ML23025A036
| ML23025A036 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 01/16/2023 |
| From: | Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML23025A036 (1) | |
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MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2023-1172 USER INFORMATION:
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of 2
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL I
I Table Of Contents Issue Date:
01/15/2023 Procedure Name Rev Issue Date Change ID Change NUI!lber TEXT LOES 138 01/03/2019
Title:
LIST OF EFFECTIVE SECTIONS TEXT TOC 25 03/05/2019
Title:
TABLE OF CONTENTS TEXT 2.1.1 6
03/31/2021
Title:
SAFETY LIMITS (SLS) REACTOR CORE SLS' TEXT 2.1.2 1
TEXT 3.1.1 TEXT 3.1.2 TEXT 3.1.4 5
11/16/2016
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM TIMES TEXT 3.1.5 2
11(16/2016
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS TEXT 3.1. 6 6
01/15/2023
Title:
REACTIVITY CONTROL SYSTEMS ROD PATTERN CONTROL Page 1 of 8
Report Date: 01/16/23
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS*BASES UNIT 2 MANUAL TEXT 3.1. 7 5
01/05/2023
Title:
REACTIVITY CGNTROL SYSTEMS STANDBY LIQUID CONTROL (SLC) SYSTEM I
TEXT 3.1. 8 4
11/16/2016
Title:
REACTIVITY CONTROL SYSTEMS SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVES TEXT 3.2*.1 6
03/31/2021
Title:
POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
TEXT 3.2.2 5
03/31/2021
Title:
POWER DISTRIBUTION LIMITS MINIMUM CRITICAL POWER RATIO (MCPR)
TEXT 3.2.3 4
03/31/2021
Title:
POWER DISTRIBUTION LIMITS LINEAR HEAT GENERATION RATE LHGR TEXT 3.3.1.l 7
01/05/202_3
Title:
INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS) INSTRUMENTATION TEXT 3. 3.1. 2 4
01/23/2018 I
Title:
INSTRUMENTATION SOURCE RANGE MONITOR (SRM) INSTRUMENTATION TEXT 3.3.2.1 6
01/15/2023
Title:
INSTRUMENTATION CONTROL ROD BLOCK INSTRUMENTATION TEXT 3.3.2.2 4
01/05/2023
Title:
INSTRUMENTATION FEEDWATER -
MAIN TURBINE HIGH WATER LEVEL TRIP INSTRUMENTATION TEXT 3.3.3.1 9
11/16/2016
Title:
INSTRUMENTATION POST ACCIDENT MONITORING (PAM) INSTRUMENTATION TEXT 3.3.3.2 2
11/16/2016
Title:
INSTRUMENTATION REMOTE SHUTDOWN SYSTEM TEXT 3.3.4.1 3
01/05/2023
Title:
INSTRUMENTATION END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT) INSTRUMENTATION Page 2 of 8
Report Date: 01/16/23
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.3.4.2 2
01/05/2023
Title:
INSTRUMENTATION ANTICIPATED TRANSIENT WITHOUT SCRAM RECIRCULATION PUMP TRIP (ATWS-RPT) INSTRUMENTATION TEXT 3.3.5.1 8
01/05/2023
Title:
INSTRUMENTATION EMERGENCY CORE COOLING SYSTEM (ECCS) INSTRUMENTATION TEXT 3.3.5.2 3
03/18/2021
Title:
REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL INSTRUMENTATION TEXT 3.3.5.3 1
01/05/2_023
Title:
UNIT 1 REACTOR PRESSURE VESSEL WIC TS CHANGES TEXT 3.3.6.1 10 01/05/2023
Title:
INSTRUMENTATION PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TEXT 3.3.6.2 6
03/05/2019
Title:
INSTRUMENTATION SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION TEXT 3.3.7.1 4
03/05/2019
Title:
INSTRUMENTATION CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS) SYSTEM INSTRUMENTATION TEXT 3.3.8.1 7
01/05/2023
Title:
INSTRUMENTATION LOSS OF POWER (LOP) INSTRUMENTATION TEXT 3.3.8.2 1
11/16/2016
Title:
INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS) ELECTRIC POWER MONITORING TEXT 3.4.1 6
03/31/2021
Title:
REACTOR COOLANT SYSTEM (RCS) RECIRCULATION LOOPS OPERATING TEXT 3. 4. 2 4
11 / 16 / 2 0 16
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: 01/16/23
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.4.4 1
11/16/2016
Title:
REACTOR COOLANT SYSTEM (RCS) RCS OPERATIONAL LEAKAGE TEXT 3.4.5 3
03/10/2010
Title:
REACTOR COOLANT SYSTEM (RCS) RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE TEXT 3.4.6 5
11/16/2016
Title:
REACTOR COOLANT SYSTEM (RCS) RCS LEAKAGE DETECTION INSTRUMENTATION TEXT 3.4.7 3
11/16/2016
Title:
REACTOR COOLANT SYSTEM (RCS) RCS SPECIFIC ACTIVITY TEXT 3.4.8 3
11/16/2016
Title:
REACTOR COOLANT SYSTEM (RCS) RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM HOT'SHUTDOWN TEXT 3.4.9 2
Title:
REACTOR COOLANT SYSTEM (RCS) 11/16/2016 RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM COLD SHUTDOWN TEXT 3.4.10 6
05/14/2019
Title:
REACTOR COOLANT SYSTEM (RCS) RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TEXT 3.4.11 1
11/16/2016
Title:
REACTOR COOLANT SYSTEM (RCS) REACTOR STEAM DOME PRESSURE TEXT 3.5.1 9
01/05/2023
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL AND REACTOR CORE ISOLATION COOLING (~CIC) SYSTEM ECCS OPERATING TEXT 3.5.2 7
06/09/2022
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ECCS OPERATING TEXT 3.5.3 7
01/05/2023
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) REACTOR PRESSURE VESSEL (RPV) WATER INVENTORY CONTROL AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ECCS OPERATING Page 4 of 8
Report Date: 01/16/23
- SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.6.1.1 6
11/16/2016
Title:
PRIMARY CONTAINMENT TEXT 3. 6. 1. 2 3
01/05/2023
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCK TEXT 3. 6. 1. 3 20 01/05/2023
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT ISOLATION VALVES (PCIVS)
TEXT 3.6.1.4 2
11/16/2016
Title:
CONTAINMENT SYSTEMS CONTAINMENT PRESSURE TEXT 3.6.1.5 2
11/16/2016
Title:
CONTAINMENT SYSTEMS DRYWELL AIR TEMPERATURE TEXT 3. 6. 1. 6 2
01/05/2023
Title:
CONTAINMENT SYSTEMS SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS TEXT 3.6.2.1 3
11/16/2016
Title:
CONTAINMENT SYSTEMS SUPPRESSION POOL AVERAGE TEMPERATURE TEXT 3.6.2.2 2
03/05/2019
Title:
CONTAINMENT SYSTEMS SUPPRESSION POOL WATER LEVEL TEXT 3.6.2.3 3
01/05/2023
Title:
CONTAINMENT SYSTEMS RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING TEXT 3.6.2.4 2
01/05/2023
Title:
CONTAINMENT SYSTEMS RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL SPRAY TEXT 3.6.3.1
Title:
CONTAINMENT SYSTEMS TEXT 3.6.3.2 2
4 06/13/2006 INTENTIONALLY LEFT BLANK 08/02/2021
Title:
CONTAINMENT SYSTEMS DRYWELL AIR *FLOW SYSTEM Page 5 of 8
