cooling at low RPV pressure. Upon receipt of an initiation signal, all four LPCI pumps are automatically started. RHR System valves in the LPCI flow path are automatically positioned to ensure the proper flow path for water from the suppression pool to inject into the recirculation loops. When the RPV pressure drops sufficiently, the LPCI flow to the RPV, via the corresponding recirculation loop, begins. The water then enters the reactor through the jet pumps.
Full flow test lines are provided for each LPCI subsystem to route water from the suppression pool, to allow testing of the LPCI pumps without injecting water into the RPV. These test lines also provide suppression pool cooling capability, as described in LCO 3.6.2.3, "RHR Suppression Pool Cooling."
SUSQUEHANNA - UNIT 1 3.5-2
BASES BACKGROUND (continued)
Rev. 7 ECCS-Operating B 3.5.1 The HPCI System (Ref. 3) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger. Suction piping for the system is provided from the CST and the suppression pool. Pump suction for HPCI is normally aligned to the CST source to minimize injection of suppression pool water into the RPV. Whenever the CST water supply is low, an automatic transfer to the suppression pool water source ensures an adequate suction head for the pump and an uninterrupted water supply for continuous operation of the HPCI System. The steam supply to the HPCI turbine is piped from a main steam line upstream of the associated inboard main steam isolation valve.
The HPCI System is designed to provide core cooling for a wide range of reactor pressures (165 psia to 1225,psia). Upon receipt of an initiation signal, the HPCI turbine stop valve and turbine control valve open and the turbine accelerates to a specified speed. As the HPCI flow increases, the turbine control valve is automatically adjusted to maintain design flow.
Exhaust steam from the HPCI turbine is discharged to the suppression pool. A full flow test line is provided to route water to the CST to allow testing of the HPCI System during normal operation without injecting water into the RPV.
The ECCS pumps are provided with minimum flow bypass lines, which discharge to the suppression pool. The valves in these lines automatically open to prevent pump damage due to overheating when other discharge line valves are closed. To ensure rapid delivery of water to the RPVand to minimize water hammer effects, all ECCS pump discharge lines are filled with water. The HPCI, LPCI and CS System discharge lines are kept full of water using a "keep fill" system that is supplied using the condensate transfer system.
The ADS (Ref. 4) consists of 6 of the 16 S/RVs. It is designed to provide depressurization of the RCS during a small break LOCA if HPCI fails or is unable to maintain required water level in the RPV. ADS operation reduces the RPV pressure to within the operating pressure range of the low pressure ECCS subsystems (CS and LPCI), so that these subsystems can provide coolant inventory makeup. Each of the S/RVs used for automatic depressurization is equipped with two gas accumulators and associated inlet check valves. The accumulators provide the pneumatic power to actuate the valves.
SUSQUEHANNA - UNIT 1 3.5-3
BASES APPLICABLE SAFETY ANALYSES Rev. 7 ECCS-Operating B 3.5.1 The ECCS performance is evaluated for the entire spectrum of break sizes for a postulated LOCA. The accidents for which ECCS operation is required are presented in References 5, 6, and 7. The required analyses and assumptions are defined in Reference 8. The results of these analyses are also described in Reference 9.
This LCO helps to ensure that the following acceptance criteria for the ECCS, established by 10 CFR 50.46 (Ref. 10), will be met following a LOCA, assuming the worst case single active component failure in the ECCS:
- a. Maximum fuel element cladding temperature is ::::: 2200°F;
- b. Maximum cladding oxidation is ::::: 0.17 times the total cladding thickness before oxidation;
- 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react;
- d. The core is maintained in a coolable geometry; and
- e. Adequate long term cooling capability is maintained.
The fuel vendor performed LOCA calculations for the ATRIUM 10 and ATRIUM 11 fuel designs. The limiting single failures for the analyses are discussed in Reference 9. The LOCA analyses examine both recirculation pipe breaks and non-recirculation pipe breaks. For the recirculation pipe breaks, breaks on both the discharge and suction side of the recirculation pump are performed for two geometries; double-ended guillotine and split break.
