Text

Proposed Tech Specs,Submitting Revised Copy of TS Bases Which Include Minor Changes & Clarifications Made Per Requirements of 10CFR50.59
	ML17291B249
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Site:	Columbia
Issue date:	02/26/1996
From:	; WASHINGTON PUBLIC POWER SUPPLY SYSTEM
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ML17291B248	List: ML17291B247; ML17291B249;
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NOTE The BASES contained in succeeding pages summarize the reasons for the Specifications in Section 2.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications.

9603050453 960226 PDR ADQCK 05000397 P PDR

BASES FOR

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SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2.0 SAF TY LIMITS and LIMITING SAFETY SYSTEM SETTINGS BASES INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.07 for two recirculation loop operation and 1.08 for single recirculation loop operation for all nuclear fuel in WNP-2. MCPR greater than 1.07 for two recirculation loop operation and 1.08 for single .recirculation loop operation represents 'a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.

The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. 'While'ission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding integrity Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity safety limit assures that during normal operation and during anticipated operational occurrences, at least 99.9 percent of the fuel rods in the core do not experience transition boiling (

Reference:

ANF-524(P)(A), Rev. 2); ABB Atom Report UK90-126; GEll Lead Fuel Assembly Report for Washington Public Power Supply System Nuclear Project No. 2, Reload 5, Cycle 6). The latter two references support application of the above established safety limit to GEll and SVEA-96 LFA fuel in WNP-2.

2.1 SAF Y LIMITS

2. 1. I THERMAL POW R Low Pressure or ow Flow For certain conditions of pressure and flow, the ANFB correlation is not valid for all critical power calculations. The ANFB correlation is not valid for bundle mass velocities less than 0. 10 x 10'bs/hr-ft or pressures less than 590 psia. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following, basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pr essure drop at low power WASHINGTON NUCLEAR- UNIT 2 B 2-1 Revision No. 0

SAF Y MITS BASES THERMAL POW R ow Pressure or Low Flow (Continued) and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10 lbs/h (approximately a mass velocity of 0.25 x 10 lbs/hr-ft ), bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 10'bs/h. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50/ of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25K of RATED THERMAL POWER for reactor pressure below.,785.psig.-is-conservative"."- -'"'.

1.2 TH RMA POW i h Pressure and Hi h low The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occu~ if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9X of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is determined using the ANF Critical Power Methodology for boiling water reactors" which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the ANF nuclear critical heat fluxenthalpy ANFB correlation.

The ANFB correlation is valid over the range of, conditions used in the tests of the data used to develop the correlation.

a. Advanced Nuclear Fuels Critical Power Methodology for Boiling Water Reactors, ANF-524(P)(A), Rev. 2.

WASHINGTON NUCLEAR UNIT 2 B 2-2 Revision No. 0

SAF Y L ITS BASES TH RMA POWER ow Pressure or Low Flow (Continued)

The required input to the statistical model are the uncertainties listed in Bases Table B2.1.2-1.

The bases for the reactor system and fuel uncertainties are given in ANF-524(P)(A), Rev. 2'". The power distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power levels. The worst distribution during any fuel cycle would not be as severe as the distribution used in the analysis.

2. 1.3 REACTOR COOLANT SYST PRESSURE The Safety Limit for the reactor coolant system pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel Code 1971 Edition, including Addenda through Summer 1971, which permits a maximum pressure transient of 1105, 1375 psig, of design pressure, 1250 psig. The Safety Limit of 1325 psig, as measured by the reactor vessel steam dome pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The reactor coolant system is designed to the ASME Boiler and Pressure Vessel Code, 1971 Edition, including Addenda through Winter 1971 for the reactor recirculation piping, which permits a maximum. pressure transient of 125%, 1565 psig, of design pressure, 1250 psig for suction piping and 1550 psig for discharge piping. The pressure Safety Limit is selected to be the lowest transient overpressure allowed by the applicable Codes.

. 1.4 REACTOR V SS L WAT R LEVE With fuel in the reactor vessel during periods when the reactor is shutdown, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level became less than two-thirds of the core height. The Safety Limit has been established at the top of the active irradiated fuel to provide a point which can be monitored and also provide adequate margin for effective action.

a. Advanced Nuclear Fuels Critical Power Methodology for Boiling Water Reactors, ANF-524(P)(A), Rev. 2.

WASHINGTON NUCLEAR UNIT 2 8 2-3 Revision No. 0

BASES TABLE B 2.1.2-1 UNC RTAINTIES CONSIDERED IN TH MCPR SAFETY IMIT STANDARD Para et DEVIATION*

Feedwater Flow Rate .0176 Feedwater Temperature .0076 Core Pressure .0050 Total Core Flow Rate .0250 Assembly Flow Rate .0280 Power Distribution:

Radial Assembly Power .0409 Local Power~* .0229 ANFB Correlation Additive Constants 8x8 FUEL .0200

. 9x9-2 FUEL .0200 9x9-9X FUEL .0080

Fraction of Nominal Value.

~Relative Local Rod Power.

WASHINGTON NUCLEAR UNIT 2 B 2-4 Revision No. 0

Sa T M TS BASES 2.2 IMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR PROT CTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System instrumentation setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each parameter.

The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist in mitigating the consequences of accidents. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference. between, each Trip Setpoint and the Allowable Valiie is equal to or less than the drift allowance assumed for each trip in the safety analyses.

l. Intermediate R n e Monitor Neutron Flux - Hi h The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges. Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up. The IRM instruments provide for overlap with both the APRM and SRM systems.

The most significant source of reactivity. changes during the power increase is due to control rod withdrawal. In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed. The results of these analyses are in Section 15.4 of the FSAR. The most severe case involves an initial condition in which THERMAL POWER is at approximately 1% of RATED THERMAL POWER. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the control rod being withdrawn is bypassed.

The results of this analysis show that the reactor is shutdown and peak power is limited to 21% of RATED THERMAL POWER with the peak fuel enthalpy well below the fuel failure threshold of 170 cal/gm. Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

2. Avera e Power Ran e Monitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 151 of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Temperature coefficients are WASHINGTON NUCLEAR UNIT 2 B 2-5 Revision No..O

LIMITING SAFETY SYS EM SETTINGS BASES R ACTOR PROTECT ON SYST M INSTRUMENTATION S POINTS (Continued)

Avera e Power Ran e Monitor (Continued) small and control rod patterns are constrained by the RSCS and RWM. Of all the possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase.

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission In. an'. assumed uniform-rod-withdrawal- approach'o the trip level, 'ate.,

the rate of power rise is not more than 51 of RATED THERMAL PO'WER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit. The 15/

neutron flux trip remains active until the mode switch is placed in the Run position.

The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Fixed Neutron Flux-High setpoint; i.e, for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel. For the Flow'iased Simulated Thermal Power-High setpoint, a time constant of 6 il seconds is introduced into the flow biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-).

The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown. The flow referenced trip setpoint must be adjusted by the specified formula in Specification 3.2.2 in order to maintain these margins when the design TOTAL PEAKING FACTOR is exceeded (MFLPD is greater than or equal to FRTP).

3. Reactor Vessel Steam Dome Pressure-Hi h High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity." The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and WASHINGTON NUCLEAR UNIT 2 B 2-6 Revision No. 0

LIMITING SA ETY SYSTEM SETTINGS BASES REACTOR ROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Reactor Vessel Steam Oome Pressure-Hi h (Continued) takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power/flow conditions when the turbine control valve fast closure and turbine. stop valve closure trips are bypassed. For a load rejection or a turbine trip under these conditions, the transient 'analysis indicated an adequate margin to the thermal hydraulic limit.

4. Reactor'Vessel'ater"'Lev'el=Low' The reactor vessel water level trip setpoint was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection for the fuel and pressure limits.

5. in Steam ine solation Valve-Closure, The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain, postulated events.

The MSIV's are closed automatically from measured parameters such as high steam flow, low reactor water level, high steam tunnel temperature, and low steam line pressure. The MSIV's closure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal/hydraulic Safety Limits.

6. DELETED

7. Primar Containment Pressure-Hi h High pressure in the drywell could indicate a break in the primary pressure boundary systems. The reactor is tripped in order to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant. The trip setting was selected as low as possible without causing spurious trips.

WASHINGTON NUCLEAR UNIT 2 B 2-7 Revision No. I

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECT ON SYST M INSTRUMENTATION SETPOINTS (Continued)

8. Scram 0'sc ar e Vo ume Water Level-Hi h The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram. Should this volume fill up to a point where there is insufficient volume to accept the displaced water at pressures below 65 psig, control rod insertion would be hindered. The reactor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the .volume is still,great enough to accommodate the water from. the. movement-of--the-rods-at- pressures- below"65 psig when they are tripped. The scram discharge volume high level alarm setpoint for scram discharge volume 'A'525' 1/2" elevation) provides 87.1 gallons of margin above the required 617.9 gallons of free volume required for a reactor scram. The scram discharge volume high level alarm setpoint for scram discharge 'B'524'", elevation) provides 91.3 gallons of margin above the required 617.9 gallons of free volume required for a reactor scram. The rod block setpoint for scram discharge volume 'A'nd elevation) provides 89.6 gallons of margin above the required

'B'527'"

617.9 gallons of free volume required for a reactor scram. The scram setpoint for scram discharge volume 'A'nd 'B'529'" elevation) provides 64.9 gallons of margin above the required 617.9 gallons of free volume for a reactor scram.

9. Turbine Sto Valve-Clos re The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves. With a trip setting of 5f. of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained during the worst case transient assuming the turbine bypass valves fail to operate and an RPT occurs.

IO. Turbine Control Valve Fast Closure Tri Oil Pressure-Low The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection with or without coincident failure of the turbine bypass valves. The Reactor Protection System initiates a trip when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than 30 milliseconds after the start of control valve fast closure. This is achieved by the action of the fast acting solenoid valves in rapidly r educing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves. This loss of pressure is sensed by pressure WASHINGTON NUCLEAR UNIT 2 B 2-& Revision No. 0

IHITING SAF Y SYSTEH SETTING BASES REACTOR PROT CT ON SYST H INSTRUH NTATION SETPOINTS (Continued)

Turbine Control Valve Fast Closure Tri Oil Pressure-Low switches whose contacts form the one-out-of-two-twice logic input to the Reactor Protection System. This trip setting, a faster closure time, and a different valve characteristic from that of the turbine stop valve, combine to produce transients which are very similar to that for the stop valve. Relevant transient analyses are discussed in Section 15.2.2 of the Final Safety Analysis Report.

11. Reactor Hode Switch"Shutdow "Position""- "

The reactor mode switch Shutdown position is a redundant channel to the automatic protective instrumentation channels and provides additional manual reactor trip capability.

