Text

Amend 57 to License DPR-51,providing Redefinition of Term Operable & Adding General Limiting Conditions of Operation
	ML19351A047
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/16/1981
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Arkansas Power & Light Co
Shared Package
ML19351A048	List: ML19351A047; ML19351A049; ML19351A050;
References
	DPR-51-A-057 NUDOCS 8106250116
	Download: ML19351A047 (35)
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b UNITED STATES g

NUCLEAR REGULATORY COMMisslON g

j wAshlNGTON, D. C. 20685 e

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ARKANSAS POWER & LIGHT COMPANY DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE - UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Anendnent No.57 License No. DPR-51 1.

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

A..The application for amendment by Arkansas Power and Light Company.

(the licensee) dated liovember 28, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Connissi,on; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be I

conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical' to the comon I

defense and security or to the health and safety of the public; l

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The issuance of this amendment is in accordance with 10 CPR Part E.

51 of the Consnission's regulations.and all applicable regirements have been satisfied.

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8106250 l

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2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.c.(2) of Facility Operating License i

No. DPR-51 is hereby amended to read as follows:

(2) Technical Spe'cifications The Technical Specifications contained in I.ppendices A.and B, as revised through Amendment No.57 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the' Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY C3MMISSION oh h;esto1F, Chief-Opyrating Reactors Branch M ivision of L'icensing

Attachment:

Changes to the Technical Specifications Date of Issuance: June 16,1981 l

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. ATTACHMENT TO LICENSE AMENDMENT NO. 57 -

FACILITY OPERATING LICENSE NO. DPR-51 DOCKET NO. 50-313 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.. Tha revised pages ~are identified by amendment number and contain vertical lines indicating the area of change.

Pages 2

52*

15a (new page) 53d 15b (new page) 54 15c (new page) 55 16 59 18 59a 18a (new page) 59b 19 h

66 21 66b 27 66g 28*

66m i

30 66n 32 66o 46 66p 49 66q 51 66r

No change on these pages, included for document completeness only.

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refueling t(1perature (normally 140F).

Pressure is defined by Specifi-cat. ion 3.1.2.

A refueling shutdown refers to a shutdown to replace or rearrange all or a portion.of the fuel assemblies and/or control rods.-

1.2.7 Refueling Operation An 6peration involving a change in core geometry.by manipulation of fuel or control. rods when the reactor vessel head is removed.

1.2.8 Startup The reactor shall be considered in the startup mode when the shutdown margin is reduced with the intent of going critical.

1. 3 OPERABLE - OPERABILITY A system, subsystem, train', component or device shall be OPERABLE nr have OPERABILITY when it is capable of performing its specified func-tion (s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of per-g forming their related support function (s).

1.4 PROTECTION INSTRUMENTATION LOGIC 1.4.1 Instrument Channel' An instrument channel is the combination of sensor, wires, amplifiers.

and output devices which are connected for the purpose of measuring the value of a process variable for the purpose of observation, control and/

or protection.

An instrument channel:may be either analog or digital.

1.4.2 Reactor Protection System The reactor protection system is shown in Figures 7-1 and 7-9 of the FSAR.

It is that combination of protect je channels and associated circuitry which forms the automatic system that protects the reactor by control rod trip.

It includes the four protection channels, their associated instru-ment channel inputs, manual trip switch, all rod drive control protective trip breakers and activating relays or coils.

A protection channel, as shown in Figure 7-1 of the FSAR (one of. three or one of four independent channels, complete with sensors, sensor power supply 2

Amendment No. 25,57

3.

LIMITING CONDITIONS FOR OPERATION 3.0 LIMITING CONDITION FOR OPERATION (GENERAL)-

3.0.1 The Limiting Conditions for Operation requirements shall be

' applicable during-the Reactor Operating Conditions or other condi-tions specified for each specification.

3.0.2 Adherence to,the requirements of the Limiting Condition for 4

Operation within the specified time interval shall constitute com-pliance with the specification.

In the event the Limiting. Condition i

for. Operation is restored prior to expiration of the specified time interval, no further actions need be taken.

3.0.3 Whe'n'a Limiting Condition for.0peration is not met, except as provided in the associated Action' requirements, within one hour.

action shall be initiated to place the unit in an OPERATING CONDITION le in which the Specification does not apply by placing ic, as applicable, in:-

1 1.

At least Hot STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , i

2.. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and 3.

At least COLD SHUTDOWN within the subs'equent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4 Where corrective measures arc completed that permit operation under the Action requirements, the Action may.be taken in accordance with the specified time limits as measured from the time of failure to i

meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual Specifications.

3.0.4 Entry into an Operating Condition or other specified applicability L

~ condition shall not be made unless the conditions of the Limiting Con-dition for Operation are set without reliance on provisions contained in the Action statements unless.otherwise excepted..This provision shall l.

not prevent psssage through' Operating Conditions as-requbed by a Limit-ing Condition for Operation.

l 3.0.5 When a system, subsystem, train, component or device is determined s

to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inopereble, it may be considered OPERABLE for the purpose of sctisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s),

train (s), component (s) and device (s) are OPERABLE; or likewise satisfy the re Unless both conditions (1)quirements of this specification.and (2) are satisfied, within 2 j

be initiated to place the unit in an OPERATING CONDITION in which the applicable Limiting Condition for Operation do's not apply e

by placing it, as Applicable, in:

1.. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , l

2.

At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and 3.

At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

f This Specification is not applicable in Cold Shutdown or Refueling Shutdown.

15a CTordiirgt No. 57 a---.-.-....-

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BASES 3.0.1 This specification defines the applicability of each specifica-tion in terms of defined Operating Conditions or other specified con-ditions and is provided to delineate specifically when each specification

'is applicable.

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3.0.2 This specification defines.those conditions necessary to consti-l tute. compliance with the terms of an individual Limiting Condition for Operation.

3.0.3 This specification delineates the Action to be taken for circum-stances not directly provided for in the Action statements and whose occ'urrence would violate the intent of the specification.

3.0.4 This specification provides that entry into an Operating Condition ur other specified applicability condition must be made with (a) the full

- complement of. required systems, equipment or components OPERABLE and (b) all other parameters as specified in the Limiting Conditions for Operation being met.

The intent of this provision is to insure that facility operation is not initiated with either required equipment or systems inoperable or other

-specified limits being exceeded.

Exceptions to this provision have' heen provided for a limited number of specificatio's when startup with inoperable equipment would not affect plant safety.

These exceptions are stated in the Limiting Condition for Operation statements of the appropriate specifications.

3.0.5 This specification delineates what additional conditions must be

. satisfied to permit operation to continue when a normal or emergency power source is not OPERABLE.

It specifica1ly prohibits operation when one division is inoperable because its normal or emergency power source is inoperable and a system, subsystem, train, component or device in another I

division is inoperable for another reason.

l The provisions of this specification permit the Limiting Condition for Operation statements associated with individual systems, subsystems, trains, compo a,?s or devices to be consistent with the Limiting Condi-l-

tion for Operation statements of the associated electrical' power source.

It allows operation to be governed by the time limits of the Limiting t-Condition for Operation for the normal or emergency power source, not the individual Limiting Condition for Operation statements for each sys-tem, subsystem, train, component or device that is determined to be in-operable solely because of the inoperability of its normal or emergency power source.

