U-603708, Transmittal of Revision 9 to the Clinton Power Station Technical Specification Bases

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Transmittal of Revision 9 to the Clinton Power Station Technical Specification Bases
ML050110247
Person / Time
Site: Clinton Constellation icon.png
Issue date: 01/03/2005
From: Lliff W
AmerGen Energy Co, Executive Court Reports
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
10CFR50.36, U-603708
Download: ML050110247 (25)
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Text

AmerGen.

An Exelon Company Clinton Power Station R.R. 3, Box 228 Clinton, IL 61727 10CFR50.36 U-603708 January 3, 2005 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Clinton Power Station, Unit 1 Facility Operating License NRC Docket No. 50-461

Subject:

Transmittal of Revision 9 to the Clinton Power Station Technical Specification Bases In accordance with Clinton Power Station (CPS) Technical Specification 5.5.11, "Technical Specification (TS) Bases Control Program," AmerGen Energy Company (AmerGen), LLC is transmitting the revised pages constituting Revision 9 to the CPS TS Bases. The changes associated with this revision were processed in accordance with CPS TS 5.5.11, which became effective with Amendment No. 95 to the CPS Operating License. Compliance with CPS TS 5.5.11 requires updates to the TS Bases to be submitted to the NRC at a frequency consistent with 10CFR50.71, "Maintenance of records, making of reports," paragraph (e).

Should you have any questions concerning this information, please contact Mr. Jim Peterson at (217) 937-2810.

Respectfully, William S. Iliff (° Regulatory Assurance Manager Clinton Power Station bIf/JLP - Revision 9 to the CPS Technical Specification Bases - Summary of Commitments cc: Regional Administrator - NRC Region IlIl NRC Senior Resident Inspector - Clinton Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

Attachment 1 to U-603708 Clinton Power Station, Unit 1 Revision 9 to the CPS Technical Specification Bases B 3.1-23 B 3.1-38 B 3.1-39 B 3.2-1 B 3.2-2 B 3.2-3 B 3.2-4 B 3.2-4a B 3.2-9 B 3.2-1 0 B 3.2-11 B 3.2-12 B 3.3-65 B 3.3-144 B 3.3-163 B 3.4-3 B 3.4-4 B 3.4-8 B 3.4-58 B 3.6-105a B 3.7-16 B 3.8-68b

Control Rod Scram Times B 3.1.4 BASES APPLICABLE The scram function of the CRD System protects the MCPR SAFETY ANALYSES Safety Limit (SL) (see Bases for SL 2.1.1, "Reactor Core (continued) SLs," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"),

and the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," and LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded. Above 950 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL during the analyzed limiting power transient. Below 950 psig, the scram function is assumed to perform during the control rod drop accident (Ref. 7) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "Control Rod Pattern"). For the reactor vessel overpressure protection analysis, the scram function, along with the safety/relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

Control rod scram times satisfy Criterion 3 of the NRC Policy Statement.

LCO The scram times specified in Table 3.1.4-1 (in the accompanying LCO) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met.

To account for single failure and "slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis. The scram times have a margin to allow up to 7.5% of the control rods (i.e., 145 x 7.5% = 10) to have scram times that exceed the specified limits (i.e., "slow" control rods) assuming a single stuck control rod (as allowed by LCO 3.1.3, "Control Rod OPERABILITY") and an additional control rod failing to scram per the single failure criterion. The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes ("pickup") when the index tube passes a specific location and then opens

("dropout") as the index tube travels upward. Verification of the specified scram times in Table 3.1.4-1 is I accomplished through measurement of the "dropout" times.

(continued)

CLINTON B 3.1-23 Revision No. 9-5

SLC System B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Standby Liquid Control (SLC) System BASES BACKGROUND The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive xenon free state without taking credit for control rod movement. The SLC System satisfies the requirements of 10 CFR 50.62 (Ref. 1) on anticipated transient without scram (ATWS).