12. Hanual Scram The manual scram is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

WASHINGTON NUCLEAR UNIT 2 B 2-9 Revision No. 0

"- BASES-FOR<<

SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

NOTE The BASES contained in succeeding pages summarize the reasons for the Specifications in Section 3.0 and 4.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications.

3 4 LIMITING CONDITIONS FOR OPERATION AND SURV ILLANCE RE UIREMENTS 3 4.0 APP ICABILITY BC'.::S S ecifications 3.0. 1 throu h 3.0.4 establish the general requirements applicable to Limiting Conditions for Operation. These requirements are based on the requirements for Limiting Conditions for Operation stated in the Code of Federal Regulations, 10 CFR 50.36(c)(2):

"Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the techni'cal speciffc'atio'n 'uritil"'the'condition can be met."

S ecif'cation 3.0.1 establishes the Applicability statement within each individual specification as the requirement for when (i.e., in which OPERATIONAL CONDITIONS or other specified conditions) conformance to the Limiting Conditions for Operation is required for safe operation of the facility. The ACTION requirements establish those remedial measures that must be taken within specified time limits when the requirements of a Limiting Condition for Operation are not met. It is not intended that the shutdown ACTION requirements be used as an operational convenience which permits (routine) voluntary removal of a system(s) or component(s) from service in lieu of other alternatives that would not result in redundant systems or components being inoperable.

There are two basic types of ACTION requirements. The first specifies the remedial measures that permit continued operation of the facility which is not further restricted by the time limits of the ACTION requirements. In this case, conformance to the ACTION requirements provides an acceptable level of safety for unlimited continued operation as long as the ACTION requirements continue to be met. The second type of ACTION requirement specifies a time limit in which conformance to the conditions of the Limiting Condition for Operation must be met. This time limit is the allowable outage time to restore an inoperable system or component to OPERABLE status or for restoring parameters within specified limits. If these actions are not completed within the allowable outage time limits, a shutdown is required to place the facility in an OPERATIONAL CONDITION or other specified condition in which the specification no longer applies.

The specified time limits of the ACTION requirements are applicable from the point in time it is identified that a Limiting Condition for Operation is not met. The time limits of the ACTION requirements are also applicable when a system or component is removed from service for surveillance testing or investigation of operational problems. Individual specifications may include a specified time limit for the completion of a Surveillance Requirement when equipment is removed from service. In this case, the allowable outage time WASHINGTON NUCLEAR UNIT 2 B 3/4 0-1 Revision No. 0

IA TY BASES (Continued) limits of the ACTION requirements are applicable when this limit expires if the surveillance has not been completed. When a shutdown is required to comply with ACTION requirements, the plant may have entered an OPERATIONAL CONDITION in which a new specification becomes applicable. In this case, the time limits of. the ACTION requirements would apply from the point in time that the new specification becomes applicable if the requirements of the Limiting Condition for Operation are not met.

S ecif cation 3.0.2 establishes that noncompliance with a specification exists when the requirements of the Limiting Condition for Operation are not met and

,the associated ACTION requirements have not been implemented within the specified time interval... The,purpose of this-specification-is'o clarify that (I) implementation of the ACTION requirements within the specified time interval constitutes compliance with a specification and (2) completion of the remedial measures of the ACTION requirements is not required when compliance with a Limiting Condition for Operation is restored within the time interval specified in the associated ACTION requirements.

I' ecification 3.0.3 establishes the shutdown ACTION requirements that must be implemented when a Limiting Condition for Operation is not met and condition is not specifically addressed by the associated ACTION requirements. The purpose of this specification is to delineate the time limits for placing the unit in a safe shutdown CONDITION when plant operation cannot be maintained within the limits for safe operation defined by the Limiting Conditions for Operation and its ACTION requirements. It is not intended to be used as an operational convenience which permits {routine) voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. One hour is allowed to prepare for an orderly shutdown before initiating a change in plant operation. This time permits the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to reach lower CONDITIONS of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the cooldown capabilities of the facility assuming only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the primary coolant system and the potential. for a plant upset that could challenge safety systems under conditions for which this specification applies.

If remedial measures permitting limited continued operation of the facility under the provisions of the ACTION requirements are completed, the shutdown may be terminated. The time limits of the ACTION requirements are applicable from the point in time there was a failure to meet a Limiting Condition for Operation. Therefore, the shutdown may be terminated if the ACTION requirements have been met or the time limits of the ACTION requirements have not expired, thus providing an allowance for the completion of the required actions.

WASHINGTON NUCLEAR UNIT 2 B 3/4 0-2 Revision No. 0

APPLICABI ITY BASES (Continued)

The time limits of Specification 3.0.3 allow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months for the plant to be in COLD SHUTDOWN when a shutdown is required during POWER operation. If the plant is in a lower CONDITION of operation when a shutdown is required, the time limit for reaching the next lower CONDITION of operation applies.

However, if a lower CONDITION of operation is reached in less time than allowed, the total allowable time to reach COLD SHUTDOWN, or other OPERATIONAL condition, is not reduced. For example, if STARTUP is reached in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the time allowed to reach HOT SHUTDOWN is the next Il hours because the total time to reach HOT SHUTDOWN is not reduced from the allowable limit of 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months .

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 CONDITION of operation in less than the total time allowed.

The same principle applies with regard to the allowable outage time limits of the ACTION requirements, if compliance with the ACTION requirements for one specification results in entry into an OPERATIONAL CONDITION or condition of oper ation for another specification in which the requirements of the Limiting Condition for Operation are not met. If the new specification becomes applicable in less time than specified, the difference may be added to the allowable outage time limits of the second specification. However, the allowable outage time limits of ACTION requirements for a higher CONDITION of operation may not be used to extend the allowable outage time that is applicable when a Limiting Condition for Operation is not met in a lower CONDITION of operation.

The shutdown requirements of Specification 3.0.3 do not apply in CONDITIONS 4 and 5, because the ACTION requirements of individual specifications define the remedial measures to be taken.

S ecification 3.0.4 provides that entry into an OPERATIONAL 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 ensure that unit operation is not initiated with either required equipment or systems inoperable or other 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.

WASHINGTON NUCLEAR UNIT 2 B 3/4 0-3 Revision No. 0

PP CAB L TY BASES (Continued)

S ecifications 4.0. 1 throu h 4.0.5 establish the general requirements applicable to Surveillance Requirements. These requirements are based on the Surveillance Requirements stated in the Code of Federal Regulations, 10 CFR 50.36(c)(3):

"Surveillance requirements are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions of operation will be met.

S ecific tion 4.0. 1 establishes the requirement that surveillances must be performed during the OPERATIONAL .CONDITIONS or..other.'conditions for-which the requirements of"the"'Limi*ting Conditions for Operation apply unless otherwise stated in an individual Surveillance requirement. The purpose of this specification is to ensure that surveillances are performed to verify the operational status of systems and components and that parameters are within specified limits to ensure safe operation of the facility when the plant is in an OPERATIONAL CONDITION or other specified condition for which the individual Limiting Conditions for Operation are applicable. Surveillance Requirements do not have to be performed when the facility is in an OPERATIONAL CONDITION for which the requirements of the associated Limiting Condition for Operation do not apply unless otherwise specified. The Surveillance Requirements associated with a Special Test Exception are only applicable when the Special Test Exception is used as an allowable exception to the requirements of a specification.

S ecification 4.0.2 establishes the conditions under which the specified time interval for Surveillance Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. It also provides flexibility to accommodate the length of a fuel cycle for surveillances that are performed at each refueling outage and are specified with an 18 month surveillance interval. It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for surveillances that are not performed during refueling outages. The limitation of Specification 4.0.2 are based on engineering judgement and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. Thi" provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly .degraded beyond that obtained from the specified surveillance interval.

MASHINGTON NUCLEAR UNIT 2 B 3/4 0-4 Revision No. 0

APPLICABILITY BASES (continued)

S ecification 4.0.3 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 they are found or known to be inoperable although still meeting the Surveillance Requirements. This specification also clarifies that the ACTION requirements are applicable when Surveillance Requirements have not been completed within the allowed suryeil.lance,interval and that, the time limits of theACTION 're'quirements apply from the point in time it is identified that a surveillance has not been performed and not at the time that the allowed surveillance interval was exceeded. Completion of the Surveillance requirement within the allowable outage time limits of the ACTION requirements restores compliance with the requirements of Specification 4.0.3.

However, this does not negate the fact that the failure to'ave performed the surveillance within the allowed surveillance interval, defined by the provisions of Specification 4.0.2, was a violation of the OPERABILITY requirements of a Limiting Condition for Operation that is subject to enforcement action. Further, the failure to perform a surveillance within the provisions of Specification 4.0.2 is a violation of a Technical Specification requirement and is, therefore, a reportable event under the requirements of 10 CFR 50.73(a)(2)(i)(B) because it is a condition prohibited by the plant's Technical Specifications.

If the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or a shutdown is required to comply with ACTION requirements, e.g.,

Specification 3.0.3., a 24-hour allowance is provided to permit a delay in implementing the ACTION requirements. This provides an adequate time limit to complete Surveillance Requirements that have not been performed. The purpose of this allowance is to permit the completion of,a surveillance before a shutdown would be required to comoly with ACTION requirements or before other remedial measures would be required that may preclude the completion of a surveillance. The basis for this allowance includes consideration for plant conditions, adequate planning, ava:.lability of personnel, the time required to perform the surveillance, and the -afety significance of the delay in completing the required surveillance. This provision also provides a time limit for the completion of Surveillance Requirements that become applicable as a consequence of CONDITION changes imposed by ACTION requirements and for completing Surveillance Requirements that are applicable when an exception to the requirements of Specification 4.0.4 is allowed. If a surveillance is not completed within the 24-hour allowance, the time limits of the ACTION requirements are applicable at that time. When a surveillance is performed within the 24-hour allowance and the Surveillance Requirements are not met, the time limits of the ACTION requirements are applicable at the time that the surveillance is terminated.

WASHINGTON NUCLEAR UNIT 2 B 3/4 0-5 Revision No. 0

APP ICAB TY BASES (Continued)

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.

S ecification 4.0.4 establishes the requirement that all applicable surveillances must be met before entry into an OPERATIONAL CONDITION or other condition of operation specified in the Applicability statement. The purpose of this specification is to ensure that system and component OPERABILITY requirements or parameter limits are met before entry into an OPERATIONAL CONDITION or other specified condition for which these systems and components ensure safe. operationof, the'facility-.--This. provision"appli'es to changes in OPERATIONAL CONDITIONS or other specified conditions associated with plant shutdown as well as startup.

Under the provisions of this specification, the applicable Surveillance Requirements must be performed within the specified surveillance interval to assume that the Limiting Conditions for Operation are met during initial plant startup or following a plant outage.

bfhen a shutdown is required to comply with ACTION requirements, the provisions of Specification 4.0.4 do not apply because this would delay placing the facility in a lower CONDITION of operation.

