ML19345C785

Proposed Tech Specs,Sections 1.0 & 3.0,incorporating NRC Std Definition of Operable & Changing Limiting Conditions for Operation,Respectively

ML19345C785
Person / Time
Site:	Arkansas Nuclear
Issue date:	11/28/1980
From:	ARKANSAS POWER & LIGHT CO.
To:
Shared Package
ML19345C782	List: ML19345C781 ML19345C785
References
IEB-79-05A, IEB-79-5A, NUDOCS 8012080347
Download: ML19345C785 (30)

Text

.

BASIS FOR CHANGE ~

This change is to be made to incorporate the NRC's Standard Technical

. Specification definition of OPERABLE and its associated Limiting Con-dition for Operation as required in the NRC's letter of April 10, 1980.

The effects of this change'will be to specifically limit, for each Lim-iting Condition for Operation, the time available to take corrective action prior to placing the unit in a lower power condition. If an ex-ception is not stated in the Limiting Condition for Operation, Specifi-

. cation 3.0.3 will apply.

4 e

e i

l l

f 90l%090347

- -x .

refueling. temperature-(normally-140F). Pressure-is defined by Specifi-cation 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.

~

11.2.7 Refueling 0peration-

~

An operation 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 '0PERABLE or 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-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 protective channels and associated circuitry which' forms the automatic system thot 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 9

2 s

l

3. LIMITING CONDITIONS FOR OPERATION 3.0 LIMITIN3 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 Operation within the specified time interval shall constitute com-pliance'with the specification. -In the event the Limiting Condition for Operation is restored prior to expiration of the specified time interval, no further actions-need be taken.

3.0.3 In the event a Limiting Condition fo. Operation cannot be sat-isfied because of circumstances in excess of those addressed in the specification, the unit shall be placed in at least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least-COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless corrective measures are completed that would permit operation or until the reactor

~

is placed in a condition in which the specification is not applicable.

Exceptions to these requirements shall be stated in the individual specification.

3.0.4 Entry into an Operating Condition or other specified applicability condition shall not be made unless the conditions of the Limiting Con-dition for Operation are met without reliance on provisions contained in the ACT10N statements unless otherwise excepted. This provision shall

. not preveat passage through Operating Conditions as required by a Limit-ing Cond'. tion for Operation.

3.0.5 vhen a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or. sole'y because its normal power source is inoperable, it may be cc .-

sidered OPERABLE for the purpose of satisfying the requirements of its .

applicable Limiting Condition for Operation, provided: (1) its corres-ponding no nal or emergency power source is OPERABLE, and (2) all of its redundant sys um(s), subsystem (s), train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of this specification.

Unless both condiv. ions (1) and (2) are satisfied, th' e unit shall be placed in at least HOT SlANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in at least HOT SHUTDOWN within

'the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least COLD SHUTDOWN within the following 3C hours. This specification is not applicable in Cold Shutdown or Retuel-ing Shutdown.

. 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 whea each specification is applicable.

15a

G

• 3.0.'2 - This specification defines those conditions .necessary to consti-

. 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

_ occurrence would violate the~ intent of the specification.

.3.0.4. lThis specification provides that entry into an Operating Condition ~

or other specified applicability condition must be made with (a) the full

complement of required systems, equipment or bomponents OPERABLE _and (b) all other parameters asispecified in the Limiting Conditions for Operation i . 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 i

specified limits being exceeded.

Exceptions to this'provisien have been provided for a limited number of specifications-when startu; with inoperable equipment would not affect

plant. safety. -These exceptions are stated in the Limiting Condition for Operet. ion n,t.atements 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 l source is not OPERABLE. It specifically prohibits operation when one -

division is inoperable because its normal or emergency power source is *

~

inop'erable and a system, subsystem, train, component or device in another

. division is inoperable for another rcason. .

. - -

1 The provisions of this specification permit the Limiting Condition for

] Operation statements associated with individual systems, subsystems, trains, components or: devices to be consistent with the-Limiting Condi-i tion for Operation statements of the associated electrical power source.

It allows operation to be governe'd by the time limits of the: Limiting 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 l power source. .

