ML20133D249

Proposed Tech Specs Re Inservice Insp Program

ML20133D249
Person / Time
Site:	Hatch
Issue date:	08/01/1985
From:	GEORGIA POWER CO.
To:
Shared Package
ML20133D197	List: ML20133D195 ML20133D249
References
TAC-59334, NUDOCS 8508070271
Download: ML20133D249 (56)

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1 ENCLOSURE 1 NRC DOCKET 50-321 OPERATING LICENSE OPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS The proposed changes to the Technical Specifications (Appendix A to Operating License DPR-57) would be incorporated as follows:

Remove Page Insert Page iv iv viii viii 3.4-1 3.4-1 3.4-2 3.4-2 3.4-5 3.4-5 3.5-1 3.5-1 3.5-2 3.5-2 3.5-3 3.5-3 3.5-5 3.5-5 3.5-6 3.5-6 3.5-7 3.5-7 3.5-8 3.5-8 3.5-10 3.5-10 3.5-12 3.5-12 3.5-13 3.5-13 3.5-14 3.5-14 3.5-15 3.5-15 3.5-17 3.5-17 3.5-18 3.5-18 3.5-21 3.5-21 3.6-9b 3.6-9b

- 3.6-9c 3.6-10 3.6-10 3.6-11 thru 3.6-14 3.6-11 thru 3.6-14 3.6-23 3.6-23 3.6-24 thru 3.6-30 3.6-24 thru 3.6-30 3.7-13 3.7-13 8500070271 050801 PDR ADOCK 05000321 p PDR

" 1 Section Section Pag 3 LIMITING CDNDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY 3.6-1 A. Reactor Coolant Heatup A. Reactor Coolant Heatup 3.6-1 and Cooldown and Cooldown B. Reactor Vessel Temperature B. Reactor Vessel Temperature 3.6-1 and Pressure and Presaire C. Reactor Vessel Head Stud C. Reactor Vessel Head Stud 3.6-2 Tensioning Tensioning D. Idle Recirculation Imop D. Idle Recirculation Loop 3.6-2 Startup Startup E. Recirculation Rimp Start E. Recirculation Rimp Start 3.6-3 F. Reactor Coolant Chemistry F. Reactor Coolant Chemistry 3.6-4 G. Reactor Coolant Leakage G. Reactor Coolant Leakage 3.6-7 H. Safety and Relief Valves H. Safety and Relief Valves 3.6-9 I. Jet Rimps I. Jet Rimps 3.6-9b l J. Recirculation Rimp Speeds J. Recirculation Rimp Speeds 3.6-9b l K. Structural Integrity K. Structural Integrity 3.6-10 L. Shock Suppressors L. Shock Suppressors 3.6-10a 3.7 CDNTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS 3.7-1 A. Primary Containment A. Primary Containment 3.7-1 B. Starxiby Gas Treatment System B. Standby Gas Treatment System 3.7-10 C. Secondary Containment C. Secondary Containment 3.7-12 D. Primary Containment D. Primary Containment 3.7-13 Isolation Valves Isolation Values 3.8 RADIOACTIVE MATERIAIS 4.8 RADIOACTIVE MATERIALS 3.8-1 A. Miscellaneous Radioactive A. Miscellaneous Radioactive 3.8-1 Materials Sources Materials Scurces 3.9 AUXILIARY ELECTRICAL SYSTEMS 4.9 AUXILIARY ELEcr'RICAL SYSTDS 3.9-1 A. Recpirements for Reactor A. Auxiliary Electrical Systems 3.9-1 Startup Ecuipnent iv

f ^

l l LIST OF TABLES i

, ..cluded)

I Table Title Page l

l 4.2-7 Check, Rinctional hst, and Calibration Mininum Freq2ency 3.2-40 l For Naltron Monitoring Instrumentation Which Initiates Control Rod Blocks i 4.2-8 Check, Rinctional Test, and Calibration Mininum Frecuency 3.2-42 l for Radiation Monitoring Systems Which Limit Radioactivity Release i

4.2-9 Check and Calibration Mininum Freglency for Instrumentation 3.2-45 Which Initiates Recirculation Ring Trip 4.2-10 Check, Rinctional Test, and Calibration Mininum Freq1ency 3.2-46 for Instrumentation Which Monitors Leakage into the Drywell 4.2-11 Check and Calibration Mininum Freq1ency for Instrumentation 3.2-48 Which Provides Surveillance Information 4.2-12 Instrumentation Which Initiates the Disconnection of Offsite 3.2-49a Power Sources 4.2-13 Instrumentation Which Initiates Energization by Onsite Power 3.2-49b Sources 3.6.1 Safety Related Shock Suppressors (Srubbers) 3.6-10c 3.7-1 Primary Containment Isolation Valves 3.7-16 3.7-2 Testable Penetrations with Double O-Ring Seals 3.7-21 3.7-3 Testable Penetrations with Testable Bellows 3.7-22 3.7-4 Primary Containment Testable Isolation Valves 3.7-23 3.13-1 Fire Detectors 3.13-2 3.13-2 Fire Hose Stations 3.13-9 6.2.2-1 Mininum Shift Crew Carposition 6-4 6.9.2-1 Special Reporting Req 11rements 6-19 r ,

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Amendment No. 37, 50, 65, 88, viii

1 LIMITING CONDITIONS IOR OPERATION SURVEILLANCE REQUIREMENIS 3.4 STANDBY LIQUID CDtTI'ROL SYSTEM 4.4 STANDBY LIOUID CDNTROL SYSTEM Applicability Applicability The Limiting Conditions for Operation The Surveillance Recnirements apply apply to the operating status of the to the periodic test and examination Standby Licpid Control System. of the Standby Licuid Control System.

Objective Objective The objective of the Limiting The objective of the Surveillance (bnditions for Operation is to assure Recnirements is to verify the the availability of a system with operability of the Standby Licuid the capability to slut down the Control System.

reactor and maintain the stutdown corrlition without the use of control rods.

Specifications Specifications A. Normal System Availability A. Normal Operational Tests During periods when fuel is in 'Ibe operability of the Stardby Licpid the reactor and prior to startup Control System shall be verified by the -

from the Cold Stutdown Condition performance of the following tests:

the standby licuid control system shall be operable except: 1. Monthly Verify the contiruity of the

1. When performing control rod explosive charge in each loop.

drive maintenance, at which time Specification 3.10.E 2. As recuired by Specification 4.6.K shall be met, Each pump loop shall be locally started and functionally tested or, by recirculating demineralized water to the test tank.

2. When operating with an inoperable component, at which 3. Each Operating Cycle l

time Specification 3.4.B shall At least once during each operating be met, cycle:

or, a. Check that the setting of the system relief valve is

3. When the reactor is in the 1325 + 75 psig.

Cold Stutdown Condition and all control rods capable of normal b. Verify that each pump will de-insertion are inserted and the liver 43 gpm against a system recpirements of Specification head of at least 1190 psig.

3.3.A are met.

c. Initiate one of the Standby Licuid Control System loops from the control room after arranging suction from the test tank and purp demineralized water into the reactor 3.4-1 k

r-LIMITING CEDITIONS FOR OPERATION SURVEILLANCE REQUIREMENIS 4.4.A.3 Each Operating Cycle (Contirued)

c. vessel. This test checks the explosive charge, proper c,peration of the associated valves and selected pump operability. The replacement charge to be installed will be selected from a marufactured batch which has been tested.

d. Both loops including both explosive valves should be tested in the course of two operating cycles.

3,4.B Operating with Inoperable Components If one Standby Licpid Control redundant component is inoperable the reactor may remain in operation for a period not to exceed seven (7) days provided the redundant component is operable.

C. Sodium Pentaborate Solution C. Sodium Pentaborate Solution At all times when the Standby The following tests shall be performed Licpid Control System is recnired to verify the availability of the to be operable the following licnid control solution:

conditions shall be met:

1. Volume 1. Volume The volume of the licnid control Check the standby licnid control solution in the licuid control tank volume at least once per day.

tank shall be maintained as recnited in .?igure 3.4-1.

2. Concentration 2. Concentration The concentration of the licuid Check the concentration of the control tank shall be maintained licnid in the standby licnid as recpired in Figure 3.4-1. control tank by chemical analysis:

3.4-2 W_______________ _ _ _ _ _ _ _ . _

BASES FOR LIMITING CONDITIONS FOR OPEP"7 TION AND SURVEILLANCE REQUIREMENIS 3.4.B Operation with Inoperable Components (Contimed) redandant component upstream of the explosive valves may be out of operation should be consistent with the very small probability of failure of both the control rod shatdown capability and the alternate component in the system, together with the fact that mclear system cooldown takes several hours while licpid control solution injection takes about two hcurs. This indicates the considerable time available for testing and restoring the Standby Licuid Control System to an operable corrlition after testing, while reactor operation contimes. Assurance that the system will still fulfill its function claring repairs is obtained by the wrveillance testing recuired by Specifications 4.4.A and 4.6.K.

Each positive displacement p2mp is sized to inject the solution into the reactor in 50 to 125 mimtes, independent of the amcunt of solution in the tank. The slower rate asmres that the boron gets into the reactor considerably cuicker than the cooldown rate. The faster injection rate limit asmres that there is mfficient mixing so that the boron does not recirculate thrcugh the core in uneven concentrations which could possibly cause the mclear power to rise and fall cyclically.

The maxinum solution volume in Figure 3.4-1 is determined by the tank size.

A mininum recpired p2mp flow rate of 41.2 gpm has been determined by the rate recpired using one pimp to inject the maxinum volume of control solution (5150 gallons) within the maxinum allowed time caf 125 mimtes. Using the mininum punp rate of 41.2 gpn and the fastest injection time of 50 mimtes, a mininum cuantity of 2060 gallons of solution having a 20.2 porcent sodium pentaborate concentration is recuired to meet the shutdown tecuirement. For the maxinum expected pump capacity of 43 gpm a mininum volume of 2150 gallons is that volume which co21d be injected in the mininum allowed time of 50 mimtes.

C. Sodium Fentaborate Solution Limiting Conditions for Operation:

The licpid centrol solution is acceptable if the combination of volume and concentratien of the colution is maintained in the region recuired as shown in Figure 3.4-1 atxl the solution temperature is maintained 10oF above the corresponding sattration temperature (Figure 3.4-2) to guard against boron precipitation.

Surveillance Recnirements:

Level indication and an alarm indicate whether the solution volume has changed, which might indicate a possible solution concentration change. The combination of volume and concentration recuired of the solution is mch that shculd evaporation occur from any point within the acceptable region, a low level alarm will anmnciate before the combination of volume and concentration recnirements are unacceptable. The test interval has been established in consideration of these factors. The solution temperature and volume are checked at a high enough f recpency to asmre a high reliability of acceptability of the solution should it ever be recnited. Temperature and licpid level alarms for the syctem are anmnciated in the control room.

3.4-5 l C----_--------- ---- - - - - .

f LIM m NG W ND m ONS FUR OPERATIN SURVEILLANG REQUIRENENIS 3.5 CORE AND CONTAIfMENP 030 LING 4.5 GRE AND GJNTADNENT COOLING SYSTEMS SYSTD4S Applicability Applicability The Limiting Conditions for The Surveillance Req 11rements operation apply to the apply to the core and containment operational status of the core cooling systems when the corres-and containment cooling systems. ponding limiting conditions for operation are in effect.