For example, Specification 3.7.2.C provides for a 2 day out-of-service

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time when one emergency diesel generator is not OPERABLE.

If the deft-nition of OPERABLE were applied without consideration of Specification 3.0.5, all systems, subsystems, trains,. components and devices supplied by the inoperable emergency power source would also be inoperable.

This would dictate invoking the applicable Action statements for each of the applicable Limiting Conditions for Operation. However, the provisions 1

of S,ecification 3.0.5 permit the time limits for continued operation to Amendment No. 57 15b

be consistent with the [Liniting Condition for Operat'on state ent for

- the inoperable emergency diesel generator instead, provided the 'other specified conditions are satisfied.

In this case, this would mean that the corresponding ~ normal power source must be OPERABLE, and all redundant systems, subsystems, trains, components and devices must be OPERABLE, or

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otherwise satisfy Specification 3.0.5 (i.e., be capable of performing their design function and have at least one normal or one emergency power source OPERABLE).

If they are not satisfied, shutdown is required in accordance with this specification.

As 'a further example, Specification 3.7.1.A requires in part that two-physically independent circuits between the offsite transmission network and the onsite Class IE distribution system be OPERABLE.

Specification 3.7.2.8 provides a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months out-of-service time when both required offsite circuits are not OPERA 8LE.

If the definition of OPERABLE were applied

- without consideration of Specification 3.0.5, all systems, subsystems, trains, components and devices supplied by the inoperable normal power sources, both of the offsite circuits would also be inoperable.

This '

would dictate invoking the applicable Limiting Condition for Operation statements for each of the applicable LCOs.

However, the provisions of Specification 3.0.5 permit the time limits for continued operation to be consistent with the Limiting Condition for Operation statement for the i

inoperable normal power sources instead, provided the other specified conditions are satisfied.

In this case, this weuld mean that for ona division the emergency power source must be OPERABLE (as must be the components supplied by the emergency power source) and all redundant systems, subsystems, trains, components and devices in the other divi-sion must be OPERABLE, or likewise satisfy Specification 3.0.5 (i.e.,

be capable of performing their design functions and have an emergency power source OPERA 8LE).

In other words, both emergency power sources must be OPERABLE and all redundant systems, subsystems, trains, compo-i-

nents and devices in both divisions must also be OPERABLE.

If these conditions are not satisfied, shutdown is required in accurdance with l-

- this specification.-

During Cold Shutdown and Refuelin Shutdown, Specification 3.0.5 is not

- applicable and thus the individua Action statements for each appilcable

- Limiting Condition for Operation in these MODES must be adhered to.

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l l-l Amendment No. 57 15c

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3.1 REACTOR COOLANT SYSTEM

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Applicability Applies to the operating status of the reactor coolant system.

Objective To specify those limiting conditions for operation of tb reactor coolant system which must be met to ensure safe reactor operations.

3.1.1 Operaticnal Components 1

Specification 3.1.1.1 Reactor Coolant Pumps A.

Pump combinations permissible for given power levels shall be as shown in Table 2.3-1.

B.

The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or 1

one decay heat removal pemp is circulating reactor coclant.

With no reactor coolant pumps or decay heat removal pumps running, immediately suspend all operations involving a reduction of boron concentration in the reactor coolant system.

c, 3.1.1.2 Steam Generator A.

Two steam generators shall be operable whenever the reactor i

coolant average temperature is above 280F.

3.1.1. 3 Pressurizer Safety Valves A.

Both pressurizer code safety valves shall be operable when the reactor is critical.

With one pressurizer code safety j

valve operable, either restore the valve to operable status within 15 minutes or be in Hot Shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

. hen the reactor is suberitical, at least one pressurizer

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B.

W code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code,Section III. The provisions of Specification 3.0.3 are not l

applicable.

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l 3.1.1.4 Reactor Internals Vent Valves l

The structural integrity and operability 'of the reacto:- internals vent valves shall be maintained at a level consistent with the acceptance criteria in Specification 4.1.

The provisions of Specification 3.0.3 are not applicable.

3.1.1. 5 Reactor Coolant Loops A.

With the reactor coolant average temperature above 280'F, the reactor coolant loops listed below shall ba operable:

Amendment No. 77, N,57 16

3.1.2 Pressurization, Heatup and Cooldown Limitations Specification 3.1.2.1 Hydro Tests For thermal-steady state system hydro tests, the system may be pressurized to the limits set ferth in Specification 2.2 when there are fuel assemblies in the core, under the provisions of 2

3.1.2.3, and to ASME Code limits when no fuel assemblies are present provided the reactor coolant system limits are to tbo right of and below the limit line in Figure 3.1.2-1.

The pro-visions of Specifications 3.0.3 and 3.0.4 are not applicable.

i 3.1.2.2 Leak Tests Leak tests required by Specification 4.3 shall be conducted under the provision of 3.1.2.3.

The provisions of Specifi-cation 3.0.3 are not applicable.

3.1.2.3 The reactor coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figure 3.1.2-2 and 3.1.2-3, and are as follows:

Heatup:

Allowable combinations of pressure and temperature shall bs to the right of and below the limit ifne in Figure 3.1.2-2.

The heatup rates shall not exceed those shown in Figure 3.1.2-2.

Cooldown: '

Allowable combinations of pressure and temperature for a specific cooldown shall be to the right of and below the Ifmit line in Figure 3.1.2-3.

Cooldown rates shall not exceed those shown in Figure 3.1.2-3.

3.1.2.4 The secondary side of the steam generator shall not be pressurized above 200 psig if the temperature of the steam generator shell is above 100F.

3.1.2.5 The pressurizer heatup and cooldown rates shall not exceed 100F/hr.

The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 430F.

3.1.2.6' With the limits of Specifications 3.1.2.3 or 3.1.2.4 or 3.1.2.5 exceeded, restore the temperature and/or. pressure to within the limit within 30 minutes; perform an engineering eval:1ation to determine the effects of the out-of-limit condition on the frac-

>ture toughness properties of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS Tavg and pressure to less than 200F and 500 psia, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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Amendment No. 21, #. 57 18 i

_ _ __ _.,.,..-.._- -._-, -..- ---..?- - - --.,

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3.1.2.7 Prior to reaching six effective full power years of operation, Figures 3.1.2-1, 3.1.2-2 and 3.1.2-3 shall be updated for the next service period in accordance with 10CFR50, Appendix G, Section V.8.

The service period shall be of sufficient dura-tion to permit the scheduled evaluation of a portion of the surveillance data scheduled in accordance with Specification 4.2.7.

The highest predicted adjusted reference temperature of all the beltline region materials shall be used to determine the adjusted reference temperature at the end of the service period..The basis for this prediction shall be submitted for NRC staff review in accordance with Specification 3.1.2.8.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

3.1.2.8 The updated proposed technical specifications referred to in

? '. 2.7 shall be submitted for NRC review at 'least 90 days prior to the end of the service period. Appropriate additional NRC review time shall be allowed for proposed technical specifications submitted in accordance with 10 CFR Part 50, f.ppendix G,Section V.C.