Report Date: 01/16/23
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.6.3.3 3
09/29/2017
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT OXYGEN CONCENTRATION TEXT 3.6.4.1 17 12/16/2020
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT TEXT 3.6.4.2 14 03/05/2019
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT ISOLATION VALVES (SCIVS)
TEXT 3.6.4.3 7
03/05/2019
Title:
CONTAINMENT SYSTEMS STANDBY GAS TREATMENT (SGT) SYSTEM TEXT 3.7.1 10 01/05/2023
Title:
PLANT SYSTEMS RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SYSTEM AND THE ULTIMATE HEAT SINK (UHS)
TEXT 3.7.2 6
01/05/2023
Title:
PLANT SYSTEMS EMERGENCY SERVICE WATER (ESW) SYSTEM TEXT 3.7.3 4
03/05/2019
Title:
PLANT SYSTEMS CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS) SYSTEM l
TEXT 3.7.4 2
03/05/2019
Title:
PLANT SYSTEMS CONTROL ROOM FLOOR COOLING SYSTEM TEXT 3.7.5 2
11/16/2016
Title:
PLANT SYSTEMS MAIN CONDENSER OFFGAS TEXT 3.7.6 4
11/16/2016
Title:
PLANT SYSTEMS MAIN TURBINE BYPASS SYSTEM TEXT 3.7.7 2
11/16/2016
Title:
PLANT SYSTEMS SPENT FUEL STORAGE POOL WATER LEVEL TEXT 3.7.8 1
11/16/2016
Title:
MAINE TURBINE PRESSURE REGULATION SYSTEM Page 6 of 8
Report Date: 01/16/23
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.8.1 17 01/05/2023
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES -
OPERATING TEXT 3.8.2 2
03/18/2021
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES -
SHUTDOWN TEXT 3.8.3 7
08/07/2019
Title:
ELECTRICAL POWER SYSTEMS DIESEL FUEL OIL LUBE OIL AND STARTING AIR TEXT 3.8.4 5
01/05/2023
Title:
ELECTRICAL POWER SYSTEMS DC SOURCES -
OPERATING TEXT 3.8.5 2
03/05/2019
Title:
ELECTRICAL POWER SYSTEMS DC SOURCES -
SHUTDOWN TEXT 3.8.6 2
11/16/2016
Title:
ELECTRICAL POWER SYSTEMS BATTERY CELL PARAMETERS TEXT 3.8.7 9
01/05/2023
Title:
ELECTRICAL POWER SYSTEMS DISTRIBUTION SYSTEMS -
OPERATING TEXT 3.8.8 2
03/05/2019
Title:
ELECTRICAL POWER SYSTEMS DISTRIBUTION SYSTEMS -
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
11/18/2002
Title:
REFUELING OPERATIONS CONTROL ROD POSITION INDICATION I
Page 7 of 8
Report Date: 01/16/23
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.9.5 1
11/16/2016
Title:
REFUELING OPERATIONS CONTROL ROD OPERABILITY - REFUELING TEXT 3.9.6 2
11/16/2016
Title:
REFUELING OPERATIONS REACTOR PRESSURE VESSEL (RPV) WATER LEVEL TEXT 3.9.7 1
11/16/2016
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL TEXT 3.9.8 1
11/16/2016
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) -
LOW WATER LEVEL
_.I TEXT 3.10.1 2
03/05/2019
Title:
SPECIAL OPERATIONS INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION TEXT 3.10.2 1
11/16/2016
Title:
SPECIAL OPERATIONS REACTOR MODE SWITCH INTERLOCK TESTING TEXT 3.10.3 1
11/16/2016
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD WITHDRAWAL -
HOT SHUTDOWN TEXT 3.10.4 1
11/16/2016 Tit.le: SPECIAL OPERATIONS SINGLE CONTROL ROD WITHDRAWAL -
COLD SHUTDOWN TEXT 3.10.5 1
11/16/2016
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD DRIVE (CRD) REMOVAL - REFUELING TEXT 3.10.6 1
11/16/2016
Title:
SPECIAL OPERATIONS MULTIPLE CONTROL ROD WITHDRAWAL - REFUELING TEXT 3.10.7 2
03/31/2021
Title:
SPECIAL OPERATIONS CONTROL ROD TESTING -
OPERATING TEXT 3.10.8 4
03/31/2021
Title:
SPECIAL OPERATIONS SHUTDOWN MARGIN (SDM) TEST - REFUELING Page 8 of 8 Report Date: 01/16/23
Rev. 4 Control Rod OPERABILITY B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Control Rod OPERABILITY BASES BACKGROUND Control rods are components of the control rod drive (CRD) System, which is the primary reactivity control system for the reactor. In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes to ensure under conditions of normal operation, including anticipated operational occurrences, that specified acceptable fuel design limits are not exceeded. In addition, the control rods provide the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System. The CRD System is designed to satisfy the requirements of GDC 26, GDC 27, GDC 28, and GDC 29 (Ref. 1).
The CRD System consists of 185 locking piston control rod drive mechanisms (CRDMs) and a hydraulic control unit for each drive
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provide additional energy for scram. An index tube and piston, coupled to the control rod, are locked at fixed increments by a collet mechanism. The collet fingers engage notches in the index tube to prevent unintentional withdrawal of the control rod, but without restricting insertion.
APPLICABLE SAFETY ANALYSES This Specification, along with LCO 3.1.4, "Control Rod Scram Times," and LCO 3.1.5, "Control Rod Scram Accumulators," ensure that the performance of the control rods in the event of a Design Basis Accident (OBA) or transient meets the assumptions used in the safety analyses of References 2, 3, and 4.
The analytical methods and assumptions used in the evaluations involving control rods are presented in References 2, 3, and 4. The control rods provide the primary means for rapid reactivity control (reactor scram), for maintaining the reactor subcritical and for limiting the potential effects of reactivity insertion events caused by malfunctions in the CRD System.
SUSQUEHANNA - UNIT 2 3.1-13
BASES APPLICABLE SAFETY ANALYSES (continued)
Rev. 4 Control Rod OPERABILITY B3.1.3 The capability to insert the control rods provides assurance that the assumptions for scram reactivity in the OBA and transient analyses are not violated. Since the SOM ensures the reactor will be subcritical with the highest worth control rod withdrawn (assumed single failure), the additional failure of a second control rod to insert, if required, could invalidate the demonstrated SOM and potentially limit the ability of the CRO System to hold the reactor subcritical. If the control rod is stuck at an inserted position and becomes decoupled from the CRO, a control rod drop accident (CROA) can possibly occur. Therefore, the requirement that all control rods be OPERABLE ensures the CRO System can perform its intended function.
