Text

Amend 42 to License NPF-30,reducing Required RHR Sys Flowrate in Mode 6,deleting RHR ACI Function & Allowing Safety Injection Pumps to Be Energized W/Head on & W/Water Level Not Above Top of Reactor Vessel Flange in Modes 5 & 6
	ML20247E504
Person / Time
Site:	Callaway
Issue date:	03/27/1989
From:	Hannon J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20247E507	List: ML20247E504;
References
	NUDOCS 8904030123
	Download: ML20247E504 (18)
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UtJITED STATES g

NUCLEAR REGULATORY COMMISSION lin

j' WASHINGTON, D. C. 20665

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UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO. STN 50-483 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 42 License No. NPF-30 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment filed by Union Electric Company (UE, the licensee) dated January 9,1989 as sup)lemented by a letter dated February 10, 1989, complies with tie standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

Thereisreasonableassurance(1)thattheactivitiesauthorizedby this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment.will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and para-graph 2.C.(2) of Facility Operating License No. NPF-30 is hereby amended to read as follows:

8904030123 890327 PDR ADOCK 050004S3 P

PDC

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, (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 42, and the Environmental Protection Plan l

contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license. UE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

l 3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION bMl4 n N.Hannon, Director l

Project Directorate III-3 Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 27, 1989

r, J

ATTACHMENT TO LICENSE AMENDMENT NO. 42 OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Corresponding overleaf pages are provided to maintain document completeness.

REMOVE INSERT 3/4 4-35 3/4 4-35 3/4 5-4 3/4 5-4 3/4 5-9 3/4 5-9 3/4 9-9 3/4 9-9 3/4 9-10 3/4 9-10 B 3/4 4-15 B 3/4 4-15 B 3/4 5-2 B 3/4 5-2 B 3/4 5-3 B 3/4 9-2 B 3/4 9-2

d REACTOR COOLANT SYS~EM SURVEILLANCE REQL'IREMENTS l

4.4.9.3.1 Each PORV thall be demonstrated OPERABLE by:

l a.

Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel', but excluding valve operation, within 31 days prior.to entering a condition in which the PORY is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE; I

b.

Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and j

c.

Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORV is being used for overpressure protection.

)

1 4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection ac follows:

a.

For RHR suction relief valve 8708B:

By verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that RHR RCS suction isolation valves (RRSIV) EJ-HV-8701B and BB-PV-8702B are open.

b.

For RHR suction relief valve 8708A:

By verifying at least once per 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s-that RRSIV EJ-HV-8701A and BB-PV-8702A are open.

)

c.

Testing pursuant to Specification 4.0.5.

4.4.9.3.3 TheRCSvent(s)shallbeverifiedtobeopenatleastonceper 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

when the vent (s) is being used for overpressure protection.

Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

CALLAWAY - UNIT 1 3/4 4-35 Amendment No. 42

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5'O 100 200 300 400 MEASURED RTD TEMPERATURE, F CALLAWAY - UNIT 1 3/4 4-36 Amendment No. 36

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i MI R(it.NCY CORI C00l LNG SYSIEMS 3/4.b.2 ECCS SUBSYSTEMS - T,yg > 250*F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be i

OPL9ABLE with each subsystem comprised of:

a.

One OPERABLE centrifugal charging pump, Une OPERABLE Safety Injection pump, u.

c.

One OPERABLE RHR heat exchanger,

~

d.

One OPERABLE RHR pump, and An OPERABLE ilow path capable of taking suction from the refueling e.

water storage tank on a Safety Injection signal and automatically transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: MODES 1, 2, and 3.*

ACTION:

With one ECCS subsystem inoperable, restore the inoperable subsystem a.

to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUT 00WN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated dCluation Cycles to date.

The Current Value of the usage factor f or each af fected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

l h[.

[vi olon, of Specif icatinns 3.0.4 and 4.0.4 are not applicable f or entry into MODI.I for the centrifugal charging pump and the Safety injection pumps

<lertareil inoperable pursuant. to Specification 4.5.3.2 provided the centrifugal c.harging pump and the Safety injection pumps are restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months priur to the temperature of one or more of the RCS cold legs exceeding 3/5'F.

I CAIt.AWAY UNil 1 3/4 5-3

a-EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the following valves are in the indicated positions with power to the valve operators removed:

Valve Number Valve Function Valve Position BN-HV-8813 Safety Injection to Open RWST Isolation Viv EM-HV-8802A(B)

SI Pump Discharge Closed Hot Leg Iso Vivs EM-HV-8835 Safety Injection Open Cold Leg Iso yalve EJ-HV-8840 RHR/SI Hot Leg Closed Recirc Isc Yalve EJ-HV-8809A RHR to Accum Inj Open Loops 1 & 2 Iso Viv EJ-HV-8809B RHR to Accum Inj Open Loops 3 & 4 Iso Viv b.

