Text

Proposed Tech Specs Re RHR Sys
	ML20043H124
Person / Time
Site:	North Anna
Issue date:	06/13/1990
From:	; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML20043H122	List: ML20043H121; ML20043H124;
References
	NUDOCS 9006220037
	Download: ML20043H124 (26)
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ATTACHMENT ONE .

i PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS RESIDUAL HEAT REMOVAL SYSTEM ,

NORTH ANNA POWER STATION :

UNITS 1 AND 2 l'- !

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> 4 'o REACTOR COOLANT SYSTEM SHUTDOWN l SURVElLLANCE REQUIREMENTS 4.4.1.3.1 The required RHR subsystems shall be demonstrated OPERABLE por Specification 4.7.9.2.

4.4.1.3.2 The required reactor coolant pump (s), it not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignment and indicated power availability.

4.4.1.3.3 The required steam generator (s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 17% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.3.4 At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , verify at least one coolant loop to be in operation and circulating reactor coolant by:

a. Verifying at least one Reactor Coolant Pump is in operation.

or

b. Verifying at least one RHR Loop is in operation and,

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1. If the RCS temperature >140' F or the time since entry into MODE 3 is

<100 hours, circulating reactor coolant at a flow rate 23000 ppm.

or

2. if the RCS temperature 5140' F and the time since entry into MODE 3 is 2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months , circulating reactor coolant at a flow rate 22000 gpm to remove decay heat.

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REFUELING. OPERATIONS 3/4.9.8 RESIDUAL HE AT REMOVAL (RHR) AND COOLANT CIRCULATION l ;

NORMAL WATFR LEVEL l .

LIMITING CONDITIONS FOR OPERATION 3.9.8.1 At least one RHR loop shall be OPERABLE

and at least one RHR loop shall be in operation.

APPLICABILITY: MODE 6 With the reactor vessel water level greater than or equal to 23 feet above the top of the reactor pressure vessel flange.

ACTlON' a. With less than one RHR loop OPERABLE,immediately initiate corrective

actions to retum the required RHR loops to OPERABLE status as soon as possible.

b. With less than one RHR loop in operation, except as provided in c. below. l suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close "

all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

c. The RHR loop may be removed f rom operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months l penod during the performance of CORE ALTURATIONS in the vicinity I of the reactor pressure vessel hot legs.

d. The provisions of Specification 3.0.3 are not applicable. l SURVEILL ANCE REQUIREMENTS 4.9.8.1.1 Verify the required RHR loop to be OPERABLE per Specification 4.0.5. l 4.9.8.1.2 At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , verify at least one RHR Loop is in operation and,

a. If the RCS temperature >140' F or the time since entry into MODE 3 is

<100 hours, circulating reactor coolant at a flow rate 23000 gpm.

b ll the RCS temperature 5140' F and the time since entry into MODE 3 is 2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months , circulating reactor coolant at a flow rate 22000 ppm to remove decay.

The normal or emergency power source may be inoperable for each RHR loop.

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FIEFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION l LOW WATER LEVELS LIMITING CONDITION FOR OPERATION r 3.9.8.2 Two independent RHR loops shall be OPERABLE' with at least one loop in operation.

APPLIC ABILITY: MODE 6 with the reactor vessel water levelless than 23 f eet above the top of the reactor pressure vesselflange.

A Q J 10 E l a. With less than the required RHR loops OPERABLE, immediately initiate corrective actions to return the required RHR loops to OPERABLE status as soon as possible,

b. With less than one RHR loop in operation, suspend all operations involving an increase in the reactor decay heat load of a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

c. The provisions of Specification 3.0.3 are not applicable. l SURVEIL L ANCE REQUIREMENTS 3 4.9.8.2.1 Verify the required RHR loops to be OPERABLE por Specification 4.0.5. l 4.9.8.2.2 At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , verify at least one RHR Loop is in operation and,

a. If the RCS temperature >140* F or the time since entry into MODE 3 is

<100 hours, circulating reactor coolant at a flow rate 23000 gpm.

b ll the RCS temperature 5140' F and the time since entry into MODE 3 is 2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months , circulating reactor coolant at a flow rate 22000 gpm to remove decay.

The normal or emergency power source may be inoperable for each RHR loop.

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3/4.4 RE ACTOR COOLANT SYSTEM BASES i 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normat operations arxl anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least .

HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , '

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

In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat, but single f ailure considerations require that at least two loops be OPERABLE. Thus,if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE. >

After the reactor has shutdown and entered into MODE 3 for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , a minimum RHR system flow rate of 2000 gpm in MODE 5 is permitted, provided there is sufficient decay heat removal to maintain the RCS temperature less than or equal to 140'F Since the decay heat power production rate decreases with time after reactor shutdown, the requirements for RHR system decay heat removal also decrease. Adequate decay heat removalls provided as long as the reactor has been shutdown for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after entry into MODE 3 and RHR flow is sufficient to maintain the RCS temperature less than or equal to 140'F. The reduced flow rate provides additional margin to vortexing at the RHR pump suction while in Mid Loop Operation.

During a reduction in reactor coolant system boron concentration the Specification 3.1.1.3.1 requirement to maintain a 3000 gpm flow rate provides sufficient coolant circulation to minimize the effect of a boron d:lution incident and to prevent boron stratification.

The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 324'F are provided to prevent RCS pressure transients, caused by energy additions f rom the secondary system which could exceed the limits of Appendix G to 10 CFR Part

50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPS to when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures.

The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification, and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The requirement to maintain the boron concentration of an isolated loop greater than or equal to the boron concentration of the operating loops ensures that no reactivity addition to the core could occur during stanup of an isotated loop. Venfication of the boron concentration in an l

idle loop prior to opening the cold leg stop valve provides a reassurance of the adequacy of the boron concentration in the isolated loop. Operating the isolated loop on recirculating flow for at least 90 minutes prior to opening its cold leg stop valve ensures adequate mixing of the coolant in this loop and prevents any reactivity effects due to boron concentration stratification.

l Startup of an idle loop will inject cool water from the loop into the core. The reactivity l transient resulting from this cool water injection is minimized by delaying isolated loop startup until its temperature is NORTH ANNA- UNIT 1 B 3/4/ 41

.* s REFUELING OPERATIONS BASES 3/4.9.6 MANIPULATOR CRANE OPERABILITY The OPERABILITY requirements for the manipulator cranes ensure that: 1) manipulator cranes will be used for movement of control rodt and fuel assemblies,2) each crane has sufficient load capacity to I:f t a control rod or f uel assembly, and 3) the core intomals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL SPENT FUEt_ PIT The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped,1) the activity release shall be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuelin the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident '

analyses, t 3/4.9.8 RESIDUAL HE AT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140'F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effe1 of a boron dilution incident and prevent boron stratification.