'For example, Specifi. cation 3.7.2.C provides for a 2 day out-of-service time when one emergency diesel generator is not OPERABLE. If the defi -

[ nition of OPERABLE were applied without consideration of Specification i

3.0.5, all systems, subsystems, trains, components and devices. supplied

. by the ' inoperable emergency power source would also be troperable. This would dictate invoking the appl.icable ACTION statements for each of the applicable Limiting Conditions for. Operation. However, the provisions 2 of- Specification 3.0.5 permit the time limits for continued operation to j be consistent with the Limiting Cendition for Operation statement for

.the inoperable emergency d h .'. generator instead, provided the'other I,

specified conditions are satisfied. In this case, this would mean that

< the' corresponding normal power source must be OPERABLE, and all redundant e

l 15b.

1-

., ,.w., <,.m . , . - 4 ,,.- - - , .+- c. , - .,m

n .

~

a'

^

. systems, subsys. ems,ftrains,Leomponents and'd2 vices must-be OPERABLE, or

. otherwise sati.fy -Specification'3.0.5 (i.e. , be . capable of p rforming .

'theiridesign fu'nction_and have at least one normal or one emergency power source OPERABLE). If they are ~not satisfied, shutdown is required in

' ~

-accordance.with this1 specification.

- As-a furtherfexample, Specif? cation 3.7.1.A requires in part that two

, physica11yjindependent circuits between the offsite transmission network-and the.onsite Class IE distribution system be OPERABLE. Specification-

-3.~7.2.8 provides a 24-hour _ out-of-service time when both required offsite circuits are not OPERABLE. 11 tha definition of OPERABLE were applied l

without consideration of Specification 3.0.5, all systems, subsystems, -

_ trains, components and devices supplied by.the inoperable normal pov i

sources, both of the offsite circuits would also be inoperable. -This would _ dictate invoking the applicable Limiting Condition for Operatio statements for each of the' applicable LCOs. However, the provisions of t

, ' Specification 3.0.5 permit the time limits for continued operation to be l

consistent with the Limiting Condition far Operation statement for the.

inoperable rormal power sources instead, provided the other specified

.. conditions are satisfied. In this cate, this would mean that for one

' - 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-i --

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 OPERABLE). In other words, both emergency power sources-must be OPERABLE and all redundant systems, subsystems, ~ trains, compo-nents and devices in.both divisions must also be.0PERABLE. 'If these

- conditions are not satisfied . shutdown is required in accordance with this specification.

During Cold Shutdown and' Refueling Shutdown,. Specification 3.0.5 is not applicable and'thus the individual ACTION statements for each applicable

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

0 1

4 U

l 6

f f

k'

, 15c

j .~

e+---- --y e -9 + F-v- T -- n$' r v 1--= ---

• u- *wer--+1 - ee **' "

4

' 3.^1L REACTOR COOLANT SYSTEM

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

Objective To specify_those limiting conditions for operation of the rea;ter coolant

~

system which must be' met to ensure safe reactor operations.

3.1.1 Operational Com onents 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 unle:s at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.

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. ,

- 3.1.1.2 Steam Generator_

r A. One steam; generator shall be operable whenever the reactor coolant average temperature is above 280F.

3.1.1.3 Pre'ssurizer Safety Valves A. Both pressurizer code safety valves shall be operable when the reactor is critical . With one pressurizer code safety valve operable, either restore the valve to operable status within 15 minutes or be in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. .

When the reactor is subcritical, at least one pressurizer 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 applicable.

3.1.1.4 Reactor' Internals Vent Valves The structural integrity and operability of the reactor 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.

e .

L 16

, ,e n.-, n . - e,-

r- - - ,. - ---..e,-

3.1.2 Pressurization, Heatup and Cooldown Limit:tions Specification 3.1.2.1 Hydro Tests

. For_ thermal steady state system hydro _ tests,- the system may be pressurized to the limits set forth in Specification 2.2 when-there are fuel assemblies in the core, under the provisit ,s of

- 3.1.2.3, and to ASME Code limits when no fuel assemblies are present provided the reactor coolant system limits are to the 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.