The LOCA calculations demonstrate that the limiting fuel type (highest PCT) is ATRIUM 10 fuel. The most limiting (highest PCT) break is a double-ended guillotine break in the recirculation pump suction piping.
The limiting single failure is the failure of the LPCI injection valve in the intact recirculation loop to open.
One ADS valve failure is analyzed as a limiting single failure for events requiring ADS operation. The remaining OPERABLE ECCS subsystems provide the capability to adequately cool the core and prevent excessive fuel damage.
The ECCS satisfy Criterion 3 of the NRC Policy Statement (Ref. 15).
SUSQUEHANNA~ UNIT 1 3.5-4
BASES LCO APPLICABILITY ACTIONS Rev. 7 ECCS-Operating B 3.5.1 Each ECCS injection/spray subsystem and six ADS valves are required to be OPERABLE. The ECCS injection/spray subsystems are defined as the two CS subsystems, the two LPCI subsystems, and one HPCI System.
The low pressure ECCS injection/spray subsystems are defined as the two CS subsystems and the two LPCI subsystems.
With less than the required number of ECCS subsystems OPERABLE, the potential exists that during a limiting design basis LOCA concurrent with the worst case single failure, the limits specified in Reference 1 O could be exceeded. All ECCS subsystems must therefore be OPERABLE to satisfy the single failure criterion required by Reference 10.
LPCI subsystems may be considered OPERABLE during alignment and operation for decay heat removal when below the actual RHR cut in permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core cooling, thereby allowing operation of RHR shutdown cooling when necessary.
All ECCS subsystems are required to be OPERABLE during MODES 1, 2, and 3, when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a break in the primary system piping. In MODES 2 and 3, when reactor steam dome pressure is ~ 150 psig, ADS and HPCI are not required to be OPERABLE because the low pressure ECCS subsystems can provide sufficient flow below this pressure. Requirements for MODES 4 and 5 are specified in LCO 3.5.2, "Reactor Pressure Vessel (RPV) Water Inventory Control."
A Note prohibits the application of LCO 3.0.4.b to an inoperable HPCI subsystem. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable HPCI subsystem and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
SUSQUEHANNA - UNIT 1 3.5-5
BASES ACTIONS (continued)
A.1 Rev. 7 EGGS-Operating B 3.5.1 If any one low pressure ECCS injection/spray subsystem is inoperable for reasons other than Condition B, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA. However, overall ECCS reliability is reduced, because a single failure in one of the remaining OPERABLE subsystems, concurrent with a LOCA, may result in the ECCS not being able to perform its intended safety function. The 7 day Completion Time is based on a reliability study (Ref. 12) that evaluated the impact on ECCS availability, assuming various components and subsystems were taken out of service. The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowed outage times (i.e., Completion Times).
B.1 If one LPCI pump in one or both LPCI subsystems is inoperable, the inoperable LPCI pumps must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE LPCI pumps and at least one CS subsystem provide adequate core cooling during a LOCA.
However, overall ECCS reliability is reduced, because a single failure in one of the remaining OPERABLE subsystems, concurrent with a LOCA, may result in the ECCS not being able to perform its iritended safety function. A 7 day Completion Time is based on a reliability study cited in Reference 12 and has been found to be acceptable through operating experience.
C.1 and C.2 If the inoperable low pressure ECCS subsystem or LPCI pump(s) cannot be restored to OPERABLE status within the associated Completion Time, 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 at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SUSQUEHANNA - UNIT 1 3.5-6
BASES ACTIONS (continued)
D.1 and D.2 Rev. 7 ECCS-Operating.
B 3.5.1 If the HPCI System is inoperable and the RCIC System is verified to be OPERABLE, the HPCI System must be restored to OPERABLE status within 14 days. In this Condition, adequate core cooling is ensured by the OPERABILITY of the redundant and diverse low pressure ECCS injection/spray subsystems in conjunction with ADS. Also, the RCIC System will automatically provide makeup water at most reactor operating pressures. Verification of RCIC OPERABILITY is therefore required when HPCI is inoperable. This may be performed as an administrative check by examining logs or other information to determine if RCIC is out of service for maintenance or other reasons. It does not mean to perform the Surveillances needed to demonstrate the OPERABILITY of the RCIC System. If the OPERABILITY of the RCIC System cannot be verified, however, Condition H must be immediately entered. If a single active component fails concurrent with a design basis LOCA, there is a potential, depending on the specific failure, that the minimum required ECCS equipment will not be available. A 14 day Completion Time is based on a reliability study cited in Reference 12 and has been found to be acceptable through operating experience.