Objective Obiective The objective of the Limiting 'Ite objective of the Surveillance Conditions for Operation is to Req 11rements is to verify the asaire the operability of the operability of the core and con- i core and containment cooling tainment cooling systems under systems under all conditions all conditions for which this for which this cooling capa- cooling capability is an bility is an essential response essential response to plant to plant abnormalities. abnormalities.

Specifications Specifications A. Core Spray (G) System A. Core Spray (G) System

1. Normal System Availability 1. Normal System Availability

a. The G System shall be operable: CS system testing shall be performed as follows:

(1) Prior to reactor startup from a cold condition, or Item Freolency

a. Sinulated once/ Operating (2) When irradiated fuel is in the Automatic Cycle the reactor vessel and the Actuation reactor presaire is greater than Test atmospheric presaire, except as stated in Specification 3.5.A.2. b. System flow Once/3 months rate:

Each loop shall deliver at least 4625 gpn against a system head corresponding to a reactor presrure of at i

least 113 psig.

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Amendment No. 12, 3.5-1 t

E

(

LIMITING C0tOITIONS FOR OPERATION SURVEILLANCE REQUIRDENIS 3.5.A.2. Operation with Inoperable 4.5.A.2. Surveillance with Inoperable Conponents Coroponents If one G system loop is When it is determined that one inoperable, the reactor may core spray loop is inoperable remain in operation for a at a time when operability is period not to exceed seven reg 2 ired, the diesel generators (7) days providing all active associated with the remaining conponents in the other G operable core spray loop system loop, the RIR system shall be demonstrated to be LPCI mode and the diesel operable innediately.

generators are operable.

3. Stutdown Req 11rements If Specification 3.5.A.l.a or 3.5.A.2 cannot be met the reactor shall be placed in the Cold Stutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. Residlal lleat Removal (RIR) B. ResidJal Heat Removal (RIR)

System (LPCI and Containment System (LPCI and Containment Cooling Mode) Cooling Mode)

1. Normal System Availability 1. Normal Operational Tests RIR system testing shall be performed as follows:

Item Freq1ency

a. The RIR System shall be operable: a. Air test on Once/10 years l drywell headers (1) Prior to reactor startup and nozzles and from a cold condition, or air or water test on torus headers (2) When irradiated fuel is in and nozzles.

the reactor vessel and the reactor presaire is greater than atmospheric except as stated in Specification

3. 5.B . 2.

3.5-2

LIMITING WNDITIONS EOR OPERATION SURVEILLANCE REQUIRENENIS

3. 5.B.1 Normal System Availability (Cont.) 4.5.B.l. Normal Operational 'Ibsts (Cont.)

Item Frecquency

b. One ER loop with two pugs or two b. Sinulated Once/ Operating loops with one pu w per loop shall be Automatic Cycle operable in the shutdown cooling mode Actuation Test when irradiated fuel is in the reactor vessel and the reactor presaire is atmospheric except prior to a reactor startup as stated in Specification

3. 5.B. l. a.

c. 'Ihe reactor shall not be started up c. System flow Once/3 months with the E R system s2pplying rate: Each m R cooling to the fuel pool, pig shall deliver at least

d. During reactor power operation, the 7700 gpm against LPCI system discharge cross-tie a system head valve, Ell-F010, shall be in the corresponding closed position and the associated to a reactor valve motor starter circuit pressure of breaker shall be locked in the at least off position. In addition, an 20 psig.

anrunciator which indicates that the cross-tie valve is not in the fully closed position shall be available in the control room,

e. Both recircx11ation pump discharge d. Both recirculation p2mp valves shall be operable prior to discharge valves shall be reactor startup (or closed if tested for operability permitted elsewhere in these during any outage exceeding specifications) . 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if operability tests have not been performed during the preceding nonth.

2. Operation with Inoperable 2. Surveillance with Inoperable components components

a. One LPCI Rimp Inoperable a. One LPCI Rimp Inoperable If one LPCI pump is inoperable, the When one LPCI is inoperable, the reactor may remain in operation for the diesel generators associated a period not to exceed seven (7) days with the remaining LPCI pigs shall provided that the remaining LPCI p2g s, be demonstrated to be operable both LPCI s1bsystem flow paths, the immediately and daily thereafter, Cbre Spray system, and the associated until the inoperable LPCI pl@ is diesel generators are operable. restored to normal service.

b. One LPCI Subsystem Inoperable b. One LPCI Subsystem Inoperable A LPCI albsystem is considered to be When one LPCI albsystem is inoper-inoperable if (1) both of the LPCI able, the diesel generators pu g s within that system are inoperable associated with the remaining or (2) the active valves in the alb- LPCI subsystem shall be demon- l strated to be operable, inmediately system flow path are inoperable.

Amendment No. 12, 31, 33, 41, 3.5-3

LIMITING C0t0ITIONS FOR OPERATION SURVEILLANCE REQUIRENENTS 3.5.B.3 Shitdown Reglirements If Specification 3.5.B.l.a or 3.5.B.2 cannot be met, tie reactor shall be placed in the Cold Shit-down Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C. M R Service Water System 4.5.C. M R Service Water System

1. Normal System Availability 1. Normal Operational Tests The E R service water system RIR service water system testing shall be operable: shall be performed as follows:

Item Freq2ency

a. Prior to reactor startup from a Rinp Capacity Once/3 Cold Condition, or l Test: months Each W R ser-

) b. when irradiated fuel is in the vice water pump reactor vessel and the reactor shall deliver vessel presaire is greater at least 4000 than atmospheric presaire gpm at a system l except as stated in head of at. least l Specification 3.5.C.2., or 847 feet.

c. when irradiated fuel is in i the reactor vessel and the reactor is depresatrized at i

least one RIR service water loop shall be operable.

2. One Rimp Inoperable i If one RIR service water ging is inoperable the reactor may remain in operation for a period not to exceed thirty (30) days provided all l other active coroponents of both albsystems are operable.

kaendment No.17, 72, 3.5-5 l

LIMITING ENDITIONS EOR OPERATION SURVEILLANCE REQUIREM NFS 3.5.C.3 Two Rinos Inoperable 4.5.C.3. Two Rams Inoperable If two mR service water pinps are When two ER service water punps inoperable, the reactor may remain are inoperable, the diesel gener-in operation for a period not to ators associated with the exceed seven (7) days provided all remaining operable E R service redundant active conponents in both water albsystems shall be demon-of the E R service water albsystems strated to be operable inmed-are operable. iately and daily thereafter for seven (7) days or until the inoperable conponents are returned to normal operation.

4. Stutdown Req 11rements If Specifications 3.5.C cannot be met, the reactor shall be placed in the Cold Stutdown Condition ,

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (

D. High Presaire Coolant Injection GIPCI) D. High Presaire Coolant Injection .

System (HPCI) System (

i l

1. Normal System Availability 1. Normal Operational Tests HPCI system testing shall be performed as follows:

Item Freq2ency

a. The HPCI System shall be a. Sinulated Once/ Operating operable: Altomatic Cycle Actuation (1) Prior to reactor startup Test from a cold condition, or

b. Flow rate at Once/3 months (2) When irradiated fuel is in normal reactor the reactor vessel and the vessel oper-reactor presaire is greater ating presaire than 150 psig, except as and stated in Specification Flow rate at Once/ Operating 3.5.D.2. 150 psig Cycle reactor presaire I

Amendment No. 107, 3.5-6

( l l

LIMITING CI)NDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.5.D.l.b Normal Operational Tests (Contirued) l

%e HPCI punps shall deliver at least 4250 gpn during each flow rate test.

3.5.D.2 Operation with Inoperable li (bmponents If the HPCI system is inoperable, the reactor may remain in operation for a period not to exceed fourteel.

(14) days provided the ADS, CS systen, '

RIR system LPCI node, and RCIC system are operable. 2. Surveillance with Inoperable Components with the alrveillance reg 2irements of Specification 4.5.D.1 not performed When the HPCI system is inoperable, at the recalired freglencies clie to the ADS acblation logic shall be l low reactor steam presaire, reactor denonstrated to be operable imed-startup is permitted and the lately. %e ADS logic shall be l appropriate alrveillance will be demonstrated to be operable daily performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after thereafter until the HPCI system reactor steam presaire is is returned to normi operation.

adeglate to perform the tests.

3. Stutdown Rectlirements If Specification 3.5.D.l. or 3.5.D.2 cannot be met, an orderly stutdown shall be initiated and the reactor vessel presaire shall be recticed to 150 psig or less within 24 hcurs.

E. Reactor Core Isolation Cooling (RCIC) E. Reactor Core Isolation' Cooling System (RCIC) System

1. Normal System Availability 1. Normal Operational Tests RCIC system testing shall be performed as frllows:

Item Frecuency

a. The RCIC system shall be operable a.Sim11ated Once/ Operating with an operable flow path capable Automated Cycle of (automatically) taking alction Actuation from the alppression pool and trans- (and restart *)

ferring the water to the reactor Test presaire vessel:

(1) Prior to reactor startup from a cold condition, or

• Automatic Restart on a Low Water Level Which is Subseglent to a High Level Trip.

Amendment No. 77, 101, 107, 3.5-7

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENfS 3.5.E.1. Normal System Availability (Cont.) 4.5.E.1 Normal Operational 'Ibsts (Cont.)

Item Freq2ency

a. (2) when there is irradiated fuel b. Verifying that Once/Dperating in the reactor vessel and the mction for the Cycle reactor presm re is above 150 RCIC system is psig, except as stated in altamatically trans-Specification 3.5.E.2. ferred from the CST to the alppression pool on a sinulated low CST level or high alppression pool level signal.

Item Freq1ency

c. Flow rate at Once/3 months normal reactor vessel operating presaire and Once/ Operating Flow rate at Cycle 150 psig reactor presaire The RCIC punp shall deliver at least 400 gpm during each flow test.

I

2. Operation with Inoperable l Components l j If the RCIC system is inoperable, the reactor may remain in operation for a period not to exceed seven (7) days if the HPCI system is operable daring such time.

3. Shatdown Reglirements l If Specification 3.5.E.1 or 3.5.E.2 is not met, an orderly shitdown shall be initiated and the reactor shall be depresairized to less than 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Amendment No. 101, 107, 3.5-8

+-

n-LIMITING (IDIDITIONS POR OPERATION SURVEILIANCE REQUIRDENTS 3.5.G Minim 2m Core and Containment 4.5.G Surveillance of Core'and Contain-Cooling Systems Availability ment Cooling Systems During any period when one of When it is determined that one the standby diesel generators is of the standby diesel generators inoperable, contirued reactor is inoperable, the remaining operation is limited to seven (7) diesels shall be demonstrated l days unless operability of the to be operable insnediately diesel generator is restored and daily thereafter.

within this period. During alch seven (7) days all of the com-ponents in the MIR system LPCI mode and contairunent cooling mode shall be operable. If this reg 2irement cannot be met, an orderly shutdown shall be ini-tiated and the reactor shall be in the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Specification 3.9 provides further guidance on electrical system availability.

Any combination of inoperable. '

conponents in the core and con-tainment cooling systems shall not defeat the capability of the remaining operable components to fulfill the core and contain-ment cooling functions.