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 Specif.i-cation 3.0.3 are not applicable.

3.1.2.3 The reactor coolant pressure and the system heatup and cooldown rate: (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 follos.r; Heatup:

Allowable combinations of pressure and temperature shall be to the right of and below the limit line 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 limit 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 evaluation to -

determine the effects of the out-of-limit condition on the frac-ture_ toughness properties of the Reactor Coolant System; deterrine that_ the Reactor Coolant System remains acceptable for contint.ed operations or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS Tavg and pressure to less than 200F and 500 psia,~respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

18

' , J

3.1.2.7 Prior to reaching six effective full power years of cperation, Figures 3.1.2-1, '.1.2-2 and 3.1.2-3 shall be updated for the

.next service period in accordance with 10CFR50,- Appendix G,Section V.B. The service period shall be of sufficient dura-tion to permit the scheduled evaluation ~of a portion of the strveillance 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 3.1.2.7. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

O i

l

-2 18a i

~~ " - - - - v.., ,.,.

- 3.1.3 iMinfmum Conc'itions 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 shal'1 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 Specification 3.1.8 shall apply, the reactor shall be subtritical by an amount equal to or greater than the calculated reactivity insertion due to depressurization.

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 established 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 With any of the above limits violated, restore the reactor to within the limit in 15 minutes or be in at least Hot Shutdown within the next 15 minutes.

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

ture coefficient is expected to be slightly positive at operating tempera-tures with the operating configuration of control rods. (1) Calculations show that above 525F the positive moderator coefficient is acceptable.

Since tne moderator temperature coefficient at lower temperatures will be less negative or 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.

The potential reactivity insertion due to the mo.derator pressure coeffi-cient (2) that could result from depressuri:ing the 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 addi-tion, the strong negative Doppler coefficient (1) and the small integrated Ak/k would limit .the magnitude of a power excursion resulting from a re-duction of moderator density.

'A 21

-The requirement that the reactor is not to be mi itical below the limits

~

of Figure 3.1.2-2 provides increased assurances tho, the proper relationship between primary coolant. pressure and temperatures will be maintained rela-tive .to the NDTT of the primary coolant system. Heatup to this temperature will be accomplished by operating the reactor coolant pumps.'

4 e

D w

t g e O

e d

e e

e e

• 21a l

• ~

. 3.1. 6 Leakaga

~

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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.

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

3.1.6.4 If reactor shutdown 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of detection. The nature, as well as the magnitude of the leak, shall be consid-ered-in this evaluation. The safety evaluation shall assure that the exposure of offsite personnel to radiation is withi.n the guidelines of 10CFR20.

3.1.6.6 If reactor shutdown is require'd by Specification 3.1.6.1, 3.1.6.2, or 3.1.6.3, the reactor shall not be restarted until 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided two other means are available to detect leakage and reactor building air samples are taken and analyzed in the laboratory at leas.t once per shift; other-wise, be in at lent Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and' in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3.1.6.8 Less of reactor coolant through reactor coolant pump seals and system valves to connecting systems which 3

27

o. .

' 3.1.7 Moderator Terperature Coefficient of' Reactivity Specification-T The moderator temperature coefficient (MTC) shall bs non-

~

. 3.1.7.1 positive whenever thermal power is- 195% of rated thermal power and shall be less positive ~than 0.5 x 10-' lik/k/*F

~

- lwhenever~ 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 limits 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 to permit' direct comparison'with the limits in 3.1.7.1 above.

~

13.1.7.3 Withi he MTC outside any one of the above limits', be in at least slot Standby _within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

i

' Bases A:non positive moderator coefficient at power-levelsrabove 95% of: rated

}

power is specified;such that the maximu'm 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 temperature coefficient of +0.5 x 10-'tik/k/*F corrected to 95% of' rated power. All other accident analyses as reported in.the -

FSAR;have been performed for.a range.of moderatorftemperature coefficients including +0.5 x 10-' 4k/k/* F.

4 n

I'

+

j- .

~

7 i

M l

30 Y ,_

3'.1.9- Control Rod Operation .

