ML20147B662

Forwards Proposed Amend to Tech Specs Which Will Upgrade the Insvc Inspec Requirements Per 10CFR55a(g)(4)(ii).Proposed Changes Rely on Standard Tech Specs

ML20147B662
Person / Time
Site:	Hatch
Issue date:	10/03/1978
From:	Whitmer C GEORGIA POWER CO.
To:	Office of Nuclear Reactor Regulation
References
TAC-07811, TAC-08019, TAC-59334, TAC-7811, TAC-8019, NUDOCS 7810110089
Download: ML20147B662 (31)

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'" ' ( tet )l X : P()Wel October 3, 1978 Direct 6r of Nuclear Reactor Regulation U.~ S. Nuclear Regulatory Commission Washington, D. C. 20555 NRC DOCIGT 50-321 OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 IN-SERVICE INSPECTION Gentlemen:

Pursuant to 10 CRF 50.90, as required by 10 CFR 50.59(c)(1), Georgia Power Company hereby proposes an amendment to the Technical Specifica-tions (Appendix A to the Operating License). The proposed change will update the in-service inspection requirements in accordance with 10 CFR 55a(g)(4)(ii).

Our letter dated August 3,1978, provided you with a description of the in-service inspection and testing program for Plant Match Unit 1.

We also advised you that proposed changes to the Technical Specifications would be submitted to reflect the provisions of the revised program.

We have conducted a review of the requirements of the applicable ASME Section XI Code and the Technical Specifications presently in effect in l

order to eliminate any conflicts which may exist between the two. You

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l will find enclosed proposed changes to the Technical Specifications which l

conform them to the updated program.

In accordance with your recommendations, we have incorporated stan-dard language referencing 10 CFR 50.55a(g) in place of existing in-service inspection and testing requirements. Houever, in lieu of simply referencing l

10 CFR 50.55a(g), which would consequently incorporate the additional sur-veillance requirements required by ASME Section XI in conjunction with those that presently exist in the Technical Specifications, we took the approach of preparing the proposed changes in the direetion of the requirements of Standard Technical Specifications (STS). Such an approach introduces requirements some of which are more restrictive while others are less re-

- strictive than the requirements currently in the Technical Specifications.

As a point of clarification of our intentions, STS do not require safety-related components (e.g. , ECCS, service ' water, etc. ) to be demonstrated  !

operable when a redundant or associaLed safety-related component is declared inoperable. The only surveillance requirements which are applicable are

.those normally performed in accordance with ASME Section XI purcuant to 10 CFR 50.55a(g). Although the STS requirements are less restrictive than h

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Director of Nuclear Reactor Regulation October 3, 1978 Page 2 the existing Technical Specifications in that respect, it must be emphasized that the normal testing required by ASME Section XI is significantly more comprehensive than our presene testing requirements.

Therefore, the net result using the STS approach should be an equivalent or more reliable testing program.

Based on the above discussion, the Plant Review Board and Safety Review Board have reviewed and approved the proposed changes to the Technical Specifications. The probability of an accident or possibility of failures of a type not previously considered are not increased since the proposed changes are based on sufficiently conservative requirements set forth in STS. Margins of safety are not reduced. Therefore, it can be concluded that the proposed changes do not constitute an unreviewed safety question.

Yours very truly,

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l F. Whitmer WE3/JMM/mb Enclosure

1. Proposed Determination of Amendment Class

2. Proposed changes to Technical Specifications

'Secro to and subscribed before me this 3rd day of October, 1978.

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Wb t:cf ruuc, cecry, m38otory3 Public b%cammwen bpres Stat. s. lssi xe: Ruble A. Thomas George F. Trowbridge, Esquire 4

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ATTACHMENT 1 f NRC DOCKFT 50-321' l OPERATING LICENSE DPR-57 7 EDWIN 1. HATCH MUCLEAR PLANT UNIT 1 PROPOSED DETERMINATION OF AMENDMENT CLASS 1

Pursuant to 10 CFR 170.12 (c), Cecrgik Power Company has evaluated the attached proposed amendmentLto Operating License DPR-57 and have determined that:

a) The proposed; amendment does not require the evaluation of a new Safety Analysis Report or rewrite of the facility license; b) The proposed amendment does not contain several complex issues,  ;

does not involve ACES review, or does not require an environmoutal '

impact statement; c) The proposed amendment does not involve a complex issue, an environ-mental ~ issue or more than one safety. issue; d) The proposed amendment does involve a single issue; namely the up-dating of in-service inspection and testing surveillance to meet current codes and should thus be acceptable; e) The proposed amendment is therefore a Class III amendment.  :

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NRC DOCKET 50-321 'l OPERATING LICENSE DPR-57 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS The proposed changes to the Technical Specifications (Appendix A to Operating Licence DPR-57) would be incorporated ns follows:

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3.5-8 3.5-8 l 3.5-10 3.5-10 '

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3.6-11 thru 3.6-14 3.6-11 thru 3.6 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 3.7-14 3.7-14 1

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Section- 'Sectiog ,

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_i LIMIT 1NG CONDITIONS FOR OPERATION- SURVEILLANCE REQUIREMENTS i3. 6' PRIMARY SYSTEM B0UNDARY 4.6- PRIMARY SYSTEM BOUNDARY 3.6-l'  !