The SLC System consists of a boron solution storage tank, two positive displacement pumps, two explosive valves, which are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The preferred flow path of the boron neutron absorber solution to the reactor vessel is by the High Pressure Core Spray (HPCS) System sparger. The SLC piping is connected to the HPCS System just downstream of the HPCS manual injection isolation valve. An alternate flow path to the reactor vessel is provided by the SLC sparger near the bottom of the core shroud. This flow path is normally locked out of service by the SLC manual injection valve.

APPLICABLE The SLC System is manually initiated from the main control SAFETY ANALYSES room, as directed by the emergency operating procedures, if the operator believes the reactor cannot be shut down, or kept shut down, with the control rods. The SLC System is used in the event that not enough control rods can be inserted to accomplish shutdown and cooldown in the normal manner. The SLC System injects borated water into the reactor core to compensate for all of the various reactivity effects that could occur during plant operation. To meet this objective, it is necessary to inject a quantity of boron that produces a concentration equivalent to at least 1000 ppm of natural boron in the reactor core at 680F. This is accomplished by the use of enriched boron (2 30 atom%

boron 10). To allow for potential leakage and imperfect mixing in the reactor system, an additional amount of boron equal to 25% of the amount cited above is added (Ref. 2).

The concentration versus volume limits in Figure 3.1.7-1 are calculated such that the required concentration is achieved accounting for dilution in the RPV with normal water level (continued)

CLINTON B 3.1-38 Revision No. 9-4

SLC System B 3.1.7 BASES I APPLICABLE and including the water volume in the residual heat removal SAFETY ANALYSES shutdown cooling piping and in the recirculation loop (continued) piping. This quantity of borated solution is the amount that is above the storage tank level instrument zero. (The instrument zero is based on ensuring sufficient net positive suction head and includes additional margin to preclude air entrainment in the pump suction piping due to vortexing during two pump operation.)

The SLC System satisfies the requirements of the NRC Policy Statement because operating experience and probabilistic risk assessment have generally shown it to be important to public health and safety.

LCO The OPERABILITY of the SLC System provides backup capability for reactivity control, independent of normal reactivity control provisions provided by the control rods. The OPERABILITY of the SLC System is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the RPV, including the OPERABILITY of the pumps and valves. Two SLC subsystems are required to be OPERABLE, each containing an OPERABLE pump, an explosive valve and associated piping, valves, and instruments and controls to ensure an OPERABLE flow path.

APPLICABILITY In MODES 1 and 2, shutdown capability is required. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in Shutdown and a control rod block is applied. This provides adequate controls to ensure the reactor remains subcritical. In MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Demonstration of adequate SDM (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") ensures that the reactor will not become critical. Therefore, the SLC System is not required to be OPERABLE during these conditions, when only a single control rod can be withdrawn.

ACTIONS A.1 If one SLC subsystem is inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE subsystem is adequate to perform the shutdown function. However, the overall reliability is reduced because a single failure in the (continued)

CLINTON B 3.1-39 Revision No. 9-4

APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

BASES BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on I the APLHGR are specified to ensure that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES Design Basis Accidents (DBAs) and normal operations that determine APLHGR limits are presented in USAR Chapters 6 and 15, and in References 1, 2, 3, and 4.

APLHGR limits are developed as a function of exposure to ensure adherence to 10 CFR 50.46 during a LOCA (Refs. 3 and 4).

(continued)

CLINTON B 3.2-1 Revision No. 9-3

APLHGR B 3.2.1 BASES APPLICABLE LOCA analyses are performed to ensure that the above SAFETY ANALYSES determined APLHGR limits are adequate to meet the PCT and (continued) maximum oxidation limits of 10 CFR 50.46. The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. A complete discussion of the analysis code is provided in Reference 5.

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 an assembly.

Clinton has no additional MAPLHGR setdown requirements in the LOCA evaluation to achieve acceptable ECCS-LOCA performance. The LHGRFAC multipliers of Section 3.2.3 are sufficient to provide adequate protection for the off-rated conditions from an ECCS-LOCA perspective.