The control rods also protect the fuel from damage which could result in release of radioactivity. The limits protected are the MCPR Safety Limit
- (SL) {see Bases for SL 2.1.1, "Reactor Core SLs," and LCO 3.2.2, "MINIMUM CRITICAL POWER RA TIO (MCPR)"), the 1 % cladding plastic strain fuel design limit (see Bases for LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," and LCO 3.2.4, "Average Power Range Monitor (APRM) Gain and Setpoints"), and the fuel damage limit (see Bases for LCO 3.1.6, "Rod Pattern Control") during reactivity insertion
::;::==============::fllli:ffi~================================================
LCO The negative reactivity insertion (scram) provided by the CRO System provides the analytical basis for determination of plant thermal limits and provides protection against fuel damage limits during a CROA. The Bases for LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6 discuss in more detail how the SLs are protected by the CRO System.
Control rod OPERABILITY satisfies Criterion 3 of the NRC Policy Statement (Ref. 5).
The OPERABILITY of an individual control rod is based on a combination of factors, primarily, the scram insertion times, the control rod coupling integrity, and the ability to determine the control rod position. Accumulator OPERABILITY is addressed by LCO 3.1.5. The associated scram accumulator status for a control rod only affects the scram insertion times; therefore, an inoperable accumulator does not immediately require declaring a control rod inoperable. Although not all control rods are required to be OPERABLE to satisfy the intended reactivity control requirements, strict control over the number and distribution of inoperable control rods is required to satisfy the assumptions of the OBA and transient analyses.
SUSQUEHANNA - UNIT 2 3.1-14
BASES APPLICABILITY ACTIONS Rev.4 Control Rod OPERABILITY B 3.1.3 In MODES 1 and 2, the control rods are assumed to function during a OBA or transient and are therefore required to be OPERABLE in these MODES.
In MODES 3 and 4, control rods are not able to be withdrawn (except as permitted by LCO 3.10.3 and LCO 3.10.4) since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements for control rod OPERABILITY during these conditions.
Control rod requirements in MODE 5 are located in LCO 3.9.5, "Control Rod OPERABILITY-Refueling."
The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each control rod. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable control rod. Complying with the Required Actions may allow for continued operation, and* subsequent inoperable control rods are governed by subsequent Condition entry and application of associated Required Actions.
A.-1, A.2, A.3 and A.4
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water or scram pressure. With a fully inserted control rod stuck, no actions are required as long as the control rod remains fully inserted. The Required Actions are modified by a Note, which allows the rod worth minimizer (RWM) to be bypassed if required to allow continued operation.
LCO 3.3.2.1, "Control Rod Block Instrumentation," provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis. With one withdrawn control rod stuck, the local scram reactivity rate assumptions may not be met if the stuck control rod separation criteria are not met. This separation criteria stipulates that a stuck control rod is equivalent to a "slow" control rod for purposes of separation requirements between "slow control rods.
Therefore, a verification that the separation criteria are met must be performed immediately. The separation criteria are not met if a) the stuck control rod occupies a position adjacent to two "slow" control rods, b) the stuck control rod occupies a position adjacent to one "slow" control rod and the one "slow" control rod is also adjacent to another "slow" control rod, or, c) if the stuck control rod occupies a location adjacent to one "slow" control rod when there is another pair of "slow" control rods adjacent to one another. Adjacent control rods include control rods that are either face or diagonally adjacent. The description of "slow" control rods is provided in LCO 3.1.4, "Control Rod Scram Times." In addition, the associated control rod drive must be disarmed in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The allowed Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is acceptable, considering the reactor can still be shut down, SUSQUEHANNA - UNIT 2 3.1-15
BASES ACTIONS (continued)
A.1, A.2, A.3 and A.4 (continued)
Rev. 4 Control Rod OPERABILITY B 3.1.3 assuming no additional control rods fail to insert, and provides a reasonable time to perform the Required Action in an orderly manner.
Isolating the control rod from scram prevents damage to the CROM. The control rod can be isolated from scram and normal insert and withdraw pressure, yet still maintain cooling water to the CRD.
Monitoring of the insertion capability of each withdrawn control rod must also be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM. SR 3.1.3.3 performs periodic tests of the control rod insertion capability of withdrawn control rods. Testing each withdrawn control rod ensures that a generic problem does not exist. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." The Required Action A.3 Completion Time only begins upon discovery of Condition A concurrent with THERMAL POWER greater than the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of rod pattern control (LCO 3.1.6) and the RWM (LCO 3.3.2.1 ). The allowed Completion
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potential for a need to reduce power to perform the tests. To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SOM is also required within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Should a OBA or transient require a shutdown, to preserve the single failure criterion, an additional control rod would have to be assumed to fail to insert when required.
Therefore, the original SOM demonstration may not be valid. The SOM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn.
The allowed Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to verify SOM is adequate, considering that with a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram. Even with the postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control remains to reach and maintain MODE 3 conditions.
SUSQUEHANNA - UNIT 2 3.1-16
BASES Rev. 4 Control Rod OPERABILITY 133.1.3 ACTIONS 8.1 (continued)
With two or more withdrawn control rods stuck, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The occurrence of more than one control rod stuck at a withdrawn position increases the probability that the reactor cannot be shut down if required. Insertion of all insertable control rods eliminates tl:te possibility of an additional failure of a control rod to insert.
The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
C.1 and C.2 With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, operation may continue, provided the control rods are fully inserted within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and disarmed (electrically or hydraulically) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations.
The control rods can be hydraulically disarmed by closing the drive water fl{j---ffi(Aal:JS-t-water isolation \\'alves. The-eoAtfel-fOEis-c--aR-ee-eleet-r~saUy,_ ____ _
disarmed by disconnecting power from all four directional control valve solenoids. Required Action C.1 is modified by a Note, which allows the RWM to be bypassed if required to allow insertion of the inoperable control rods and continued operation. LCO 3.3.2.1 provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis.
The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging plant systems.
D.1. D.2. and D.3 Out of sequence control rods may increase the potential reactivity worth of a dropped control rod during a CRDA. At =:;; 10% RTP, the generic banked position withdrawal sequence (BPWS) analysis requires inserted control rods not in compliance with BPWS to be separated by at least two OPERABLE control rods in all directions, including the diagonal.
Therefore, if two or more inoperable control rods are not in compliance with BPWS and not separated by at least two OPERABLE control rods, action must be taken to restore compliance with BPWS or restore the control rods to OPERABLE status. Condition D is modified by a Note indicating that the Condition is not applicable when > 10% RTP, since the BPWS is not SUSQUEHANNA-UNIT 2 3.1-17
BASES ACTIONS (continued)
D.1, D.2, and D.3 (continued)
Rev.4 Control Rod OPERABILITY B 3.1.3 required to be followed under these conditions, as described in the Bases for LCO 3.1.6. The allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is acceptable, considering the low probability of a CRDA occurring.