E.1 and E.2 If Condition A or Condition B exists in addition to an inoperable HPCI System, the inoperable low pressure ECCS injection/spray subsystem or the LPCI pump(s) or the HPCI System must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, adequate core cooling is ensured by the OPERABILITY of the ADS and the remaining low pressure ECCS subsystems. However, the overall ECCS reliability is significantly reduced because a single failure in one of the remaining OPERABLE subsystems concurrent with a design basis LOCA may result in the ECCS not being able to perform its intended safety function. Since both a high pressure system (HPCI) and a low pressure subsystem are inoperable, a more restrictive Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is required to restore either the HPCI System or the low pressure ECCS injection/spray subsystem to OPERABLE status. This Completion Time is based on a reliability study cited in Reference 12 and has been found to be acceptable through operating experience.
SUSQUEHANNA - UNIT 1 3.5-7
BASES ACTIONS
- ( continued)
Ll Rev. 7 ECCS-Operating B 3.5.1 The LCO requires six ADS valves to be OPERABLE in order to provide the ADS function. Reference 9 contains the results of an analysis that evaluated the effect of one ADS valve being out of service. Per this analysis, operation of only five ADS valves will provide the required depressurization. However, overall reliability of the ADS is reduced, because a single failure in the OPERABLE ADS valves could result in a reduction in depressurization capability. Therefore, operation is only allowed for a limited time. The 14 day Completion Time is based on a reliability study cited in Reference 12 and has been found to be acceptable through operating experience.
G.1 and G.2 If Condition A or Condition B exists in addition to one inoperable ADS valve, adequate core cooling is ensured by the OPERABILITY of HPCI and the remaining low pressure ECCS injection/spray subsystem. However, overall ECCS reliability is reduced because a single active component failure concurrent with a design basis LOCA could result in the minimum required ECCS equipment not being available. Since both a high pressure system (ADS) and a low pressure subsystem are inoperable, a more restrictive Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is required to restore either the low pressure ECCS subsystem or the ADS valve to OPERABLE status. This Completion Time is based on a reliability study cited in Reference 12 and has been found to be acceptable through ope~ating experience.
H.1 and H.2 If any Required Action and associated Completion Time of Condition D, E, F, or G is not met, or if two or more ADS valves are inoperable, the plant must be brought to a condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reactor steam dome pressure reduced to :s; 150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
L1 When multiple ECCS subsystems are inoperable, as stated in Condition I, LCO 3.0.3 must be entered immediately.
SUSQUEHANNA - UNIT 1 3.5-8
BASES SURVEILLANCE REQUIREMENTS SR 3.5.1.1 Rev. 7 ECCS-Operating B 3.5.1 The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the HPCI System, CS System, and LPCI subsystems full of water ensures that the ECCS will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring that the lines are full is to vent at the high points. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.1.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the HPCI System, this SR also includes the steam flow path for the turbine and the flow controller position.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that allows LPCI subsystems to be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the RHR cut in permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. This.
allows operation in the RHR shutdown cooling mode during MODE 3, if necessary.
SUSQUEHANNA - UNIT 1 3.5-9
BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.3 Rev. 7 ECCS-Operating B 3.5.1 Verification that ADS gas supply header pressure is z 135 psig ensures adequate gas pressure for reliable ADS operation. The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The design pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at least one valve actuations can occur with the drywell at 70% of design pressure.
The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of z 135 psig is provided by the containment instrument gas system. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.1.4 Verification every 31 days that at least one RHR System cross tie valve is closed and power to its operator is disconnected ensures that each LPCI subsystem remains independent and a failure of the flow path in one subsystem will not affect the flow path of the other LPCI suosystem.