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Amendment No. 57 18a

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O Bases All reactor coolant system components are designed to withstand the effects of cyclic loads due' to system temperature and pressure changes.(1) These cyclic loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations. The number of thermal and loading The maxi-cycles used (or design purposes are shown in Table 4-8 of the FSAR.

mum unit heatup and cooldown rate of 100F per hour satisfies stress limits for cyclic operation. (2) The 200 psig pressure limit for the secondary side of the steam generator at a temperature less than 100F satisfies stress levels for temperatures below the DTT.(3)

The major components of the reactor coolant pressure boundary have been analyzed in accordance.with Appendix G to 10CFR50. Results of this analysis, including the actual pressure-temperature limitations of the reactor coolant pressure boundary, are given in BAW-1440(4).

Figures 3.1.2-1, 3.1.2-2, and 3.1.2-3 present the pressure-temperature limit curves for hydrostatic test, normal heatup, and normal cooldown respectively. The limit curves are applicable through the fifth effective full power year of operation. These curves are adjusted by 25 psi and 10F tor possible errors in the pressure and temperature sensing instruments. The pressure limit is also adjusted for the pressure differea-tial between the point of system pressure nessurement and the limiting com-ponent for all operating reactor coolant pump combinations.

The pressure-temperature limit lines shown on. Figure 3.1.2-2 for reactor criticality and on Figure 3.1.2-1 for hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50 for reactor criticality and for inservice hydrostatic testing.

The attual shift in RTNDT of the beltline region material will be established todically during operation by removing and evaluating, in accordance (h Appendix il to.10CFR50, reactor vessel material irradiation surveillance specimens which are installed near the inside wall of this or a similar reactor vessel in the core region'.

The spray temperature difference restriction based on a stress analysis of l

the spray line nozzle is bnposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit. Temperature require-ments for the steam generator correspond with the sensured NPTT for the shell.

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l Anendment No. 77, 78, 57 19 l

3.1.3 Minimio Conditiens for Criticality Specification 3.1. 3.1 The reactor coolant temperature shall be above 525F'except for portions of low power physics testing when the requirements of Specification 3.1.8 shall apply.

3.1.3.2 Reactor coolant temperature shall be to the right of the criti-cality limit of Figure 3.1.2-2.

3.1.3.3 When the reactor coolant temperature is below the minimum temp-erature specified in 3.1.3.1 above, except for portions of low power physics testing when the requirements of Spcification 3.1.8 shall apply, the reactor shall be subcritical by an amount equal to or greater than the calculated reactivity insertion due to depressurizat' ion.

3.1.3.4 The reactor shall be. maintained subcritical by at least 1 percent Ak/k until a steam bubble is formed and an indicated water level between 45 and 305 inches is est.iblished in the pressurizer.

3.1.3.5 Except for physics tests and as limited by 3.5.2.1, safety rod groups shall be fully withdrawn and the regulating rods shall be positioned within their position, limits as defined by Specifica-tion 3 5.2.5 prior to any other reduction in shutdown margin by deboration or regulating rod withdrawal during the approach to criticality.

3.1.3.6 The reactor shall not be made critical until at least 2 of the 3 emergency-powered pressurizer heater groups are operable.

With less than 2 of the 3 required heater groups ope,'able, restore the required heater groups to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

If the required heater groups are not restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , be in hot shutdown within the followinp 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.1.3.7 With any of the above limits violated, restore the reactor t,o within the limit in 15 minutes or be in at least Hot Shutdown w hin the next 15 minutes.

Bases l

At the beginning of life of the initial fuel cycle, the moderator tempera.

l ture coefficient is expected to be slightly positive at operating tempera-tures with the operating configuration of control rods.

(1) Calculations show that above E25F the positive moderator coefficient is acceptable.

Sincethemoderatohtemperaturecoefficientatloertemperatureswillbe less negative nr more positive than at operating temperature, (2) startup and operation of the reactor when reactor coolant temperature is less than 525F is prohibited except where necessary for low power physics. tests.

l The potential reactivity insertion due to the mo.derator pressure coeffi-l cient (2) that could result from depressurizing t'ie coolant from 2100 psia to saturation pressure of 900 psia is approximately 0.1 percent A k/k.

During physics tests, special operating precautions will be taken. In.addf-tion, the strong negative Doppler coefficient (1) and the sniall integrated Ak/k would limit the magnitude of a power excursion resulting from a r,e-duction of moderator density, i

Amendnent floe 7, LT, Ed,57 21

3.1.6 Leakage Specification 3.1.6.1 If the total reactor coolant leakage rate exceeds 10 gpm, the reactor shall be shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of detection.

3.1.6.2 If unidentified reactor coolant leakage (exceeding normal evaporative losses) exceeds 1 gpm or if any reactor coolant leakage is evaluated as unsafe, the reactor shall be shut-down within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of detection.

3.1.6.3 If it is determined that any reactor coolant leakage exists through a non-isolable fault in a reactor coolant system strength boundary (such as the reactor vessel, piping, valve body, etc., except steam generator tubes), the reactor shall be shutdown and a cooldown to the cold shutdown condition shall be initiated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of detection.

i 3.1.6.4 If reactor shutdowr. is required by Specification 3.1.6.1, 3.1.6.2, or 3.1.6.3, the rate of cooldown and the conditions of shutdown shall be determined by the safety evaluation for each case and reported as required by Specification 6.12.3.

3.1.6.5 Action to evaluate the safety implication of reactor coolant leakage shall be initiated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of detection.

The nature, as well as the magnitude of the leak, shall be consid-ered in this evaluation.

The safety eWJation shall 45sure that the exposure of offsite personnel to radiation is within the guidelines of 10CFR20, 3.1.6.6 If reactor shutdown is required by Specification 3.1.6.1, 3.1'.6.2, or 3.1.G.3, the reactor shall not be restarted unti.1.the leak is repaired or until the problem is otherwise corrected.

3.1.6.7 When the reactor is at power operation, three reactor cool-ant leak detection systems of different operating principles shall be in operation.

One of these-systems is sensitive to radioactivity and consists of a radioactive gas detector and an air particulate activity detector.

Both of these instru-ments may be out-of-service simultaneously for a period of no more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided two othe means are available to detect leakage and reactor building air samples are taken and analyzed in the laboratory at least once per shift; other-wis:e, 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 Cold Shutdown within the following 30' hours.

3.1.6.8 Loss of reactor coolant through reactor coolant pump seals and system valves to connecting systems which Amendment No. 57 27

s vent to the gas vent header and 'from which coolant can be returned i

to the reactor coolant system shall not be considered as reactor coolant leakage and shall not be subject to the consideration of Specifications 3.1.6.1 and 3.1.6.6 except that such losses when added to leakage shall not axceed 30 gpm.

3.1.6.9 If the reactor coolant system pressure isolation valve leakage is greater than the values given in Table 3.1.6.9, isolate (by having at least two valves in the high pressure piping closed *)-the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ind apply Specification 3.3.6, or be in at least hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Bases Every reasonable effort will be made to reduce reactor coolant leakage, fr.cluding evatorative losses (which may be on the order of 0.5 gpm), to prevent a large Reacto' building suno level, leak from masking the presence of a smaller leak.

r cater inventory balances, radiation monitoring equipment, boric acid crystalline depcsits, and physical inspections can disclose reactor coolant leaks. Any leak of radioactive fluid, whether from the reactor coolant syctem primary boundary or not can be a serious problem with respect to in-plant radioactive contamina-tion and cleanup or it could develop into a still more serious problem; and therefore, the first indication of such leakage will be followed up as soon as practicable.