Alternatively, Required Action D.3 allows action to be taken to confirm compliance with the analyzed rod position sequence within four hours. The analyzed rod position sequence shall be established consistent with Ref. 6, and may or may not be in compliance with the BPWS. The analyzed rod position sequence will ensure that all licensing requirements continue to be met with respect to the CRDA analyses. This is a temporary allowance and applicable only during Cycle 21. Upon completion of Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
E.1 and E.2 In addition to the separation requirements for inoperable control rods, a BPWS assumption requires that no more than three inoperable control rods
=-c================J:;li°~-leweEl-iR-a:EW-GAe IW----WS--Qr-0up.
Therefore, with one or more BPWS groups having four or more inoperable control rods, control rods must be restored to OPERABLE status so that no BPWS group has four or more inoperable control rods. Required Action E.1 is modified by a Note indicating that the Condition is not applicable when THERMAL POWER is > 10% RTP since the BPWS is not required to be followed under these conditions, as described in the Bases for LCO 3.1.6. The allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is acceptable, considering the low probability of a CRDA occurring.
Alternatively, Required Action E.2 allows action to be taken to confirm compliance with the analyzed rod position sequence within four hours. The analyzed rod position sequence shall be established consistent with Ref. 6, and may or may not be in compliance with the BPWS. The analyzed rod position sequence will ensure that all licensing requirements continue to be met with respect to the CRDA analyses. This is a temporary allowance and
- applicable only during Cycle 21. Upon completion of Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
SUSQUEHANNA - UNIT 2 3.1-18
BASES ACTIONS (continued)
SURVEILLANCE REQUIREMENTS E..1 Rev. 4 Control Rod OPERABILITY B 3.1.3 If any Required Action and associated Completion Time of Condition A, C, D, or E are not met, or there are nine or more inoperable control rods, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i.e., scram) of the control rods. The number of control rods permitted to be inoperable when operating above 10% RTP (e.g., no CRDA considerations) could be more than the value specified, but the occurrence of a large number of inoperable control rods could be indicative of a generic problem, and investigation and resolution of the potential problem should be undertaken.
The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant systems.
=~============+/-he=p_osi1i:0n=of:eaeh:_eonrro-m_d::m:tls-1=b_e.=de_te_rffl:tlle_d::t~...=e~deg=t1a=t~=======-
information on control rod position is available to the operator for determining CRD OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.1.3.2 NOT USED SR 3.1.3.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal. These Surveillances are not required when THERMAL POWER is less than or equal to the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the Banked Position Withdrawal Sequence (BPWS) (LCO 3.1.6) and the RWM (LCO 3.3.2.1). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SUSQUEHANNA - UNIT 2 3.1-19
BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.1.3.4 Rev.4 Control Rod OPERABILITY B 3.1.3 Verifying that the scram time for each control rod to notch position 05 is
- ,:; 7 seconds provides reasonable assurance that the control rod will insert when required during a OBA or transient, thereby completing its shutdown function. This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," and the functional testing of SDV vent and drain valves in LCO 3.1.8, "Scram Discharge Volume (SDV) Vent and Drain Valves," overlap this Surveillance to provide complete testing of the assumed safety function. The associated Frequencies are acceptable,,
considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.
SR 3.1.3.5 Coupling verification is performed to ensure the control rod is connected to the CROM and will perform its intended function when necessary. The
=-~=============S-runieiUance=i:equir.eS::.\\f.erif1:ing:::a:cont-i:-oJ::ro.ebio.e_S=r)_o_:t=g:ett-e-tMe-witMdr-awA------
overtravel position. The overtravel position feature provides a positive check on the coupling integrity since only an uncoupled CRD can reach the overtravel position. The verification is required to be performed any time a control rod is withdrawn to the "full out" position (notch position 48) or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This includes control rods inserted one notch and then returned to the "full out" position during the performance of SR 3.1.3.3. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved and operating experience related to uncoupling events.
SUSQUEHANNA - UNIT 2 3.1-20
BASES REFERENCES Rev.4 Control Rod OPERABILITY B 3.1.3
- 1.
1 O CFR 50, Appendix A GDC 26, GDC 27, GDC 28, and GDC 29.
- 2.
FSAR, Section 4.3.2
. 3.
FSAR, Section 4.6
- 4.
FSAR, Section 15.
- 5.
Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
- 6.
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 2 3.1-21
Rev. 6 Rod Pattern Control B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Rod Pattern Control BASES BACKGROUND 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 aU 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-defi~the potentiai initial-eondit1ons fer the GR0A-analys1s.
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 2 3.1-34
BASES APPLICABLE SAFETY ANALYSES (continued)
Rev.6 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 :s;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.
=~=============I:Ec0£Eler-t0-H-Se--t-Ae--RefeFeRse-9-BP-WS-sl'.JutooW13=ProGess,a-rt-extr-a-el:teck~---
is required in order to consider a control rod to be "confirmed" to be coupled. This extra check ensures that no Single Operator Error can result in an incorrect coupling check. For purposes of this shutdown process, the method for confirming that control rods are coupled varies depending on the position of the control rod in the core. Details on this coupling confirmation requirement are provided in Reference 9, which requires that any partially inserted control rods, which have not been confirmed to be coupled since their last withdrawal, be fully inserted prior to reaching THERMAL POWER of :=;;10% RTP. If a control rod has been checked for coupling at notch 48 and the rod has since only been moved inward, this rod is in contact with it's drive and is not required to be fully inserted prior to reaching THERMAL POWER of :s;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
- =;;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 2 3.1-35
BASES LCO APPLICABILITY Rev. 6 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.
The LCO is modified by a Note which states OPERABLE control rods may comply with the requirements of the analyzed rod position sequence in lieu of the BPWS. The analyzed rod position sequence shall be established consistent with Ref. 1, and may or may not be in compliance with the BPWS. The analyzed rod position sequence will ensure that all licensing requirements continue to be met with respect to the CRDA analyses. This is a temporary allowance and applicable only during Cycle 21. Upon completion of Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
In MODES 1 and 2, when THERMAL POWER is~ 10% RTP, the CRDA is a Design Basis Accident and, therefore, compliance with the assumptions
~=============o-f::tl:re=safety:ana*lysis=is:r:eq11i-r-ed:=Wben=El=l-s::==============
> 10% RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Ref. 2). In MODES 3, 4, and 5, since the reactor is shut down and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SOM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn.
ACTIONS A.1 and A.2 Condition A is modified by a footnote which states that for Cycle 21 only, one or more OPERABLE control rods not in compliance with the analyzed rod position sequence requires entry into Condition A rather than one or more OPERABLE control rods not in compliance with the BPWS. The analyzed rod position sequence shall be established consistent with Ref. 1, and may or may not be in compliance with the BPWS. The analyzed rod position sequence will ensure that all licensing requirements continue to be met with respect to the CRDA analyses. This is a temporary allowance and applicable only during Cycle 21. Upon completion of Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023. Required Actions A.1 and A.2 remain unchanged for this temporary requirement.