Speci fication

~

. 3.1.9.1 ' The concentration of dissolved gases in tSe reactor coolant shall be limited to 100 std. cc/ liter of water at the 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 ab.ove the limi. ting pressure'versus temperature curve for a dissol.ved gas concen-tration of 100 std. cc/ liter of water as shown in Figure 3.1.9-1.

~

~

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 dis' solved gas concentration shall be restored to within their limits Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least Hot Standby i

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

Bas ~es .

~ ~

By maintaining th'e reactor toolant temperature and p'ressure as . spec _i[fied

- above, a'ny dissolved gase's in the reactor ' coolant system are maintained

~ ' ~

~

in solution. _ .-

~

Although the dissolve ~d gas' concentration is expected to b'e approximate 1y' 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 vesse.1 outlet temperature. -

The limiting pressure versus temperature curve for dissolved gases is -

determined by the equilibrium pressure versus temperature curve for the dissolved gas concentration of 100 std. cc/ liter of water. The equili-brium total pressure is the sum of the partial pressure of the dissolved .

gases plus the partial pressure of water at a given temperature. The margin of error consists of the maximum pressure 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-erature gage error. .

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 CROM should be checked for accumulation of undissolved gases.

1. 32 4

3 3.5.2 Control Rod Group and Power Distribution Limits Applicability This specification applies to power distribution and~operaton of control rod,s during power operation.

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

Specification 3.5.2.1 The avAlilable sh'utdown margin shall be not less than 1% Ak/k with the highest worth control rod fully withdrawn. 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 regu1 &

ing rod groups shall not be permitted.

~

2. If a control rod in the regulating or safety rod groups is declared inoperable in the withdrawn position as defined in. Specification 4.J.1.1 and 4.7.1.3, an evaluation shall be initiated immediately to verify the existance of 1%Ak/k available shutdown margin. Boration may be initiated either to the worth of the inoperable rod or until'the regulating aid 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 and safety rods shall be initiated to verify operability.

3. If within one (1) hour of the determination of an inoper-able rod as defined in Specification 4.7.1, it is not determined that a 1%Ak/k available shutdown margin exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be brought to the Hot Standby condition until this margin is established.

4. Following the determination of an inoperable rod as definad in Specification 4.7.1, all remaining rods shall be exer-cised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised weekly until the rod problem is sohed.

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 60% of the thermal power allowable for the reactor coolant

• pump combination.

46

h

- 3.5.3 Safety Features Actuatfon System Setpoints

- Applicability - -

- This specification applies to the safety features ac'tuation system act,uation setpoints.

Objective ,

To providt for automatic initiation of the safety features actuation system in the event of a' breach of reactor coolant system integrity.

$pecification The safety features 7.ctuation setpoints and permissable oypasses shall be as follows: ,

Functional Unit ' Action Setpoint High Reactor Building Reactor Building Spray- 4 30 psig (44.7 psia)

. Pressure

• High Pressure Injection L4 psig (18.7 psia)

Start of Reactor Building Cooling and Reactor Building Isolation 64 psig (18.7 psia)

Reactor Bldg. Ventilation $3 psig (18.7 psia)

- Low Pressure Injection <4 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 > 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 actuation setpoints less 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.

Bases High Reactor Building Pressure

' The basis for the 30 psig and 4 psig setpoints for the high pressure-signal is to establish a setting ~which would be . reached in adequate time in the event of a DBA, cover a spectrum of break sizes'and yet be far enough above normal operation maximum internal pressure to prevent spurious initiation.

4 49

3'.5.4' Incore-Instrumentation

. Applicability

, Applies to the operability of the incore instrt: mentation system.

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

Specification

' bove 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 _imbal'ance' trip limits. Th'e detectors shall be arranged. as.

follows and may be a part of both basic arrangements.

3.S.4.1 Axial ~ Imbalance A. Three detectors, one in each of three strings,- shall lie

'in the same axial plane with one plane in- each axial core half.

B. The axial planes in each core half sh'all be symmetrical about the core mid 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.

i B. Detectors in the same plane shall have quarter core radial symmetry.