Although some leak rates on the order of GPfi may be tolerable from a dose point of view, especially if they are to closed systems, it must be recognized that leaks on,the order of drops per minute through any'of the walls of the primary system could be indicative of materials failure such as by stress corrosion cracking. If depressurization, isolation and/or other safety measures are not taken promptly, these small leaks could develop into much larger leaks, possibly into a gross pipe rupture. Therefore, the nature of the leak, as well as the magnitude of the leakage must be considered in the safety evaluation.

When the source of leakage has been identified, the situation can be evaluated to determine if operation can safely continue. This evaluation will be performed by the Operating Staff and will be documented in writing and approvec by. tne Superintendent. Under these conditions, an allowable reactor coolant system leakage rate of 10 This explained leakage rate of 10 gpm is also availabIe;;m has been established.

even during a loss of off-site power.

If' leakage is to the reactor building it may be identified by one or more of th. following methods:

a.

Leakage is monitored by a level indicator in the reactor building sumo.

Changes in normal sump level may be indicative of leakage from any of the systems located inside the reactor building such as the reactor coolant system, service water system, intermediate cooling system and steam and feedwater lines or condensation of humidity within the reactor building atmosphere. ~he reactor building sump contains 63.5 gallons per incn of heignt. A 1 gpm leak would be detected in less than I hour.

Th2 motor operated va ve s ah ll remain closed and power supplies deenergi:ed.

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' Crder dtd.4/20/81 3

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O 3.1.7 Moderator Temperature Coefficient of Reactivity Specification 3.1.7.1 The moderator temperature coefficient (MTC) shall be non-pcsitive whenever thermal power is 2 95% of rated thermal power and shall be less positive than 0.5 x 10"' A k/k/*F whenever thermal power is <95% of rated thermal power and the reactor is not shutdown.

3.1.7.2 The MTC shall be determined to be within its lioits by con-firmatory measurements prior to initial operation above 5%

of rated thermal power after each fuel loading.

MTC measured values shall be extrapolated and/or compensated t o permit direct comparison with the limits in 3.1.7.1 above.

3.1.7.3 With the MTC outside any one of the above limits, be in at least liot Standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Bases A non positive moderator coefficient at p6wer levels above 95% of rated power is specified such that the maximum clad temperatures will not ex-ceed the Final Acceptance Criteria based on LOCA analyses. Below 95% of rated power, the Final Acceptance Criteria will not be exceeded with a positive moderator temperaturo coefficient of +0.5 x 10sik/k/*F corrected to 95% of rated power. All other accident analyses as reported in the FSAR have been performed for a range of moderator *emperature coefficients including +0.5 x 10-' Ak/k/'F.

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l Amendment No. 21. JJ. 57 30 l

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3.1.9 Control Rod Operation Specification

-3.1.9.1' The concentration of dissolved gases in the reactor coolant shall be limited to 100 std. cc/ liter of water at tne reactor vessel outlet temperature.

3.1.9.2 llowable combinations of pressure and temperature for control

. rod operation shall be to the left of and abcve the limiting pressureversustemperaturecurveforadissolved,Iigure3.1.9-1..

gas concen-tration of,100 std. cc/ liter of watcr as shown in 3.1.9.3 In the event the limits of Specifications 3.1.9.1 or 3.1.9.2

~

are exceeded, the center control rod drive mechanism shall be checked for accumulation of undissolved gases.

The temperature, pressure and dissolved gas concentration shall be-restored to within their limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 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 Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Bases By maintaining the reactor coolant-temperature and pressure as specified

'above, any dissolved gases in the reactor coolant system are maintained in solution.

Although the dissolved gas concentration is expected to be approximately

_20-40 std. cc/ liter of water, the dissolved gas concentration is conser-vatively assumed to be.100 std. cc/ liter of water at the reactor vessel outlet temperature.

The limiting pressure versus temperature curve for dissolved gases is determined by the equilibrium pressure versus temperature curve for the i

dissolved gas concentration of 100 std. cc/ liter of water.

The equili-I brium total pressure is.the sum of the partial pressure of the dissolved gases plus the partial pressure cf water at a given temperature. The margin of error consists of the maximum pressurre difference between the pressure sensing tap and lowest pressure point in the system, the maximum pressure gage error, and the pressure difference due to the maximum temp-l erature gage error.

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If either the maximum dissolved gas concentration.(100 std. cc/ liter of water) is exceeded or the operating pressure falls below the limiting pressure versus temperature curve, the center CRDM should be checked for accumulation of undissolved gases.

Amendment No. 57 3?.

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_.__..._.___.-,___._.-..___.__...__._._...-._l_..._.______.

3.5.2 Control Rod Group and Power Distribution Limits Appilcability This specification applies-to power distribution and onoration of control rods during power operation.

Objective To assure 'an acceptable core power distribution during pt.,wer operation, to set a limit on potential _ reactivity insertion from a hypothetical control rod ejection, and to assure core subcriticality after a reactor trip.

Specification 3.5.2.1

. The available shutdown margin shall be not less than 1% Ak/k with the highest worth control rod fully sithdrawn. With the-shutdown margin.less ' than 1%Ak/k, immediately initiate and continue boration injection until the required shutdown margin is restored.

3.5.2.2 Operation with inoperable rods:

1.

Operation with more'than one inoperable rod, as defined in Specification 4.7.1 and 4.7.2.3, in the safety or regulat-ing rod groups shall not be permitted.

2.

If a control rod in the regulating or safety rod groups is declared incperable in the withdrawn position as defined in Specification 4.7.1.1 and 4.7.1.3, an evaluation shall t

l be initiated immediately to verify the existence of 1%Ak/k l

available shutdown margin.

Boration may he initiated either l

c*

to the worth of the inoperable rod or until the regulating

..and transient rod groups are withdrawn to the limits of Specification 3.5.2.5.3, whichever occurs first.

Simultan-eously, a program of exercising the remaining regulating I

'and safety rods shall be initiated to verify operability.

l 3.

If within one (1) hour of the detennination of an inoper-l able rod as defined in Specification 4.7.1, it is not j

determined that a 1%Ak/k available shutdown' margin exists combining the worth of the inoperable rod with each of the i

other rods, the reactor shall be brought to the Hot Standby condition until this margin is established.

l 4.

Following the determination of an inoperable rod as defined in Specificaticn 4.7.1, all remaining rods shall be exer-cised within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months cnd exercised weekly until the rod problem is solved.

t 5.

If a control rod in the regulating or safety rod groups is declared inoperable per 4.7.1.2, power shall be reduced to l

60% of the thermal power allor ble for the reactor ccolant pump combination.

ge ent No. 57 46

3.5.3 Safety Features Actuaticn System Setpoints Applicability This specificatian applies to the safety features actuation system actuation setpoints.

Objective To provide for automatic initiation of the safety features actuation system in the event of a breath of rei.ctor coolant system integrity.