SUSQUEHANNA - UNIT 2 3.1-36
BASES ACTIONS (continued)
A.1 and A.2 (continued)
Rev.6 Rod Pattern Control 83.1.6 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
=-~=============Wln,ilU,e-mmre~J't:::Aei>mel:red--fl0HA--e0m~li-aRee-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, "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.
8.1 and 8.2 Condition Bis modified by a footnote which states that for Cycle 21 only, nine or more OPERABLE control rods not in compliance with the analyzed rod position sequence requires entry into Condition 8 rather than nine or more OPERABLE control rods not in compliance with the BPWS. The analyzed rod position sequence shall be established consistent with Ref. 1, and may or may not be in compliance with the BPWS. The analyzed rod position sequence will ensure that all licensing requirements continue to be met with respect to the CRDA analyses. This is a temporary allowance and applicable only during Cycle 21. Upon completion of Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023. Required Actions 8.1 and B.2 remain unchanged for this temporary requirement.
SUSQUEHANNA - UNIT 2 3.1-37
BASES ACTIONS (continued)
SURVEILLANCE REQUIREMENTS 8.1 and 8.2 (continued)
Rev.6 Rod Pattern Control 83.1.6 If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has less impact on control rod worth than withdrawals have. Required Action B.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3.3.2.1 requires verification of control rod movement by a qualified member of the technical staff.
When nine or more OPERABLE control rods are not in compliance with the prescribed control rod sequence-, the reactor mode switch must be placed in the shutdown position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. With the mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the
-RtFeH-ees-oot-ef.se§J~eR~;M,--------------------
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 provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at :S; 10% RTP.
SR 3.1.6.1 is modified by a footnote which allows verification of the control rod sequence against the analyzed rod position sequence in lieu of the BPWS. The analyzed rod position sequence shall be established consistent with Ref. 1, and may or may not be in compliance with the BPWS. The analyzed rod position sequence will ensure that all licensing requirements continue to be met with respect to the CRDA analyses. This is a temporary allowance and applicable only during Cycle 21. Upon completion of Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
SUSQUEHANNA - UNIT 2 3.1-38
BASES REFERENCES Rev. 6 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,
.1:1~y-2-2,'.!-99a--@g_i;-_R-a9-1-32}*r.-. ----------------
- 9.
NEDO 33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.
SUSQUEHANNA - UNIT 2 3.1-38a
Rev.6 Control Rod Block Instrumentation B 3.3.2.1 B 3.3 INSTRUMENTATION B 3.3.2.1 Control Rod Block Instrumentation BASES BACKGROUND Control rods provide the primary means for control of reactivity changes.
Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events. During low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions, control rod blocks from the Reactor Mode Switch-Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities.
The Nominal Trip Setpoint (NTSP) is a predetermined setting for a protective device chosen to ensure automatic actuation prior to the process
----=:::::::::=============:::\\rar-iab1e=reachi11g=the=-Ana1y_ticai=l::imit:and:.tb-ttS=enst1:ring=f:hat-::tbe~i-eJt:::::========
Limit (SL) would not be exceeded. The NTSP accounts for various uncertainties. As such, the NTSP meets the definition of a Limiting Safety System Setting (LSSS) because the protective instrument channel actuates to protect a reactor core or RCS pressure boundary Safety Limit. Rod Block Monitor functions 1 a, 1 b and 1 c are LSSSs.
Technical Specifications contain values related to the OPERABILITY of equipment required for safe operation of the facility. OPERABLE is defined in Technical Specifications as "... being capable of performing its specified safety function(s)." For automatic protective devices related to SLs, the required safety function is to ensure that a SL is not exceeded and therefore the NTSP is the LSSS, as defined by 10 CFR 50.36. However, use of the NTSP to define OPERABILITY in Technical Specifications would be an overly restrictive requirement if it were applied as an OPERABILITY limit for the "as-found" value during a Surveillance. This would result in Technical Specification compliance problems, as well as reports and corrective actions required by the rule, which are not necessary to ensure safety.
Use of the NTSP to define "as-found" OPERABILITY under the expected circumstances described above would result in actions required by both rule and Technical Specifications that are not warranted. However, there is also some point beyond which the device would have not been able to perform its function due, for example, to greater than expected drift. This SUSQUEHANNA-UNIT 2 3.3-44
BASES BACKGROUND (continued)
Rev. 6 Control Rod Block Instrumentation B 3.3.2.1 value needs to be specified in the Technical Specifications in order to define OPERABILITY of the devices and is designated as the Allowable Value, which is the least conservative value of the as-found setpoint that a channel can have during testing.
The Allowable Value specified in SR 3.3.2.1.7 is the least conservative value of the as-found setpoint that a channel can have when tested, such that a channel is OPERABLE if the setpoint is found conservative with respect to the Allowable Value during the CHANNEL CALIBRATION.
The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude MCPR Safety Limit violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above the low power range setpoint. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit. The
=~============+/-RB~M=eJIBflnel sigA-aH~eAer--ate1-by---av-era§ifil)---a---set-ef-L-eeal-P-0weF-Ra-Rg....._---
Monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn. A simulated thermal power signal from one of the four redundant Average Power Range Monitor (APRM) channels supplies a reference signal for one of the RBM channels and a simulated thermal power signal from another of the APRM channels supplies the reference signal to the second RBM channel. This reference signal is used to determine which RBM range setpoint (low, intermediate, or high) is enabled. If the APRM simulated thermal power is indicating less than the low power range setpoint, the RBM is automatically bypassed. The RBM is also automatically bypassed if a peripheral control rod is selected (Ref. 2).
The purpose of the RWM is to control rod patterns during startup, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP. The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA. Prescribed control rod sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses steam flow signals to determine when the reactor power is above the preset power level at which the RWM is automatically bypassed (Ref. 1). The RWM is a single channel system that provides input into RMCS rod block channel 2.
SUSQUEHANNA - UNIT 2 3.3-45
BASES BACKGROUND (continued)
Rev.6 Control Rod Block Instrumentation B 3.3.2.1 The function of the individual rod sequence steps (banking steps) is to minimize the potential reactivity increase from postulated CRDA at low power levels. However, if the possibility for a control rod to drop can be eliminated, then banking steps at low power levels are not needed to ensure the applicable event limits can not be exceeded. The rods may be inserted without the need to stop at intermediate positions since the possibility of a CRDA is eliminated by the confirmation that withdrawn control rods are coupled.
To eliminate the possibility of a CRDA, administrative controls require that any partially inserted control rods, which have not been confirmed to be coupled since their last withdrawal, be fully inserted prior to reaching the THERMAL POWER of :-s;;10% RTP. If a control rod has been checked for coupling at notch 48 and the rod has not been moved inward, this rod is in contact with its drive and is not required to be fully inserted prior to reaching the THERMAL POWER of ~10% RTP. However, if it cannot be confirmed that the control rod has been moved inward, then that rod shall be fully inserted prior to reaching the THERMAL POWER of :-s;;10% RTP.