S ecification 4.0.5 establishes the requirement that inservice inspection of ASME Code Class l, 2, and 3 components and inservice testing of ASME Code Class I, 2, and 3 pumps and valves shall be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. This Specification also contains the requirements for the additional inspection program established in Generic Letter 88-01, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping." An alternative schedule to these requirements was provided by the NRC. This alternative schedule allows for Category D 5 E welds to be inspected every three years, as opposed to every two cycles as specified in the Generic Letter. (Letter, JW Clifford (NRC) to GC Sorensen (Supply System), dated January 19, 1993, "Alternate Schedule for IGSCC Inspections (TAC,No. M84714)").

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 applicable Addenda. This clarification is provided to ensure consistency in surveillance intervals throughout the Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.

WASHINGTON NUCLEAR UNIT 2 B 3/4 0-6 Revision No. 0

c~

~PTY BASES (Continued)

Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASHE Boiler and Pressure Vessel Code and applicable Addenda. The requirements of Specification 4.0.4 to perform surveillance activities before entry into an OPERATIONAL CONDITION or other specified condition takes precedence over the ASHE Boiler and Pressure Vessel Code provision that allows pumps and valves to be tested up to one week after return to norma1 operation. The Technical Specification definition of OPERABLE does not allow a grace period before a component, which is not capable of performing its specified function, is declared inoperable and takes precedence over the ASHE Boiler and Pressure Vessel Code provision that allows a valve to be incapable of performing its specified function for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before being declared inoperable...,.

WASHINGTON NUCLEAR UNIT 2 B 3/4 0-7 Revision No. 0

3 4. 1 R ACTIVITY CONTROL SYSTEMS BASES 3 4..1 SHUTDOWN M RGIN A sufficient SHUTDOWN MARGIN ensures that (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be subcritical by at least,R +..0.38K delta..k/k or..R + .0.28% delta. k/k, as appropriate. The value of R in units of f delta k/k is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of R must be positive or zero and must be determined for each fuel loading cycle.

Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN..

The highest worth rod may be determined analytically or by test. The SHUTDOWN MARGIN is demonstrated by an insequence control rod withdrawal at the beginning of life fuel cycle conditions, and, in the cycle if if necessary, at any future time the first demonstration indicates that the required margin could be reduced as a function of exposure., Observation of subcriticality in this condition assures subcriticality with the most reactive control rod fully withdrawn.

This reactivity characteristic has been a basic assumption in the analysis 'of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion.

3 4.1.2 REACT VITY ANOMALIES Since the SHUTDOWN MARGIN requirement is small, a careful check on actual reactor conditions compared to the predicted conditions is necessary. Any changes in reactivity from that predicted (predicted core K,<<) can be determined from the core monitoring system (monitored core K,<<). In the absence of any deviation in plant operating conditions of reactivity anomaly, these values should be essentially equal since the calculational methodologies are consistent. The predicted core K,<< is. calculated by a 3D core simulation code as a function of cycle, exposure. This calculation is performed for projected or anticipated reactor operating states/conditions throughout the cycle and is usually done prior to cycle operation. The monitored core K.<< is the K~ as calculated by the core monitoring system for actual plant conditions.

WASHINGTON NUCLEAR UNIT 2 B 3/4 1-1 Revision No. 0

R ACTIVITY CONTROL SYST MS BASES REACTIV TY ANOMALI S (Continued)

Since the comparisons are easily done, frequent checks are not an imposition on normal operation. A I percent deviation in reactivity from that of the predicted is larger than expected for normal operation and, therefor e, should be thoroughly evaluated. A deviation as large as I percent would not exceed the design conditions of the reactor.

3 4. 1.3 CONTRO RODS The specification of this section ensure that (I) the minimum SHUTDOW MARGIN is maintained,,(2) the, control rod=insertion"time's"ar'e'onsistent with those

'used in the safety analyses, and (3) limit the potential effects of the rod drop accident. The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued oper ation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum.

The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully inserted position are consistent with the SHUTDQMN MARGIN requirements.

The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.

The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than the fuel cladding safety limit during the core wide transient analyzed in the cycle specific transient analysis report. This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than the fuel cladding safety limit. The occurrence of scram times longer then those specified should be viewed as an indication of a systemic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.

'ASHINGTON NUCLEAR-UNIT 2 B 3/4 1-2 Revxsson No. 0

REACTIVITY CONTROL SYSTEMS BASES

~TII I (C ti d)

The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required.

Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with..normal,.drive.,water. pressure.'-.Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.

Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have affected the control rod coupling integrity. The subsequent check is performed as a backup to the initial demonstration.

In order to'ensure that the control rod patterns can be followed and therefore that other parameters are within their limits, the control rod position indication system must be OPERABLE.

The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.

The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.

3 4.1.4 CONTRO ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal/gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 2N'f RATED THERMAL POWER, there is no possible rod worth WASHINGTON NUCLEAR-UNIT 2 B 3/4 1-3 Revision No. 0

REACTIVITY CONTROL SYSTEMS BASES CONTRO ROD PROGRAM CONTROLS (Continued) which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal/gm. Thus requiring the RSCS and RWM to be OPERABLE when THERMAL POWER is less than or equal to 20% of RATED THERMAL POWER provides adequate control.

The RSCS.and RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

Parametric Control Rod Drop Accident analyses have shown that for a wide range of key:-.reactor parameters "(which -envel'ope"th'e"ope'rating ranges of these parameters) the fuel enthalpy rise during a postulated control rod drop accident remains considerably lower than the 280 cal/gm limit. For each operating cycle, cycle-specific parameters such as maximum control rod worth, Doppler coefficient, effective delayed neutron fraction and maximum four-bundle local peaking factor are compared with the inputs to the parametric analyses to determine the peak fuel rod enthalpy rise. This value is then compared against the 280 cal/gm design limit to demonstrate compliance for each operating cycle. If cycle-specific values of the above parameters are outside the range assumed in the parametric analysis, an extension of the analysis or a cycle-specific analysis may be required. Conservatism present in the analysis, results of the parametric studies and a detailed description of the methodology for performing the Control Rod Drop Accident analysis are provided in XN-NF-80-19 Volume l.

The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods.

3 4. 1.5 STANDBY I UID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that none of the withdrawn control rods can be inserted. To meet this objective, it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core. To account for imperfect mixing and leakage, an additional margin of 165 ppm boron equivalent is added. For RHR shutdown dilution, an additional quantity of boron equivalent to 275 ppm is added to ensure the final concentration will not be less than 660 ppm in the reactor core. To achieve this shutdown requii ement, a minimum solution of 4587 gallons containing a minimum of 5500 pounds of sodium pentaborate decahydrate is required. This quantity of sodium pentabor ate decahydrate will provide an undiluted concentration of 1100 ppm of boron in the reactor core.

WASHINGTON NUCLEAR UNIT 2 B 3/4 1-4 Revision No. 0

REACTIVIT CONTRO SYSTE S BASES STANDBY I UID CONTROL SYSTEM (Continued)

A minimum of 41.2 gpm per pump injection rate has been selected to override the reactivity insertion'rate due to cool down and xenon decay.

The minimum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted and the filling of other piping systems connected to the reactor vessel. The temperature requirement on the sodium pentaborate solution is necessary to ensure that the sodium pentaborate remains in solution.

With two pumps and explosive,,injectionvalves..and.,with .a...highly. reliable control r'od scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.

Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron concentration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months assures that the solution is available for use.

Replacement of the explosive charges in the valves at regular intervals will ensure that these valves will not fail because of deterioration of the charges.

For Anticipated Transient Without Scram (ATWS), mitigation requires an equivalent 86 gpm -- 13% sodium pentaborate decahydrate pumping rate. This requirement is met by running both standby liquid control (SLC) pumps simultaneously. At the minimum allowable single pump pumping rate (41.2 gpm),

and with two pumps operating, a 13.6% sodium pentaborate decahydrate concentration will provide the required equivalent injection rate.

Maintenance of the SLC solution volume .and concentration within the limits established ensure that the single pump shutdown and ATWS requirements can be met.

3 4. 1.6 F EDWATER TEMPERATURE For the purpose of extending the cycle, feedwater temperature may be used for reactivity addition to compensate for the reactivity loss due to fuel depletion. The analysis performed is applicable to core flow values up to the maximum attainable (106 percent of rated core flow) and to feedwater temperature reductions to as low as 355'F. It is anticipated that a thermal coastdown from rated power with feedwater temperature held at 355'F would follow the rated run. This analysis also supports thermal coastdown followed by feedwater temperature reduction if this order is desirable. The analysis covers a reduction in power by thermal coastdown to 47 percent of rated thermal power with feedwater t'emperature held at or above 355'F.

WASHINGTON NUCLEAR-UNIT 2 B 3/4 1-5 Revision No. 0

REACTIVITY CONTRO SYSTEMS BASES FE DWATER TEMPERATURE (Continued)

It should be noted that during a normal feedwater lineup, a feedwater temperature at 355'F -entering the-reactor-vessel is achieved at approximately 47 percent of rated thermal power. The Limiting Condition for Operations clearly does not apply during reactor startups and shutdowns when reactor power is below the point at which a feedwater temperature of 355'F is attainable with a normal feedwater system lineup.

Prior to reaching the end-of-cycle exposure, operation with an abnormal feedwater heater lineup is permissible as the short-term effect of increased subcooling.is tomore-.strongly-bottom- peak"the-axial 'power shape allowing a scram to suppress the flux faster. Compensation for the long-term effect of a pronounced bottom burn can be made by rod pattern adjustments and axial flux shape monitoring.

WASHINGTON NUCLEAR-UNIT 2 B 3/4 1-6 Revision No. 0

3 4.2 POWER DISTRIBUTION IMITS BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200 F limit specified in 10 CFR 50.46.

3 4.2. AV RAGE P ANAR LIN AR HEAT GENERATION RATE The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. For GE fuel, the peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal-to-or-less-than"the"design-LHGR.corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady-state gap conductance and rod-to-rod local peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for GE fuel is this LHGR of the highest powered rod divided by its local peaking factor which results in a calculated LOCA PCT much less than 2200 F. The Technical Specification APLHGR for ANF fuel is specified to assure the PCT following a postulated LOCA will not exceed the 2200'F limit.

The limiting value for APLHGR is specified in the Core Operating Limits Report.

The calculational procedure used to establish the APLHGR specified in the Core Operating Limits Report is based on a loss-of-coolant accident analysis. The analysis was performed using calculational models which are consistent with the requirements of Appendix K to 10 CFR Part 50. These models are referenced in Specification 6.9.3.