Specification The safety features actuation setpoints and permissible bypasses shall be as follows:

F_urctional Unit Action Setpoint High Reactor Building Reactor Building Spray 130 psig (44.7 psia)

Pressure

High Pressure Injection 14 psig (18.7 psia)

Start of Reactor Building Cooling and Reactor Building Isolation

$4 psig (18.7 psia)

Reactor Bldg. Ventilation

.54 psig (18.7 psia)

Low Pressure Injection

'14 psig (18.7 psia)

Penetration Room Ventilation

<4 psig (18.7 psia)

Low Reactor Coolant High Pressure Injection 2 1500 psig System Pressure **

Low Pressure Injection 2 1500 psig Start of Reat. tor Building Cooling and Reactor Building Isolation

> 1500 psig

May be bypassed during reactor building leak rate test.

- May be bypassed below 1750 psig and is automatically reinstated above 1750 psig.

With the safety features actuatioit setpoints lese conservative than the above values,_ declare the channel inoperable and apply the applicable Action requirements of Table 3.5.1-1 until the channel is restored to

-OPERABLE status with the trip setpoint adjusted consistent with the trip setpoint value.

Amendment No. 0,57 49

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^l 13.5.4 Incore Instrumentation

~

Applicability Applies to the operability of the incore instrumentation system.

Objective

,To specify the functional and operational requirements of the incore instrumentation system.

Specification

.Above 80 percent of operating power determined by the reactor coolant pump combination (Table 2.3-1) at'least 23 individual incore detectors shall be operable to check gross core power distribution and to assist in the periodic calibration of the out-of-core detectors in regard to the core imbalance trip limits. The detectors shall be arranged as follows and may be'a part of both basic arrangements.-

3.5,4.1 Axial Imbalance A.

Three detectors, one in each of three strings, sf. ail-lie in the same axial' plane with one plane in each axial core half.

B.

The axial planes in each core half shall be symmetrical about the core.nid plane.

C.

The detector shall not have radial symmetry.

3.5.4.2 Radial Tilt A. ' Two sets of four detectors shall lie in each core half.

..Each set of four shall lie in the same axial plane.

The two sets in the 'same core half may lie in the same axial plane.

B.

Detectors in the same plane shall have quarter core radial symnetry.

With the incore detector system inoperable, do.not use the system for the aoove applicable monitoring function.

The provi,s, ions of Specifications 3.0.3 and 3.0.4 are 'not applicat;1e.

Bases A system of 52.incore flux detector assemblies wit'.h 7 detectors per assem-bly has been provided primarily for fuel management purposes.

The system includes data display and record functions and is also used for out-of-core nuclear instrumentation calibration and for core power distribution verifi-cetion.

A.

The out-of-core nuciear instrumentation calibration includes:

l 1.

Calibrationofthesplitdetectorsatinitialreactor startup, during the power escalation program, and periodically thereafter.

Amendment No. 57 51

2.

A comparison check with the incore' instrumentation in the event c 1e of the four out-of-core power rance detector assemblies gives abnormal readings during ' operation.

3.

Confirmation that the oui ~of-core axial power splits are as expected.

B.

Core power distribution verification includes:

1.

Measurement at low power initial reactor startup 1 check that power dictribution is consistent with calculations.

2.

Subsequent checks during operation to insure that power distribucion is consistent with calculations.

3.

Indication of power distribution in the event that abnormal situations occur during reactor operation.

C.

The syf ty of unit operation at or below 80 percent of operating power \\1 for the reactor coolant pump combinations without the core imbalance trip systeli bas been determined ly extensive 3-D calculations. This will be verified during the physics startup testing program.

D.

The minimu:'i requirement for 23 individual incore detectors is bcsed on the following:

4 1.

An adequate axial imbalance indication can be obtained with 9 individual detectors. Figure 3 5.h-1 shows a typical set j

of three detector strings with 3 detectora per string that will indicate an axisl imbalance. The three detector strings are the center one, one from the inner ring of symmetrical strings and one from the outer ring of symmetrical strings.

2.

Figure 3.5.k-2 shows a typical detection scheme which will indicate the radial pcwer distribution with 16 individual detectors. The readings from 2 detectors in a radial quad-rant at either plane can be compared with readings _ from t'ae other quadrants to measure a radial flux tilt.

3.

Finure 3 5.h-3 combines Figures 3.5.h.1"and 3 5.h-2 to illus-trate a typical set of 23 individual detectors that can be specified as a ainimum for e.xial imbalance determir.ation and radial tilt indication, as well as for the htermination of gross pore power distributions. Startup i,esting will verify the adequacy of this set of dettctors for the above func-tions.

E.

At least E3 specified incore detectors will be operable to check power distribution above 80 percent power determined by reactor coolant p imp combination. 'Ihese incore detectors will be read.

out either on the computer or on a recorder. If a set of 23 52

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. 3.5.5 Fire Detection Instrumentation Applicability This specification applies to fire detection inctrumentation utilized within fire areas containing safety related equipment or circuitry for the purposes of protecting that safety related equipment or circuitry.

Objective To provide immediate notification of fires in areas where there exists a potential.for a fire to disable safety related systems.

Specification 3.5.5.1 A minimum of 50% of the heat / smoke detectors in the locations specified in Table 3.5-5 shall be operable.

3.5.5.2 If less than 50% of the fire detectors in any of the locations designated in Table 3.5-5 are operable, within one hour esta-blish-a fire watch patrol to inspect the zone (s) with the in-operable instrument (s) at least once per hour and restore the equipment to operable status within 14 days or prepare and submit a report to the Commission pursuant to Specification 6.12.3.2(b) within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the instrument (s) to operable status. The provisions Of Specifications 3.0.3 and 3.0.4 are not applicable.

Bases The various detectors provide alarms that notify the operators of the existence of a fire in its early stages thus providing early initiation of fire protection. The detectors in the main and auxiliary control rooms also provide automatic fire ~ protection nitiation.

The detectors required to be operable in the various areas represent one half of those installed.

Operability of the fire detection instrumentation ensures that operable warning capability is available for the prompt detection of. fires. This capability is required in order to detect and locate fires in their early stages.

Prompt detection of fires will reduce the pctential for damage to safety related equipment.

In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected area (s) is required to provide detection capability until the inoperable instrumentation is' restored to operability.

Amendment No. 37, #3, 57 53d J

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3.6 REACTOR BUILDING

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Applicability.

l Applies to the integrity of the reactor building.

L Objective l

To assure reactor building integrity.

_ Specification i

3.6.1 Reactor building integrity shall be maintained whenever all three (3) of the following conditions exist:

a.

Reactor coolant pressure is 300 psig or greater, b.

Reactor coolant temperature is 200F or greater.

i c.

Nuclear fuel is in the core.

Without reactor building integrity, restore the integrity within one hour 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 Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.6.2 Reactor building integrity shall be maintained when the reactor coolant system is open to the reactor building atmosphere and the requirements for a refualing shutdown are not met. The provisions of Specification 3.0.3 are not applicable.

3.6.3 Positive reactivity insertions which would result in the reactor being subcritical by less than 1% Ak/k shall not be.'nade by control rod ' motion or boron dilution whenever reactor building integrity is not in force. The provisions of Specification 3.0.3 are not applicable.

3.6.4 The reactor shall not ce taken critical or, remain critical l

if the reactor. building internal pressure exceeds 3.0 psig or a vacuum of 5.5 inches Hg. With the reactor critical,

, restore the c.ontainment pressure to within its limits with-l in one hour or be in at least Hot Standby within the next l

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l 3.6.5 Prior to criticality following a refueling shutdown, a ch'eck shall be made to confirm that all manual. reactor building isolation valves which should be closed are closed and locked, as required. The provisions of Specification 3.0.3 are not l

I applicable.