At leasi. once per 31 days by:

1)

Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping high points, and 2)

Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

c.

By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions.

This visual inspection shall be performed:

.)

For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and 2)

Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.

d.

At least once per 18 months by:

1)

Verifying automatic isolation action of the RHR System from the Reactor Coolant System by ensuring that, with a simulated or actual Reactor Coolant System pressure signal greater than or equal to 425 psig, the interlocks prevent the valves from oeing opened.

CALLAWAY-UNIT 1 3/4 5-4 Amendment No. 42

_ - _ = _ _ _ _ _ _ _ _ _ _ _ - _____

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EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 ECCS SUBSYSTEMS - T

< 200*F ava LIMITING CONDITION FOR OPERATION 3.5.4 All Safety Injection pumps shall be inoperable.

APPLICABILITY: MODE 5 with the water level above the top of the reactor vessel flange, and MODE 6 with the reactor vessel head on and with the water level above the top of the reactor vessel flange.

ACTION:

With a Safety Injection pump OPERABLE, restore all Safety Injection pumps to an inoperable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.4 All Safety Injection pumps shall be demonstrated inoperable

by verifying that the motor circuit breakers are secured in the open position at least once per 31 days.

l l

l

An inoperable pump may be energized for testing or for filling accumulators provided the discharge at the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.

CALLAWAY-UNIT 1 3/4 5-9 Amendment No. 42

e' EMERGENCY CORE (00 LING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:

a.

A minimum contained borated water volume of 394,000 gallons, b.

A boron concentration of between 2000 and 2100 ppm of boron, c.

A minimum solution temperature of 37*F, and d.

A maximum solution temperature of 100*F.

APPLICABILITY: MODES 1, 2, 3, and 4.

AC110.4:

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the f ollowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the conta;..2d borated water volume in the tank, and 2)

Verifying the boron concentration of the water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside air temperature is either less than 37*F or greater than 100*F.

.s CALLAWAY - UNIT 1 3/4 5-10

)

i d

REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION

\\

3.9.8.1 At least one residual heat remova'l (RHR) loop shall be OPERABLE and in operation.*

APPLICABILITY: MODE 6 when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet.

ACTION:

With no RHR loop OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentra-tion of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible.

Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.9.8.1 At least once per twelve hours one RHR loop shall be verified in operation and circulating coolant at a flow-rate of:

a) greater than or equal to 1000 gpm, and b) sufficient to maintain the RCS temperature at less than or equal to 140 F.

The RHR loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.

CALLAWAY - UNIT 1 3/4 9-9 Amendment No. 42

r REFUELING OPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in operation.

I APPLICABILITY:

MODE 6 when the water level above the top of the reactor vessel flange is less than 23 feet.

ACTION:

a.

With less than the required RHR loops OPERABLE, immediately initiate corrective tction to return the required RHR loops to OPERABLE status or to establish greater than or equal to 23 feet of water e

above the reactor vessel flange, as soon as possible.

b.

With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 1

4.9.8.2 At leac. once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months one RHR loop shall be verified in operation and circulating coolant at a flow-rate of a) greater than or equal to 1000 gpm, and b) sufficient to maintain the RCS temperature at less than or equal to 140 F.

CALLAWAY - UNIT 1 3/4 9-10 Amendment No. 42

_a

s e

l REACTOR COOLANT SYSTEM BASES HEATUP(Continued)

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the i

course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs, or two RHR suction relief valves, or an RCS vent opening of at least 2 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 368 F.

Either PORV or either RHR suction relief valve has adequate relieving capabil-ity to protect the RCS from overpressurization when the transient is limited to either:

(1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50 F above the RCS cold leg temperatures, Dr (2) the start of a centrifugal charging pump and its injection into a water-solid RCS.

I In addition to opening RCS vents to meet the requirement of Specifica-tion 3.4.9.3c., it is acceptable to remove a pressurizer Code safety valve, open a PORV block valve and remove power from the valve operator in conjunction with disassembly of a PORV and removal of its internals, or otherwise open the RCS.

COLD OVERPRESSURE The Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation System (COMS) is derived by analysis which models the performance of the COMS assuming various mass input and heat input transients.

Operation with a PORV setpoint less than or equal to the maximum setpoint ensures that Appendix G criteria will not be violated with consideration for 1) a maximum pressure overshoot beyond the PORV setpoint which can occur as a result of time delays in signal processing and valve opening; 2) a 50 F heat transport effect made CALLAWAY - UNIT 1 B 3/4 4-15 Amendment No.

