ML17263A728

Proposed Tech Specs Providing NRC W/Opportunity to Communicate at Early Stage Any Concerns W/Respect to Differences from NUREG-1431

ML17263A728
Person / Time
Site:	Ginna
Issue date:	07/15/1994
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17263A729	List: ML17263A728 ML17309A550
References
RTR-NUREG-1431 NUDOCS 9407190408
Download: ML17263A728 (175)

Text

1.12 Frecruenc'otation The frequency notation specified for the performance of surveillance requirements shall correspond to the intervals defined below.

Notation Precruenc S, Each Shift At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> D, Daily At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Twice per week At least once per 4 days and at least twice per 7 days W, Weekly A< least once per 7 days B/W, Biweekly At least once per 14 days M, Monthly At least once per 31 days B/M, Bimonthly At least once per 62 days Q, Quarterly At least once per 92 days SA, Semiannually At least once per 6 months A, Annually At least once pe 12 months At least once per 18 months Prior to each startup N.A. Not Applicable Prior to each startup if not done previous week Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to each release

1. 13 Offsite Dose Calculation Manual ODCY'he ODCM is a manual containing the methodology and

.parameters to be used for calculating the offsite qq07>cocoa esO7>s PDR ADOCK 05000244 p PDR

3 ~ Emer enc Core Coolin S stem Auxil a Cool n S ste A'r Recirculation Fan Coolers Containment S ra and arcoal HEPA Filters Ob'e 've To define se conditions for o ation that are neces-sary:(1) to remove ecay hea rom the core in emergency or normal shutdown situat,'2) to remove heat from contain-ment in normal crating and e rgency situations, (3) to remove ai orne iodine from the c ainment atmosphere foll xnq a postulated Design Basis Acciden and (4) to minimize containment leakage to the environment subse ~nt to a Design Basis Accident.

~/

3.3.1 Safet In'ection and Residual Heat Removal S stems 3.3.1.1 The reactor shall not be taken above the mode indicated unless the following conditions are met:

a ~ Above cold shutdown, the refueling water storage tank CQ 3, 5.4 contains not less than 300,000 gallons of water, with a SR, o .5.'t <

boron concentration of at least 2000 ppm.

~

SQ. Z.S.~.R.

b. Above a reactor coolant system pressure of 1600 psig, except during performance of RCS hydro test, each accumulator, is pressurized to at least 700 psig with an indicated level of at least 50< and a maximum of 824 with (3,vi a boron concentration of at least 1800 ppm.

c ~ At or above a reactor coolant system temperature of 4C.u 3.S.Z.

350 F, three safety injection pumps are operable.

L3 ~ L.I i

At or above an RCS temperature of 350oF, two residual heat removal pumps are operable.

At or above an RCS temperature of 350'F, two residual heat removal heat exchangers are operable.

At the conditions required in a through e above, all valves," interlocks and piping associated with the above components which are required to function during accident conditions are operable.

At or above an RCS temperature of 350 F, A.C. power shall be removed from the following valves with the valves in the open position: safety injection cold leg injection valves 878B and D. A.C. power shall be removed from safety injection hot leg injection valves 878A and C with the valves closed. D.C. control power shall be removed from refueling water storage tank delivery valves 896A, 896B and 856 with the valves open.

At or above an RCS temperature of 350 F, check valves 853A, 853B, 867A, 867B, 878G, and 8788 shall be operable with less than 5.0 gpm leakage each. The leakage requirements of Technical Specification 3.1.5.2.1 are still applicable.

Above a reactor coolant system pressure of 1600 psig, except during performance of RCS hydro test, A.C. poweri shall be removed from accumulator isolation valves 841 and 865 with the valves open.

At or above an RCS temperature of 350 F, A.C. power shall be removed from Safety Injection suction valves 825A and B with the valves in the open position, and from valves 826A, B, C, D with the valves in the closed position.

3 ' 1 2 If the conditions of 3.3.1.1a are not met, then satisfy the o ~.~.9 condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be at hot shutdown in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at least cold shutdown within an additional 30 i 3 i.v

~

hours.

3.3. 1.3 The requirements of 3.3.l.lb and 3.3.l.li may be modified to L,C-O 'Z. s. i allow one accumulator to be inoperable or isolated for up to one hour. If the accumulator is not operable or is still (iQ isolated after one hour, the reactor shall be placed in hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and below a RCS pressure of 1600 psig within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3.3. 1.4 The requirements of 3. 3. 1. lc may be modified to allow one safety injection pump to be inoperable for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If L.c.o ,s,2 the pump is not operable after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor shall be placed in hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and below a RCS temperature less than 350oF within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3.3.1.5 ~ The requirements of 3.3.1.1d through h. may be modified to allow components to be inoperable at any one time- More than l Co 3.5.2 one component may be inoperable at any one time provided that one train of the ECCS is operable. If the requirements of 3.3.1.1d through h. are not satisfied within the time period specified below, the reactor shall be placed in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at an RCS temperature less than 350 F in an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ae One residual heat removal pump may be out of service

~

the pump is restored to operable status within

-'q+~-q+~cayg.'rovided 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

residual heat removal heat exchanger may be out of 4'. 5.2

b. one service for a period of no more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

co Any valve, interlock, or piping required for the func-tioning of one safety injection train and/or one low head safety injection train (RHR) may be inoperable provided repairs are completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (except as speci-fied in e. below) .

Power may be restored to any valve referenced in 3.3.1.1g

". s.2. Cuow')

for the purposes of valve testing provided no more than Lc.o one such valve has power restored and provided testing is completed and power removed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

e. Those check valves specified in 3.3.1.1h may be inopera-ble (greater than 5. 0 gpm leakage) provided the inline Movs are de-energized closed and repairs are completed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

a~

The facility has four service water pumps. Only on i needed during the injection phase, and two re ired during the recirculation phase postu ated loss-of-coolant accident.' The ontrol room e rgency air treatment system is d igned to filter th control room atmosphere during eriods when the control oom is isolated and to mai ain radiation I

levels in t control room at a ceptable levels following "he esign Basis Acc 'nt. ' Reactor operation may co tinue for a limited time while repairs are being de to t air treatment system since it. is unlikely t at t e system would, be needed.

'Zechnical Specification 3.3.5 applies only to the equipmen" necessary o ilter the control room atmosphere. Equipme t neces ry to initiate isolation of. t!>e control room is cove)ed hy another specfication.

The limits r the accumulator p essure and volume assure the required amount of water 'njection'uring an acci nt, and are based on value used for the accid t analyses. The indicated 1 vel of 50%

cor esponds to 1108 cubic feet of wat r in the a .cumulator and the indicated level of 82% co esponds to 1134 cubic feet.

The limitation of no more than .one safety inje tion pump to be operable when overpressure protection

's being provided by a RCS vent of > 1.1 sq. in. insures 3.3-13 Amendmen't No. gg, 48

hat the mass addition from the inadvertent operation s ety injection will not result in RHR system pres re exc eding design limits. The limitation on no afety injec ion pumps operable and the discharge lines solated when ov rpressure protection is provided by t pressur-izer POR 's removes mass injection from inadvertent safety inje tion as an event for which thi configuration of overpress e protection must be des'gned to protect.

i inoperability a safety injection mp may be verified from the main co trol board with t e pump control switch in pull stop, or t e pump breake in the test racked out position such that the pump could not start from an inadvertent safety i 'ecti n signal. Isolation of a safety injection pump 'arge path to the RCS may be verified from the main rol board by the discharge MOV switch position indic ting osed, or the discharge valve closed with A.C. wer remo ed, or a manual discharge path isolation v lve closed s h that operation of the associated saf ty injection pump ould not result in mass injection t the RCS.

High conc tration boric acid is no needed to mitigate the con equences of a design basis ac 'dent. Reference (10) emonstrates that the design basis ccidents can be mi gated by safety injection flow of RWST oncentration.

erefore, SI pump suction is taken from th RWST.

Requiring that the safety injection suction v ves (825A and 8, 826A, B, C and D) are aligned with A. . power removed insures that the safety injection syste would not be exposed to high concentration boric acid an the assumptions of the accident analysis are satisfied.

Amendment No. , 57 3.3-14

eferences

( ) Deleted (2) UFSAR Section 6.3.3.1 (3) UFSAR Section 6.2.2.1 (4) FSAR Section 15.6.4.3 (5) U AR Section 9.2.2.4 (6) UFS Section 9.2.2.4 (7) Dele ed (8) UFSAR ection 9.2.1.2 (9) UFSAR S ction 6.2.1.1 UFSAR Se tion 6.4 (Containm (CR Emergency t Integrity)

Air Treatment) and (10) Mestingho se Report, "R.E. G'a Boric Acid Storage Tank Boro Concentration eduction Study" dated Nov. 1992 C.J. McHugh d J.J. Sprysbak endment No. P, 57 3 3. 14a

TABLE 4.1-1 (Continued)

Channel esc i tio ~C eck Calibrate Test liemarks

10. Rod Position Bank S(1,2) H.A. H.A. 1) With rod position indication Counters 2) Log rod position indications each 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when rod deviation monitor is out of service ll. Steam Generator Level

12. Charging Flow H.A. H.A.

13. Residual Heat Removal fl. A. H.A.

Pump Flow

14. Boric Acid Storage Tank Level D H.A. Note 4 s a z.s.~. > 15. Ref ue ling Water H.A. H.A.

Storage Tank Level

16. Volume Control Tank N.A. N.A.

Level

17. Reactor Containment M(1) 1) Isolation Valve signal Pressure

18. Radiation Monitoring Area Monitors Rl to R9, System System Monitor R17

19. Boric Acid Control N.A. N.A.

SR a.~eiS.z 20. Containment Drain H.A. N.A.

Sump Level

21. Valve Temperature H.A. H.A.

Interlocks

22. Pump-Valve Interlock N.A. N.AD

23. Turbine Trip N.A. M(1) 1) Block Trip Set-Point sa s.s.i.2. 24. Accumulator Level and N.A.

sa 3.s. i.2, pressure Amendment Nc. g 57 4.1-6

2'R.i i.

~

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TABLE 4 1-2 MINIMUM FRE UENCIFS FOR E UIPM NT AND SAM LING TESTS

~Fro u~onc

1. Reactor Coolant Chloride and Fluoride 3 times/week and at least Chemistry Samples every third day Oxygen 5 times/week and at least every second day except when below 2504F

2. Reactor Coolant Boron Concentration Meekly Boron

3. Refueling Water Boron Concentration Weekly Sp p.g q 2, Storage Tank Water Z.Jt, 'a1,$ Sample

4. Boric Acid Storage Boron Concentration Twice/Week"'.

Tank Control Rods Rod drop times of all After vessel head removal full length rods and at least once por 18 months ( 1) 6a. Full Length Hove any rod not fully Monthly Control Rod inserted a sufficient number of steps in any ona direction to cause a change of posi.tion as indicated by the rod position indication system 6b. Full Length Hove each rod through Each Refueling Shutdown Control Rod its full length to verify that tho rod position i.ndication system transitions occur

7. Pressurizer Safety Sot point Each Refueling Shutdown Valves

8. Hain Steam Safety Set point Each Refueli.ng Shutdown Valves

9. Containment Functioning Each Refueling Shutdown Zsolation Trip

10. Refueling System Functioning Prior to Refueling Znterlocks Operations Amendment No. , 57 4. 1-8

T69t ~Fe ~enc

11. Service Mater Functioning Each Refueling Shutdown System

12. Fire Protection Functioning Monthly Pump and Power Supply

13. Spray Addi.tive NaOH Concent Monthly Tank Z.t '~'i Qg, S,S.i."( 14. Accumulator Boron Concentration Bi-Monthly KQ 'at h

<P.'i i.c.

15. Primary System Evaluate Daily Leakage

16. Diesel Fuel Supply Fuel Xnventory Daily

17. Spent Fuel Pit Boron Concentration Monthly

18. Secondary Coolant Gross Activity 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (2) (3)

Samples

19. Circulating Mater Calibrate Each Refueling Shutdown Flood Protection Equipment Notes:

Also required for specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the dro time of th specific rods.

Not required during a cold or refueling shutdown.

(3) An isotopic analysis for I-131 equivalent activity is required at least monthly whenever the gross activity determination indicates iodine concentration greater than 104 of the allowable limit but only once per 6 months wh enever th e gross activit,y, dete e ermination indicates iodine concentration below 104 of the allowable limit.

(4) When BAST is required to be operable.

Amendment Nn. g, 57 4 1-9

4. 5 Safet Iniection, Containment Sora and Iodine Removal S stems Tests Ao licabilit 4 Applies to testing of the Safety Injection System, the Contain-ment Spray System, and the Air Ziltratior System inside Ccn-V ~

tainment.

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~V 7

Obiective:

To verify that the subject systems will respond promptly and perform their intended functions, if required.

4. 5. l. 1 Safet In ection S stem

\J I

a. System tests shall be performed at each reactor refuel.ing

~ t interval. The test shall be performed in accordance with the following:

With the reactor coolant system pressure less than or, equal to 350 psig and temperature less than or equal to 350 F, a test safety injection signal will be applied to initiate operation of the system. The safety injection and residual heat removal pump motors are prevented from starting during the test.

4. 5-1

Thc system test will bc considered satisfactory if control board indicati;n and visual observations ZR indicate that all valves have received the Safety injection signal and have complctcd their travel.

Thc proper scnucnce and timing of thc rotating

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components are to bc verified in conjunction xvith

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Section 4. 6.1 b.

4. 5. l. 2 Containmcnt Snra S~ stem

a. System tests shall bc performed at.each rc" ctor re-fueling interval. The test shall. bc performed with tnc ssolatacn valves,in thc spray supply lines,at thc con-tainmcnt blocltcd closed. Ppcration of the system f' ~

ts )nitiatcd by tripping thc normal actuation instrumcn-e ~ ~

mc I

tation.

b. Thc spray no@ales shall bc checked for proper functioninr; at least every five years.

C~ Thc test will bc considered satisfactory if visual obser-vations indicate all c'Opponents have operated satisfac-tor ily.

% ~

4.5.2 Com onent Tests

4. 5. 2. 1 Pumps Except during cold or refueling shutdown)s thh safety pumps, residual heat rea'oval pu~s,l'njectxon and S'a 3.g.z.."I containmcnt spray pumps shall be sa,arted at intervals QQ ~ c c not to exceed one month. The pumps shall be tested prior to startup if the time since the last test exceeds 1 month.

4.S-2

0

b. Acceptable levels of performance for the pumps shall be that the pumps start, operate, and develop 3.e.i.

~ ~ l the minimum discharge pressure for the flows listed 3'X.4. L.

in the table below:

DISCHARGE PRES&JRE .

Containment Spray Pumps 35 .gpm 240 psig Notes Residual Heat Removal Pumps [200 gpmj [140 psigJ 450 gpm 138 psig Safety Inject:ion Pumps [50 gpml [1420 psigj (2) 150 gpm 1356 psig Table 4.5-3.

Notes (1) Items in square brackets are effective until the installation of t:he new residual heat removal minimum flow recirculation system.

(2) Items in square brackets are effective until installation of the new safety injection minimum flow recirculation system.

4. 5.2. 2 Ve3.ves

a. Except during cold or refueling shutdowns the spray 4

additive valves shall be tested at intervals not to exceed one month. With the pumps shut down and the valves upstream and downstream Aaandramt No. 33 4 5-3

of "he spray additive valves closed, each valve will be opened and closed bv operator act'on.

This test shall be per armed prior to startup i =he time since the last test exceeds one mon&.

he accumulator check valves shall be checked or 3 Q,ill operabili"y during each refueling shutdown.

4.5.2.3 Air . ' "rat an System 4.5.2.3.1 At 'eas" =nce evezv 18 months or a te every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> af cha c"a tra"ion svstem operat'on since the last test, allowing painting, i e or chemical release in any ve.-

or ~o1 1 a-'".". "one communicating with the svstem, the "ost accident char""a'ystem shall have the following cond'ians demonstr ted

a. The "ressure drop aczoss the charcoal adsorber bank is less than 3 inches az water at des'gn zlow rate (~ 105) .

b. In place Freon testing, under ambient conc't'ons, shall show at least 99% removal.