3 4.2.2 APRM SETPO NTS The flow biased simulated thermal power-upscale scram setting and control rod block functions of the APRM instruments limit plant operations to the region covered by the transient and accident analysis. In addition, the APRM setpoints must be adjusted for both two recirculation loop operation and single recirculation loop operation to ensure that the MCPR does not become less than the fuel cladding, safety limit or that > I/ plastic strain does not occur in the degraded situation. The scram settings and rod block settings are adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and MFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.

WASHINGTON NUCLEAR UNIT 2 B 3/4 2-1 Revision No. 0

POWER D STRIBVTION LIMITS BASES 3 4.2.3 MINIMUM CRITICA POWER RATIO The required operating limit HCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition .of the reactor being at the steady-state operating limit, it is required that the resulting HCPR does not decrease below the Safety Limit HCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.'o assure that,.the, fuel, cladding-integrity"Safety"Limit "is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta HCPR. When added to the Safety Limit HCPR, the required minimum operating limit HCPR of Specification 3.2.3- is spe ified in the Core Operating Limits Report.

The evaluation of a given transient begins with the system initial parameters shown in the cycle specific transient analysis report that are input to an ANF core dynamic behavior transient compute; program. The outputs of this program along with the initial HCPR form the input for further analyses of the thermally limiting bundle. The codes and methodology to evaluate pressurization and nonpressurization events are referenced in Specification 6.9.3. The principal result of this evaluation is the reduction in HCPR caused by the transient.

The purpose of the flow dependent HCPRf specified in the Core Operating Limits Report is to define operating limits at other than rated core flow conditions.

At less than 100%%u. of rated flow the required 'HCPR, is the maximum of the rated flow MCPR and the reduced flow MCPR both specified in the Core Operating Limits Report. HCPQ assures that the Safety Limit HCPR will not be violated.

HCPR, is only calculated for the manual flow control mode. Automatic flow control operation is not permitted.

Lead Fuel Assemblies (LFAs) from Advanced Nuclear Fuels (ANF), General Electric (GE) and ABB Atom (ABB) reside in the reactor core. Analyses performed by the three vendors indicate that the transient CPR changes for the LFAs are greater than the CPR change calculated for the dominant ANF 8x8 fuel, due primarily to the shorter thermal time constants of the smaller diameter rods. All vendors state that their LFAs have inherently higher thermal margins (by design) than the dominant 8>8 'uel. Each vendor chose to address the CPR limit in a slightly different fan;.ion. These methods are discussed as follows.

WASHINGTON NUCLEAR UNIT 2 B 3/4 2-2 Revision No. 0

POWER DISTRIBUTION IMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

GE concludes that the inherent high thermal margin of the LFAs is sufficient to compensate for the larger CPR change associated with the shorter time constant and that the ANF Bx8 limits can be conservatively applied to the GEll LFAs.

The XN-3 CHF correlation used by ANF in the analysis was developed for the ANF Bx8 fuel. A review of the correlation and comparison to CHF data obtained for the 9x9 LFAs concluded XN-3 is conservative when applied to the ANF LFAs and the LFAs should be conservatively assumed to have a CPR performance at least equal to that calculated for an 8x8 assembly for the samepower and inlet conditions'. 'n'ddition, due to the water canister in the interior of the bundle, ANF modified the S-factors for the 9x9"LFAs to improve the XN-3 predictive capability. The'se S-factors were used in the analysis and were provided for use in monitoring the LFAs.

ABB Atom chose to take a more conservative approach and performed analyses which established conservative and unique MCPR values for the SVEA-96. The resulting MCPRs are included in the Technical Specifications.

In addition to the conservatisms discussed, the Supply System has committed to load the LFAs in core locations which have been analyzed to have sufficient margins such that the LFAs are not expected to be the limiting assemblies in the core on either a nodal or a bundle power basis. This approach is to prevent the possibility of the LFAs from ever being the limiting fuel bundle and adds additional margin to the event of a plant transient.

At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial start-up testing of the plant, a MCPR evaluation will be made at 25K of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary.

The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25K of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

At EOC during FFTR, the LOAD REJECTION WITHOUT BYPASS transient is slightly more severe when compared to the same transient without FFTR, which is accounted for by an increased MCPR operating limit. The analysis WASHINGTON NUCLEAR UNIT 2 B 3/4 2-3 Revision No; 0

POM R 0 STR BVTION IHITS BASES HINIHVM CRITICA POWER RATIO (Continued) conservatively reduces the feedwater temperature by 65'F and burns the produced power shape to achieve the final core conditions used in the transient analysis. This depletion causes the power peak to shift upwards, slightly increasing the time required for the normal scram to suppress the flux.

3 4..4 N A HEAT GENERATION RAT This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than,.the. design-linear"heat"generation-ev'eri'"i'f fuel pellet densification is postulated..

3 4.2.6 POWER FLOW INSTABILITY At the high power/low flow corner of the operating domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating conditions (e.g., power shape, bundle power, and bundle flow).

In 1984, GE issued SIL 380 addressing boiling instability and made several recommendations. In this SIL, the power/flow map was divided into several regions of varying concern. It also discussed the objectives and philosophy of "detect and suppress." The SIL recommends that REGION A be bounded by the 100% rod line and REGION C be bounded by the 80% rod line.

The NRC General Letter 86-02 discussed both the GE and SIEMENS (then EXXON) stability methodology and stated that due to uncertainties, General Design Criteria 10 and 12 could not be met using available analytical procedures on a BWR. The letter discussed SIL 380 and stated that General Design Criteria 10 and 12 could be met by imposing SIL 380 recommendations in operating regions of potential instabilities. The NRC concluded that regions of potential instability constituted decay ratios of 0.8 and greater by the GE methodology and 0.75 by the SIEMENS methodology which existed at that time.

SIEMENS Power Corporation has recently developed an improved stability computer code STAIF. A topical report (EHF-CC-074P) which describes the STAIF stability code and provides benchmarking against reactor data was submitted to the NRC in l993. The NRC issued a SER approving the STAIF stability code for establishing stability boundaries on April 14, 1994. In the SER on STAIF the NRC stated the uncertainty in the STAIF code was 20%.

The STAIF stability code has been used to establish the stability region boundaries for MNP-2. The lower boundary of REGION A was defined'o assure it bounds a decay ratio of 0.9. REGION C was conservatively defined to bound a decay ratio of 0.75.

MASHINGTON NUCLEAR UNIT 2 B 3/4 2-4 Revision No. 0

POWER 0 ST IBUTION HITS BASES POWER FLOW INSTABILITY (Continued)

Stability REGION A is shown in Figure 3.2.6-1. REGION A conforms to the recommendations of SIL 380 in that REGION A bounds a calculated decay ratio of 0.9. Operation in REGION A is prohibited.

3 4.2.7 STAB LITY MONITORING - TWO LOOP OPERATION At the high power/low flow corner of the operating domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating conditions (e.g., rod patterns, power shape). To -provide assurance that neutron flux, limit, cycleoscillations.aredetected and suppressed, APRH and LPRH neutron flux signal decay ratios should be monitored while operating this region (Region C).

'n Stability tests at operating BWRs were reviewed to determine a generic region of the power/flow map in which surveillance of neutron flux noise levels should be performed. A conservative decay ratio of 0.75 was chosen as the basis for determining the generic region for surveillance to account for the plant to plant variability of decay ratio with core and fuel designs. This generic region has been determined to correspond to a core flow of less than or equal to 45K of rated core flow and a thermal power at which the calculated decay ratio is less than 0.75.

Stability monitoring is performed utilizing the ANNA system. The system shall be used to monitor APRH and LPRH signal decay ratio and peak-to-peak noise values when operating in the region of concern. A minimum number of LPRH and APRH signals are required to be monitored in order to assure that both global (in-phase) and regional (out-of-phase) oscillations are detectable. Decay ratios are calculated from 30 seconds worth of data at a sample rate of 10 samples/second. This sample interval results in some inaccuracy in the decay ratio calculation, but provides rapid update in decay ratio data. A decay ratio of 0.75 is selected as a decay ratio limit for operator response such that sufficient margin to an instability occurrence is maintained. When operating in the region of applicability, decay ratio and peak-to-peak information shall be continuously calculated and displayed. A survei11ance requirement to continuously monitor decay ratio and peak-to-peak noise values ensures rapid response such that changes in core conditions do not result in approaching a point of instability.

3 4.2.8 STABI ITY MONITORING - SINGLE LOOP OPERATION The basis for stability monitoring during single loop operation is consistent with that given above for two loop operation. The smaller size of the region of allowable operation, Region C, is due to a limit on the allowed flow above the 8(5 rodline. When operating above the 801 rodline in single loop operation, the core flow is required to be greater than 39K. Continuous operation in Region B is not permitted. Should Region B be entered the actions required by Technical Specification 3/4.4.l.l are to be complied with.

WASHINGTON NUCLEAR UNIT 2 B 3/4 2-5 Revision No. 0

3 4.3 INSTRUMENTATION BASES 3 4.3; R ACTOR PROTECTION SYSTEM INSTRUM NTATION The reactor protection system automatically initiates a reactor scram to:

a. Preserve the integrity of the fuel cladding.

b. Preserve the integrity of the reactor coolant system.

c. Minimize the energy which must be adsorbed following a'oss-of-coolant accident, and d.. -

Prevent inadvertent"critical-ity:- -- - -" "-"-.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessa'ry, one channel may be made inoperable for brief-intervals to conduct required surveillance.

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip .system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC 30851 P, "Technical Specification Improvement Analyses for BWR Reactor Protection System," as approved by the NRC and documented in the SER (letter to T. A. Pickens from A. Thadani dated July 15, 1987). The bases for the trip settings of the RPS are discussed in the bases for Specification 2.2.1.

The RPS instrumentation that provides I) the Turbine Throttle Valve-Closure and 2) Turbine Governor Valve Fast Closure,:Valve Trip System Oil Pressure-Low trip signals measures first stage turbine pressure to initiate a trip signal. The Load Rejection safety analysis (FSAR 15.2.2) bases initial conditions on rated'ower and specifies turbine bypass operability at greater than or equal to 3(C or rated thermal power. Because first stage pressure can vary depending on operating conditions, the qualifying notes describing when the turbine bypass feature is to be disabled specify a turbine first stage pressure corresponding to less than 30% RTP (turbine first stage pressure is dependent on the operating parameters of the reactor, turbine, and condenser).

Therefore, because a value fot turbine first stage pressure cannot be precisely fixed and because pressure measurement initiates the trip, the Technical Specification refers to a pressure associated with a specific Rated Thermal Power value rather than a value for pressure.

WASHINGTON NUCLEAR UNIT 2 B 3/4 3-1 Revision No. 0

INSTRUMENTATION BASES R CTOR P OT TION SYST M INSTRUMENTATION (Continued)

The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the safety analyses. No credit was taken for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times. The response time limits are, contained-in-FSAR"Chapter-7-. ""' "

3 4.3. 2 SOLATION CTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints for isolation of the reactor systems. Nen necessary, one channel may be inoperable for brief intervals to conduct required surveillance. Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.

Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connected. For D.C.-operated valves, a 3-second delay is assumed before the valve starts to move. For A.C.-operated valves, it is assumed that the A.C. power supply is lost and is restored by startup of'he emergency diesel generators. In this event, a time of 13 seconds is assumed before the valve starts to move. In addition to the pipe break, the failure of the D.C.-operated valve is assumed; thus the signal delay (sensor response) is concurrent with the 13-second diesel startup. The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 13-second delay. It follows that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isolation functions. However, to enhance overall system reliability and to monitor instrument channel response time trends, the isolation actuation instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME. The response time limits are contained in FSAR Chapter 7.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

NSHINGTON NUCLEAR-UNIT 2 B 3/4 3-2 Revision No. 0

NSTRUMENT T ON BASES 4.3.3 MERGENCY CORE COO NG SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERABILITY requirements and trip setpoints that will ensure effectiveness of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time. The response time limits are contained in FSAR Chapter 7.

Operation with a trip-set--less-conservative"than"its"Trip-Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety anal>ses.

3 4.3.4 CIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NED0-10349, .dated March 1971, and NEDO-24222, dated December 1979.

The end-of-cycle recirculation pump trip (EOC-RPT) system is a part of the.

reactor protection system and is an essential safety supplement to the reactor trip. The purpose of the EOC-RPT is to recover the loss of thermal margin which occurs at the end-of-cycle. The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivity. Each EOC-RPT system trips both recirculation pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most .limiting pressurization events.

The two events for which the EOC-RPT protective feature will function are closure of the turbine throttle valves and fast closure of the turbine governor valves.

A fast closure sensor from each of two turbine governor valves provides input to the EOC-RPT system; a fast closure sensor from each of the other two turbine governor valves provides input to the second EOC-RPT system.

Similarly, a position switch for each of two turbine throttle valves provides input to one EOC-RPT system; a position switch from each of the other two throttle valves provides input to the other EOC-RPT system. For each EOC-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine governor valves and a 2-out-of-2 logic for the turbine throttle valves. The operation of either logic will actuate the EOC-RPT system and trip both recirculation pumps.

MASHINGTON NUCLEAR UNIT 2 B 3/4 3-3 Revision No. 0

INSTRUMENTATION BASES R CIRCU TION PUMP RIP ACTUAT ON INSTRUMENTATION (Continued)

Each EOC-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled. The manual bypasses and the automatic Operating Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control room. The EOC-RPT System instrumentation that provides a trip signal measures first stage turbine pressure to initiate a trip signal. The safety analysis requiring an EQC-RPT bases initial conditions on rated power and specifies turbine bypass operability at greater than or equal to 30% of rated thermal power. Because first stage pressure can vary depending on operating conditions, the qualifying notes describing when the turbine bypass feature is to disabled specify a turbine-first"stage p'ressure'"corr'esponding to less than 30% RTP (turbine first stage pressure is dependint on the operating parameters of the reactor, turbine, and condenser). Therefore, because a value for turbine first stage pressure cannot be precisely fixed and because pressure measurement initiates the trip the Technical Specification refers to a pressure associated with a specific Rated Thermal Power value rather than a value for pressure.

The EOC-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e.,

190ms, less the, time allotted for sensor response, i.e., 10ms, and less the time allotted for breaker arc suppression determined by test, as correlated to manufacturer's test results, i.e., 83ms, and plant preoperational test results.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

3 4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without providing actuation of any of the emergency core cooling equipment.

Oper ation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

WASHINGTON NUCLEAR-UNIT 2 B 3/4 3-4 Revision No. 0

INSTRUM NTATION BASES 3 4.3.6 CONTROL ROD BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the requirements of Specifications 3/4. 1.4, Control Rod Program Controls, 3/4.2, Power Distribution Limits and 3/4.3. 1 Reactor Protection System Instrumentation.

The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the, drift. allowance, assumed..for each trip .in,.the safety analyses..

The test exception to the weekly Channel Functional Test of the SRM/IRM, Detector Not Full In instrumentation noted in Table 4.3.6-1, Control Rod Block Instrumentation Requirements, is, intended to avoid cable damage and radiation exposure during operational condition 5 periods when outage work is being done in the under core region. Upon completion of all the work in this area, when access for maintenance or construction efforts is no longer required, the test will be completed per the prescribed frequency within seven days.

3 4.3.7 MONITORING INSTRUMENTATION 3 4.3.7. 1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the individual channels; (2) the alarm is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.

This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63, and 64.

The criticality monitor alarm setpoints were calculated using the criteria from 10 CFR 70.24.a.l that requires detecting a dose rate of 20 Rads per minute of combined neutron and gamma radiation at 2 meters. The alarm setpoint was determined by calculational methods using the gamma to gamma plus neutron ratios from ANSI/ANS 8.3-1979, Criticality Accident Alarm System, Appendix B and assuming a critical mass was formed from a seismic event, with a volume of 6' 6' 6't a distance of 27.7 feet from the two detectors.

The calculated dose rate using the methodology is 5.05 R/hr. The allowable value for the alarm setpoint was, therefore, established at 5R/hr.

3 4.3.7.2 Deleted WASHINGTON NUCLEAR UNIT 2 B 3/4 3-5 Revision No. 0

INSTRUMENTATION BASES 3 4.3.7.3 METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the meteorological monitoring instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public. This instrumentation is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs,"

February 1972.

3 4.3.7. 4 REMOTE -SHUTDOWN'MONITORING.-INSTRUMENTATION - ..

The OPERABILITY of the remote shutdown monitoring instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTDOWN of the unit from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of Appendix A to 10 CFR Part 50.

3 4.3. 7. 5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor .

and assess important variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980. Regulatory Guide 1.97 commitment compliance is summarized in FSAR Table 7.5-1, "Safety-Related Display Instrumentation."

3 4.3.7.6 SOURCE RANGE MONITORS The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown. At these power levels, reactivity additions shall not be made without this flux level information available to the operator. When the intermediate range monitors are on scale, adequate information is available without the SRMs and they can be retracted.

3 4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM The OPERABILITY of the traversing in-core probe system with the specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core.

3/4.3.7.8 NOT USED 3/4.3.7.9 NOT USED WASHINGTON NUCLEAR-UNIT 2 B 3/4 3-6 Revision No. 1

INSTRUM NTATION BASES 3 4. 3.7. 10 LOOSE- PART DE ECTION SYSTEM The OPERABILITY of the loose-part detection system ensures that sufficient capability is available to detect loose metallic parts in the primary system and avoid or mitigate damage to primary system components. The allowable out-of-service times and surveillance requirements are consistent with the recommendations of Regulatory Guide 1.133, "Loose-Part Detection Program for the Primary System of Light-Mater-Cooled Reactors," May 1981.

3/4.3.7.11 NOT USED 3 4.3.7. XP OS V GAS MONITORING INSTRUMENTA ION This instrumentation provides for monitoring the concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM to ensure that the concentration of potentially explosive gas mixtures contained in the offgas holdup system is maintained below the flammability limits of hydrogen.

Maintaining the concentration of hydrogen below its flammability limit in accordance with Specification 3/4 11.2.6 provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3 4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM This specification is provided to ensure that the turbine overspeed protection system instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety-related components, equipment or structures.

3 4.3.9 FEEDWATER SYSTEM MAIN TURBINE TRIP SYSTEH ACTUATION INSTRUMENTATION The feedwater system/main turbine trip system actuation 'instrumentation is provided to initiate the feedwater system/main turbine trip system in the event of reactor vessel water level equal to or greater than the level 8 setpoint associated with a feedwater controller failure.

WASHINGTON NUCLEAR UNIT 2 B 3/4 3-7 Revision No. 0

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510 AECNc 101.0 O15OH4AOE 1TZ.5 NOZZLE NOZZLE 100 BASES FIGURE 8 3/4.3<> Reactor operati on and resul tant fast neutron i rradi at i on, E greater than I MeV, will cause an increase in the RT>>. Therefore, an adjusted reference temperature, based upon the fluence, nickel content, and copper content of the material in question, can be predicted using the fluence for 109.2% of original rated power and the recommendations of Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials." The pressure/temperature limit curves, Figures 3.4.6.1A, 3.4.6.1B, and 3.4.6.1C include predicted adjustments for this shift in RTo~ for the end of-life fluence and are effective for 10 EFPY and 8 EFPY, respectively.

The actual shift in RT>> of the vessel material will be established periodically during operation by removing arid evaluating, in accordance with ASTM E185-73 and'0 CFR Part 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area. The irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limit curves of Figures 3.4.6.1A, 3.4.6.1B, and 3.4.6.1C shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99, Revision 2.

The pressure-temperature limit lines shown in Figures 3.4.6.1A, 3.4.6.1B, and 3.4.6.1C for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

3 4.4.7 MAIN STEAM LIN SO ATION VALV S Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case'of a line break. Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating MASHINGTON NUCLEAR-UNIT 2 B 3/4 4-5 Revision No. 0

R ACTOR COOLANT SYST M BASES M IN ST M ISO A ION VALVES (Continued) 0 history of this type valve. The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3 4 4.8 STRUCTURAL NTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable.leve]-throughout-the"1'ife o'f"the plant.

Access to permit inservice inspections of components of the reactor coolant system is in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1974 Edition and Addenda through Summer 1975.

The inservice inspection program for ASME Code Class 1, 2 and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a.

3 4.4.9 RESIDUAL H AT REMOVA A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assuage accurate temperature indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an al'ternate method of coolant mixing be in operation.

WASHINGTON NUCLEAR-UNIT 2 B 3/4 4-6 Revision No. 0

3 4.5 H RG NCY COR COOLING SYST BASES 3 4.5.1 nd 3 4.5.2 ECCS - OPERATING and SHUTDOWN ECCS division I consists of the low pressure core spray system and low pressure coolant injection subsystem "A" of the RHR system and the automatic depressurization system (ADS) as actuated by ADS trip system "A". ECCS division 2 consists of low pressure coolant injection subsystems "B"'nd "C" of the RHR system and the automatic depressurization system as actuated by ADS trip system "B".

The low pressure core spray (LPCS) system is provided to assure that the core is adequately cooled following a loss-of-coolant accident and provides adequate core cool.ing capaci.ty...for all break. sizes..up,to..and .including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the ADS.

The LPCS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.

The surveillance requirements provide adequate assur ance that the LPCS system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. ,The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

The low pressure'oolant injection (LPCI) mode of the RHR system is provided to assure that the core is adequately cooled following a loss-of-coolant accident. Three subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor line break, and for small breaks following depressurization by 'ecirculation the ADS.