For single recirculation loop operation, the APLHGR limit I must be multiplied by the MAPLHGR single loop multiplier, which is specified in the COLR. This multiplier is due to the conservative analysis assumption of an earlier departure from nucleate boiling with one recirculation loop available, resulting in a more severe cladding heatup during a LOCA.

The APLHGR satisfies Criterion 2 of the NRC Policy Statement.

LCO The APLHGR limits specified in the COLR are the result of fuel design and LOCA analyses. For two recirculation loops I operating, the limit is determined by the exposure dependent APLHGR limits. With only one (continued)

CLINTON B 3.2-2 Revision No. 9-3

APLHGR B 3.2.1 I .

BASES LCO recirculation loop in operation, in conformance with the (continued) requirements of LCO 3.4.1, "Recirculation Loops Operating,"

the limit is determined by multiplying the exposure I dependent APLHGR limit by the MAPLHGR single loop multiplier value specified for single recirculation loop operation in the COLR, which has been determined by a specific single I recirculation loop analysis (Ref. 3).

APPLICABILITY The APLHGR limits are primarily derived from fuel design I evaluations and 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. This trend continues down to the power range of 5% to 15% RTP when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor (IRM) scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2. Therefore, at THERMAL POWER levels

< 21.6% RTP, the reactor operates with substantial margin to the APLHGR limits; thus, this LCO is not required.

ACTIONS A.1 If any APLHGR exceeds the required limit, an assumption I regarding an initial condition of the LOCA analyses may not be met. Therefore, prompt action is taken to restore the APLHGR(s) to within the required limit(s) such that the plant will be operated within analyzed conditions and within the design limits of the fuel rods. 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 limit and is acceptable based on the low probability of a I LOCA occurring simultaneously with the APLHGR out of specification.

B.1 If the APLHGR cannot be restored to within its required limit 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 (continued)

CLINTON B 3.2-3 Revision No. 9-3

APLHGR B 3.2.1 BASES ACTIONS B.1 (continued)

POWER must be reduced to < 21.6% 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 < 21.6% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.1.1 REQUIREMENTS APLHGRs are required to be initially calculated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is 2 21.6% RTP and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. They 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 /> Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution under normal conditions. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER 2 21.6% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.

With regard to APLHGR values obtained pursuant to this SR, as determined from plant indication instrumentation, the specified limit is considered to be a nominal value and therefore does not require compensation for instrument I indication uncertainties (Ref. 6).

(continued)

CLINTON .B 3.2-4 Revision No. 9-3

APLHGR B 3.2.1 BASES REFERENCES 1. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel, GESTAR-II" (latest approved revision).

2. USAR, Chapter 15, Section 15.0.
3. USAR, Chapter 15, Appendix 15B.
4. USAR, Chapter 15, Appendix 15C.

I 5. NEDE-20566, "General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR5O Appendix K," November 1975.

I 6. Calculation IP-0-0002.

CLINTON B 3.2-4a Revision No. 9-3

LHGR B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

BASES BACKGROUND The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location. Limits on the LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, including anticipated operational occurrences (AOOs), and to ensure that the peak clad temperature (PCT) during a postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46. 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 anticipated operating conditions identified in I USAR Chapters 4 and 15.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the fuel system design are presented in USAR Chapter 4. The analytical methods and assumptions used in evaluating AOOs and normal operation that determine the LHGR limits are presented in References 1 and 2. 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 the guidelines of 10 CFR, Parts 20, 50, and 100. 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 U0 2 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 (Refs. 1 and 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. The analysis also (continued)

CLINTON B 3.2-9 Revision No. 9-3

LHGR B 3.2.3 BASES APPLICABLE includes allowances for short term transient operation above SAFETY ANALYSES the operating limit to account for AOOs, plus an allowance (continued) for densification power spiking.

LHGR limits are developed as a function of exposure and the various operating core flow and power states to ensure adherence to fuel design limits during the limiting AQOs (Ref. 2). Flow dependent LHGR limits are determined using the three dimensional BWR simulator code (Ref. 1) to analyze slow flow runout transients. The flow dependent multiplier, LHGRFACf, is dependent on the maximum core flow runout capability. LHGRFACf curves are provided based on the maximum credible flow runout transient. The result of a single failure or single operator error is the runout of only one loop because both recirculation loops are under independent control.