42

O i '.

e REACTOR COOLANT SYSTEM BASES COLD OVERPRES$URE (Continued) possible by the geometrical relationship of the RHR suction line and the RCS wide range temperature indicator used for COMS; 3) instrument uncertainties; and 4) single failure.

To ensure mass and heat input transients more severe than those assumed cannot occur, technical specifications require lockout of both safety injection pumps and all but one centrifugal charging pump while in MODES 4, 5 and 6 with the' reactor vessel head installed and disallow start of an RCP if secondary temperature is more than 50*F above primary temperature.

i Exceptions to these mode requirements are acceptable as described below.

l l

Operation above 350 F but less than 375 F with only one centifugal charging i

pump OPERABLE and no safety injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

I As shown by analysis LOCA's occurring at low temperature, low pressure conditions can be successfully mitigated by the operation of a single centrifugal charging pump and a single RHR pump with no credit for accumulator injection. Given the shurt time duration that the condition of having only one centrifugal charging pump OPERABLE is allowed and the probability of a LOCA occurring during this time, the failure of the single centrifugal charging pump is not assumed.

Operation below 350 F but greater than 325*F with all centrifugal charging and safety injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . During low prenure, low temperature operation all automatic safety injection actuation signals except Containment Pressure - High are blocked.

In normal conditions a

t. ingle f ailure of the ESF actuation circuitry will result in the starting of at most one train of safety injection (one centrifugal charging pump, and one safety injection pump).

For temperatures above 325 F, an overpressure event occurring as a result of starting two pumps can be successfully mitigated by operation of both PORV's without exceeding Appendix G limit.

Given the short time duration that this condition is allowed and the low probability of a single failure causing an overpressure event during this time, the single failure of a PORV is not assumed.

Initiation of both trains of safety injection during this 4-hour time frame due to operator error or a single failure occurring during testing of a redundant channel dre not Considered to be Credible accidents.

Although COMS is required to be OPERABLE when RCS temperature is less than 368"F, operation with all centrifugal charging pumps and both safety injection pumps OPERABL E is acceptable when RCS temperature is greater than 350 F.

Should an inadvertent safety injection occur above 350 F, a single PORV has sufficient eapacity to relieve the combined flow rate of all pumps.

Above 350 F, two RCP and all pressurizer safety valves are required to be OPERABLE.

Operation of an RCP climinates the possibility of a 50 F dif ference existing between indicated and attnai RCS temperature as a result of heat transport ef fects.

Considering i ni, t rumen t uncertainties only, an indicated RCS temperature of 350 F is suffi-tiently high to allow full RCS pressurization in accordance with Appendix G limitations.

Should an overpressure event occur in these conditions, the pres-surizer safety valves provide acceptable and redundant overpressure protection.

The Maximum Allowed PORV setpoint for the Cold Overpressure Mitigation System will be undated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H and in accordance with the schedule in Table 4.4-L CALLAWAY - UNIT 1 8 3/4 4-16

i

'3/4'.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4. 5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below s

the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.

The accumulator power operated isolation valves are considered to be

" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met.

In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required.

The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures.

If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a MODE where this capability is not required.

In order to perform check valve surveillance testing per 4.0.5 or 4.4.6.2.2 above 1000 psig RCS pressure, one accumulator isolation valve may be closed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in mode 3 only.

The requirement to verify accumulator isolation valves shut with power removed from the valve operator when the pressurizer is solid ensures the accumulators will not inject water and cause a pressure transient when the Reactor Coolant System is on solid plant pressure control.

3/4.5.2, 3/4.5.3, and 3/4.5.4 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that st.fficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.

Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the a

double ended break of the largest RCS cold leg pipe downward.

In addition, each ECCS subsystem provides long-tenn core cooling capability in the recircula-tion mode during the accident recovery period.

With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

CALLAWAY - UNIT 1 B 3/4 5-1 Amendment No. 40

EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued)

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE charging pump to be inoperable in MODES 4 and 5 and in MODE 6 with the reactor vessel head on, provides assurance that a mass addi-tion pressure transient can be relieved by the operation of a single PORV or RHR suction relief valve.

In addition, the requirement to verify all Safety Injection pumps to be inoperable in MODE 4, in MODE 5 with the water level above the top of the reactor vessel flange, and in MODE 6 with the reactor vessel head on and with the wate level above the top of the reactor vessel flange, provides assurance that tne mass addition can be relieved by a single PORV or RHR suction relief valve.

With the water level not above the top of the reactor vessel flange and with the vessel head on, Safety Injection pumps may be available to mitigate the effects of a loss of decay heat removal during partially drained conditions.

