ML20041E955

Amend 29 to License NPF-6,changing Listed Tech Specs

ML20041E955
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/04/1982
From:	Clark R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20041E956	List: ML20041E955 ML20041E957 ML20041E959
References
NUDOCS 8203160011
Download: ML20041E955 (38)

Text

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UNITED STATES y,

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E WASHINGTON, D C. 20555

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l ARKANSAS POWER AND LIGHT COMPANY

'l DOCKET NO. 50-368 ARKANSAS NUCLEAR ONE, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. E 3

- License No. NPF-6 1.

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

A.

The applications' for amendment by Arkansas Power and Light Company (the licensee) dated September 22, 1981; October 8, 1981;-July 24, 1979 as supplemented August 1, 1980 and May 6, 1981; October 31',

1980 and March 5, 1981 as supplemented May 6, 1981, comply 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 f acility will operate in confermity with the applications, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authori:ed by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

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

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

a B203160011 820304

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-PDR ADOCK 05000368 P

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b 2-i 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, by d letion of license condition 2.C.3.'[, and by amending i

paragraph 2.C.(2) of Facility Operating License No. NPF-6 to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 29, are hereby incorporated in the license. The licensee shall operate the f acility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION 1

l Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: March 4, 1982 i

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1 ATTACHMENT TO LICENSE AMENDMENT NO.2 9 FACILITY OPERATING LICENSE NO. NPF-6 DOCKET NO. 50-368 Replace the following pages of the Appendix "A" and "B" Technical The revised pages are. identified Specifications with the enclosed pages.

by Amendment number and contain vertical lines. indicating the area of Corresponding overleaf pages are provided to maintain docurent change.

completeness.

Paces - Accendix A V

VIII X

3/4 2-14 3/4 3-20 3/4 4-1 3/4 4-2a 3/4 6-14

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3/4 6-15 3/4 9-9 3/4 9-9a B 3/4 3-3 8 3/4 4-1 i

3 3/4 9-2 6-2 6-3 6-5 6-21 4

Paces Accendix B 5-10 t

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INDEX LIttITING C0flDITIONS FOR OPERATION AND SURVEILLAtlCE REQUIREMEtlTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION..............

3/4 4-1 3/4s4.2 SAFETY VALVES - SHUTD0sti..................................

3/4 4-3 3/4.4.3 SAFETY VALVES - OPERATING.................................

3/4 4-4 i

3/4.4.4 PRESSURIZER...............................................

3/4 4-5 3/4.4.5 STEAM GENERATORS...........................................'3/4 4-6 3/4.4.6 REACTOR COOLANT SYSTEtt LEAKAGE i'

Lea ka ge Detec ti o n Sys t ems................................. 3/4 4-13 Reactor Coolant Sys tem Lea kage............................

3/4 4-14 3/4.4.7 C H Ef t ! S T R Y.................................................

3 / 4 4 - 1 5 3/4.4.8 SPECIFIC ACTIVITY.........................................

3/4 4-18 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System....................................

3/4 4-22 Pressurizer...............................................

3/4 4-25 1

3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2 and 3 Components.....................

3/4 4-26 il

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3/4.5 EMERGEtlCY CORE COOLING SYSTEMS (ECCS) 3 / 4. 5.1 S AFETY IflJECTI0ri TAN KS....................................

3/ 4 5-1 t

3/4.5.2 ECCS SUBSYSTEMS - T

• 300*F............................

3/4 5-3 avg -

3/4.5.3 ECCS SUBSYSTEMS - T,yg <,300*F............................

3/4 5-6 t

3/4.5.4 REFUELIllG WATER TANK......................................

3/4 5-7 1

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i ARKAt!SAS - UtlIT 2 V

Amendment No. 1o i ;

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE-REQUIREMENTS

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SECTION PAGE 3/4.6 CONTa":"CNT SYSTEMS 3/4. 6.1 MinARY CONTAINMENT Containment Integrity..............................

3/46-1 Co n ta i nme n t L ea ka g e...................'............. 3/4 6-2 Co n ta i nmen t Ai r Loc ks.............................. 3/4 6-4 Internal Pressure, Air Temperature and Relative Hu mi d i ty.........................

3/4 6-6 Containment Structural Integrity.................. 3/4 6-8 Containment Ventilation System.....................

3/4 6-9a 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS C on ta i nmen t S p ray Sy s tem........................... 3/4 6-10 Sodium Hydroxide Addition System...................

3/4 6-12

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Con ta inment Cooling System......................... 3/4 6-14 3/4.6.3 CONTAINMENT ISOLATION VALVES.......................

3/4 6-16 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers.................................

3/4 6-22 El ectric Hydrogen Recombiners - h[..................