When irradiated fuel is in the reactor vessel and the reactor is in the Cold Shutdown Condition, both core spray systems and the LPCI and contairunent cooling subsystems of the RIR system may be inoperable provided that the shatdown cooling subsystem of the RIR system is ,

operable in accordance with Specification 3.5.B.l.b and that no work is being done which has the potential for draining the reactor vessel.

H. Maintenance of Filled Discharge H. Maintenence of Filled Discharge Pipes Pipes Whenever the core spray system, The following surveillance LPCI, HPCI, or RCIC are regaired reojirements shall be performed to be operable, the discharge' to assare that the discharge piping from the ging discharge piping of the core spray system, of these systems to the last LPCI, HPCI, and RCIC are filled block valve shall be filled., when renaired:

The anction of the HPCI punps shall be aligned to the 1. Every month prior to the condensate storage tank. testing of the LPCI and core spray systems, the discharge piping of these systems shall be vented Amendment No. 12, 3.5-10 x __ - - - _ - _ _ _ _ _ _ - _ - - _ - _ _ _ _ _ _ _ - _ _ L

LIMITING CDNDITIONS FOR OPERATION SURVEILLANCE REQUIRDENIS 3.5.J Plant Service Water System 4.5.J Plant Service Water Systems

1. Normal Availability 1. 'Ihe automatic pl@ start i functions and altamatic

'Ite reactor shall not be made isolation functions shall critical from the cold slut- be tested once per operating down condition unless the cycle.

Plant Service Water System l (includire 3 plant service water punps and the standby service water pinp) is operable.

2. Inoperable Conponents 2. Inoparable Components

! a. The standby service water a. With the standby service water >

pump may be inoperable for a subsystem inoperable for up to period not to exceed 60 60 days, provide Unit 1 service l days provided that an alternate water cooling to the 1B diesel Unit 1 plant service water E nerator by verifying cooling source to the IB OPERABILITY of an alternate Unit diesel generator is OPERABLE. 1 service water cooling source I within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Otherwise, declare the 1B diesel generator inoperable and take the action req 2 ired by Specification 3.9.B.2.

b. One PSW punp may be inoperable b. When one PSW punp is made or found for a period not to exceed to be inoperable, all diesel gener-30 days provided all diesel ators associated with the operable generators associated with the PSW cogonents shall be demon-operable PSW conponents are strated to be operable immediately operable. and weekly thereafter.

c. One PSW punp and the standby c. When one PSW pump and the standby service water pimp may be service water pimp are made or inoperable for a period not to found to be inopeyable, all diesel exceed 30 days provided generators associated with the all diesel generators associated operable PSW conponents shall be with the operable PSW components demonstrated to be operable are operable. imediately and weekly thereafter.

d. Two PSW pimps or one PSW d. When two PSW pumps or one PSW division may be inoperable for division are made or found to be a period not to exceed 7 days inoperable, the diesel generators l provided the diesel generators associated with the operable PSW associated with the operable components shall be demonstrated PSW conponents are operable. to be operable inmediately and daily thereafter.

Amendment No. 56, 80, 3.5-12

~ __________

LIMITING CONDITIONS EUR OPERATION SURVEILLANCE REQU1HEMENIb _

3.5.J Plant Service Water System 4.5.J Plant Service Water System _s

2. Inoperable Couponents (Cont'd) 2. Inoperable Conponents (Cbnt'd)

e. Two PSW pumps or one PSW e. When two PSW p2mps or one division, and the standby PSW division, and the service water pump may be in- standby service water pamp operable for a period not to are made or found to be exceed 7 days provided the inoperable, the diesel generators l diesel generators associated associated with the operable PSW with the operable PSW com- components shall be demonstrated ponents are operable. to be operable inmediately and daily thereafter.

For each condition above in which When cooling water to diesel the standby service water pump is generator 1B is intertied with inoperable, cooling water to diesel the PSW divisional piping alpply, generator IB shall be intertied operability of the divisional with the PSW divisional piping supply. interlock valves shall be demonstrated.

3. Shutdown Req 1irements If the reglirements of Specifications 3.5.J.1 and 3.5.J.2 cannot be met the reactor shall be placed in the cold shatdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.5.K Fatignent Area Coolers 4.5.K FA11pment Area Coolers l

1. The eglipnent area coolers 1. Each eglipment area cooler serving the Reactor Core is operated in conjunction l Isolation Cooling (RCIC), High with the eglipment served by Presaire Coolant Injection that partiallar cooler; i (HPCI), Core Spray or Residual therefore, the eglipment area  !

Heat Removal (IER) pumps nust coolers are tested at the be operable at all times when same freq2ency as the punps the pamp or pimps served by which they serve.

that specific cooler is considered to be operable.

j 2. When an eg2ipment area cooler is not operable, the punp(s) served by that cooler nust be considered inoperable for Technical Specification purposes.

I w ^"" '**"' ' -*

BASES -FOR LIMITING (DNDITIONS FOR OPERATION AND SURVEILIANCE REQUIREMENTS 3.5 CORE AND CONPADNENT COOLING SYSTENS A. Core Spray'(G) System

1. Normal System Availability Analyses presented in Section 6 of the FSAR and Appendix I of the HNP-2 PSAR demonstrated that the core spray system provides adecpate cooling to the core to dissipate the energy associated with the loss-of-coolant accident and to limit fuel clad temerature to below 2,300oF which asmres that core geometry remains intact and to limit any clad roetal-water reaction to less than one percent. Core spray distribution has been shown in tests of systems similar in design to HNP-1 to exceed the mininum recuirements. In addition, cooling effectiveness has been demonstrated at less than half the rated flow in sinulated fuel assemblies with heater rods to &plicate the decay heat characteristics of irradiated fuel.

The intent of the G system specifications is to prevent operation above atmospheric pressure without all associated ecpipment being operable.

However, & ring operation, certain cogonents may be out of service for the specified allowable repair times. The allowable repair times have been selected using engineering judgement based on experiences and mpported by availability analysis. Assurance of the availability of the remaining systems is increased by demonstrating operability immediately and by regliring selected testing & ring the outage period.

When the reactor vessel pressure is atmospheric, the limiting conditions for operation are less restrictive. At atmospheric presmre, the mininum reg 2irement is for one supply of makeup water to the core. Recuiring two operable ER pmps and one G pq provides reclindancy to ensure makeup water availability.

2. Operation with Inoperable Components Should one core spray loop become inoperable, the diesel generators associated with the remaining operable core spray loop are demonstrated to be operable to ensure their availability shculd the need for core cooling arise. The mrveillance testing recnited by Specifications 4.5.A, 4.5.H, and 4.6.K enm res the availability of the remaining core spray loop. '1he mrveillance testing recnired by Specifications 4.5.B, 4.5.C, 4.5.H, and 4.6.K ensures the availability of the E R system. These provide extensive margin over the operable ecnipment needed for adecpate core cooling. With

&e regard for this margin, the allowable repair time of 7 days was chosen.

B. Resicisal Heat Removal (ER) System (LPCI and Containment Cooling Mode)

1. Normal System Availability The ER system LPCI mode is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant accident. This system is cogletely independent of the core spray system; however, it does function in combination with the core spray system to prevent excessive fuel clad te@erature. The LPCI mode of the E R system and the core spray system provide adecuate cooling for break areas of approximately 0.2 scuare feet up to and including the double-ended recirculation line break without assistance from the high-presmre emergency core cooling systems.

3.5-14 N  ;

BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIRENENIS 3.5.B.l. Normal System Availability (Contimed)

Observation of the stated reglirements for the contaiment cooling mode asaares that the alppression pool and the drywell will be alfficiently cooled, following a loss-of-coolant accident, to prevent primary contain-ment overpresatrization. The containment cooling function of the ER system is permitted only after the core has been reflooded to the two-thirds core height level. This prevents inadvertently diverting water needed for core flooding to the less urgent task of containment cooling.

The two-thirds core height level interlock may be mamally bypassed by a.

keylock switch.

'Ihe intent of the ER system specifications is to prevent operation above atmospheric presaire without all associated eglipnent being operable.

However, daring operation, certain cmponents may be out of service for the specified allowable repair times. '1he allowable repair times have been selected using engineering judgement based on experiences and alpported by availability analysis. Assurance of the availability of the remaining systems is increased by demonstrating operability innediately and by renairing selected testing daring the outage period.

When the reactor vessel presaire is atmospheric, the limiting conditions for operation are less restrictive. At atmospheric pressure, the mininum req 11rement is for one alpply of makeup water to the core.

2. Operation with Inoperable Comonents With one LPCI pinp inoperable or one LPCI albsystem inoperable, adeq1 ate core flooding is asaired. The alrveillance testing reglired by Specifications 4.5.B, 4.5.C, 4.5.H, and 4.6.K enaires the availability of the redandant LPCI punps and LPCI albsystems. The alrveillance testing req 11 red by Specifications 4.5.A, 4.5.H, and 4.6.K enaires the availability of the Core Spray system. In addition, the associated diesel generators are demonstrated to be operable.

Amendment No. 12, 33, 3.5-15 N

. . _ _ _ _- _ = _ . .. . ___ . . _ _ _ _ _ _ . __

BASES FOR LIMITING CDNDITIONS POR OPERATION AND SURVEILIANCE REQUIREMENTS

3. 5.D. 2 Operation with Inoperable Congonents The HPCI system serves as a backup to the RCIC system as a source of feedwater makalp & ring primary system isolation conditions. The ADS serves as a backup to the HPCI system for reactor depresairization for posb11ated transients and accidents. The ADS is checked for operability if the HPCI system is determined to be inoperable. In addition, the airveillance testing required by Specifications 4.5.E, 4.5.H, and 4.6.K enaires the operability of the RCIC system. Considering the reclandant systems, an allowable repair tine of seven (7) days was selected.

E. Reactor Core Isolation Cooling (RCIC) System

1. Normal System Availability The various conditions under which the RCIC system plays an essential role in providing makeup water to the reactor vessel have been identified by evaluating the various plant events over the full range of planned operations. The specifications enmre that the function for which the RCIC system was designed will be available when needed.

Beca1se the low-presaire cooling systems (LPCI and core spray) are capable of providing all the cooling reg.lired for any plant event when mclear system l

presaire is below 150 psig, the RCIC system is not reglired below this presaire. RCIC system design flow (400 gpm) is alfficient to maintain water level above the top of the active fuel for a couplete loss of feedwater flow at i the design power.

'I%o sources of water are available to the RCIC system. Suction is initially l taken from the condensate storage tank and is altamatically transferred to the l mppression pool upon low GT level or high alppression pool level.

l

2. Operation With Inoperable Congonents l

l l Consideration of the availability of the RCIC system reveals that the average l risk associated with failure of the RCIC system to cool the core when reglired is not increased if the RCIC system is inoperable for no longer than seven (7) days, provided that the HPCI system is operable during this period. 'Ihe alrveillance testing reglired by Specifications 4.5.D, 4.5.H, and 4.6.K enaires the operability of the HPCI system.

F. Automatic Depresairization System (ADS)

1. Normal System Availability

This specification enaires the operability of the ADS under all conditions for l which the depresairization of the mclear system is an essential response to Unit abnormalities.