-A. Reactor Coolant Heatup- A. . Reactor Coolant Heatup 2.6-1 and Cooldown and Cooldown ,

B.. Reactor Vessel Temperature B. Reactor Vessel Temperature 3.6-1 i and Pressure and Pressure LC. Reactor Vessel Head Stud C. Reactor Vessel Head Stud 3.6 '

Tensioning Tensioning D. Idle Recirculation Loop D. Idle Recirculation Loop 3.6-2  ;

.Startup Startup E. Recirculation Pump Start E. Recirculation Pump 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 l

H. ' Safety and Relief Valves H. Safety and Relief Valves 3.6-9 1

I. Jet' Pumps I. Jet Pumps 3.6-9 l

.J. Recirculation Pump Speeds J. Recirculation Pump Speeds 3.6-9a >

K. Structural Integrity K. Structural Integrity 3.6-10 L. Shock Suppressors L. Shock Suppressors 3.6-10a 3.7 ' CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS 3.7-1 1

'A. Primary Containment A. Primary Containment 3.7-1 B. Standby Gas Treatment System B. Standby Gas Treatment System 3.7-10 1 C. Secondary Containment C. Secondary Containment 3.7-12 ,

D. Primary Containment D. Primary Containment 3.7-13' Isolation Valves - Isolation Valves 3.8 RADI0 ACTIVE MATERIALS 4.8 RADI0 ACTIVE MATERIALS 3.8-1 A. ' Miscellaneous Radioactive A. Miscellaneous Radioactive 3.8-1 i Materials Sources- Materials Sources

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3.9 AUXILARY ELECTRIC _ 43TEMS 4.9 AUXILIARY ELECTRICAL SYSTEMS 3.9-1 A.' Requirements for Reactor A. Quxiliary Electrical Systems 3.9-1 ,

Startup Equipment

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. . LIST OF TABLES (Concluded)

'Page Table Title 4.2 Check, Functional Test, and ralibration Minimum Frequency 3.2-40 for Neutron Menitoring Instrumentation Which Initiates- Con- t trol Rod Blocks 4.2-8 Check, Functional Test, and Calibration Minimum Frequency 3.2-42 for Radiation Monitoring Systems Which Limit Radioactivity Release 4.2-9 Check and. Calibration . Minimum Frequency for Instrumentation- 3.2-45 Which Initiates Recirculation Pump Trip 4.2-10 Check, Functional Test, and Calibration Minimum Frequency 3.2-46 for Instrumentation Which Monitors Leakage into the Drywell 4.2-11 Check and Calibration Minimum Frequency for Instrumentation 3.2-48 Which Provides Surveillance Information 3.6.1 Safety Related Shock Suppressors (Snubbers) 3.6-10c 3.7-1 Primary Containment' Isolation Valves 3.7-16 3.7-2 Testable Penetrations with Double 0-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 6.0-1 Shift Manning Chart for Unit Operation 6.0-23 6.0-3 Special Test Report Requirements 6.0-25 ,

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. LIMITING CONDITIONS FOR OPERATION SURl!EILLANCE REQUIREMENTS 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 ' STANDBY LIQUID CONTROL SYSTEM Applicability Applicability The limiting Conditions for Opera- The Surveillance Requirements apply-

' tion apply to the operating status to the periodic test and examination of the Standby Liquid Control of the Staridby Liquid Control System. ,

System. l 1

Objective Objective l l

The objective of the limiting The objective of the Surveillance j Conditions for Operation is to Requirements is to verify the oper- I assure the availability of a ability of the Standby Liquid Control 4 system with the capability to System. l shut down the reactor and main-tain the shutdown condition  ;

without the use of control rods.

Specifications Specifications A. Normal System Availability A. Normal Operational Tests During periods when fuel is The operability of the Stendby in the reactor and prior to Liquid Control System shall be startup from the Cold Shut- verified by the performance of down Condition the standby the following tests- I liquid control system shall  !

be operable except:

1. When performing control rod 1. Monthly l drive maintenance, at which Verify the continuity of the l time Specification 3.10.E explosive change in each loop. i shall be met, 2. As required by Specification 4.6.} l Each pump loop shall be locally 1 or started and functionally tested by recirculating de-

2. When operating with an inoper- mineralized water to the test i able component, at which time tank. l Specification 3.4.B shall be 3. Each Operating Cycle l met, At least once during each oper-ating cycle:

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a. Check that the setting of l

3. When the reactor is in the the system relief valve is I I

Cold Shutdown Condition and 1325 + 75 psig.

all control rods capable of b. Verify that each pump will I normal insertion are inserted deliver 43 gpm against a and the requirements of Speci- system head of at least 1190 fication 3.3.A are met, psig.

c. Initiate one of the Standby ,

Liquid Control System Loops  !

from the control room after arranging suction from the test tank and pump deminer- i alized water into the reactor. l 3.4-1 0 . d

e LIMITlNG CONDITIONS FOR OPERAT10N- SURVEILLANCE REQUIREMENTS , 4.4.A.2 Each Operating Cycle (Continued)_

c. vessel. This test checks the explosive charge, proper opera-tion of the associated valves and selected pump operability.

The replacement charge to be instulled will be selected from a manufactured batch which has been tested,

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

3.4.B. Operating with Inoperable Components.

If one Standby Liquid 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 Liquid Control System is re- performed to verify the avail-quired to be operable the ability of the liquid control following conditions shall be solution.

met:

1. Volume 1. Volume The volume of the liquid Check _the standby liquid control solution in the control tank volume at least liquid control tank shall once per day.

be maintained as required in Figure 3.4-1.