After the reactor has shutdown and entered into MODE 3 for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , a minimum RHR system flow rate of 2000 gpm in MODE 6 is permitted, provided there is sufficient decay heat removal to maintain the RCS temperature less than or equal to 140*F. Since the decay heat power production rate decreases with time after reactor shutdown, the requirements for RHR system decay heat removal also decrease. Adequate decay heat removalls provided as long as the reactor has been shutdown for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after entry into MODE 3 and RHR flow is sufficient to maintain the RCS temperature less than or equal to 140*F. The reduced flow rate provides additional mar 0i n to vortexing at the RHR pump suction while in Mid Loop Operation.

During a reduction in reactor coolant system boron concentration the Specification 3.1.1.3.1 requirement to maintain a 3000 ppm flow rate provides sufficient coolant circulation to minimize the effect of a boron dilution incident and to prevent boron stratification. ;

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vesselflange ensures that a single f ailure of the operating RHR loop will not result in a complete loss of residual heat removal capabiGly. With the reactor vessel head removed and 23 f eet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus,in the event of a f ailure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

3/4.9.9 CONTAINMENT PURGE AND EXHAUST ISOL ATION SYSTEM The OPERABILITY of this system ensures that the containment vont and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

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1 INDEX j i

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAQR )

i 34 0 A P P L I C AB i L I T Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 0 1 1 341 REACTIVITY CONTROL SYSTEMS 1 3/4.1.1 BORATION CONTROL i

i Shutdown Margin Tavg > 200* F...... ............. .... ....... ................. .......... 3/4 1 1 !

Shutdown M argin Tavg s: 200* F.... . .............................................. ...... ... 3/4 1 3 ;

Boron Dilution Re actor Coola nt Flow........................... ............. . .............. 3/4 1 4 Boron Dilution Valve Position.. .. ....................................... .. ................. 3/4 1 4a Moderator Temperature Coellicient................................. . . . . . . . . . . . . . . . . .. . . . . . 3 /4 1 5 Minimum Te mpe rature f or Criticality................................. ............ ... .......... 3/4 1 7 f

i 3/4.1.2 BORATION SYSTEMS f

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F low Pa t h s S h u tdown . . . . . .. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . .. . . 3/4 1 8 ,

Flo w Pa t h s Ope rating . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . .. .. . . 3/4 1 9 Ch a rging Pu mps S h u tdow n... . ...... ....... ............ .... ........... .... ................ 3/4 1 11 Ch arging Pu mps Ope rating............. . ..... ... .. ................... . .......... ............ 3/4 1 12 Borated Water Sources Shutdown..... ........... ...... .................................. 3/4 1 13 Borated Water Sources Operating...... .., ...... . ........... .... .... .. ... ... ..... 3/41 14 r

3/4.1.3 MOVABLE CONTROL ASSEMBLIES !

G ro u p H eig bl . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . 3/4 1 16 Position Indicalor Channels Operating...... ......... .......... ........... ........ .... 3/4.1 19 Position Indicator Channels Shutdown.... ....... ........................................ 3/4 1 20 :

R o d D ro p Ti m e . .. . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1 21 Shutdown Rod inse rtion Limits.............................. ..... ... ...... ............... 3/4 1 22

. Control Rod Insertion Limits.. .. .. .. .... . . . . . . . . . . . . . . . . . . . , . . . . . . ... .. 3/4123 l - NORTH ANNA UN T 2 III l&

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REACTIVITY CONTROL SYSTEM 3 /4.1.1. 3 BOROW O!LOTlQB REACTOR COOLANT FLQM LIMITING CONDITIONS FOR OPERATION 3.1.1.3.1 The flow rate of reactor coolant through the reactor coolant system shall be 23000 gpm whenever a reduction in Reactor Coolant System boron concentration is being made. ,

APPLICABILITY: All MODES I

ACTION. With the flow rate of reactor coolant through the reactor coolant system <3000 gpm, immediately suspend all operations involving a reduction in boron concentration of the Reactor Coolant System.

SURVEILL ANCE REQUIREMENTS 4.1.1.3.1 The flow of reactor coolant through the reactor coolant system shall be determined to be 2 3000 gam within one hour prior to the start of and at least once per hour during a reduction in the Reactor Coolant System boron concentration by either: !

a. Verifying at least one reactor coolant pump is in operation, '

or

b. Verifying that at least one RHR pump is in operation and supplying 23000 gpm ,

through the reactor coolant system.

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REACTIVITY CONTROL SYSTEM l BORON DILUTION >

VALVE POSITION LIMITING CONDITIONS FOR OPERATION 3.1.1.3.2 The following valves shall be locked, sealed or otherwise secured in the closed position except during planned boron dilution or makeup activities: l

a. 2 CH-140 -

or

b. 2 CH 160, 2 CH 156, FCV-2114B and FCV 2113B APPLIC A BILITY: MODES 3,4,5, and 6 ACTION: With the above valves not locked, sealed, or otherwise secured in the closed position:

1) suspend all operations involving positive reactivity changes or CORE ALTERATIONS,

2) lock, seal or otherwise secure the valves in the closed position within 15 minutes, and

3) verify the SHUTDOWN MARGIN is greater than or equal to 1.77% delta k/k within 60 minutes.

P SURVEILL ANCE REQUIREMENTS 4.1.1.3.2 The above listed valves shall be verified to be locked, sealed or otherwise secured l In the closed position within 15 minutes after a planned boron dilution or makeup activity.

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REACTOR COOLANT SYSTEM SHUTDOWN l SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required RHR subsystems shall be demonstrated OPERABLE PER SPECIFICATION 4,7.9.2.

