ML061740343

Davis-Besse Nuclear Power Station Technical Specification Bases Update

ML061740343
Person / Time
Site:	Davis Besse
Issue date:	06/21/2006
From:	Bezilla M B FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3274
Download: ML061740343 (47)

Text

-FENOC IOC 5501 North State Route 2 FirstEnergy Nuclear Operating Company Oak Harbor, Ohio 43449 Mark B. Bezilla 419-321-7676 Vice President

-Nuclear Fax: 419-321-7582 Docket Number 50-346 License Number NPF-3 Serial Number 3274 June ?-I , 2006 U.S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, D.C. 20555-0001

Subject:

Davis-Besse Nuclear Power Station Technical Specification Bases Update Ladies and Gentlemen:

Davis-Besse Nuclear Power Station (DBNPS) Technical Specification 6.17, "Technical Specifications (TS) Bases Control Program," requires the FirstEnergy Nuclear Operating Company (FENOC) to submit changes made to the TS Bases implemented without NRC prior approval.

Enclosure 1 reflects the changes made to the TS Bases and implemented by FENOC without prior approval from May 21, 2004 through June 12, 2006. The last report was submitted on June 21, 2004.Attachment 1 provides a list of Technical Specification Bases changes. Attachment 2, Commitment List, identifies that there are no commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr.Gregory A. Dunn, Manager -FENOC Fleet Licensing, at (330) 315-7243.Very truly yours, Mark B. Bezilla, Vice-President

-Nue ar A0oi Docket Number 50-346 License Number NPF-3 Serial Number 3274 Page 2 of 2 MSH Attachments Enclosure cc: Regional Administrator, NRC Region III DB-1 NRC/NRR Project Manager DB-1 NRC Senior Resident Inspector Utility Radiological Safety Board, without attachments Docket Number 50-346 License Number NPF-3 Serial Number 3274 Attachment 1 Page 1 of 3 List of Technical Specification Bases Changes LAR' Affected TS Bases Section (s) Description 01-0014 3.0.5 and 3.8.2.1 Clarified the purpose and application of TS 3.0.5 when a normal or emergency power source is inoperable.

Clarified "tie breaker" as used in TS 3/4.8.2.1.

02-0003 3/4.9.7 and 3/4.9.13 Deleted TS Bases 3/4.9.7 to reflect that TS 3/4.9.7 has been deleted. Revised TS Bases 3/4.9.13 to remove information regarding low density spent fuel pool racks and the cask and transfer pits, consistent with Amendment 266.02-0006 3/4.9.2 Relocated information regarding source range neutron flux monitor visual and audible indications from Limiting Condition for Operation 3.9.2 to TS Bases, consistent with reference Amendment 269.03-0004 3/4.5.2 and 3/4.5.3 Deleted discussion of Decay Heat Pit Watertight Enclosure, consistent with Amendment 263. The associated surveillance requirement was relocated to the Updated Safety Analysis Report Technical Requirements Manual.03-0005 3/4.6.1.4 Clarified the discussion of containment maximum pressure, and revised the accident analysis description to include updated results from the containment pressure transient analysis.03-0021 3/4.3.3.6 and 3/4.6.4 Removed discussions of the combustible gas control systems, consistent with Amendment 265.04-0001 3/4.5.2 and 3/4.5.3 Added discussion of operability considerations regarding the Low Pressure Injection crossover path.04-0004 3/4.6.4 Removed discussion of the hydrogen recombiner, consistent with rule changes to 10 CFR 50.44.04-0005 3.0.2, 3.0.3, and 3.0.5 Added guidance on allowable outage time.1 TS Bases changes made under the DBNPS Bases Control Program are assigned License Amendment Request (LAR) numbers, even though the changes are not submitted to the NRC prior to issuance.

Docket Number 50-346 License Number NPF-3 Serial Number 3274 Attachment 1 Page 2 of 3 List of Technical Specification Bases Changes LAR Affected TS Bases Section (s) Description 04-0007 3/4.3.1 and 3/4.3.2 Added guidance on actions to take when the Safety Features Actuation System response time requirement can not be met.04-0008 4.0.2 Clarified application of staggered test intervals.

04-0009 3/4.3.3.6 Clarified expectations regarding the auxiliary feedwater flow rate monthly channel check.04-0010 3/4.3.1 and 3/4.3.2 Added guidance regarding actions to take if a Reactor Trip Module or Control Rod Drive Trip Breaker is inoperable.

04-0011 3/4.3.1 and 3/4.3.2 Added guidance for the condition where the Shutdown Bypass High Pressure Trip is enabled with the Control Rod Drive breakers open.04-0012 3/4.6.3 Clarified that TS 3.6.3.1 applies to all containment isolation valves, regardless of whether the valve(s) have a Safety Features Actuation Signal automatic closure feature.Provided guidance on valves with multiple functions.

04-0013 3/4.8 Added discussion regarding energization of Vital Instrument Buses and NRC guidance dated May 30, 1984 stating that Action Statement 3.8.2.1 is not invoked when the AC Vital Buses are powered from non-Class 1 E sources.04-0014 3/4.8 Added NRC guidance dated May 31, 2000 regarding operability of switchyard equipment when that equipment is vulnerable to a single point failure.04-0017 3/4.3.3.1 Specified the actions to take when one or more channels of Fuel Storage Pool Area Emergency Ventilation System Monitors or Containment Activity Monitors are inoperable.

04-0018 3/4.7.1.2 Removed incorrect discussion of the Auxiliary Feedwater Pump low-low suction pressure interlock.

Docket Number 50-346 License Number NPF-3 Serial Number 3274 Attachment 1 Page 3 of 3 List of Technical Specification Bases Changes LAR Affected TS Bases Section (s) Description 04-0021 3/4.3.3.6 Clarified which Containment Vessel Post-Accident Radiation monitors are preferred for complying with Limiting Condition for Operation 3.3.3.6.04-0022 3/4.5.2 and 3/4.5.3 Clarified that one Low Pressure Injection pump can supply the flow requirements of both High Pressure Injection Pumps, provided that the necessary flow paths are maintained.

04-0026 3/4.5.2 and 3/4.5.3 With regard to containment inspections, added the following: "As discussed in the NRC Safety Evaluation for License Amendment No.196, it is intended that visual inspections conducted in accordance with SR 4.5.2.c.2 be limited to those areas in containment where work was performed, so as to limit the potential for radiation dose to people performing the inspections." 04-0027 3/4.7.1.2 Clarified exceptions to TS 4.0.4 for certain Auxiliary Feedwater requirements for entry into Mode 3. Noted that there must be a reasonable assurance of Auxiliary Feedwater System operability prior to entry into Mode 3.05-0002 2.1 and 3/4.3 Incorporated information for Mark B-HTP fuel consistent with Amendment 274.05-0008 3/4.8 Revised Electrical Power Bases to reflect changes made to the Emergency Diesel Generator load sequencing logic and time delays.06-0001 3/4.5.4 Clarified discussion of Borated Water Storage Tank available volume Docket Number 50-346 License Number NPF-3 Serial Number 3274 Attachment 2 Page 1 of 1 COMMITMENT LIST The following list identifies those actions committed to by the Davis-Besse Nuclear Power Station (DBNPS) in this document.

Any other actions discussed in the submittal represent intended or planned actions by the DBNPS. They are described only for information and are not regulatory commitments.

Please contact Mr. Gregory A. Dunn, Manager -FENOC Fleet Licensing, at (330) 315-7243 of any questions regarding this document or any associated regulatory commitments.

COMMITMENT DUE DATE None N/A Docket Number 50-346 License Number NPF-3 Serial Number 3274 Enclosure 1 Revised Technical Specification Bases Pages (40 pages follow)Note: The revised pages incorporate all changes made within the time period for this update. Only the most recent change to a page is identified with revision bars.

2.1 SAFETY

LIMITS BASES 2.1.1 AND 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant.Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime would result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB)and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNTB using critical heat flux (CHF) correlations.

The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.The B&W-2, BWC and BHTP CHF correlations have been developed to predict DN]B for axially uniform and non-uniform heat flux distributions.

The B&W-2 correlation applies to Mark-B fuel. The BWC correlation applies to Mark-B fuel with zircaloy or M5 spacer grids. The BHTP correlation applies to the Mark-B-HTP fuel. The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30 (B&W-2), 1.18 (BWC) and 1.132 (BHTP). The value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR equal to or greater than the correlation limit is predicted for the maximum possible thermal power 112% when the reactor coolant flow is 380,000 GPM, which is approximately 108% of design flow rate for four operating reactor coolant pumps. (The minimum required measured flow is 389,500 GPM). This curve is based on the design hot channel factors with potential fuel densification and fuel rod bowing effects.The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum allowable control rod withdrawal, and form the core DNBR design basis.DAVIS-BESSE, UNIT I B 2-1 Amendment No. 11,33,91,123,149, 189,239, LAR 05-0002

2.1 SAFETY

LIMITS BASES The CORE OPERATING LIiMITS REPORT includes curves for protective limits for AXIAL POWER IMBALANCE and for nuclear overpower based on reactor coolant system flow. A protective limit is a cycle-specific limit that ensures that a safety limit is not exceeded by requiring operation within both the cycle design (operating) limits and the Reactor Protection System setpoints.

These protective limit curves reflect the more restrictive of two thermal limits and account for the effects of potential fuel densification and potential fuel rod bow: 1. The DNBR limit produced by a design nuclear power peaking factor as described in the CORE OPERATING LIMITS REPORT or the combination of the radial peak, axial peak, and position of the axial peak that yields no less than the DNBR limit.2. The combination of radial and axial peak that causes central fuel melting at the hot spot. The limits for all fuel designs during the operating cycle are listed in the CORE OPERATING LIMITS REPORT.Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.The specified flow rates for the CORE OPERATING LIMITS REPORT curves for protective limits for AXIAL POWER IMBALANCE and for nuclear overpower based on reactor coolant system flow correspond to the analyzed minimum flow rates with four pumps and three pumps, respectively.

The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in BASES Figure 2.1. The curves of BASES Figure 2.1 represent the conditions at which a minimum DNBR equal to the DNBR limit is predicted at the maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to the corresponding DNB correlation quality limit (+22% (B&W-2), +26% (BWC) or +34% (BHTP)), whichever condition is more restrictive.

DAVIS-BESSE, UNIT 1 B 2-2 Amendment No. 11,33,45,61,80,123, 149,189, LAR 05-0002

2.1 SAFETY

LIMITS BASES For the curve of BASES Figure 2.1, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 (B&W-2), 1.18 (BWC) or 1.132 (BHTP) and a local quality at the point of minimum DNBR less than +22% (B&W-2), +26% (BWC) or +34%(BHTP) for that particular reactor coolant pump situation.