2. Concentration 2. Concentration The concentration to the Check the concentration of the liquid control tank shall liquid in the standby liquid be maintained as required control tank by chemical in Figure 3.4-1. analysis.

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G BASES FOR LlMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.4.8 Operation with -Inoperable components (Continue'd)-

redundant' 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 shutdown capability and the alternate component in the system, together with the fact that nuclear system cooldown takes several hours while liquid control solution injection takes about two hours. This indicates the considerable time avail-able for testing and restoring the Standby Liquid Control System to an operable condition after testing, while reactor operation. continues.

Assurance that the system will still fulfill its function during repairs is obtained by the surveillance testing required by Specifications 4.4.A and 4.6.K.

Each positive displacement pump is sized to inject the solution into the reactor in 50 to 125 minutes, independent of the amount of solution in the tank. The slower rate assures that the boron gets into the reactor considerably quicker than the cooldown rate. The faster injection rate limit assures that there is sufficient mixing so the baron does not recirculate through the core.in uneven concentrations which could possibly-cause the nuclear power to rise and fall cyclically.

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

A minimum required pump flow rate of 41.2 gpm has t"en determined by the rate reouired using one pump to inject the maximum volume of control solution (5150 gallons) within the maximum allowed time of 125 minutes.

Using the minimum pump rate of 41.2 gpm and the fastest injection time of 50 minutes, a minimum quantity of 2060 gallons of solution having a 20.2 percent sodium pentaborate concentration is required to meet the shutdown requirements. For the maximum expected pump capacity of 43 gpm a minimum volume of 2150 gallons is that volume which could be injected in the minimum allowed time of 50 minutes.

C. Sodium Pentaborate Solution Limiting Conditions for Operation:

The liquid control solution is acceptable if the combination of volume and concentration of the solution is maintained in the region required i as shown in Figure 3.4-1 and the solution temperature is maintained i 100F above the corresponding saturation temperature (Figure 3.4-2) 1 to guard against boron precipitation.

Surveillance Requirements l

L Level indication and an alarm indicate whether the solution volume has l- changed, which might indicate a possible solution concentration change.

The combination of volume and concentration required of the solution is such that should evaporation occur from any point within the acceptable region,. a -low level alarm will annunciate before the combination of volume and concentration requirements are unacceptable. The test interval has-been established in consideration of these factors. The solution temperature and volume are checked at a high enough frequency to assure a high reliability of acceptability of the solution should it ever be required. Temperature and liquid level alarms for the system are annunciated in the control room.

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m LIMITING CONDITIONS FOR OPERATION SURl/EXLLANCE RE0VIREMENTS 3.5.-CORE AND CONTAINMENT COOLING 4.5 CORE AND CONTAINMENT COOLING SYSTEMS SYSTEMS Applicability Applicability

-The Liminting Conditions for. The Surveillance Requirements 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.

Objecti ve ' Objective The objective of the Limiting The objective of the Surveillance Conditions for 0peration is to Requirements is to. verify the assure the operability of the operability of the core and con-core and containment cooling tainment cooling systems under all systems under all conditions conditions for which this cooling for which this. cooling capa- capability is an essential response bility is an essential to plant abnormalities.

response to plant abnormalities.

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

l. Normal System Availability 1. Normal Operational Tests

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a. The CS System shall be operable: CS system testing shall be  ;

(1) Prior to reactor startup from a cold condition, or Item Frequency (2) When irradiated fuel is in a. Simulated Once/ Operating the reactor vessel and the Automatic Cycle i i reactor pressure is greater Actuation  !

than atmospheric pressure, Test except as stated in Specifi-cation 3.5.A.2. b. System flow Oncc/3 months rate:

Each loop shall deliver at least 4625 gpm against a system head of at least 113 psig.

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LIMITING CON 0lTI0iiS FOR 0PERATION - SURVEILLANCE REQUIREMENTS 3.5.A.2. Operation with I'noperable'- 4.5.'A.2. Surveillance with Inoperable O Components Components l i

If'one CS system loop'is inoper- When it is determined'that one i able,1the reactor may remain in core spray loop is inoperable operation for a at a time when operability is exceed seven (7)pariod not todays provided required,-the diesel generators all active components in the shall be demonstrated to be other CS system loop, the RHR operable immediately, system LPCI mode and the diesel generators are operable.

3. Shutdown Requirements If Specification'3.5.A.1.a or '

3.5.A.2 cannot be met, the reactor shall be placed in'the Ccid Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. Residual Heat Removal (RHR) B. Residual Heat Removal (RHR)

System (LPCI and Containment System (LPCI and Containment ~

Cooling Mode) Cooling Mode)

1. Normal System Availability 1. Normal Operational Test RHR system testing shall be performed as follows:

Item Frequency a.- The RHR System shall be operable: a. Air test on Once/5 years drywell head-(1) Prior to reactor startup ers and-nozzles from a cold condition, or and air or water test on (2) When irradiated fuel is in torus headers the reactor vessel and the and nozzles reactor pressure is greater than atmospheric except as

. stated in Specification 3.5.B.2.