Acceptable methods of removing power to the operator include opening the breaker, or racking out the breaker, or removing the breaker. If both RHR System cross tie valves are open or power has not been removed from at least one closed valve operator, both LPCI subsystems must be considered inoperable. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.1.5 Verification that each 480 volt AC swing bus transfers automatically from the normal source to the alternate source on loss of power while supplying its respective bus demonstrates that electrical power is available to ensure proper operation of the associated LPCI inboard injection and minimum flow valves and the recirculation pump discharge and bypass valves. Therefore, each 480 volt AC swing bus must be OPERABLE for the associated LPCI subsystem to be OPERABLE. The test is performed by actuating the load test switch or by disconnecting the preferred power source to the transfer switch and verifying that swing bus automatic transfer is accomplished. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SUSQUEHANNA - UNIT 1 3.5-10
BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.6 Rev. 7 ECCS-Operating B 3.5.1 Cycling the recirculation pump discharge and bypass valves through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and provides assurance that the valves will close when required to ensure the proper LPCI flow path is established. Upon initiation of an automatic LPCI subsystem injection signal, these valves are required to be closed to ensure full LPCI subsystem flow injection in the reactor via the recirculation jet pumps. De-energizing the valve in the closed position will also ensure the proper flow path for the LPCI subsystem. Acceptable methods of de-energizing the valve include opening the breaker, or racking out the breaker, or removing the breaker.
The specified Frequency is once during reactor startup before THERMAL POWER is> 25% RTP. However, this SR is modified by a Note that states the Surveillance is only required to be performed if the last performance was more than 31 days ago. Therefore, implementation of this Note requires this test to be performed during reactor startup before exceeding 25% RTP. Verification during reactor startup prior to reaching> 25% RTP is an exception to the normal lnservice Testing Program generic valve cycling Frequency, but is considered acceptable due to the demonstrated reliability of these valves. If the valve is inoperable and in the open position, the associated LPCI subsystem must be declared inoperable.
SR 3.5.1.7, SR 3.5.1.8, and SR 3.5.1.9 The performance requirements of the low pressure ECCS pumps are determined through application of the 1 O CFR 50, Appendix K criteria (Ref. 8). This periodic Surveillance is performed (in accordance with the ASME OM Code requirements for the ECCS pumps) to verify that the ECCS pumps will develop the flow rates required by the respective analyses. The low pressure ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of Reference 10. The pump flow rates are verified against a system head equivalent to the RPV pressure expected during a LOCA. The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during a LOCA. These values may be established during preoperational testing.
SUSQUEHANNA - UNIT 1 3.5-11
BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.7, SR 3.5.1.8, and SR 3.5.1.9 (continued)
Rev. 7 ECCS-Operating B 3.5.1 The flow tests for the HPCI System are performed at two different pressure ranges such that system capability to provide rated flow is tested at both the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the HPCI System diverts steam flow. Reactor steam pressure is considered adequate when :2'. 920 psig to perform SR 3.5.1.8 and :2'. 150 psig to perform SR 3.5.1.9. However, the requirements of SR 3.5.1.9 are met by a successful performance at any pressure :5 165 psig. Adequate steam flow is represented by at least 1.25 turbine bypass valves open. Therefore, sufficient time is allowed after adequate pressure and flow are achieved to perform these tests. Reactor startup is allowed prior to performing the low pressure Surveillance test because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance test is short. The reactor pressure is allowed to be increased to normal operating pressure since it is
. assumed that the low pressure test has been satisfactorily completed and there is no indication or reason to believe that HPCI is inoperable.
Therefore, SR 3.5.1.8 and SR 3.5.1.9 are modified by Notes that state the Surveillances are not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the reactor steam pressure and flow are adequate to perform the test.