The mclear system presaire relief system provides altomatic mclear system depresairization for small breaks in the mclear system so that the low-presaire coolant injection (LPCI) and the core spray system can operate to protect the fission proc 11ct barrier. Note that this Specification applies only to the altomatic feature of the presaire relief system.

Amendment No. 107, 3.5-17 Y

BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.5.F.1. Normal System Availability (contirued) l Specification 3.6 states the reglirements for the presaire relief function of the valves. It is possible for any mmber of the valves assigned to the ADS to be incapable of performing their ADS functions becalse of instrumentation failures yet be fully capable of performing their presaire relief function. l Because the automatic depresairization system does not provide makalp to the reactor primary vessel, no credit is taken for the steam cooling of the core caised by the system actuation to provide further conservatism to the Core Standby Cooling Systems.

2. Operation with Inoperable Couponents With one ADS valve known to be incapable of altomatic operation six valves remain operable to perform their ADS function. However, since the ECG Ioss of Coolant Accident analysis for small line breaks asalmed that all seven ADS valves were operable, reactor operation with one ADS valve inoperable is only allowed to contirue for seven (7) days provided that the actuation logic for the (remaining) six ADS valves is demonstrated to be operable. The surveillance testing reglired by Specifications 4.5.D, 4.5.H, and 4.6.K enaires the availability of the HPCI system.

3. Mininum Core and Containnent Cooling Systems Availability

'Ihe purpose of this Specification is to asaire that adenlate core cooling eglipnent is available at all times. If, for exanple, one core spray loop were out of service and the diesel which powered the opposite core spray were out of service, only 2 RIR pinps would be available. Specification 3.9 nust also be cona11ted to determine other reglirements for the diesel generators. In addition, refer to definition 1.0.00 for Qinulative Downtime realirements.

' Itis specification establishes conditions for the performance of major maintenance, alch as draining of the alppression poul. The availability of the shutdown cooling albsystem of the RIR system and the IIIR. service water system enaire adegiate alpplies of reector cooling and emergency makeup water when the reactor is in the Cold Stutdown condition. In addition, this specification provides that, should major maintenance be performed, no work will be performed which could lead to draining the water from the reactor vessel.

Amendment No. 12, 21, 27, 3.5-18 Y

BASES FOR LIMITING G)NDITIONS EUR OPERATION AND SURVEILLANCE REQUIREMENTS 3.5.J/4.5.J Plant Service Water System The Plant Service Water (PSW) system consists of two subsystems (divisions) of two pumps each and a separate standby service water pump system for diesel generator 1B. During normal full power operation the two subsystems function as a 3 out of 4 p2mp cross connected system supplying cooling water to the turbine and reactor b2ilding cooling systems. In the event of an accident signal, non-safety-related cooling loads are isolated and the PSW pumps in the two m bsystems supply cooling water to diesel generators lA and 1C, the reactor building cooling system and the control room air conditioners, while the standby service water pum is available to automatically supply cooling water to diesel generator 1B should it be needed. Additionally, diesel 1B has a mamal backup water supply available from the Unit 1 Division 1 or Division 2 PSW mbsystems so that during maintenance on the standby diesel service water pump, either division of the PSW system can mamally be aligned to supply cooling water to the 1B diesel. The two mbsystems and the standby service water pump system are split in the accident mode for greater reliability with one pug in each of the two subsystems aitomatically starting while a start signal from diesel generator 1B initiates standby service water pa@ operation. Only one of the Division 1 PSW p2mps and one of the Division 2 PSW pumps are required for cooling diesel generators lA and 1C, respectively, while the standby service water pump provides adegaate cooling water to diesel generator 1B. In the event that the standby service water pump is inoperable, the HNP-1 Division 1-Division 2 intertie supply piping can be aligned to cool the 1B diesel. In this condition, one PSN pump is capable of mpplying the cooling regairements for the reactor bailding cooling system, the control room air conditioners, and the lA,1B, and 1C diesel generators.

The PSW system can mpply all power generation systems at full load and the diesel generators with redandancy if one PSW p2mp and/or the standby service water pump are inoperable. Hence, a 60-day outage time is justified if the standby service water pump is inoperable since all four PSW pumps are available (divisonal intertie to 1B diesel reglired). In addition, a 30-day outage is justified if one PSW pump is inoperable, or if one PSW pump and the standby service water pump are inoperable (divisional intertie to 1B diesel reg 2 ired). Should two PSW pamps (or one subsystem) become inoperable, or should two PSW pugs (or one mbsystem) and the standby service water pump become inoperable (division intertie to IB diesel req 2 ired) plant operation will probably only contime at less than full power. However, safety-related loads are still adeglately powered for these conditions. Therefore, a 7 day outage time is justified for mch events.

The mrveillance testing regaired by Specification 4.6.K will provide adequate assurance that the PSW system will be operable when reglired.

K. Engineering Safety Features Eq2ipment Area Coolers The egaipment area cooler in each pump compartment is capable of providing adeg2 ate ventilation flow and cooling. Ergineering analyses indicate that the temperature rise in safeguard compartments without adeglate ventilation flow or cooling is such that contin 2ed operation of the safeguard eclipment or associated a2xiliary eglipment cannot be assured.

The mrveillance and testing of the egaipment area coolers in each of their various modes is accomplished during the testing of the egaipment served by these coolers.

The testing is adeg2 ate to assure the operability of the equipment area coolers.

L. References

1. FSAR Section 6, Core Standby Cooling System.

2. HNP-2 PSAR Appendix I, Conformance to NRC Interim Acceptance Criteria for Emergency Core Cooling Systems.

Amendment No. 56, 3.5-21 L

r

' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

b. With the relinf v21ve function and/or the low low set function of more than one of the above I req 11 red reactor coolant system l relief / safety valves inoperable, l be in at least HCTf SHUIDOWN with-  ;

in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in (DLD SIUIDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.6.I Jet Rimps 4.6.I Jet Rimps Whenever the reactor is in the Whenever both recira11ating pu@s ,

Start & Hot Standby or Rin Mode are operating with the reactor in I with both reciro11ating pumps the Start & Hot Standby or Rin operating, all jet pumps shall be Mode, jet pump operability shall operable. If it is determined that be checked daily by verifying that a jet pimp is inoperable, an orderly the following conditions do not stutdown shall be initiated and the occar sim21tanea2 sly.

reactor shall be in the Cold Shit-down Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 1. The two recirculation loops have a flow imbalance of 15%

or more when the pumps are operated at the same speed.

2. The indicated value of core flow rate varies from the value derived from loop flow measurements by more than 10%.

3. The diffuser to lower plerum differential pressure reading on an indivichial jet pimp vary from the mean of all jet pl@ differential presaires by more than 10%.

3.6.J Recira11ation Rimps Speeds - 4.6.J Recirculation Rimp Speeds

1. Core thermal power shall not exceed Recira11ation pump speeds shall be 1% of rated thermal power without recorded at least once per day, forced recira11ation.

2. Operation with a single recirculation pump is permitted for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the recirculation pump is sooner made operable. With one recirculation pl@ not in operation, initiate action wihin 15 mirutes or contirue action to reclice reactor power to or below the limit specified in Figure 3.6-5 within 2 hairs. If the gl@

cannot be made operable or the limit of Fig 2re 3.6-5 cannot be met within the reglired time, the reactor shall be in cold shitdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Amendment No. 27, 31, 42, 103, 3.6-9b k

p; LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l 3.6.J Recirculation Punp Speeds (contimed)

3. Following one pung operation the discharge valve of the low speed punp may not-be opened unless the speed of the faster pung is less than 50% of its rated speed, l

l 3.6-9c l

LIMITING CONDITIONS EDR OPERATION SURVEILLANCE REQUIREMENIS 3.6.K STRUCIURAL INTEGRITY 4.6.K STRUCIURAL INTEGRITY

1. Normal Condition Surveillance Req 1irements for in-service inspection and testing of ASME Code The structural integrity of ASME Class 1, 2, ard 3 (eglivalent) components Code Class 1, 2, and 3 (egliva- shall be applicable as follows:

lent) conponents shall be maintained in accordance with the 1. In-service inspection of ASME Surveillance Req 1irements of Code Class 1, 2, and 3 (eq11 valent)

Specification 4.6.K. components and in-service testing of ASME Code Class 1, 2, and 3

1. Off-Normal Coniitions (eglivalent) punps and valves shall be performed in accordance with

a. With the structural Section XI of the ASME Boiler and integrity of any ASME Code Pressure Vessel Code and @plicable Class 1 couponent not Addenda as realired by 10CFR50, conforming to the above Section 50.55a(g), except where realirements, restore the specific relief has been structural integrity of granted by the Conmission paralant the affected conponent(s) to 10CFR50, Section 50.55a(g) to within its limit or (6) (i) .

isolate the affected component (s) prior to increasing the Reactor 2. Performance of the above in-service Coolant System tenperature inspection and testing activities more than 500F above the shall be in addition to other mininum temperature reglired specified Surveillance by NDT considerations. Realirements.

b. With the structural integrity 3. Nothing in the ASME Boiler and of any ASME Code Class 2 Presaire Vessel Code shall be component (s) not conforming construed to supersede the to the above reglirements, reglirements of any Technical restore the structural Specification.

integrity of the affected couponent(s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 2120F.

c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above reglirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected conponent(s) from service.

3.6-10 k

i These pages have been left blank. l l

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3.6-11 through 3.6-14 l

BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENIS 3.6.K STRUCIURAL INI'EGRITY In-service inspection of ASME Code Class 1, 2, and 3 (ecuivalent) components and in-service testing of ASPE Code Class 1, 2, and 3 (ecuivalent) pumps and valves shall be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Presaire Vessel Code and Addenda as recpired by 10CFR50.55a(g). This objective will maintain the structural integrity of safety-related conponents, punps, and valves which are necessary to safely slut down the plant or mitigate the consecuences of an accident.

3.6-23 k

These pages have been left blank. l 3.6-24 through 3.6-30 l N.

LIMITING CDNDITIONS FOR OPERATIOF SURVEILLANCE RB2UIREMENI'S 4.7.C.l. Surveillance While Integrity Maintained (Cont'd)

b. Secondary containment capability to maintain a mininum 1/4-inch of water vaalum under calm wind (< 5 nph) l conditions with each filter train flow rate not nore than 4000 cfm shall be demonstrated at each refueling outage, prior to refueling.

3.7.C.2 Violation of Secondary 2. Surveillance After Integrity Violated Containment Integrity

a. Without Hatch-Unit 1 secon- After a secondary containment viola-dary containment integrity, tion is determined the standby gas restore Hatch - Unit 1 se- treatment system will be operated condary containment inte- inmediately after the affected zones grity within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or per- are isolated from the remainder of form the following (as appli- the secondary containment. The cable) : ability to maintain the remainder (1) Suspend irradiated fuel of the secondary containment at and/or fuel cask handling 1/4-inch of water vacuum pressure a in the Hatch-Unit 1 se- under calm ((5 nph) wind conditions- l condary containment. shall be confirmed.

(2) Be in at least Hot Stutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and meet the Conditions of 3.7.C.l.a within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.'

b. Without Hatch-Unit 1 secondary containment, refer to the follow-ing Hatch-Unit 2 Technical Specifications, for IID's to be followed for Hatch-Unit 2:

(1) Section 3.6.5.1.

(2) Section 3.9.5.1.