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LIMITIN CLONDITONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B.1 Normal System Availability (Cont'd). 4.5.B.1 Normal Operational Tests One RHR Loop with two pumps or' two Item Frequency-

' loops with one pump per loop shall be. operable-in the shutdown cooling b.. Simulated Once/ Operating mode when irradiated fuel is in the Automatic Cycle reactor vessel and the reactor pres- Actuation sure is atmospheric except prior to Test a reactor startup as stated in Speci-fication 3.5.B.1.a.

c. The reactor shall not be started up c. System flow Once/3 months with the RHR system supplying rate: Each cooling to the fuel pool. RHR pump shall deliver

d. During reactor power operation, the at least 7700 LPCI system discharge cross-tie gpm against a valve, E11-F010, shall be in the system head of closed position and the associated at least 20 psig.

valve motor starter circuit breaker shall be locked in the off position. In addition, an annunciator which indicates that the cross-tie valve is not-in the fully closed nosition shall be available in the control room.

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

2. Operation with Inoperable 2. Surveillance with Inoperable ,

Components gmponents

a. One LPCI Pump Inorerable a. One LPCI Pump Inoperable If one LPCI pump is inoperable, the When one LPCI pump is inoperable, reactor may remain in operation for the associated diesel generators a period not to exceed seven (7) shall be demonstrated to be opera-days provided that the remaining ble immediately and daily LPCI pumps, both LPCI subsystem thereafter, until the inoperable flow paths, the Core Spray System, LPCI pump is restored to normal and.the associated diesel; genera- service.

tors are operable.

b. One LPCI- Subsystem Inoperable b. One LPCI Subsystem Inoperable A LPCI subsystem is considered to When one LPCI subsystem is inopera-be. inoperable if (1) both of the ble, the associated diesel generato LPCI pumps within' that system are shall be demonstrated to be

. inoperable or (2) the active operable, immediately.

valves in the subsystem flow path '

are inoperable.

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LIMIT 1NG CONDlTIONS FOR OPERATION- SURVEILLANCE REQU1REMENTS

3. 5. B. 3 Shutdown Requirements.

If Specification 3.5.B.1.a or 3.5.B.2 cannot be met, the reactor shall be placed 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 />.

3.5.C. RHR Service Water System 4.5.C. RHR Service Water System

1. Normal-System Availability '1. Normal System Availability The RHR. service water system RHR service water system testing shall be operable: shall be performed as follows:

Item Frequency

a. Prior to reactor startup Pump Capacity After pump from a Cold Condition, or Test: maintenance Each RHR ser- ~

and once/3

b. When irradiated fuel is in vice water months the reactor vessel and the pump shall i reactor vessel pressure is deliver at I greater than atmospheric least 4000 gpm pressure except as stated in at a system head Specification 3.5.C.2. of at least 938 feet,

c. When irradiated fuel is in the reactor vessel and the reactor is depressurized at least one RHR service water loop shall be '

operable.

2. One Pump Inoperable If one RHR service water pump is inoperable the reactor may remain in operation i for a period not to exceed. j thrity (30) days provided all ]

other active components of both i subsystems are operable. l l

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LIMIT 1f(G CONDlTONS FOR OPERAT10N' SURVEILLANCE REOUIREMENTS 3.5.C.3. . Two Pumps Inoperable 4.5.C.3. Two Pumps Inoperable If two RHR; service water pumps are- When two RHR service water pumps

. inoperable, the reactor may re- are inoperable, the diesel gen-main in operation for a period erators associated with the re-not to exceed seven (7) days pro- maining operable RHR service water vided all redundant active com- subsystems shall be demonstrated .

ponents in both of the RHR to be operable immediately and service water subsystems are daily thereafter for seven (7) operable. days or.until the inoperable components are returned to normal operation.

4. Shutdown Requirements If Specifications 3.5.C ,

cannot be met, the reactor-shall be placed in the Cold Shut-down Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. High Pressure Coolant Injection D. High Pressure Coolant Injection (HPCI) System (HPCI) System

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

Item Frequency

a. The HPCI System shall be a. Simulated Or.ce/ Operating operable: Automatic Cycle Actuation (1) Prior to reactor startup Test from a cold condition, or  ;

1 (2) when irradiated fuel is in b. Flow rate at Once/3 months I the reactor vessel and the normal reactor reactor pressure is greater vessel oper-than 113 psig, except as ating pressure, stated in Specification and 3.5.D.2. Flow rate at Once/ Operating 150 psig Cycle i reactor i pressure l I

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7-LIMITING CONDITIONS FOR OPERATION SURVE1LLA' ICE REQUIREMENTS 4.5.D.1.b. Normal Operational Tests (Continued)

The HPCI pumps shall deliver at-least 4250 gpm during each flow rate test, e

3.5.D.2. Operation with Inoperable 2. Surveillance with inoperable-Components Components  !

If the HPCI system is inoperable, When the HPCI system is in- '

the reactor may remain in. opera- operable, the ADS actuation L tion for a period not to exceed logic shall be demonstrated seven-(7)' days provided the ADS, to be operable immediately. '

-SC system, RHR system LPCI mode, The ADS. logic shall be and SCIC system are operable. demonstrated to be operable daily thereafter until the HPCI system is returned to normal operation. ]

3. Shutdown Requirements If Specification 3.5.D.1 or 3.5.D.2 cannot be met, an ,

orderly shutdown shall be initiated and the reactor '

vessel ' pressure shall be re .

duced to 113 psig or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. i j

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

1. Normal System Availability 1. Normal Operational Tests l

RCIC system testing shall be ~

performed as follows: '

Item Frequency

a. ,The RCIC System shall be operable: a. Simulated Once/ Operating Automated Cycle  !