3/4 6-23 Conta inment Reci rculation System...................

3/4 6-24 Z

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t ARKANSAS - UNIT 2 VI l

INDEX 4

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

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j SECTION PAGE

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3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE l

Safety Valves...............................~..............

3/4 7-1 Eme rg ency Fe edwa t er Sys t em................................ 3 /4 7-5 Condensate Storage Tank...................................

3/4 7-7 A c t i v i ty.............................. ~....................

3 / 4 7 - 8 Mai n Steam Isolation Val ves...............................

3/4 7-10 l

Seco ndary Wa ter Ch emi s try.................................

3/ 4 7-11 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION...........

3/4 7-14 3/4.7.3 S ERV I C E WAT ER S YST EM......................................

3/ 4 7-15 3/4.7.4 EMERGEN CY COOLING POND....................................

3/ 4 7-16 i

3/4.7.5 FLOO D PROTECT I ON..........................................

3/4 7-16a 3/4.7.6 CONTROL ROOM EMERGENCY AIR CONDITIONING AND AIR F I LTRAT I O N SY ST EM.......................................

3 / 4 7 -17 3/4.7.8 HYDRAULIC SHOCK SUPPRESSORS...............................

3/4 7-22 3/4.7.9 S EALED SOURCE CONTAMINATION...............................

3/4 7-27 3/4.7.10 FIRE SUPPRESSION SYSTEMS Fi re Suppressio n Wa ter Sys tem.............................

3/4 7-29 Spray and/or Spri nkl er Systems............................. 3/4 7-33 Fi re Ho s e S ta ti o ns........................................

3/ 4 7-3 5 3/4.7.11 PENETRATION FIRE BARRIERS.................................

3/4 7-37 i

3/4.7.12 S PENT FUEL P00L STRUCTURAL INTEGRITY......................

3/4 7-38 3/4.8 ELECTRICAL. POWER SYSTEMS l

3 / 4. 8.1 A.C. SOURCES O p e r a t i n g.......................................'........... 3 / 4 8 - 1 Shu$down.............................................;....

3/4 8-5 1

ARKANSAS - UNIT '2 VII i-i l

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS PAGE SECTION t

3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEM 3 A.C. Di s tri bu tion - Operati ng............................. 3/4 8-6 A.C. Di s tri bution - Shu tdown..............................

3/4 8-7 D.C. Distribution - Operating...............'."............. 3/4 8-8 D.C. Di s tri butio n - Shu tdown..............................

3/4 8-10 Containment Penetration Conductor Overcurrent Pro tec ti o n Dev i c es...................................... 3/ 4 8-11 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.......................................

3/4 9-1 3/4.9.2 INSTRUMENTATION...........................................

3/4 9-2 3/4.9.3 DECAY TIME................................................

3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.........................

3/4 9-4 3/4.9.5 COMMUNICATIONS............................................

3/4 9-6 3/4.9.6 REFUELING MACHINE OPERABILITY.............................

3/4 9-7 3/4.9.7 CRANE TRAVEL - SPENT FUEL POOL BUILDIRG...................

3/4.9-8 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION.................. 3/4 9-9 l

3/4.9.9 WATER L EVEL - REACTOR V ESS EL..............................

3/4 9-10 3/4.9.10 S PENT FUEL POOL WATER L EVEL...............................

3/4 9-11~

3/4.9.11 FUEL HANDLING AREA VENTILATION SYSTEM.....................

3/4 9/12 3/4.10 SPECIAL TEST EXCEPTIONS

,ii 3/4.10.1 SHUTDOWN MARGIN.........................................

3/4 10-1 j

3/4.10.2 GROUP HEIGHT, INSERTION AND PCWER DISTRIBUTION LIMITS.....

3/4 10-2 3/4.10.3 REACTOR COOLANT LOOPS.....................................

3/4 10-3 3/4.10.4 CENTER CEA MISALIGNMENT...................................

3/4 10-4 3/4.10.5 MINIMUM TEMPERATURE FOR CRITICALITY.......................

3/4 10-5 II ARKANSAS - UNIT 2 VIII Amendment No'. 2 9 f-5 I

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INDEX 1

BASES PAGE SECTION 3/4.0 APPLICABILITY............................................ B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL.....*.................................

B 3/4 1-1 i

3/4.1.2 B O R A T I O N S Y S T EM S...................................... B 3/4 1-2 3/ 4.1. 3 MOVABLE CONTROL ASSEMBLIES............................ B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE......................................

B 3/4 2-1 3/4.2.2 RADIAL PEAKING FACTORS................................

B 3/4 2-2 3/4.2.3 AZIMUTHAL POWER TILT..................................