54 Amendment No. 57

- -- J -

3.6.6 If.. while the reactor is critical, a reactor building isola-tion valve is determined to be inoperable in a position other than the closed position, the other reactor building isolation i

. valve (except for check valves) in the line shall be tested to insure operability.

If the inoperable valve is not re-stored within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be brought to the cold shutdown condition within an additional Li lours or the operable valve will be closed.

Bases The reactor coolant system conditions of cold shutdown assure that ao steam will be formed and hence there will be no pressure buildup in the reactor building if the reactor coolant system ruptures.

The selected shutdown conditions are bas'ed on the type of activities that are being carried out and will preclude criticality in any occurrence.

The reactor building is designed for an internal pressure of 59 psig and an \\

external pressure 3.0 psi greater than the internal pressure. The design external pressure of 3.0 psi corresponds to a margin of 0 5 psi above the differential pressure that could be deve16 ped if the building is sealed with 0

an internal temperature of 110 F and the building is subsequently cooled to an internal temperature of less than 50*F.

When reactor building integrity is established, the limits of 10 CFR 100 vill' not be exceeded should the maximum hypothetical accident occur.

Re ferences FSAR, Section 5 l

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4 Amendment No.57 55

, _ _.. -. ~. _ _ _. _ _ _,, _..

3.8.6

,During the handling of irradiated fuel in the reactor building, at least one door on the personnel and emergency hatches shall be closed.

The equipment hatch cover shall be in place with a minimum of four bolts securing the cover to'the sealing surfaces.

.3.8.7 Isolation valves in lines containing automatic containment iso-lation valves shall be operable, or at least one shall be closed.

3.8.8 When two irradiated fuel assemblies are being moved simultaneously by the bridges within the fuel transfer canal, a minimum of 10 feet separation shall be~ maintained between the assemblies at all times.

3.8.9 If anv of the above specified limiting conditions for fuel loading and refueling are not met, movement of fuel into the reactor core shall cease; action shall be initiated to correct the conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be made.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

3.3.10 The reactor building purge isolation system, including the radiation monitors, shall be tes'ted and verified to be oper-able within 7 days prior to refueling operations.

The pro-visions of Specifications 3.0.3 and 3.0.4 are not. applicable.

3.8.11' Irradiated fuel shall not be removed from the reactor until-the unit has been subcritical for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The provisions of Specifications 3.0.3 and 3.0.4 are not appli-cable.

l 3.8.12 All fuel handling in the. Auxiliary Building shall cease upon I

notification of the issuance of a tornado watch for Pope, Yell, Johnson, or Logan counties in Arkansas.

Fuel headling opera-tions in progress will be completed to the extent necessary to place the fuel handling bridge and crane in their normal parked i

and locked position. :The provisions of Specifications 3.0.3 l

and 3.0.4 are not applicable.

l 3.8.13 No loaded spent-fuel shipping cask shall be carried above or l

into the Auxiliary Building equipment shaft unless atmospheric dispersion conditions are equal to or better than those pro-l duced by Pasquill Type D stability accompanied by a wind velo-city of 2 m/sec.

In addition, the railroad spur door of the Turbine Building shall be closed and the fuel handling area ventilation system shall.be'in operation.

The provisions of

^

Specifications 3.0.3 and 3.0.4 are not applicable.

3.8.14 For the maximum fuel pool heat load capacity (i.e., seven re-load batches - 413 assemblies - stored in the pool at the time of discharge of the full core), the full core to be discharged i

shall be cooled in the reactor vessel a minimum of 175 hours0.00203 days 0.0486 hours 2.893519e-4 weeks 6.65875e-5 months l

prior to discharge.

The preisions of Specifications 3.0.3 and-3.0.4 are not applicable.

Amendment No. IS, 77, y,57 59 l

.t.

,3.8.15 Loads _in. excess of 2000S pounds shall be prohibited from travel over fuel assemblies in the storage pool.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

3.8.16 The spent fuel shipping cask shall not' be carried by the Auxil-iary Building crane pending the' evaluation _of the spent fuel-cask drop accident and'the crane design by AP&L and NRC-review

~and approval.

The provisions of Specifications 3.0.3 and 3.0.4 i

are notiapplicable.

~

BASES:

~ Detailed written procedures _will be available for use by refueling personnel. These procedures, the above specifications, and.the design

.of the fuel handling equipment as_' described in Section 9.7 of the'FSAR incorporating built-in interlocks and safety' features, provide assurance that. c s. incident could occur during the refueling ~ operations that would

--result in a hazard to public health and safety.

If no change is'being made in core geometry, one flux monitor is sufficient. This permits maintenance on1the-instrumentation. Continuous monitoring of radiation 1evels and neutron flux provides immediate indication of an unsafe condition.

The requirement that at least one decay he'at removal loop be in opera-tion ensures that (1) sufficient cooling capacity is available to remove

' decay heat and maintain the water.in the reactor pressure vessel at.the refueling temperature-(normally.140*F), and (2) sufficient coolant circulation is maintained through-the reactor core to minimize the effects of a boron dilution incident and-prevent boron stratification.(3) 1 The requirement to have two decay heat removal loops operable when there is'less than 23 feet of water above the core, ensures that a single failure of the operating decay heat removai loop will not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling,'thus in the event of a failure of the operat-ing decay heat removal loop, adequate time is provided to initiate emergency procedures to cool the core.

'The shutdown margin indicated in Specification 3.8.4 will keep core I

suberitical, even with all control rods withdrawn from the core The boron concentration will be maintained above 1,800 ppe. Although this concentration is sufficient to maintain the core k

< 0.99 if all the

~

.controlrods'wereremovedfromthecore,onlyafe0kontrolrodswillbe removed at any one time during fuel shuffling and replacement. The k,ff L

with all rods in the core and with refueling boron concentration is approximately 0.9.

Specification 3.8.5 allows the control room operator to inform the reactor building personnel of any impending unsafe condi-tion detected from the main control board indicators during fuel move-ment.

The specification requiring testing reactor building purge termination is to verify that these cosionents will function as required should a

~

fuel handling accident occur which resulted in the release of rignifi-cant fission products.

Amendment No.17, $$, 57 59a j

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_. _. -.. _. _ _ _ _ _... _ _......., -. _. _. _. _.. _,, _.,. -. _. _, _ _. ~.. _... ~., _.. -.

Because of physical. dimensions of the fuel bridges, it is piysically impossible for fuel assemblics to be within 10 feet of cach other while being handled.

Spacification 3.8.11 is required as the safety analysis for the fuel handlingaccidentwasggjedon.theassumptionthatthereactorhadbeen shutdown for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Specification 3.8.14, which requires cooling of the full core for 175 hours0.00203 days 0.0486 hours 2.893519e-4 weeks 6.65875e-5 months prior to discharge to the spent fuel pool when seven reload batches ~ are aircady stored in the pool, is necessary to assure t hat the maximum design heat. load of the spent fuel pool cooling system will not be exceeded.

Specification 3.8.15 will assure that damage to fuel in the spent fuel pool will not be, caused by dropping heavy objects onto the fuel.