4.4.1.3.2 The required reactor coolant pump (s),if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignment and indicated power avaltability.

4.4.1.3.3 The required steam generator (s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 17% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . t 4,4.1.3.4 At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . verify at least one coolant loop to be in operation and circulating reactor coolant by:

a. Verifying at least one Reactor Coolant Pump is in operation. ,

or

b. Verifying at least one RHR Loop is in operation and,

1. If the RCS temperature >140* F or the time since entry into MODE 3 is

< 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , circulating reactor coolant at a flow rate ddu00 gpm.

or

2. 11 the RCS temperature s140' F and the time since entry into MODE 3 is 2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months , circulating reactor coolant at a flow rate 22000 rpm to remove decay heat.

5 NORTH ANNA UNIT 2 3/4 4 3a

9 o-REFUELING OPE R ATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION l NORMAL WATER LEVEL l LIMITING CONDITIONS FOR OPERATION 3.9.8.1 At least one RHR loop shall be OPERABLE' and at least one RHR loop shall be in operation.

APPLICABILITY: MODE 6 With the reactor vessel water level greater than or equal to 23 feet above the top of the reactor pressure vessel flange, ACTlON: a. With less than one RHR loop OPERABLE,immediately initiate corrective actions to retum the required RHR loops to OPERABLE status as soon as possible,

b. With less than one RHR loop in operation, except as provided in c. below, l suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System. Close all containment penetrMions providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

c. The RHR loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months por 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months l period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

d. The provisions of Specification 3.0.3 are not applicable. l SURVEILL ANCE REQUIREMENTS 4.9.8.1.1 Verify the required RHR loop to be OPERABLE per Specification 4.0.5. l 4.9.8.1.2 At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , verify at least one RHR Loop is in operation and,

a. If the RCS temperature >140' F or the time since entry into MODE 3 is

<100 hours, circulating reactor coolant at a flow rate 23000 gpm.

b. If the RCS temperature 5140* F and the time since entry into MODE 3 is !

2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months , circulating reactor coolant at a flow rate 22000 gpm to remove decay heat.

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The normal or emergency power source may be inoperable for each RHR loop.

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p REFUELING OPE R ATIONS ;!

RESIDUAL HEAT REMOVAL (RHR) AND COOLANT CIRCULATION l l l

LOW WATER LEVELS LIMITING CONDITION FOR OPERATION _ _ _

3.9.8.2 Two independent RHR loops shall be OPERA 0LE' with at least one loop in operation.

4 APPLChulLITY: MODE 6 with the reactor vessel water levelless than 23 feet above -

the top of the reactor pressure vessel flange.

t A.CTIO N : a. With less than the required RHR loops OPERABLE,immediately initiate corrective actions to retum the required RHR loops to OPERABLE status as soon as possible.

b. With less than one RHR loop in operation, suspend all operations invoMng an -(

increase in the reactor decay heat load or a reduction in boron concentration of j-the Reactor Coolant System. Close all containment penetrations providing -!

direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , q

c. The provisions of Specification 3.0.3 are not applicable. l SURVEILLANCE REQUIREMENTS i

P 4.9.8.2.1 Verify the required RHR loops to be OPERABLE per Specification 4.0.5. l I

4.9.8.2.2 At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , verify at least one RHR Loop is in operation and, l a. If the RCS temperature >140* F or the time since entry into MODE 3 is

<100 hours, circulating reactor coolant at a flow rate 23000 gpm. !

b. if the RCS temperature s140' F and the time since entry into MODE 3 is 2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months , circulating reactor coolant at a flow rate 22000 gpm to ,

remove decay heat.

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The normal or emergency power source may be Inoperable for each RHR loop.

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3/14 REACTOR COOLANT SYSTEM BASES ,

3/4.4.1 REACTOR COOLANT LOOP,3 The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least -

HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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

in MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat ,

removal capability for removing decay heat, but single f ailure considerations require that at least .I two loops be OPERABLE; Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR toops to be OPERABLE.

After the reactor has shutdown and entered into MODE 3 for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , a minimum RHR system flow rate of 2000 gpm in MODE 5 is permitted, provided there is sufficient decay heat removat to maintain the RCS temperature less than or equal to 140'F. Since the decay heat power production rato decreases with time after reactor shutdown, the requirements for RHR system decay heat removal also decrease. Adequate decay heat removalis provided as long as the reactor has been shutdown for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after entry into MODE 3 and RHR flow is sufficient to maintain the RCS temperature less than or equal to 140'F. The reduced flow rate provides additional margin to vortexing at the RHR pump suction while in Mid Loop Operation.

During a reduction in reactor coolant system boron concentration the Specification 11.1.3.1 requirement to maintain a 3000 gpm flow rate provides sulficient coolant circulation to minimize the effect of a boron dilution incident and to prevent boron stratification.

l strictions on starting a Reactor Coolant Pump with one or more RCS cold legs less tar ' e aual to 324 F are provided to prevent RCS pressure transients, caused by energy Mdiu from the secondary system which could exceed the limits of Appendix G to 10 CFR Part

. - J.' TN RCS will be protected against overpressure transients and will not exceed the limits of y Ar$nendix G by *ier (1) restricting the water volume in the pressunzer and thereby providina a

, 6. : 1e for a e */ coolant to expand into or (2) by restricting starting from the RCPs to when

.5 .e y %m iemperature of each steam generator is less than 50 F above each of the 11 -

s 4 % v Oratures.

! ' ^ re n ement to maintain the boron concentration of an isolated loop greater than or q i ; N non concentration of the operating toops ensures that no reactivity addition to the core cou,a occur during startup of an isolated loop. Verification of the boron concentration in an idle loop prior to opening the cold leg stop valve provides a reassurance of the adequacy of the ;

o boron concentration in the isolated loop. Operating the isolated loop on recirculating flow for at l' least 90 minutes prior to opening its cold leg stop valve ensures adequate rnixing of the coolant in 7 thb loop and prevents any reactivity effects due to boron concentration stratification.