B 3/4 2-2 3/4.2.4 DNBR MARGIN........................................... B 3/4 2-3 3/4.2.5 RCS FLOW RATE.......................-................. B 3/4 2-4 i

3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE..................

B 3/4 2-4 5

3/4.2.7 AXIAL SHAPE INDEX....................................,.

B 3/4 2-4 3/4.2.8 PRESSURIZER PRESSURE................................. B 3/4 2-4 l3/4.3 INSTRUMENTATION 1

3/ 4. 3.1 PROTECTIVE INSTRUMENTATION............................

B 3/4 3 1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION.............

B 3/4 3-1 1

3/4.3.3 MONITORING INSTRUMENTATION............................ B 3/4 3-1 3/4.3.4 TURBINE OVERSPEED PROTECTION.......................... B 3/4 3-1 s

ARKANSAS - UNIT 2 IX Amendment No.24 i

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INDEX BASES PAGE SECTION 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION.............. B 3/4 3/4.4.2 and 3/4.4.3 SAFETY VALVES................................... B 3/4 4-1 3/4.4.4 PR ES SU R I ZER................................................ B 3 / 4 4-3/4.4.5 S T EAM G EN ERATORS........................................... B 3/ 4 4-3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................. B 3/4 4-3 i

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3/4.4.7 CHEMISTRY.................................................. B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY...........................................B 3/4'4-4 3/4.4.9 PRESSURE / TEMPERATURE LIMITS................................ B 3/4 4-5 3/4.4.10 STRUCTURAL INTEGR ITY....................................... B 3/

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/ 4. 5.1 SAFETY INJECTION TANKS..................................... B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBS YSTEMS................................. B 3/4 5-1 3/4.5.4 REFUELING WATER TANK (RWT)................................. B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT........................................ B 3/4 6-1 1

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEh

....................... B 3/4 6-3 l

3/4.6.3 CONTAINMENT ISOLATION V ALV ES................. '.............. B 3/ 4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTROL.................................... B 3/4 6-4 l

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X Amendment No. 2 9 ARKANSAS - UNIT 2 e

f J.

POWER DISTRIBUTION LIMITS AXIAL SHAPE INDEX LIMITING CONDITION FOR OPERATION 3.2.7 The core average AXIAL SHAPE INDEX (ASI) shall be maintained.within the following limits:

a.

COLSS OPERABLE

-0.28 < ASI < + 0.28 b.

COLSS OUT OF SERVICE (CPC)

-0.20 < ASI < +0.20 1

APPLICABILITY:

MODE 1 above 20% of RATED THERMAL POWER

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ACTION:

With the core average AXIAL SHAPE INDEX (ASI) exceeding its limit, restore the ASI to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to o

less than 20% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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i SURVEILLANCE REQUIREMENTS r

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4.2.6 The core average AXIAL SHAPE INDEX shall be determined to be within its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> using the COLSS or any OPERABLE Core Protection Calculator channel.

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• See Special Test Exception 3.10.2.

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ARKANSAS - UNIT 2 3/4 2,13 Amendment No. 24' l

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POWER DISTRIBUTION LIMITS 4

PRESSURIZER PRESSURE LIMITING CONDITION FOR OPERATION 3.2.8 The average pressurizer pressure shall be maintained between 2225 psia and 2275 psia.

APRLICABILITY:

MODE 1 ACTION:

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With the average' pressurizer pressure exceeding its limits, restore the pressure to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to l

less than 57. of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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o SURVEILLANCE REQUIREMENTS 4.2.6 The average pressurizer pressure shall be determined to be within its liinit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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P TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES

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INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS I

1.

Manual i a.

SIAS Safety Injection No*$ Applicable b.

CSAS r

Containment Spray Not Applicable c.

CIAS L

F i

Containment Isolation Not Applicable f.

d.

MSIS Main Steam Isolation Not Applicable l

e.

CCAS l

Containment Cooling Not Applicable f.

RA5 Containment Sump Recirculation Not Applicable t

g.

EFAS Train A Not Applicable Train B Not Applicable

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1 2.

Pressurizer Pressure-Low a.

Safety Injection 1)

High Pressure Safety Injection 1 30*

2)

Low Pressure Safety Injection 1 35*

3.

Containment Pressure-Hich_.

a.

' Safety Injection i

1)

High Pressure Safety Injection-1 31.6*

2)

Low Pressure Safety Injection 1 51.6*

b.

Containment Isolation 1 52.l*/37.l**

l c.

Containment Cooling

< 43.1*/28.1**

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ARKANSAS - UNIT 2 3/4 3-19 f

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TABLE 3.3-5 (Continued)

ENGINEEREDSAFETYFEATURESRESPdNSETINES t

j INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS i

4.