Ad-ministrative' controls will prohibit the storage of fuel in locationa adjoining the walls at the north and south ends of the pool, in the vicinity of cask storage area and fuel tilt pool access gates,.until the review specified in 3.8.16 is completed.

Specification 3.8.16 assures that the spent fuel cask drop accident c

cannot occur prior to completion of.the NRC staff's review of this potential accident and the completion of any modifications that may be necessary to preclude the accident or mitigate the consequences. Upon satisfactory completion of the NRC's review, Specification 3.8.16 shall be deleted.

REFERENCES (1)~ FSAR, Section 9.5 (2) FSAR, Section 14.2.2.3 (3) FSAR, S ction 14.2.2.3.3 Amendment No.EJ, 57 59b

3.10 SECONDARY SYSTEM ACTIVITY Aoplicability

-Applies to the limiting conditions of secondary system activity for operation of the reactor.

Objective To limit the maximum secondary system activity.

Specification The I-131 dose equivalent of the radioiodine activity in the secondary coolant shall not exceed 0.17 uCi/gm. With the secondary coolant ac-tivity in excess of 0.17 uCi/gm I-131, be in at least Hot Standby with-in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Bases For the purpose of determining a maximum allowable secondary coolcint activity, the activity contained in the mass released followin0 the rupture of a steam generator tube, a steam lice break outside contain-ment and a' loss of load incident were considered.

The whole body dose is negligible since any noble gases entering the secondary coolant system are continuously vented to the atmosphera by the condenser vacuum pumps.

Thus, in the event of a loss of load in-cident or steam line break, there are only small quantities of these gases which would be released.

The dose analysis performed to determine the maximum allowable reactor coolant activity assuming the maximum allowable primary to secondary leakaga of 1 gpm as given in the Bases for Specification 3.1.4.1 indi-cated that the. controlling accident to determine the allowable secon-dary coolant activity would be the rupture of a steam generator tube.

For the loss of load incident with a loss of 205,000 pounds of water released to the atmosphere via the relief valves, the resulting thyroid dose at the I-131 dose equivalent activity limit of 0.17 uCi/gm would be 0.6 Rem with the same meteorological and iodine release assumptions used for the steam generator tube rupture as given in the Bases for Specification 3.1.4.1.

For the less probable'accidpnt of a steam line break, the assumption is made that a loss of 1 x 10" pounds of water or the contents of one loop in' the secondary coolant system occurs and is released directly to the atmosphere.

Since the water will flash to steam, the total radioiodine activity is assumed to be released to the atmosphere. The resulting thyroid dose at the I-131 dose equivalent activity limit of 0.17 uCi/gm would be less than 28 Rem with the same meteorological assumptions used for the steam generator tube rupture and loss of load incident.

Amendment No. 57 66

3.12 MISCELLANEOUS RADIOACTIVE MATERIALS SOURCES Applicability Applies to byproduct, source, and special nuclear radioactive material sources.

i Objective

~

To ' assure that leakage from byproduct, source, and special nuclear radio-active material sources does not exceed allowable limits.

Specification 3.12.1 The source leakage test performed pursuant to Specification 4.14 shall be capable of detecting the presence of 0.005 uCi of radio-active material ~on the test sample.

If the test reveals the pre-sence of 0.005 uCi or more of removable contamination, it shall immediately be withdrawn.from use, decontaminated and repaired, or be disposed of in accordance with Commission regulations.

Sealed sources are exempt f rom such leak tests when the source contains 100 uti or less of beta and/or gamma emitting material or 5 uCi or less of alpha emitting material.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

3.12.2 A special report shall be prepared and substitted to the Commis-sion within 90 days if source leakage tests reveal the presence of 2 0.005 microcuries of removable contamination.

3.12.3 A complete inventory of licensed radioactive materials in posses-sion shas! be maintained current at all times.

l l

Amendment No. 73, 38,57 66b l

_,y*

' 3.15 ' FUEL HANDLING AREA VENTILATION SYSTEM Applicability Applies to the operability of the fuel handling area ventilation system.

Objectivb, To ensure that the fuel handling area ventilation system will perfo m within acceptable levels of efficiency and reliability.

Specification 3.15.1

-The fuel handling area ventilation system shall be in operation whenever irradiated fuel hanciling operations are in progress in the fuel handling area of the ' auxiliary building and shall have

~

the following performance capabilities:

a.

The results of the in place cold 00P and halogenated hydro-carbon tests at design flows (+ 10%) on HEra filters and.

charcoal adsorber banks shall show > 99% 00P removal and

> 99% halogenated hydrocarbon removal.

b.

The results of laboratory carbon sample ' analysis shall show

> 90% radioactive methyl iodide removal at a velocity within T 20% of system design, 0.03 to.0.15 mg/m 3 inlet methyl io-Hide concentration,> 70% R. H. ~ and > 125F. _

c.

Fans shall be shown to operate within 1 10% design flow.

d.

The pressure drop across the combined HEPA filters and

~

charcoal adscrber banks shall be less than 6 inches of water at system design flow rate (1 10%).

e.

Air distribution shall be uniform within + 20% across HEPA filters and charcoal adsorbers when testeli initially and after any maintenance or testing that could affect the air distribution within the fuel handling area ventilation system.

l 3.15.2 If the' requirements-of Specification 3.15.1 cannot be met, irra-diated fuel movement shall not be started (any irradiated ft.el i

assembly movement in progress may be completed).

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

Bases

(

The fuel handling area ventilation system is designed to filter the auxil-iary building atmosphere during fuel handling operations to limit the re-lease of activity should a fuel handling accident occur.

The system con-sists of one circuit containing two exhaust fans and a filter train.

The fans are redtndant and only one is required to be operating.

The filter train consists of a prefilter,.a HEPA filter and a charcoal adsorber in series.

~ Amendment No. JD,- 75, 57 66g i

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,s-3.17 FIRE SUPPRESSION WATER SYSTEM A>plicability J

-This specification applies to the portions of the fi're suppression water systtm necessary to provide fire protection to safety related equipment.

Objecti_ve_

To assure'that. fire suppression is available to safety related equipment.

Sp_eci fication 3.17.1 The fire suppression water system shall be operable at all times with.two high pressure pumps, each with a capacity of at least 2500 gpm, with their discharges aligned to the fire suppression header, and with an operable flow path capable of transferring water through distribution piping with operable sectionalizing control valves to the shutoff valve ahead of each hose standpipe and the water flow alarm device on each sprinkler system.

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3.17.2 With one pump inoperable, restore the inoperable equipment to operable status within seven days or prepare and submit a report to the Commission pursuant to Specification 6.12.3.2(b) within the next 30 days outlining the plans and precedures to be used to provide for the loss of redundancy in this system.

3.17.3 With the fire suppression water system inoperable:

Establish'a backup fire suppression water system within 24 a.

hours; and b.

submit a report in accordance with Specification 6.12.3.1(b).

.If "a." above cannot be fulfilled, place the reactor in Hot c.

Standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Cold Shutdown within l

the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The provisions of Specification 3.0.4 are not applicable.

i Bases The fire pumps supply the on1v rource of water for all fire suppression systems utilizing water.

I we pump is indiviJually capable of providing full flow required for proper fire suppression. water system operation.

The pumps start automatically on low system pressure, In the event that the fire suppression water sys' tem becomes inoperable, I

immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant.