With the incore detector system inoperable, do not'use the system for the above applicable monitoring function. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

Bases _

A system of 52 incore flux detector assemblies with 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-cation.

A. The out-of-core nuclear instrumentation calibration includes:

4

1. Calibration of the split detectors at initial reactor startup, during the power escalation program, and

-periodically thereafter.

51

3'.5.5- Fire Detection Instrumentation Applicability This specification applies to fire detection instrumentation utilized within. fire areas.containing safety related equipment or circuitry for the' purposes of protecting that safety related equipment of 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 Tab'le 3.5-5 shall be operable.

3.5.5.2 If less than 50% of the f. ire detectors in any~ of the Tocations 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 i,tatus. 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 auxiliaiy control rooms also provide automatic fire protection initiation.

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 potential 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.

e 53d

- 3. 6 REACTOR BUILDING Applicability ,

Applies to the integrity of the reactor building.

Objective To assure' reactor building integrity.'

Specification 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.

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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30

. hours.

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 refueling 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% a k/k 'shall not be made 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 be taken critical or remain critical

- if the reactor building internal pressure exceeds 3.0 psig or a vacuum of 5.5 inches Hg. With the reactor critical, restore the containment pressure to'within its limits with-in one' hour or be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3.6.5 Prior to criticality following a refueling shutdown, a check.

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 8

. applicable.

1 i

48 f 54 l l

4 f 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 valve (except for check v'alves) in the line shall be tested to insure' operability; If the inoperable ' valve is not re-stored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be brought to the

_. cold shutdown condition within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the operable valve will be closed.

Bases The reactoi'c'oolant system conditions of cold shutdown assure that no steam will be formed and hence there will be no pressure buildup in the

< - reactor building if the reactor coolant system ruptures.

9 i

e 4

n 4

e 54a

3'. 8 FUEL LOADING AND REFUELING Applicability

. Applies to fuel loading and refueling operations.

Obfective To assure 'that fuel loading, refueling and fuel handling operations are performed in a responsible manner.

Specification 3.8.1 Radiation levels in the reactor building refueling area shall be monitored by instrument RE-8017. Radiation levels in the spent fuel storage area shall be monitored by instrument RE-8009. If any of these instruments become inoperable, portable survey instrumentation, having the appropriate ranges and sen-sitivity to fully protect individuals involved in refueling operation, shall be used until the permanent instrumentation is returned to service.

3.8.2 Core subcritical neutron flux shall be continuously monitored by at least two neutron flux monitors, each with continuous

- indication available, whenever core geometry is being changed.

- When core geometry is not being changed, at least one neutron flux monitor shall be in service.

3.8.3 At least one decay heat removal pump and cooler and its cooling water supply shall be operable.

3.8.4 During reactor vessel head removal and while loading and unload-ing fuel from the reactor, the boron concentration shall be-maintained at not less than that required for refueling shutdown.

3.8.5 Direct communications between the control room and the refuel .

ing personnel in the reactor building shall exist whenever changes in core geometry are taking place.

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 contai[iment 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.

. g 58

3.8.9 If any of the above specified limiting conditians for fuel loading and refueling are not act, 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.

3.8.10 The reactor building purge isolation system, including the radiation monitors, shall be tested and verified to be oper-able within 7 days prior to refueling operations. The pro-visiens 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The provisions of Specifications 3.0.3 and 3.0.4 are pot appli-cable.

3.8.12 All fuel handling in the Auxiliary Building shall cease upon notification of the issuance of a tornado watch for Pope, Yell, Johnson, or Logan counties in Arkansas. Fuel handling opera-tions in progress will be completed to the extent necessary to place the fuel handling bridge and crane in their normal parked and locked position. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

3.8.13 No loaded spent fuel shipping cask shall be carried above or into the Auxiliary Building equipment shaft unless atmospheric

- dispersion conditions are equal to or better than those pro-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 ossemblies - stored in the pool at the time of discharge of the full core), the full core to be discharged shall be cooled in the reactor vessel a minimum of 175 hours0.00203 days <br />0.0486 hours <br />2.893519e-4 weeks <br />6.65875e-5 months <br /> prior to discharge. The provisions of Specifications'3.0.3 and 3.0.4 are not applicable.