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

Surveillance Requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injec-tion point is necessary to:

'(1) prevent total pump flow from exceedim runout conditions when the system is in its minimum resistance configu-ation, (2) provide the proper flow split between injection points in accordar.ce with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in 1

the ECCS-LOCA analyses.

The Surveillance Requirements for leakage testing of ECCS check valves ensure that a failure of one valve will not cause an inter-i system LOCA.

The Surveillance Requirement to vent the ECCS pump casings and accessible, i.e., can be reached without personnel hazard or high radiatic..

dose, discharge piping ensures against inoperable pumps caused by gas binding or water hammer in ECCS piping.

3/4.5.5 REFUELING WATER STORAGE RANK The OPERABILITY c/ the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injec-tion by the ECCS in the event of a LOCA.

The limits on RWST minimum volume and boron concentration ensure that:

(1) sufficient water is available within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are censistent with the LOCA analyses.

I CALLAWAY - UNIT 1 B 3/4 5-2 Amendment No. 42

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EMERGENCY CORE C0OLING SYSTEMS BASES REFUELING WATER STORAGE TANK (Continued) l The contained water volume limit includes an allowance for water not I

usable because of tank discharge line location or other physical character 1stics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion en mechanical systems and components.

t l

l CALLAWAY - UNIT 1 B 3/4 5-3 Amendment No. 42

v 3/4.9 REFUELINGOPERATIOg BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain suberitical during CORE ALTERATIONS, and (2) a uniform boroa concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel.

The limitation on K nogreaterthan0.95issufficienttopreventreactorcriticalitydurin$y,'of refueling ope ations.

The locking closed of the required valves during refueling oper&tions precludes the possibility of uncontrolled boron dilution of the filled partions of the Reactor Coolant System.

This action prevents flow to the RCS of unberated water by closing flow paths from sources of unborat(d water.

These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses.

3/4.9.2 INSTRUMENTATION

'the OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring ca condition of the core. pability is available to detect changes in the reactivity 3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient tim has elapsed to allow the radioactive decay of the short-lived fission products.

Thif, deca analyses.y time is consistent with the assumptic,s used in the safety 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containm will be restricted from leakage to the environment.

The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupturs based upon the lack of containment pressurization potential while in the REFUELING MODE.

The OPERABILITY of this system ensures the containment purge penetrations will be automatically isolated upon detection of high radiation levels within containment.

The OPERABILITY of this system is required to restrict the release of radioactive materials from the containment atmosphere to the environment.

The restriction on the setpoint for GT-RE-22 and GT-RE-33 is baseo or a fuel handling accident inside the Containment Building with resulting damage to pne fuel red and subsequent release of 0.1% ef the noble gas gap activity, except for 0.3% of the Kr-85 gap activity.

The setpoint concentration of SE-3 uCi/cc is equivalent to approximately 150 mR/hr submersion dose rate.

3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling statien personnel can be promptly informed of significant changes in the facility statue or core reactivity conditions during CORE ALTERATIONS.

CALLAWAY - UNIT 1 B 3/4 9-1 Amendment No.

20 APR 101987

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'i REFUELING OPERATIONS _

BASES 3/4.9.6 REFUELING MACHINE

~

The OPERABILITY requirements for the refueling machine and auxiliary hoist ensure that:

(1) manipulator cranes will be used for movement of drive rods and fuel assemblies, (2) each crane has sufficient load capacity to lift a drive rod or fuel assembly, and (3) the core internals and reactor vessel are protected from excessive lif ting force in the event they are inadvertently

).

engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool areas ensures that in the event this load is dropped: (1) the activity release will be limited to that contained in a single fuel assspbly, 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 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that:

(1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140 F as required during the REFUELING MODE, and (2). sufficient coolant circulation is maintained through the core to minimize the effect of a boron dilution incident and 5

prevent boron stratification.

The requirement to maintain a 1000 gpm flowrate ensures that there is adequate flow to prevent boron stratification. The RHR flow to the RCS will provide adequate cooling to prevent exceeding 140 F and to allow flowrates which provide additional margin against vortexing at the RHR pump suction while in partial drain operation.

e The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure a

of the operating RHR loop will not resu in a complete loss of residual heat removal e.apability.

With the reactor ve. el head removed and at least 23 feet of water above the reactor vessel flange, a large heat sink is available for core cooling.

Thus, in the event of a failure of the operating RHR loop, T'

adequate time is provided to initiate emergency procedures to cool the core.

3/4.9.9 CONTAINMENT VENTILATION SYSTEM The OPERABILITY of this system ensures that the containment purge penetra-tions will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

CALLAWAY - UNIT 1 B 3/4 9-2 Amendment No. 42

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