The remaining control rods may then be inserted without the need to stop at intermediate positions since the possibility of a CRDA has been
=-=~==============~imir:iate1.1-..--------------------------------
APPLICABLE SAFETY
- ANALYSES, LCO, and APPLICABILITY With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is required to be in the sh.utdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods.
Allowable Values are specified for each applicable Rod Block Function listed in Table 3.3.2.1-1. The NTSPs (actual trip setpoints) are selected to ensure that the setpoints are conservative with respect to the Allowable Value. A channel is inoperable if its actual trip setpoint is non-conservative with respect to its required Allowable Value.
NTSPs are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The Analytical Limits are derived from the limiting values of the process parpmeters obtained from the safety analysis. The Allowable Values are derived from the Analytical Limits, corrected for calibration, process, and some of the instrument errors. The NTSPs are then SUSQUEHANNA - UNIT 2 3.3-46
BASES APPLICABLE SAFETY
- ANALYSES, LCO, and APPLICABILITY (continued)
Rev. 6 Control Rod Block Instrumentation B 3.3.2.1 determined, accounting for the remaining channel uncertainties. The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, and drift are accounted for.
The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.
- 1. Rod Block Monitor The RBM is designed to prevent violation of the MCPR SL and the cladding 1 % strain fuel design limit that may result from a Single Control Rod Withdrawal (RWE) event.
The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 14. The fuel thermal performance as a function of RBM Allowable Value is determined from the analysis. The NTSP and Allowable Values are chosen as a function of power level.
NTSP operating limits are established based on the specified Allowable Values.
The RBM function satisfies Criterion 3 of the NRC Policy Statement (Ref. 7).
Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Value for the associated power range, to ensure that no single instrument failure can preclude a rod block for this Function. The actual setpoints are calibrated consistent with applicable setpoint methodology.
Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS.
Nominal trip setpoints are those predetermined values of output at which an action should take place. The trip setpoints are compared to the actual process parameter, the calculated RBM flux (RBM channel signal). When the normalized RBM flux value exceeds the applicable trip setpoint, the RBM provides a trip output.
The analytic limits are derived from the limiting values of the process parameters. Using the GE setpoint methodology, based on ISA RP 67.04, Part II "Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation" setpoint calculation Method 2, the RBM Allowable Values are determined from the analytical limits using the statistical combination of the RBM input signal calibration error, process measurement error, primary element accuracy and instrument accuracy SUSQUEHANNA - UNIT 2 3.3-47
BASES APPLICABLE SAFETY
- ANALYSES, LCO, and APPLICABILITY (continued)
- 1. Rod Block Monitor (continued)
Rev.6 Control Rod Block Instrumentation B 3.3.2.1 under trip conditions. Accounting for these errors assures that a setpoint found during calibration at the Allowable Value has adequate margin to protect the analytical limit thereby protecting the Safety Limit.
For the digital RBM, there is a normalization process initiated upon rod selection, so that only RBM input signal drift over the interval from the rod selection to rod movement needs to be considered in determining the nominal trip setpoints. The RBM has no drift characteristic with no as-left or as-found tolerances since it only performs digital calculations on the digitized input signals provided by the APRMs.
The RBM Allowable Value demonstrates that the analytical limit would not be exceeded, thereby protecting the safety limit. The nominal trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and the environment errors are accounted for and appropriately applied for the RBM. There are no margins applied to the RBM nominal trip setpoint calculations which could
=~=============mas-k:::HBM:::fiegfadatieA-.--------------------
The RBM will function when operating greater than or equal to 28% RTP.
Below this power level, the RBM is not required to be OPERABLE.
The RBM selects one of three different RBM flux trip setpoints to be applied based on the current value of THERMAL POWER. THERMAL POWER is indicated to each RBM channel by a simulated thermal power (STP) reference signal input from an associated reference APRM channel. The OPERABLE range is divided into three "power ranges," a "low power range," an "intermediate power range," and a "high power range." The RBM flux trip setpoint applied within each of these three power ranges is, respectively, the "low trip setpoint," the "intermediate trip setpoint," and the "high trip setpoint" (Allowable Values for which are defined in the COLR).
To determine the current power range, each RBM channel compares its current STP input value to three power setpoints, the "low power setpoint",
(28%), the "intermediate power setpoint" (current value defined in the COLR), and the "high power setpoint" (current value defined in the COLR),
which define, respectively, the lower limit of the low power range, the lower limit of the intermediate power range, and the lower limit of the high power range. The trip setpoint applicable for each power range is more restrictive than the corresponding setpoint for the lower power range(s). When STP is below the low power setpoint, the RBM flux trip outputs are automatically bypassed but the low trip setpoint continues to be applied to indicate the RBM flux setpoint on the NUMAC RBM displays.
SUSQUEHANNA - UNIT 2 3.3-48
BASES APPLICABLE SAFETY
- ANALYSES, LCO, and APPLICABILITY (continued)
- 1. Rod Block Monitor (continued)
Rev. 6 Control Rod Block Instrumentation B 3.3.2.1 The calculated setpoints and applicable power ranges are bounding values. In the equipment implementation, it is necessary to apply a "deadband" to each setpoint. The deadband is applied to the RBM trip setpoint selection logic and the RBM trip automatic bypass logic such that the setpoint being applied is always equal to or more conservative than the required setpoint. Since the RBM flux trip setpoint applicable to the higher power ranges are more conservative than the corresponding trip setpoints for lower power ranges, the trip setpoint applicable to the higher power range (high power range or intermediate power range) continues to be applied when STP decreases below the lower limit of that range until STP is below the power range setpoint by a value exceeding the deadband.
Similarly, when STP decreases below the low power setpoint, the automatic bypass of RBM flux trip outputs will not be applied until STP decreases below the trip setpoint a value exceeding the deadband.
The RBM channel uses THERMAL POWER, as represented by the STP input value from its reference APRM channel, to automatically enable RBM flux trip outputs (remove the automatic bypass) and to select the RBM flux
=~===========:::tr"ii:p=setpotru::t0=be::ap:pl:ied:+/-io.we11.e£;:::t-he~BM:_l::lp_s-ea~e-ft1F1eti-efl-is--ooni--* -----
reqI uiredd tofi bedC?PtEhRACBOLLERwdhen thde_ MCPRthvaTluHesERarMeAleLsspO thWanEtRheI I
va ues e me in e
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eve.
Therefore, even though the RBM Upscale Function is implemented in each RBM channel as a single trip function with a selected trip setpoint, it is characterized in Table 3.3.2.1-1 as three Functions, the Low Power Range
- Upscale Function, the Intermediate Power Range-Upscale Function, and the High Power Range - Upscale Function, to facilitate correct definition of the OPERABILITY requirements for the Functions. Each Function corresponds to one of the RBM power ranges. Due to the dead band effects on the determination of the current power range, the transition between these three Functions will occur at slightly different THERMAL POWER levels for increasing power versus decreasing power.
- 2. Rod Worth Minimizer The RWM enforces the banked position withdrawal sequence (BPWS) to ensure that the initial conditions of the CRDA analysis are not violated.