Startup of an idle loop will inject cool water from the loop into the core. The reactivity transient resuming from this cool water injection is minimized by delaying isolated loop startup until

l. Its temperature is within 20*F of the operating loops. Making the reactor subcritical prior to loop

l. startup prevents any power spike which could result from this cool water induced reactivity .

transient, i L NORTH ANNA - UNIT 2 B 3/4/ 4-1

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REFUELING OPERATIONS BASES' l 3/4.9.6~ MANIPULATOR CRANE OPERABILITY The OPERABILITY requirements for the manipulator cranes ensure that: 1) manipulator cranes will be used for movernc'11 of control rods and fuel assemblies,2) each crane has sufficient ;

load capacity to ill a control rou or fuel assembly, and 3) the core intemals and pressure vessel are proWW.i fr6M excessive lif ting force in the event they are inadvertently engaged during lifting got.%#or t

{Mj CHANE TRAyik SPENT FUEL PIT ,

The restriction on movers! of loads in excess of the nominal weight that of a fuel and . ,

control rod assemblies and associated fit.ndling tool over other fuel assemblies in the storage L pool ensures that in the event this load is dropped,1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuelin the storage racks will '

not result in a critical array. This assumption is consistent with the activity release assumed in the ;

accident. J

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3/4.9.8 RESIDU_AL HEAT REMOVAL AND CO.QLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation

, ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the L water in the reactor pressure vessel below 140*F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the etfect of a boron dilution incident and prevent boron stratification.

After the reactor has shutdown and entered into MODE 3 for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , a minimum RHR system flow rate of 2000 gpm in MODE 6 is permitted, provided there is sufficient ,

l decay heat removal to maintain the RCS temperature less than or equal to 140'F. Since the decay (

heat power production rate decreases with time after reactor shutdown, the requirements for RHR system decay heat removal also decrease. . Adequate decay heat removal is provided as long as the reactor has been shutdown for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after entry into MODE 3 and RHR flow is 1 sufficient to maintain the RCS temperature less than or equal to 140 F. The reduced flow rate i provides additional margin to vortexing at the RHR pump suction while in Mid Loop Operation.

During a reduction in reactor coolant system boron concentration the Specification 3.1.1.3.1 requirement to maintain e 3000 gpm flow rate provides sulficient coolant circulation to minimize the effect of a boron dilu'4on incident and to prevent boron stratification. +

1 The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange ensures that a single failure of the operating RHR

loop will not result in a complete loss of residual heat removal capabiity. With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is aval:able for core cooling. Thus,in the event of a f ailure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

3/4.9.9 CONTAINMENT + PURGE AND EXH AUST ISOL ATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge -

penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

NORTH ANNA UNIT 2 8 3/4 9-2

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i ATTACHMENT TWO DISCUSSION AND SAFETY EVALUATION FOR RESIDUAL HEAT REMOVAL

TECHNICAL SPECIFICATIONS CHANGES

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l Virginia Electric and Power Company

o * . sN 90 340A PROPOSED RHR Ts CHANGE NAPS UNITS 1 AND2 PROPOSED TECHNICAL SPECIFICATION CHANGES:

1 : T.S. 3/4.1.1.3.1 REACTIVITY CONTROL SYSTEM BORON DILUTION - REACTOR COOLANT FLOW (Unit 2 only), is being added to the Unit 2 Technical Specifications to insure that adequate mixing will be provided during Boron Concentration reductions.

'The requirement established in this specification are the same as those in the Unit 1 Technical Specifications.

' DISCUSSION The addition of this specification to the Unit 2 Technical Specifications requiring at least a minimum flow of 3000 gpm through the reactor coolant system whenever a reduction in Reactor Coolant System boron concentration is being made, is consistent with the existent Unit 1 specification and Standard Technical Specifications. Also this restriction is specifically cited in the' Basis to Unit 2 Specification 3/4.1.1.3 BORON DILUTION as being effected by specification 3/4.4.1 which requires at least either one RHR pump or one Reactor Coolant Pump to be circulating the reactor coolant system in all modes of operation.

Therefore this change is of an administrative nature and results in additional controls ~

and restrictions that are not specifically required but which are stated in the Bases of the Technical Specification.

BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION Thel proposed change adding specification 3/4.1.1.3.1 to the Unit ~2 Technical Specifications ' requiring at least a minimum ' flow of 3000 gpm through the Reactor-Coolant System whenever a reduction in Reactor Coolant System boron concentration is being made, doe not involve a significant hazards consideration as defined in-10 CFR 50.92 because operation of North Anna Unit 2 in accordance :with this change will not be affected and will not:

1. result in a significant increase in the probability or consequence of an accident previously evaluated. The addition of the specification would impose additional controls and restrictions on the operation of the plant which is consistent with the Unit 1 specifications and is consistent with, and cited in, the Basis of the Unit 2 specifications. Also the change is consistent with the present operating practices.

2. create the possibility of a new or different kind of accident from any accident previously identified. The addition of the specification would impose additional controls and restrictions on the operation of the plant which is consistent with the Unit 1 specifications and is consistent with, and cited in, the Basis of the Unit 2 specifications. Also the change is consistent with the present operating practices.

3. result in a significant reduction in a margin of safety. This change does not alter the conditions or assumptions of the accident analysis or the basis of the current page 1 of 11

4 sN 90 340A PROPOSED RHR Ts CHANGE j NAPS uNrrs 1 AND 2 y

Technical Specification. The addition of the specification would impose additional controls and restrictions on the operation of the plant which is consistent with the Unit 1 specifications and is consistent with, and cited in, the Basis of the Unit 2 specifications. Also the change is consistent with the present operating practices.

Therefore, pursuant to 10 CFR 50.92, based on the above considerations, it has been determined that this change does not involve a significant hazards consideration.

2: T.S. 3/4.1.1.3 REACTIVITY CONTROL SYSTEM BORON DILUTION VALVE POSITION (Unit 2 only), which is equivalent to Unit 1 specification 3/4.1.1.3.2, will be renumbered as 3/4.1.1.3.2 in order to enhance consistency and reduce confusion between Unit 1 and Unit 2 Technical Specifications.

INDEX page ill for Unit 2 is also updated to indicate the insertion of the new specification 3/4.1.1.3.1 and the renumbering of 3/4.1.1.3 to 3/4.1.1.3.2.