Based on analyses of limiting plant transients (other than core flow increases) over range of power and flow conditions, power dependent multipliers, LHGRFAC., are also generated. Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which turbine control valve fast closure scram signals are bypassed, both high and low core flow LHGRFACp limits are provided for operation at power levels between 21.6t RTP and the previously mentioned bypass power level. The exposure dependent LHGR limits are reduced by LHGRFACf and LHGRFACp at various operating conditions to ensure that all fuel design criteria are met for normal operation and AOOs. A complete discussion of the analysis code is provided in Reference 5.

The LHGRFAC multipliers are sufficient to provide adequate protection for the off-rated conditions from an ECCS-LOCA analysis perspective.

For single recirculation loop operation, the LHGRFAC multiplier is limited to a maximum value that is specified in the COLR. This maximum limit is due to the conservative analysis assumption of an earlier departure from nucleate boiling with one recirculation loop available, resulting in a more severe cladding heatup during a LOCA.

The LHGR satisfies Criterion 2 of the NRC Policy Statement.

LCO The LHGR is a basic assumption in the fuel design analysis.

The fuel has been designed to operate at rated core power I with sufficient design margin to the LHGR limit calculated to cause a It fuel cladding plastic strain. The operating limit to accomplish this objective is specified in the COLR.

The LHGR limits specified in the COLR are the result of fuel design and transient analyses. For two recirculation loops operating, the limit is determined by multiplying the smaller of the LHGRFACf and LHGRFACp factors times the exposure dependent LHGR limits.

(continued)

CLINTON B 3.2-10 Revision No. 9-3

LHGR B 3.2.3 BASES LCO With only one recirculation loop in operation, in (continued) conformance with the requirements of LCO 3.4.1, "Recirculation Loops Operating," the limit is determined by multiplying the exposure dependent LHGR limit by the smaller of either LHGRFACf, LHGRFACp, and the LHGR single loop operation multiplier, where the single loop operation multiplier has been determined by a specific single recirculation loop analysis (Refs. 6 and 7).

APPLICABILITY The LHGR limits are derived from fuel design analysis that is limiting at high power level conditions. At core thermal power levels < 21.6% 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 2 21.6% RTP.

ACTIONS A.I 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 limit(s) such that the plant is operating within analyzed conditions and within the design limits of the fuel rods. 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 limit and is acceptable based on the I low probability of a transient occurring simultaneously with the LHGR out of specification.

B.l If the LHGR cannot be restored to within its required limit 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 < 21.6% 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 < 21.6% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.3.1 REQUIREMENTS The LHGRs are required to be initially calculated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is 2 21.6% RTP and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. They are compared with 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 /> Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution I during normal conditions. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER 2 21.6t RTP is achieved is acceptable given the large inherent margin to operating limits at lower power levels.

(continued)

CLINTON B 3.2-11 Revision No. 9-3

LHGR B 3.2.3 BASES SURVEILLANCE SR 3.2.3.1 REQUIREMENTS (continued) With regard to LHGR values obtained pursuant to this SR, as determined from plant indication instrumentation, the specified limit is considered to be a nominal value and therefore does not require compensation for instrument indication uncertainties (Ref. 4).

REFERENCES 1. NEDE-24011-P-A, "General Electric Standard Application for Reactor-Fuel, GESTAR-II," (latest approved revision).

2. USAR, Section 15.0.
3. NUREG-0800, "Standard Review Plan," Section 4.2, II.A.2(g), Revision 2, July 1981.
4. Calculation IP-0-0002.
5. NEDO-24154-A, "Qualification of the One Dimensional I Core Transient Model for Boiling Water Reactors,"

August 1986.

6. USAR, Chapter 15, Appendix 15B
7. "Clinton Power Station SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-32945P, June 2000.