C ~ The iad'ne removal ef fic'ncy oz at least one charcoal fi'er cell sha'1 be measured. The fter cel'o be tested shall be selected randomly from those cells with the longest in-ban3c residence time. The.;mi~a~4 py414 acceptable value for ilter ef iciency is 90%'r re-moval o" methyl iod'de when tested at at least 2S6'F and 954 RH and at 1.5 ta 2.0 mg/m3 load'..g with taggec.

CK3I.

4 5-4

ATTACHMENT C Proposed Revised R.E. Ginna Nuclear Power Plant Improved Technical Specifications Revise the pages as follows:

Remove Inser t Table of Contents Ginna Station ITS Table of Contents Entire Section 1.0 Ginna Station ITS Section 1.0 Entire Section 2.0 Ginna Station ITS Section 2.0 Entire Section 3.0 Ginna Station ITS Section 3.0 Entire Section 4.0 Entire Section 5.0 Ginna Station ITS Section 4.0 Entire Section 6.0 Ginna Station ITS Section 5.0 ONLY SECTION 3.5 IS PROVIDED AT THIS TIME

Accumulators 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators LCO 3.5.1 Two ECCS accumulators shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODE 3 with pressurizer pressure > 1600 psig.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One accumulator A, 1 Restore boron 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable due to concentration to boron concentration within limits.

not within limits.

B. One accumulator B. 1 Restore accumulator 1 hour inoperable for reasons to OPERABLE status.

other than Condition A.

C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time of Condition A AND or B not met.

C.2 Reduce pressurizer 12 hours pressure to < 1600 pslg.

D. Two accumulators D.1 Enter LCO 3.0.3. Immediately inoperable.

R.E. Ginna Nuclear Power Plant 3.5-1

J t l

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5. 1. 1 Verify each accumulator motor operated 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> isolation valve is fully open.

SR 3.5. 1.2 Verify borated water volume in each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> accumulator is z 1120 cubic feet (50%) and g 1190 cubic feet (82%).

SR 3.5. 1.3 Verify nitrogen cover pressure in each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> accumulator is > 700 psig and S 790 psig.

SR 3.5. 1.4 Verify boron concentration in each 31 days on a accumulator is Z 1800 ppm and g 2900 ppm. STAGGERED TEST BASIS SR 3.5. 1.5 Verify power is removed from each 31 days accumulator motor operated isolation valve operator when pressurizer pressure is

> 1600 psig, R.E. Ginna Nuclear Power Plant 3.5-2

ECCS Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

-NOTES

1. In MODE 3, both safety injection (SI) pump flow paths may be isolated by closing the isolation valves for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to perform pressure isolation valve testing per SR 3.4. 14. 1. Power may be restored to motor operated isolation valves 878A, 878B, 878C, and 878D for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the purpose of testing per SR 3.4, 14. 1 provided that power is restored to only one valve at a time.

2. Operation in MODE 3 with ECCS pumps declared inoperable pursuant to LCO 3.4. 12, "Low Temperature Overpressure Protection (LTOP) System," is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or until the temperature of both RCS cold legs exceeds 375'F, whichever comes first.

ACTIONS CONDITION RE(UIRED ACTION COMPLETION TIME A. One train inoperable. A.1 Restore train to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status.

AND At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available, (continued)

R.E. Ginna Nuclear Power Plant 3.5-3

II V 41

~a u

ECCS Operating 3.5.2 ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B,l Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours C. Two trains inoperable. C.1 Enter LCO 3.0.3 Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.

~ ~ ~ 1 Verify the folloWing valves are in the 12 hours listed position.~

Number Position Function 825A Open RWST Suction to SI Pumps 825B Open RWST Suction to SI Pumps 826A Closed BAST Suction to SI Pumps 826B Closed BAST Suction to SI Pumps 826C Closed BAST Suction to SI Pumps 8260 Closed BAST Suction to SI Pumps 851A Open Sump B to RHR Pumps 851B Open Sump B to RHR Pumps 856 Open RWST Suction to RHR Pumps 878A Closed SI Injection to RCS Hot Leg 878B Open SI Injection to RCS Cold Leg 878C Closed SI Injection to RCS Hot Leg 878D Open SI Injection to RCS Cold Leg 896A Open RWST Suction to SI and Containment Spray 896B Open RWST Suction to SI and Containment Spray (continued)

R.E. Ginna Nuclear Power Plant 3.5-4

' '4 -

~ "5L 3, ~ '~ I I . II

ECCS Operating 3.5.2 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3.5.2.2 Verify each ECCS manual, power operated, 31 days and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.2.3 Verify each breaker or key switch, as 31 days applicable, for each valve listed in SR 3.5.2. 1, is in the correct position, SR 3.5.2.4 Verify each ECCS pump's develop'ed head at In accordance the test flow point is greater than or with the equal to the required developed head. Inservice Testing Program SR 3.5.2.5 Verify each ECCS automatic valve in the 24 months flow path that is not locked, sealed, or otherwise secured in position actuates to the correct position on an actual or simulated actuation signal.

SR 3.5.2.6 Verify each ECCS pump starts automatically 24 months on an actual or simulated actuation signal.

R.E. Ginna Nuclear Power Plant 3.5-5

ECCS Shutdown 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.3 ECCS Shutdown LCO 3.5.3 One ECCS train shall be OPERABLE.

'PPLICABILITY: MODE 4.

ACTIONS CONDITION RE(UIRED ACTION COMPLETION TIME A. Required ECCS residual A. 1 Initiate action to Immediately heat removal (RHR) restore required ECCS subsystem inoperable. RHR subsystem to OPERABLE status.

B.

~ Required ECCS Safety B. 1 Restore required ECCS 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> injection

~

SI subsystem to (SI)subsystem OPERABLE status.

inoperable. ~

C. Required Action and C.1 Be in MODE 5. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> associated Completion Time of Condition B not met.

R.E. Ginna Nuclear Power Plant 3.5-6

ECCS Shutdown 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3. I NOTE-An RHR train may be considered OPERABLE during alignment and operation for decay heat removal, if capablemodeofofbeing manually operation.

realigned to the ECCS The following SR is applicable for all In accordance equipment required to be OPERABLE: with applicable SR SR 3.5.2.4 R.E. Ginna Nuclear Power Plant 3.5-7

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RWST 3.5.4 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.4 Refueling Water Storage Tank (RWST)

LCO 3.5.4 The RWST shall be OPERABLE.

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

ACTIONS CONDITION RE(UIRED ACTION COMPLETION TIME A. RWST boron A.l Restore RWST to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> concentration not OPERABLE status.

within limits.

B.~ RWST water volume not B. I Restore RWST to I hour within limits.

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~ OPERABLE status.

C. Required Action and C.l Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or AND B not met.

C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> R.E. Ginna Nuclear Power Plant 3.5-8

RWST 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4. 1 Verify RWST borated water volume is Z 7 days 300,000 gallons (88%).

SR 3.5.4.2 Verify RWST boron concentration is 7 days Z 2000 ppm and g 2900 ppm.

R.E. Ginna Nuclear Power Plant 3.5-9

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Accumulators B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5. 1 Accumulators BASES BACKGROUND The functions of the ECCS accumulators are to supply water to the reactor vessel during the blowdown phase of a large break loss of coolant accident (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide Reactor Coolant System (RCS) makeup for a small break LOCA.

The blowdown phase of a large break LOCA is the initial period of the transient during which the RCS departs from equilibrium conditions, and heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. The reactor coolant inventory is vacating the core during this phase through steam flashing and ejection out through the break. The blowdown phase of the transient ends when the RCS pressure falls to a value approaching that of the containment atmosphere.

In the refill phase of a LOCA, which immediately follows the blowdown phase, the core is essentially in adiabatic heatup.

The balance of accumulator inventory is available to reflood the core and help fill voids in the lower plenum and reactor vessel downcomer so as to establish a recovery level at the bottom of the core.

The accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas. The accumulators are passive components, since no operator or control actions are required in order for them to perform their function. Internal accumulator tank pressure is sufficient to discharge the accumulator contents to the RCS, if RCS pressure decreases below the accumulator pressure.

(continued)

R.E. Ginna Nuclear Power Plant 8 3.5-1

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Accumulators B 3.5.1 BASES (continued)

BACKGROUND Each accumulator is piped into an RCS cold leg via an (continued) accumulator line and is isolated from the RCS by a motor operated isolation valve and two check valves in series.

The motor operated isolation valves (841 and 865) are maintained open with AC power removed under administrative control when pressurizer pressure is ) 1600 psig. This feature ensures that the valves meet the single failure criterion of manually-controlled electrically operated valves per Branch Technical Position (BTP) ICSB-18 (Ref. 1).

This is also discussed in References 2 and 3.

r The accumulator size, water volume, and nitrogen cover pressure are selected so that one of the two accumulators is sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA, The need to ensure that one accumulator is adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA.

APPLICABLE The accumulators are assumed OPERABLE in both the large and SAFETY ANALYSES small break LOCA analyses at full power (Ref. 4). These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits.

In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a large break LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is required by regulations and conservatively imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-2

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Accumulators B 3.5.1 BASES (continued)

APPLICABLE The limiting large break LOCA is a double ended guillotine SAFETY ANALYSES break at the discharge of the reactor coolant pump. During (continued) this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure. As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. This delay accounts for SI signal generation, the diesels starting, and the pumps being loaded and delivering full flow. During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA.

The worst case small break LOCA analyses also assume a time delay before pumped flow reaches the core. For the larger range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators and safety injection pumps both play a part in terminating the rise in clad temperature. As break size continues to decrease, the role of the accumulators continues to decrease until they are not required and the safety injection pumps become solely responsible for terminating the temperature increase.

This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 5) will be met following a LOCA:

a. Haximum fuel element cladding temperature is g 2200'F;

b. Haximum cladding oxidation is g 0. 17 times the total cladding thickness before oxidation;

c. Haximum hydrogen generation from a zirconium water reaction is < 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; and

d. Core is maintained in a eoolable geometry.

Since the accumulators discharge during the blowdown phase of a LOCA, they do not contribute to the long term cooling requirements of 10 CFR 50.46.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-3

Accumulators B 3.5.1 BASES (continued)

APPLICABLE For both the large and small break LOCA analyses, a nominal SAFETY ANALYSES contained accumulator water volume is used. The contained (continued) water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged. for small breaks, an increase in water volume is a peak clad temperature penalty due to the reduced gas volume. A peak clad temperature penalty is an assumed increase in the calculated peak clad temperature due to a change in an input parameter. For large breaks, an increase in water volume can be either a peak clad temperature penalty or benefit, depending on downcomer filling and subsequent spill through the break during the core reflooding portion of the transient. The analysis uses a nominal accumulator volume and includes the line water volume from the accumulator to the check valve due to these competing effects.

The minimum boron concentration setpoint is used in the post LOCA boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment. Of particular interest is the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the time frame in which boron precipitation is addressed post LOCA.

The maximum boron concentration limit is based on the coldest expected temperature of the accumulator water volume and on chemical effects resulting from operation of the ECCS and the Containment Spray System. A value of 2,900 ppm would not create the potential for boron precipitation in the accumulator assuming a containment temperature of 40'F (Ref. 6). Analyses performed in response to 10 CFR 50.49 (Ref. 7) assumed a chemical spray solution of 2000 to 3000 ppm boron concentration (Ref. 6) which provides a margin of 100 ppm. The chemical spray solution impacts sump pH and the resulting effect of chloride and caustic stress corrosion on mechanical systems and components. The sump pH also affects the rate of hydrogen generation within containment due to the interaction of Containment Spray and sump fluid with aluminum components.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-4

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Accumulators B 3.5.1 BASES (continued)

APPLICABLE The large and small break LOCA analyses are performed at the SAFETY ANALYSES minimum nitrogen cover pressure, since sensitivity analyses (continued) have demonstrated that higher nitrogen cover pressure results in a computed peak clad temperature benefit. The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation at 800 psig, and ultimately preserves accumulator integrity.

The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Refs. 8 and 9).

The accumulators satisfy Criterion 3 of the NRC Policy Statement.

LCO The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Two accumulators are required to ensure that 100% of the contents of one accumulator will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than one accumulator is injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 5) could be violated.

For an accumulator to be considered OPERABLE, the motor-operated isolation valve must be fully open, power removed above 1600 psig, and the limits established in the SRs for contained volume, boron concentration, and nitrogen cover pressure must be met.

APPLICABILITY In NODES 1 and 2, and in NODE 3 with RCS pressure > 1600 psig, the accumulator OPERABILITY requirements are based on full power operation. Although cooling requirements decrease as power decreases, the accumulators are still required to provide core cooling as long as elevated RCS pressures and temperatures exist.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-5

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Accumulators B 3.5.1 BASES (continued)

APPLICABILITY This LCO is only applicable at pressures > 1600 psig. At (continued) pressures g 1600 psig, the rate of RCS blowdown is such that the ECCS pumps can provide adequate injection to ensure that peak clad temperature remains below the 10 CFR 50.46 (Ref. 5) limit of 2200'F.

In NODE 3, with RCS pressure < 1600 psig, and in NODES 4, 5, and 6, the accumulator motor operated isolation valves are closed to isolate the accumulators from the RCS. This allows RCS cooldown and depressurization without discharging the accumulators into the RCS or requiring depressurization of the accumulators.

ACTIONS A,1 If the boron concentration of one accumulator is not within limits, it must be returned to within the limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />'n this Condition, ability to maintain subcriticality or minimum boron precipitation time may be reduced. The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially recovered core during the early reflooding phase of a large break LOCA is sufficient to keep that portion of the core subcritical. One accumulator below the minimum boron concentration limit, however, will have no effect on available ECCS water and an insignificant effect on core subcriticality during reflood since the accumulator water volume is very small when compared to RCS and RWST inventory. Boiling of ECCS water in the core during reflood concentrates boron in the saturated liquid that remains in the core. In addition, current analysis techniques demonstrate that the accumulators are not expected to discharge following a large main steam line break. Even if they do discharge, their impact is minor and not a design limiting event. Thus, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to return the boron concentration to within limits.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-6

Accumulators B 3.5.1 BASES (continued)

ACTIONS B.1 (continued)

If one accumulator is inoperable for a reason other than boron concentration, the accumulator must be returned to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In this Condition, the required contents of one accumulator cannot be assumed to reach the core during a LOCA. Due to the severity of the consequences should a LOCA occur in these conditions, the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time to open the valve, remove power to the valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt action will be taken to return the inoperable accumulator to OPERABLE status. The Completion Time minimizes the potential for exposure of the plant'o a LOCA under these conditions.

C. 1 and C.2 If the accumulator cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and pressurizer pressure reduced to ~ 1600 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

D.1 If both accumulators are inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-7

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Accumulators B 3.5.1 BASES (continued)

SURVEILLANCE SR 3.5.1.1 RE(UIREHENTS Each accumulator motor-operated isolation valve should be verified to be fully open every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Use of control board indication for valve position is an acceptable verification. This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the RCS would be reduced. Although a motor operated valve position should not change with power removed, a closed valve could result in not meeting accident analyses assumptions. This Frequency is considered reasonable in view of other administrative controls that ensure a mispositioned isolation valve is unlikely.

SR 3.5.1.2 and SR 3.5.1.3 The borated water volume and nitrogen cover pressure should be verified every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for each accumulator. This Frequency is sufficient to ensure adequate injection during a LOCA. Because of the static design of the accumulator, a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency usually allows the operator to identify changes before limits are reached. Hain control board alarms are also available for these accumulator parameters.

Operating experience has shown this Frequency to be appropriate for early detection and correction of off normal trends.

SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulator every 31 days on a STAGGERED TEST Frequency since the static design of the accumulators limits the ways in which the concentration can be changed. The 31 day STAGGERED TEST Frequency is adequate to identify changes that could occur from mechanisms such as stratification or inleakage.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-8

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Accumulators B 3.5.1 BASES (continued)

SURVEILLANCE SR 3.5.1.5 RE(UIREHENTS (continued) Verification every 31 days that power is removed from each accumulator isolation valve operator when the pressurizer pressure is > 1600 psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, no accumulators would be available for injection if the LOCA were to occur in the cold leg containing the only OPERABLE accumulator. Since power is removed under administrative control and valve position is verified every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the 31 day Frequency will provide adequate assurance that power is removed.