Containment Pressure--High-High I

a.

Containment Spray 1 42.l*/27.l**

l 5.

Steam Generator Pressure-Low 1

9-3 h.

Main Steam Isolation b.

Feedwater Isolation 1 36.4*/21.4**

6.

Refueling Water Tank-Low a.

Containment Sump Valve Open 1 145.0 l

7.

Steam Generator Level-Low a.

Emergency Feedwater - Train A 1 97.4 l

b.

Emergency Feedwater - Train B 1 112.4*/97.4**

8.

Steam Generator e.P-High Coincident With Steam Generator Level-Low a.

Emergency Feedwater - Train A

,1 97.4 l

b.

Emergency Feedwater - Train B 1 112.4*/97.4**

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.i TABLE NOTATION Diesel generator starting and. sequence loading delays included.

Diesel generator starting delays not included, sequence loadin: delays l

included..0ffsite power available.

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i ARKANSAS - UNIT 2 3/4 3-20

~ Amendment No. 2 9 j.

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r 3/4.4 REACTOR COOLANT SYSTEM 3/4.%.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION

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STARTUP AND POWER OPERATION l

LIttITING CONDITION FOR OPERATION Both reactor coolant loops and both reactor coblant pumps in each loop i

3. 4.1.1 shall be in operation.

APPLICABILITY: MODES 1 and 2.*

ACTION':

With less than the above required reactor coolant pumps in operation, be in at l

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 />.

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SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least cnce per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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ARKANSAS - UNIT 2 3/4 4-1 Amendment No. 24, P3 6

REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 a.

The reactor coolant loops listed below shall be operable:

1.

Reactor Coolant Loop (A) and at least,o,ne associated reactor coolant. pump.

2.

Reactor Coolant Loop (B)-and at least one associated reactor coolant pump.

b.

At least one of the above Reactor Coolant Loops shall be in operation.*

APPLICABILITY: MODE 3.

ACTION:

With less than the above required reactor coolant loops operable, a.

restore the required loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

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

SURVEILLANCE REQUIREMENTS

4. 4.1. 2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct' breaker alignments and indicated power availability.

4. 4.1. 2. 2 At least one cooling loop shall be verified to be in ooeration and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• All reactor coolant pumps may be de-energi:ed for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the reactor coolant syste* Soron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

ARKANSAS - UNIT 2 3/4 4-2 Amendment No. 24, 29

REACTOR COOLANT SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION t

At least two of the coolant loops liste,d below shall be OPERABLE:

3.4.1.3 a.

1.

Reactor Cool' ant Loop (A) and it's' associated steam generator and at least or.e associated reactor coolant pump.

2.

Reactor Coolant Loop (B) and its associated steam generator and at least one associated reactor coolant pump.

3.

Shutdown Cooling Loop (A)#,

4 Shutdown Cooling Loop (B)#.

b.

At least one of the above coolant loops shall be in operation.*

APPLICABILITY: Modes 4 and 5.

ACTION:

a.

With less than the above required coolant loops OPERABLE, immediately initiate corrective action to return the required i

coolant loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

b.

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

SURVEILLANCE REQUIREMENTS 7

4. 4.1. 3.1 The required shutdown cooling loop (s) shall be determined OPERABLE per Specification 4.0.5.

4. 4.1. 3. 2 The required reactor coolant pump (s), if not.in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments I

and indicated power availability.

I 4.4.1.3.3 The requi' red steam generator (s) shall be determined OPERAELE by verifying the secondary side water level to be > 235 indicated level at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l 4.4.1.3.4 At least one coolant loop shall be verified to be in operation and ill circulating reactor coolant at le.ast once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• All reactor coolant pumps and decay heat removal pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature.

1. The normal or emergency power source may be inoperable in Mode 5.

ARKANSAS - UNIT 2 3/4 4-2a Amendment No. 27, 23

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CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) l t

At least once per 18. months, during shutdedd, by:

d.

Verifying that each automatic valve in the flow path 1.

actuates to its correct position on a CSAS test signal.

2.

Verifying that each sodium hydroxide addition pump starts automatically on a CSAS test signal.

At least once per 5 years by verifying the flow rate through i

e.

each component and pipe section in each sodium hydroxide injec-tion path from the tank to the containment spray pump discharge piping to be at least 14 gpm.

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ARKANSAS-UNIT 2 3/4 6-13 1

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CONTAINMENT SYSTEMS.

CONTAINMENT COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.3 Two independent containment cooling groups shall be OPERABLE with at least one operational cooling unit in,pach group.

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

ACTION:

With one group of the above required containment cooling units a.

inoperable and both containment spray systems OPERABLE, restore the inoperable group of cooling units to OPERABLE status within 7 da,s 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 />.