CLINTON B 3.2-12 Revision No. 9-3

EOC-RPT Instrumentation B 3.3.4.1 B 3.3 INSTRUMENTATION B 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation BASES BACKGROUND The EOC-RPT instrumentation initiates a recirculation pump trip (RPT) to reduce the peak reactor pressure and power resulting from turbine trip or generator load rejection transients to provide additional margin to core thermal MCPR Safety Limits (SLs).

The need for the additional negative reactivity in excess of that normally inserted on a scram reflects end of cycle reactivity considerations. Flux shapes at the end of cycle are such that the control rods may not be able to ensure that thermal limits are maintained by inserting sufficient negative reactivity during the first few feet of rod travel upon a scram caused by Turbine Control Valve (TCV) Fast Closure, Trip Oil Pressure-Low, or Turbine Stop Valve (TSV)

Closure. The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity at a faster rate than the control rods can add negative reactivity.

The EOC-RPT instrumentation as described in Reference 1 is comprised of sensors that detect initiation of closure of the TSVs, or fast closure of the TCVs, combined with logic circuits, load drivers, and fast acting circuit breakers that interrupt the fast speed power supply to each of the recirculation pump motors. The channels consist of pressure

  • switches and limit switches. When the setpoint is exceeded, the switch closes which then inputs a signal to the EOC-RPT trip logic. Actuation of the EOC-RPT system causes each division of the RPS to energize a trip coil in its associated RPT breaker. When the EOC-RPT breakers trip open, the recirculation pumps downshift to slow speed.

The EOC-RPT system is a two-out-of-four logic for each Function; thus, either two TSV Closure or two TCV Fast Closure, Trip Oil Pressure-Low signals are required to actuate tripping both recirculation pumps from fast speed operation. There are two EOC-RPT breakers in series per recirculation pump. A trip in Division 1 (or 4) will cause a trip of the 'A' recirculation pump. A trip in Division 2 (or 3) will cause a trip of the 'B' recirculation pump.

Both EOC-RPT breakers for each recirculation pump trip upon actuation of the EOC-RPT system. Placing an EOC-RPT bypass switch in "bypass" will allow the EOC-RPT trip capability to be maintained, however, an additional single failure cannot be accommodated (refer to Required Action B.1 Bases).

(continued)

CLINTON B 3.3-65 Revision No. 9-8

Primary Containment and Drywell Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 2. Primary Containment and Drywell Isolation SAFETY ANALYSES, LCO, and 2.a, and 2.e. Reactor Vessel Water Level-Low Low, APPLICABILITY Level 2 (continued)

Low RPV water level indicates the capability to cool the fuel may be threatened. The valves whose penetrations communicate with the primary containment are isolated to limit the release of fission products. The isolation of the primary containment on Level 2 supports actions to ensure that offsite dose limits of 10 CFR 100 are not exceeded.

The Reactor Vessel Water Level-Low Low, Level 2 Function associated with isolation is implicitly assumed in the USAR analysis as these leakage paths are assumed to be isolated post LOCA. In addition, Function 2.a provides an isolation signal to certain drywell isolation valves. The isolation of drywell isolation valves, in combination with other accident mitigation systems, functions to ensure that steam and water releases to the drywell are channeled to the suppression pool to maintain the pressure suppression function of the drywell.

In addition to providing automatic isolation capability for primary containment and drywell isolation valves, Function 2.a provides signals for automatic actuation of the Division 1 and 2 SX subsystems, including automatic start of the Division 1 and 2 SX pumps and automatic actuation of the associated subsystem isolation valves (as required to support automatic operation of the SX subsystems). The equipment involved with the SX subsystems is described in LCO 3.7.1, "Division 1 and 2 SX Subsystems."

Reactor Vessel Water Level-Low Low, Level 2 signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level-Low Low, Level 2 Function are available and are required to be OPERABLE to ensure no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level-Low Low, Level 2 Allowable Value was chosen to be the same as the ECCS Reactor Vessel Water Level-Low Low, Level 2 Allowable Value (LCO 3.3.5.1),

(continued)

CLINTON B 3.3-144 Revision No. 9-7

i Primary Containment and Drywell Isolation Instrumentation B 3.3.6.1 BASES ACTIONS A.1 (continued) inoperable channel and the Function still maintains isolation capability (refer to Required Action B.1 and C.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.l. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate two failures, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an undesired isolation), Condition F must be entered and its Required Action taken.