The DNBR curve for three pump operation is less restrictive than the four pump curve.2.1.3 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from.reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Boiler and Pressure Vessel Code which permits a maximum transient pressure of 110%, 2750 psig, of design pressure.

The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.7, 1968 Edition, which permits a maximum transient pressure of 110%, 2750 psig, of component design pressure.

The Safety Limit of 2750 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3125 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

DAVIS-BESSE, UNIT I B 2-3 Amendment No. 11,33,45,123,149, LAR 05-0002 LIMITING SAFETY SYSTEM SETTINGS BASES The AXIAL POWER IMBALANCE boundaries are established in order to prevent reactor thermal limits from being exceeded.

These thermal limits are either power peaking kW/ft limits or DNBR limits. The AXIAL POWER IMBALANCE reduces the power level trip produced by a flux-to-flow ratio such that the boundaries of the figure in the CORE OPERATING LIMITS REPORT are produced.RC Pressure -Low, High, and Pressure Temperature The high and low trips are provided to limit the pressure range in which reactor operation is permitted.

During a slow reactivity insertion startup accident from low power or a slow reactivity insertion from high power, the RC high pressure setpoint is reached before the high flux trip setpoint.

The Allowable Value for RC high pressure, 2355 psig, has been established to maintain the system pressure below the safety limit, 2750 psig, for any design transient.

The RC high pressure trip is backed up by the pressurizer code safety valves for RCS over pressure protection, and is therefore set lower than the set pressure for these valves, 2500 psig (nominal), even when accounting for the RPS RC pressure instrument string uncertainty.

The RC high pressure trip also backs up the high flux trip.The RC low pressure, 1900.0 psig, and RC pressure-temperature (16.25 Tou. -7899.0) psig, Allowable Values have been established to maintain the DNB ratio greater than or equal to the minimum allowable DNB ratio for those design accidents that result in a pressure reduction.

It also prevents reactor operation at pressures below the valid range of DNB correlation limits, protecting against DNB.High Flux/Number of Reactor Coolant Pumps On In conjunction with the flux -Aflux/flow trip the high flux/number of reactor coolant pumps on trip prevents the minimum core DNBR from decreasing below the minimum allowable DNB ratio by tripping the reactor due to the loss of reactor coolant pump(s). The pump monitors also restrict the power level for the number of pumps in operation.

DAVIS-BESSE, UNIT 1 B 2-6 Amendment No. 23,45,60,61,149, 189,218, LAR 05-0002 Revised by NRC letter dated May 7, 1999 Bases Figure 2.1 Pressure/Temperature Limits at Maximum Allowable Power for Minimum DNBR 2300 2200 2100 S2000 Cn 1900 1800 1700 595 605 615 625 Reactor Outlet Temperature, OF 635 Pumps 4 3 Flow, qpm 380,000 283,860 Power 112%90.1%Required Measured Flow to Ensure Compliance, qpm 389,500 290,957 I DAVIS-BESSE, UNIT 1 B 2-8 Amendment No. 11, 33, 45, 91, 123, LAR 05-0002 3/4.0 APPLICABILITY BASES-The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operation and Surveillance Requirements within Section 3/4.3.0.1 This specification defines the applicability of each specification in terms of defined OPERATIONAL MODES or other specified conditions and is provided to delineate specifically when each specification is applicable.

3.0.2 This specification defines those conditions necessary to constitute compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement.

An NRC internal memorandum provided in an NRC letter dated January 20, 1987 (Log Number 1-1548) provided NRC guidance on shutdown time allowances when a LCO is not met. This memorandum stated in part: It is our position that it is acceptable for a licensee to initiate and complete a reduction in operational modes in a shorter time interval than required by the allowable out-of-service time specified in an Action Statement and then to add the unused portion of this allowable out-of-service time to that provided for operation in a lower operational mode.Furthermore, it is our position that a stated allowable out-of-service time (frequently 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or 7 days) should be applicable regardless of the operational mode in which the inoperability is discovered.

However, the times provided for achieving a reduction in operational modes (e.g., generally 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from Modes 1 or 2 to Mode 3, and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from Mode 3 to Mode 4) should not be applicable if the inoperability is discovered in a lower operation mode.3.0.3 This specification delineates the ACTION to be taken for circumstances not directly provided for in the ACTION statements and whose occurrence would violate the intent of the specification.

For example, Specification

3.5.1 requires

each Reactor Coolant System core flooding tank to be OPERABLE and provides explicit ACTION requirements if one tank is inoperable.

Under the terms of the Specification 3.0.3, if more than one tank is inoperable, within one hour measures must be initiated to place the unit in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. As a further example, Specification 3.6.2.1 requires two Containment Spray Systems to be OPERABLE and provides explicit ACTION requirements if one spray system is inoperable.

Under the terms of Specification 3.0.3, if both of the required Containment Spray Systems are inoperable, within one hour measures must be initiated to place the unit in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least COLD SHUTDOWN in the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. It is assmned that the unit is brought to the required MODE within the required times by promptly initiating and carrying out the appropriate ACTION statement.

DAVIS-BESSE, UNIT 1 B 3/4 0-1 Amendment No. 71, 163, LAR No. 04-0005 APPLICABILITY BASES As described in NRC Generic Letter 87-09, the time limits of Specification

3.0.3 allow

37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> for the unit to be in COLD SHUTDOWN when a shutdown is required during POWER OPERATION.

If the unit is in a lower OPERATIONAL MODE when a shutdown is required, the time limit for reaching the next lower Mode applies. If a lower Mode is reached in less time than allowed, however, the total allowable time to reach COLD SHUTDOWN, or other applicable Mode, is not reduced. For example, if HOT STANDBY is reached in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, then the time allowed for reaching HOT SHUTDOWN is the next 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />, because the total time for reaching HOT SHUTDOWN is not reduced from the allowable limit of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Therefore, if remedial measures are completed that would permit a return to POWER OPERATION, a penalty is not incurred by having to reach a lower OPERATIONAL MODE in less than the total time allowed.3.0.4 This specification provides that entry into an OPERATIONAL MODE or other specified applicability condition must be made with (a) the full complement of required systems, equipment or components OPERABLE and (b) all other parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out of service provisions contained in the ACTION statements.

The intent of this provision is to insure that facility operation is not initiated with either required equipment or systems inoperable or other specified limits being exceeded.Exceptions to this provision have been provided for a limited number of specifications when startup with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION statements of the appropriate specifications.

3.0.5 establishes

an exception to the Specification

1.6 definition

of OPERABILITY for the condition when a system, subsystem, train, component, or device is inoperable either solely because its emergency electrical power source is inoperable or solely because it normal electrical power source is inoperable.

This exception, when applied, permits entry only into the ACTION(S) for the inoperable electrical power source(s).

This exception is provided because the definition of OPERABILITY would otherwise require that the ACTIONS for all systems, subsystems, trains, components, and devices powered by an inoperable electrical power source be entered.Application of the exception contained in Specification 3.0.5 to the OPERABILITY definition is an operational allowance.

Application of the exception of Specification 3.0.5 is an alternative that may be chosen in lieu of entering the ACTIONS for a system, subsystem, train, component, or device which is inoperable solely because either its emergency electrical power source or its normal electrical power source is inoperable.

In order to apply the exception contained in Specification 3.0.5, certain criteria must be satisfied.

Specifically, (1) the corresponding normal or emergency electrical power source must be OPERABLE, and (2) all redundant systems, subsystems, trains, components, and devices must be OPERABLE having at least an DAVIS-BESSE, UNIT 1 B 3/4 0-la Amendment No. 71 LAR No. 04-0005, 01-0014 APPLICABILITY BASES OPERABLE normal electrical power source or an OPERABLE emergency power source. If these criteria are not met for a system, subsystem, train, component, or device, then it must be considered inoperable, and action is required in accordance with-Specification 3.0.5. This specification allows plant operation to be governed by the time limits of the ACTION statement associated with the Limiting Condition for Operation of the normal or emergency electrical power source, rather than the individual ACTION statements for each system, subsystem, train, component, or device that is determined to be inoperable solely because of the inoperability of its normal or emergency electrical power source.For example, Specification 3.8.1.1 requires, in part, that two emergency diesel generators be OPERABLE.

The ACTION statement provides for a 7 day out-of-service time when one emergency diesel generator is not OPERABLE.

If the definition of OPERABLE were applied without application of Specification 3.0.5, all systems, subsystems, trains, components and devices supplied by the inoperable emergency diesel generator would also be inoperable.

This would dictate invoking the applicable ACTION statements for each of the applicable Limiting Conditions for Operation.

However, the provisions of Specification 3.0.5, when applied, permit the time limits for continued operation to be consistent with the ACTION statement for the inoperable emergency diesel generator instead, provided the other specified conditions are satisfied.

In this case, this would mean that the corresponding normal power source must be OPERABLE, and all redundant systems, subsystems, trains, components, and devices must be OPERABLE, or otherwise satisfy Specification 3.0.5 (i.e., be capable of performing their design function and have at least one normal or one emergency power source OPERABLE).

If they are not satisfied, the system, subsystem, train, component, or device is inoperable and action is required in accordance with Specification 3.0.5.As a further example, Specification 3.8.1.1 requires, in part, that two qualified circuits between the offsite transmission network and the onsite Class IE distribution system be OPERABLE.

The ACTION statement provides a 24-hour out-of-service time when both required qualified circuits are not OPERABLE.

If the definition of OPERABLE were applied without application of Specification 3.0.5; all systems, subsystems, trains, components and devices supplied by the inoperable normal electrical power sources, both of the qualified circuits, would also be_inoperable.

This would dictate invoking the applicable ACTION statements for each of the applicable LCOs. However, the provisions of Specification 3.0.5, when applied, permit the time limits for continued operation to be consistent with the ACTION statement for the inoperable normal electrical power sources instead, provided the other specified conditions are satisfied.

In this case, this would mean that for one train the emergency electrical power source must be OPERABLE (as must be the components supplied by the emergency electrical power source)and all redundant systems, subsystems, trains, components and devices in the other train must be OPERABLE, or likewise satisfy Specification 3.0.5 (i.e., be capable of performing their design functions and have an emergency electrical power source OPERABLE).

In other words, both emergency electrical power sources must be OPERABLE and all redundant systems, subsystems, trains, components and devices in both trains must also be OPERABLE.