The Frequency for SR 3.5.1.7 and SR 3.5.1.8 is in accordance with the lnservice Testing Program requirements. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.1.10 The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCI, CS, and LPCI will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions. This functional test includes the LPCI
- and CS interlocks between Unit 1 and Unit 2 and specifically requires the following:
A functional test of the interlocks associated with the LPCI and CS pump starts in response to an automatic initiation signal in Unit 1 followed by a false automatic initiation signal in Unit 2; SUSQUEHANNA - UNIT 1 3.5-12
BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.10 (continued)
Rev. 7 ECCS-Operating B 3.5.1 A functional test of the interlocks associated with the LPCI and CS pump starts in response to an automatic initiation signal in Unit 2 followed by a false automatic initiation signal in Unit 1; and A functional test of the interlocks associated with the LPCI and CS pump starts in response to simultaneous occurrences of an automatic initiation signal in both Unit 1 and Unit 2 and a loss of Offsite power condition affecting both Unit 1 and Unit 2.
The purpose of this functional test (preferred pump logic) is to assure that if a false LOCA signal were to be received on one Unit simultaneously with an actual LOCA signal on the second Unit, the preferred LPCI and CS pumps are started and the non-preferred LPCI and CS pumps are tripped for each Unit. This functional test is performed by verifying that the non-preferred LPCI and CS pumps are tripped. The verification that preferred LPCI and CS pumps start is performed under a separate surveillance test. Only one division of LPCI preferred pump logic is required to be OPERABLE for each Unit, because no additional failures needs to be postulated with a false LOCA signal. If the preferred or non-preferred pump logic for CS is inoperable, the associated CS pumps shall be declared inoperable and the pumps should not be operated to ensure that the opposite Unit's CS pumps or 4.16 kV E;SS Buses are protected.
This SR also ensures that the HPCI System will automatically restart on an RPV low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool. The LOG IC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlaps this Surveillance.
This SR can be accomplished by any series of sequential overlapping'or total steps such that the entire channel is tested.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that excludes vessel injection/spray during the Surveillanc~. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
SUSQUEHANNA - UNIT 1 3.5-13
BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.11 Rev. 7 ECCS-Operating B 3.5.1 The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e.,
solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SR 3.5.1.12 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that excludes valve actuation. This prevents an RPV pressure blowdown.
SR 3.5.1.12 A manual actuation of each ADS valve actuator is performed to verify that the valve and solenoid are functioning properly. This is demonstrated by the methods described below. Proper operation of the valve tailpipes is ensured through the use of foreign material exclusion during maintenance.
Valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation.
Manual actuation of the actuator at atmospheric temperature and pressure during cold shutdown is performed. Proper functioning of the valve actuator and solenoid is demonstrated by visual observation of actuator movement. The ADS actuator will be disconnected from the valve to ensure no damage is done to the valve seat or to the valve internals. Each valve shall be bench-tested prior to reinstallation. The bench-test along with the test on the ADS actuator establishes the OPERABILITY of the valves.
SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SUSQUEHANNA - UNIT 1 3.5-14
BASES SURVEILLANCE REQUIREMENTS
( continued)
REFERENCES SR 3.5.1.13 Rev. 7 ECCS-Operating B 3.5.1 This SR ensures that the ECCS RESPONSE TIME for each ECCS injection/spray subsystem is less than or equal to the maximum value assumed in the accident analysis. Response Time testing acceptance criteria are included in Reference 13. This SR is modified by a Note that allows the instrumentation portion of the response time to be assumed to be based on historical response time data and therefore, is excluded from the ECCS RESPONSE TIME testing. This is allowed since the instrumentation response time is a small part of the ECCS RESPONSE TIME (e.g., sufficient margin exists in the diesel generator start time when compared to the instrumentation response time) (Ref. 14).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
- 1.
FSAR, Section 6.3.2.2.3.
2.
FSAR, Section 6.3.2.2.4.
- 3.
FSAR, Section 6.3.2.2.1.
- 4.
FSAR, Section 6.3.2.2.2.
- 5.
FSAR, Section 15.2.8.
- 6.
FSAR, Section 15.6.4..
- 7.
FSAR, Section 15.6.5.
- 8.
- 9.
FSAR, Section 6.3.3.
- 10.
- 11.