DISCUSSION Renumbering specification T.S. 3/4.1.1.3 to T.S. 3/4.1.1.3.2, in the Unit 2 Technical Specifications and updating INDEX page 111 are-required with the addition of T.S.

3/4.1.1.3.1 as stated above. These changes are consistent with the existent Unit 1 specification.

.Therefore these changes are only of administrative nature and results in consistency between the Unit 1 and Unit 2 Technical Specifications.

BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION The proposed change renumbering Specification T.S. 3/4.1.1.3 to'T.S. 3/4.1.1.3.2, in the Unit 2 Technical Specifications and updating INDEX page ill are' required for consistency with the existent Unit 1 specification. These changes- are only of administrative nature and do not involve a significant hazards consideration. as-

_ defined in 10 CFR 50.92 because operation of North Anna Unit 2 in accordance with these changes will not be affected and will not:

1. result in a significant increase in the probability or consequence of an accident previously evaluated because these changes are only of administrative nature and are consistent with the existent Unit 1 specification.

2. create the possibility of a new or different kind of accident from any accident previously identified because these changes are only of administrative nature and-l are consistent with the existent Unit 1 specification.

l 3. result in a significant reduction in a margin of safety. These changes do not alter j the conditions or assumptions-of the accident analysis or the basis of the current L

page 2 of 11 1

. ;+ Y e- sN 90 340A PROPOSED RHR TS CHANGE NAPS uNRs 1 AND2 Technical Specification because these changes are only of administrative nature and are consistent with the existent Unit 1 specification.-

Therefore, pursuant to 10 CFR 50.92, based on the above considerations, it has been determined that these changes do not involve a significant hazards consideration, 3: T.S. 4.4.1.3.1, REACTOR COOLANT SYSTEM - SHUTDOWN (Surveillance)

(applicable in Modes 4 and 5), T.S. 3/4,9.8.1, REFUELING OPERATIONS -

RESIDUAL HEAT REMOVAL - NORMAL WATER LEVEL (applicable in MODE 6), and T.S. 3/4.9.8.2, REFUELING OPERATIONS RESIDUAL HEAT REMOVAL LOW WATER LEVEL-(applicable in MODE 6), are being changed to allow for a reduced RHR flow from 3000 to 2000 gpm when the Reactor Coolant System Temperature is 4 less than or equal to 140 F and the reactor has been shutdown greater than or equal to 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .

DISCUSSION The proposed changes revise the current min! mum required Residual Heat Removal (RHR) system flow rate of 3,000 gpm to 2,000 gpm in Modes 5 and 6 if the reactor has '

been shutdown for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , and if there is sufficient decay heat removal to maintain the Reactor Coolant System (RCS) temperature s140 F. At the current 3,000 gpm flow rate, the RHR system is more susceptible to vortex formation at the RHR pump suction piping during mid-loop operation. Vortexing can cause RHR system air ingestion and subsequent pump cavitation, a contributor to loss of decay heat removal events.

Domestic nuclear power plants have experienced degradation or loss of RHR flow:

during mid loop operation. industry and NRC documents have addressed the loss of decay heat removal (DHR) during nonpower operation. Two of these documents specifically discuss vortexing in RHR pump suction piping during mid loop operation as a contributor to loss of RHR flow. These are NRC IE Information Notice 86-101 and INPO Significant Operating Experience Report (SOER) 85 4.

Concerning the April 10,1987 Diablo Canyon Power Plant Unit 2 event in which i operators secured the RHR pumps during Mode 5 mid loop operation to prevent pump

cavitation damage due to air entrainment caused by vortexing at the pump suction-piping, NUREG-1269 states, "Vortexing at the suction of the RHR pumps is a function

, of RHR-flowrate, and, thus, could have been reduced by reducing the RHR flowrate.

The reduction in RHR flow would, however, require an amendment to the technical specifications."

The licensee, PG&E, responded to this event, in part, by submitting a proposed Technical Specification change to decrease the susceptibility of the RHR system to ,

vortexing and thus enhance the reliability and availability of the RHR system during mid-loop operations. The NRC has authorized this change via a license amendment.

I page 3 of 11 1

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'sN 90-340A PROPOSED RHR Ts CHANGE I NAPS UNITS 1 AND 2 ;

j NRC administrative actions have also addressed RHR pump vortexing during mid loop operation as a contributor to loss of decay heat removal events. Generic Letter 8817 discusses this vortexing phenomenon, and specifically notes, " Operation of equipment in a manner that would increase the likelihood of its malfunction should be addressed.

For example, many TSs require a high DHR system flow rate when core cooling-requirements can be met at a lower rate. The high rate contributes to the likelihood that air will be ingested and cause a loss of DHR. Such operating techniques are inconsistent with reliable operation and should be addressed in meeting the longer term recommendations of this letter." 1 North Anna Power Station has experienced RHR flow degradations from ingesting air into the RHR pumps during mid loop operation, as reported in LERs 339 82 026,_339- i

'82-049, 338 82 067, 338 83 009, 339 83 038 01, and 339 84 008. Although corrective action has been enacted to both raise the RCS level during mid loop operation and to improve the RCS level measurement, reducing RHR flow provides an :

Inherent safety improvement of increased pump reliability through reduced susceptibility to vortexing in the RHR pump suction piping.

To permit a reduced Residual Heat Removal (RHR) system flow rate in Modes 5 and 6, Virginia Electric and Power Company is proposing changes to the Technical Specification Sections 4.4.1.3 (4.4.1.3.4),4.9.8.1 and 4.9.8.2 to allow RHR flow rate to be reduced to 2,000 gpm if the reactor has been shutdown at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months and ,

there is sufficient decay heat removal to maintain the RCS temperature less than or equal to 140 F. These specifications will require a reactor coolant flow rate greater than or equal to 3,000 gpm for all operation if the reactor has been shutdown less than 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months or the RCS temperature is greater than 140'F. This change will allow reduced reactor coolant flow rate only if there is sufficient heat transfer to maintain the reactor coolant temperature s140 F, which ensures adequate decay heat removal.