If these conditions are not satisfied, the affected systems, subsystems, trains, components, or devices are inoperable and action is required in accordance with Specification 3.0.5.DAVIS-BESSE, UNIT 1 B 3/4 0-lb Amendment No. 71, 135, 203, 206 LAR No. 01-0014 APPLICABILITY BASES As another example, Specification 3.1.2.7 requires, in part, one OPERABLE boric acid pump, and its ACTION statement provides 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> out-of-service time when no boric acid pump is OPERABLE.

If one of the boric acid pumps is inoperable for maintenance, and the emergency diesel generator for the opposite train becomes inoperable, the ACTION statement of Specification 3.1.2.7 could be applied since it allows for continued plant operation up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with no OPERABLE boric acid pump.As another example, Specification 3.7.6.1 requires, in part, two station vent radiation monitors (RE4598AA and RE4598BA) be OPERABLE, and its ACTION statement provides for limited out-of-service times should one or both of the monitors become inoperable.

In cases where an inoperable monitor cannot be restored within the out-of-service time, the ACTION statement provides for continued plant operation by isolating the control room normal ventilation system and placing at least one control room emergency ventilation system train in operation.

If the sample pump for one monitor was removed from service for filter replacement, ACTION b of Specification 3.7.6.1 would be entered since one monitor would be inoperable, and the allowable out-of-service time of 7 days would be applied. If concurrently, the emergency diesel generator associated with the second monitor's sample pump was removed from service, ACTION c of Specification 3.7.6.1 would be entered. Since ACTION c of Specification 3.7.6.1 addresses the loss of both monitors and allows continued plant operation under compensatory actions, it would be inappropriate to apply Specification 3.0.5 to the station vent radiation monitors.As a final example, Specification 3.4.6.1 requires, in part, one containment atmosphere radioactivity monitor (RE4597AA or RE4597BA) be OPERABLE, and its ACTION statement (provides for a 30 day out-of-service time and compensatory actions should the monitor become inoperable.

If one of the monitor's sample pumps became inoperable due to its filter plugging, and the emergency diesel generator associated with the other monitor's sample pump was already inoperable due to maintenance being performed, ACTION b of Specification 3.4.6.1 would be entered. Since the ACTIONS of Specification 3.4.6.1 address the loss of both monitors and allow limited continued plant operation, it would be inappropriate to apply Specification 3.0.5 to the containment atmosphere radioactivity monitors.The time limits of Specification

3.0.5 allow

38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br /> for the unit to be in COLD SHUTDOWN when a shutdown is required during POWER OPERATION.

If the unit is in a lower OPERATIONAL MODE when a shutdown is required, the time limit for reaching the next lower Mode applies. If a lower Mode is reached in less time than allowed, however, the total allowable time to reach COLD SHUTDOWN, or other applicable Mode, is not reduced. For example, if HOT STANDBY is reached in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, then the time allowed for reaching HOT SHUTDOWN is the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, because the total time for reaching HOT SHUTDOWN is not reduced from the allowable limit of 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />. Therefore, if remedial measures are completed that would permit a return to POWER OPERATION, a penalty is not incurred by having to reach a lower OPERATIONAL MODE in less than the total time allowed.In MODES 5 or 6, Specification 3.0.5 is not applicable, and thus the individual ACTION statements for each applicable Limiting Condition for Operation in these MODES must be adhered to.DAVIS-BESSE, UNIT I B 3/4 0-ic Amendment No. 71 LAR No. 04-0005, 01-0014 APPLICABILITY BASES 3.0.6 Specification

3.0.6 establishes

the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to Specification 3.0.2 (e.g., to not comply with the applicable Required Actions(s))

to allow the performance of required testing to demonstrate:

a. The OPERABILITY of the equipment being returned to service; or b. The OPERABILITY of other equipment.

The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to perform the required testing to demonstrate OPERABILITY.

This Specification does not provide time to perform any other preventive or corrective maintenance.

An example of demonstrating the OPERABILITY of the equipment being returned to service is reopening a containment isolation valve that has been closed to comply with Required Actions, and must be reopened to perform the required testing.An example of demonstrating the OPERABILITY of other equipment being returned to service is taking an inoperable channel or trip system out of the tripped condition to prevent the trip function from occurring during the performance of required testing on another channel in the other trip system. A similar example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to permit the logic to function and indicate the appropriate response during the performance of required testing on another channel in the same trip system.DAVIS-BESSE, UNIT 1 B 3/4 0-1d Amendment No. 224 LAR No. 01-0014 APPLICABILITY BASES 4.0.1 This specification provides that surveillance activities necessary to insure the Limiting Conditions for Operation are met and will be performed during the OPERATIONAL MODES or other conditions for which the Limiting Conditions for Operation are applicable.

Provisions for additional surveillance activities to be performed without regard to the applicable OPERATIONAL MODES or other conditions are provided in the individual Surveillance Requirements.

4.0.2 The provisions of this specification provide allowable tolerances for performing surveillance activities beyond those specified in the nominal surveillance interval.

These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations.

The phrase "at least" associated with a surveillance frequency does not negate this allowable tolerance value and permits the performance of more frequent surveillance activities.

The allowable tolerance for performing surveillance activities is sufficiently restrictive to ensure that the reliability associated with the surveillance activity is not significantly degraded beyond that obtained from the nominal specified interval.

It is not intended that the allowable tolerance be used as a convenience to repeatedly schedule the performance of surveillances at the allowable tolerance limit.The allowable tolerance for performing surveillance activities also provides flexibility to accommodate the length of a fuel cycle for surveillances that are specified to be performed at least once each REFUELING INTERVAL.

It is the intent that REFUELING INTERVAL surveillances be performed in an OPERATIONAL MODE consistent with safe plant operation.

In the case where the surveillance interval applies to tests performed on a STAGGERED TEST BASIS, as defined by Technical Specification (TS) Definition 1.21, the 25 percent allowance may be applied to the particular channel of interest, irrespective of the test completion history for the other channels.

In NRC letter Log Number 5208, dated February 4, 1998, the NRC stated, "...the 25 percent extension of TS 4.0.2 addresses the overall length of the surveillance interval, while the subintervals of the staggered testing definition only deal with how the beginning of the overall intervals for different subsystems are arranged.

Therefore, TS 4.0.2 should be applied to the overall interval requirement...

Since TS 4.0.2 cannot'be used on a routine basis, a test (say, on Channel 2) conducted subsequent to the application of TS 4.0.2 would still have to performed again in the next subinterval as previously scheduled (for Channel 2)." The following example was provided by the NRC for a four channel system with channel functional testing performed once per 32 days on a staggered test basis: DAVIS-BESSE, UNIT I B 3/4 0-2 Amendment No. 140. 213 I°LAR No. 04-0008 APPLICABILITY Day Channel 8 1 completed (normal schedule)16 (2 scheduled, not completed, TS 4.0.2 invoked)24 2 completed (including TS 4.0.2 25 percent extension) 3 completed (normal schedule)32 4 completed (normal schedule)40 1 completed (normal schedule)48 2 completed (return to normal schedule)56 3 completed. (normal schedule)4.0.3 This specification establishes the failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by the provisions of Specification 4.0.2, as a condition that constitutes a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation.

Under the provisions of this specification, systems and components are assumed to be OPERABLE when Surveillance Requirements have been satisfactorily performed within the specified time interval.

However, nothing in this provision is to be construed as implying that systems or components are OPERABLE when either they are found or known to be inoperable although still meeting the Surveillance Requirements, or the requirements of the surveillance(s) are known to be not met between required surveillance performances.

Specification

4.0.3 establishes

the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency.

A delay period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with Specification 4.0.2, and not at the time that the specified Frequency was not met.This delay period provides an adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with required ACTIONS or other remedial measures that might preclude completion of the Surveillance.

The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

DAVIS-BESSE, UNIT I B 3/4 0-3 Amendment No. 145 LAR No. 01-0017, LAR No. 04-0008 APPLICABILITY When a Surveillance with a Frequency based not on time intervals, but upon specified unit conditions, operating situations, or requirements of regulations (e.g., prior to entering MODE I after each fuel loading, or in accordance with 10 CFR 5.0, Appendix J, as modified by approved exemptions, etc.) is discovered to not have been performed when specified, Specification

4.0.3 allows

for the full delay period up to the specified Frequency to perform the Surveillance.

However, since there is not a time interval specified, the missed Surveillance should be performed at the first reasonable opportunity.

Specification

4.0.3 provides

a time limit for, and allowances for the performance of, Surveillances that become applicable as a consequence of MODE changes imposed by required ACTIONS.Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence.

Use of the delay period established by Specification 4.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals.

While up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity.

The determination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration changes required or shutting the plant down to perform the Surveillance) and impact on any analysis assumptions, in addition to unit conditions, planning, availability of personnel, and time required to perform the Surveillance.

This risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, NRC Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to and including plant shutdown.The missed surveillance should be treated as an emergent condition as discussed in the Regulatory Guide. The risk evaluation may use quantitative, qualitative, or blended methods. The degree of depth and rigor of the evaluation should be commensurate with the importance of the component.

Missed Surveillances for important components should be analyzed quantitatively.

If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action. All missed Surveillances will be placed in the Corrective Action Program.If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the ACTIONS forthe applicable LCO Conditions must be entered immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the ACTIONS for the applicable'LCO Conditions must be entered immediately upon the failure of the Surveillance.

Completion of the Surveillance within the delay period allowed by this Specification, or within the time limits of the ACTIONS, restores compliance with this specification.

Surveillance Requirements do not have to be performed on inoperable equipment because the ACTION requirements define the remedial measures that apply. However, the Surveillance Requirements have to be met to demonstrate that inoperable equipment has been restored to OPERABLE status.DAVIS-BESSE, UNIT I B 3/4 0-4 Amendment No. 145, 248 LAR No. 01-0017, LAR No. 04-0008 APPLICABILITY BASES 4.0.4 This specification ensures that the surveillance activities associated with a Limiting Condition for Operation have been performed within the specified time interval prior to entry into an OPERATIONAL MODE or other applicable condition.

The intent of this provision is to ensure that surveillance activities have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Limiting Condition for Operation.

Under the terms of'this specification, for example, during initial plant startup or following extended plant outages, the applicable surveillance activities must be performed within the stated surveillance interval prior to placing or returning the system or equipment into OPERABLE status.4.0.5 This specification ensures that 1) inservice inspection of ASME Code Class 1, 2 and 3 components will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a, and 2) inservice testing of ASME Code Class 1, 2 and 3 pumps and valves will be performed in accordance with a periodically updated version of the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) and applicable Addenda as required by 10 CFR 50.55a.This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and the ASME OM Code and their applicable Addenda. This clarification is provided to ensure consistency in surveillance intervals throughout these Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.

Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and the ASME OM Code and their applicable Addenda. For example, the requirements of Specification 4.0.4 to perform surveillance activities prior to entry into an OPERATIONAL MODE or other specified applicability condition takes precedence over the ASME OM Code provison which allows pumps to be tested up to one week after return to normal operation.

DAVIS-BESSE, UNIT I B 3/4 0-5 Amendment No. 145, 197, 250 LAR No. 04-0008 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM AND SAFETY SYSTEM INSTRUMENTATION OPERABILITY of the RPS, SFAS and SFRCS instrumentation systems ensures that 1) the associated action and/or trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2) the specified coincidence logic is maintained, 3)sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for RPS, SFAS and SFRCS purposes from diverse parameters.

OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the required design standards.

The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

The response time limits for these instrumentation systems are located in the Updated Safety Analysis Report and are used to demonstrate OPERABILITY in accordance with each system's response time surveillance requirements.

As indicated in RPS Table 4.3-1 for Functional Units 1 and 12, a CHANNEL FUNCTIONAL TEST is required to be performed for the Manual Reactor Trip function and the CRD Trip Breakers function once prior to each reactor startup, i.e., prior to Mode 2 entry, if not performed within the previous 7 days. These surveillance requirements ensure the OPERABILITY of these Functional Units prior to achieving criticality.

If the plant is in MODE 2 or if the plant is in MODE 3, 4, or 5 with the control rod drive trip breakers in the closed position and the control rod drive system capable of rod withdrawal, then Functional Unit 11.A of Table 3.3-1 applies. With THERMAL POWER level > 10-1i amps on both Intermediate Range channels, high voltage to the Source Range detectors may be de-energized.

If the plant is in MODE 3, 4, or 5 with the control rod drive trip breakers not in the closed position or the control rod drive system not capable of withdrawal, Functional Unit 11 .B of Table 3.3-1 applies. Applicability of Functional Units 1 .A and 1 1.B is dependent upon the plant MODE and control rod drive system status and is not dependent on if the plant is in the process of a startup or a shutdown.DAVIS-BESSE, UNIT 1 B 3/4 3-1 Amendment No. 225 LAR No. 03-0006, 04-0006, 05-0002 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM AND SAFETY SYSTEM INSTRUMENTATION (Continued)

To comply with TS 3.3.1.1, all four CRD Trip Breakers and all four RPS Reactor Trip Modules (RTMs) must be OPERABLE, or entry into RPS Table 3.3-1, ACTION 7 and/or ACTION 8 is required: If an RTM is inoperable, depending on the reason, it may be appropriate to apply Action 7.a.1, 7.a.2, or both, as needed to ensure that a valid reactor trip signal will cause a reactor trip when needed, despite the inoperability of the RTM. The most conservative action would be to open the CRD Trip Breaker and physically remove the RTM from its cabinet, however, this may make the plant susceptible to an unnecessary reactor trip and transient due to a spurious signal..If a CRD Trip Breaker is inoperable, it is appropriate to apply Action 7.a.2.Action 7.b affords limited flexibility with respect to surveillance testing in the Reactor Trip System opposite the Reactor Trip System with the inoperable channel.The Shutdown Bypass High Pressure trip provides the means for removing four automatic RPS trip parameters (i.e., RC Low Pressure, RC Pressure-Temperature, Flux-AFlux-Flow, High Flux/Number of Reactor Coolant Pumps On) from the RPS circuitry during a plant shutdown and heatup, or control rod drive testing and zero power physics testing, and inserts a lower RC high pressure trip in the circuit. If the CRD breakers are open and the CRD system is not capable of rod withdrawal, the trip function of the Shutdown Bypass is not considered actuated with the keyswitch initiated since it will not cause tripping of the CRD breakers.

Therefore, under this condition the "Applicable MODES" of Functional Unit 14 of Table 3.3-1 and the "MODES in Which Surveillance Required" for Functional Unit 14 of Table 4.3-1 are not entered.SFAS Table 3.3-3, ACTION 10, allows entry into this ACTION statement to be delayed for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when a functional unit is placed in an inoperable status solely for performance of a CHANNEL FUNCTIONAL TEST, provided at least two other corresponding functional units remain OPERABLE.

The term "corresponding functional units" refers to the functional units (Total No. of Units column in Table 3.3-3) for the same trip setpoint in the other SFAS channels.For example, the corresponding functional units for a Containment Pressure -High unit are the other three Containment Pressure -High units. This 8-hour allowance provides a reasonable time to perform the required surveillance testing without having to enter the ACTION statement and implement the required ACTIONS.DAVIS-BESSE, UNIT 1 B 3/4 3-2 Amendment No. 259 LAR No. 99-0004, 04-0010, 04-0011, 05-0002 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM AND SAFETY SYSTEM INSTRUMENTATION (Continued)

SFRCS Table 3.3-11, ACTION 16, allows entry into this ACTION statement to be delayed for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when a channel is placed in an inoperable status solely for performance of a CHANNEL FUNCTIONAL TEST, provided the remaining actuation channel remains OPERABLE.

This 8-hour allowance provides a reasonable time to perform the required surveillance testing without having to enter the ACTION statement and implement the required ACTIONS.For the RPS, SFAS Table 3.3-4 Functional Unit Instrument Strings b, c, d, e, and f, and Interlock Channel a, and SFRCS Table 3.3-12 Functional Unit 2: [Note: Setpoint methodology for RPS Functional Unit 7 is discussed on page B 3/4 3-4.]Only the Allowable Value is specified for each Function.

Nominal trip setpoints are specified in the setpoint analysis.

The nominal trip setpoints are selected to ensure the setpoints measured by CHANNEL FUNCTIONAL TESTS do not exceed the Allowable Value if the bistable is performing as required.

Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable provided that operation and testing are consistent with the assumptions of the specific setpoint calculations.

Each Allowable Value specified is more conservative than the analytical limit assumed in the safety analysis to account for instrument uncertainties appropriate to the trip parameter.

These uncertainties are defined in the specific setpoint analysis.A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.

Setpoints must be found within the specified Allowable Values. Any setpoint adjustment shall be consistent with the assumptions of the current specific setpoint analysis.A CHANNEL CALIBRATION is a complete check of the instrument channel, including the sensor. The test verifies that the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift to ensure that the instrument channel remains operational between successive tests. CHANNEL CALIBRATION shall find that measurement errors and bistable setpoint errors are within the assumptions of the setpoint analysis.

CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the setpoint analysis.The frequency is justified by the assumption of an 18 or 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.DAVIS-BESSE, UNIT I B 3/4 3-3 Amendment 218, 241, 243, 259 LAR No. 99-0004, LAR 05-0002 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM AND SAFETY SYSTEM INSTRUMENTATION (Continued)

For RPS Functional Unit 7, the Limiting Trip Setpoint is specified in the USAR Technical Requirements Manual. The Limiting Trip Setpoint is based on the calculated total loop uncertainty per the plant-specific methodology identified below. The Limiting Trip Setpoint may be established using Method I or Method 2 from Section 7 of ISA RP67.04.02-2000,"Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation." Additional information is contained in the Technical Requirements Manual.The purpose of a Limiting Safety System Setting is to ensure that protective action is initiated before the process conditions reach the analytical limit, thereby limiting the consequences of a design-basis event to those predicted by the safety analyses.

For RPS Functional Unit 7, the Limiting Trip Setpoint is the Limiting Safety System Setting required by 10 CFR 50.36. This definition of the LSSS is consistent with the guidance issued to the industry through correspondence with NEI (Reference NRC-NEI Letter dated September 7, 2005). The definition of LSSS values continues to be discussed between the industry and the NRC, and further modifications to these TS Bases will be implemented as guidance is provided.TS Table 4.3-1 Note 10 ensures that unexpected as-found conditions are evaluated prior to returning the channel to service, and that as-left settings provide sufficient margin for uncertainties.

Specifically, for RPS Functional Unit 7, the following additional requirements are added by TS Table 4.3-1 Note 10: 1) If the as-found Trip Setpoint is non-conservative with respect to the Allowable Value specified in the Technical Specification, the channel shall be declared inoperable and the associated Technical Specification ACTION statement shall be followed.2) If the as-found Trip Setpoint is conservative with respect to the Allowable Value, and outside the pre-defined as-found acceptance criteria band, but the instrument is functioning as required and the channel can be reset to within the setting tolerance of the Limiting Trip Setpoint, or a value more conservative than the Limiting Trip Setpoint, then the channel may be considered OPERABLE.

If it cannot be determined that the instrument channel is functioning as required, the channel shall be declared inoperable and the associated Technical Specification actions shall be followed.3) If the as-found Trip Setpoint is outside the predefined as-found acceptance criteria band, the condition must-be entered into the corrective action program for further evaluation.

DAVIS-BESSE, UNIT I B 3/4 3-4 LAR No. 05-0002 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM AND SAFETY SYSTEM INSTRUMENTATION (Continued)

The pre-defined as-found acceptance criteria band is determined by adding or subtracting (as applicable) credible uncertainties to the as-left value from the most recently completed surveillance test. Credible uncertainties include instrument uncertainties during normal operation including drift and measurement and test equipment uncertainties.

In no case shall the pre-defined as-found acceptance criteria band overlap the Allowable Value. If one end of the pre-defined as-found acceptance criteria band is truncated due to its proximity to the Allowable Value, this does not affect the other end of the pre-defined as-found acceptance criteria band.. If equipment is replaced, such that the previous as-left value is not applicable to the current configuration, the as-found acceptance criteria band is not applicable to calibration activities performed immediately following the equipment replacement.

Measurement of response time at the specified frequencies provides assurance that the RPS, SFAS, and SFRCS action function associated with each channel is completed within the time limit assumed in the safety analyses.Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined.Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.The SFRCS RESPONSE TIME for the turbine stop valve closure is based on the combined response times of main steam line low pressure sensors, logic cabinet delay for main steam line low pressure signals and closure time of the turbine, stop valves. This SFRCS RESPONSE TIME ensures that the auxiliary feedwater to the unaffected steam generator will not be isolated due to a SFRCS low pressure trip during a main steam line break accident.DAVIS-BESSE, UNIT I B 3/4 3-5 Amendment 125, 225, 246, LAR No. 05-0002 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM AND SAFETY SYSTEM INSTRUMENTATION (Continued)

Surveillance Requirement 4.3.2.1.3 requires demonstration that each SFAS function can be performed within the applicable SAFETY FEATURES RESPONSE TIME. If an inoperable SFAS-actuated component is placed in its actuated position, this surveillance requirement is met.When this surveillance requirement can not be met due to an inoperable SFAS-actuated component, the LCO ACTION associated with the inoperable actuated component should be entered. The actuated component, except for the contacts at the actuator, is not part of the SFAS output logic. When the SAFETY FEATURES RESPONSE TIME surveillance requirement can not be met due to inoperable components within the SFAS,. the applicable Table 3.3-3 ACTION should be followed.