D. Primary Containment Isolation Valves D. E imarv Containment Isolation Valves l

1. Valves P wired to be Operable 1. Surveillance of Operable Valves During rez.ctor power operation, Surveillance of the primary con-all primary containment isolation tainment isolation valves shall be valves listed in Table 3.7-1 and performed as follows:

all reactor coolant system instruc-ment line excess flow check valves a. At least once per operating shall be operable except as stated cycle the operable isolation in Specification 3.7.D.2. valves that are power operated and automatically initiated shall be tested for sinulated automatic initiation and the closure times specified in Table 3.7-1.

Amendment No. 40, 56, 91, 100, 3.7-13

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ENCLOSURE 2 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS In addition to the obvious changes made as a result of amendments issued to sone of the affected Technical Specification pages since the October 3, 1978 submittal, the following are significant changes made to " modernize" the aforementioned submittal:

Page Tech Spec Change Reason for Change iv Table of Contents Page numbers for Editorial 3.6/4.6.I and 3.6/4.6.J change from pg. 3.6-9 and 3.6-9a, respectively to 3.6-9b for both 3.6/4.6.I and 3.6/4.6.J.

3.4-5 3.5.C Add degree sign to LC0 for Editorial sodium pantaborate solu-tion relative to solution temperature maintenance.

3.5-1 4.5.A.l.b Add the words " correspond- Proposed wording addi-ing to a reactor pressure" tion consistent with relative to system head. Hatch 2 Tech Specs.

3.5-2 4.5.B.l.a Test frequency changed Proposed testing fre-from "Once/5 years" quency change consis-to "Once/10 years". tent with requirements of ASME Section XI Code.

3.5-3 4.5.B.l.c Add the words " correspond- Proposed wording addi-ing to a reactor pressure" tion consistent with relative to system head. Hatch 2 Tech Specs.

3.5-3 4.5.B.2.a Add the words " associated Editorial in nature with the remaining LPCI to be consistent with pumps" relative to diesel proposed wording in generators with one LPCI other system Tech Spec pump inoperable. sections, e.g., RHR SW.

3.5-3 4.5.B.2.b Add the words " associated Editorial. Proposed with the remaining LPCI wording similar to that subsystem" relative to proposed in other sys-diesel generators with tem Tech Spec sections, one LPCI subsystem in- RHR SW.

operable.

3.5-5 3.5.C.1.b Word "or" added Editorial

ENCLOSURE 2 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATI0 g Page Tech Spec Change Reason for Change 3.5-5 4.5.C.1 Deleted the words "After Pump test required by pump maintenance and" ASME Section XI Code under " Frequency". and by Hatch plant procedures after pump maintenance. Tech Spec redundancy not neces-sary due to Code re-quirements and plant procedural require-ments.

3.5-13 3.5.K.2 Existing wording con- Plant desires retention cerning equipment area of existing Tech Spec.

cooler (s) inoperability relative to pump (s) served by cooler (s) retained.

3.6-9b 4.6.I, 3.6.J, 4.6.I.2 and 3, 3.6.J.1 and Editorial in nature to and 4.6.J 2, and 4.6.J relocated accommodate new wording from pg. 3.6-10. for 3.6.K/4.6.K on pg.

3.6-10.

3.6-9c 3.6.J 3.6.J.3 relocated from Editorial in nature to pg. 3.6-10. accommodate new wording for 3.6.K/4.6.K on pg.

3.6-10.

3.6-10 4.6.K.1 Word " written" deleted Should emergency Code relative to specific relief ever be neces-relief. sary, verbal relief could suffice to sup-port inspections, etc.

pending receipt of written relief from NRC.

3.6-10 4.6.K Deleted 4.6.K.4 which Submittal of report to concerned reporting NRC in 1980 (i.e., 5 of inservice inspection years after start of results to NRC after commercial operation) 5 years. completed this Tech Spec requirement; thus, 4.6.K.4 proposed previously is no longer required.

'\w 2

ENCLOSURE 2 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS Page Tech Spec Change Reason for Change 3.6-13 4.7.0.1 Deleted references to Awaiting NRC review and

" Specification 4.6.K" approval of proposed previously shown in Appendix J Tech Specs 4.7.D.l.b and 4.7.D.1.d (as discussed in GPC on pgs. 3.6-13 and 3.6-14, June 25, 1985 letter respectively, in October 3, NED-85-483). Possible 1978 submittal. addition of reference to Specification 4.6.K in affected sections at later date after NRC review and approval of previously proposed Appendix J Tech Spec changes.

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i da

1 ENCLOSURE 3 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS Pursuant to 10 CFR 50.59, the Plant Review Board has reviewed the attached proposed amendment to the Pl ant Hatch Unit 1 Technical Specifications and has determined that implementation of the proposed amendment does not constitute an unreviewed safety question.

PROPOSED-CHANGES The proposed Technical Specification changes relative to inservice inspection / inservice testing provide for the following:

1. Deletion of individual surveillance specifications for pumps and valves by adding new wording to Technical Specification sections 3.6.k and 4.6.k which requires inservice testing of ASME Code C1 ass 1, 2, and 3 (equivalent) pumps and valves per the ASME Section XI Code in accordance with 10 CFR 50.55a;

2. Deletion of Technical Specification Table 4.6-1 since Technical Specifications are being changed to require inservice inspection / inservice' testing per an external document (i .e., ASME Section XI Code);

3. -Deletion of requirement to demonstrate operabilit of safety-related components (e.g., ECCS, service water, etc.) ywhen a redundant or associated safety-related component is declared inoperable. This change is consistent with that of Standard Technical Specifications as discussed in the transmittal letter for this Enclosure. The only surveillance requirements which are applicable are those normally performed in accordance with ASME Section XI pursuant to 10 CFR 50.55a;

4. Clarification of Technical Specifications 4.5.A.l.b and 4.5.B.l.c for Core Spray and RHR, respectively, to indicate that flow rate testing will be conducted at a system pressure corresponding to a reactor pressure as given in the aforementioned sections. This will eliminate any confusion that may arise over point of pressure measurement and is consistent with Hatch Unit 2 Technical Specifications wording;

5. Change of frequency of RHR drywell and torus spray headers and nozzles air / water test from once/5 years to once/10 years.

Proposed testing frequency is consistent with ASME Section XI Code requirements;

6. Deletion of the requirement in existing Technical Specification 4.5.C.l.b for a pump capacity test following pump maintenance.

Post-maintenance testing of the affected pump (s) is required by the ASME Section XI Code and existing plant surveillance procedures; thus, redundancy in Technical Specifications is not necessary; h

ENCLOSURE 3 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS

7. Revision of Technical Specification " Bases" sections to reflect inservice inspection / inservice testing per ASME Section XI Code requirements by Technical Specification section reference (i.e.,

4.6.K) and other Technical Specification references, as appropriate; and,

8. Editorial changes in " Table of Contents", etc., as appropriate, to support inservice inspection / inservice testing-related Technical Specification changes.

BASIS The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety are not increased above those analyzed in the Final Safety Analysis Report (FSAR) due to these changes because the changes do not change equipment operation but only testing requirements. The possibility of an accident or malfuncton of a different type than analyzed in the FSAR does not result form these changes because the changes have no effect on equipment operation, thus, no new modes of failure are created. The margin of safety as defined in Technical Specifications is not reduced as a result of the changes because equipment operability is adequately assured by inservice inspection / inservice testing in accordance with ASME Boiler and Pressure Vessel Code requirements pursuant to 10 CFR 50.55a.

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ENCLOSURE 4 t

NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS Pursuant to 10 CFR 50.92, Georgia Power Company has evaluated the attached proposed amendment for Plant Hatch Unit 1 and has determined that its adoption would not involve a significant hazard. The basis for this determination is as follows:

PROPOSED CHANGES On page iv of the Table of Contents, change the page numbers for items 3.6.I/4.6.I and 3.6.J/4.6.J and omit the reference to " Primary Pressure Boundary" under items 3.6.K/4.6.K.

BASIS The changes are consistent with Item (i) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register. The page numbers for items 3.6.I/4.6.I and 3.6.J/4.6.J are revised due to text processing to accommodate new proposed wording for 3.6.K/4.6.K on page 3.6-10 of the Technical Specifications. The deletion of the reference to " Primary Pressure Boundary" is necessary since the proposed inservice inspection / inservice testing-related changes to the Technical Specifications covers a broader scope than the prinary pressure boundary.

As a result, the existing title for item 3.6.K/4.6.K 'is modified to reflect this.

PROPOSED CHANGE Delete all references to Table 4.6-1 in the List of Tables on page viii of the Technical Specifications.

BASIS This change is consistent with Item (1) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the

_ Federal Register. The change is administrative in nature in that it removes reference to the old Inservice Inspection Program (Table 4.6-1) since the o'd program was superceded by a new inspection program as required by 10 CFR 51.55a. In addition, the subject table is deleted since it does not reflect carrent inservice inspection / inservice testing requirements.

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ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND-INSERVICE INSPECTION TECHNICAL SPECIFICATIONS PROPOSED CHANGES Revise Technical Specification 4.4.A to reflect the following for the Standby Liquid Control System:

1. Add new Technical Specification 4.4.A.1 for monthly verification of explosive charge in each loop.

2. Install 10 CFR 50.55a testing requirements (i.e., ASME Section XI Code) in new Technical Specification 4.4.A.2 for pump testing.

3. Re-number existing Technical Specification 4.4.A.2 to read 4.4.A.3 to acco wodate new Technical Specification 4.4.A.l.

BASIS The following changes as identified above are listed by Technical Specification section number (proposed revision). These changes are consister<t with the appropriate Item number of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register for the reasons given below:

1. Proposed Technical Specification 4.4.A.1: The change is consistent with Item (ii) of the " Examples" because it adds the requirement of additional control of monthly continuity checks of the explosive charges to the existing surveillance.

2. Proposed Technical Specification 4.4.A.2: This change is consistent with Item (vi) of the " Examples" and installs the 10 CFR 50.55a testing requirements for the pumps and are of a lesser frequency than the existing Technical Specifications, 3 months vice 1 month. As a result, this change may appear to slightly reduce an existing safety margin. However this change still lies within the criteria of Standard Review Plan (SRP) section 9.3.5 for periodic testing of components.

3. Proposed Technical Specification 4.4.A.3: The re-numbering of existing Technical Specification 4.4.A.2 to 4.4.A.3 is consistent with Item (i) of the " Examples" since it is an administrative change necessary to accommodate new Technical Specification 4.4.A.l.

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS PROPOSED CHANGE Delete Technical Specification 4.4.8 concerning surveillance with Standby Liquid Control System inoperable components.

BASIS This change is consistent with Item (vi) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register. This change may possibly reduce a safety margin if the redundant portion of the Standby Liquid Control System fails in conjunction with a drastic reduction of control rod shutdown capability. Due to the low probability of such an occurrence, and the consistency of system testing and inspection required by 10 CFR 50.55a (as referenced by proposed wording in Technical Specification 4.6.K in the attached amendment), the results of this change still lie within criteria defined in SRP section 9.3.5. One portion of 4.4.8 concerned itself with explosive charge continuity testing.

Removal of that portion of Technical Specification 4.4.B is consistent with Item (i) since the continuity testing requirements is now covered by proposed Technical Specification 4.4.A.1.