(1) Prior to reactor startup Actuation l from a cold condition, or Test I i

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LIMITING CONDITIONS FOR OPERATION. SURVEILLANCE REQUIREMENTS 3.5.E.1 Normal System' Availability (Cont'd) 4.5.E.1 Normal Operational Tests (Cont'd) a.(2)whenthere'isirradiated ,

fuel in the reactor vessel and the reactor pressure is above 113 psig, except as stated in Specification 3.5.E.2.

Item Frequency

b. Flow rate at

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Once/3 months normal reactor vessel oper-ating p~ressure ,

and Flow rate at Once/ Operating 150 psig Cycle reactor pressure 1 The RCIC pump shall deliver at least 400 gpm during each flow test. ,

2. Operation with Inoperable t Components

• If the RCIC system is inoperable, the reactor may remain in opera- +

tion for a period not to exceed seven (7) days if the HPCI system is operable during such time.

3. Shutdown Requirements  ;

If Specification 3.5.E.1 or 3.5.E.2 is not met, an orderly shutdown shall be initiated and the reactor shall be depressurized .

to less than 113 psig within ,

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. LIATTING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 2

3. 5. G. - Minimum 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 of the standby diesel generators is inoperable, continued reactor is inoperable, the remaining die-

' operation is limited to seven (7) sels shall be demonstrated to be days unless operability of the operable immediately and daily diesel generator is restored thereafter. '

within this period. During such seven (7) days all of the com-ponents in the RHR system LPCI mode and containment cooling  ;

mode shall be operable. If this requirement 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 incperable components in the core and con-tainment cooling systems shall not defeat the capability of the remaining operable components to fulfil'. 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 containment cooling subsystems of the RHR system may be inoperable provided that the

. shutdown cooling subsystem of the RHR system is operable in accordance with Specification 3.5.B.l.b and that no work 4 is being done which has the potential .

i for draining the reactor vessel.

H. Maintenance of Filled Discharge H. ~ Maintenance of Filled Discharge '

Pipes- Pipes Whenever the core spray system, The following surveillance re-LPCI, HPCI, or RCIC are required quirements shall be performed j to be operable,'the discharge pip- to assure that the discharge i ing from the pump discharge of these piping of the core spray sys- i systems to the last block valve tem, LPCI, HPCI, and RCIC are ,

shall be filled. The sur. tion of the filled when required- I HPCI pureps shall be aligned to the condensate storage tank. 1. Every month prior to the testing of the LPCI and i core spray systems, the dis-charge piping of these sys-tems shall be vented.

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Lit 1ITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.S.J. Plant Service Water System 4.5.J Plant Service Water System j

.l . Normal Availability 1. The automatic pump start func-tions and automatic isolation

The reactor shall not be functions shall be tested once ,

made critical'from the per operating cycle.  !

cold shutdown condition unless the Plant Service Water System (including 3 .

plant service water pumps '

and the standby service water pump) is operable.

2. Inoperable Components 2. Inoperable _ Components ,

a. The standby service water a. 'When the standby service pump may be inoperable for water pump is made or found a period not to exceed 60 to be inoperable, .all three days provided all diesel diesel generators shall be generators are operable. demonstrated to be operable immediately.

b. One PSW pump may be in- b. When one PSW pump is made operable for a period not or found to be inoperable, 1

to exceed 30 days provided all diesel generators all diesel generators associated associated with the opera-with the operable PSW components ble PSW components shall are operable. be demonstrated to be operable immediately and weekly thereafter.

c. One PSW pump and the stand- c. When one PSW pump and the l by service water pump may standby service water pump  :

be inoperable for a period are made or found to be not to exceed 30 days pro- inoperable, all diesel vided all diesel generators generators associated with associated with the operable the operable PSW components PSW components are operable. shall be demonstrated to to be operable immediately and weekly thereafter.

l

d. Two PSW pumps or one PSW d. When two PSW pumps or one I division may be inoperable for PSW division are made or l a period not to exceed 7 found to be inoperable. i days provided the diesel the diesel generators asso-  !

generators associated with the ciated with the operable i operable PSW components are PSW components shall be 1 operable. demonstrated to be operable l immediately and daily there-after. j l

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LIMITING-CONDITIONS FOR OPERAT ON SURVEILLANCE REQUIREMENTS-3.'5 J " Plant Service Water System 4.5.J Plant Service Water System

2. Inoperable Components (Cont'd) 2. Inoperable Components (Cont'd)

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

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

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

3.5.K Equipment Area Coolers 4.5.K Equipment Area Coolers

1. The equipment area coolers 1. Each equipment cooler servino the Reactor Core Iso- is operated in conjunction lation Cooling (RCIC), High with the equipment served by Pressure Coolant Injection that particular cooler; (HPCI), Core Spray or Residual therefore, the equipment area ,

Heat Removal (RHR) pumps must coolers are tested at the i be operable at all times when same frequency as the pumps

-the pump or pumps served by which they serve. l that specific cooler is con-  !

sidered to be operable. 1 1

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. BASES FOR LIMITING CONDIT0NS FOR OPERATION AND SURVEILLANCE REQUIREMEfHS 3.5 CORE AND CONTAINMENT COOLING SYSTEMS A. ' Core Spray-(CS) System

1. Normal System Availability Analyses presented in Section 6 of the FSAR and Appendix I of the HNP PSAR demonstrated that the core spray system provides adequate cooling to the core to dissipate the energy associated with the loss-of-coolant accident and to limit fuel clad temperature to below 2,300 F which assures that core geometry r o ains intact and to limit any clad metal-water reaction to less than one percent. Corc spray distribution has been shown in tests of systems similar in design to HNP 1 to exceed the minimum requirements.