REFERENCES 1. Branch Technical Position (BTP) ICSB-18 "Application of the Single Failure Criterion to Hanually-Controlled Electrically Operated Valves."

2. Letter from D. H. Crutchfield, NRC, to J. E, Haier, RGLE,

Subject:

"SEP Topics VI-7.F, VII-3, VII-6, and VIII-2," dated June 24, 1981.

3. Letter from R. A. Purple, NRC, to L, D. White, RG&E,

Subject:

" Issuance of Amendment 7 to Provisional Operating License No. DPR-18," dated Hay 14, 1975.

4. UFSAR, Section 6.3.

5. 10 CFR 50.46.

6. UFSAR, Section 3. 11.

7. 10 CFR 50.49.

8. UFSAR, Section 6.2.

9. UFSAR, Section 15.6.

R.E. Ginna Nuclear Power Plant B 3.5-9

ECCS Operating 8 3.5.2 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.2 ECCS Operating BASES BACKGROUND The function of the ECCS is to provide core cooling and negative reactivity to ensure that the reactor core is protected after any of the following accidents:

a. Loss of coolant accident (LOCA) and coolant leakage greater than the capability of the normal charging

=

system;

b. Rod ejection accident;

c. Loss of secondary coolant accident, including uncontrolled steam release or loss of feedwater; and

d. Steam generator tube rupture (SGTR).

The addition of negative reactivity is designed primarily for the loss of secondary coolant accident where primary cooldown could add enough positive reactivity to achieve criticality and return to significant power.

There are two phases of ECCS operation: cold leg injection and cold leg recirculation. In the injection phase, water is taken from the refueling water storage tank (RWST) and injected into the Reactor Coolant System (RCS) through the cold legs and reactor vessel upper plenum. When sufficient water is removed from the RWST to ensure that enough boron has been added to maintain the reactor subcritical and the containment sump has enough water to supply the required net positive suction head to the ECCS pumps, suction is switched to Containment Sump B for cold leg recirculation. After approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, simultaneous ECCS injection is used to reduce the potential for boiling in the top of the core and any resulting boron precipitation.

The ECCS consists of two separate subsystems: safety injection (SI) and residual heat removal (RHR). Each subsystem consists of two redundant, 100% capacity trains.

The ECCS accumulators and the RWST are also part of the ECCS, but are not considered part of an ECCS flow path as described by this LCO.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-10

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ECCS Operating B 3.5.2 BASES (continued)

BACKGROUND The ECCS flow paths which comprise the redundant trains (continued) consist of piping, valves, heat exchangers, and pumps such that water from the RWST can be injected into the RCS following the accidents described in this LCO. The major components of each subsystem are the RHR pumps, heat exchangers, and the SI pumps. The RHR subsystem consists of two 100% capacity trains that are interconnected and redundant such that either train is capable of supplying 100% of the flow required to mitigate the accident consequences. The SI subsystem consists of three redundant, 50% capacity pumps which supply two RCS cold leg injection lines. Each injection line is capable of providing 100% of the flow required to mitigate the consequences of an accident. These interconnecting and redundant subsystem designs provide the operators with the ability to utilize components from opposite trains to achieve the required 100%

flow to the core.

During the injection phase of LOCA recovery, suction headers

'upply water from the RWST to the ECCS pumps. A common supply header is used from the RWST to the safety injection (SI) and Containment Spray System pumps. This common supply header is provided with two in-series motor-operated isolation valves (896A and 896B) that receive power from separate sources for single failure considerations. These isolation valves are maintained open with DC control power removed via a key switch located in the control room. The removal of DC control power eliminates the most likely causes for spurious valve actuation while maintaining the capability to manually close the valves from the control room during the recirculation phase of the accident (Ref.

1). The SI pump supply header also contains two parallel motor-operated isolation valves (825A and 825B) which are maintained open by removing AC power. The removal of AC power to these isolation valves is an acceptable design against single failures that could result in undesirable component actuation (Ref. 2).

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-11

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ECCS Operating B 3.5.2 BASES (continued)

BACKGROUND A separate supply header is used for the residual heat (continued) removal (RHR) pumps. This supply header is provided with a check valve (854) and motor operated isolation valve (856) which is maintained open with DC control power removed via a key switch located in the control room. The removal of DC control power eliminates the most likely causes for spurious valve actuation while maintaining the capability to manually close the valve from the control room during the recirculation phase of the accident (Ref. 3).

The three SI pumps feed two RCS cold leg injection lines.

SI Pumps A and B each feeds one of the two injection lines while SI Pump C can feed both injection lines. The discharge of SI Pump C is controlled through use of two normally open parallel motor operated isolation valves (871A and 871B). These isolation valves are designed to close based on the operating status of SI Pumps A and B to ensure that SI Pump C provides the necessary flow through the RCS cold leg injection line containing the failed pump.

The discharges of the two RHR pumps and heat exchangers feed a common injection line which penetrates containment. This line then divides into two redundant core deluge flow paths each containing a normally closed motor operated isolation valve (852A and 852B) and check valve (853A and 853B) which provide injection into the reactor vessel upper plenum.

For LOCAs that are too small to depressurize the RCS below the shutoff head of the SI pumps, the steam generators provide core cooling until the RCS pressure decreases below the SI pump shutoff head.

During the recirculation phase of LOCA recovery, RHR pump suction is manually transferred to Containment Sump B (Refs.

4 and 5). This transfer is accomplished by stopping the RHR pumps, isolating RHR from the RWST by closing motor operated isolation valve 856, opening the Containment Sump B motor operated isolation valves to RHR (850A and 850B) and then starting the RHR pumps. The SI and Containment Spray System pumps are then stopped and the RWST isolated by closing motor operated isolation valve 896A and 896B for the SI and Containment Spray System pump common supply header and closing motor operated isolation valve 897 or 898 for the SI pumps recirculation line.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-12

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ECCS Operating B 3.5.2 BASES (continued)

BACKGROUND The RHR pumps then supply the SI and Containment Spray (continued) System pumps (as needed for pressure control purposes) if the RCS pressure remains above the RHR pump shutoff head (Ref. 6). This high-head recirculation path is provided through RHR motor operated isolation valves 857A, 857B, and 857C. These isolation valves are interlocked with valves 896A, 896B, 897, and 898. This interlock prevents opening of the RHR high-head recirculation isolation valves unless either 896A or 896B are closed and either 897 or 898 are closed. If RCS pressure is less than approximately 140 psig, the SI and Containment Spray pumps remain in pull-stop and only RHR is used to provide core cooling. During recirculation, flow is discharged through the same paths as the injection phase. After approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, simultaneous injection by the SI and RHR pumps is used to prevent boron precipitation (Ref. 7). This consists of providing SI through the RCS cold legs and into the lower plenum while providing RHR through the core deluge valves into the upper plenum.

The two redundant flow paths from Containment Sump B to the RHR pumps also contain a motor operated isolation valve located within the sump (851A and 851B). These isolation valves are maintained open with power removed to improve the reliability of switchover to the recirculation phase. The operators for isolation valves 851A and 851B are also not qualified for containment post accident conditions. The removal of AC power to these isolation valves is an acceptable design against single failures that could result in an undesirable actuation (Ref. 2).

The SI subsystem of the ECCS also functions to supply borated water to the reactor core following increased heat removal events, such as a main steam line break (HSLB). The limiting design conditions occur when the negative moderator temperature coefficient is highly negative, such as at the end of each cycle.

During low temperature conditions in the RCS, limitations are placed on the maximum number of ECCS pumps that may be OPERABLE. Refer to the Bases for LCO 3.4. 12, "Low Temperature Overpressure Protection (LTOP) System," for the basis of these requirements.

(continued)

R.E. Ginna Nuclear Power Plant

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ECCS Operating B 3.5.2 BASES (continued)

BACKGROUND The ECCS subsystems are actuated upon receipt of an SI (continued) signal. The actuation of safeguard loads is accomplished in a programmed time sequence. If offsite power is available, the safeguard loads start immediately in the programmed sequence. If offsite power is not available, the Engineered Safety Feature (ESF) buses shed normal operating loads and are connected to the emergency diesel generators (EDGs).

Safeguard loads are then actuated in the programmed time sequence. The time delay associated with diesel starting, sequenced loading, and pump starting determines the time required before pumped flow is available to the core following a LOCA.

The active ECCS components, along with the passive accumulators and the RWST covered in LCO 3.5. 1, "Accumulators," and LCO 3.5.4, "Refueling Water Storage Tank (RWST)," provide the cooling water necessary to meet AIF-GDC 44 (Ref. 8).

APPLICABLE The LCO helps to ensure that the following acceptance SAFETY ANALYSIS criteria for the ECCS, established by 10 CFR 50.46 (Ref. 9),

will be met following a LOCA:

a. Maximum fuel element cladding temperature is g 2200'F;

b. Maximum cladding oxidation is < 0. 17 times the total cladding thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is g 0.01 times the hypothetical amount generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react;

d. Core is maintained in a eoolable geometry; and

e. Adequate long term core cooling capability is maintained.

The LCO also limits the potential for a post trip return to power following an MSLB event and helps ensure that containment temperature limits are met post accident.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-14

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ECCS Operating 8 3.5.2 BASES (continued)

APPLICABLE Both ECCS subsystems are taken credit for in a large break SAFETY ANALYSIS LOCA event at full power (Refs. 6 and 10), This event (continued) establishes the requirement for runout flow for the ECCS pumps, as well as the maximum response time for their actuation. The SI pumps are credited in a small break LOCA event. This event establishes the flow and discharge head at the design point for the pumps. The SGTR and HSLB events also credit the SI pumps. The OPERABILITY requirements for the ECCS are based on the following LOCA analysis assumptions:

a ~ A large break LOCA event, with loss of offsite power and a single failure disabling one RHR pump (both EDG trains are assumed to operate due to requirements for modeling full active containment heat removal system operation); and

b. A small break LOCA event, with a loss of offsite power and a single failure disabling one ECCS train.

During the blowdown stage of a LOCA, the RCS depressurizes as primary coolant is ejected through the break into the containment. The nuclear reaction is terminated either by moderator voiding during large breaks or control rod insertion for small breaks. Following depressurization, emergency cooling water is injected by the SI pumps into the cold legs, flows into the downcomer, fills the lower plenum, and refloods the core. The RHR pumps inject directly into the core barrel by upper plenum injection.

The effects on containment mass and energy releases are accounted for in appropriate analyses (Refs. 10 and 11).

The LCO ensures that an ECCS train will deliver sufficient water to match boiloff rates quickly enough to minimize the consequences of the core being uncovered following a large LOCA. It also ensures that the SI pumps will deliver sufficient water and boron during a small LOCA to maintain core subcriticality. For smaller LOCAs, the SI pumps deliver sufficient fluid to maintain RCS inventory. For a small break LOCA, the steam generators continue to serve as the heat sink, providing part of the required core cooling.

The ECCS trains satisfy Criterion 3 of the NRC Policy Statement.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-15

ECCS Operating B 3.5.2 BASES (continued)

LCO In MODES 1, 2, and 3, two independent (and redundant) ECCS trains are required to ensure that sufficient ECCS flow is available, assuming a single failure affecting either train.

Additionally, individual components within the ECCS trains may be called upon to mitigate the consequences of other transients and accidents.

In MODES 1, 2, and 3, an ECCS train consists of an SI subsystem and an RHR subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST upon an SI signal and transferring suction to Containment Sump B. This includes securing the motor operated isolation valves as specified in SR 3.5.2. 1 in position by removing the power sources as listed below.

E IN Position Secured in Position B 825A Open Removal of AC Power 825B Open Removal of AC Power 826A Closed Removal of AC power 826B Closed Removal of AC Power 826C Closed Removal of AC Power 826D Closed Removal of AC Power 851A Open Removal of AC power 851B Open Removal of AC Power 856 Open Removal of DC Control Power 878A Closed Removal of AC Power 878B Open Removal of AC Power 878C Closed Removal of AC Power 878D Open Removal of AC Power 896A Open Removal of DC Control Power 896B Open Removal of DC Control Power The major components of an ECCS train consists of an RHR pump and heat exchanger taking suction from the RWST (and eventually Containment Sump B), and capable of injecting through one of the two isolation valves to the reactor vessel upper plenum and one of the two lines which provide high-head reci} culation to the SI and Containment Spray System pumps.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-16

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ECCS Operating B 3.5.2 BASES (continued)

LCO Also included within the ECCS train are two of three SI (continued) pumps capable of taking suction from the RWST and Containment Sump B (via. RHR), and injecting through one of the two RCS cold leg injection lines. In the case where SI Pump C is inoperable, both RCS cold leg injection lines must be OPERABLE to provide 100% of the ECCS flow equivalent to to a single train of SI due to the location of check valves 870A and 870B.

The flow path for each train must maintain its designed independence to ensure that no single failure can disable both ECCS trains.

APPLICABILITY In NODES 1, 2, and 3, the ECCS OPERABILITY requirements for the limiting Design Basis Accident, a large break LOCA, are based on full power operation. Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The SI pump performance requirements are based on a small break LOCA. MODE 2 and MODE 3 requirements are bounded by the MODE 1 analysis.

This LCO is only applicable in MODE 3 and above. Below MODE 3, the SI signal setpoint is manually bypassed by operator control, and system functional requirements are relaxed as described in LCO 3.5.3, "ECCS Shutdown."

As indicated in Note 1, the flow path may be isolated for 2 hours in MODE 3, under controlled conditions, to perform pressure isolation valve testing per SR 3.4. 14. 1. The flow path is readily restorable from the control room or field test personnel. The note also allows an SI isolation MOV to be powered for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the performance of this testing.

As indicated in Note 2, operation in MODE 3 with ECCS trains declared inoperable pursuant to LCO 3.4. 12, "Low Temperature Overpressure Protection (LTOP) System," is necessary since the LTOP arming temperature is near the MODE 3 boundary temperature of 350'F. LCO 3.4. 12 requires that certain pumps be rendered inoperable at and below the LTOP arming temperature. When this temperature is near the MODE 3 (continued)

R.E. Ginna Nuclear Power Plant B 3.5-17

ECCS Operating B 3.5.2 BASES (continued)

APPLICABILITY boundary temperature, time is needed to restore the (continued) inoperable pumps to OPERABLE status.

In MODES 4, 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Mode 4 core cooling requirements are addressed by LCO 3.4.6, "RCS Loops - Mode 4," and LCO 3.5.3, "ECCS - Shutdown." Core cooling requirements in MODE 5 are addressed by LCO 3.4,7, "RCS Loops MODE 5, Loops Filled,"

and LCO 3.4.8, "RCS Loops MODE 5, Loops Not Filled."

MODE 6 core cooling requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation High Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation Low Water Level."

ACTIONS A.1 With one train inoperable and at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available, the inoperable components must be returned to OPERABLE status within 72 hours. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on an NRC reliability evaluation (Ref. 12) and is a reasonable time for repair of many ECCS components.

An ECCS train is inoperable if it is not capable of delivering 100% design flow to the RCS. Individual components are inoperable if they are not capable of performing their design function or necessary supporting systems are not available.

The LCO requires the OPERABILITY of a number of independent subsystems. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does not render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. The intent of this Condition is to maintain a combination of equipment such that 100% of the ECCS flow equivalent to a single OPERABLE ECCS train remains available. This allows increased flexibility in plant operations under circumstances when components in opposite trains are inoperable.

(continued)

R.E. Ginna Nuclear Power Plant B 3,5-18

ECCS Operating B 3.5.2 BASES (continued)

ACTIONS In the case where SI Pump C is inoperable, both RCS cold leg (continued) injection lines must be OPERABLE to provide 100% of the ECCS flow equivalent to a single train of SI due to the location of check valves 870A and 870B.

An event accompanied by a loss of offsite power and the failure of an EDG can disable one ECCS train until power is restored. A reliability analysis (Ref. 2) has shown that the impact of having one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours.