The requirement for a 24-hour report to the Commission provides for prompt evaluation of the acceptability of the corrective measures to provide adequate fire suppres-sion capability for the continued operation of the plant.

Amendment No. M, 57 66m

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3.18 TIRE SUPPRESSION SPRINKLER SYSTEM Applicability This specification applies to the following fire sup~pression sprinkler systems protecting' safety related areas:

a.

Each ~of the four-reactor building cable penetration areas.

b.

Each.of_the four cable penetration rooms.

c.

Each of the two emergency diesel generator rooms.

d.

Cable spreading room.

e.

.Each of the two diesel generator fuel vaults.

.f.

Hallway El 372 (Zone 98-J).

g.

Condensate demineralizer area.'*

Objective To assure that fire suppression is available to safety related equipment located in the above-listed areas.

Specification 3.18.1 ihe above-listed sprinkler systems shall be operable at all times.

3.18.2 With one or more of the.above-listed sprinkler systems inoperable, establish a continuous fire watch (or operable smoke and/or heat detection equipment with ' control room alarm) with backup fire sup-pression equipment for the applicable area (s) within one hour.

Restore the system (s) to operable status within 14 days or prepare and submit a report to the Commission pursuant to Specification-6.12.3.2(b) within the next 30 days outlining-the action taken, the cause of the inoperability and the plans and schedule for re-storing the system (s) to operable status.

The provisions of Spe-cifications 3.0.3 and 3.0.4 are not applicable.

Bases Safety related equipment located in various areas'is protected by sprinkler systems.

The operability of these systems ensures that anequate fire sup-pression capability is available to confine and. extinguish a fire occurring in.the applicable areas.

In the event a system is. inoperable, alternate backup fire fighting equipment or operable detection equipment is required to be made available until the inoperable equipment is restored to service.

To be implemented no later than July 30, 1979.

Amendment No. 79, #7, 57 66n

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1 s-l (3.19 CONTROL ROOM AND AUXILIARY CONTROL ROOM hat 0N SYSTEMS Applicability This specification applies to the llalon systems utilized as the fire suppression system for the control room and auxilia"y control room.

Objective To assure that fire suppression is available to the safety related equipment'in the control room and auxiliary control room.

Specification 3.19.1

-The three control room and auxiliary control room Halon

. systems shall be operable at all times with'the storage tanks having at'least 95% of full charge weight and 90%

of full charge pressure.

3.19.2 With any of the control room-and auxiliary control room Halon systems inoperable, establish backup fire suppres-sion equipment for the affected area within one hour.

Restore the-system (s) to operable status within 14 days or prepare and submit a report to the Commission pursu-ant tu Specification 6.12.3.2(b) within 30 days outlining the. action taken, the cause of the inoperability, and the

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plans and schedule for restoring the system to operable status. The provisions of Specifications 3.0.3 and 3.J.4 are not applicable.

Bases Safety related circuitry located in portions of the control room and auxiliary control room is protected by the Halon systems.

The opera-bility of these, systems ensures that adequate fire suppression capa-bility is available'to confine and extinguish a fire occurring in the control room or the auxiliary control room.~

In the event that the system (s) is inoperable, a' ternate backup fire fighting equipment is required to be made available in the affected area until the inoperable equipment is restored to service.

4 660 Amendment No. 30,57

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i 3.20 FIRE HOSE S'ATIONS T

Applicability J

This specification applies to the fire hose stations protecting areas containing safety.related equipment.

Objective To assure that manual. fire suppression capability is available to all

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safety related equipment.

Specification 3.20.1 All fire hose stat' ions protecting areas containing safety related equipment shall be operable whenever the equipment in these areas i's required to be operable.

3.20.2 With one or more of the fire hose stations of 3.20.1 inoper-able, route an additional equivalent capacity fire hose to the unprotected area (s) from an operable hose station within opa hour. -The-provisioas of Specifications 3.0.3 and 3.0.4 are not. applicable.

Bases The operability of the. fire hose stations adds additional assurance that adequate fire suppression capability is available to confine and'extin-guish fires occurring in any portion of the facility where safety related equipment is' located.

In'the event that portions of this system are inoperable, backup fire hose equipment is required to-be made available in the affected area (s) until the inoperable equipment is restored to service.

Amendment No. 39, 57 66p n

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3.21 PENETRATION FIRE BARRIE."S Applicability This specification applies to penetration fire barriers relied on for restriction of fire damage such that safety related equipment ir areas other than the main fire area are not affected.

Objective To assure that penetration fire barriers protecting safety related areas perform their separation function.

Specification 3.21'.1 All penetration fire barriers protecting safety related areas shall be intact'at all times.

3. 2?. 2

With ca or more of the required penetration fire barriers not intact, establish a continuout fire watch (or operable snoke and/or heat detection equipment.with control room alarm; on at least one side of the affected penetration within one hour.

The provisions of Specifications 3.0.3 and 3.0.4 are not appli-l cable.

Bases-The functional integrity of the penetration fire ba.riers ensu m that fires will be confined or adequately retarded from spreading to adjacent portions of'the facility.

This design feature minimizes the possibility of a single fire rapidly involving several areas of the facility prior to detection and extinguishment.

The penetration fire barriers are a passive eieraent in the. facility fire protection program and are subject to. periodic inspections.

During periods"of time when the barriers are c)t functional, a continuous.

fire watch or operable detection equipment is required to be maintained l

in the vicinity _of the affected barrier until the barrier is restored to functional status.

i i-Amendment No. $$, 57 66q

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.3.22 REACTOR BUILDING PURGE FILTRATION SYST'EM

. Applicability-This specification applies to the operability-of the reactor building purge filtration system.

[0bjective

. To-assure that the reactor building purge filtration system will perform within acceptable levels of efficiency and reliability.

Specification 3.22.1 The-reactor building purge filtration system shall be operable whenever irradiated fuel handling operations are in progress in the reactor b'uilding and shall have the following performance capabilities:

The results of the l'n place cold 00P and halogenated hydro-a.

carbon tests at design flows (+ 10%) on HEPA filters and charcoal adsorber banks sha'll show > 99% DOP removal and

> 99% halogenated hydrocarbon removal.

b.

The-results of laboratory carbon sample analysis shall show

> 90% radioactive methyl iodide removal at a velocity within

' T 20% of system design, 0.05 to 0.15 mg/m3 inlet methyl io-Hide concentration,> 70% R. H. and >125F.

c.

Fans shall be shown to. operate within 1 10% design flow.

d.

The pressure drop across the combined HEPA filters and charcoal adsorber banks shall be less than 6 inches of-water at system design. flow rate (1 10%).

" Air distribution shall be unifc.rm within + 20% across HEPA e.

filters and charcoal adscrbers when tested initially and after any maintenance or testing that could affect the air distribution within the reactor building purge filtration system.

3.22.2 If the requirements of Specification 3.22.1 cannot be met, eithcr:

a.

Irradiated fuel movement shall not-be started (any irradiated fuel assembly movement in progress may be completed); or b.

Isolate the reactor building purge system.

3.22.3 The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.l Bases 4

The reactor building purge filtration system is designed to filter the're-actor building atmosphere'during normal operations for ease of personnel

^ entry into the reactor buildi.ng.

This specification is intended to re-

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quire the system operable during fuel handling operations, if the system

^ Amendmcnt No. $$,57 66r

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ML19351A047

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