In addition, cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiated fuel.

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

However, during operation, certain components may be out of service for the specified allowable repair times. The allowable repair times have been selected using engineering judgment based on experience and supported by availability analysis. Assurance of the availability of the remaining systems is increased by demonstrating operability immediately and by requiring selected testing during the outage period.

When the reactor vessel pressure is atmospheric, the limiting conditions for operation are less restrictive. At atmospheric pressure, the minimum requirement is for one supply of makeup water to the core. Requiring two operable RHR pumps and one CS pump provides redundancy to ensure makeup water availability.

2. Operation with Inoperable Components Should one core spray loop become inoperable, the diesel generators are demonstrated to be operable to ensure their availability should the need for core cooling arise. The surveillance testing required by Specifications 4.6.K, 4.5.A, and 4.5.H ensures the availability of the remaining core spray loop. The surveillance testing required by Specifications 4.6.K.

4.5.B, 4.5.C, and 4.5.H ensures the availability of the RHR sy, stem. These provide extensive margin over the operable equipment needed for adequate core cooling. With due regard for this margin, the allowable repair time of 7 days was chosen.

r B. Residual Heat Removal (RHR)~ System (LPCI and Containment Cooling Mode)

1. Normal System Availability +

The RHR 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 completely independent of the core spray system; however, it does function in combination with the core spray system to prevent excessive fuel clad temperature. The LPCI mode of the RHR system and the core spray system provide adequate cooling for b W ak areas of approximately 0.2 square feet up to and including the double-ended recirculation line break without assistance from the high-pressure emergency core cooling systems.

3.5-14

BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.5.8.1. Normal System Availability (Continued)

Observation of the stated requirements for the containment'coolino mode assures ,

that the suppression pool and the drywell will be sufficiently cooled, follow-ing a~ loss-of-coolant accident, to prevent primary containment overpressur iza-tion. The containment cooling function of the RHR system is permitted only after the core has 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 manually bypassed by a keylock switch.

The intent of the RHR. system Gpecifications is to prevent operation above atmospheric pressure without all associated equipment being operable. How-ever, during operation, certain components may be out of service for the specified allowable repair times., The allowable repair times have been selected using engineering judgment based on experiences and supported by availability analysis. Assurance of the availability of the remaining systems is increased by demonstrating operability immediately and by requiring selected testing during the outage period.

When the reactor vessel pressure is atmospheric, the limitng conditions for operation are less restrictive. At atmospheric pressure, the minimum require- ,

ment is for one supply of makeup water to the core.

2. Operation with Inoperable Components With one LPCI pump inoperable or one LPCI subsystem inoperable, adequate core flooding is assured. The surveillance testing required by Specifications 4.6.K 4.5.B, 4.5.C, and 4.5.H ensures the availability of the redundant LPCI pumps and LPCI subsystem. The surveillance testing required by Specifications 4.6.K, 4.5.A, and 4.5.H ensures the availability of the Core Spray System. In addition, the associated diesel generators are demonstrated to be operable.

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4 3.5-15 b

u BASES FOR LIMITING CONDITXONS FOTOPERATION AND SURVEILLANCE REQUIREMENTS 3.5.D.2. Operation With Inoperable Components ,

The HPCI system serves as a backup to the RCIC system as a source of feedwater makeup during primary system isolatiun conditions. The ADS

-serves as a backup to the HPCI system for reactor depressurization for postulated transients and accidents. 'The ADS is checked for operability if the HPCI system is determined to be inoperable. In addition, the surveillance testing required by Specifications 4.6.K, 4.5.E, and 4.5.H ensures the availability of the RCIC system. Considering the redundant systems, an allowable repair time 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 ensure that the function for which the e RCIC system was designed will be available when needed.

Because the low-pressure cooling systems (LPCI and core spray) are capable of providing all the cooling required for any plant even when nuclear system pressure is below 113 psig, the RCIC system is not required below this pressure. Between 113 psig and 150 psig the RCIC system need not provide its design flow, but reduced flow is required for certain events. RCIC system design flow (400 gpm) is sufficient to maintain water level above the top of the active fuel for a complete loss of feedwater flow at the design power.

2. Operation With Inoperable Components Consideration of the availability of the RCIC system reveals that the average risk associated with failure of the RCIC system to cool the core when required is not increased if the RCIC system is inoperable for no longer than seven (7) days, provided that the HPCI system is operable.

The surveillance testing required by Specifications 4.6.K, 4.5.D, and 4.5.H ensures the availability of the HPCI system.

F. Automatic Depressurization System (ADS)

1. Normal System Availability This specification ensures the operability of the ADS under all con-ditions for which the depressurization of the nuclear system is an l essential response to Unit abnormalities. I The nuclear system pressure relief system provides automatic nuclear system depressurization for small breaks in the nuclear system so that the low-pressure coolant injection (LPCI) and the core spray system can operate to protect the fission product barrier. Note that this Specifica- 1 tion applies only to the automatic feature of the pressure relief system. j i

3.5-17

L , BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS I'

l 3.5.F.1. Normal System Availability.(Continued) l.

Specification 3.6 states the requirements for the pressure relief function of the valves. It is possible for any number of the valves assigned to the ADS to be incapable of performing their- ADS functions because of instrumentation failures yet be fully capable _ of performing their pressure relief function.

Because the automatic depressurization system does not provide makeup to the reactor primary vessel, no credit is taken for thesteam coolir.g of the core caused by the system actuation to provide further conservatism to the Core Standby Cooling Systems.