PROPOSED CHANGES Change the " Bases" section of Technical Specification to reflect the following:

1. In Section 3.4.B, add references to inservice testing criteria; and,

2. In Section 3.4.C, correct typographical error relative to sodium pentaborate solution maintenance temperature.

BASIS The above changes are consistent with Item (1) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register. The change to section 3.4.B incorporates the stipulations of 10 CFR 50.55a, Inservice Testing criteria (as referenced by new wording proposed for Technical Specification 4.6.K). As a result, the change to 3.4.B is consistent with the proposed changes to Technical Specification 4.4.A. The change to section 3.4.C provides for a correction of a typographical error (10F to 100F). As a result, it is administrative in nature.

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r ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL -SPECIFICATIONS PROPOSED CHANGES Revise Technical Specification 4.5.A to reflect the following for the Core Spray System:

1. Add clarification to the term " system head" in Technical Specification 4.5.A.l.b.

2. Delete existing Technical Specification 4.5.A.1.c concerning pump operability testing and frequency thereof.

3. Delete existing Technical Specification 4.5.A.l.d concerning motor operated valve operability testing and frequency thereof.

4. Revise Technical Specification 4.5.A.2 to remove the requirement to demonstrate the operability of the remaining Core Spray loop and the RHR LPCI mode when one Core Spray loop is inoperable. In addition, remove from section 4.5.A.2 the requirement to test the operable Core Spray loop immediately (with one loop inoperable) and daily _thereafter until both loops are operable.

BASIS The following changes as ider.tified above are listed by Technical Specification section number. These changes are consistent with the appropriate item number of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register for the reasons given below:

1. Technical Specification 4.5.A.l.b: The change is administrative in nature and is consistent with Item (1) of the " Examples" because it provides clarification of the term " system head". The proposed wording is similar to that which exists in the Hatch Unit 2 Technical Specifications.

2. Technical Specification 4.5.A.l.c: The deletion of the monthly pump operability testing from Technical Specifications is deemed to be consistent with Item (vi) of the " Examples" since it may appear to reduce slightly an existing safety margin. Pump operability will be conducted under the auspices of the ASME Section XI Code as referenced in 10 CFR 50.55a and at the Code-specified frequency.

Technical Specification 4.6.K is being modified by the attached amendment to reference the conduct of inservice inspection / inservice testing per 10 CFR 50.55a. The change lies within the criteria of SRP 6.2.2 for periodic testing of components.

1 b

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS BASIS (Continued)

3. Technical Specification 4.5.A.l.d: The deletion of the monthly motor- operated valve operability testing from Technical Specifications is deemed to be consistent with Item (vi) of the

" Examples" since the testing will be conducted under the auspices of the ASME Section XI Code as referenced in 10 CFR 50.55a.

Technical Specification 4.6.K is being modified by the attached amendment to reference the conduct of inservice inspection / inservice testing per 10 CFR 50.55a. The test frequency remains unchanged due to Code requirements even though the test frequency is deleted in Technical Specifications.

4. Technical Specification 4.5. A.2: Removal of the requirements to demonstrate the operability of remaining Core Spray loop and the RHR LPCI mode when one Core Spray loop is inoperable and the removal of the requirement to test the operable Core Spray loop immediately (with one loop inoperable) and daily thereafter until both loops are operable are consistent with Item (vi) of the

" Examples." The reasons for this determination are as follows:

a. The removal of the requirement to demonstrate the operability of the remaining Core Spray loop and the RHR LPCI mode when one Core Spray loop is inoperable may reduce slightly an existing safety margin if a loss of RHR LPCI occurs in conjunction with the loss of one Core Spray loop. The chances of such an event are small however, and even with such an occurrence, the redundant Core Spray loop is still available. Based on these considerations and the testing requirements imposed as a result of 10 CFR 50.55a, the change still lies within the criteria of SRP sections 6.3 and 5.4.7.

b. The removal of the requirement to test the remaining operable Core Spray loop immediately and daily thereafter until both Core Spray loops are operable may reduce slightly an existing margin of safety by limiting the I assurance of operability. However, this change still i lies within the criteria for Core Spray systems as found '

in SRP Section 6.2.2 and its proposed revision. This is because the testing requirements of General Design Criteria (GDC)-40 (10 CFR 50 - Appendix A) are adherred

~

to. In addition, system functionality is assured by the criteria of 10 CFR 50.55a which follows ASME Section XI {

l l

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r i-.

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE OPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS BASIS (Continued)

Code guidance for inservice inspection / inservice testing. As a result, necessary assurance of the functionality of the redundant Cere Spray loop is provided by consistent testing and monitoring of any degradation within the subject system.

PROPOSED CHANGES Revise Technical Specification 4.5.B to reflect the following for the Residual Heat Removal System:

1. Change the frequency of the air / water tests for the drywell and torus spray headers and nozzles in Technical Specification 4.5.8.1.a from once/5 years to once/10 years.

2. Add clarification to the term " system head" in Technical Specification 4.5.B.1.c.

3. Delete existing Technical Specification 4.5.B.l.d concerning pump operability testing and frequency thereof.

4. Delete existing Technical Specification 4.5.B.l.e concerning motor operated valve operability and frequency thereof.

5. Re-number existing Technical Specification 4.5.8.1.f to read 4.5.B.l.d as a result of the proposed deletion of Technical Specifications 4.5.B.1.c and 4.5.B.l.d.

6. Remove the requirement to verify operability of Core Spray system and the remaining LPCI pumps and associated flow paths in Technical Specification 4.5.B.2.a in the event of loss of one LPCI pump.

7. Remove the requirement to verify operability of the active components of the remaining LPCI subsystem and the Core Spray system in Technical Specification 4.5.B.2.b in the event of loss of one LPCI subsystem.

BASIS The following changes as identified above are listed by Technical Specification section number. These changes are consistent wi>5 the l appropriate Item number of the " Examples of Amendments that are Cor idered I Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register for the reasons given below:

p j

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE OPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS BASIS (Continued)

1. Technical Specification 4.5.B.l.a: Changing of the frequency of the drywell and torus spray header and nozzle air / water testing from once/5 years to once/10 years is consistent'with item (vii) of the " Examples" because the change is being made in order to maintain consistency with the requirements of tne ASME Section XI Code as referenced by 10 CFR 50.55a. Technical Specification 4.6.K is being revised by the attached amendment to reference the subject Code and 10 CFR 50.55a.

2. Technical Specification 4.5.B.l.c: The change is administrative in nature and is consistent with Item (i) of the " Examples" because it provides clarification of the term " system head" . The proposed wording is similar to that which exists in the Hatch Unit 2 Technical Specifications.

3. Technical Specification 4.5.B.1.d: The deletion of the monthly pump operability testing from Technical Specifications is deemed to be consistent with Item (vi) _ of the " Examples" since it may appear to reduce slightly an existing safety margin. Pump operability will be conducted under the auspices of the ASME Section XI Code as referenced in 10 CFR 50.55a and at the Code-specified frequency.

Technical Specification 4.6.K is being modified by the attached amendment to reference the conduct of the inservice ,

inspection / inservice testing per 10 CFR 50.55a. The change lies within the criteria of SRP section 5.4.7.

4. Technical Specification 4.5.B.l.e: The deletion of the monthly motor operated valve operability testing from Technical Specifications is deemed to be consistent with Item (vi) of the

" Examples" since the testing will be conducted under the auspices of the ASME Section XI Code as referenced in 10 CFR 50.55a.

Technical Specification 4.6.K is being modified by the attached amendment to reference the conduct of inservice inspection / inservice testing per 10 CFR 50.55a. The test frequency remains unchanged due to Code requirements even though the test frequency is deleted in Technical Specifications.

5. Technical Specification 4.5.B.l.f: The renumbering of Technical Specification 4.5.B.l.f to 4.5.B.l.d is consistent with Item (1) of the " Examples" since it is an administrative change necessitated by the deletion of Technical Specifications 4.5.B.1.d and 4.5.B.l.e.

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS BASIS (Continued)

6. Technical Specification 4.5.B.2.a: Removal of the requirement from Technical Specification to verify the operability of the remaining LPCI pumps and associated flow paths and Core Spray system with one LPCI pump inoperable is deemed a change consistent with Item (vi) of the " Examples" . The change may reduce an existing margin of safety if a loss of the Core Spray system occurs in conjunction with a loss of one LPCI pump. However, this is an unlikely occurrence, and even in such a case, the redundant- LPCI system is available. Based on these considerations, and the testing requirements imposed as a result of 10 CFR 50.55a (as referenced in Technical Specification 4.6.K in the attached amendment), this change still lies within-the criteria of SRP sections 5.4.7 and 6.3..  ;

7. Technical Specification 4.5.B.2.b: Removal of the requirement from Technical Specifications to verify operability of the active components of the remaining LPCI subsystem and the Core Spray system immediately and daily thereafter with one LPCI subsystem inoperable is deemed a change consistent with Item (vi) of the

" Examples". The change may slightly reduce an existing safety margin. However, this change still lies within the criteria of SRP sections 5.4.7 and 6.3. In addition, testing requirements invoked by 10 CFR 50.55a (as referenced in Technical Specification 4.6.K in the attached amendment) add additional assurance of system operability.

PROPOSED CHANGE Add the word "or" to Technical Specification 3.5.C.l.b which specifies a condition when the RHR service water system shall be operable.

BASIS This change is consistent with Item (1) of the " Examples of Amendments that are Considered Not likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register. The wording change results from a change in nomenclature to distinguish the two cases considered (i.e., Technical Specification 3.5.C.b - reactor vessel pressurized, Technical Specification 3.5.C.c -

reactor vessel depressurized).

(

i

s P

EHCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS PROPOSED CHANGE Revise Technical Specification 4.5.C to reflect the following for the RHR Service Water System:

1. Delete existing Technical Specification 4.5.C.l.a concerning pump and valve operability testing and frequency thereof.

2. Revise Technical Specification 4.5.C.l.b " nomenclature" to accommodate deletion of 4.5.C.l.a.

3. Delete requirement in Technical Specification 4.5.C.1.b to perform testing after pump maintenance.

4. Delete Technical Specification 4.5.C.2 concerning surveillance with

-one pump inoperable.

5. Delete the requirement in Technical Specification 4.5.C.3 to test the remaining operable RHR service water subsystems immediately and daily thereafter with two RHR service water pumps inoperable.

BASIS The following changes as identified above are listed by Technical Specification section number. These changes are consistent with the appropriate Item number of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register for the reasons given below:

1. Technical Specification 4.5.C.l.a: The deletion of the quarterly pump and valve operability testing (4.5.C.l.a) from Technical Specifications is deemed to be a change consistent with Item (vi) of the " Examples" since it may appear to reduce slightly an existing safety margin. Pump and valve operability will be conducted under the auspices of the ASME Section XI code as referenced in 10 CFR 50.55a and at the Code-specified test frequency. Technical Specification 4.6.K is being modified by the attached amendment to reference the conduct of the inservice inspection / inservice testing per 10 CFR 50.55a.

2. Technical Specification 4.5.C.l.b: As a result of the deletion of Technical Specification 4.5.C.1.a, it is no longer necessary to have an item "b" suffix for Technical Specification 4.5.C.l.

Deletion of the letter "b" is consistent with Item (i) of the

" Examples" since it is an administrative change.