Similarly, Surveillance Requirement 4.3.2.2.3 requires demonstration that each SFRCS function can be performed within the applicable SFRCS RESPONSE TIME. When this surveillance requirement can not be met due to an inoperable SFRCS-actuated component, the LCO ACTION associated with the inoperable actuated component should be entered. When the SFRCS RESPONSE TIME surveillance requirement can not be met due to inoperable components within the SFRCS, ACTION 16 of Table 3.3-11 should be followed.The actuation logic for Functional Units 4.a., 4.b., and 4.c. of Table 3.3-3, Safety Features Actuation System Instrumentation, is designed to provide protection and actuation of a single train of safety features equipment, essential bus or emergency diesel generator.

Collectively, Functional Units 4.a., 4.b., and 4.c. function to detect a degraded voltage condition on either of the two 4160 volt essential buses, shed connected loads, disconnect the affected bus(es) from the offsite power source and start the associated emergency diesel generator.

In addition, if an SFAS actuation signal is present under these conditions, the sequencer channels for the two SFAS channels which actuate the train of safety features equipment powered by the affected bus will automatically sequence these loads onto the bus to prevent overloading of the emergency diesel generator.

Functional Unit 4.a. has a total of four units, one associated with each SFAS channel (i.e., two for each essential bus). Functional Units 4.b. and 4.c. each have a total of four units, (two associated with each essential bus); each unit consisting of two undervoltage relays and an auxiliary relay.DAVIS-BESSE, UNIT 1 B 3/4 3-6 Amendment No. 211, 246, LAR No. 04-0007, 05-0002 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR PROTECTION SYSTEM AND SAFETY SYSTEM INSTRUMENTATION (Continued)

An SFRCS channel consists of 1) the sensing device(s), 2) associated logic and output relays, and 3) power sources. The SFRCS output signals that close the Main Feedwater Block Valves (FW-779 and FW-780) and trip the Anticipatory Reactor Trip System (ARTS) are not required to mitigate any accident and are not credited in any safety analysis.

Therefore, LCO 3.3.2.2 does not apply to these functions.

Safety-grade anticipatory reactor trip is initiated by a turbine trip (above 45 percent of RATED THERMAL POWER) or trip of both main feedwater pump turbines.

This anticipatory trip will operate in advance of the reactor coolant system high pressure reactor trip to reduce the peak reactor coolant system pressure and thus reduce challenges to the pilot operated relief valve. This anticipatory reactor trip system was installed to satisfy Item II.K.2. 10 of NUREG-0737.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.There are two redundant Fuel Storage Pool Area EVS Area Monitors.

With one channel of Fuel Storage Pool Area EVS Area Monitors operable and one channel inoperable, the requirements of TS LCO 3.3.3.1 and TS Table 3.3-6 are satisfied without reliance on the associated actions.Therefore, entry into TS 3.3.3.1 Action b is not required.

Appropriate actions with respect to TS 3.9.12 must still be taken.With zero channels of Fuel Storage Pool Area EVS Area Monitors operable, the requirements of TS LCO 3.J.3.1 and TS Table 3.3-6 are not satisfied, so the TS 3.3.3.1 Action b must be entered.Therefore, Action 22 of Table 3.3-6 must be satisfied.

With one or more of the Containment Activity Monitors (either gaseous or particulate) operable, the requirements of TS LCO 3.3.3.1 and TS Table 3.3-6 are satisfied without reliance on the associated actions. Therefore, entry into TS 3.3.3.1 Action b is not required.With no Containment Activity Monitors operable, the requirements of TS LCO 3.3.3.1 and TS Table 3.3-6 are not satisfied, so the TS 3.3.3.1 Action b must be entered. Action 21 of Table 3.3-6 must be satisfied.

3/4.3.3.2 INCORE DETECTORS

-Deleted DAVIS-BESSE, UNIT 1 B 3/4 3-7 Amendment No 73, 128, 135, 234, 246 LAR No. 01-0001, 04-0017, 05-0002 3/4.3 INSTRUMENTATION BASES 3/4.3.3.3 SEISMIC INSTRUMENTATION

-Deleted 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION

-Deleted 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost.SR 4.3.3.5.2 verifies that each Remote Shutdown System transfer switch and control circuit required for a serious control room or cable spreading room fire performs its intended function.This verification is performed from the remote shutdown panel and locally, as appropriate.

This.will ensure that if the control room becomes inaccessible, the unit can be safely shutdown from the remote shutdown panel and the local control stations.3/4.3.3.6 POST-ACCIDENT MONITORING INSTRUMENTATION OPERABILITY of the post-accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.LCO 3.3.3.6, Table 3.3-10 requires only two "Containment Vessel Post-Accident Radiation" channels be operable, i.e., any two of RE4596A, RE4596B, RE4597AB, and RE4597BB can be utilized to comply with the LCO. Since RE4596A and RE4596B were specifically credited for meeting NUREG-0737 Item II.F. 1.3, and LCO 3.3.3.6 was credited as providing requirements for their operability, RE4596A and RE4596B are the preferred monitors for complying with LCO.3.3.3.6.Regarding the Auxiliary Feedwater Flow Rate Monthly CHANNEL CHECK, verifying that the AFW flow meters read zero, within an acceptable tolerance, and checking that both meters for each steam generator are reading acceptably provides assurance that there is no flow to the steam generators, which is normal.. The fact that the flow meters are in an idle system does not reduce the importance or necessity for ensuring the meters are in an expected state and providing expected and confirmed information.

3/4.3.3.7 CHLORINE DETECTION SYSTEMS -Deleted 3/4.3.3.8 FIRE DETECTION INSTRUMENTATION

-Deleted 3/4.3.3.9 WASTE GAS SYSTEM OXYGEN MONITOR -Deleted DAVIS-BESSE, UNIT 1 B 3/4 3-8 Amendment No. 9, 50, 58, 86, 134, 156, 167, 170, 174, 187,201,234 LAR No. 03-0021, 04-0009, 04-0021, 05-0002 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)BASES (Continued)

Tank (BWST) or the Containment Emergency Sump to the reactor vessel. Either subsystem operating in conjunction with the core flooding tanks is capable of supplying sufficient core cooling to maintain the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward.

With RCS average temperature

>_ 280'F, the Limiting Condition for Operation (LCO) requires the OPERABILITY of a number of independent trains, the inoperability of one component in a train does not necessarily render the ECCS incapable of performing its function.

Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. The intent of this LCO is to maintain a combination of equipment such that 100% of the safety injection flow equivalent to 100% of a single subsystem remains available.

This allows increased flexibility in plant operations under circumstances when components in opposite subsystems are inoperable.

With one or more components inoperable such that 100% of the flow equivalent to a single OPERABLE ECCS subsystem is not available, the facility is in a condition outside the accident analyses.

Therefore, LCO 3.0.3 must be immediately entered.In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period.Cross-connects are provided between the two LPI trains, downstream of the decay heat (DH).cooler in each train. Included in these cross-connects are two normally closed motor-operated valves, DH-830 and DH-83 1. As described in USAR Section 6.3.2.11, the crossover lines can be actuated to provide additional protection in the unlikely event of a double-ended rupture of a Core Flooding Tank line, however the accident analysis performed for this break does not take credit for operator action to open the cross-connects.

Rather, the analysis assumes only minimum ECCS is available consisting of one HPI train and one Core Flooding Tank (i.e., no LPI was included).

The USAR further states that if the crossover lines were utilized, the core would recover with liquid at a faster rate, and would fulfill the requirements of abundant core cooling. Based on this information, the LPI crossover piping and valves need not be available for the required flow path to be OPERABLE.

However, if maintenance activities are being conducted which make a crossover valve inoperable, either the inoperable valve (DH-830/DH-83

1) or the associated flow path stop-check valve (DH-127/DH-125) should be maintained closed such that train separation is maintained.

In order to facilitate online maintenance, License Amendment No. 253 revised the TS LCO 3.5.2 Action statement to increase the allowable outage time for an inoperable LPI train or its associated decay heat cooler from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days. The allowable outage time for an inoperable HPI train was maintained at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, recognizing that if one LPI pump was removed from service (entering the 7 day Action statement), as long as both HPI trains could be supplied by the remaining LPI pump, both HPI trains would remain operable and not require entry into the associated 72-hour HPI train Action statement.

An example of when the 72-hour HPI train Action statement would apply is removing an LPIIHPI piggyback valve (DH-63 or DH-64) from service. Another example is removing an LPI train from service such that the cross connect valve DH830 or DH 831 would not be available.

In both of these examples, the affected HPI train would be declared inoperable in that the activity would not support operability of both HPI trains. Therefore, caution must be taken during the plann'ing and execution of HPI and LPI train outages with respect to the effect the removal of a HPI or a LPI train and associated flow pathways (e.g., piggyback valves DH-63 and DH-64, or the LPI cross connect valves DH-830 and 831) will have on the remaining HPI and LPI trains to ensure compliance with TS LCO 3.5.2.DAVIS-BESSE, UNIT 1 B 3/4 5-1a Amendment No. 20, 191, 253, LAR No. 04-0001, 04-0022 EMERGENCY CORE COOLING SYSTEMS BASES With the RCS temperature below 280'F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.

The function of the trisodium phosphate dodecahydrate (TSP) contained in baskets located in the containment normal sump or on the 565' elevation of containment adjacent to the normal sump, is to neutralize the acidity of the post-LOCA borated water mixture during containment emergency sump recirculation.

The borated water storage tank (BWST) borated water has a nominal pH value of approximately

5. Raising the borated water mixture to a pH value of 7 will ensure that chloride stress corrosion does not occur in austenitic stainless steels in the event that chloride levels increase as a result of contamination on the surfaces of the reactor containment building.

Also, a pH of 7 is assumed for the containment emergency sump for iodine retention and removal post-LOCA by the containment spray system.The Surveillance Requirement (SR) associated with TSP ensures that the minimum required volume of TSP is stored in the baskets. The minimum required volume of TSP is the volume that will achieve a post-LOCA borated water mixture pH of -7.0, conservatively considering the maximum possible sump water volume and the maximum possible boron concentration.