The surveillance requirements provide adequate assurance that the LPCI system will be OPERABLE when required. Although all active components are testable and, full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

ECCS division 3 consists of the high pressure core spray system. The high pressure core spray (HPCS) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCS system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HPCS system operates over a range of 1210 psid, differential pressure between reactor vessel and HPCS suction source, to 0 psid.

WASHINGTON NUCLEAR UNIT 2 B3/4 5-1 Revision No. 0

EM RGENCY CORE COOL NG SYST M BASES ECCS - OP RATING and SHUTDOWN (continued)

The capacity of the system is selected to provide the required core cooling.

The HPCS pump is designed to del-iver greater than or equal to 516/1550/6350 gpm at differential pressures of 1160/1130/200 psig. Initially, water from the condensate storage tank is used instead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water.

With the HPCS system inoperable, adequate core cooling is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the. LPCS .and- LPCI-systems . "In"addition,the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the safety analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition. The HPCS out-of-service period of 14 days is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems.

The surveillance requirements provide adequate assurance that the HPCS system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.

Upon failure of the HPCS system to function properly after a small break loss-of-coolant accident, the automatic depressurization system (ADS) automatically causes selected safety/relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200'F. ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 100 psig. This pressure is substantially below that for which the low pressure core cooling systems can provide adequate core cooling for events requiring ADS.

ADS automatically controls seven selected safety/relief valves although the safety analysis only takes credit for five valves. It is therefore appropriate to permit two valves to be out-of-service for up to 14 days without materially reducing system reliability.

3 4.5.3 SUPPR SSION CHAMB R The suppression chamber is required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of water is available to the HPCS; LPCS, and LPCI systems in the event of a LOCA. This limit on suppression chamber minimum water volume ensures that sufficient water is available to permit recirculation cooling flow to the core. The OPERABILITY of the suppression chamber in OPERATIONAL CONDITION 1, 2, or 3 is required by Specification 3.6.2.1.

WASHINGTON NUCLEAR UNIT 2 B 3/4 5-2 Revision No. 0

HERG NCY C R COOLING SYSTEM BASES SUPPR SSION CHAMB R (Continued)

Repair work might require making the suppression chamber inoperable. This specification will permit those repairs to be made and at the same time give assurance that the irradiated fuel has an adequate cooling water supply when the suppression chamber must be made inoperable, including draining, in

'PERATIONAL CONDITION 4 or 5.

MASHINGTON NUCLEAR UNIT 2 B 3/4 5-3 Revision No. 0

3 4.6 CONTAINM NT SYSTEMS BASES 3 4.6.1 PRIMARY CONTAINMENT 3 4.6.1.1 PRIMARY CONTA NM NT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

3 4.6.1. PRIMARY CONTAINMENT LEAKAGE The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the calculated peak accident pressure. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L. during performance of the, periodic tests'to account for possible degradation of the containment leakage barriers between leakage tests.

Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore the special requirement for testing these valves.

i The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR Part 50 with the exception of exemptions granted for main steam isolation valve leak testing and testing the air locks after each opening.

3 4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the primary containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the primary containment leakage rate given in Specifications 3.6.1.1 and 3.6. 1.2.

The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor operation. Only one closed door in each air lock is required to maintain the integrity of the containment.

3 4.6.1.4 MSIV EAKAGE CONTROL SYSTEM Calculated doses resulting from the maximum leakage allowance for the main steamline isolation valves in the postulated LOCA situations would be a small fraction of the 10 CFR Part 100 guidelines, provided the main steam line system from the isolation valves up to and including the turbine condenser remains intact. Operating experience has indicated that degradation has occasionally occurred in the leak tightness of the MSIVs such that the WASHINGTON NUCLEAR UNIT 2 B 3/4 6-1 Revision No. 0

CONTAINH NT SYSTEHS BASES HS V AKAGE CONTRO SYSTEH (Continued) specified leakage requirements have no always been maintained continuously.

The requirement for the leakage control system will reduce the untreated leakage from the HSIVs when isolation of the primary system and containment is required.

Design specifications require the system to accommodate a leak rate of five times the Technical Specification leakage allowed for the HSIVs while maintaining a negative pressure downstream of the HSIVs. The allowed leakage value per each valve is 11.5 scfm, or a total of 230 scfm (3.8 scfm)'". Mhen corrected, for..worst,.case"pressure-,--temperature-and"hbmidity expected to be seen during surveillance testing conditions, the flow would never exceed an indicated value (uncorrected reading from local flow instrumentation) of 5 cfm. The 30 cfm acceptance criterion provides significant margin to this design basis requirement and provides a benchmark for evaluating long term blower performance. The Technical Specification limit for pressure of -17" H~O M.C. was also established based on a benchmark of the installed system performance capability. This -17" H,O M.C. provides assurance that the negative pressure criterion can be met.

3 .6.1.5 PRIHARY CONTAINMENT STRUCTURA INT GRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the unit. Structural integrity is required to ensure that the containment will withstand the maximum calculated pressure in the event of a LOCA. A visual inspection in conjunction with Type A leakage tests is to demonstrate this capability. 'ufficient 3 4.6. l.6 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE The limitations on drywell and suppression chamber internal pressure ensure that the calculated containment peak pressure does not exceed the design pressure of 45 psig during LOCA conditions or that the external pressure differential does not exceed the design maximum external pressure differential of 2 psid. The limit of 1.75 psig for initial positive containment pressure will limit the peack pressure to be less than the design pressure and is consistent with the safety analysis.

(a) Letter, G02-75-238, dated August 18, 1975, NO Strand (SS) to OD Parr (NRC), "Response to Request for Information Hain Steam Isolation Valve Leakage Control System" WASHINGTON NUCLEAR-UNIT 2 B 3/4 6-2 Revision No. 0

CONTAINMENT SYSTEMS BASES 3 4.6.1.7 DRYW L AV RAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 340'F during LOCA conditions and is consistent with the safety analysis.

3 4.6. 1.8 DRYW L AND SUPPRESSION CHAMBER PURGE SYSTEM The 24-inch and 30-inch drywell and suppression chamber purge supply and exhaust isolation valves are required to be closed during plant operation except as required for inerting, de-inerting and pressure control. Until all the drywell andsuppression chamber .val.ves. have,.been qual.i.fied. as, capable of closing within the times assumed in the safety analysis, they shall not be .

open more than 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months in any consecutive 365 days. Valves not capable of closing from a full open position during a LOCA or steam line br eak accident shall be blocked so as not to open more than 70'.

The time limit on use of the drywell and suppression chamber purge lines is not restricted when using the 2-inch purge supply and exhaust isolation valves since the 2-inch valves will close during a LOCA or steam line break accident and therefore the SITE BOUNDARY dose guidelines of 10 CFR Part 100 would not be exceeded in the event of an accident during PURGING operations. The design of the 2-inch purge supply and exhaust isolation valves meets the requir erne'nts of Branch Technical Position CSB 6-4, "Containment Purging During Normal Plant Operations."

Leakage integrity tests with a maximum allowable leakage rate for purge supply and exhaust isolation valves will provide early indication of resilient material seal degradation and will allow the opportunity for repair before gross leakage failure develops. Valves with metal to metal seals will be tested on a Type C schedule in accordance with Surveillance 4.6. 1.2.d to assure allowable'leakage rates are not exceeded. The 0.60 L. leakage limit shall not be exceeded when the leakage rates determined by the leakage integrity tests of those valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.

3 4.6.2 DEPRESSURIZATION SYST MS The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 45 psig during primary system blowdown from full operating pressure.

The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system. The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1040 psig. Since all of the gases in the drywell are purged into the suppression chamber air space during a loss-of-coolant accident, the pressure WASHINGTON NUCLEAR UNIT 2 B 3/4 6-3 Revision No. 0

CONTAINMENT SYSTEMS BASES DEPRESSU I ATION SYSTEMS (Continued) of the liquid must not exceed 45 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant and to be considered is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

Using the minimum or maximum water volumes given in this specification, containment pressure during the design basis accident is below the design pressure of 45 psig. Maximum water volume of 128,827 ft'esults in a downcomer..submergence-of-l2-ft-and "the minimum 'volume of 127, 197 ft'esults in a submergence approximately 4 inches less. The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation. Thus, with respect to the downcomer submergence, this specification is adequate. The maximum temperature at the end of the blowdown tested during the Humboldt Bay,and Bodega Bay tests was 170 F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above 170'F.

Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3.

Under full power operating conditions, blowdown from an initial suppression chamber water temperature of 90'F results in a water temperature of approximately l45'F immediately following blowdown which is below the 200'F used for complete condensation via quencher devices. At this temperature and atmospheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus, there is no dependency on containment overpressure during the accident injection phase. If both RHR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations.

data indicate that excessive steam condensing loads can be if

'xperimental avoided the peak bulk temperature of the suppression pool is maintained below 200 F during any period of relief valve operation with sonic conditions at the discharge exit for quencher devices. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the, regime of potentially high suppression chamber'oadings.

Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual examination following'ny event where potentially high loadings could occur provides assurance that no significant damage was encountered.

NSHINGTON NUCLEAR-UNIT 2 B 3/4 6-4 Revision No. 0

CONTAINM NT SYSTEMS

'ASES D PRESSURIZATION SYSTEMS (Continued)

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety/relief valve inadvertently opens or sticks open. 'As a minimum this action shall include: (I) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate 'reactor shutdown, and (4) if other safety/relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety/relief valve to assure mixing and uniformity of energy insertion to the pool.

3 4.6.3 PR MARY CONTAINMENT. ISO ATION..VALVES..., ..

The OPERABILITY of the primary containment isolation valves ensures that the atmosphere wil,l be isolated from the outside environment in the 'ontainment event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures for those isolation valves designed to close automatically that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

t 3 4.6.4 VACUUM RELI F Vacuum relief breakers are provided to equalize the pressure between the suppression chamber and drywell and between the reactor building and suppression chamber. This system will maintain the structural integrity of the primary containment under conditions of large differential pressures.

The vacuum breakers between the suppression chamber and the drywell, must not be inoperable in the open position since this would allow bypassing of the suppression pool in case of an accident. There are nine pairs of valves to provide redundancy and capacity so that operation may continue indefinitely with no more than two pairs of vacuum breakers inoperable in the closed position.

3 4.6.5 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The reactor building and associated structures provide secondary containment during normal operation when the drywell is sealed and in service. At other times the drywell may be open and, when required, secondary containment integrity is specified.

Establishing and maintaining a vacuum in the reactor building with the standby gas treatment system once per 18 months, along with the surveillance of the doors, hatches, dampers, and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.

MASHINGTON NUCLEAR-UNIT 2 B 3/4 6-5 Revision No. 0

CONTAI M NT SYSTEMS BASES SECOND RY CONTAINM NT (Continued)

The OPERABILITY of the standby gas tr eatment systems ensures that sufficient iodine removal capability will be available in the event of a LOCA. The reduction in containment iodine inventory reduces the resul,ting SITE BOUNDARY radiation'oses associated with containment leakage. The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses. Continuous operation of the system with the heaters OPERABLE for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months during each 31 day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.