Not used
- 12. Memorandum from R.L. Baer (NRC) to V. Stello, Jr. (NRC),
"Recommended Interim Revisions to LCOs for ECCS Components,"
December 1, 1975.
SUSQUEHANNA - UNIT 1 3.5-15
BASES REFERENCES (continued)
- 13. FSAR, Section 6.3.3.3.
Rev. 7 ECCS-Operating B 3.5.1
- 14. NEDO 32291-A, "System Analysis for the Elimination of Selected Response Time Testing Requirements, October 1995.
- 15. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
SUSQUEHANNA - UNIT 1 3.5-16
Rev. 2 Control Rod Testing-Operating B 3.10.7 B 3.10 SPECIAL OPERATIONS B 3.10.7 Control Rod Testing-Operating BASES BACKGROUND APPLICABLE SAFETY ANALYSES The purpose of this Special *Operations LCO is to permit control rod testing, while in MODES 1 and 2, by imposing certain administrative controls. 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"), such that only the specified control rod sequences and relative positions required by LCO 3.1.6, "Rod Pattern Control," are allowed over the operating range from all control rods inserted to the low power setpoint (LPSP) of the RWM. The sequences effectively limit the potential amount and rate of reactivity increase that could occur during a control rod drop accident (CRDA). During these conditions, control rod testing is sometimes required that may result in control rod patterns not in compliance with the prescribed sequences of LCO 3.1.6. These tests include SOM demonstrations, control rod scram time testing, control rod friction testing, and testing performed during the Startup Test Program (e.g. local criticality). This Special Operations LCO provides the necessary exemption to the requirements of LCO 3.1.6 and provides additional administrative controls to allow the deviations in such tests from the prescribed sequences in LCO 3.1.6.
The analytical methods and assumptions used in evaluating the CRDA area summarized in References 1 and 2. CRDA analyses assume the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analyses.
The RWM provides backup to operator control of the withdrawal sequences to ensure the initial conditions of the CRDA analyses are not violated. For special sequences developed for control rod testing, the initial control rod patterns assumed in the safety analysis of References 1 and 2 may not be preserved. Therefore, special CRDA analyses are required to demonstrate that these special sequences will not result in unacceptable consequences, should a CRDA occur during the testing. These analyses, performed in accordance with an NRC approved methodology, are dependent on the specific test being performed.
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.
SUSQUEHANNA - UNIT 1 3.10-29
BASES LCO APPLICABILITY Rev. 2 Control Rod Testing-Operating B 3.10.7 As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Control rod testing may be performed in compliance with the prescribed sequences of LCO 3.1.6, and during these tests, no exceptions to the requirements of LCO 3.1.6 are necessary. For testing performed with a sequence not in compliance with LCO 3.1.6, the requirements of LCO 3.1.6 may be suspended, provided additional administrative controls are placed on the test to ensure that the assumptions of the special safety analysis for the test sequence are satisfied. Assurances that the test sequence is followed can be provided by either programming the test sequence into the RWM, with conformance verified as specified in SR 3.3.2.1.8 and allowing the RWM to monitor control rod withdrawal and provide appropriate control rod blocks if necessary, or by verifying conformance to the approved test sequence by a second licensed operator or other qualified member of the technical staff. These controls are consistent with those normally applied to operation in the startup range as defined in the SRs and ACTIONS of LCO 3.3.2.1, "Control Rod Block Instrumentation."
Control rod testing, while in MODES 1 and 2, with THERMAL POWER greater than the LPSP of the RWM, is adequately controlled by the existing LCOs on power distribution limits and control rod block instrumentation. Control rod movement during these conditions is not restricted to prescribed sequences and can be performed within the constraints of LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," LCO 3.2.2, 'MINIMUM CRITICAL POWER RATIO (MCPR)," LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," and LCO 3.3.2.1. With THERMAL POWER less than or equal to the LPSP of the RWM, the provisions of this Special Operations LCO are necessary to perform special tests that are not in conformance with the prescribed sequences of LCO 3.1.6.