The amount of TSP required is based on the mass of TSP needed to achieve the required pH.However, a required volume is verified by the SR, rather than the mass, since it is not feasible to weigh the entire amount of TSP in containment.

The minimum required volume is based on the manufactured density of TSP (53 lb/ft 3). Since TSP can have a tendency to agglomerate from high humidity in the containment, the density may increase and the volume decrease during normal plant operation, however, solubility characteristics are not expected to change.Therefore, considering possible agglomeration and increase in density, verifying the minimum volume of TSP in containment is conservative with respect to ensuring the capability to achieve the minimum required pH. The minimum required volume of TSP to meet all analytical requirements is 250 ft 3.The surveillance requirement of 290 ft 3 includes 40 ft 3 of spare TSP as margin. Total basket capacity is 325 ft 3.DAVIS-BESSE, UNIT 1 B 3/4 5-2 Amendment No. 20,123,182,191, 195,207,215, LAR No. 03-0004 EMERGENCY CORE COOLING SYSTEMS BASES (Continued)

The surveillance requirement for throttle valve position stops provides assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME OM Code.This type of testing may be accomplished by measuring the pump's developed head at only one point of the pump's characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant accident analysis.Surveillance Requirements are specified in the Inservice Testing Program, which encompasses the ASME OM Code. The ASME OM Code and Technical Specification

4.0.5 provide

the activities and frequencies necessary to satisfy the requirements.

Containment Emergency Sump Recirculation Valves DH-9A and DH-9B are de-energized during MODES 1, 2, 3 and 4 to preclude postulated inadvertent opening of the valves in the event of a Control Room fire, which could result in draining the Borated Water Storage Tank to the Containment Emergency Sump and the loss of this water source for normal plant shutdown.Re-energization of DH-9A and DH-9B is permitted on an intermittent basis during MODES 1, 2, 3 and 4 under administrative controls.

Station procedures identify the precautions which must be taken when re-energizing these valves under such controls.Borated Water Storage Tank (BWST) outlet isolation valves DH-7A and DH-7B are de-energized during MODES 1, 2, 3, and 4 to preclude postulated inadvertent closure of the valves in the event of a fire, which could result in a loss of the availability of the BWST.Re'-energization of valves DH-7A and DH-7B is permitted on an intermittent basis during MODES 1, 2, 3, and 4 under administrative controls.

Station procedures identify the precautions which must be taken when re-energizing these valves under such controls.The Decay Heat Isolation Valve and Pressurizer Heater Interlock setpoint is based on preventing over-pressurization of the Decay Heat Removal System normal suction line piping. The value stated is the RCS pressure at the sensing instrument's tap. It has been adjusted to reflect the elevation difference between the sensor's location and the pipe of concern.As discussed in the NRC Safety Evaluation for License Amendment No. 196, it is intended that visual inspections conducted in accordance with SR 4.5.2.c.2 be limited to those areas in containment where work was performed, so as to limit the potential for radiation dose to people performing the inspections.

DAVIS-BESSE, UNIT 1 B 3/4 5-2a Amendment No. 191,207,215,218,241 LAR No. 03-0002, 04-0026 EMERGENCY CORE COOLING SYSTEMS BASES (Continued) 3/4.5.4 BORATED WATER STORAGE TANK The OPERABILITY of the borated water storage tank (BWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on the BWST minimum volume (500,100 gallons of borated water, conservatively rounded up from the calculated value of 500,051 gallons)and boron concentration ensure that: 1) sufficient water is available within containment to permit recirculation cooling flow to the core following manual switchover to the recirculation mode, and 2) The reactor will remain at least 1% Ak/k subcritical in the cold condition at 70'F, xenon free, while only crediting 50% of the control rods' worth following mixing of the BWST and the :RCS water volumes.These assumptions ensure that the reactor remains subcritical in the cold condition following mixing of the BWST and the RCS water volumes.With either the BWST boron concentration or BWST borated water temperature not within limits, the condition must be corrected in eight hours. The eight hour limit to restore the temperature or boron concentration to within limits was developed considering the time required to change boron concentration or temperature and assuming that the contents of the BWST are still available for injection.

A portion of the contained vo:lume of the BWST is not usable because of the tank discharge configuration.

Therefore, the limits on water volume reflect the available volume. The limits onwater volume, and boron concentration ensure a pH value of between 7.0 and 11.0 of the solution sprayed within the containment after a design basis accident.

The pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion cracking on mechanical systems and components.

DAVIS-BESSE, UNIT 1 B 3/4 5-2b Amendment No. 191, 215, 218 LAR No. 03-0002, 06-0001 CONTAINMENT SYSTEMS BASES 3/4.6.1.3 CONTAINMENT AIR LOCKS (Continued)

The surveillance requirement which verifies that only one door in each air lock can be opened at a time is not part of the Containment Leakage Rate Testing Program. Therefore, its test frequency is subject to the provisions of Specification 4.0.2.3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.5 psi and 2) the containment peak pressure does not exceed the maximum internal pressure of 40 psig during LOCA conditions.

The initial pressure condition used in the containment analysis was 1 psig. This resulted in a maximum peak pressure from a LOCA of 38 psig.3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial temperature condition assumed in the accident analysis for a LOCA.3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY Deleted 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM Maintaining the containment purge supply and exhaust isolation valves closed with control power removed at all times during MODES 1,2, 3and 4 provides assurance that the safety function of containment isolation is maintained in the event of a LOCA.The ACTION statement assures that at least one containment purge supply and exhaust isolation valve is closed in each containment penetration and provides reasonable time to permit closure of an open valve.3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the containment spray system ensures that containment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment DAVIS-BESSE, UNIT 1 B 3/4 6-2 Amendment No. 135, 205, 221, 240 LAR No. 03-0005 CONTArNMENT SYSTEMS BASES ..leakage rate are consistent with the assumptions used in the safety analyses.Borated Water Storage Tank (BWST) outlet isolation valves DH-7A and DH-7B are de-energized during MODES 1, 2, 3, and 4 to preclude postulated inadvertent closure of the valves in the event of a fire, which could result in a loss of the availability of the BWST. Re-energization of valves DH-7A and DH-7B is permitted on an intermittent basis during MODES 1, 2, 3 and 4 under administrative controls.

Station procedures identify the precautions which must be taken when re-energizing these valves under such controls.Containment Emergency Sump Recirculation Valves DH-9A and DH-9B are de-energized during MODES 1, 2, 3, and 4 to preclude postulated inadvertent opening of the valves in the event of a fire, which could result in draining the Borated Water Storage Tank to the Containment Emergency Sump and the loss of this water source for normal plant shutdown.

Re-energization of valves DHt-9A and DH-9B is permitted on an intermittent basis during MODES 1, 2, 3, and 4 under administrative controls.

Station procedures identify the precautions which must be taken when re-energizing these valves under such controls.3/4.6.2.2 CONTAIN'MENT COOLING SYSTEM The OPERABILITY of the containment cooling system ensures that 1) the containment air temperature.

will be maintained within limits during normal operation, and 2) adequate heat removal capacity is available when operated in conjunction with the containment spray systems during post-LOCA conditions.

3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment.

Containment isolation within the required time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.Containment isolation valves and their required isolation times are addressed in the USA.R. The opening of a closed inoperable containment isolation valve on an intermitt~nt basis during plant operation is permitted under administrative control. Operating procedures identify.those valves which may be opened under administrative control as well as the safety precautions which must be taken when opening valves under such controls.The containment purge supply and exhaust system isolation valves are considered OPERABLE with respect to containment isolation when they meet the requirements of Specification 3.6.1.7.Technical Specification (TS) 3.6.3.1 applies to all containment isolation valves (as identified in USA.R Table 6.2-23), regardless of whether or not an SFAS automatic closure feature exists for the valve.The OPERABILITY of systems which are required to have an open.flowpath during accident conditions, but which have this flowpath secured due to~irnplementation of a TS 3.6.3.1 Action statement, will be governed by the LCOs for those systems.DAVIS-BESSE, UNIT I B 3/4 6-3 A.mendment No. 147, 182, 195, 221 LAP No. 04-0012 CONTAINMENT SYSTEMS BASES An inoperable containment isolation valve that has been disabled in the closed position should still be declared inoperable and the TS 3.6.3.1 Action statement entered. For example, an inoperable motor-operated containment isolation valve that is closed with power removed meets TS Action 3.6.3. lb, however the TS Action statement should not be exited until the valve is returned to OPERABLE status. Closure of the valve could impact other TS LCOs.3/4.6.4 Deleted 3/4.6.5 SHIELD BUILDING 3/4.6.5.1 EMERGENCY VENTILATION SYSTEM The OPERABILITY of the emergency ventilation systems ensures that containment vessel leakage occurring during LOCA conditions into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere.

This requirement is necessary to meet the assumptions used in the safety analyses and limit the site boundary radiation doses to within the limits of 10 CFR 100 during LOCA conditions.

The proper functioning of the EVS fans, dampers, filters, adsorbers, etc., as a system is verified by the ability of each train to produce the required system flow rate.The required emergency ventilation filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

(DAVIS-BESSE, UNIT 1 B 3/4 6-4 Amendment No. 66, 167, 183,233, 244 Correction Ltr dated 5/17/94 LAR No. 04-0004, 04-0012, 03-0021 PLANT SYSTEMS BASES 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 280'F from normal operating conditions in the event of a total loss of offsite power. The OPERABILITY of the Auxiliary Feed Pump Turbine Inlet Steam Pressure Interlocks is required only for high energy line break concerns and does not affect Auxiliary Feedwater System OPERABILITY.

The Condensate Storage Tanks are the non-safety-related primary source of the water for the Auxiliary Feedwater System. When the auxiliary feedwater pumps are needed and either the Condensate Storage Tanks are not available or have been emptied by the Auxiliary Feedwater System, a safety-related transfer system transfers the suction from the Condensate Storage Tanks to the Service Water System. The Service Water System is the safety-related secondary source of the water and must be available for the associated Auxiliary Feedwater System train to be OPERABLE.

The transfer is initiated upon detection of a low suction pressure at the suction of the auxiliary feedwater pumps by suction pressure interlock switches.

These pressure switches, upon sensing low suction pressure, will automatically transfer the suction of the auxiliary feedwater pumps to the Service Water System. On a sustained low-low suction pressure, additional Auxiliary Feedwater Pump Suction Pressure Interlocks will operate to close the steam supply valves to protect the turbine driven auxiliary feedwater pumps from cavitation.