B.l and B.2 If the inoperable train cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a HODE in which the LCO does not apply. To achieve this status, the plant must be brought to HODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1 If both trains of ECCS are inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be immediately entered. With one or more component(s) inoperable such that 100% of the flow equivalent to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analysis. Therefore, LCO 3.0.3 must be immediately entered.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-19

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ECCS Operating B 3.5.2 BASES (continued)

SURVEILLANCE SR 3.5.2.1 RE(UIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained. Use of control board indication for valve position is an acceptable verification. Misalignment of these valves could render both ECCS trains inoperable. The listed valves are secured in position by removal of AC power or key locking the DC control power. These valves are operated under administrative controls such that any changes with respect to the position of the valve breakers or key locks is unlikely. The verification of the valve breakers and key locks is performed by SR 3.5.2.3. Mispositioning of these valves can disable the function of both ECCS trains and invalidate the accident analyses. A 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered reasonable in view of other administrative controls that ensure a mispositioned valve is unlikely.

SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position, The 31 day Frequency is appropriate because the valves are operated under administrative control, and an improper valve position in most cases, would only affect a single train. This Frequency has been shown to be acceptable through operating experience.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-20

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ECCS Operating 8 3.5.2 BASES (continued)

SURVEILLANCE SR 3.5.2.3 REQUIREHENTS (continued) Verification every 31 days that AC or DC power is removed, as appropriate, for each valve specified in SR 3.5.2. 1 ensures that an active failure could not result in an undetected misposition of a valve which affects both trains of ECCS. If this were to occur, no ECCS injection or recirculation would be available. Since power is removed under administrative control and valve position is verified every 12 hours, the 31 day Frequency will provide adequate assurance that power is removed.

SR 3.5.2.4 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by Section XI of the ASHE Code. This type of testing may be accomplished by measuring the pump developed head at a single point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant safety analysis. SRs are specified in the Inservice Testing Program, which encompassesSection XI of the ASHE Code. Section XI of the ASHE Code provides the activities and Frequencies necessary to satisfy the requirements.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-21

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ECCS Operating B 3.5.2 BASES (continued)

SURVEILLANCE SR 3.5.2.5 and SR 3.5.2.6 RE(UIREMENTS (continued) These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SI signal and that each ECCS pump starts on receipt of an actual or simulated SI signal. This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 24 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned plant transients if the Surveillances were performed with the reactor at power. The 24 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of ESF Actuation System testing, and equipment performance is monitored as part of the Inservice Testing Program.

REFERENCES 1. Letter from R. A. Purple, NRC, to L. D. White,.RG&E,

Subject:

"Issuance of Amendment 7 to Provisional Operating License No. DPR-18," dated Hay 14, 1975.

2. Branch Technical Position (BTP) ICSB-18, "Application of the Single Failure Criterion to Manually-Controlled Electrically Operated Valves."

3. Letter from A. R. Johnson, NRC, to R. C. Hecredy, RG&E,

Subject:

"Issuance of Amendment No. 42 to Facility Operating License No. DPR-18, R. E. Ginna Nuclear Power Plant (TAC No. 79829)," dated June 3, 1991.

4. Letter from D. H. Crutchfield, NRC, to J. E. Haier, RG&E,

Subject:

"SEP Topic VI-7.B: ESF Switchover from Injection to Recirculation Mode, Automatic ECCS Realignment, Ginna," dated December 31, 1981.

5. NUREG-0821.

6. UFSAR, Section 6.3.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-22

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ECCS Operating B 3.5.2 BASES (continued)

REFERENCES 7. Letter from D. H. Crutchfield, NRC, to J. E. Haier, (continued) RGLE,

Subject:

"SEP Topic IX-4, Boron Addition System, R. E. Ginna," dated August 26, 1981.

8. Atomic Industrial Forum (AIF) GDC 44, Issued forcomment July 10, 1967.

9. 10 CFR 50.46.

10. UFSAR, Section 15.6.

ll. UFSAR, Section 6.2.

12. NRC Hemorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

R.E. Ginna Nuclear Power Plant B 3.5-23

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ECCS Shutdown B 3.5.3 8 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.3 ECCS Shutdown BASES BACKGROUND The Background section for Bases 3.5.2, "ECCS Operating,"

is applicable to these Bases, with the following modifications.

In MODE 4, the required ECCS train consists of two separate subsystems: safety injecti'on (SI) and residual heat removal (RHR).

The ECCS flow paths consist of piping, valves, heat exchangers, and pumps such that water from the refueling water storage tank (RWST) can be injected into the Reactor Coolant System (RCS) following the accidents described in Bases 3.5.2. The RHR subsystem must also be capable of taking suction from containment Sump B to provide recirculation.

APPLICABLE The Applicable Safety Analyses section of Bases 3.5.2 also SAFETY ANALYSES applies to this Bases section.

Due to the stable conditions associated with operation in MODE 4 and the reduced probability of occurrence of a Design Basis Accident (DBA), the ECCS operational requirements are reduced. It is understood in these reductions that certain automatic safety injection (SI) actuation is not available.

In this MODE, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a DBA (Ref. I).

Only one train of ECCS is required for MODE 4. This requirement dictates that single failures are not considered during this MODE of operation. The ECCS trains satisfy Criterion 3 of the NRC Policy Statement.

LCO In MODE 4, one of the two independent (and redundant) ECCS trains is required to be OPERABLE to ensure that sufficient ECCS flow is available to the core following a DBA.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-24

ECCS Shutdown B 3.5.3 BASES (continued)

LCO In MODE 4, an ECCS train consists of an SI subsystem and an (continued) RHR subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST and transferring suction to the containment sump. The major components of an ECCS train during MODE 4 consists of an RHR pump and heat exchanger, capable of taking suction from the RWST (and eventually Containment Sump B), and able to inject through one of the two isolation valves to the reactor vessel upper plenum.

Also included within the ECCS train are one of three SI pumps capable of taking suction from the RWST and injecting through one of the two RCS cold leg injection lines. The high-head recirculation flow path from RHR to the SI pumps is not required in the MODE 4 since there is no accident scenario which prevents depressurization to the RHR pump shutoff head prior to depletion of the RWST.

Based on the time available to respond to accident conditions during MODE 4, ECCS components are OPERABLE if they are capable of being reconfigured to the injection mode from the control room within 10 minutes. This includes taking credit for an RHR pump and heat exchanger as being OPERABLE if they are being used for shutdown cooling purposes.

APPLICABILITY In MODES 1, 2, and 3, the OPERABILITY requirements for ECCS are covered by LCO 3.5.2.

In MODE 4 with RCS temperature below 350'F, one OPERABLE ECCS train is acceptable without single failure consideration, on the basis of the stable reactivity of the reactor and the limited core cooling requirements.

In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops MODE 5, Loops Filled,"

and LCO 3.4.8, "RCS Loops MODE 5, Loops Not Filled."

MODE 6 core cooling requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation High Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation Low Water Level."

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-25

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ECCS Shutdown B 3.5.3 BASES (continued)

ACTIONS A.1 With no ECCS RHR subsystem OPERABLE, the plant is not prepared to respond to a loss of coolant accident or to continue a cooldown using the RHR pumps and heat exchangers.

The Completion Time of immediately to initiate actions that would restore at least one ECCS RHR subsystem to OPERABLE status ensures that prompt action is taken to restore the required cooling capacity. Normally, in MODE 4, reactor decay heat is removed from the RCS by an RHR loop. If no RHR loop is OPERABLE for this function, reactor decay heat must be removed by some alternate method, such as use of the steam generators. The alternate means of heat removal must continue until the inoperable RHR loop components can be restored to operation so that decay heat removal is continuous.

With both RHR pumps and heat exchangers inoperable, it would be unwise to require the plant to go to MODE 5, where the only available heat removal system is the RHR. Therefore, the appropriate action is to initiate measures to restore one ECCS RHR subsystem and to continue the actions until the subsystem is restored to OPERABLE status.

B.l With no ECCS SI subsystem OPERABLE, due to the inoperability of the SI pump or flow path from the RWST, the plant is not prepared to provide high pressure response to Design Basis Events requiring SI. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time to restore at least one SI subsystem to OPERABLE status ensures that prompt action is taken to provide the required cooling capacity or to initiate actions to place the plant in MODE 5, where an ECCS train is not required.

C.1 When the Required Actions of Condition B cannot be completed within the required Completion Time, a controlled shutdown should be initiated. Twenty-four hours is a reasonable time, based on operating experience, to reach MODE 5 in an orderly manner and without challenging plant systems or operators.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-26

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ECCS Shutdown B 3.5.3 BASES (continued)

SURVEILLANCE SR 3.5.3.1 REOUIREHENTS The applicable Surveillance description from Bases 3.5.2 apply. This SR is modified by a Note that allows an RHR train to be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned (remote or local) to the ECCS mode of operation and not otherwise inoperable. This allows operation in the RHR mode during HODE 4, if necessary.

REFERENCES The applicable references from Bases 3.5.2 apply.

1. WCAP-12476, "Evaluation of LOCA During Hode 3 and Hode 4 Operation for Westinghouse NSSS," November 1991.

R.E. Ginna Nuclear Power Plant B 3.5-27

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RWST B 3.5.4 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.4 Refueling Water Storage Tank (RWST)

BASES BACKGROUND The RWST supplies borated water to both trains of the ECCS and the Containment Spray System during the injection phase of a loss of coolant accident (LOCA) recovery. A common supply header is used from the RWST to the safety injection (SI) and Containment Spray System pumps. A separate supply header is used for the residual heat removal (RHR) pumps.

Isolation valves and check valves are used to isolate the RWST from the ECCS and Containment Spray System prior to transferring to the recirculation mode. The recirculation mode is entered when pump suction is transferred to the containment sump based on RWST level. Use of a single RWST to supply both trains of the ECCS and Containment Spray System is acceptable since the RWST is a passive component, and passive failures are not required to be assumed to occur coincidentally with Design Basis Events.

The RWST is located in the Auxiliary Building which is normally maintained between 50'F and 104'F (Ref. 1). These moderate temperatures provide adequate margin with respect to potential freezing or overheating of the borated water contained in the RWST.

During normal operation in NODES 1, 2, and 3, the safety injection (SI) and residual heat removal (RHR) and Containment Spray System pumps are aligned to take suction from the RWST.

The ECCS and Containment Spray System pumps are provided with recirculation lines that ensure each pump can maintain minimum flow requirements when operating at or near shutoff head conditions. The recirculation lines for the RHR and Containment Spray System pumps are directed from the discharge of the pumps to the pump suction. The recirculation lines for the SI pumps are directed back to the RWST.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-28

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RWST B 3.5.4 BASES (continued)

BACKGROUND When the suction for the ECCS and Containment Spray System (continued) pumps is transferred to the containment sump, the RWST and SI pump recirculation flow paths must be isolated to prevent a release of the containment sump contents to the RWST, which could result in a release of contaminants to the Auxiliary Building and the eventual loss of suction head for the ECCS pumps.

This LCO ensures that:

a. The RWST contains sufficient borated water to support the ECCS during the injection phase;

b. Sufficient water volume exists in the containment sump to support continued operation of the ECCS and Containment Spray System pumps at the time of transfer to the recirculation mode of cooling; and

c. The reactor remains subcritical following a LOCA.

Insufficient water in the RWST could result in inadequate NPSH for the RHR pumps when the transfer to the recirculation mode occurs. Improper boron concentrations could result in a reduction of SDM or excessive boric acid precipitation in the core following the LOCA, as well as excessive caustic stress corrosion of mechanical components and systems inside the containment.

APPLICABLE During accident conditions, the RWST provides a source of SAFETY ANALYSES borated water to the ECCS and Containment Spray System pumps. As such, it provides containment cooling and depressurization, core cooling, and replacement inventory and is a source of negative reactivity for reactor shutdown (Ref. 3). The design basis transients and applicable safety analyses concerning each of these systems are discussed in the Applicable Safety Analyses section of B 3.5.2, "ECCS-Operating"; B 3.5.3, "ECCS Shutdown"; and B 3.6.6, "Containment Spray and Cooling Systems." These analyses are used to assess changes to the RWST in order to evaluate their effects in relation to the acceptance limits in the analyses.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-29

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RWST B 3.5.4 BASES (continued)

APPLICABLE The RWST must also meet volume, boron concentration, and SAFETY ANALYSIS temperature requirements for non-LOCA events. The volume is (continued) not an explicit assumption in non-LOCA events since the volume required for Reactor Coolant System (RCS) makeup is a small fraction of the available RCS volume. The deliverable volume limit is set by the LOCA and containment analyses.

For the RWST, the deliverable volume is selected such that switchover to recirculation does not occur until sufficient water has been pumped into containment to provide necessary NPSH for the RHR pumps. The minimum boron concentration is an explicit assumption in the main steam line break (HSLB) analysis to ensure the required shutdown capability. The maximum boron concentration is an explicit assumption in the evaluation of chemical effects resulting from the operation of the Containment Spray System.

For a large break LOCA analysis, the minimum water volume limit of 300,000 gallons and the lower boron concentration limit of 2000 ppm are used to compute the post LOCA sump boron concentration necessary to assure subcriticality. The large break LOCA is the limiting case since the safety analysis assumes that all control rods are out of the core.

The upper limit on boron concentration of 2900 ppm is used to determine the time frame in which boron precipitation is addressed post LOCA. The maximum boron concentration limit is based on the coldest expected temperature of the RWST water volume and on chemical effects resulting from operation of the ECCS and the Containment Spray System. A value of 2,900 ppm would not create the potential for boron precipitation in the RWST assuming an Auxiliary Building temperature of 50'F (Ref. 1). Analyses performed in response to 10 CFR 50,49 (Ref. 2) assumed a chemical spray solution of 2000 to 3000 ppm boron concentration (Ref, 1) which provides a margin of 100 ppm. The chemical spray solution impacts sump pH and the resulting effect of chloride and caustic stress corrosion on mechanical systems and components. The sump pH also affects the rate of hydrogen generation within containment due to the interaction of Containment Spray and sump fluid with aluminum components.

The RWST satisfies Criterion 3 of the NRC Policy Statement.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-30

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RWST B 3.5.4 BASES (continued)

LCO The RWST ensures that an adequate supply of borated water is available to cool and depressurize the containment in the event of a Design Basis Accident (DBA), to cool and cover the core in the event of a LOCA, to maintain the reactor subcritical following a DBA, and to ensure adequate level in the containment sump to support ECCS and Containment Spray System pump operation in the recirculation mode.

To be considered OPERABLE, the RWST must meet the water volume and boron concentration limits established in the SRs.

APPLICABILITY In MODES 1, 2, 3, and 4, RWST OPERABILITY requirements are dictated by ECCS and Containment Spray System OPERABILITY requirements. Since both the ECCS and the Containment Spray System must be OPERABLE in MODES 1, 2, 3, and 4, the RWST must also be OPERABLE to support their operation. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation High Water Level,"

and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation Low Water Level."

ACTIONS A.l With RWST boron concentration not within limits, it must be returned to within limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Under these conditions neither the ECCS nor the Containment Spray System can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE condition.

The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> limit to restore the RWST boron concentration to within limits was developed considering the time required to change the boron concentration and the fact that the contents of the tank are still available for injection.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-31

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RWST 8 3.5.4 BASES (continued)

ACTIONS B.l (continued)

With the RWST water volume not within limits, it must be restored to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In this Condition, neither the ECCS nor the Containment Spray System can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE status or to place the plant in a HODE in which the RWST is not required, The short time limit of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to restore the RWST to OPERABLE status is based on this condition simultaneously affecting redundant trains.

C. 1 and C.2 If the RWST cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a NODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least HODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to HODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.5.4.1 REgUIREHENTS The RWST water volume should be verified every 7 days to be above the required minimum level in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation. Since the RWST volume is normally stable and the RWST is located in the Auxiliary Building which provides sufficient leak detection capability, a 7 day Frequency is appropriate and has been shown to be acceptable through operating experience.