2. Operation with Inoperable-Components With one ADS valve known to be incapable of automatic operation six valves remain operable to perform their ADS function. However, since the ECCS Loss of Cooland Accident analysis for small line breaks assumed that all seven ADS valves were operable, reactor operation with one ADS valve inoperable is only allowed to continue for seven (7) days provided that the actuation logic for the (remaining) six ADS valves is demonstrated 3 to be operable. The surveillance testing required by Specifications 4.6.K, 4.5.D, and 4.5.H ensures the availability of the HPCI system.

3. Minimum Core and Containment Cooling Systems Availability The purpose of this specification is to assure that adequate core cooling equipment is available at all times. If, for example, one core spray loop were out of service and the diesel which powered the opposite core spray were out of service, only 2 RHR pumps would be available. Specification  ;

3.9 must also be consulted to determine other requirements for the diesel  !

generators. In addition, refer to definition 1.0.00 for Cumulative Downtime requirements.

1 This specification establishes conditions for the performance of major i maintenance, such as draining of the suppression pool. The availability I of the shutdown cooling subsystem of the RHR system and the RHR service water system ensure adequate supplies of reactor cooling and emergency makeup water when the reactor is in the Cold Shutdown condition. In addition, this specification provides that, should major maintenance be performed, no work will be performed which could lead tb draining the water from the reactor vessel.

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. BASES FOR LlMlTING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS )

1 3.5.J/4.5.J' Plant' Service Water System  ;

1 The Plant Service Water (PSW) system consists of two subsystems (division) of two pumps each and a separate standby service water pump system for diesel generator 18. During ,

normal full power operation'the two subsystems function as a 3 out of 4 pump cross 1 connected system supplying cooling water to the turbine and reactor building cooling "

I systems. In the event of an accident signal, non safety-related cooling loads are iso-  !

lated and the PSW pumps in~'the two subsystems supply cooling water to diesel generators  !

-1A and 1C, the reactor building cooling system and the control room air conditioners, j while the standby service water pump is available to automatically supply cooling water to diesel generator 1B should it be needed. Additionally, diesel 1B has a manual back- '

up water supply available from the Unit 1 Division 1 or Division 2 PSW subsystems so . .

that during maintenance on. the standby diesel service water pump, either division of the PWS system can manually be aligned to supply cooling water to the 1B diesel. The two subsystems and the standby service water pump system are split in the accident mode for greater reliability with one pump in each of the two subsystems automatically starting while a start signal from diesel generator 1B initiates standby service water pump operation. Only one'of the Division 1 PSW pumps and one of the Division 2 PSW pumps are  ;

required for cooling diesel generators lA and 1C, respectively, while the standby ser-vice water pump provides adequate cooling water to diesel generator 18. 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 PSW pump is capable of supplying the cooling requirements for the reactor building cooling system, the control room air conditioners, and the 1 A,18, and 1C diesel generators.

The PSW system can supply all power generation systems at full load and the diesel ,

generators with redundancy if one PSW pump and/or the standby service water pump are inoperable. Hence, a 60-day outage time is justified if the standby service water pump is inoperabla since all four PSW pumps are available (divisional intertie and 1B diesel required). In addition, a 30-day outage is justified if one PSW pump is inoperable, or if one to PSW required).

1B diesel pump and the standby Should service two PSW pumpswater p(ump or one are inoperable subsystem) (divisional become inoperable, or intertie inoperable (division intertie to 1B diesal required) plant operation will probably only <

continue at less than full power. However, safety-related loads are still adequately p0wered for these conditions. Therefore, a 7 day outage time is justified for such events.

The surveillance testir.g required by Specification 4.6.K will provide adequate assur-ance that the PSW system will be operable when required.

K. Engineering Safety Features Equipment Area Coolers The equipment area cooler in each pump compartment is capable of providing adequate ventilation flow and cooling. Engineering analyses indicate that the temperature rise in safeguard compartments without adequate ventilation flow or cooling is such that continued operation of the safeguard equipment or associated auxiliary equip-ment cannot be assured.

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

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

L. References- r 1.. FSAR Section 6, Core Standby Cooling System.

2. HNP-2 PSAR Appendix I, Conformance to NRC Interim Acceptance Criteria for  ;

Emergency Core Cnoling. Systems. >

3.5-21

'f LIMITTWCONDITIdiiS FOR OPERATION SURVEILLANCE REQUIREMENTS

4. 6. I' -Jet Pumps'(Continued)

2. The indicated value of core 3 flow rate varies from the value derived from loop flow measurements by more than 10%.

, 3. The diffuser to lower' plenum differential pressure reading on an individual jet pump vary from the mean of all jet pump differential pressures by more than 10%.

P 3.6.J. Recirculation Pump Speeds 4.6.J. Recirculation Pump Speeds

1. Core thermal power shall not Recirculation pump speeds shall be exceed 1% of rated thermal recorded at least once per day, power without forced .recircula-tion. .i

2. Operatioc with a single recircu-lation pump is permitted for 24 ,

hours unless the recirculation pump is sooner made operable.