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ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL-5PECIFICATIONS BASIS (Continued)

3. Technical Specification 4.5.C.l.b: Deletion of the pump capacity test from the Technical Specifications after pump maintenance is performed is a change consistent with Item (vi) of the " Examples" since it may appear that an existing safety margin is being reduced. Pump testing will be conducted under the auspices of the ASME Section XI code as referenced in 10 CFR 50.55a. Technical Specification 4.6.K is being revised by the attached amendment to reference the conduct of the inservice inspection / inservice testing per 10 CFR 50.55a. The Code specifically requires a pump test after the performance of maintenance on the subject pump (s). In addition, the plant surveillance procedures also require that a l pump test be conducted following pump maintenance.

4. Technical Specification 4.5.C.2: The deletion of the requirement i l to verify operation of the remaining RHR service water j l pumps / subsystems upon failure of one RHR service water pump is a i

change consistent with Item (vi) of the " Examples" since it may appear to be a reduction in an existing margin of safety only if the failure of. the other pumps, etc. is considered.

4 Since there l

(- '

are four pumps total, sufficient redundancy is provided to make l such a consideration a highly unlikely event. This change does not j alter the consistency of the RHR Service Water System design, )

operability, inspection, and testing with that of SRP Section 5.4.7.

t

5. Technical Specification 4.5.C.3: The removal of the Technical l Specification requirements to verify immediately and daily l thereafter, proper operation of the remaining two functioning RHR l service water subsystems upon loss of two RHR service water pumps

) -

is considered a change consistent with-Item (vi) of the " Examples" since there may be a slight reduction in an existing safety margin. However, this change does not alter the consistency of the RHR Service Water System design, 9perability, inspection, and testing with that of SRP section 5.4.7. This is because the testing requirements not only remain in accordance with SRP 5.4.7, but, are covered by the details of 10 CFR 50.55a as well.

Technical 20ecification 4.6.K is being modified by the attached amendment to reference the conduct of the inservice inspection / inservice testing per 10 CFR 50.55a.

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ENCLOSURE 4 (Continued) l NRC DOCKET 50-321 OPERATING LICENSE DPR-57 l EDWIN I. HATCH NUCLEAR PLANT UNIT 1 l

REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS PROPOSED CHANGES Revise Technical Specification 4.5.D to reflect the following for the High Pres ure Coolant Injection (HPCI) System:

1. Add the word " continued" to the title line for the remainder of Technical Specification 4.5.D.l.b located on Technical l Specification page 3.5-7.

2. Delete existing Technical Specification 4.5.D.l.d concerning pump operability testing and frequency thereof.

3. Delete existing Technical Specification 4.5.D.l.e concerning motor operated valve operability testing and frequency thereof.

4. Delete the requirements in Technical Specification 4.5.D.2 to test the RCIC System, the RHR system LPCI mode, and the Core Spray system immediately with HPCI inoperable. In addition, in the aforementioned specification, delete the requirement to test the RCIC system daily until the HPCI system is returned to normal.

I BASIS The following changes as identified above are listed by Technical Specification section number. These changes are consistent with the appropriate Item number of the " Examples of Amendments that are Considered Not Likely to involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register for the reasons

.- gives below:

1. Technical Specification 4.5.D.l.b: The change is consistent with Item (i) of the " Examples" since it provides for syntax correction by adding a " continued" statement to the continuation of Technical Specification 4.5.D.l.b on page 3.5-7 of the Technical Specification.

2. Technical Specification 4.5.D.l.d: Deletion of the pump operability test requirements from the Technical Specification is a change consistont with Item (vi) of the "Examoles" since it may appear that an existing margin of safety is being reduced slightly. However, the testing requirements will be covered by the requirements of the ASME Section XI code per 10 CFR 50.55a and at the Code-specified test frequency. Testing in this manner is still in keeping with the testing and inspection criteria of SRP section 9.6.3.

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ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-7 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS BASIS (Continued)

3. Technical Specification 4.5.0.1.e: Deletion of the motor operated valve operability test requirements from the Technical Specifications is a change consistent with Item (vi) of the

" Examples" since it may appear that an existing margin of safety,is reduced slightly. However, the testing requirements will be covered by the requirements of the ASME Section XI Code per 10 CFR 50.55a and at the Code-specified test frequency. Testing in this manner is still in keeping with the testing and inspection criteria of SRP section 9.6.3.

4. Technical Specification 4.5.D.2: The removal of the requirement to verify immediate cperability of the Core Spray, the RHR LPCI mode, and the RCIC system with HPCI inoperable is a change consistent with Item (vi) of the " Examples" since it may reduce a possible

. safety margin. Similarly, deletion of the requirement to test the RCIC system daily until the HPCI system is returned to normal if HPCI system components are inoperable may reduce a possible safety margin. A reduction in safety margin would occur if RCIC became unavailable in conjunction with the loss of HPCI, as well as, the low pressure ECCS capability of the redundantly configured Core Spray and LPCI systems. The redundant capabilities of Core Spray and LPCI covers postulated losses of low pressure cooling. The specific testing and inspection requirements introduced through 10 CFR 50.55a (which references the ASME Section XI Code) will help assure system operability. The change still remains within the criteria of SRP section 6.3.

PROPOSED CHANGES Revise Technical Specification 4.5.E to reflect the following for the Reactor Core Isolation Cooling (RCIC) System:

1. Delete existing Technical Specification 4.5.E.1.d concerning pump operability testing and frequency thereof.

2. Delete existing Technical Specification 4.5.E.1.e concerning motor operated valve operability testing and frequency thereof.

3. Delete existing Technical Specification 4.5.E.2 concerning surveillance with inoperable components.

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-7 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTIGN TECHNICAL SPECIFICATIONS BASIS The following changes as identified above are listed by Technical Specification section number. These changes are consistent with the appropriate Item number of the " Examples of Amendments that are Considered Not Likely to involve Significant Hazards Considerations" as listed on page 14870 of the April 6, 1983 issue of the Federal Register for the reasons given below:

1. Technical Specification 4.5.E.1.d: Deletion of the pump operability test requirements from the Technical Specifications is a change consistent with Item (vi) of the " Examples" since it may appear that an existing margin of safety is being reduced.

However, the testing requirements will be covered by the requirements of the ASME Section XI Code per 10 CFR 50.55a and at the Code-specified test frequency. The change is still in keeping with the testing and inspection criteria of SRP section 5.4.6.

2. Technical Specification 4.5.E.1.e: Deletion of the motor operated valve operability testing requirements from Technical Specifications is a change consistent with Item (vi) of the

" Examples" since it may appear that an existing margin of safety is being reduced. However, the testing requirements will be covered by the requirements of the ASME Section XI Code per 10 CFR 50.55a and at the Code-specified frequency. The change is still in keeping with the testing and inspection criteria of SRP section 5.4.6.

3. Technical Specification 4.5.E.2: The removal of the requirement to verify immediately and daily thereafter the proper operation of the HPCI system when RCIC is inoperable is a change consistent with Item (vi) of the " Examples" since it may slightly reduce a safety margin. However, this change is still in keeping with SRP section 5.4.6. HPCI, as an alternate or backup system to RCIC, is subjected to similar testing criteria as RCIC. Testing criteria are now based on 10 CFR 50.55a (as referenced in Technical Specification 4.6.K in the attached amendment) and provides adequate assurance of system availability in the event of RCIC inoperability.

PROPOSED CHANGE In Technical Specification 4.5.G, remove the reference to testing immediately and daily thereafter the components of the RHR LPCI mode and containment cooling mode connected to the operable diesel generators, upon loss of one of the diesel generators.

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ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-7 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS BASIS The proposed change has been determined not to involve a significant hazard because:

1. The proposed change does not involve a significant increase in the probability or consequences of an accident because:

a. The components under consideration are redundant and have separate sources of electrical power. Therefore, the operable diesels would be able to power at least one of the redundant systems in the event power is shifted to the diesel generators;

b. Actual availability of the components under consideration is verified through other Technical Specification surveillance requirements and/or through the requirements of 10 CFR 50.55a;

c. Single component failure in redundant safety systems has been evaluated; and,

d. The proposed change will still include the requirement for testing all remaining operable diesels daily, until the inoperable unit is placed back in service. ,

2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because the change does not modify the original intent of the surveillance requirement. The intent is to ensure adequate and reliable core and containment cooling capacity in the event that one source of backup power (i.e., one diesel generator) is lost.

The change merely simplifies this intent by focusing attention on the problem source which is an inoperable diesel generator and taking 'the originally intended corrective measure of increased diesel testing frequency. Verification of operability of the systems omitted in this change is accomplished in other Technical Specification surveillance requirements and/or the ASME Section XI Code requirements which are referenced in 10 CFR 50.55a.

3. The proposed change does not involve a significant reduction in the margin of safety because of the adequate operability assurance provided for by other surveillance requirements and/or 10 CFR 50.55a for the omitted systems (i.e., RHR LPCI and Containment Cooling).

Y____________-__-_____-.-_-_______-__--_

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-7 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO ' AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS PROPOSED CHANGES Revise Technical Specification 3.5.J to reflect the following for the Plant Service Water System:

1. In Technical Specification 3.5.J.1, change the number of plant service water pumps from 4 to 3.

2. In Technical Specification 3.5.J.2.b, add wording such that the diesel generators are those associated with the operable Plant Service Water (PSW) components.

3. In Technical Specification 3.5.J.2.c, add wording such that the diesel generators are those associated with the operable PSW components.

BASIS The following changes as identified above are listed by Technical Specification section number. These changes are consistent with the appropriate Item number of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" as listed on page 14870 of the April 6, 1983 issue of the Federal Register for the reasons given below:

1.- Technical Specification 3.5.J.1: The change is consistent with item (i) of the " Examples" since it corrects the number of plant service water pumps.

2. Technical Specification 3.5.J.2.b: The ' change is consistent with Item (i) of the " Examples" since it was made to provide consistency with other sections of the Technical Specifications which are concerned with equipment that can be diesel powered. This section references operability of only those particular diesels which can power the equipment in question.

3. Technical Specification 3.5.J.2.c: The change is consistent with Item (i) of the " Examples" since it was made to provide consistency with other sections of the Technical Specifications which are concerned with equipment that can be diesel powered. This section references operability of only those particular diesels which can power the equipment in question.

I

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-7 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO- AMEND -INSERVICE -INSPECTION TECHNICAL SPECIFICATIONS PROPOSED-CHANGES Revise Technical Specification 4.5.J to reflect the following for the Plant Service Water System:

.l. In Technical' Specification 4.5.J.2.b, delete the requirements to demonstrate operability of the standby service water pump, the remaining PSW pumps, and both PSW divisions when one PSW pump is made or found to be inoperable.

2. In Technical Specification 4.5.J.2.b, add wording such that the diesel generators are those associated with the operable PSW components.

3. In Technical Specification 4.5.J.2.c, delete the requirement to demonstrate operability of the remaining PSW pumps and both PSW divisions when one PSW pump and the standby service water pump are made or found to be inoperable.

-4. In Technical Specification 4.5.J.2.c, add wording such that the diesel generators are those associated with the operable PSW components.

5. In Technical Specifications 4.5.J.2.d, delete the requirement to demonstrate operability of the standby service water pump and all '

active components of the operable PSW division (s) when two PSW pumps or one PSW division are made or found to be inoperable.