Both the low and the low-low suction Auxiliary Feed Pump Suction Pressure Interlocks are required to be OPERABLE for OPERABILITY of the associated auxiliary feedwater train.Each steam driven auxiliary feedwater pump is capable of delivering the required feedwater flow at the full open pressure of the Main Steam Safety Valves as assumed in the Updated Safety Analysis Report. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 280'F where the Decay Heat Removal System may be placed in operation.

Each train of auxiliary feedwater must be capable of providing feedwater flow to each steam generator in order to be OPERABLE.However, the design of the system does not provide for feeding both steam generators simultaneously from one train.When conducting tests of an auxiliary feedwater train in MODES 1, 2, or 3 which require local manual realignment of valves that make the train inoperable, a dedicated individual shall be stationed at the valves, in communication with the control room, able to restore the valves to normal system OPERABLE status. However, it is not required to have this dedicated individual stationed if the other train of the Auxiliary Feedwater System is OPERABLE and the Motor Driven Feedwater Pump System is OPERABLE pursuant to Technical Specification 3/4.7.1.7 because two sources of auxiliary feedwater to the steam generators are OPERABLE.

In either situation, the Auxiliary Feedwater System train with the local manual realigned valves is inoperable and the Limiting Condition for Operation ACTION must be followed.Closure of valve AF 599 or AF 608 will render both trains of the Auxiliary Feedwater System and the Motor Driven Feedwater Pump System inoperable.

This is because closure of these valves Would result in a complete loss of auxiliary feedwater to the steam generators for certain postulated.feedwater line, and steam line breaks.(3/4.7.1.2 is continued on page B 3/4 7-2.)DAVIS-BESSE, UNIT I B 3/4 7-lb Amendment No. 117, 122, 131,153, 193, 200, LAR 04-0018 Next page is B 3/4 7-2 PLANT SYSTEMS BASES 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM (Continued)

Following any modifications or repairs to the Auxiliary Feedwater System piping from the Condensate Storage Tank through auxiliary feed pumps to the steam generators that could affect the system's capability to deliver water to the steam generators, following extended cold shutdown, a flow path verification test shall be performed.

This test may be conducted in MODES 4, 5 or 6 using auxiliary steam to drive the auxiliary feed pumps turbine to demonstrate that the flow path exists from the Condensate Storage Tank to the steam generators via auxiliary feed pumps.Verification of the turbine plant cooling water valves (CW 196 and CW 197), the startup feedwater pump suction valves (FW 32 and FW 91), and the startup feedwater pump discharge valve (FW 106) in the closed position is required to address the concerns associated with potential pipe failures in the auxiliary feedwater pump rooms, that could occur during operation of the startup feedwater pump.Exceptions to Specification 4.0.4 are provided for Surveillance Requirements 4.7.1.2.1 .a. 1, 4.7.1.2.1.c.2, and 4.7.1.2.1.g.1 for entry into MODE 3. Upon entering MODE 3, the requirements of Specification 4.0.3 are immediately applicable.

Surveillance Requirements provided with exceptions to Specification 4.0.4 that have exceeded their required interval may be treated as missed surveillances in accordance with Specification 4.0.3. The time of discovery of the missed surveillance should be considered the time at which MODE 3 was entered. All other Surveillance Requirements, to which Specification 4.0.4 is applicable, must have been performed within the required surveillance interval or as otherwise specified prior to entry into MODE 3.In all cases, reasonable assurance of Auxiliary Feedwater System OPERABILITY must exist prior to entry into MODE 3 including a reasonable expectation that any Surveillance Requirements provided with exceptions to Specification

4.0.4 which

may not have been performed within the time interval required by Specification 4.0.2 will successfully demonstrate OPERABILITY when performed.

3/4.7.1.3 CONDENSATE STORAGE TANKS The OPERABILITY of the Condensate Storage Tanks with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> with steam discharge to atmosphere and to cooldown the Reactor Coolant System to less than 2800 under normal conditions (i.e., no loss of offsite power). The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant offsite radiation does will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.DAVIS-BESSE, UNIT 1 B 3/4 7-2 Amendment No. 96, 122, Ltr. 5/1/91, 164 LAR No. 04-0027 PLANT SYSTEMS BASES 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1)minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment.

The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the safety analyses.3/4.7.1.6 SECONDARY WATER CHEMISTRY

-Deleted 3/4.7.1.7 MOTOR DRIVEN FEEDWATER PUMP SYSTEM The OPERABILITY of the Motor Driven Feedwater Pump System ensures that the Reactor Coolant System can be cooled down from normal operating conditions in the event of the total loss of Main Feedwater and Auxiliary Feedwater Pumps.The Motor Driven Feedwater Pump System must be capable of providing feedwater flow to each steam generator in order to be OPERABLE.The Motor Driven Feedwater Pump flow capability ensures that adequate feedwater flow is available to remove Decay Heat and reduce the Reactor Coolant System temperature to where the Decay Heat System may be placed into operation.

When conducting tests of the Motor Driven Feedwater Pump System in MODE 1 at greater than 40% RATED THERMAL POWER which requires local manual realignment of valves which make the system inoperable, a dedicated individual shall be stationed at the realigned train's valves, in communication with the control room, able to restore the valves to normal system OPERABLE status. However, it is not required to have this dedicated individual stationed if both trains of the Auxiliary Feedwater System are OPERABLE pursuant to Technical Specification 3/4.7.1.2 because two sources of auxiliary feedwater to the steam generators are OPERABLE.

In either situation, the Motor Driven Feedwater Pump System with the local manual realigned valves is inoperable and the Limiting Condition for Operation ACTION must be followed.When at 40% RATED THERMAL POWER or less and in MODES 1, 2, or 3, the Motor Driven Feedwater Pump System may be aligned to provide a flow path from the Deaerator Storage Tank through the Motor Driven Feedwater Pump to the Main Feedwater System. During this Motor Driven Feedwater Pump mode of operation, a flow path from the Condensate Storage Tanks through the Motor Driven Feedwater Pump to the Auxiliary Feedwater System shall be maintained with the ability for manual positioning

• of valves such that the flow path can be established.

The ability for local, manual operation is demonstrated by verifying the presence of the handwheels for all manual valves and the presence of either handwheels or available power supply for motor operated valves.DAVIS-BESSE, UNIT 1 B 3/4 7-3 Amendment No. 103, 135, 193, 200, 246 LAR No. 04-0027 3/4.8 ELECTRICAL POWER SYSTEMS BASES The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety related equipment required for 1) the safe shutdown of the facility and 2) the mitigation and control of accident conditions within the facility.

The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix "A" to 10 CFR 50.Qualified offsite to onsite circuits are those that are described in the USAR and are part of the licensing basis for the plant.An OPERABLE qualified offsite to onsite circuit consists of all breakers, transformers, switches, interrupting devices, cabling, and controls required to transmit power from the offsite transmission network to the onsite Class I E essential buses.An OPERABLE qualified offsite to onsite circuit consists of: 1. One OPERABLE 345 kV transmission line 2. One OPERABLE 345 -13.8 kV startup transformer, or an OPERABLE main transformer and unit auxiliary transformer with the generator links removed ("backfeed" alignment), as described below 3. One OPERABLE 13.8 kV bus, and 4. One OPERABLE 13.8 -4.16 kV bus tie transformer as described below.An OPERABLE qualified circuit from the transmission line through the main transformer and unit auxiliary transformer exists when the following conditions are met: 1. The plant is in MODE 3, 4, 5, or 6.2. Each OPERABLE 13.8 kV bus is powered from the unit auxiliary transformer, and 3. If a startup transformer is OPERABLE, each OPERABLE 13.8 kV bus is configured to permit automatic transfer to an OPERABLE 345 -13.8 kV startup transformer.

Typically, the electrical power reserve source selector switches are selected to the two different startup transformers.

However, under certain conditions it is appropriate to select both switches to the same startup transformer.

The circuit in which the startup transformer does not have a reserve source selector switch pre-selected to it must still meet the requirements of having its 345 kV transmission line, startup transformer, 13.8 kV bus and bus tie transformer OPERABLE (unless backfeeding through the unit auxiliary transformer).

DAVIS-BESSE, UNIT I B 3/4 8-1 Amendment No. 100, 203 LAR 03-0018, 04-0002, 05-0008 3/4.8 ELECTRICAL POWER SYSTEMS BASES In the case where a 13.8 kV bus is powered from a startup transformer, the reserve source selector switch should be selected to the opposite startup transformer (when that transformer is OPERABLE).

In MODES 1, 2, 3, and 4, additional restrictions apply to the configuration of the electrical power distribution system to ensure that adequate voltage is available for each of the required loads. Any time less than two 345 kV -13.8 kV transformer circuits or less than two 13.8 kV -4.16 kV transformers are OPERABLE, at least one qualified offsite to onsite circuit is not OPERABLE, and the appropriate ACTION statement must be entered: Number of Number of Number of Number of OPERABLE OPERABLE OPERABLE OPERABLE 345 kV -13.8 kV 13.8 kV -4.16 kV Qualified Offsite Qualified Offsite Transformer Bus Tie to Onsite Circuits to Onsite Circuits Circuits Transformers MODES 1, 2, 3, 4 MODES 5, 6 2 1 1 1 2 0. 0 0 1 2 1 1 1 0 1 0 0 0 0 0,1,2 0 0 The essential 4.16 kV buses remain OPERABLE while energized with one 13.8 kV -4.16 kV bus tie transformer inoperable.

The electrical circuit created between essential 4.16 kV buses Cl and D 1, when both buses are powered off the same bus tie transformer, is not a bus tie under LCO 3.8.2.1. Therefore, the breakers supporting this circuit are not "tie breakers" and need not be open in order to satisfy the requirements of LCO 3.8.2.1.The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation.

The OPERABILITY of the power sources are consistent with the. initial condition assumptions of the safety analyses and are based upon maintaining at least one of each of the onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure of the other onsite A.C. source.DAVIS-BESSE, UNIT I B 3/4 8-2 Amendment No. 203, LAR 03-0018, 04-0002, 05-0008 3/4.8 ELECTRICAL POWER SYSTEMS BASES Surveillance Requirement 4.8. 1.11 .b is performed at least once each REFUELING INTERVAL during shutdown by (1) demonstrating the capability of transferring (both manually from the control room and automatically) each 13.8 kV bus power supply from the unit auxiliary transformer to each startup transformer circuit, and from each startup transformer circuit to the other startup transformer circuit, and (2) demonstrating the capability of transferring manually from each startup transformer circuit to the circuit through the main transformer and unit auxiliary transformer.

Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.c.4 verify proper starting of the Emergency Diesel Generators from standby conditions.