The analytical methods and assumptions used in evaluating the CRDA are summarized in References 2, 3, 4, and 5. The BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6, "Rod Pattern Control."
SUSQUEHANNA - UNIT 2 3.3-49
BASES APPLICABLE SAFETY
- ANALYSES, LCO, and APPLICABILITY (continued)
~
- 2. Rod Worth Minimizer (continued)
Rev. 6 Control Rod Block Instrumentation B 3.3.2.1 When performing a shutdown of the plant, an optional BPWS control rod sequence (Ref. 7) may be used if the coupling of each withdrawn control rod has been confirmed. The rods may be inserted without the need to stop at intermediate positions. When using the Reference 11 control rod insertion sequence for shutdown, the rod worth minimizer may be reprogrammed to enforce the requirements of the improved BPWS control rod insertion, or may be bypassed and the improved BPWS shutdown sequence implemented under the controls in Condition D.
The RWM Function satisfies Criterion 3 of the NRC Policy Statement.
(Ref. 7)
- Since the RWM is designed to act as a backup to operator control of the rod sequences, only one channel of the RWM is available and required to be OPERABLE (Ref. 6). Special circumstances provided for in the Required Action of LCO 3.1.3, "Control Rod OPERABILITY," and LCO 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control rod
=---==:::::=============tplctiatter~e_e=Withlbe--BPW:&----IMe-RWM-may-0e--byJ3assedK-a¥.-is----
required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed.
Compliance with the BPWS, and therefore OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is< 10% RTP.
When THERMAL POWER is> 10% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Refs. 4 and 6). In MODES 3 and 4, all control rods are required to be inserted into the core (except as provided in 3.10 "Special Operations"); therefore, a CRDA cannot occur. In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SOM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.
- 3. Reactor Mode Switch - Shutdown Position During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed.
The Reactor Mode Switch-=..Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.
The Reactor Mode Switch-=..Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement. (Ref. 7)
SUSQUEHANNA - UNIT 2 3.3-50
BASES APPLICABLE SAFETY
- ANALYSES, LCO, and APPLICABILITY (continued)
ACTIONS Rev. 6 Control Rod Block Instrumentation B 3.3.2.1
- 3. Reactor Mode Switch-Shutdown Position (continued)
Two channels are required to be OPERABLE to ensure that no single channel failure will preclude a rod block when required. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on reactor mode switch position.
During shutdown conditions (MODE 3, 4, or 5), no positive reactivity insertion events are analyzed because assumptions are that control rod withdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch is in the shutdown position, the control rod withdrawal bloc!< is required to be OPERABLE. During MODE 5 with the reactor mode switch in the refueling position, the refuel position one-rod-out interlock (LCO 3.9.2) provides the required control rod withdrawal blocks.
A.1 With one RBM channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod block function; however, overall
=*==============reJ~
tta" ~-edtte~rl::be.eat1se-a~§le-f-aHttre-ifl-the-r-emairufl§::9PERABl::E~---
channel can resultin no control rod block capability for the RBM. For this reason, Required Action A.1 requires restoration of the inoperable channel to OPERABLE status. The Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on the low probability of an event occurring coincident with a failure in the remaining OPERABLE channel. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program. However, because Function 3, Reactor Mode Switch -
Shutdown Position, is only applicable in MODES 3, 4, and 5, the Risk Informed Completion Time Program may not be entered for inoperable channel(s) of Function 3.
B.1 If Required Action A.1 is not met and the associated Completion Time has expired, the inoperable channel must be placed in trip within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. If both RBM channels are inoperable, the RBM is not capable of performing its intended function; thus, one channel must also be placed in trip. This initiates a control rod withdrawal block, thereby ensuring that the RBM function is met.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities and is acceptable because it minimizes risk while allowing time for restoration or tripping of inoperable channels.
SUSQUEHANNA - UNIT 2 3.3-51
BASES ACTIONS (continued)
C.1, C.2.1.1, C.2.1.2, and C.2.2 Rev.6 Control Rod Block Instrumentation B 3.3.2.1 With the RWM inoperable during a reactor startup, the operator is still capable of enforcing the prescribed control rod sequence. However, the overall reliability is reduced because a single operator error can result in violating the control rod sequence. Therefore, control rod movement must be immediately suspended except by scram. Alternatively, startup may continue if at least 12 control rods have already been withdrawn, or a reactor startup with an inoperable RWM was not performed in the last calendar year, i.e. the last 12 months. Required Actions C.2.1.1 and C.2.1.2 require verification of these conditions by review of plant logs and control room indications. A reactor startup with an inoperable RWM is defined as rod withdrawal during startup when the RWM is required to be OPERABLE. Once Required Action C.2.1.1 or C.2.1.2 is satisfactorily completed, control rod withdrawal may proceed in accordance with the restrictions imposed by Required Action C.2.2. Required Action C.2.2 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff. The RWM may be bypassed under
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Actions of LCO 3.1.3 and LCO 3.1.6 may require bypassing the RWM, during which time the RWM must be considered inoperable with Condition Centered and its Required Actions taken.
Required Action C.2.2 is modified by a footnote which still allows for manual performance of the RWM function and requires a verification of compliance with the analyzed rod position sequence by a second licensed operator. The analyzed rod position sequence shall be established consistent with Ref. 15, and may or may not be in compliance with the BPWS. The analyzed rod position sequence will ensure that all licensing requirements continue to be met with respect to the CRDA analyses. This is a temporary allowance and applicably only during Cycle
- 21. Upon completion of Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
D.1 With the RWM inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence. Required Action D.1 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff. The RWM may be bypassed under these conditions to allow the reactor shutdown to continue.
SUSQUEHANNA - UNIT 2 3.3-52
BASES ACTIONS (continued)
D.1 (continued)
Rev. 6 Control Rod Block Instrumentation B 3.3.2.1 Required Action D.1 is modified by a footnote which still allows for manual performance of the RWM function and requires a verification of compliance with the analyzed rod position sequence by a second licensed operator. The analyzed rod position sequence shall be established consistent with Ref. 15, and may or may not be in compliance with the BPWS. The analyzed rod position sequence will ensure that all licensing requirements continue to be met with respect to the CRDA analyses. This is a temporary allowance and applicably only during Cycle
- 21. Upon completion of Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
E.1 and E.2 With one Reactor Mode Switch-Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod withdrawal block function. However, since the Required Actions are consistent with the normal action of an OPERABLE Reactor Mode Switch-Shutdown Position Function (i.e., maintaining all
=-~;============:::£c:eooiifiifiIEel::mas--iAseFtea}d-here is no distinctiOA--BetweeR---Ravi-At:OR&-0F-two-----
channels inoperable.
SURVEILLANCE REQUIREMENTS In both cases (one or both channels inoperable), suspending all control rod withdrawal and initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies will ensure that the core is subcritical with adequate SOM ensured by LCO 3.1.1. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are therefore not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.
As noted at the beginning of the SRs, the SRs for each Control Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1.