The intent of the reduced RHR flow rate proposed in Specifications 4.4.1.3,4.9.8.1 and-4.9.8.2 is to decrease the susceptibility of the RHR system to vortexing and enhance t the RHR pump reliability without affecting the ability of the system to provide residual heat removal. An evaluation of the pertinent operational and safety criteria concluded that the proposed specification changes will support the decay heat _ removal requirements of 10 CFR 50 Appendix A Criterion 34, will reduce the susceptibility to vortexing in mid loop operation, and will not adversely impact applicable UFSAR accident analyses. The following provides additional material to support these conclusions.

DECAY HEAT REMOVAL REQUIREMENTS General Design Criterion 34, " Residual Heat Removal" requires a system, '...to transfer i fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. .' The proposed specifications require L that the temperature of the reactor pressure vessel water be maintained 5140 F, so i that there are no concerns for fuel or pressure boundary design limits. Since the decay heat power production rate decreases with time after reactor shutdown, the page 4 of 11

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. .u *- sN 90 340A PROPOSED RHRTs CHANGE NAPS UNITS 1 ANC .?

requirements for RHR system decay heat removal also decrease in a similar manner.

To model the North Anna RHR system decay heat removal requirements, a relation was created combining the decay heat Power curve over time, the RCS water inlet (RHR return to RX vessel) temperature, and the Component Cooling system water temperature. As expected, because the Component Cooling system provides the heat sink for the Residual Heat Removal system, the lower Component Cooling water temperatures provide greater heat removal due to the larger temperature difference between the RHR and Component Cooling water, therefore, less RHR flow is required to remove the decay heat power.

The table below lilustrates the approximate minimum RHR flow rates for specific Component Cooling water temperature conditions at 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after shutdown. These calculations assume design conditions for the heat transfer components, and utilize a decay heat power curve based upon ANSI /ANS 5.1-1979.

CC Water Minimum RHR Flow Rate Temperature at 100 Hours After Shutdown 65 F 1,290 gpm 75 F 1,770 gpm 85 F 2,300 gpm 95'F- 3,340 gpm 105'F >4,000 gpm From these conditions, a minimum RHR system flow rate of 2,000 gpm during mid loop operation in Modes 5 or 6 can be justified, provided there is sufficient decay heat removal to maintain the RCS temperature 5140 F. The proposed changes to Specifications 4.4.1.3,4.9.8.1 and 4.9.8.2 will provide reasonable vortexing relief at a reasonable time after shutdown, however, if the Component Cooling water temperature is tooJ11gh, heat transfer to the Component Cooling system could be insufficient to maintain a 140 F RCS water temperature and a higher RHR flow will be required until the decay heat production diminishes further.

The proposed changes to Specifications 4.4.1.3, 4.9.8.1 and 4.9.8.2 also prohibit

' reduced RHR flow operation if the reactor has not been_ shutdown at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .

This requirement furth s. ensures that decay heat will be reduced sufficiently to permit adequate heat removal by the RHR with the reduced flow rate. The 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months specification corresponds to the typical time from power operation to mid loop, operation.

page 5 of 11

h;"e- sN 90-340A PROPOSED RHR Ts CHANGE -

NAPS uNrrs 1 AND 2,

' BORON DILUTION- !

The proposed reduction in RHR system flow to 2,000 gpm will not impact the boron

~ dilut!on accident analysis as addressed in the North Anna UFSAR. As long as RCS flow is sufficient to provide adequate boron mixing, the time to reduce.the shutdown margin to criticality will not be affected by the RHR flow rate. Furthermore, Specification 3/4.1.1.3.1 requires that RCS flow rate be greater than or equal to 3,000 gpm during a planned reduction in RCS boron concentration. Nevertheless, boron dilution is not expected to be affected by boron stratification for the following reasons.

First, boron concentration is relatively stable during Modes 5 and 6 operation. t Second, fluid temperature is well above precipitation values for the range of boron ;

concentration experienced. Third, turbulent flow Is expected in the RCS cold leg ensuring adequate boron mixing during boron dilution evolutions. Fourth, a minimum t RCS flow rate of 2,000 gpm ensures that the- RCS volume will circulate in ,

approximately 30 minutes or less during Modes 5 and 6 operation, l y

The proposed Technical Specifications do not affect the evaluation or conclusions of the boron dilution accident for Modes 5 and 6. In Modes 5 and 6, the primary grade water source is completely isolated from the_ reactor coolant system, in accordance ;

with the Technical Specifications, unless there is a planned dilution underway. Thus, ~,

inadvertent criticality due to an unplanned dilution at cold shutdown or refueling is not a credible event at North Anna. The North Anna UFSAR refers to this as a procedural prevention of RCS dilution. ,

VORTEXING l t T.he intent of the reduced RHR flow rate proposed for Specifications 4.4.1.3., 4.9.8.1 and 4 9.8.2 is to decrease the. susceptibility of vortexing at the RHR pump suction piping. ~Vortexing can cause RHR system alr ingestion and subsequent pump .

cavitation, a contributor to loss of decay heat removal events. A Westinghouse d Owners Group (WOG) project evaluated the effects of system geometry, RHR flow rate j and water level within the RCS hot leg piping, in order to predict the onset of detrimental vortexing. For the North Anna RCS configuration, at a 2,000 gpm RHR flow rate, the WOG evaluation predicts that detrimental vortexing begins at a RCS hot 1 le0 water level of about 1/4 inch above reactor vessel nozzle centerline. The normal I

p RCS-hot leg water level at North Anna during mid-loop operation is 10 inches above

! , reactor vessel nozzle centerline, yielding adequate operational flexibility and margin l to vortexing for normal plant mid-loop operation.

L

!; Mechanical Equipment Performance 1

We have evaluated the mechanical equipment in the North Anna RHR system trains, ;

l from the RCS hot leg piping to the RCS cold leg piping. Adequate long term L component performance can be assured as long as the minimum RHR flow in any p RHR loop is 2,000 gpm. The RHR pump discharge check valves 1 RH 7, I RH-15, l 2 RH-7 and 2-RH 15 require a minimum RHR flow through any one valve of 2,000 gpm to keep the check valve fully open. Since only one RHR pump is normally operated during Modes 5 and 6 operation, long term check valve performance can be assured l'

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sN 90-340A PROPOSED RHR Ts CHANGE .l NAPG UNITS 1 AND 2- J l

by maintaining RHR flow at or above 2,000 gpm. Note that short periods of lower flow can be tolerated without component degradation, which may facilitate loop swap over

. or similar evolutions.