4 b.

With two groups of the above required containment cooling units inoperable end both containment spray systems OPERABLE, restore at least one group of cooling units to OPERABLE status within I

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 />.

Restore both above required groups of cooling units to OPERABLE status within 7 days of initial loss 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 wichin the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

I c.

With one group of the above required contcinment cooling units inoperable and one containment spray system inoperable, restore the inoperable spray system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 />. Restore the j

inoperable group of containment cooling units 1, OPERABLE status i

within 7 days of initial loss 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 SH'JTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

l F

ARKANSAS - UNIT 2 3/4 6-14 Amendment No.J$, 29 l

l l

l

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.2.3 Each containment cooling group shall be demonstrated OPERABLE:

4 a.

At least once per 14 days by:

1.

Verifying a service water flow rate of > 1450 gpm to each group of cooling units; each unit within the group having i

an operable fan, or by verifying a service water flow rate of 2 1250 gpm to one unit within the group; that unit having an operable fan.

2.

Chlorinating the service water during the surveillance in 4.6.2.3.a.1 above, whenever service water temperature is between 60 F and 80*F.

b.

At least once per 31 days by:

1.

Starting (unless already operating) each operational cooling unit from the control room.

2.

Verifying that each operational cooling unit operates for at least 15 minutes.

At least once per 18 months by verifying that each cooling unit c.

starts automatically on a CCAS test signal.

ARKANSAS - UNIT 2 3/4 6-15 Amendment No. If, 76, 29 I

f i

l

REFUELING OPET!ATIONS SHUTDOWN COOLING AND' COOLANT CIRCULATION SHUTDOWN COOLING - ONE LOOP LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one shutdown cooling loop shall be irr' operation.

l APPLICABILITY: MODE 6.

ACTION:

With less than one shutdown cooling loop in operation, except a.

as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all t

containment penetrations providing direct access from the 3

containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

The shutdown cooling loop may be removed frcm operation for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period during the perfomance of CORE ALTERATIONS.

The provisions of Specification 3.0.3' are not' applicable.

c.

f l

SURVEILLANCE REQUIREMENTS J. 9. S.1 A shutdown cooling 1c00 shall be determined to be in operation i

and circulating reactor coolant at a flow rate of > 3000 gpm at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

i ARKANSAS - UNIT 2 3/4 9-9 Amendment No. 2 9 j

REFUE' LING OPERATIONS ____

SHUTDOWN COOLING - TWO LOOPS I

LIMITING CONDITION FOR OPERATION 3.9 8.2 Two independent shutdown cooling loops shall be OPERABLE.*

APPLICABILITY: MODE 6 when the water level above the top of the irradiated fuel assemblies seated within the reactor pressure vessel is less than 23 feet.

ACTION:

With less than the required shutdown cooling loops OPERABLE, a.

immediately initiate corrective action to return the loops to OPERABLE status as soon as possible.

The provisions of Specification 3.0.3 are not applicable.

b.

SURVEILLANCE REQUIREMENTS 4.9.8.2 The required shutdown cooling loops shall be determined 0PERABLE per Specification 4.0.5.

The normal or emergency power source may be inoperable for each shutdown i

cooling loop, i

,t t

i 1

l!

l<

l l

ARKANSAS - UNIT 2 3/4 9-9a Amendment No. 29 l

,-. +

4 REFUELING OPERATIONS -

WATER LEVEL - REACTOR VESSEL LIMITING CONDITION FOR OPERATION At least 23 feet of water shall be maintained ov'Ar the top of 3.9.9 irradiated fuel assemblies seated within the reactor pressure vessel.

APPLICABILITY: During movement of fuel assemblies or CEAs within the reactor pressure vessel while in t10DE 6.

ACTION:

With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or CEAs within the pressure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REOUIREMEilTS 4.9.9 The water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to the start of and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter during movement of fuel assemblies or CEAs.

i 4

b I

i ARKANSAS UNIT 2 3/4 9-10 i

INSTRUMENTATION BASES 3/4.3.3.6 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.

This capability is consistent with the recommen-dations of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term l

Recommendations."

3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the chlorine detection system ensures that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," February 1975.

]

3/4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrur.entation ensures that adequate warning capability is available for the prompt detection of fires. This capabil-ity 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 equip-ment aad is an integral element in the overall facility. fire protection program.

In the event that a portion of the fire detection instrumentation is in-operable, except for detectors located in the containment during Modes 1 and 2, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

l 3/4.3.4 TURSINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protec-

' tion instrumentation and the turbine speed control valves are OPERABLE and will l

protect the turbine from excessive overspeed.

Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety related l

components, equipment or structures.

l ARKANSAS - UNIT 2 3 3/4 3-3 Amendment No. 22, 2 9 I

l I

i

4 3/4.4 REACTOR COOLANT SYSTEM BASES l

3/ 4. 4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above the limits specified by TS 3.2.4 during all normal operatiocs and anticipated transients.

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.

In MODES 4 and 5, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.

Thus, if the reactor coolant loops are not OPERABLE, this specification requires two shutdown cooling loops to be OPERABLE.

The operation of one Reactor Coolant Pump or one shutdown cooling pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor The reactivity change rate associated with boron reductions Coolant System.

esill, therefore, be within the capability of. operator recognition and control.

3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves' operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia.

Each safety valve is' designed to relieve 420,000 lbs per hour of saturated steam at the valve setpoint. The relief capacity of a single safety valve is adequate to relieve any over-pressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating shutdown cooling loop, connected to the RCS, provides overpressure relief capability and will prevent RCS over-pressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia.

The combined relief capacity of these valves is sufficient to limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a cceplete loss of turbine generator load while operating at RATED THEPRAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the steam dump valves.

ARKANSAS - UNIT 2 B 3/4 4-1 Amendment No. 24, 00

4 I

REACTOR COOLANT T.': iEM BASES Demonstration of the safety valves' lift.s'ettings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.

3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCC is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The steam bubble also protects the pressurizer code safety valves against water relief.

The ' steam bubble functions to relieve RCS pressure during all design transients.

The requirement that 150 KW of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency ous provides assurance that these heaters can be-energized during 'a loss.-of-offsite power condition to maintain natural' circulation at HOT STANDBY.

i lt3/4.4.5 STEAM GENERATORS I

The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservic~e inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, tiRevision 1.

Inservice inspection of steam generator tubing is essential

! in order-to maintain surveillance of the conditions of the tubes in the

!,eventthatthereisevidenceofmechanicaldamageorprogressivedegra-; d Inservice inspection of steam generator tubing also

!jleadtocorrosion.l;provides a means of characterizing the nature and cause of a iicegradation so that corrective measures can be taken.

i{l; The plant is expected to be operated in a manner such that the

! {i secondary coolant will be maintained within those chemistry limi

! to result in negligible corros-ion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these limits, The i j iocalized corrosion may likely result in stress corrosion cracking.

' extent of cracking during plant operation would be limited by the limitation,of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage =

0.5 GPM per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.

Operating plants have demonstrated that

, primary-to-secondary leakage of 0.5 GPM per steam generator can readily t ' be detected by radiation monitors of steam generator blowdown.

Leakage

. l in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

ARKANSAS - UNIT 2 B 3/4 4-2 Amendment No. 20

3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION _

The limitations on reactivity conditions during REFUELING ensure that:

1) the reactor will remain subcritical during CORE ALTER,ATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limita-tions are consistent with the initial conditions assumed for the boron dilution incident in the accident analyses.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redondant. monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

3/4.9.4 CONTAIN"ENT PENETRATIONS i

The requirements on containment penetration closure and OPERABILITY of the containment purge and exhaust system HEPA filters and charcoal adsorbers ensure that a release of radioactive material within contain-ment will be restricted from leakage to the environment or filtered through the HEPA filters and charcoal adsorbers prior +o discharge to the atmo-sphere.

The OPERABILITY and closure restrictions arc sufficient to restrict radioactive material release from a fuel element rupture based l

upon the lack,of containment pressuri:ation potential while in the REFUELING MODE. Operation of the contair. ment purge and exhaust system l

HEPA filters.and charcoal adsorbers and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.

[

l I

ARKANSAS - UNIT 2 B 3/4 9-i

,0-

REFUELING OPERATIONS

~~

BASES 3/4.9.5 COMMUNICATIONS

. The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS.

3/4.9.6 REFUELING MACHINE OPERABILITY The OPERABILITY requirements for the refueling machine ensure that:

1) the refueling machine will be used for movement of CEAs with fuel assemblies i

and that it has sufficient load capacity to lift a fuel assembly, and 2) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel assembly, CEA and associated handling tool over other fuel assemblies in j

the storage pool ensures that in the event this-load is ' dropped (1) the

' activity release will be limited to that contained in a single fuel assembly, l

and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses.

.3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION i,

The requirement that at least one shutdown cooling loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140 F as required l-during the REFUELING MODE, and (2) sufficient coolant circulation is maintained j

through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.

The requirenent to have two shutdown cooling loops OPERABLE when there is less than 23 feet of water above the core ensures that a single failure of the operating shutdown cooling loop will not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling, thus in the event of a failure of the operating shutdown cooling loop, adequate time is provided to initiate emergency procedures to cool the core.