3 4.6.6 PRIMARY CONTAINM NT..ATMOSPHERE-CONTRO--

The OPERABILITY of the systems required for the detection and control of oxygen/hydrogen gas ensures that these systems will be available to maintain the oxygen concentration within the primary containment below the lower oxygen limit for oxygen/hydrogen mixture during post-LOCA conditions. .Either drywell and suppression chamber oxygen/hydrogen recombiner system is capable of controlling the expected oxygen generation associated with radiolytic decomposition of water. The oxygen/hydrogen control system is consistent with the recommendations of Regulatory Guide 1.7, "Control of Combustible Gas Concentrations in Containment Following a LOCA," September 1976.

Following an accident the inerted primary containment oxygen level is controlled to not exceed 4.8'A volume with the catalytic recombiner system. By FSAR Figure 6.2-26 the containment will reach 4.8% oxygen approximately 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months after the accident if either recombiner system is operating.

To provide assurance that recombiners are capable of achieving the required oxygen removal, the feed and effluent streams will be sampled during surveillance testing to establish that the effluent hydrogen concentration is less than 25 ppm by volume for a feed of at least 1% hydrogen by volume. This will confirm a minimum efficiency of 99.7N for the expected range of post-accident conditions. This efficiency will be adequate to maintain the post-accident oxygen level below 4.8% by volume.

The CAC system employs a platinum on alumina catalyst to recombine the oxygen and hydrogen flow from the containment. During accident conditions, the gas mixture is preheated to approximately 450 to 550'F prior to entering the catalyst. This preheat increases the effectiveness of the hydrogen/oxygen recombination because it limits the potential for bed poisoning. In the test configuration, the blower is used as the only source of gas stream heating and the catalyst preheaters are not energized. The blowers are capable of heating the gas stream by compression. Temperatures at the blower exit are limited for test purposes to approximately 300'F due to the blower gas exit temperature trip setpoint.

NSHINGTON NUCLEAR UNIT 2 B 3/4 6-6 Revision No. 0

CONTAINM NT SYSTEMS BASES

~PR"::-'.Y CONTAINM NT ATMOSPHERE CONTRO (Continued)

The capacity of the catalyst bed can be reduced through mechanical, thermal, or chemical (poisoning) deactivation. Poison can be introduced through the environment or the precess. To protect the catalyst, the CAC skid is maintained isolated with a pressurized nitrogen blanket. During the performance of the required surveillance testing, the catalyst is exposed to air (and the potential for poisoning from the environment). However, the process of testing establishes that the bed is operable and has not been damaged. The process stream, in a design bases event, may include iodine.

Iodine can chemically poison the platinum catalyst; however, the CAC skid scrubs the process gas to remove,.iodine .from...the..process stream.and heats it to reduce effects on the catalyst.

It is important to note that a catalyst bed and its ability to recombine hydrogen and oxygen does not deplete simply from use. Any reduction in recombination capability is caused by poisoning or other damage to or loss of catalyst, or by insufficient activation energy (low inlet temperature). Given adequate inlet temperature, the presence of poisoning in the top few inches of the bed will simply move the peak reaction further down in the bed with very little'effect on the percent completion of the reaction. Any such downward movement in the site of the majority of the recombination should be evaluated to determine any actions that may be necessary. Measuring the hydrogen concentration in the effluent stream provides the necessary information that in fact the catalyst is able to recombine hydrogen and oxygen at greater than the 99% efficiency assumed in the containment analysis. Verification that the maximum temperature rise occurs near the top of the bed (i.e., as seen on the first three RTDs) assures that no damage to the bed is preventing proper operation. If the maximum temperature rise occurs near the bottom of the bed (i.e., on the lowest RTD), verification that at least 75% of the increase was achieved above that RTD indicates that the lower portion of the bed is still capable of providing the necessary catalytic function. However, the change of location of that recombination process provides indication of the potential degradation of the catalyst.

Degradation of the catalyst bed will also be indicated by the decreased ability to recombine hydrogen and oxygen. This indication can be determined through the evaluation of the hydrogen content of the influent and effluent.

The catalyst bed should maintain a relatively constant capacity for recombination. If the comparison of the influent and effluent hydrogen concentrations begins to indicate a degradation of the catalyst bed, replacement of the bed wi11 be evaluated.

MASHINGTON NUCLEAR UNIT 2 B 3/4 6-7 Revision No. 0

rp 1

4.7 PLANT S STEMS BASES p

3 4.7.1 SERVICE WATER SYST MS The OPERABILITY of the service water systems ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of these systems, assuming a single failure, is consistent with the assumptions used in. the accident conditions within acceptable limits.

During periods of low ambient temperatures, when the possibility of freezing exists if the sprays were to be operated, the discharge of each spray cooling division is typically aligned directly into the pond (spray bypass mode).

Safety. analysis:has..shown"that-several. hours. are available..for..realignment to spray following the design basis LOCA accident in conjunction with extreme meteorological conditions. A 72'F alarm requiring action for realignment provides 2 1/2 hours before 77'F would be exceeded, based on accident heat loads. With the pond temperature below 77'F and the spray headers in service the safety analysis provided in FSAR Section 9.2.5 is bounding and the system therefore remains operable in the spray or bypass mode of operation.

3 4.7.2 CONTROL ROOM MERG NCY FI TRATION SYSTEM The OPERABILITY of the control room emergency filtration system ensures that (1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating'or the equipment and instrumentation cooled by this system and (2) the control room will remain habitable for operations personnel during and following all design basis accident conditions. Continuous operatio'n of the. system with the heaters OPERABLE for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months during each 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. Th'e OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rems or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criterion 19 of Appendix A, 10 CFR Part 50.

3 4.7.3 REACTOR CORE ISOLATION COO ING SYSTEM The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater. flow to the reactor vessel without requiring actuation of any of the emergency core cooling system (ECCS) equipment. The RCIC system is conservatively required to be OPERABLE whenever reactor pressure exceeds '150 psig. This pressure is substantially below that for-which the low pressure core cooling systems can provide adequate core cooling for events requiring the RCIC system.

The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, 2, and 3 when reactor vessel, pressure exceeds 150 psig because RCIC is the primary non-ECCS source of emergency core cooling when the reactor is pressurized.

WASHINGTON NUCLEAR UNIT 2 B 3/4 7-1 Revision No. 0

P ANT SYS HS BASES 3 4.7.3 R ACTOR COR ISOLATION COO ING SYSTEM (Continued)

With the RCIC system inoperable, adequate core cooling is assured by the OPERABILITY of the HPCS system and justifies the specified 14 day out-of-service period.

The surveillance requirements provide adequate assurance that RCIC will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage and to start cooling at the earliest possible moment-.-":"

3 4.7.4 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integri+.'f the Reactor Coolant System and all other safety-related systems is maintained during and following a, seismic or other event initiating dynamic loads.

Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system. During shutdown, snubbers which are redundant per engineering analysis can be removed for maintenance and/or testing and are excluded from the operability requirements.

Snubbers are classified and grouped by design and manufacturer but not by size. For example, mechanical snubbers utilizing the same design features of the 2-kip, 10-kip, and 100-kip capacity manufactured by Company "A" are of the same type. The same design mechanical snubbers manufactured by Company "B" for the purposes of this Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer.

A list of individual snubbers with detailed information of snubber location and size and of system affected shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50. The accessibility of each snubber shall be determined and approved by the Plant Operations Committee. The determination shall be based upon the existing radiation levels and the expected time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (e.g.,

temperature, atmosphere, location, etc.), and the recommendations of Regulatory Guides 8.8 and 8.l0. The addition or deletion of any hydraulic or mechanical snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50.

The visual inspection schedule is based on the number of unacceptable snubbers found during the previous inspection in proportion to the sizes of the various snubber populations or categories. A snubber is considered unacceptable if it fails the acceptance criteria of the visual inspection. Snubbers may be WASHINGTON NUCLEAR-UNIT 2 B 3/4 7-2 Revision No. 0

~NNT NT T NN BASES SNUBBERS (Continued) categorized, based upon their accessibility during power operation, as accessible or inaccessible. These categories may be examined separately or jointly. The decision to examine these categories separately or jointly shall be made and documented before the examination begins, and cannot be changed during the examination. The inspection interval is .based on a fuel cycle of up to 24 months and may be as long as two fuel cycles, or 48 months for other fuel cycles, depending on the number of unacceptable snubbers found during the previous visual inspection. The examination interval may vary by 25 percent to coincide with the actual outage.

i To provide assurance of"snubber'unc'tional reliability, one of two functional testing methods are used with the stated acceptance criteria:

I. Functionally test IOX of a type of snubber with an additional 5/. tested for each functional testing failur'e, or

2. Functionally test a sample size and determine sample acceptance or continue testing using Figure 4.7-1.
Figure 4.7-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in "guality Control and Industrial Statistics" by Acheson J. Duncan.
Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubbers for the applicable design conditions at either the completion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions.
The service life of a snubber is establ .;;ed via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life.
3 4.7.5 SEALE SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.
This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources are classified into three group- according to their use, with WASHINGTON NUCLEAR UNIT 2 B 3/4 7-3 Revision No 0
P ANT SYSTE S BASES SEALED SOURCE CONTAMINATION (Continued) surveillance requirements commensurate with the probability of damage to a source in that group. Those sources which're frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism, i.e., sealed sources within radiation monitoring devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
3 4.7.6 Deleted 3 4.7.7 Deleted 3 4.7.8 AR A T MP RATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause loss of its OPERABILITY.
3 4.7.9 MAIN TURBINE BY ASS SYSTEM The main turbine bypass system is required to be OPERABLE consistent with the assumptions of the feedwater controller failure analysis of the cycle specific analysis. The main turbine bypass system provides pressure relief during the feedwater controller failure event so that the safety limit MCPR is not violated.
WASHINGTON NUCLEAR-UNIT 2 B 3/4 7-4 Revision No. 0
3 4.8 E ECTRICAL POWER SYSTEMS BASES 3 4.8. 1 3 4.8.2 and 3 4.8.3 A.C. SOURCES D.~. SOURCES and ONSITE POWER DISTRIBUTION SYSTEMS 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 Part 50.
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 Division 1 or 2 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. or D.C. source. Division 3 supplies the high pressure core spray (HPCS) system only.
The A.C. and D.C. source allowable out-of-service times are based on Regulatory Guide 1.93, "Availability of Electrical Power Sources," December 1974. When diesel generator (1) or (2) is inoperable, there is an additional ACTION requirement to verify that all required systems, subsystems, trains, components and devices, that depend on the remaining OPERABLE diesel generator (1) or (2) as a source of emergency power, are also OPERABLE. This requirement is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period diesel generator (1) or (2) is inoperable. The term verify as used in this context means to administratively check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons. It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component.
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 unit status.
The surveillance requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guide 1.9, "Selection of Diesel Generator Set Capacity for Standby Power Supplies," March 10, 1971, Regulatory Guide 1. 108, "Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, August 1977, and Regulatory Guide 1. 137, "fuel-Oil Systems for Standby Diesel Generators," Revision 1, October'1979.
WASHINGTON NUCLEAR UNIT 2 B 3/4 8-1 Revision No. 0
LECTRICA POWER SYSTEMS BASES A.C. SOURCES O.C. SOURCES and ONSIT POW R DISTRIBUTION SYSTEMS (Continued)
The diesel generator fast start surveillance requirements, based on a PRA study, are sufficient to demonstrate the onsite A.C. power system capability to mitigate the consequences of the design basis event for the plant, i.e.,
large LOCA coincident with a loss-of-offsite power, while minimizing the mechanical stress and wear on the diesel engine.
The surveillance requirements for demonstrating the OPERABILITY of the unit batteries are in accordance with the recommendations of Regulatory Guide 1.129, "Maintenance Testing and Replacement of Large Lead Storage Batteries for NuclearPower. Plants;-" February-l978-,'and IEEE'td 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."
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.
The simulated emergency load profile used for the battery service test of Surveillance Requirement 4.8.2.l.d is verified to be at least equivalent to the actual emergency load profile and is based on anticipated operations required during loss-of-offsite power (LQQP) and loss-of-coolant accident (LOCA) conditions as described in the WNP-2 Final Safety Analysis Report (FSAR). The simulated emergency load profiles for the batteries are defined and located in the FSAR Section 8.3.
Table 4.8.2.1-1 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 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, 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 0.020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than 0.010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery.
Operation with a battery cell's parameter outside the normal limit but within, the allowable value specified in Table 4.8.2.1-1 is permitted for'up to 7 days. During this 7-day period: (I) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable v=lue for the average specific gravity of all the cells, not more than 0.020 be ow the manufacturer's recommended full charge specific gravity ensures tha-. the decrease in rating will be less WASHINGTON NUCLEAR-UNIT 2 B 3/4 8-2 Revision No. 0
c i ~j ELECTRICAL POWER SYSTEMS BASES A.C. SOURCES O.C. SOURC S and ONSITE POWER OISTRIBUTION SYST MS (Continued) than the safety margin 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 0.040 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.
3 4.8. 4 LECTR ICAL E UIPMENT PROTECTIVE OEVICES
~ a ~e Primary containment electrical penetrations and penetration conductors are protected by either deenergizing circuits not required during reactor operation or demonstrating the OPERABILITY of primary and backup overcurrent protection circuit breakers by periodic surveillance.
-The surveillance requirements applicable to 'lower voltage circuit breakers provide assurance of breaker reliability by testing at least one representative sample of each manufacturers brand of circuit breaker. Each manufacturer's molded case and metal case circuit breakers are grouped into representative samples which are than tested on a rotating basis to ensure that all breakers are tested. If a wide variety exists within any manufacturer's brand of circuit breakers, it is necessary to divide that manufacturer's breakers into groups and treat each group as a separate type of breaker for surveillance purposes.
The bypassing of the motor-operated valve thermal overload protection continuously or during accident conditions ensures that the thermal overload protection will not prevent safety-related valves from performing their function. The surveillance requirements for demonstrating the bypassing of the thermal overload protection continuously and during accident conditions are in accordance with Regulatory Guide 1.106 "Thermal Overload Protection for Electric Motors on Motor Operated Valves," Revision I, March 1977.
WASHINGTON NUCLEAR UNIT 2 B 3/4 8-3 Revision No. 0
'C ~if~
3 4.9 R FUELING OPERATIONS BASES 3 4.9.1 REACTOR MODE SWITCH Locking the OPERABLE reactor mode switch in the Shutdown or Refuel position, as specified, ensures that the restrictions on control rod withdrawal and refueling platform movement during the refueling operations are properly activated. These conditions reinforce the refueling procedures and reduce the probability of inadvertent criticality, damage to reactor internals or fuel assemblies, and exposure of personnel to excessive radioactivity.
3 4.9.2 INSTRUMENTATION The OPERABILITYof at l,east...two..source range monitors..ensure. that, redundant monitoring capability is'vailable to detect changes in the reactivity condition of the core. SRMs are not required to be OPERABLE when less than or equal to 4 bundles are inserted around the SRM and no other fuel assemblies are in the associated core quadrant since this configuration critical even with all control rods withdrawn. Additionally,will this not be configuration (four bundles inserted around each SRM) provides significantly more SHUTDOWN MARGIN than is required by LCO 3.1.1 (SHUTDOWN MARGIN).
3 4.9.3 CONTROL ROD POSITION The requirement that all control rods be inserted during other CORE ALTERATIONS ensures that fuel will not be loaded into a cell without a control rod.
OE AY TIN The minimum requirement for reactor subcriticality prior to fuel movement 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. 5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during movement of fuel within the reactor pressure vessel.
3 4.9.6 R FU I G P ATFORM The OPERABILITY requirements ensure that (I) the refueling platform will be used for handling control rods and fuel assemblies within the reactor pressure vessel, (2) each crane and hoist has sufficient load capacity for handling fuel assemblies and control rods, and (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
WASHINGTON NUCLEAR UNIT 2 B 3/4 9-1 Revision No. 0
a R FUELING OP RATIONS BASES 3 4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL The restriction on movement of loads in excess of the nominal weight of a fuel assembly over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses.
3 4.9.8 and 3 .9.9 MATER EVEL - REACTOR VESSE and WATER L Y - SP NT FU
~IN E C I The restrictions on minimum water level ensure that sufficient water depth is available to remove 991 of the assumed 105 iodine gap activity released from the rupture of an irradiated fuel assembly. This minimum water depth is consistent with the assumptions of the safety analysis.
3 4.9. Q CONTROL OD REMOVAL These specifications ensure that maintenance or repair of control rods or control rod drives will be performed under conditions that limit the probability of inadvertent criticality. The requirements for simultaneous removal of more than one control rod are more stringent since the SHUTDOWN MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn.
3 4.9.11 RESIDUAL HEAT REMOVA AND COOLANT CIRCULATION The requirement that at least one residual heat removal loop be OPERABLE or that an alternate method capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pr essure vessel below 140 F as required during REFUELING, and (2) sufficient coolant circulation would be available through the reactor core to assure accurate temperature indication and to distribute and prevent stratification of the poison in the event it becomes necessary to actuate the standby liquid control system.
The requirement to have two shutdown cooling mode loops OPERABLE when there is less than 22 feet of water above the reactor vessel flange ensures that a single failure of the operating loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and 22 feet of water above the reactor vessel flange, a large heat sink is available for core cooling. Thus, in the event a failure of the operating RHR loop, adequate time is provided to initiate alternate methods capable of decay heat removal or emergency procedures to cool the core.
WASHINGTON NUCLEAR-UNIT 2 B 3/4 9-2 Revision No. 0
) +r 3 4.10 SPEC A T ST C PTIONS BASES 3 4.10.1 PRIMARY CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable during the
.period when open vessel tests are being performed during the low power PHYSICS TESTS.
3 . 0. 0 S U C CONTRO SYST In order to perform the tests required in the technical specifications necessary to bypass the sequence restraints on control rod movement. The it is additional surveillance requirements ensure that the specifications on heat generation rates and.shutdown.margin. requirements,.are..not,.exceeded-during the period when these tests are being performed and that individual rod worths do not exceed the values assumed in the safety analysis.
3 4. 0.3 SHUTOOWN MARGIN OEMONSTRATIONS Performance of shutdown margin demonstrations with the vessel head removed requires additional restrictions in order to ensure that criticality does not occur. These additional restrictions are specified in this LCO.
3 4.10.4 RECIRCULATION LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.
3 4.10.5 OXYGEN CONC NTRATION Relief from the oxygen concentration specifications is necessary in order to provide access to the primary containment during the initial startup and testing phase of operation. Without this access the startup and test program could be restricted and delayed.
3 4.10.6 TRAINING STARTUPS This special test exception permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperature while controlling RCS temperature with one RHR subsystem aligned in the shutdown cooling mode in order to minimize contaminated water discharge to the radioactive waste disposal system.
WASHINGTON NUCLEAR UNIT 2 B 3/4 10-1 Revision No. 0
SPEC L T ST XCEPTIONS BASES 3 4.10 7 NSERVIC LEAK AND HYDROSTATIC TESTING OPERATION This special test exception allows reactor vessel inservice leak and hydrostatic testing to be performed in OPERATIONAL CONDITION 4 with the maximum reactor coolant temperature not exceeding 212'F. The additionally imposed OPERATIONAL CONDITION 3 requirement for secondary containment operability provides conservatism in the response of the unit to an operational event. This allows flexibility since temperatures of the reactor vessel metal will be Z 180'F during the testing and a higher reactor coolant temperature will be necessary to sustain the .vessel metal temperature. The flexibility is provided so that there is margin to allow temperature drift due to decay and,mechanical"heat:-
WASHINGTON NUCLEAR UNIT 2 B 3/4 10-2 Revision No. 0
a ~e 3 4.11 RADIOACTIVE EFFLUENTS BASES 3 4.11.1 LI UID EFFLUENTS 3/4.11.1.1 Relocated to ODCM 3/4.11.1.2 Relocated to ODCH 3/4.11.1.3 Relocated to ODCM 3 4.11, .4 LI OlD HO DUP ANKS The tanks listed in this specification include all those outdoor radwaste tanks that are.not surrounded-by-liners;- dikes,- or'walls'apable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.
Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks'ontents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.
3 4. 11.2 GASEOUS FFLUENTS 3/4.11.2.1 Relocated to ODCM 3/4.11.2.2 Relocated to ODCM 3/4.11.2.3 Relocated to ODCM 3/4.11.2.4 Relocated to ODCM 3/4. 11.2.5 Relocated to ODCM 3 4.11.2.6 EXPLOSIVE GAS MIXTUR This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the offgas system is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.
MASHINGTON NUCLEAR-UNIT 2 B 3/4 11-1 Revision No. 0
RAOIOACTIV EF LU NTS BASES 3 4. 1. . HAIN CONOENS R Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.
WASHINGTON NUCLEAR UNIT 2 B 3/4 11-2 Revision No. 0
3 4. 12 RAD IOLOG I CA I ENV RONMENTA MONITORING BASES 3 4.1 .1 Relocated to ODCM 3 4. 2.2 Relocated to ODC 3 4. 2.3 Relocated to ODC MASHINGTON NUCLEAR UNIT 2 B 3/4 12-1 Revision No. 0
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ML17291B249

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