B.1 Condition B exists when, for any one or more MSL isolation Functions, two required channels are inoperable. (For example, a failure of a coincidence logic card (i.e., a two-out-of-four logic card) in one division may affect two channels). In this condition, the MSL isolation system still maintains isolation capability for that Function, but cannot accommodate an additional single failure in that Function.

Required Action B.1 limits the time the MSL isolation logic for any Function would not accommodate a single failure.

Within the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the associated Function will have at least one inoperable channel in trip. Completing this Required Action restores the MSL isolation system to an equivalent reliability level as that evaluated in Reference 6, which justified a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> allowable out of service time.

The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of MSL isolation. Placing one of the two inoperable channels in trip satisfies both Required Actions A.1 and B.1 for that Function. If one channel is already in trip for the Function when a second channel is determined to be inoperable, Required Action B.1 is met by the one channel already in trip for that Function and no additional action is required.

Alternately, if it is not desired to place one inoperable channel in trip (e.g., as in the case where placing the (continued)

CLINTON B 3.3.-163 Revision No. 9-8

Recirculation Loops Operating B 3.4.1 BASES APPLICABLE operating at the lower flow rate), a small mismatch has been SAFETY ANALYSES determined to be acceptable based on engineering judgement.

(continued) The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational occurrences (AOOs)

(Ref. 2), which are analyzed in Chapter 15 of the USAR.

A plant specific LOCA analysis has been performed assuming only one operating recirculation loop. This analysis has 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 the APLHGR requirements are modified accordingly (Refs. 3 and 7).

The transient analyses of Chapter 15 of the USAR have also been performed for single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR, APLHGR and LHGR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System 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 MCPR, APLHGR and LHGR limits for single loop operation are specified in the COLR. The APRM flow biased simulated thermal power setpoint is in LCO 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation."

Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement.

LCO Two recirculation loops are normally 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. In addition, the total core flow must be 2 45% of rated core flow or total core flow expressed as a function of THERMAL POWER must be in Region C as identified in Figure 3.4.1-1, "THERMAL POWER/Core Flow Stability Regions." Alternatively, with only one recirculation loop in operation, THERMAL POWER must be S 58% RTP, total core flow must be 2 45% of rated (continued)

CLINTON B 3.4-3 Revision No. 9-3

Recirculation Loops Operating B 3.4.1 BASES LCO core flow or total core flow expressed as a function of (continued) THERMAL POWER must be in Region C of Figure 3.4.1-1, and modifications to the required APLHGR limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"),

LHGR limits (LCO 3.2.3, "LINEAR HEAT GENERATION RATE l (LHGR)"), and APRM Flow Biased Simulated Thermal Power-High setpoint (LCO 3.3.1.1) must be applied to allow continued operation consistent with the assumptions of References 3 and 7.

The LCO is modified by a Note which allows up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> before having to put in effect the required modifications to required limits and setpoints after a change in the reactor operating conditions from two recirculation loops operating to single recirculation loop operation. If the required limits and setpoints are not in compliance with the applicable requirements at the end of this period, the associated equipment must be declared inoperable or the limits "not satisfied," and the ACTIONS required by nonconformance with the applicable Specifications implemented. This time is provided due to the need to stabilize operation with one recirculation loop, including the procedural steps necessary to limit flow (to less than the volumetric recirculation loop flow) in the operating loop, monitor for excessive APRM and local power range monitor (LPRM) neutron flux noise levels; and the complexity and detail required to fully implement and confirm the required limit and setpoint modifications.

APPLICABILITY 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 I AOOs and LOCAs are assumed to occur.

I In MODES 3, 4, and 5, the consequences of an accident or AOO are reduced and the coastdown characteristics of the recirculation loops are not important.

ACTIONS A.1 With both recirculation loops operating but the flows not matched, the recirculation loops must be restored to operation with matched flows within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. If the flow mismatch cannot be restored to within limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, one recirculation loop must be shut down.