ARKANSAS - UNIT 2 B 3/4 9-2 Amendme.nt No. 2A, 2 9 i

6.0 ADMINISTRATIVE CONTROLS

~

6.1 RESPONSIBILITY The General Manager shall be responsible fcr overall facility l

6.1.1 operation and shall delegate in writing the succession to this responsi-bility during his absence.

6.2 ORGANIZATION OFFSITE 6.2.l' The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1.

FACILITY. STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:

Each en duty shift shall be composed of at least the minimum a.

shif t crew composition shown in Table 6.2-1.

b.

At least one licensed Operator shall be in the control room when fuel is in the reactor.

At least two licensed Operators shall :e present in the c.

control room during reactor start-up, scheduled reactor snutdown and during recovery from reactor trips.

d.

An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.

All CORE ALTERATIONS shall be directly supervised by either a e.

licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operaticn.

l

.A site Fire Brigade of at least 5 members shall be maintained f.

onsite at all times. The Fire Brigade shall not include 3 members of tne minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency.

ARKANSAS - UNIT 2 6-1 Amendment No. 5

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ARKAflSAS POWER & LIGitT COMPANY' ARKAtlSAS flVCLEAR OfiE 2,

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SENIOR VICE PRESIDENT 5

EllERGY SUPPLY m

SAFETY REVIEW C0!!MITTEE I

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DIRECTOR DIRECTOR ASSISTAllT DIRECTOR DIRECTOR DIRECTOR DIRECTOR.

m GEllERATION FOSSIL VICE PRESIDEll_.

GENERATION PROJECTS TECilNICAL &

ADMINISTRATION m

TEClifl0 LOGY OPERATI0fl5 L AR EllGINEERING MANAGEMEflT ENVIRONMENTAL SERVICES &

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SERVICES PROJECT SUPPORT j

MANAGER GEllERAL PLANT NUCLEAR MAllAGER SAFETY Corporate Responsibility fhr SERVICES C0ft1ITTEE Fire Protection Program y

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• 0flSITE RESPollSillil.lTY FOR FIRE PROTECriori PROGRAM s

(1) Tlic Ilealth Physica Superintentlent rel. ort s t o the Haiiater, Engineering and Technical Support in administrative uut tern anil roint ine licalt h physcis concerns and he t eports to tlic Ceneral Hanager in mat teru of radiological heal.tli, safety and policy.

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(2)

'the Heal th Phys ica Siiperint endent han direct interface with the Corporate llealth Physicist in mat ters of radiological health aiid safet y.

'the Corporate llealth Physicist repo rt s to the Manager, Ndelear Services. lie will lielp formislate Corporate IIcalth Physics Policy and ensure that it is prope rly i mplement'ed.

FIGURE 6.2-2* Fnidt ional Organi zat ion for Plant Operation

TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION #'

APPLICABLE MODE,5, LICENSE CATEGORY 1, 2, 3 & 4 5&6 I

I*

SOL OL 2

1 Non-Licensed 2

1 Shift Technical Advisor i

None Required

• Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS.

1. Shift crew composition may be less than the. minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in trder to' accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

ARKANSAS - UNIT 2 6-4 Amendment No. 20

4 i

ADMINISTRATIVE CONTROLS 6.3 UNIT STAFF QUALIFICATIONS

~6. 3.1.

Each member of the unit staff shall mee't o'r exceed the minimum

~

qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Health Physics Superintendent who shall meet or exceed the qualifications of l

Regulatory Guide 1.8, September 1975, and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and. response and analysis of the plant for transients and accidents.

6.4 TRAINING

6. 4.1.

A retraining and replacement training p'rogram for the unit staff shall be maintained under the direction of the General Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.

6.4.2 A training program for the Fire Brigade shall be maintained undar the direction of the General Manager and shall meet or exceed the require-ments of Section 27 of the NFPA Code - 1975, except for Fire Brigade train-ing sessions which shall be held at least quarterly.

6.5 REVIEW AND AUDIT l

l 6.5.1 PLANT SAFETY COMMITTEE (PSC)

FUNCTION t

6. 5.1.1 The Plant Safety Committee shall function to ' advise the Plant I

'4anager on all matters related to nuclear safety.

'C0" POSITION 6.5.1.2 The Plant Safety Committee shall be composed of the:

ei Chairman: Manager of Special Projects I

!! ember:

Operations : tanager I,

lember:

Maintenance l tanager e

ltember:

Engineering and Technical Support : tanager I

llember:

Administrative Manager Member:

Technical Analysis Superintendent Member:

Plant Analysis Superintendent Member:

Plant Engineering Superintendent Member:

Health Physics Superintendent l

Member:

Nuclear Software Expert

• The General Manager shall designate in writing the Alternate Chairman in the

+ absence of the PSC Chairman.