The Surveillances are modified by a Note to indicate that when an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains control rod block capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis Refs. 9, 12, and 13 assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> testing allowance SUSQUEHANNA-UNIT 2 3.3-53
BASES SURVEILLANCE REQUIREMENTS (continued)
Rev. 6 Control Rod Block Instrumentation B 3.3.2.1 does not significantly reduce the probability that a control rod block will be initiated when necessary.
SR 3.3.2.1.1 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. It includes the Reactor Manual Control Multiplexing System input. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.2.1.2 and SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function: The CHANNEL FUNCTIONAL TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs and by verifying proper indication of the selection error of at least one out-of-sequence control rod. As noted in the SRs, SR 3.3.2.1.2 is not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after any control
=-~============::ffl-: fl::is::witlwr-awfr:ie::M_8.0E2,:=As netee,----SR3/4-3-+/-.-4-;-3-is-Ret-rea1:1ir-eEl-te-0-----
performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after THERMAL POWER is ::;; 10% RTP in MODE 1.
This allows entry into MODE 2 for SR 3.3.2.1.2, and entry into MODE 1 when THERMAL POWER is ::;; 10% RTP for SR 3.3.2.1.3, to perform the required Surveillance if the Frequency is not met per SR 3.0.2. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowance is based on operating e)(perience and in*consideration of providing a reasonable time in which to complete the SRs. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.2.1.4 The RBM setpoints are automatically varied as a function of Simulated Thermal Power. Three control rod block Allowable Values are specified in Table 3.3.2.1-1, each within a specific power range. The power at which the control rod bloc!< Allowable Values automatically change are based on the APRM signal's input to each RBM channel. Below the minimum power setpoint, the RBM is automatically bypassed. These control rod block NTSPs must be verified periodically to be less than or equal to the specified Allowable Values. If any power range setpoint is non-conservative, then the affected RBM channel is considered inoperable. As noted, neutron detectors are excluded from the Surveillance because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.3 and SR 3.3.1.1.B. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SUSQUEHANNA - UNIT 2 3.3-54
BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.1.5 Rev. 6 Control Rod Block Instrumentation B 3.3.2.1 The RWM is automatically bypassed when power is above a specified value. The power level is determined from steam flow signals. The automatic bypass setpoint must be verified periodically to be not bypassed
~ 10% RTP. This is performed by a Functional check. If the RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power setpoint channel can be placed in the conservative condition (nonbypass). If placed in the nonbypassed condition, the SR is met and the RWM is not considered inoperable. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.2.1.6 A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch - Shutdown Position Function to ensure that the entire channel will perform the intended function. The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch - Shutdown Position Function is performed by attempting to withdraw any control rod with the reactor mode switch in the
:::::::::::::=============s*.tl: t1tdown....--posittori::a11el::1Lerifyiftg::a:.e_C:ll:l!reJi.-e.fl::blt'l_e.l~:f>...eet1tr:r-s~.=================
As noted in the SR, the Surveillance is not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links.
This allows entry into MODES 3 and 4 if the Frequency is not met per SR 3.0.2. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs.
The Surveillance Frequency _is controlled under the Surveillance Frequency Control Program.
SR 3.3.2.1.7 CHANNEL CALIBRATION is a test that verifies the channel responds to the measured parameter with the necessary range and accuracy.
CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibration consistent with the plant specific setpoint methodology.
As noted, neutron detectors are e>Ccluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8.
SUSQUEHANNA - UNIT 2 3.3-54a
BASES SURVEILLANCE REQUIREMENTS
{continued)
SR 3.3.2.1.7 {continued)
Rev.6 Control Rod Block Instrumentation 8 3.3.2.1 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.2.1. 7 for the RBM Functions is modified by two Notes as identified in Table 3.3.2.1-1. The RBM Functions are Functions that are LSSSs for reactor core Safety Limits. The first Note requires evaluation of channel performance for the condition where the as-found setting for the channel setpoint is not the NTSP but is conservative with respect to the Allowable Value. For digital channel components, no as-found tolerance or as-left tolerance can be specified. Evaluation of instrument performance will verify that the instrument will continue to behave in accordance with design-basis assumptions. The purpose of the assessment is to ensure confidence in the instrument performance prior to returning the instrument to service.
These channels will also be identified in the Corrective Action Program.
Entry into the Corrective Action Program will ensure required review and documentation of the condition for continued OPERABILITY. The second Note requires that the as-left setting for the instrument be returned to the NTSP. If the as-left instrument setting cannot be returned to the NTSP, tbeA:--tbe-iflStfl:lfflem-channel shaH--be-aeelarea---iooperael&.---+Ae-seee-nd-----
Note also requires that the NTSP and NTSP methodology are to be contained in a document controlled by 10 CFR 50.59.
, SR 3.3.2.1.8 The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors are possible.
SR 3.3.2.1.8 is modified by a footnote which allows verification of the control rod sequence against the analyzed rod position sequence in lieu of the BPWS prior to declaring the RWM operable following loading of the sequence into the RWM. The analyzed rod position sequence shall be established consistent with Ref. 15, and may or may not be in compliance with the BPWS. The analyzed rod position sequence will ensure that all licensing requirements continue to be met with respect to the CRDA analyses. This is a temporary allowance and applicably only during Cycle
- 21. Upon completion of Cycle 21, this temporary allowance is no longer applicable and will expire on April 15, 2023.
SUSQUEHANNA - UNIT 2 3.3-54b
BASES REFERENCES
- 1.
FSAR, Section 7.7.1.2.8.
- 2.
FSAR, Section 7.6.1.a.5.7 Rev.6 Control Rod Block Instrumentation B 3.3.2.1
- 3.
NEDE-24011-P-A-9-US, "General Electrical Standard Application for Reload Fuel," Supplement for United States, Section S 2.2.3.1, September 1988.
- 4.
"Modifications to the Requirements for Control Rod Drop Accident Mitigating Systems," BWR Owners' Group, July 1986.
- 5.
NEDO-21231, "Banked Position Withdrawal Sequence,"
January 1977.
- 6.
NRC SER, "Acceptance of Referencing of Licensing Topical Report NEDE-24011-P-A," "General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27, 1987.
- 7.
Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 32193)
- 8.
NEDC-30851-P-A, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation," October 1988.
- 9.
GENE-770-06-1, "Addendum to Bases for changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation, Technical Specifications," February 1991.
- 10.
FSAR, Section 15.4.2.
- 11.
NEDO 33091-A, Revision 2, "Improved BPWS Control Rod Insertion Process," July 2004.
- 12.
NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function," October 1995. *
- 13.
NEDC-3241 OP-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option Ill Stability Trip Function," November 1997.
- 14.
XN-NF-90-1 0(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.
SUSQUEHANNA-UNIT 2 3.3-54c
BASES REFERENCES (continued)
Rev. 6 Control Rod Bloc!< Instrumentation B 3.3.2.1
- 15. ANP-10333P-A, "AURORA-B: An Evaluation Model for Boiling Water Reactors; Application to Control Rod Drop Accident (CRDA)" (as identified in the COLR).
SUSQUEHANNA-UNIT 2 3.3-54d