3.8.15 Loads in excess of 2000 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 are not applicable.

58a

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 4 incorporating built-in interlocks and safety features, provide assur-ante that no 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 per-mits maintenance on the instrumentation. Continuous monitoring of radi-ation levels and neutron flux provides immediate indication of an unsafe condition. The decay heat removal pump is used to maintain 'a uniform boron concentration.(1) The shutdown margin indicated in Specificati n

~

3.8.4 will keep the core subcritical, even with all control rods with-drawn from the core.(2) The boron concentration will be maintained.

above 1800 ppm. Although this concentration is sufficient to maintain the core keff 3:0.99 if all the control rods were removed from the core,

.only a few control rods wi.11 be removed at any one time during fuel .

-shuffling and replacement.- The keff 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 condition detected from the. main cor; trol board .

indicators during fuel movement. _ , _,_

• 0n a one time basis during Cycle 3 operation, a 25-ton NAC-1 cask shall be allowed to be carried by the Auxiliary Building crane for the purpose

of r'emoving the fuel storage pool irradiated B4C test rods. The cask

- handling shall be in accordance with restriction set forth in AP&L letter

~

dated September 25, 1978.  :

O e

e e

y 6

e 4

59

3.10 SECONDARY SYSTEM ACTIVITY Applicability ,

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 'radioioiine activity in the secondary coolant shall not exceed 0.17 uCi/gm. W)'.h the secondary coolant ac-tivity in excess of 0.17 uti/gm I-131, be in at least Hot Standby with-in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Bases -

For the purpose of determining a maximum allowable secondary coolant activity, the activi.ty contained in the mass released following the rupture of a ste'am generator tube, a steam line 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 atmosphere 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 the_se gases which would:be, released.

The dose analysis performed to determine the maximum allowable reactor coolant activity assuming the maximum allowable primary to secondary leakage of 1 gpm as given in the Bases for Specification 3.1.4.1 indi-cated that the controlling accide'nt 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 va; /es, 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 accidgnt 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.

66 .

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

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 from such leak tests when the source contains 100 uCi 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 submitted 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 shall be maintained current at all times.

66b

r 3.15 FUEL HANDLING AREA VENTILATION D iter 4

-Applicability ,

)-

l' ,

Applies to the 'perability o of the fuel handling area ventilation system.

! Objective.

To ensure that the fuel handling area ventilation sys. tem will perform l

within acceptable levels of effi_ciency and reliability.

Specification

~

3.15.1 The fuel handling area ventilation system shall be in operation whenever irradiated fuel handling operations are in progress in i

the fuel handling area of the auxiliary building and shall-have

the following performance cap Qilities:

a. ' The'results of th~e in place col.d D0PTand halog'enated hydro-

- carbon te'sts at design flows (+ 10%) on HEPA filters and

charcoal adsorber banks shall show 99% D0P removal-and j 99% halogenated hydrocarbon removal,

~

b. - The~ results of -laboratory carbon sample 'a~nalysis shall show

90% radioactive methyl iodide removal at a velocity within j l'ide concentration,'20%

d 70% R.ofH..sy' andstem 125F.design, 0.05 to 0.15 mg/m

- c. Fans, shall b,e shown .to operate within 1,10% design f, low.

~ ~

i d. The pressure dr'op across the combined HEPA filters and

. . charcoal adsorber banks shall be less than'6 inches of

water at system design flow rate (1'10%).

e. Air. distribution s' hall be uniform within + 20% across HEPA

! filters and charcoal adsorbers when testeil initially and

after.any maintenance or te' sting that could affect the air

' distribution within the fuel handling area ventilation system. .

3.15.2 If the requirement's of Specification 3.15.1 cannot be met, irra-diated fuel movement shall not be started (any irradiated fuel.