(continued)

R.E. Ginna Nuclear Power Plant B 3.5-32

I H ~ y~ 0 H 'I a 1

RWST B 3.5.4 BASES (continued)

SURVEILLANCE SR 3.5.4.2 RE(UIREMENTS (continued) The boron concentration of the RWST should be verified every 7 days to be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA.

Further, it assures that the resulting sump pH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized. Since the RWST volume is normally stable, a 7 day sampling Frequency to verify boron concentration is appropriate and has been shown to be acceptable through operating experience, REFERENCES 1. UFSAR, Section 3. 11.

2. 10 CFR 50.49.

3. ,UFSAR, Section 6.3 and Chapter 15.

R.E. Ginna Nuclear Power Plant B 3.5-33

ATTACHMENT D Marked Up Copy of Improved Technical Specifications (NUREG-1431)

Included pages:

All pages contained in NUREG-1431.

ONLY ITS 3.5 IS PROVIDED AT THIS TIME

Accumulators 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators

~~0 LCO 3.5.1 +Fet+PECCS accumulators shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, 1 400 MODE 3 with pressurizer pressure > +NO~ psig.

h ACTIONS CONDITION RE(U I RED ACTION COMPLETION TIME A. One accumulator A. 1 Restore boron 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable due to concentration to boron concentration within limits.

not within limits.

B. One accumulator B.l Restore accumulator 1 hour inoperable for reasons to OPERABLE status.

other than Condition A.

C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time of Condition A AND or B not met.

C.2 Reduce pressurizer 12 hours pressure to

~ @40@ psig.

~4oo D. Two e~m~ 0.1 Enter LCO 3.0.3. Immediately accumulators inoperable.

3.5-1

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS

~

SURVEILLANCE FREQUENCY 5i.'t 3.5. 1. 1 Mac o ~c4 SR Verify each accumulator isolation valve is 12 hours fully open.

SR 3.5.1.2 Verify borated water volume in each 12 hours t t d Ill.a ~Q;~~P (54 Ta)

>>RO ~S'c C + (ZX I>)

SR 3.5. 1.3 Verify nitrogen cover ressure in each 12 hours accumulator is > psig and gPRR-Tpssg. ~no 790 SR 3.5.1.4 Verify boron concentration in each 31 days on ~ Rw~ re cava accumulator is > -@9~pm and e, Rs'<f

< ~~pm. iso+ AND GL. cil.

?.,0 oo

---

NOTE------

nly required be performe fr affected ac umulators Once ithi 6 hou ter each s l tion volume ncrease of

> (( ga lons,

( )% of ind'cated le el] that is n the resu t addition rom the refueling water storage tank (continued) 3.5-2

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FRE(UENCY SR 3.5.1.5 Verify power is removed from each 31 days accumulator isolation valve o erator when pressurize pressure is pslg.

>/4~

(Q ~+r o~dk 3.5-3

ECCS Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

NOTES 1 In MODE 3, both safety injection (SI) pump flow paths may be isolated by 'closing the isolation valves for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to perform pressure isolation valve testing per SR 3.4.14.1.v-sa~~>s <<

S gg. V a.'e Operation in MODE 3 with pumps declared inoperable pursuant to LCO 3.4. 12, "Low Temperature Overpressure Protection (LTOP) System," is allowed for up to 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> s or until the temperature of ~E ~w RCS cold legs exceeds +75@F, whichever comes first.

ACTIONS CONDITION RE(UI RED ACTION COMPLETION TIME 4

sz.vi.~ A. One inoperab e.

trai~ A.l Restore OPERABLE train status.

to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.

B. Required Action and B.l Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in NODE 4. 12 hours 3a 5 3.5-4

Insert 3.5.3 (sQ C. Two trains C. 1 Enter LCO '3.0.3 immediately ino erable.

Insert 3.5.14 Power may be restored to motor operated isolation valves 878A, 878B, 878C, and 878D for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the purpose of testing per SR 3.4. 14. 1 provided that power is restored to only one valve at a time.

I I ~

I I I I I

~ ~ N ~ ~

II ~ I I

I ~

' ~ I I I I I I I I I ~ I I I I ~

~ I ~

I ~ ~

' I ~ ~

~

' ~

I I ~ ~ I I I

~ I %AIL i~a i ~

~ ~

Insert 3.5.4 EIN Position Function 825A Open RWST Suction to SI Pumps 825B Open RWST Suction to SI Pumps 826A Closed BAST Suction to SI Pumps 826B Closed BAST Suction to SI Pumps 5Z.v ii'i 826C Closed BAST Suction to SI Pumps 826D Closed BAST Suction to SI Pumps 851A Open Sump B Suction to RHR Pumps 851B Open Sump B Suction to RHR Pumps 856 Open RWST Suction to RHR Pumps 878A Closed SI Injection to RCS Hot Leg 878B Open SI Injection to RCS Cold Leg 878C Closed SI Injection to RCS Hot Leg 878D Open SI Injection to RCS Cold Leg 896A Open RWST Suction to SI and Spray 896B Open RWST Suction to SI and Spray Insert 3.5.15 SR 3.5.2.3 Verify the breaker or key switch, as 31 days applicable, for each valve listed in SR 3.5.2. 1, is in the correct position.

ECCS Operating 3.5.2 SURVEILLANCE REQUIREMENTS continued SURVEILLANCE FREQUENCY SR 3. . .7 Verify, for each ECCS throttle valve [18] m s listed below, each position stop is e correct position.

Valve Numbe

[ ]

SR 3.5.2.8 isual inspection, each '

ECCS onths train containmen etio '

not restri cted by d e let t s and screens show no evidence o structural distress or abnormal corrosion.

3.5-6

ECCS Shutdown 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEHS (ECCS) 3.5.3 ECCS Shutdown LCO 3.5.3 One ECCS train shall be OPERABLE.

APPLICABILITY: MODE 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

. Required ECCS residual A. I Initiate action to Immediately heat removal (RHR) restore required ECCS subsystem inoperable. RHR subsystem to OPERABLE status.

s'<4'n)~o~ ()T)

B. Required ECCS B. I Restore required ECCS I hour head. subsystem+ S'Z subsyste~

inoperable. to OPERABLE status.

C. Required Action and C.l Be in MODE 5. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> associated Completion Time +of Condition EP not met.

3.5-7

ECCS Shutdown 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 -NOTE An RHR train may be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned to the ECCS mode of operation.

The fol1owiog Sibr aapplicable for all In accordance equipment required to be OPERABLE: with applicable S~

SR 3.5.2.4 3.5-8

RWST 3.5.4 3.5 EHERGENCY CORE COOLING SYSTEHS (ECCS) 3.5.4 Refueling Water Storage Tank (RWST)

LCO 3.5.4 The RWST shall be OPERABLE.

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

ACTIONS CONDITION RE(UIRED ACTION COHPLETION TIHE A. RWST boron A. 1 Restore RWST to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> concentration not OPERABLE status.

within limits.

RWST boor temperature not

'ts.

ater

'in i valgus~

B. RWST rabl B.l Restore RWST to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> reason e OPERABLE status.

ition A

~ ~ ~sW

~( ~ s~>~

C. Required Action and C.l Be in HODE 3. 6 hours associated Completion Time not met. AND o4 c ~&in~ 4 oc- <

C.2 Be in HODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 3.5-9

RWST 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. NOTE uired to ' be performed

' wh ent air tempera 3 [100]'F.

i y RWST borated water temperature is hours

> [35]'F and g [100] F.

SR 3.5.4~ Verify RWST borated water volume is 7 days 7

3oo, ooo q~~g (3K I,)

SR 3.5.4.3 Veri y RWST boron concentration is 7 days 000~pm and Z~98j ppm.

Z,QOO 3.5-10

Seal Injection Flow 3.5.

3.5 ERGENCY CORE COOLING 'SYSTEMS (ECCS) 3.5.5 S 1 Injection Flow LCO 3.5.5 Reactor coolant pump seal injection flow shall be [40] gpm with [centrifugal charging pump discharge header pressure

> [2480] psig and the [charging flow] control v ve full open.

APPLICABILITY: MODE I, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Seal injection flow A. Adjust man al seal 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> not within limit. injection throttle valves give a flow within imit with cent ifugal charging p mp discharge he er] pressure 80] psig and the cha ing flow]

contro valve full open.

B. Required Action and . I Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 4. l2 hours 3.5-11

(Qs Seal Injection Flow 3.5.

SUR ILLANCE REQUIREMENTS SURVEILLANCE FREQU CY SR 3.5.5. NOTE--

Not required to be performed until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the Reactor Coolant System pressure stabilizes at > [2215 psig and 2255 psig].

Veri manual seal injection throttle 31 days valve are adjusted to give a flow withi limit 'th [centrifugal charging pump dischar header] pressure > [2480] p g and the [ arging flow] control valv full open.

3.5-12

3.5 E RGENCY CORE COOLING SYSTEMS (ECCS) 3.5.6 Bo on Injection Tank (BIT)

LCO 3.5.6 The BIT shall be OPERABLE.

APPLICABILITY: M ES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTIO COMPLETION TIME A. BIT inoperable. Restore B to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> OPERABL status.

B. Required Action and 8.1 8 in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND not met.

B.2 Bora e to an SDM 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> equiv lent to

[l]% h k at 200'F.

AND B.3 Restore BIT o 7 days OPERABLE stat s.

C. Required Acti n and C.1 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated mpletion Time of Co dition B not met.

3.5-13

8 3..6 0

SURVEI SR l

ANCE REQUIREMENTS 3.5.6.1 SURVEILLANCE Verify BIT borated water temperature is 24 FRE our s ENCY Z [145]'F.

SR 3.5.6.2 Veri y BIT borated water volume is 7 days z [11 0] gallons.

SR 3.5.6.3 Verify BI~T )oron concentration is 7 days

> [20,000] p m and < [22,500] pp .

3.5-14

Accumulators B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.1 Accumulators BASES BACKGROUND The functions of the ECCS accumulators are to supply water to the reactor vessel during the blowdown phase of a loss of coolant accident (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide Reactor Coolant System (RCS) makeup for a small break LOCA.

The blowdown phase of a large break LOCA is the initial period of the transient during which the RCS departs from

~~ ~~E in,e~ru

<<a~ ~g Cc of a.n%

W Va.~a.4wq equilibrium conditions, and heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. The blowdown phase of the transient ends when the RCS pressure falls to a value approaching that of the containment atmosphere.

@ho.su ~~+ c,w

~4 ~~%a+. refill QL~4,~q ou+ ~au N ~ o&~ In the blowdown phase, phase of a LOCA, which immediately follows the

~

n-ed@-@he-eugh-4haa, 5'I. <v. ~ kremlin.

balance of

~e core is essentially in adiabatic heatup.

accumulator inventory is available to elp fill voids in the lower plenum and reactor vessel downcomer so as,to establish a recovery level at the bottom core.

f the The accumulators are pressure vessels partially filled wi.th borated water and pressurized with nitrogen gas. The accumulators are passive components, since no operator or control actions are required in order for them to perform their function. Internal accumulator tank pressure is sufficient to discharge the accumulator contents to the RCS, if RCS pressure decreases below the accumulator pressure.

$ 'l. i v.h w1 that the valves i

w's Each accumulator is piped into an RCS cold leg via an accumulator line and is isolated from the RCS by a motor operated isolation valve and two check valves in serie e mo or o era e iso a ion va ve are interloc

'zer pressure me a ove the permissive circuit P-annels to ensure RCS pressure (continued)

B 3.5-1

Insert 3.5.1 (841 and 865) are maintained open with AC power removed under administrative control when pressurizer pressure is > 1600 psig. This feature ensures that the valves meet the single failure criterion of manually-controlled electrically operated valves per Branch Technical Position (BTP) ICSB-18 (Ref. 1) This is also discussed in References 2 and 3 ~

Accumulators B 3.5.1 BASES BACKGROUND T>>i 'rlock also prevents inadvertent closure e (continued) valves dur valves will automa 'pen, rmal operation prior to SI signal. These features h

)dent. The

, as a result of an e that the valves meet the requirements of the astute of 'cal and Electronic Engineers I tandard 279-1971 (Ref. "operating bypa and that the accumulators will be availa r

'ection without reliance on operator action.

one Mo The accumulator size, water(volume, and nitro en cover ressure are selected so that~g~of t e accumulatorsi~

5l. iv.a. sufficient to partially cover the core before

~

significant clad melting or zirconium water reaction can

~

occur following q LOCA. The need to ensure that g~>u~~

accumulator~~'adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA.

APPLICABLE The accumulators are assumed OPERABLE in both the large and SAFETY ANALYSES small break LOCA analyses at full power (Ref. ~." These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits.

In performing the LOCA calculations, conservative assumptions are made concerning the availabilit of ECCS flow. In the early stages of a LOCA, wit or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is required by regulations and conservatively imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go.through their timed loading sequence. In cold leg break scenarios; the entire contents of one accumulator are assumed to be lost through the break.

The limiting large break LOCA is a double ended guillotine break at the discharge of the reactor coolant pump. During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure.

(continued)

B 3.5-2

Accumulators B 3.5.1

~V gS BASES SL g sgaoSL cgvm~+on APPLICABLE As a conservative estimate, no credit is taken for ECCS pump SAFETY ANALYSES flow until an effective delay has elapsed. This delay (continued) accounts fo the diesels starting, and the pumps being loaded and delivering full flow.

5'I. iV,Z.~

li

~

0 i g ti , h 1 ly d as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA.

The worst case small break LOCA analyses also assume a time delay before pumped flow reaches the core. For the larger range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued coolin . As break size decreases, the accumulators and s~~

c pumps both play a part in terminating 5 I .i. v, 4. e rise in cia temperature. As break size continues to decrease, the role of the accumulators continues to decrease until they are not required and the c pumps become solely responsible for terminating e g a&+ in$ ~i temperature increase.

This LCO helps to ensure that the following acceptance criteria established for the ECCS will be met following a LOCA:

by 10 CFR 50.46 (Ref. ~

a. Haximum fuel element cladding temperature is < 2200'F;

b. Haximum cladding oxidation is g 0. 17 times the total cladding thickness before oxidation;

c. Haximum hydrogen generation from a zirconium water reaction is g 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; and

d. Core is maintained in a eoolable geometry.

Since the accumulators discharge during the blowdown phase of a LOCA, they do not contribute to the long term cooling requirements of 10 CFR 50.46.

For both the large and small break LOCA analyses, a nominal contained accumulator water volume is used. The contained (continued)

B 3.5-3

~~c. hc ~~~~ ~~

Accumulators 2., wv.~ B 3.5.1 c3n,~c BASES VUL&~ ~Q. y~~Q~

APPLICABLE water volume is the same as the deliverable volume for the SAFETY ANALYSES accumulators, since the accumula ors are emptied, once (continued) discharged. For small breaks, an increase in water volume is a peak clad temperature penalty For large breaks, an increase in water volume can be either a peak clad temperature penalty or benefit, depending on downcomer filling and subsequent spill through the break during the core refloodin portion of the transient. The analysis line water volume from the accumulator to ec va ve T e safety ana ysis assenes-~F s o

[646 ga nd-g&X9] hm . o allow for instrument inaccuracy 20] ga ons Kmf-[6820]clams are e

The minimum boron concentration setpoint is used in the post LOCA boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment. Of particular interest is the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron 'n determinin s I. iv'. h m+A4%88~ump-pk, 4 ~~~ .s; a vere-~

The large and small break LOCA analyses are performed at the minimum nitrogen cover pressure, since sensitivity analyses have demonstrated that higher nitrogen cover pressure results in a computed peak clad temperature benefit.

Si.v The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation, and ultimately preserves 5'I, i v,Q accumulator integrity. + ck, 20Q pgsg The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Refs. ~nd 4f.

The accumulators satisfy Criterion 3 of the NRC Policy Statement.

(continued)

B 3.5-4

Insert 3.5.2 the time frame in which boron precipitation is addressed post LOCA. The maximum boron concentration limit is based on the coldest expected temperature of the accumulator water volume and on chemical effects resulting from operation of the ECCS and the Containment Spray System. A value of 2,900 ppm would not create the potential for boron precipitation in the accumulator assuming a Containment temperature of 60 F (Ref. 6).