SUMMARY

AND CONCLUSIONS Revising the current minimum required RHR system flow rate of 3,000 gpm during Modes 5 and 6 to permit a minimum RHR system flow rate of 2,000 gpm if the' reactor has shutdown and entered into Mode 3 for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , and if there is sufficient decay heat removal to maintain the RCS temperature s140 F, will reduce the susceptibility to vortex formation at the RHR pump suction piping during mid loop operation. Vortexing can cause RHR system air ingestion and subsequent pump cavitation and loss of decay heat removal capability. The change will allow greater flexibility and accuracy in RCS temperature control during all penods of low decay ,

heat generation, We have evaluated decay heat removal requirements and determined that after the reactor has shutdown and entered into Mode 3 for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , there is adequate capacity to maintain reactor pressure vessel water

. temperature $140 F during Modes 5 and 6 operation. The evaluation also determined <

that the consequences of the boron dilution accident analyses in the UFSAR, rernain acceptable and that the Specification 3/4.1.1.3.1 requirement of RHR flow 23,000 gpm ensures adequate boron mixing during planned reductions in RCS boron concentration. These revisions have been evaluated for their impact on the UFSAR safety analysis, with the conclusion that there are no changes in the UFSAR accident -

or malfunction probabilitics or consequences and that there are no changes in '

Technical Specification margins. Appropriate changes to the Bases for the affected Technical Specifications are provided.

BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION Virginia Electric And Power Company has proposed changes to the North Anna Power Station. Technical Specifications to revise the required Residual Heat Removal (RHR)

System minimum flow, it has been determined that the proposed changes do not involve a significant hazards consideration as defined in 10 CFR 50.92 because the proposed changes will not:

1. result in a significant increase in either the probability of occurrence'of any of the UFSAR accidents or their potential consequences. The boron dilution accident in.

Mode 6 was evaluated with reduced Reactor Coolant System volume and the proposed reduced RHR flow rate, and there were no increases in the probability or consequences of dilution events. Adequate mixing is provided during reductions

+

in RCS boron concentration in all modes by the 3,000 gpm reactor coolant minimum flow rate requirement of. Specification 3.1.1.3.1, so the probability of boron stratification is not increased. The requirements for the reactor to be L shutdown by at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months and to maintain reactor coolant temperature L 5140 F will ensure adequate decay heat removal capacity for the existent decay l heat generation so there is'no concern for fuel or pressure boundary design limits. :

The intent of reducing the RHR flow is to decrease the susceptibility of vortexing in the RHR pump suction piping during mid-loop operation, thereby reducing the l

L page 7 of 11

1 w I* $ * = sN 90 340A PROPOSED RHR Ts CHANGE r NAPS UNITS 1 AND2 probability of RHR pump malfunction and therefore reduce the probability of loss of decay heat removal while in mid loop operation.

L

2. create the possibility of a new or different kind of accident from any accident previously evaluated. The effect of reduced RHR flow includes the reduction of heat transfer from the RHR system to the Component Cooling System, but since this effect can be achieved through other means, the reduced heat transfer does not create the possibility of new or different accidents. The requirements for the reactor to be shutdown by at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months and to maintain reactor coolant temperature 5140 F will ensure adequate decay heat removal capacity so that there is no concern for fuel or pressure boundary design limits.

" 3. result in a significant reduction in a margin of safety. For the boron dilution basis, T. S. 3.1.1.3.1 requires a minimum reactor coolant flow rate of 3,000 gpm during ,

all modes of operation, which provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration.

reductions in the RCS. A margin of safety can be implied from the 10 CFR 50 Appendix A General Design Criterion 34, " Residual Heat Removal", wherein the 7 RHR system must provide adequate heat transfer of residual heat at a rate such that specified acceptable fuel design limits and the design conditions of the reactor-coolant pressure boundary are not exceeded. . The proposed Technical Specification revisions require that the reactor be shutdown at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months and that the RCS temperature be maintained at or s140 F during reduced RHR flow i operation, which will not challenge the fuel or reactor coolant pressure boundary design limits and the GDC-34 margin of safety for residual heat removal will not be reduced.

Therefore, pursuant to 10 CFR 50.92, based on the above considerations, it has been determined- that this. change does not involve a significant hazards.

- consideration.

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. 4: T.S. 4.4.1.3.1, REACTOR COOLANT. SYSTEM - SHUTDOWN (Surveillance),

(applicable in Modes 4 and 5) currently requires RHR to be " OPERABLE per Specification 4.0.5". This will be changed to be ' OPERABLE per Specification 4.7.9.2." t DISCUSSION The change to Specification 4.4.1.3.1 to cite the operability requirements of the RHR PLANT . SYSTEM Specification 3/4.7.9.2, rather than Specification 4.0.5 is a purely administrative change to achieve consistency throughout the specifications.

T.S. 3/4.7.9.2 provides the Surveillance Requirements for RHR System operability i determination in Modes 4 and 5 and includes the requirement for a determination of operability per Specification 4.0.5 (Reference Virginia Electric and Power Company's h - submittal for Licence Amendment dated April 27,1990). Therefore no change will be l effected to the requirements of the Technical Specifications or the operation of North Anna Units 1 and 2.

page 8 of 11

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j s sN 90 340A -

PROPOSED RHR Ts CHANGE NAPS UNITS 1 AND 2 BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION i

^

-The proposed change to Specification 4.4.1.3.1 requiring operability determination per Specification 3/4.7.9.2 rather than Specification 4.0.5, does not involve a significant ,

hazards consideration as defined in 10 CFR 50.92 because operation of North Anna Units 1 and 2 in accordance with this change will not be affected and will not:

L result in a significant increase in the probability or consequence of an accident previously evaluated. Specification 4.4.1.3.1 will still require determination of .

operability per Specification 4.0.5 by citing Specification 3/4.7.9.2 which includes operability determination per Specification 4.0.5.