Verification that an Emergency Diesel Generator has achieved a frequency of 60 Hz within the required time constraints meets the requirement for verifying the Emergency Diesel Generator has accelerated to 900 RPM.Surveillance Requirements (SR) 4.8.1.1.2.a.7 and 4.8.1.1.2.c.7 verify that the automatic load sequence timer is operable with each load sequence time within 10% of its required value. In addition to equipment directly actuated by the Safety Features Actuation System Sequence Logic Channels, the operating Makeup Pump(s) will trip following a Loss of Offsite Power, and restart approximately

2.5 seconds

after the associated Emergency Diesel Generator output breaker closes. SR 4.8.1.1.2.a.7 and 4.8.1.1.2.c.7 are applicable to the 2.5 second delay.NRC Log Number 5668, dated May 31, 2000, provides guidance relative to the operability of the offsite A.C. electrical power sources. In summary, whenever switchyard equipment is removed from service or switchyard breakers are opened that leaves the remaining switchyard equipment vulnerable to a single point failure that would result in a loss of offsite power, TS 3.8.1.1 Action a must be entered and the appropriate actions taken as specified.

For example, with either the Lemoyne line or the BayShore line out of service, the remaining two circuits are susceptible to a single event, and TS 3.8.1.1 Action a entry is appropriate.

With the Ohio Edison line out of service, unless startup transformer No. 2 is also out of service, TS 3.8.1.1 Action a entry is not required.The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that 1) the facility can be maintained in the shutdown or refueling condition for extended time periods and 2) sufficient instrumentation and control capability is available for monitoring and maintaining the facility status.DAVIS-BESSE, UNIT 1 B 3/4 8-3 Amendment No. 100, 203, LAR 03-0003, 04-0002, 04-0014, 05-0008 3/4.8 ELECTRICAL POWER SYSTEMS BASES In Modes 1-4, a single 120V A.C. Vital Instrument Bus (VIB) may be energized from a non-inverter supplied power source (either the essential alternate or non-essential alternate) for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This condition is only acceptable if all other 120V A.C.VIBs are operable and powered from their associated inverter supplied Class lE power sources. In Modes 5 and 6, at least one 120V A.C. VIB (Y1 or Y2) must be energized from an essential power source (from the inverter or the essential alternate).

This power source must be 'in the operable A.C./D.C.

train. The remaining required 120V A.C. VIBs may be energized from any power source.A May 30, 1984, NRC internal memorandum from the NRC Director, Division of Licensing, Office of Nuclear Reactor Regulation (D. G. Eisenhut to C. E. Norelius, EXT 04-00414) stated, "We conclude that Action Statement 3.8.2.1 of the Davis-Besse Technical Specifications is not invoked when the AC Vital Buses are powered from non Class 1E power sources." The Surveillance Requirements for demonstrating the OPERABILITY of the station batteries are based on the recommendations of Regulatory Guide 1.129, "Maintenance, Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants", February 1978, and IEEE Std. 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead -Acid Batteries for Stationary Applications," except that certain tests will be performed at least once each REFUELING INTERVAL.Battery degradation is indicated when the battery capacity drops more than 10% from its capacity on the previous performance discharge or modified performance discharge test, or is below 90% of the manufacturer's rated capacity.Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated capacity.DAVIS-BESSE, UNIT 1 B 3/4 8-4 Amendment No. 100, 219, 229, LAR 04-0013, 05-0008 3/4.8 ELECTRICAL POWER SYSTEMS BASES Table 4.8-1 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and .015 below the manufacturer's full charge specific gravity or a battery charger current of less than two amps is characteristic of a charged cell with adequate capacity.

The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than .020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than .010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery. Exceptions to the specific gravity requirements are taken to allow for the normal deviations experienced after a battery discharge and subsequent recharge associated with a service, performance discharge, or modified performance discharge test. The specific gravity deviations are recognized and discussed in IEEE Std. 450-1995.Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-1 is permitted for up to seven days. During this seven-day period: (1) the allowable value for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than .020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less than the safety rmargin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than.040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.DAVIS-BESSE, UNIT 1 B 3/4 8-5 Amendment No. 100, 158, 229, LAR 05-0008 3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitation on reactivity during REFUELING ensures that: 1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity conlrol in the water volumes having direct access to the reactor vessel. This limitation is consistent with the initial conditions assumed for the boron dilution incident in the accident analysis.The ACTION statement's minimum boration flow rate of 12 gpm is less than the minimum boration flow rate of 25 gpm specified in TS 3/4.1.1.1, Reactivity Control -Shutdown Margin because the lower flow rate is based on only borating the reactor vessel.3/4.9.2 INSTRUMENTATION The OPERABILITY of source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core. The monitors provide continuous visual indication in the control room, and one channel provides audible indication in containment and in the control room.3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products.

This decay time is consistent with the assumptions used in the safety analyses.3/4.9.4 CONTAINMENT PEN"ETRATIONS During CORE ALTERATIONS or movement of irradiated fuel within the containment, release of fission product radioactivity to the environment as a result of a fuel element rupture must be minimized.

During MODES 1, 2, 3, and 4, this is accomplished by maintaining CONTAINMENT INTEGRITY as described in LCO 3.6.1.1. In other situations, the potential for containment pressurization as a result of an accident is not present, and therefore less stringent requirements are needed to isolate the containment from the atmosphere outside containment.

Both containment personnel air lock doors may be open during CORE ALTERATIONS or during movement of irradiated fuel within the containment provided the conditions specified in LCO 3.9.4.b are met. The individual designated to be continuously available to close the air lock door must be stationed at the auxiliary building side of the air lock. A containment personnel air lock door is considered capable of being closed if the door is not blocked in such a way that it cannot be expeditiously closed, and any hoses and cables running through the airlock employ a means to allow safe, quick disconnect or severance, and are tagged at the airlock with specific instructions to expedite removal. The LCO 3.9.10 requirement to maintain a minimum of 23 feet of water over the top of irradiated fuel assemblies seated within the reactor pressure vessel during movement of fuel assemblies within the reactor pressure vessel while in MODE 6 ensures that sufficient water depth is available to remove 99% of the assumed iodine gap activity released from the rupture of an irradiated fuel assembly.

Further, sufficient time is available to close the personnel air lock following a loss of shutdown cooling before boiling occurs.DAVIS-BESSE, UNIT I B 3/4 9-1 Amendment 186, 202, 207, Revised by NRC letter dated March 19, 199H LAR 02-0006 REFUELING OPERATIONS Revised by NRC letter dated August 12, 1994 BASES 3/4.9.6 FUEL HANDLING BRIDGE OPERABILITY The OPERABILITY requirements of the hoist bridges used for movement of fuel assemblies ensures that: .1) fuel handling bridges will be used for movement of control rods and fuel assemblies, 2) each hoist has sufficient load capacity to lift a fuel element, and 3) the core intemals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL -FUEL HANDLING BUILDING Deleted 3/4.9.8 COOLANT CIRCULATION The requirement that at least one decay heat removal loop be in operation ensures that (1)sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140'F as required during the REFUELING MODE, and (2)sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution incident and prevent boron stratification.

The requirement to have two DHR loops OPERABLE when there is less than 23 feet of water above the core ensures that a single failure of the operating DHR loop will not result in a complete loss of decay heat removal capability.

With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating DHR loop, adequate time is provided to initiate emergency procedures to cool the core.In MODE 6, the RCS boron concentration is typically somewhat higher than the boron concentration required by Specification 3.9.1, and could be higher than the boron concentration of normal sources of water addition.

The flowrate through the decay heat system may at times be reduced to somewhat less than 2800 gpm. In this situation, if water with a boron concentration equal to or greater than the boron concentration required by Specification 3.9.1 is added to the RCS, the RCS is assured to remain above the Specification 3.9.1 requirement, and a flowrate of less than 2800 gpm is not of concern.DAVIS-BESSE, UNIT I B 3/4 9-2 Amendment No. 38, 188, 237, 247 LAR No. 02-0003 REFUELING OPERATIONS BASES 3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM Deleted 3/4.9.10 and 3/4.9.11 WATER LEVEL -REACTOR VESSEL AND STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the iodine gap activity released from the rupture of an irradiated fuel assembly.

The minimum water depth is consistent with the assumptions of the safety analysis.3/4.9.12 STORAGE POOL VENTILATION The requirements on the emergency ventilation system servicing the storage pool area to be operating or OPERABLE ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere.

The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the safety analyses.Specification 3.9.12 permits an emergency ventilation system servicing the storage pool that is incapable of meeting the acceptance criteria of Surveillance Requirement 4.9.12.1 solely because the containment equipment hatch is open and both doors of the containment personnel air lock are open to be considered OPERABLE provided at least one personnel air lock door is capable of being closed and a designated individual is available immediately outside the personnel air lock to close the door. When the containment equipment hatch is open and both doors of the containment personnel air lock are open, the emergency ventilation system servicing the fuel storage area is incapable of maintaining a negative pressure of_> 1/8 inches Water Gauge relative to the outside atmosphere during system operation.

The requirement that at least one personnel air lock door be capable of being closed and a designated individual be available immediately outside the personnel air lock to close the door ensures that the negative pressure boundary can be established in a timely mannei following a fuel handling accident in the storage pool area or containment.

Once the negative pressure boundary is established, the emergency ventilation system servicing the storage pool area will be capable of establishing the required negative pressure relative to the outside atmosphere.

3/4.9.13 SPENT FUEL ASSEMBLY STORAGE The restrictions on the placement of fuel assemblies within the spent fuel pool as dictated by Figure 3.9-1 ensure that the k-effective of the spent fuel pool will always remain less than 0.95 assuming the spent fuel pool to be flooded with non-borated water. The restrictions delineated in Figure 3.9-1 and the action statement, are consistent with the criticality safety analyses performed for the spent fuel pool.The criticality analyses qualify the high density rack modules for storage of fuel assemblies in one of three different loading patterns, subject to certain restrictions:

Mixed Zone Three Region, Checkerboard, and Homogeneous Loading. Figure 3.9-1 provides the Category-specific burnup/enrichment limitations.

Different loading patterns may be used in different rack modules, provided each rack module contains only one loading pattern. Two different loading patterns may be used in a single rack module, subject to certain additional restrictions.

The loading pattern restrictions are maintained in fuel handling administrative procedures.

The design features of the high density spent fuel storage racks are described in Specification 5.6.1.2.DAVIS-BESSE, UNIT I B 3/4 9-3 Amendment No. 130, 135, 186, 237, 247, 251 LAR No. 02-0003