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 aM1-

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 redundant and only one is required to be operating. The filter train consists of. a prefilter, a HEPA filter 'and a charcoal adsorber in

, series. # ,

. 66g v- - - y-~ n sa.- n-- W -

~ r - a evn u - - -- < - , - -,m r-- w-,~-e n-- , a v eram- v - ,c-mse-4<

3'.17 FIRE SUPPRESSION. WATER SYSTEM Applicability This specification' applies to the portions of the fi're suppre'sion s water sys, tem ne:essary to provide fire protection to safety related equipment.

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

Specification

"- 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 f. ire suppression header, and witti 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.

~

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 procedures to be used to provide for the loss of redundancy in this system. .

3.17.3 With the fire suppression water system inoperable:

a. Establish a backup fire suppression water system ~within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and . .

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

c. If "a." above cannot be fulfilled, place the reactor in Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

Bases The fire pumps supply the only source of water for all fire suppression systems utilizing water. Each pump is individually 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 system becomes. inoperable, 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 co'rrective measures to provide. adequate fire suppres-sion capability for the continued operation of the plant.

66m

3.18 FIRE-SUPPRESSION SPRINKLER SYSTEM Applicability

~

This specification applies to the fol' sing fire suppression ' 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.

t -

d. Cable spreat'ing room.

e. Each of the two didsel generator fuel vaults,

f. Hallway El 372 (Zone 98-J).

g. Condensate deminerali:er area.'*

~

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

Specification 3.18.1 The above listed sprinkler systems shall be operable at all, times.

3.18.2 k'ith one or more of the above-listed spr nkler systems ~ inoperable, i

establish a continuous fire watch (or operable smoke and/or heat detection squipment 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 ney.t 30 days out'ining 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 adequate 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.

66n

3.19 CONTROL ROOM AND AUXILIARY CONTROL ROOM HALON SYSTEMS Applicability This specification applies to the Halon systems utilized as t'he fire suppression system for the control room and auxiliary control room.

Objective To assure that fire suppression is available to the safety related erpipment 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 to Specification 6.12.3.2(b) within 30 days outlining

- the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to operable status. The provisions of Specifications 3.0.3 and 3.0.4 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.

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

e 9

e i 660

i 3.20 FIRE HOSE STATIONS.

Applicability 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 safety related equipment. .

Specification 3.20.1 All fire hose stations protecting areas cortaining safety related equipment shall be operable whenever-the equipIment in these areas is 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 one hour. The provisions 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 equipmen; is restcced to service.

-

D e

66p

.t. .

3.21 PENETRATION FIRE BARRIERS Applicability This specification applies to penetration fire barriers relieh on for restriction of fire damage such that safety related equipment'in 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.

3.21.2 With one or more of the required penetration fire barriers.not intact, establish a continuous fire watch (or operable smoke .

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 penetrati fire barriers ensures that fires will be confined or. adequately reta ~d from spreading to adjacent

' portions of the facility.1This dec gn fec.ure minimizes the possibilit of a single fire rapidly involving severe . ;.eas of the facility prior to detection and extinguishment. The penet ation fire barriers are a passive-element in the facility fire protection program and are subject to periodic inspections.

. During periods of time when the barriers are not functional, a continuous.

fire watch or operable detection equipment is required to be maintained

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

e J

66q

3.22 REACTOR BUILDING PURGE FILTRATION SYSTEM

-Applicability This specification applies to the operability of the reactor 6uilding

> purge filtration system.

Objective 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 follcwing performance capabilities.:

a. The results of the in place cold DOP and halogenated hydro-carbon tests at design flows (+ 10%) on HEPA filters and' charcoal adsorber banks shall 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

+ 20% of system design, 0.05 to 0.15 mg/m inlet methyl io-dide concentration, 70% R. H. and' 125F.

c. Fans shall be shown to operate within + 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 (+ 10%).

e. Air distribution shall be uniform within + 20% across HEPA.

filters and charcoal adsorbers when teste3 initially and after any maintenance or testing tlat could affect the air distribution within the reactor building purge filtration system.

3.22.2 If the reqt irements of Specification 3.22.1 cannot be met, either:

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 The reactor building purge filtration system is designed to filter the're-actor building atmosphere during normal operations for each of personnel entry.into the reactor building. This specification is intended to re-quire the system operable during fuel handling operations, if the system 66r

.