2. create the possibility of a new or different kind of accident from any accident 1 previously identified. Specification 4.4.1.3.1. will still require determination of ;

operability per Specification 4.0.5 by citing Specification 3/4.7.9.2 which includes operability determination per Specification 4.0.5.

3. result in a significant reduction in a margin of safety. This change does not alter the conditions or assumptions of the accident analysis or the basis _of the current l Technical Specification. Specification 4.4.1.3.1 will still require determination of j operability per Specification 4.0.5 by citing Specification 3/4.7.9.2 which includes !

operability determination per Specification 4.0.5.

Therefore, pursuant to 10 CFR 50.92, based on the above considerations, it has been- determined that this change does not involve a significant hazards consideration.

5: T.S. 3/4.9.8.1, REFUELING OPERATIONS - RESIDUAL HEAT REMOVAL - ALL *

. WATER LEVELS, will be changed from "All Water Levels" to " Normal Water Levels".

This reflects the structure of the Standard Technical Specifications.

Also' T.S. 3/4.9.8.1 LCO will be changed to read "At least one RHR loop shall be OPERABLE and at least one RHR loop shall be in operation."' This adds a requirement I for an RHR loop to be operable and will conform to the Standard Technical

-Specifications.

DISCUSSION Changing T.S. 3/4.9.8.1 from "All Water Levels" to " Normal Water Levels" reflects the structure of the Standard Technical Specifications This change is administrative in nature and achieves consistency and clarity by eliminating the need to combine the requirements of this specification with T.S. 3/4.9.8.2 during ' low' water level operations. The operation of the plant and the the requirements of the Technical Specifications are not changed.

, page 9 of 11

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sN 90-340A i PROPOSED RHR Ts CHANGE NAPS UNrrs 1 AND 2 Also although T.S. 3/4.9.8.1 currently requires that-one residual heat removal loop to be in operation, there is no requirement for an RHR loop to be operable. The change in the LCO to require at least one RHR loop to be operable insures RHR operability and conforms to the Standard Technical Specifications. The Action and Surveillance requirements added to this specification are equivalent to those currently existent for operability in Specification 3/4.9.8.2, REFUELING OPERATIONS - RESIDUAL HEAT REMOVAL - LOW WATER LEVELS. Therefore, the change to Specification 3/4.9.8.1 adding the requirement for at least one operable RHR loop constitutes an additional control or rer.triction which is not presently included in the Technical Specifications but which is cor.sistent with the current operating practices of the plant.

BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION -

The proposed changes to Specification 3/4.9.8.1 do not involve a significant hazards consideration as defined in 10 CFR 50.92 because operation of North Anna Units 1 and 2 in accordance with this change will not be affected and therefore will not:

1. result in a significant increase in the probability or consequence of an accident previously evaluated. Specification 3/4.9.8.1 will still' require the operation of at least one RHR loop and the requirement to determine at least one RHR loop to be operable constitutes an additional control which is equivalent to the requirement currently in place during ' low' water operations and is consistent with the Standard Technical Specification.

2. create the possibility of a new or different kind of accident from any accident previously identified. Specification 3/4.9.8.1 will still-require the operation of at least_one RHR loop and the requirement to determine at least one RHR loop to be operable constitutes an additional control which is equivalent to the requirement currently in place during ' low' water operations and is consistent with the Standard

Technical Specification.

3. result in a significant reduction in a margin of safety. This change does not alter the conditions or assumptions of the accident analysis or the basis of the current Technical Specification. T.S. 3/4.9.8.1 will still require.the operation of at least one RHR loop and adding the requirement to determine at least one RHR loop to be operable constitutes an additional control which is equivalent to the requirement currently in place during ' low' water operations.and is consistent with the Standard Technical Specification.

Therefore, pursuant to 10 CFR 50.92, based on the above considerations, it has been determined that this change does not involve a significant hazards consideration.

6: T.S. 3/4.9.8.2, REFUELING OPERATIONS - RESIDUAL HEAT REMOVAL - LOW WATER LEVEL, is being changed from "Two independent RHR loops shall be OPERABLE" by adding "with at least one loop in operation" to the Limiting Condition For Operation (LCO).

page 10 of 11

?o e.N *1 sN 90 340A C' PROPOSED RHRTs CHANGE '

NAPS UNrTs 1 AND 2 DISCUSSION Changing Specification 3/4.9.8.2 to require at least one operating RHR loop is consistent with the current Technical Specifications and conforms to the Standard Technical Specifications.

The: current Specification 3/4.9.8.1 applies to "ALL" water level conditions and therefore, the requirement to have at least one RHR loop in operation is in affect during -

both " Normal" and " Low" water level conditions. In order to maintain that requirement.

In light of the proposed change to 3/4.9.8.1 limiting the applicability of that specification to " Normal Water Levels" only (see above), the LCO requirement for at least one RHR loop to be in operation must be explicitly-inserted into 3/4.9.8.2. The Action and Surveillance requirements added to this specification are equivalent to those currently-existent in Specification-3/4.9.8.1. (NOTE:'the proposed change allowing reduced RHR flow is discussed separately.) This change is administrative in nature only as the requirement is not changed and is currently in effect.

BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION The' proposed changes to Specification'3/4.9.8.2 do not involve a significant hazards consideration as defined in 10 CFR 50.92 because operation of North Anna Units 1 and 2 in accordance with this change will not be affected and therefore will not:

1. result in a significant increase in the probability or consequence of an accident previously evaluated. Specification 3/4.9.8.2 will still require the operation of at least one RHR loop. The change only itemizes the requirement. There are no changes in the operation of the plant or Technical Specifications requirements.

.2. create the possibility of a new or.different kind.of accident from any accident previously identified. Specification 3/4.9.8.2 will still require the operation of at least one RHR loop. The change only itemizes the requirement. There are no changes in the operation of the plant or Technical Specifications requirements.

3. result in a significant reduction in a margin of safety. This change does not alter the conditions or assumptions of the accident analysis or the basis of the current Technical Specification. Specification 3/4.9.8.2 will still require the operation of at least one RHR loop. The change only itemizes the requirement. There are no changes in the operation of the plant or Technical Specifications requirements.

Therefore, pursuant to 10 CFR 50.92, based on the above considerations, it has been determined that this change does not involve a significant hazards consideration.

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ML20043H124

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