6. In Technical Specification 4.5.J.2.e, delete the requirements to demonstrate operability of all active components of the operable PSW division (s) when two PSW pumps or one PSW division and the standby service water pump are made or found to be inoperable.

_ BASIS The following changes as identified above are listed by Technical Specification number. These changes are consistent with the appropriate Item number of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Consideration" as listed on page 14870 of the April 6, 1983 issue of the Federal Register for the reasons given below:

1. Technical Specification 4.5.J.2.b: Deletion of the requirement to demonstrate operability of the standby service water pump, the remaining PSW pumps, and both PSW divisions immediately and weekly thereafter when one PSW pump is made or found to be inoperable is a change consistent with Item (vi) of the " Examples" since it may e____-_-____-____-____________-______-_____-__--________. . _ _ _ _ _ _ _ _ - _ - _ _ _ _ - _ _ _ _ _ _ _ . _ _ _ _ _ _

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-7 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS

, BASIS (Continued) possibly reduce an existing safety margin if failure of the remaining components is considered. Considering the redundancy involved, significant failure of the remaining components is a very small possibility. In addition, testing and inspection requirements resulting from 10 CFR 50.55a adds an extra assurance of component reliability. Based on these considerations and the fact that the change is still within SRP criteria (i.e., SRP section 9.2.1), a significant hazard is not applicable.

2. Technical Specification 4.5.J.2.b: The addition of wording such that the diesel generators tested are those associated with the operable PSW components is a change which is consistent with Item (i) of the " Examples" since it was made to provide consistency with other sections of the Technical Specifications which are concerned with equipment that can be diesel powered. This section references operability of only- those particular diesels which can power the equipment in question.

Technical Specification 4.5.J.2.c: Deletion of the requirement to 3.

demonstrate operability of the remaining PSW pumps and both PSW divisions immediately and weekly thereafter when one PSW pump and the standby service water pump are made or found to be inoperable is a change consistent with Item (vi) of the " Examples" since it may possibly reduce an existing safety margin if failure of the redundant components is considered.

Considering the redundancy involved, significant failure of the remaining components is a very small possibility. In addition, testing and inspection requirements resulting from 10 CFR 50.55a adds an extra assurance of component reliability. Based on these considerations and the fact that the change is still within SRP criteria (i.e., SRP section 9.2.1), a significant hazard is not applicable.

4. Technical Specification 4.5.J.2.c: The addition of wording such that the diesel generators tested are those associated with the operable PSW components is a change which is consistent with Item (1) of the " Examples" since it was made to provide consistency with other sections of the Technical Specifications which are concerned with equipment that can be diesel powered. This section references operability of only those particular diesels which can power the equipment in question.

W-__--_____-_-________-_____-_-_--_-______ A

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-7 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS BASIS (Continued)

5. Technical Specification 4.5.J.2.d: Deletion of the requirements to demonstrate operability of the standby service water pump and all active components in the operable PSW division (s) immediately and daily thereafter when two PSW pumps or one PSW division are made or found to be inoperable is a change consistent with Item (vi) of the

" Examples" since it may possibly reduce an existing safety margin if failure of the remaining components is considered. Considering the redundancy involved, significant failure of the remaining components is a very small possibility. In addition, testing and inspection requirements resulting from 10 CFR 50.55a adds an extra assurance of component reliability. Based on these considerations and the fact that the change is still within SRP criteria (i.e.,

SRP section 9.2.1), a significant hazard is not applicable.

6. Technical Specification 4.5.J.2.e: Deletion of the requirements demonstrate operability of all active components of the operable PSW division (s) immediately and daily thereafter when two PSW pumps or one PSW division and the standby service water pump are made or found to be inoperable is a change consistent with Item (vi) of the

" Examples" since it may possibly reduce an existing safety margin if failure of the remaining components is considered. Considering the redundancy involved, significant failure of the remaining components is a very small possibility. In addition, testing and inspection requirements resulting from 10 CFR 50.55a adds an extra

-assurance of component reliability. Based on these considerations and the fact that the change is still within SRP criteria (i.e.,

SRP section 9.2.1), a significant hazard is not applicable.

PROPOSED CHANGE Change the " Bases" section of Technical Specifications to reflect the following:

1. In Section 3.5.A.2 relative to the Core Spray System, delete the wording concerning testing of the remaining Core Spray loop and the RHR system should one Core Spray loop become inoperable;

2. In Section 3.5.A.2, add wording such that the diesel generators tested are those associated with the remaining operable Core Spray loop should one Core Spray loop become inoperable; and,

3. In Section 3.5.A.2, add references to inservice inspection / inservice testing criteria.

k__.__________ __---____ _ _-----__ _ _____ ____ _ --__ __ _ __________ _ _ _

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE OPR-7 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS BASIS The above-changes to the " Bases" section of Technical Specifications for the Core Spray System are consistent with Item (1) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register since the changes are -either administrative in nature or are provided for consistency with other Technical Specifications, existing and/or proposed by the attached amendment.

PROPOSED CHANGES Change the " Bases" section of Technical Specifications to reflect the following:

1, In Section 3.5.B.2 relative to the RHR System, revise core flooding assurance statement to delete statements concerning demonstrated operability of the redundant LPCI pumps and subsystems and the Core Spray system. In addition, remove statement concerning out-of-service period; and,

2. In section 3.5.B.2, add references to inservice inspection / inservice testing criteria.

BASIS The above changes to the " Bases" section of Technical Specifications for the RHR System (LPCI and Containment Cooling Mode) are consistent with Item (1) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register since the changes are either administrative in nature or are provided for consistency with other Technical Specifications, existing and/or proposed by the attached amendment.

1 PROPOSED CHANGE Change Technical Specification " Bases" section 3.5.D.2 relative to the HPCI System to reference inservice inspection / inservice testiiig criteria for the backup RCIC System.

BASIS The change to the subject " Bases" section is a change consistent with Item (ii) of the " Examples of Amendments that are Considered Not Likely to -

Involve Significant . Hazards Considerations" listed on page 14870 of the April 6,1983 issue of the Federal Register since it provides an additional control not presently in the Technical Specifications.

E

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-7 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS PROPOSED CHANGE Change Technical Specification " Bases" section 3.5.E.2 relative to the RCIC System to reference inservice inspection / inservice testing criteria for the HPCI System.

BASIS The change to the subject " Bases" section is a change consistent with Item (ii) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Ha.'ards Considerations" listed on page 14870 of the April 6,1983 issue of - % Icderal Register since it provides an additional control not presently fa the Technical Specifications.

PROPOSED CHANGES Change Technical Specification " Bases" section 3.5.F.1 relative to ADS to reflect the following:

1. Change the title entry "F.1" to read "3.5.F.1"; and

2. Correct the spelling of the word " failures".

BASIS

. The above changes are consistent with Item (1) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the 4

Federal Register since the changes are either administrative in nature or correct a typographical error in the Technical Specifications.

PROPOSED CHANGES Change Technical Specifications " Bases" section 3.5.F.2 relative to ADS to reference inservice inspection / inservice testing criteria for the HPCI System.

BASIS The change is consistent with Item (i) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register since it provides consistency with other Technical Specificatiori changes concerning HPCI requested in the attached amendment.

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

1 ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-7 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE' INSPECTION TECHNICAL SPECIFICATIONS PROPOSED CHANGE Change Technical Specification " Bases" section 3.5.J/4.5.J relative to the Plant Service Water System to reference inservice inspection / inservice testing criteria to provide increased assurance of PSW System operability.

BASIS The change to the subject " Bases" section is a change consistent with Item (ii) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register since it inserts additional criteria to provide increased assurance of PSW System operability.

Therefore, it constitutes an additional control not presently included in

,the Technical Specifications.

PROPOSED CHANGES Relocate existing Technical Specifications 4.6.I.2, 4.6.I.3, 3.6.J.1, 3.6.J.2, and 4.6.J from Technical Specifications page 3.6-10 to page 3.6-9b and relocate Technical Specification 3.6.J.3 from Technical Specifications page 3.6-10 to new page 3.6-9c.

BASIS The changes are consistent with Item (1) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register since they are administrative in nature to accommodate new proposed wording for Technical Specifications 3.6.K/4.6.k on page 3.6-10 of the Technical Specifications.

PROPOSED CHANGES Delete in their entirety the existing Technical Specifications 3.6.K/4.6.K and substitute new prcposed wording for the subject Technical Specifications similar to that found in Standard Technical Specifications.

BASIS This change is consistent with Item (vii) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" as listed on page 14870 of the April 6,1983 issue of the Federal Register due to changes in inservice inspection / inservice testing requirements resulting from 10 CFR 50.55a. The new proposed wording is -

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-7 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS BASIS (Continued) similar to that in Standard Technical Specifications and supercedes the existing outdated Technical Specification requirements. The existing Technical Specifications surveillance requirements for inservice inspection were for a program written to the 1971 Edition of the ASME Section XI Code with Addenda through Summer 1972. Pursuant to the requirements of 10 CFR 50.55a, the licensee is required to periodically update the inspection program to a later edition of the Code and the requirements it sets forth.

The existing Technical Specifications do not reflect the current inspection requirements. Further, deletion of the existing Technical Specifications is justified since some of the requirements (e.g, reporting of inspection results after 5 years presumably from the commercial service date upon which the inservice inspection is based) have already been met and, thus, arc.

antiquated. The new proposed Technical Specifications are written such tnat they will not require revision each time the program is revised per 10 CFR 50.55a since no specific edition / addenda of the Code is referenced la the proposed Technical Specifications.

PROPOSED CHANGE Delete in its entirety Technical Specification Table 4.6-1, " Inservice Inspection Program".

BASIS .

The change is consistent with Item (1) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" as listed on page 14870 of the April 6, 1983 issue of the Federal Register since deletion of Table 4.6-1 provides consistency with the proposed deletion of outdated Technical Specifications existing in Technical Specifications 3.6.K/4.6.K. New proposed wording for Technical Specifications 3.6.K/4.6.K are provided in the attached amendment. The subject table does not reflect the current inspection program as providea for by the requirements of 10 CFR 50.55a which requires program update' periodically.

PROPOSED CHANGE Revise Technical Specifications " Bases" section 3.6.K to reflect the current inservice inspection / inservice testing requirements per 10 CFR 50.55a and to delete any outdated material currently existing.

ENCLOSURE 4 (Continued)

NRC DOCKET 50-321 OPERATING LICENSE DPR-7 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 REQUEST TO AMEND INSERVICE INSPECTION TECHNICAL SPECIFICATIONS BASIS The change is consistent with Item (i) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" as listed on page 14870 of the April 6, 1983 issue of the Federal Register to provide consistency with the new wording proposed for Technical Specifications 3.6.K/4.6.K as shown in the attached amendment.

The changes are as a result of applying criteria set forth by the current requirements of 10 CFR 50.55a.

PROPOSED CHANGES Add "less than" symbols relative to calm wind speed to Technical Specifications 4.7.C.l.b and 4.7.C.2.

BASIS The changes are consistent with Item (i) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Cor.siderations" as listed on page 14870 of the April 6, 1983 issue of the Federal Register since the changes correct ommission of the "less than" symbols from the last Technical Specification amendment (i.e., Amendment 100) to page 3.7-13 of the Technical Specifications.

.