(continued)

CLINTON B 3.4-4 Revision No. 9-3

Recirculation Loops Operating B 3.4.1 BASES SURVEILLANCE SR 3.4.1.2 REQUIREMENTS This SR ensures the reactor THERMAL POWER and core flows are within appropriate parameter limits to prevent uncontrolled power oscillations. At low recirculation flows and high reactor power, the reactor exhibits increased susceptibility to thermal hydraulic instability. Actions have been developed based on the guidance provided in References 4 and 5 to respond to operation in these conditions. This SR identifies when the conditions requiring these actions are necessary. The Frequency is based on operating experience and the operators' inherent knowledge of reactor status, including significant changes in THERMAL POWER and core flow.

With regard to THERMAL POWER and core flow values obtained pursuant to this SR, as read from plant indication instrumentation, the specified limit is considered to be a nominal value and therefore does not require compensation for instrument indication uncertainties (Ref. 6).

REFERENCES 1. USAR, Section 6.3.3.7.

2. USAR, Section 5.4.1.1.
3. USAR, Chapter 15, Appendix 15B.
4. NRC Bulletin 88-07, Supplement 1, "Power Oscillations in Boiling Water Reactors," December 1988.
5. GE Letter, "Interim Recommendations for Stability Actions," November 1988.
6. Calculation IP-0-0029.
7. "Clinton Power Station SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-32945P, June 2000 CLINTON B 3.4-8 Revision No. 9-3

RCS P/T Limits B 3.4.11 BASES ACTIONS C.1 and C.2 (continued)

Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored.

Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 200 0F. Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components.

ASME Section XI, Appendix E (Ref. 6), may be used to support the evaluation; however, its use is restricted to evaluation of the beltline.

SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Verification that operation is within limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available-to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction of minor deviations.

Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.

This SR has been modified by a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and inservice leakage and hydrostatic testing.

(continued)

CLINTON B 3.4-58 Revision No. 9-1

Drywell B 3.6.5.1 BASES SURVEILLANCE SR 3.6.5.1.3 REQUIREMENTS The analyses in Reference 1 are based on a maximum drywell bypass leakage. This Surveillance ensures that the actual drywell bypass leakage is less than or equal to the acceptable A/6 idesign value of 1.0 ft2 assumed in the safety analysis. As left drywell bypass leakage, prior to the first startup after performing a required drywell bypass leakage test, is required to be < 10 of the drywell bypass leakage limit. At all other times between required drywell leakage rate tests, the acceptance criteria is based on the design Al/k . At the design A/Vkthe containment temperature and pressurization response are bounded by the assumptions of the safety analysis. One drywell air lock door is left open during each drywell bypass leakage test such that each drywell air lock door is leak tested during at least every other drywell bypass leakage test. This ensures that the leakage through the drywell air lock is properly accounted for in the measured bypass leakage and that each air lock door is tested periodically.

This Surveillance is performed at least once every 10 years (120 months) on a performance based frequency. The Frequency is consistent with the difficulty of performing the test, risk of high radiation exposure, and the remote possibility that sufficient component failures will occur such that the drywell bypass leakage limit will be exceeded.

This Frequency is modified by a note that allows for a one-time deferral of this surveillance until November 23, 2008.

If during the performance of this required Surveillance the drywell bypass leakage is determined to be greater than the leakage limit, the Surveillance Frequency is increased to at least once every 48 months. If during the performance of the subsequent consecutive Surveillance the drywell bypass leakage is determined to be less than or equal to the drywell bypass leakage limit, the 10-year Frequency may be resumed. If during the performance of the subsequent consecutive Surveillance the drywell bypass leakage is determined to be greater than the drywell bypass leakage limit, the Surveillance Frequency is increased to at least once every 24 months. The 24-month Frequency must be maintained until the drywell bypass leakage is determined to (continued)

CLINTON B 3.6-105a Revision No. 9-2

Control Room Ventilation System B 3.7.3 BASES SURVEILLANCE SR 3.7.3.5 REQUIREMENTS (continued) This SR verifies the integrity of the negative pressure portions of the Control Room Ventilation System ductwork located outside the main control room habitability boundary between fan OVC04CA(B) and isolation dampers OVC03YA(B) inclusive and fire dampers OVC042YA(E), OVC042YB(F),

OVC042YC(G), and OVC042YD(H). In addition, the integrity of the recirculation filter housing flexible connection to fan OVC03A(B) must be verified. This testing ensures the unfiltered inleakage (scfm) into the main control room habitability boundary is within the analysis assumptions.

Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency.

Therefore, this Frequency is concluded to be acceptable from a reliability standpoint.

With regard to unfiltered inleakage values obtained pursuant to this SR, as read from plant indication instrumentation, the specified limit is not considered to be a nominal value with respect to instrument uncertainties. This requires additional margin to be added to the limit to compensate for instrument uncertainties, for implementation in the associated plant procedures (Ref. 12).

SR 3.7.3.6 This SR verifies the integrity of the control room enclosure and the assumed inleakage rates of potentially contaminated air. The control room positive pressure, with respect to potentially contaminated adjacent areas, is periodically tested to verify proper function of the Control Room Ventilation System. During the high radiation mode of operation, the Control Room Ventilation System is designed to slightly pressurize the control room to 2 1/8 inches water gauge positive pressure with respect to adjacent areas to prevent unfiltered inleakage. The Control Room Ventilation System is designed to maintain this positive pressure at a flow rate of

  • 3000 scfm to the control room in the high radiation mode. The Frequency of 18 months on a STAGGERED TEST BASIS is consistent with industry practice and other filtration system SRs.

With regard to control room positive pressure values obtained pursuant to this SR, as read from plant indication instrumentation, the specified limit is not considered to be a nominal value with respect to instrument uncertainties.

This requires additional margin to be added to the limit to compensate for instrument uncertainties, (Ref. 13).

With regard to control room flow rate values obtained pursuant to this SR, as read from plant indication instrumentation, the specified limit is considered to be a nominal value and therefore does not require compensation (Ref. 13j.

CLINTON B 3.7-16 Revision No. 9-6

Battery Parameters B 3.8.6 BASES SURVEILLANCE SR 3.8.6.6 (continued)

REQUIREMENTS (continued) sized to meet the assumed duty cycle loads when the battery design capacity reaches this 80% limit.

The Surveillance Frequency for this test is normally 60 months. If the battery shows degradation, or if the battery has reached 85% of its expected life, the Surveillance I Frequency is reduced to 12 months. Degradation is indicated, according to IEEE Standard 450 (Ref. 3), when the battery capacity drops by more than 10% relative to its capacity on the previously performance test or when it is

> 10% below the manufacturer's rating. These Frequencies are based on the recommendations in IEEE Standard 450 (Ref.

3).

This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DC electrical power subsystem from service, perturb the electrical distribution system, and challenge safety systems. Credit may be taken for unplanned events that satisfy the Surveillance. Examples of unplanned events may include:

1) Unexpected operational events which cause the equipment to perform the function specified by this Surveillance, for which adequate documentation of the required performance is available; and
2) Post maintenance testing that requires performance of this Surveillance in order to restore the component to OPERABLE, provided the maintenance was required, or performed in conjunction with maintenance required to maintain OPERABILITY or reliability.

REFERENCES 1. USAR, Chapter 6.

2. USAR, Chapter 15.
3. IEEE Standard 450, 1995.
4. Calculation IP-0-0123.
5. IEEE Standard 485, 1983
6. USAR, Chapter 8.

CLINTON B 3.8-68b Revision No. 9-7

W Attachment 2 to U-603708 Clinton Power Station, Unit 1 Revision 9 to the CPS Technical Specification Bases

SUMMARY

OF COMMITMENTS The following table identifies commitments made in this document. (Any other actions discussed in the submittal represent intended or planned actions. They are described to the NRC for the NRC's information and are not regulatory commitments.)

COMMITMENT COMMITTED DATE NONE N/A

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