Analyses performed in response to 10 CFR 50.49 (Ref. 7) assumed a chemical spray solution of 2000 to 3000 ppm boron concentration (Ref. 6) which provides a margin of 100 ppm. The chemical spray solution impacts sump pH and the resulting effect of chloride and caustic stress corrosion on mechanical systems and components. The sump pH also affects the rate of hydrogen generation within containment due to the interaction of Containment Spray and sump fluid with aluminum components.

Accumulators B 3.5.1 BASES (continued)

LCO The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. ~H accumulators are required to ensure that 100% of the St. l.u,a contents of accumulator&will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than3M~ccumulatorW~injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. @could be violated.

62.bout. 2 (el0('s'imb v'1 For an accumulator to be considered OPERABLE,) the~isolation Sl .

valve must be fully open +sWP power remove , and the limits established in the SRs for contained volume, boron concentration, and nitrogen cover pressure must be met.

APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure l( oo > ~0 psig, the accumulator OPERABILITY requirements are ased.on full power operation. Although cooling requirements decrease as power decreases, the accumulators are still required to provide core cooling as long as elevated RCS pressures and temperatures exist.

1(o 00 <t aa S'L. i.aCR This LCO is only applicable at pressures >~psig. At pressures g 4M~sig, the rate of RCS blowdown is such that the ECCS pumps can provide adequate injection to ensure that peak clad temperature remains below the 10 CFR 50.46 (Ref.

g limit of 2200F.

i(.oo In MODE 3, with RCS pressure < P680 psig, and in MODES 4, 5, and 6, the accumulator motor operated isolation valves are closed to isolate the accumulators from the RCS. This allows RCS cooldown and depressurization without discharging the accumulators into the RCS or requiring depressurization of the accumulators.

ACTIONS A.1 If the boron concentration of one accumulator is not within limits, it must be returned to within the limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, ability to maintain subcriticality or minimum boron precipitation time may be (continued)

B 3.5-5

Accumulators 8 3.5.1 BASES ACTIONS A.l (continued) reduced. The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially recovered core during the early r eflooding phase of a large break LOCA is sufficient to keep that portion of the core subcritical. One accumulator below the minimum boron concentration limit, however, will have no effect on 8 tax Cu ~ O ~iauap4.Ml u~

available ECCS water and an insignificant effect on core subcriticalit durin refloo Boiling of ECCS water in the Suartahg ~~

Vibhu~

~~

t2,

~Q core during re lood concentrates boron in the saturated liquid that remains in the core. In addition, current

'.~a

+c R.c S ~Rl A.mR'V analysis techniques demonstrate that the accumulators 4e.not

• discharge following a large main steam line break@em-@he e..~+

Even if they do discharge, their impact is minor and not a design limiting event. Thus, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to return the boron concentration to within limits.

B.1 If one accumulator is inoperable for a reason other than boron concentration, the accumulator must be returned to OPERABLE status withi 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In this Condition, the required contents oP accumulator~annot be assumed to reach the core during a LOCA. Due to the severity of the consequences should a LOCA occur in these conditions, the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time to open the valve, remove power to the valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt action will be taken to return the inoperable accumulator to OPERABLE status. The Completion Time minimizes the potential for exposure of the plant to a LOCA under these conditions.

C.l and C.2 If the accumulator cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a NODE in which the LCO does not apply. To achieve this status, the plant must be brought to NODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and pressurizer pressure reduced to s lggd psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

B 3.5-6

Accumulators B 3.5.1 BASES ACTIONS (continued)

SL. cv, w Tf~

D.l 1t Q.~

1 a condition outside the accident analyses; LCO 3.0.3 must be entered immediately.

p Pl,th pl therefore, t

SURVEILLANCE SR 3.5.1.1 RE(UIREMENTS ~~r ~@~~+ lRbhOWOA.

L.c vA. Each accumulatorvalve should be verified to be fully open ever 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open.

isolation valve is not fully open, the rate of injection to If an the RCS would be reduced. Although a motor operated valve SL. i.v.Q position should not change with power removed, a closed valve could result in not meeting accident analyses umph,W a~( assumptions. This Frequency is considered reasonable in t5bg 9 r.

~b v~ ~i4o~i KAKl~oA, view of other administrative controls that ensure a mispositioned isolation valve is unlikely.

s~uh4 ~u~gP ~Q

~h SR is su 3.5.1.2 Because and SR 3.5.1.3 d

f h 1

1 d

t 1

g Thl icient to ensure adequate injection during a LOCA.

of the static design of the accumulator, a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 1 d Frequency usually allows the operator to identify changes Sl.i vg before limits are reached. 0 crating experience has shown this Frequency to be appropriate for early detection and Sl.h v.w correction of off normal trends.

Boos g ~S gRbbCRal~

~~~s ~~e ~<hie SR 3.5.1.4 f P pro ~>~ concentration should be verified to be within The boron si.i~i required limits for each accumulator every 31 days since the ign of the ac~um~l~t~rs limits the wa s in whi the concentration can be changed. The 31 day requency is adequate to identify changes that could occur from mechanisms such as strati ication or inleaka e g ccumu ator wit in er a 1% volume increase' will ide eakage has caused a re oron concentration to be o (continued)

'B 3.5-7

Accumulators B 3.5.1 BASES SURVEILLANCE RE(UIREMENTS SR 3.5.1.4 the added wa (continued)

It is not necessary to verify boron concentrati

'ith storage tank (RWST), beca is within the accumul This is co

'ory is from the refue on concen the recommendation of NU

'r ontained in the RWST

'quirements.

tQ ~ Lo~ 3.5.1.5 OP4RAQ m

~ o~

+ m~~e 'tv W~ ~ah4~

~A+Mxe,t~ Verification every 31 days that power is removed from each accumulator isolation valve operator when the pressurizer

> t taCIO pressure is sig ensures that an active failure could no resu >n the undetected closure of an accumulator motor Si.iv. 0 operated isolation valve. If this were to occur, accumu a ors wou e ai e for injectio Since power is removed ash. Va~afC Pthg,+Ca~ un er administrative control the 31 day Frequency will C VtA'-,@ C.U~ provide adequate assurance that power is removed.

SR allows power to be supplied to the motor opera~

isolate alves when pressurizer pressure is <' 2 8 psig,

~ Q FAR.w~

po< th~~

~,'R

~a-chart t CAX9-

" ppli~e,~

p C~('3 04 +~ S i~

TcSB- iZ thus allowing unnecessary delays startups or shutdowns.

Should clo the rational flexibility by avoi valves, inadvertent closure interlock associated i4".

nipulate the ~b ealCers during plant

'ted ith poWer supplied to the the valves.

of a valve occur in spite of the in ignal provided to the valves would open a close by the RCS pressure ck,

<a,iu~~A u~ W 4aevau~ ve in the event of a LOCA.

. ~~~VaQ. Kl~a Veal af

~~ ~~ ~ Ca ea~AcM,W <+, ~

~

REFERENCES @ ~

g.GC< +~g)~:" St=P CaP<<<

a iv t 7 F Val 3 Vtt-~f ChAaEh V<lh-2 4.0W U FSAR, Qmp4er c..a &uvve WN, t 8 ~ L .

50.46.

~s 10 CFR A-. ~FSAR, Chapter QHh q tJ PggR c~ ~ (5' 3 . M~ Weve R,FI, P~e-~ QA.+ ~ L.ath, W4a'tat Q.Ge 4, WOG STS

~a~~ Ca4

~0.

AmVh~~

bPR.

B tg' 4~~

3.5-8

~

-l +Ca Pnutaaaifta3 0 ~'th-'eatfVQ iw,tsas..

Rev. 0, 09/28/92 Cn ~ &PR lOt Q ~~~ace- B. 9,

7. tO CF'P So.HR.

ECCS Operating B 3.5.2 8 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.2 ECCS Operating BASES BACKGROUND The function of the ECCS is to provide core cooling and negative reactivity to ensure that the reactor core is protected after any of the following accidents:

5~. s.'y a. Loss of coolant accident (LOCA)>> coolant leakage greater than the capability of the normal charging system;

b. Rod ejection accident;

c. Loss of secondary coolant accident, including uncontrolled steam release or loss of feedwater; and

d. Steam generator tube rupture (SGTR).

The addition of negative reactivity is designed primarily for the loss of secondary coolant accident where primary i',Yis4, cooldown could add enough positive reactivity to achieve criticality and return to significant o er.

~o There are44veephases of ECCS operation:

~4 4.sa.

injection~cold leg recircul ation In the injection phase, water is taken from the refueling water i' ~~elf wit h.

QA +nM AA lfeAaak stora e tank RWST and in 'ected into the Reactor Coolant System (RCS) throug t e co egs. When sufficient water is removed from the RWST to ensure that enough boron has been added to maintainthe reactor subcritical and 495 containment sump'nough water to supply the required 5Z. Ya. ca.

net positive suction head to the ECCS pumps, suction is switched to 4Q-containment gumpfor cold leg recirculation.

~

After approximately Iiours, Sa ~ ~~MutahC C-~ oiling in the top of the core

~W~ ~( ta~+

as any resu ting oron precipitation.

and The ECCS consis s of @H.-e@separate subsystems:

M.vt'.. b safety injection (SI and residual heat removal (RHR) Each subsystem consists of two redundant, 100% capacity trains.

The ECCS accumulators and the RWST are also part of the (continued)

B 3.5-9

ECCS Operating B 3.5.2 BASES BACKGROUND ECCS, but are not considered part of an ECCS flow path as (continued) described by this LCO

~~~tJL ~

The ECCS flow paths consist of piping, valves, heat f4 M~4, ~ ~~ ~ a VX aQ exchangers, and pumps such that water from the RWST can be injected into the RCS following the accidents described in this The major components of 5'2..vi. ~

and LCO.

the SI pumps.

each subsystem are the e RHR pum s heat exchan er ubsyste~onsists of

~ g.~a two 100% capacity trains hat are interconnected and redundant such that either train is capable of supplying 100% of the flow required to mitigate the accident consequences. ~ThWinterconnecting and redundant subsystem design', provide+ the operators with the ability to utilize, components from opposite trains to achieve the required 100%

flow to the core.

During the head within injection phase of LOCA recovery, 4 suction l water from the RWST to the ECCS umps.

ate piping supp ies each su sys em.and eac subsystem. The discharge from the c ra

~ ~~

ifugal charging pump mbines prior to enterin SQ .vi..O injection tank (BI divides again into four s

'f the plant u 'es oron a BIT) and then s, each of which feeds the injection line to one co eg. The discharge from the SI and RHR pumps 'des and feeds 'njection line to each of the RCS d legs. Control valves ar t to balance t suff't ow to the RCS. This balance ensures flow to the core to meet the analysis assumpti owing a LOCA in one of the RCS cold legs.

For LOCAs that are too small to depres the shutoff head of the SI pumps, the

'he RCS below

~~n Mcx

'p cQv AsE m until the RCS ressure ecreases below ~~ ~WLMQ, the SI um shutoff ad L4~

@+0 ~ 2.2.7 During the recircu ation phase of LOCA recovery, fgIR pump 3.s'.9 suction is transferred to 4heMntainment s%m . ghe RHR 52.. vi..<

pumps en s

.recirculatio .

y

. 'u the same s as e inJection hase.

s.

w.s.g (continued)

B 3.5-10

Insert 3.5.5 The SI subsystem consists of. three redundant, 50% capacity pumps which 5Q V a~a~

supply two RCS cold leg injection lines. Each injection line is capable of providing 100% of the flow required to mitigate the accident consequences.

Insert 3.5.6 A common supply header is used from the RWST to the safety injection (SI) and Containment Spray System pumps. This common supply header is provided with two in-series motor-operated isolation valves (896A and 896B) that receive power from separate sources for single failure considerations.

These isolation valves are maintained open with DC control power removed 5~.V a ~ via a key switch located in the control room. The removal of DC control power eliminates the most likely causes for spurious valve actuation while maintaining the capability to manually close the valves from the control room during the recirculation phase of the accident (Ref. 1). The SI pump supply header also contains two parallel motor-operated isolation valves (825A and 825B) which are maintained open by removing AC power. The removal of AC power to these isolation valves is an acceptable design against single failures that could result in undesirable component actuation (Ref. 2).

A separate supply header is used for the residual heat removal (RHR) pumps. This supply header is provided with a check valve (854) and motor operated isolation valve (856) which is maintained open with DC control power removed via a key switch located in the control room. The removal of DC control power eliminates the most likely causes for spurious valve actuation while maintaining the capability to manually close the valve from the control room during the recirculation phase of the accident.

The three SI pumps feed two RCS cold leg injection lines. SI Pumps A and 8 each feeds one of the two injection lines while SI Pump C can feed both injection lines. The discharge of SI Pump C is controlled through use of two normally open parallel motor operated isolation valves (871A and 871B). These isolation valves are designed to close based on the operating status of SI Pumps A and B to ensure that SI Pump C provides the necessary flow through the RCS cold leg injection line containing the failed pump.

The discharges of the two RHR pumps and heat exchangers feed a common injection line which penetrates containment. This line then divides into two redundant core deluge flow paths each containing a normally closed motor operated isolation valve (852A and 852B) and check valve (853A and 853B) which provide injection into the reactor vessel upper plenum.

Insert 3.5.7 B (Refs. 4 and 5). This transfer is accomplished by stopping the RHR pumps, isolating RHR from the RWST by closing motor operated isolation valve 856, opening the Containment Sump B motor operated isolation valves to RHR (850A and 850B) and then starting the RHR pumps. The SI and Containment Spray System pumps are then stopped and the RWST isolated by closing motor operated isolation valve 896A or 896B for the SI and Containment Spray System pump common supply header and closing motor operated isolation valve 897 or 898 for the SI pump recirculation line.

Insert 3.5.8 SI and Containment Spray System pumps (as needed for pressure control purposes) if the RCS pressure remains above the RHR pump shutoff head (Ref. 6). This high-head recirculation path is provided through RHR motor operated isolation valves 857A, 857B, and 857C. These isolation valves are interlocked with valves 896A, 896B, 897, and 898. This interlock prevents opening of the, RHR high-head recirculation isolation valves unless either 896A or 896B are closed and either 897 or 898 are closed.

If RCS pressure is less than approximately 140 psig, the SI and Containment Spray System pumps remain in pull-stop and only RHR is used to provide core cooling. During Insert 3.5.9 After approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, simultaneous injection by the SI and RHR pumps is used to prevent boron precipitation (Ref. 7). This consists of providing SI through the RCS cold legs and into the lower plenum while providing RHR through the core deluge valves into the upper plenum.

The two redundant flow paths from Containment Sump B to the RHR pumps also contain a motor operated isolation valve located within the sump (851A and 851B). These isolation valves are maintained open with power removed to improve the reliability of switchover to the recirculation phase. The operators for isolation valves 851A and 851B are also not qualified for containment post accident conditions.

ECCS Operating B 3.5.2 BASES Sx. vi..t S.~

BACKGRO The subsystem of the ECCS also

.(continued) functions to supply borated water to the reactor core following increased heat removal events, such as a main steam line break (HSLB). The limiting design conditions occur when the negative moderator temperature coefficient is highly negative, such as at the end of each cycle.

During low temperature conditions in the RCS, limitations are placed on the maximum number of ECCS pumps that may be OPERABLE. Refer to the Bases for LCO 3.4. 12, "Low Temperature Overpressure Protection (LTOP) System," for the basis of these requirements.

The ECCS subsystems are actuated upon receipt of an SI signal. The actuation of safeguard loads is accomplished in a programmed time sequence. If offsite power is available, the safeguard loads start immediately in the programmed sequence. If offsite power is not available, the Engineered Safety Feature (ESF) buses shed normal operating loads and are connected to the emergency diesel generators (EDGs).

Safeguard loads are then actuated in the programmed time sequence. The time delay associated with diesel starting, sequenced loading, and pump starting determines the time required before pumped flow is available to the core following a LOCA.

The active ECCS components, along with the passive accumulators and the RWST covered in LCO 3.5. 1, "Accumulators," and LCO 3.5.4, "Refueling Water Stora e Tank (RWST)," provide the cooling water necessary to meet 5 K.V1.. C. (Ref.