If the pump cannot be made operable, the reactor shall be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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

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. LIMITING CONDITIONS FOR OPERATION SliRYCID_ANC ' RE0VIREM5NTS 3.6.K. . STRUCTURAL INTEGRITY 4.6.K. STRUCTURAL INTEGRITY

1. Normal Codition l Surveillance Requirements.for in-service

. inspection and testing of ASME Code The structural integrity of ASME Class 1, 2, and 3 (equivalent) components Code Class i, 2, and 3 (equiva- shall be applicable as follows:

lent)_ components .shall be main-tained in accordance with the 1. In-service inspection ASME Code Surveillance Requirements of Class 1, 2, and 3, (equivalent) com-Specification'4.6.K. ponents and in-service testing of ASME Code Class 1, 2, and 3 (equi-

2. Off-Nomal Conditions valent) pumps and valves shall be performed in a(cordance with Section

a. With the structural integrity XI of the ASME Boiler and Pressure of any ASME Code Class 1 com- Vessel Code and applicable ponent not conforming to the- Addenda as required by 10CFR50, above requirements, restore the Section 50.55a(g), except where structural integrity of the specific written relief has been affectedcomponent(s)towithin granted by the Commission pur-its limit or isolate the affected suant to 10CFR50, Section 50.55a(g) component (s) prior to increasing (6)(i).

the Reactor Coolant Systen tem-perature more thean 500F above the 2. Performance of the above in-minimum temperature required by service inspection and testing NDT considerations. activities shall be in addition to other specified Surveillance

b. With the structural integrity of Requirements.

any ASME Code Class 2 component (s) not confoming to the above re- 3. Nothing in the ASME Boiler and quirements, restore the structural Pressure Vessel Code shall be integrity of the affected com- construed to supersede the ponent(s) to within its limit or requirements of any Technical isolate the affected component (s) Specification.

prior to increasing the Reactor Coolant System temperature above 4. The results obtained from com-pliance with this specification

c. With the structural integrity of will be evaluated after 5 years any ASME Code Class 3 component (s) and the conclusions of this not conforming to the above re- evaluation will be reviewed with quirements, restore the structural the NRC.

integrity of the affected compon-ent(s) to within its limit or iso-late the affected component (s) from service.

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f BASES FOR LIMITING CONDITIONS OF OPERATION AND SURVEILLANCE REQUIREMENTS

3. 6. K . Strutural Integrity In-service' inspection of ASME Code Class .1, 2, and 3 (equivalent) components and in-service testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with a periodic-ally updated version.of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10CFR50.55a(g). This objective will maintain the structural integrity of safety-related components, pumps, and valves which are necessary to safety shut-down the plant or mitigate the consequences of an accident.

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3.6-24 thru 3.6-30 i

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p LIMITING CONDITIONS FOR OPERATION $DEVEILLANCE RFAUIREMENTS

' ~

4.7.C.l. Surveillance While integrity ')

Maintained (Continued)

b. Secondary containment capabil- '

ity to maintain a minimum 1/4 -

inch of water vacuum under calm wind (< 5 mph) conditions with a filter train flw rate of not more than 4000 cim, '

shall be demonstrated at each' '

refueling outage prior to refueling.

3.7.C.2. Violation of Secondary 2. Surveillance After Integrity Containment Integrity Violated If Specification 3.7.C.1 cannot After a secondary containment be met, procedures shall be violation is determined the initiated to establish conditions standby gas treatment system listed in Specification 3.7.C.l.a will be operated immediately through 3.7.C.1.d. after the affected zones are-isolated from the remainder of.

the secondary containment. The ability to maintain the re-mainder of the secondary con-tainment at 1/4-inch of water vacuum pressure under calm

(< 5 mph) wind conditions shall be confirmed.

D. Primary Containment Isolation Valves D. Primary Containment Isolation Valves

1. Valves Required to be Operable 1. Surveillance of Operable Valves I During reactor power operation, Surveillance of the primary all primary containment isolation containment isolation valves valves listed in Table 3.7-1. shall be performed as follows:  ;

and all reactor coolant system instrument line excess flow a. At least once per operating check valves shall be operable cycle the operable isolation ,

except as stated in Specification valves specified in Table l 3.7.D.2. 3.7-1 that are automatically i initiated shall be tested fcr l simulated automat 1c initiation.

b. As required by Specification 4.6.K, the closure times of '

the operable isolation valves specified in Table 3.7-1 that are power operated shall be determined to be within their limits.

3.7-13

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v. -lib 11 TING CONDITIONS FOR OPERAT10N, SURVEILLKffCE REQUIREMENTS 4.7.D.l. Surveillance of Operable Valves (Contiritied)

c. At least once per operating cycle the reactor coolant system instrument line excess flow check valves shall be tested for proper operation.

d. As required by Specification 4.6.K, all normally open '

erated isolation power-op(except valves for the main steam line power-operated isolation valves) shall be fully closed and reopened.

e. At least once per quarter, with the reactor power less than 75% of rated, the main steam line isolation valves shall be tripped (one at a time) and closure time verified.

f. At least once per week the main steam line power-oper-ated isolation valves shall be exercised one at a time by partial closure and sub-sequent reopening.

3.7.D.2. Operation with Inoperable Valves 2. Surveillance of Lines with an Inoperable Valve In the event any isolation valve Whenever an isolation valve I specified in Table 3.7-1 becomes listed in Table 3.7-1 is  !

incoerable- reactor power operation inoperable the position of.at  ;

least one other isolation valve may continue provided.at least once isolation valve in each line having in each line having an inoperable i an inoperable valve is in the mode isolation valve shall be verified corresponding to' the isolation con- to be in its isolated position uition. daily.

3. Shutdown Requirements j If Specification 3.7.D.1 and 3.7.D.2 cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the Cold Shut-down Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

)

3.7-14 I _ -