While in MODES 3 and 4, control rod withdrawal is only allowed if performed in accordance with Special Operations LCO 3.10.3, "Single Control Rod Withdrawal-Hot Shutdown," of Special Operations LCO 3.10.4, "Single Control Rod Withdrawal-Cold Shutdown," which provide adequate controls to ensure that the assumptions of the safety analyses of Reference 1 c1nd 2 are satisfied. During these Special Operations and while in MODE 5, the one-rod-out interlock (LCO 3.9.2, "Refuel Position One-Rod-Out Interlock,") and scram functions (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," and LCO 3.9.5, "Control Rod OPERABILITY-Refueling"), or the added administrative controls prescribed in the applicable Special Operations LCOs, provide mitigation of potential reactive excursions.
- SUSQUEHANNA - UNIT 1 3.10-30
BASES ACTIONS SURVEILLANCE REQUIREMENTS REFERENCE Rev. 2 Control Rod Testing-Operating B3.10.7 With the requirements of the LCO not met (e.g., the control rod pattern is not in compliance with the special test sequence, the sequence is improperly loaded in the RWM) the testing is required to be immediately suspended. Upon suspension of the special test, the provisions of LCO 3.1.6 are no longer expected, and appropriate actions are to be taken to restore the control rod sequence to the prescribed sequence of LCO 3.1.6, or to shut down the reactor, if required by LCO 3.1.6.
SR 3.10.7.1 With the special test sequence not programmed into the RWM, a second licensed operator or other qualified member of the technical staff is required to verify conformance with the approved sequence for the test.
This verification must be performed during control rod movement to prevent deviations from the specified sequence. A Note is added to indicate that this Surveillance does not need to be performed if SR 3.10.7.2 is satisfied.
SR 3.10.7.2 When the RWM provides conformance to the special test sequence, the test sequence must be verified to be correctly loaded into the RWM prior to control rod movement. This Surveillance demonstrates compliance with SR 3.3.2.1.8, thereby demonstrating that the RWM is OPERABLE. A Note has been added to indicate that this Surveillance does not need to be performed if SR 3.10.7.1 is satisfied.
- 1. FSAR 15.4.9
- 2. 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 3.10-31
BASES Rev. 2 Control Rod Testing-Operating B 3.10.7 THIS PAGE INTENTIONALLY LEFT BLANK SUSQUEHANNA - UNIT 1 3.10-32
Rev.3 SOM Test-Refueling B 3.10.8 B 3.10 SPECIAL OPERATIONS B 3.10.8 SHUTDOWN MARGIN (SOM) Test-Refueling BASES BACKGROUND APPLICABLE SAFETY ANALYSES 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 (SDM)," 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 SDM 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 SDM. 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.
Prevention and mitigation of unacceptable reactivity excursions during control rod withdrawal, with the reactor mode switch in the startup/hot 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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BASES APPLICABLE SAFETY ANALYSES
( continued)
LCO Rev. 3 SOM Test-Refueling B 3.10.8 CRDA analyses assume that the reactor operator follows prescribed withdrawal sequences. For SOM tests performed within these defined sequences, the analyses of Reference 1 is applicable. However, for 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 BPWS.
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.
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.
SUSQUEHANNA-UNIT 1 3.10-34
BASES LCO
( continued)
APPLICABILITY Rev.3 SOM Test-Refueling B 3.10.8 Local critical tests require the withdrawal of control rods in a sequence that is not in conformance with the BPWS. This testing would therefore require bypassing or reprogramming of the rod worth minimizer to allow the withdrawal of rods not in conformance with BPWS, 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 banked position withdrawal 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 AL TE RATIONS 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.
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
BASES ACTIONS Rev. 3 SOM Test-Refueling B 3.10.8 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 num.ber 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.
B.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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BASES SURVEILLANCE REQUIREMENTS SR 3.10.8.1 Rev. 3 SOM Test-Refueling B 3.10.8 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..
SR 3.10.8.4 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.
SR 3.10.8.5 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.
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BASES SURVEILLANCE REQUIREMENTS
( continued)
REFERENCE SR 3.10.8.6 Rev. 3 SOM Test-Refueling B 3.10.8 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.
- 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).
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