$ AaP-SQC.4 t APPLICABLE The LCO helps to ensure that the following acceptance SAFETY ANALYSES criteria for the ECCS, established by 10 CFR 60.46 (Ref. 2f, will be met following a LOCA: 9'.

Haximum fuel element cladding temperature is ~ 2200'F;

b. Haximum cladding oxidation is g 0.17 times the total cladding thickness before oxidation; (continued)

B 3.5-11

ECCS Operating B 3.5.2 BASES APPLICABLE C. Maximum hydrogen generation from a zirconium water SAFETY ANALYSES reaction is g 0.01 times the hypothetical amount (continued) generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react;

d. Core is maintained in a eoolable geometry; and

e. Adequate long term core cooling capability is maintained.

The LCO also limits the potential fear a post trip return to SZ.iv.c power following an HSLB event and en'sur@P that containment temperature 1 imi ts are me @ok~ ~~~,~~

EtbW G.~

"Each'CCS subsystems M-taken credit for in a large break LOCA event at full power (Refs. 3 and +.

establishes the requirement for runout flow for the ECCS This event pumps, as well as the maximum res onse time for their actuation. The SI pumps are

52. ~~.h credited in a sea 1 break DitA even . ss event establishes the flow and discharge head at the design point for the pum s. The SGT+and NSLB events also cre s t e pumps. The OPERABILITY requirements for t e ECCS are based on the following LOCA analysis assumptions:

'a ~ A large break LOCA event, with loss of offsite power and a single failure disabling one RHR pump (both EDG trains are assumed to operate due to requirements for modeling full active containment heat removal system operation); and

b. A small break LOCA event, with a loss of offsite power and a single failure disabling one ECCS train.

During the blowdown stage of a LOCA, the RCS depressurizes as primary coolant is ejected through the break into the containment. The nuclear reaction is terminated either by moderator voiding during large breaks or control rod insertion for small breaks. Following depressurization, emergency cooling water is injected into the cold legs, flows into the downcomer, fills the lower plenum, and refloods the core.

5Z, >A $~ pu~

arran P HR.

4a~~a~ mpp~u ~

ubcsxs.<>

~

/~a-vtv~ w.glean

~

Pu~g wvkgt,c.'r e (continued)

B 3.5-12

ECCS Operating B 3.5.2 BASES APPLICABLE The effects on containment mass and energy releases are SAFETY ANALYSES (continued) accounted for in appropriate analyses (Refs.'W and LCO ensures that an ECCS train will deliver sufficient water

. The Qap 1c3tC o matc oi o ra es'~ enough to minimize the consequences of the core being uncovere followin a large LOCA. It also ensures that the c SI 5z.vi ~ pumps will deliver sufficient water and boron during a small LOCA to maintain core subcriticality. For smaller LOCAs, SK p~~s t deliver&sufficient fluid to maintain RCS inventory. For a small break LOCA, the steam generators continue to serve as the heat sink, providing part of the required core cooling.

The ECCS trains satisfy Criterion 3 of the NRC Policy Statement.

LCO In NODES 1, 2, and 3, two independent (and redundant) ECCS trains are required to ensure that sufficient ECCS flow is available, assuming a single failure affecting either train.

Additionally, individual components within the ECCS trains may be called upon to mitigate the consequences of other transients and accidents.

In MODES 1 ECCS train consists of an SI

~

subsystem~and an RHR subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking 5'~ q aO suction from the RWST upon an SI signal and transferring suction to 444 containment jumpn. <~ .

a.

Duri event requiring

' ECCS actuation, a flow is required to an abundant supply of from the RWST to the RCS via CS pumps eir respective cs Q.Va. supply headers to each of t cold leg injection nozzles. In the ion m, this flow may be switched to take its s from the containment sump supply its fl the RCS hot and cold legs.

The flow path for each train must maintain its designed independence to ensure that no single failure can disable both ECCS trains.

(continued) 8 3.5-13

Inser t 3.5.10 This includes securing the motor operated isolation valves as specified in SR 3.5.2.1 in position by removing the power sources as listed below.

EIN Position Secured in Position B 825A Open Removal of AC Power 825B Open Removal of AC Power 826A Closed Removal of AC Power 826B Closed Removal of AC Power 826C Closed Removal of AC Power 826D Closed Removal of AC Power 851A Open Removal of AC Power 851B Open Removal of AC Power 856 Open Removal of DC Control Power 878A Closed Removal of AC Power 8788 Open Removal of AC Power 878C Closed Removal of AC Power 878D Open Removal of AC Power 896A Open Removal of DC Control Power 896B Open Removal of DC Control Power The major components of an ECCS train consists of an RHR pump and heat exchanger capable of taking suction from the RWST (and eventually Containment Sump B), and able to inject through one of the two isolation valves to the reactor vessel upper plenum and one of the two lines which provide high-head recirculation to the SI and Containment Spray System pumps. Also included within the ECCS train are two of three SI pumps capable of taking suction from the RWST and Containment Sump B (via RHR),

and injecting through one of the two RCS cold leg injection lines. In the case where SI Pump C is inoperable, both RCS cold leg injection lines must be OPERABLE to provide 100% of the ECCS flow equivalent to a single train of SI due to the location of check valves 870A and 870B.

ECCS Operating B 3.5.2 BASES (continued)

APPLICABILITY In MODES I, 2, and 3, the ECCS OPERABILITY requirements for the limiting Design Basis Accident, a large break LOCA, are in the lower MODE o

prov'lin based on full power operation. Although reduced power would not require the same level of performance, the accident analysis does not The centrifugal char 'er requirements which Sk.vi. ~ establishes r s

ormance c CA, ce on ower. e pump performance requirements are ase on a sma break LOCA. MODE 2 and MODE 3 requirements are bounded by the MODE I analysis.

This LCO is only applicable in MODE 3 and above. Below Nz ~tt HODE 3, the SI signal setpoint is manually bypassed by

~~ ~~4.

operator control, and system functional requirements are

~

0C rae.+ JUL CydvW-4 relaxed as described in LCO 3.5.3, "ECCS Shutdown."

vaaa& okhq3 oSRau) S csA S,Z

+O

+4 p Ca L~~u~ ~s-O~

~<rvtata.etc',

~~Q a- asp As 2

indicated in Note I, the flow path may be isolated for hours in MODE 3, under controlled conditions, to perform pressure isolation valve testing per SR 3.4. 14. 1. The flow path is readily restorable from the control roo .

As indicated in Note 2, operation in MODE 3 with ECCS trains 52..VL.I<

declared inoperable pursuant to LCO 3.4. 12, "Low Temperature Over ressure Protection (LTOP) System," is necessary CAP s *~

LTOP arming temperatur near the NOOE 3

~

oundary temperature of 350'F. LCO 3.4.12 requires that certain pumps be rendered inoperable at and below the LTOP arming temperature. When this temperature is @~near the MODE 3 boundary temperature, time is needed to restore the inoperable pumps to OPERABLE status.

In MODES, and 6, plant conditions are such that the probabilit of an event requiring ECCS injection is ex reme y ow. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops MODE 5, Loops Filled,"

g'2..i i..h and LCO 3.4.8; "RCS Loops MODE 5, Loops Not Filled."

NODE 6 core cooling requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation High Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation Low Water Level."

~tala~ ~~au ~~

3.W.C., "gC,S ~~F~

Qb4 8$ ~Q +c-t> j'.3, cccs - s~)Q (continued)

B 3.5-14

ECCS Operating 8 3.5.2 BASES (continued) 0 ACTIONS 6 X. vi..o-A. I With one ECCS

~trainer'inoperable and at least 100% of the flow equivalent to a single OPERABLE ECCS train available, the inoperable components must be returned to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

a reasonable ota'7o time The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on an NRC reliability evaluation (Ref.' and is for repair of many ECCS components.

t An EC train is inoperable if it is not capable of Sa.vi,c. delivering design flow to the RCS. Individual components are inoperable if they are not capable of performing their design function or su orting systems are not available.

met.ms The LCO requires the OPERABILITY of a number of independent subsystems. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does not render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. The intent of this Condition is to maintain a combination of equipment such that 100% of the ECCS flow equivalent'to a single OPERABLE ECCS train remains available. This allows increased flexibility in plant operations under circumstances when components in opposite trains are inoperable.

3.5'.2b An event accompanied by a loss of offsite power and the failure of an EDG can disable one ECCS train until power is restored. A reliability analysis (Ref.' has shown that the impact of having one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours.

Re 6 describes situations in which one corn such as an ssover valve, can disabl

'e ECCS trains. With one or 66 100% of the flow equivale is not available ompone acility is in a inoperable such that OPERABLE ECCS train

'on outside the accident sos. Therefore, LCO 3.0.3 must be 'ately (continued)

B 3.5-15

Insert 3.5.21 In the case where SI Pump C is inoperable, both RCS cold leg injection lines must be OPERABLE to provide 100% of the ECCS flow equivalent to a single train of SI due to the location of check valves 870A and 870B.

ECCS Operating B 3.5.2 BASES ACTIONS B.l and B.2 continued) 5l., v'I a~ If the inoperable train cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without 3,s'.ii challenging plant systems.

SURVEILLANCE SR 3.5.2. 1 REQUIREMENTS Verification of proper valve position ensures that the flow

52. i.i path from the ECCS pumps to the RCS is maintained.c, Hisalignment of these valves o re der both ECCS trains ino crab Securing these valves in position 0 wel posltlon ensure that t ange position as

52. v'i.

l ure r be inadvertentl misali 4,these valves Z..i.v C. can disable the function of both ECCS rains an invalidate the accident oW ~vt9cvah analyses. A 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is considered reasonable in

+oo c'Q view of other administrative controls that ~ensure a Q~

< ~ a.~cheka vcLQJph UcA iWm9aoe, poLl+o~ mispositioned valve is unlikely.

SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position. The 31 day Frequency is appropriate because the valves are operated (continued)

B 3.5-16

Insert 3.5.11 C.1 If both trains of ECCS are inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be immediately entered.

With one or more component(s) inoperable such that 100% of the flow equivalent to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analysis. Therefore, LCO 3.0.3 must be immediately entered.

Insert 3.5.12 The listed valves are secured in position by removal of AC power or key locking the DC control power. These valves are operated under administrative controls such that any changes with respect to the position of the valve breakers or key locks is unlikely. The verification of the valve breakers and key locks is performed by SR 3.5.2.3.

ECCS Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.2 (continued)

REQUIREMENTS under administrative control, and an improper valve position would only affect a single train. This Frequency has been shown to be acceptable through operating experience.

Z.s. ig 3.5.2.3 With the ception of the operating centrifugal ch ing pump, the E umps are normally in a standby onoperating mode. As such, w path piping has the ntial to develop voids and po ts of entraine ses. Maintaining the piping from the ECC mps to e RCS full of water ensures that the system wil orm properly, injecting its full capacity into the R pon and. This will also prevent water hammer ump cavitatio , and pumping of noncondensible e.g., air, nitrogen, hydrogen) into the reactor sel following an SI signal or ring shutdown coolin he 31 day Frequency takes into consi tion the gra nature of gas accumulation in the ECCS pipin d e procedural controls governing system operation.

SR 3.5.2.4 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by Section XI of the ASHE Code. This type of testing may be accom lished b measuring the pump developed head at . po>nt of the pump characteristic curve. This verifies oth that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant safety analysis. SRs are specified in the Inservice Testing Program, which encompassesSection XI of the ASHE Code. Section XI of the ASHE Code provides the activities and Frequencies necessary to satisfy the requirements.

SR 3.5.2.5 and SR 3.5.2.6 These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or (continued)

B 3.5-17

Insert 3.5.17 SR 3.5.2.3 Verification every 31 days that AC or OC power is removed, as appropriate, for each valve specified in SR 3.5.2. 1 ensures that an active failure could not result in an undetected misposition of a valve which affects both trains of ECCS. if this were to occur, no ECCS injection or recirculation would be available. Since power is removed under administrative control and valve position is verified every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the 31 day Frequency will provide adequate assurance that power is removed.

~~ AMWOMKGo MCA

~c~ mM, ~ ~M~~ ~ V~ ECCS Operating C3 gg p c c- c4e~~u B 3.5.2

~ BASES G KAPPA,g&~~ ~ 4, ~

SURVEILLANCE hR 3.5.2.5 and SR 3.5.2.6 (continued)

RE(UIREHENTS simulated SI signal and that each ECCS pump starts receipt of an actual or simulated SI signal. The mo h Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned plant transients if th urveillances were performed with the reactor at power.

Th month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of ESF Actuation System testing, and equipment performance is monitored as part of the Inservice Testing Program.

S 5.2.7 Realignment o necessary for proper stops to allow proper positi ruptured cold leg, ensuring t

'o es in the flow path on an SI si erformance. The ves have ricted flow to a ther cold legs receive is at least the required ' um flow. This veillance is not required fo nts with flow limiting ori . The 18 mont quency is based on the same reasons as t s in SR 3.5.2.5 and SR 3.5.2.6.

S .5.2.8 Periodic ins ions of the containment sump suction inle ensure that it is stricted and stays in proper ating condition. The 18 mon equency is based e need to perform this Surveillance un the c ions that apply during a plant outage, on the have access to the location, and because of transient if the reactor at potential i lance were performe

r. This Frequency has been found

'he an unplanned suff' to detect abnormal degradation and is confir operating experience.

~~~ 3.5'.<>

REFERENCES A~~~~ ~m~m~ ~n,~ C.A gFQ

%bc- ~0, ~ssuah. Ar ~w ~wp 10 CFR 50.46.

(continued)

B 3.5-18

Insert 3.5.13

1. Letter from R.A. Purple, NRC, to L.D. White, RG&E,

Subject:

"Issuance of Amendment 7 to Provisional Operating License No. DPR-18," dated May 14, 1975.

2. Branch Technical Position (BTP) ICSB-18, "Application of the Single Failure Criterion to Manually-Controlled Electrically Operated Valves."

3. Letter from A.R. Johnson, NRC, to R.C. Mecredy, RG&E,

Subject:

"Issuance of Amendment No. 42 to Facility Operating License No. DPR R.E. Ginna Nuclear Power Plant (TAC No. 79829)," dated June 3, 1991.

4. Letter from D.H. Crutchfield, NRC, to J.E. Haier, RG&E,

Subject:

"SEP Topic VI-7. B: ESF Switchover from Injection to Recirculation Mode, Automatic ECCS Realignment, Ginna," dated December 31, 1981.

5. NUREG-0821.

7. Letter from D.H. Crutchfield, NRC, to J.E. Haier, RG&E,

Subject:

"SEP Topic IX-4, Boron Addition System, R.E. Ginna," dated August 26, 1981, ll.~ UFSAR, Section 6.2.

~ ~

ECCS Operating B 3.5.2 BASES REFERENCES (continued) i FSAR, Section ~

4o. 3 VFShR ~ Eo is 4 A

NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

B 3.5-19

ECCS Shutdown B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.3 ECCS Shutdown BASES BACKGROUND The Background section for Bases 3.5.2, "ECCS Operating,"

is applicable to these Bases, with the following modifications.

sa.vi.4 za4.+i+)~%on (s ~)

In MODE 4 the re uired ECCS train consists of two separate subsystems: and residual heat removal (RHR) "

The ECCS flow paths consist of piping, valves, heat exchangers, and pumps such that water from the refueling water storage tank (RWST) can be injected into the Reactor 52.~v Bases 3.5.2.< ~

Coolant System (RCS) following the accidents described in a~e.~

Svc.+o~ ~~

Wcw~hoJngw

~~~

e

~~0.i~~ ~~ C.

APPLICABLE The Applicable Safety Analyses section of Bases 3.5.2 also SAFETY ANALYSES applies to this Bases section.

Due to the stable conditions associated with operation in MODE 4 and the reduced probability of occurrence of a Design Basis Accident (DBA), the ECCS operational requirements are reduced. It is understood in these reductions that certain automatic safety injection (SI) actuation is not available.

In this MODE, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a DBA.

4c.aa. 5.