See page 6-5a ARKANSAS - UNIT 2 6-5 Amendment No. 5, 72,U, 7/, ';9

\\

ADMINISTRATIVE CONTROLS n

7

• If one of the above members of the Plant Safety Consnittee meets the qualification requirements for this position, the requirement to have this member is satisfied. This membership may be filled by two appro-priately qualified individuals who shall ballot with a single combined vote.

Generic qualifications for this membership shalbbe as follows:

One Individual The Nuclear Software Expert shall have as a minimum a Bachelor's degree in Science or Engineering, Nuclear preferred (in*accordance with ANSI N18.1 )'.

In addition, he shall have a ' minimum of four years of technical experience, of which a minimum of two years shall be in Nuclear Engineer-ing and a minimum of two years shall be in Software Engineering. (Soft-ware Engineering is that branch of science and technology which deals with the. design and use of software.

Software Engineering is a discipline directed to the production and modification of computer programs that are correct, efficient, flexible, maintainable, and understandable, in reasonable time spans, and at reasonable costs). The two years of technical experience in Software Engineering may be general software experience not necessarily related to the software of the Core Protection Calculat'or System. One of these two years of experience shall be with certified computer programs.

Two Individuals One of the individuals shall meet'the requirements of the Nuclear Engineering portion of the above.

The second individual shall have a Bachelor of Science degree (digital computer speciality) and meet the Software Engineering requirements of the above.

The membership (the Nuclear Software Expert or the Digital Computer Specialist) shall be knowledgeable of the Core Protection Calculator System with regard to:

The software modules,.their interactions with each other and a.

with the data base.

b.

The, relationship between operator's module inputs and the trip variables.

Thb relationship between sensor input signals and the. trip c.

i; variable.

d.

The design basis of the Core Protection Calculator System.

I e.

The approved software change procedure and documentation require-i-

ments of a, software change.

l' f.

The security of the computer memory.and access procedures j

to the memory, i

i

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ARKANSAS --UNIT 1 6-Sa Amendment No.12 i.

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ADMINISTRATIVE CONTROLS 6.12.2 By no later than December 1,1980, complete and auditable records must be available and maintained at a central location which describe the environ-mental qualification method used for all safety-rel'ated electrical equipment

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in sufficient detail to document the degree of compliance with the DOR Guide-lines or NUREG-0588.

Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.

6.13 HIGH RADIATION AREA 6.13.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area (as defined in 20.202(b)(3)

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of 10 CFR 20) in which the intensity of radiation is 1000 mrem /hr or less shall'be barricaded and conspicuously' posted as a high radiation area and entrance thereto shall be controlled by requiring the issuance of a radiation work permit. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

A radiation monitoring device which continuously indicates the raciation dose rate in the area.

b.

A radiation monitoring device whic'h continuously integrates the

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radiation dose rate in the area and alarms when a present integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledge-able of them.

c.

An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodi.c radiation surveillance at the frequency specified in the radiation I

work permit.

6.13.2 The requirements of 6.13.1, above, shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr.

In addition, locked doors shall be provided to prevent unauthorized entry into such areas and access to these areas shall be maintained under the a'dministrative control of the Shift Supervisor on duty and/or the Health physics Super.intendent.

l ARKANSAS - UNIT 2 6-21 Ordet dafdd OttdBdt 2/1 7980 Amendment No. /), 3g

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,e SENIOR VICE PRESIDENT ENERGY SUPPLY i

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I ASSISTANT VICE PRESIDE!!i vinECTOR l

NUCLEAR OPERATIONS TECHNICAL AT;D El4VIRCNMENTAL SERVICES l

ARKANSAS NUCLEAR CriE MAfiAGER GENERAL MANAGER TECHNICAL ANALYSIS ARKANSAS NUCLEAR Oi4E CHEMISTS EtiGINEERING AND TECHNICAL SUPPORT AND MANAGER SICLOGISTS l

ARKANSAS NUCLEAR CNE

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i ARKANSAS NUCLEAR CNE ARKANSAS NUCLEAR ONE ARKANSAS NUCLEAR CNE CHEMICAL AND RADI0 CHEMISTRY HEALTH PHYSICS

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I SUPERVISOR SUPERVISOR SUPERINTENDENT

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ARKANSAS POWER & LIGHT CO.

ENVIRONMENTAL SURVEILLANCE FIGURE NC.

ARKANSAS NUCLEAR ONE GRGArilIATION CHART 5-1 5-10 AMENCMENT No. /1, 25, 2 9

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