Only one train of ECCS is required for MODE 4. This requirement dictates that single failures are not considered during this MODE of operation. The ECCS trains satisfy Criterion 3 of the NRC Policy Statement.

LCO In MODE 4, one of the two independent (and redundant) ECCS trains is required to be OPERABLE to ensure that sufficient ECCS flow is available to the core following a DBA.

5'z.v i'.~ In MODE 4, an ECCS train consists of a Z~

subsystem and an RHR subsystem. Each train inclu es e piping, instruments, and controls to ensure an OPERABLE flow (continued)

B 3.5-20

ECCS Shutdown B 3.5.3 BASES LCO path capable of taking suction from the RWST and (continued) transferring suction to the containment sump. ~

n event requiring

' ECCS actuation, a flow s required o an abundant supply of from the RWST to the RCS via S pum eir respective supply headers to each of ld leg injection nozzles. In the erm, this flow pa be switched to take pply from the containment sump and ver ow to the RCS hot and cold legs.

APPLICABILITY In MODES I, 2, and 3, the OPERABILITY requirements for ECCS are covered by LCO 3.5.2.

In MODE 4 with RCS temperature below 350'F, one OPERABLE ECCS train is acceptable without single failure consideration, on the basis of the stable reactivity of the reactor and the limited core cooling requirements.

In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops MODE 5, Loops Filled,"

and LCO 3.4.8, "RCS Loops MODE 5, Loops Not Filled."

MODE 6 core cooling requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation High Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation Low Water Level."

ACTIONS A.I With no ECCS RHR subsystem OPERABLE, the plant is not prepared to respond to a loss of coolant accident or to continue a cooldown using the RHR pumps and heat exchangers.

The Completion Time of immediately to 'initiate actions that would restore at least one ECCS RHR subsystem to OPERABLE status ensures that prompt action is taken to restore the required cooling capacity. Normally, in MODE 4, reactor decay heat is removed from the RCS by an RHR loop.

RHR loop is OPERABLE for this function, reactor decay heat If no must be removed by some alternate method, such as use of the steam generators. The alternate means of heat removal must (continued) 8 3.5-21

Insert 3.5.16 The major components of an ECCS train during NODE 4 normally consists of an RHR pump and heat exchanger, capable of taking suction from the RWST (and eventually Containment Sump B), and able to inject through one of the two isolation valves to the reactor vessel upper plenum. Also included within the ECCS train are one of three SI pumps capable of taking suction from the RWST and injecting through one of the two RCS cold leg injection lines. The high-head recirculation flowpath from RHR to the SI pumps is not required in MODE 4 since there is no accident scenario which prevents depressurization to RHR pump shutoff head prior to depletion of the RWST.

Based on the time available to respond to accident conditions during NODE 4, ECCS components are OPERABLE if they are capable of being reconfigured to the injection mode from the control room within 10 minutes. This includes taking credit for an RHR pump and heat exchanger as being OPERABLE if they are being used for shutdown cooling purposes.

ECCS Shutdown B 3.5.3 BASES ACTI0NS A.l (continued) continue until the inoperable RHR loop components can be restored to operation so that decay heat removal is continuous.

With both RHR pumps and heat exchangers inoperable, it would be unwise to require the plant to go to MODE 5, where the only available heat removal system is the RHR. Therefore, the appropriate action is to initiate measures to restore one ECCS RHR subsystem and to continue the actions until the subsystem is restored to OPERABLE status.

B.1 S~

5 Q,vt ~ Q With no ECCS ~subsystem OPERABLE Que to the inoperability of the pump or flow path from the RWST, the plant is not prepared Co, provide high pressure response to Design Basis Events requiring SI. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time to restore at least one ubsystem to OPERABLE status ensures that prompt action is taken to provide the required cooling capacity or to initiate actions to place the plant in MODE 5, where an ECCS train is not required.

C.1 When the Required Actions of Condition B cannot be completed within the required Completion Time, a controlled shutdown should be initiated. Twenty-four hours is a reasonable time, based on operating experience, to reach MODE 5 in an orderly manner and without challenging plant systems or operators.

SURVEILLANCE SR 3.5.3.1 RE(UIREMENTS

+ms.; The applicabie Survei11ance descriptiongfrom Bases 3.5.2 apply. This SR is modified by a Note that allows an RHR train to be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned (remote or local) to the ECCS mode of (continued)

B 3.5-22

ECCS Shutdown B 3.5.3 BASES SURVEILLANCE SR 3.5.3.1 (continued)

I REQU REHENTS operation and not otherwise inoperable. This allows operation in the RHR mode during MODE 4, if necessary.

REFERENCES The applicable references from Bases 3.5.2 apply.

B 3.5-23

RWST B 3.5.4 B 3.5 EHERGENCY CORE COOLING SYSTEHS (ECCS)

B 3.5.4 Refueling Water Storage Tank (RWST)

BASES BACKGROUND The RNST su pj~ie borated water to ng-eendiN ebs

+ /,t.t <0,

~y-System-Our-ing-ace Me~mMi~ns.

both trains of the ECCS and the

'5 /.C S. s4 ~~

Containment Spray System accident during the injection phase of (LOCA recovery.

~He.

a oss o coo an 3,g,ls p amdt he recirculation mode is entere war to>>

ransferred to the co tainment sump Use o a single to supply bo trains of the an Containment Spray System is acceptabl'e since the RWST is a passive component, and passive failures are not required to be assumed to occur coincidentally with Design Basis Events.

itchover from normal operation to the injectio~p ase of ECCS . ation requires changing centrifuga charging pump suction the CVCS volume controi~anR (VCT) to the 5f .Ci.h RWST through the us isolation va.'LvWs. Each set of isolation valves is inter v alves will be g in to clo are fully open. Si e the VCT is un n

e ce 'S so that the VCT isolation R t'alve T iso la ion s ressure, the preferred pum is isol ction will be from the V This will result in a delay in ob '

til the tank

. g the orated water. The effects of this delay are dis

~ sed R

~n the Applicable Safety Analyses section of these Bases.

'J.s'. Is goring normal. operation in NODES l, 2, and 3, the safety injection (SI) and residual heat removal (RHR) pumps are aligned to take suction from the RWST. Cvma ~tv.v~

The ECCS and Containment Spray System pumps are provided with recirculation lines that ensure each pump

~ ~~~~o~

minimum flow requirements when o eratin at or head conditions. it~~ ~

can

~~

ne maintain tof.

~~~ ~~~a% ~

R-Ap S'l, t.t., a

~ W~~~O~

4a.~ M M~

~

CL%W'AAJE~

~S~

C p

Y.~Ad

~ ~~a +a~~

(continued)

B 3.5-24

Insert 3.5.18 A common supply header is used from the RWST to the safety injection (SI) and Containment Spray System pumps. A separate supply header is used for the residual heat removal (RHR) pumps. Isolation valves and check valves are used to isolate the RWST from the ECCS and Containment Spray System prior to transferring to the recirculation mode.

Insert 3.5.19 The RWST is located in the Auxiliary Building which is normally maintained between 50 F and 104 F (Ref. 1). These moderate temperatures provide adequate margin with respect to potential freezing or overheating of the borated water contained in the RWST.

RWST B 3.5.4 BASES

~+ AM pu~

BACKGROUND When the suction for the ECCS and Containment Spray Sy (continued) pumps is transferred to the containment sump, the RWST flow paths must be isolated to prevent a release of the 5 8.ii 4 containment sump contents to the RWST, which could result in a release of contaminants to the here"and the eventual loss of suction head for the ECCS pumps.

Qlhihk'l~g This LCO ensures that:

a. The RWST contains sufficient borated water to support the ECCS during the injection phase;

b. Sufficient water volume exists in the containment sump to support continued operation of the ECCS and Containment Spray System pumps at the time of transfer to the recirculation mode of cooling; and

<. ii <

~

c. The reactor remains subcritical following a LOCA.

+~ ~

'~~~~~

<<M WPzH pu~ Insufficient water in the when RWST could result in the transfer to the recircula ion mode occurs. mproper boron concentrations could result in a reduction of SDM or excessive boric acid precipitation in the core following the LOCA, as well as excessive caustic stress corrosion of mechanical components and systems inside the containment.

APPLICABLE During accident conditions, the RWST provides a source of SAFETY ANALYSES borated water to the ECCS and Containment Spray System pumps. As such, it provides containment cooling and depressurization, core cooling, and replacement inventory and is a source of negative reactivity for reactor shutdown (Ref. + 2 The design basis transients and applicable safety analyses concerning each of these systems are discussed in the Applicable Safety Analyses section of B 3.5.2, "ECCS-Operating"; 8 3.5.3, "ECCS Shutdown"; and B 3.6.6, "Containment Spray and Cooling Systems." These analyses are used to assess changes to the RWST in order to evaluate their effects in relation to the acceptance limits in the analyses.

The RWST must also meet volume, boron concentration, and temperature requirements for non-LOCA events. The volume is not an explicit assumption in non-LOCA events since the (continued)

B 3.5-25

RWST B 3.5.4

&l .s.i6

~~~) ~~~ Z~<~ C <~b ma ~p BASES Vtstovaem- Rcs APPLICABLE requiredG~I& is a small fraction of the availablevolume.

SAFETY ANALYSES *The deliverable volume limit is set by the LOCA and (continued) containment anal ses. For the RWST t del'verable '

volume the design of the t he minimum o a vo ume containe nba.'an oron concentration is an explicit assumption in the main steam line break (HSL anal sis to ensure the required shutdown ca ability~ The 1 va ue is sma or units on

~So L~ injection tank (BIT wi concentration. For units with no uced BIT boron require oron concentration limit is an impor an assumption in ensuring e re uire s u own capability. The maximum pron ration is an ex ici assum io in e o4 ~w,tcgLk rtent ECCS ac ua ion ana ysis, although it is

~ opsx ~o~

~~0.v va,v~

oW ~ typica insensitive to onlimiting event and the results concentrations. T aximum very

~s.~ temperature ensures tha cooling provided from

'he amo the RWST during the heatu a feedline break is

~

consistent with sa an assumpti actu 'nalyses,

'th analysis assump the NSLB and inadverten although the inadvertent ECCS minimum is ac ent is t icall nonlimiting.

T LB analysis has considered a delay associated with interloc tween the VCT and RWST isolation valves the results show the departure from nucleate 'ng design

~

'asis is met. The

~

has been establ's

[27] seconds, with offsi wer a le, or [37] seconds without offsite power. This ~

se time includes

'or[2] seconds for electro

~

'elay, the RWST valv , and a [10] second str a second stroke time time for the VCT valves. nts with a BIT need not be conce with the de since the BIT will supply highly borated wa to RWST switchover, provided the BIT is between the pumps and the core.

g {, tt t KthaAuJ~

For a lar e break LOCA analysis, thewater volume limit of a

allons and the lower boron concentration limit of

+000~pm are used to compute the post LOCA sump boron concentration necessary to assure subcriticality. The large break LOCA is the limiting case since the safety analysis assumes that all control rods are out of the core.

g,Boo The upper limit on boron conce tration of~~ppm is used to determine

=.Za,z (continued)

B 3.5-26

Insert 3.5.20 is selected such that switchover to recirculation does not occur until sufficient water has been pumped into containment to provide necessary NPSH for the RHR pumps.

Insert 3.5.22 the time frame in which boron precipitation is addressed post LOCA. The maximum boron concentration limit is based on the coldest expected temperature of the RWST water volume and on chemical effects resulting from operation of the ECCS and the Containment Spray System. A value of 2,900 ppm would not create the potential for boron precipitation in the RWST assuming an Auxiliary Building temperature of 50 F (Ref. 1).

Analyses performed in response to 10 CFR 50.49 (Ref. 2) assumed a chemical spray solution of 2000 to 3000 ppm boron concentration (Ref. 1) which provides a margin of 100 ppm. The chemical spray solution impacts sump pH and the resulting effect of chloride and caustic stress corrosion on mechanical systems and components. The sump pH also affects the rate of hydrogen generation within containment due to the interaction of Containment Spray and sump fluid with aluminum components,

0 0

RWST B 3.5.4 BASES APPLICABLE recircu wing a LOCA. The pur o SAFETY ANALYSES from cold leg to hot s o avoid boron (continued) sn the core followi e a the ECCS analysis, the containment spray temperature ss assu to be equal to the RWST lower temperature limit

[35]'F. the lower temperature limit is violated e containment ay further reduces containment p sure, which decreases rate at which steam can vented out the break and increa peak clad tempe ure. The upper temperature limit of [I 'F is use n the small break LOCA analysis and containment OP Y analysis. Exceeding this temperature will resu n igher peak clad temperature, because t e is less h transfer from the core to the inject water for the smal eak LOCA and higher contain pressures due to reduced tainment spray cool'apacity. For the containment res se follow'n HSLB, the lower limit on boron concentr 'on a e upper limit on RWST water temperature are used aximize the total energy release to containment.

The RWST satisfies Criterion 3 of the NRC Policy Statement.

LCO The RWST ensures that an adequate supply of borated water is available to cool and depressurize the containment in the event of a Design Basis Accident (DBA), to cool and cover the core in the event of a LOCA, to maintain the reactor

~lit subcritical following a DBA, and to ensure adequate level in the containment sump to support ECCS and Containment Spray System pump operation in the recirculation mode.

To be copsidered OPERABLE, the RWST must meet the water

'1 i established in the SRs.

h APPLICABILITY In HODES I, 2, 3, and 4, RWST OPERABILITY requirements are dictated by ECCS and Containment Spray System OPERABILITY requirements. Since both the ECCS and the Containment Spray System must be OPERABLE in HODES I, 2, 3, and 4, the RWST must also be OPERABLE to support their operation. Core cooling requirements in HODE 5 are addressed by LCO 3.4.7, "RCS Loops NODE 5, Loops Filled," and LCO 3.4.8, "RCS (continued)

B 3.5-27

RWST B 3.5.4 BASES APPLICABILITY Loops MODE 5, Loops Not Filled." NODE 6 core cooling (continued) requirements are addressed by LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation High Water Level,"

and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation Low Water Level."

ACTIONS A.1 With RWST boron concentration not within limits, Cgfaust be re urne to within imMi s within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Under these conditions neither the ECCS nor the Containment Spray System can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE condition. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> limit to restore the RWST boron concentration to within limits was developed consl ering the time required to change~~

the boron concentration and the fact that the contents of the tank are st> 11 av'ailable for injection.

8.1 ma~ v otal ~h ~ ~ K i wuhan With the RWST inopera e or reasons other than Condition A e.g., wa er volume l must be res ore o saus wl 1 n our.

In this Condition, neither the ECCS nor the Containment Spray System can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE status or to place the plant in a NODE in which the RWST is not required. The short time limit of I hour to restore the RWST to OPERABLE status is based on this condition simultaneously affecting redundant trains.

C.l and C.2 If the RWST cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least NODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to NODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full (continued)

B 3.5-28

RWST B 3.5.4 BASES ACTIONS C. 1 and C.2 (continued) power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE S 5.4.1 REQUIREMENTS The RWST bor water temperature should be

'he ver'd every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to be w~ limits assumed i e accident analyses band. This ency is suff' nt to identify a temperature change that wou p either limit and has been shown to be acceptable operating experience.

The SR is modified a Note that elimina to perform thi urveillance when ambient air are withi air e operating limits of the RWST. With a eratures within the band, the RWST temperature he

't requirement ratures uld not exceed the limits.

~55 5.5.4.

The RWST water volume should be verified every 7 days to be above the required minimum level in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation. Since the RWST volume is norma y sta e an a 7 day Frequency is appropriate and has been shown to be acceptable through operating experience.

SR 3.5.4M The boron concentration of the RWST should be verified every 7 days to be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA.

Further, it assures that the resulting sump pH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized. Since the RWST volume is normally stable, a 7 day sampling Frequency to verify boron (continued)

B 3.5-29

RWST B 3.5.4 BASES SURVEILLANCE SR 3.5.4.3 (continued)

REQUIREMENTS concentration is appropriate and has been shown to be acceptable through operating experience.

s'a~~(o.3 REFERENCES FSAR, 4hay4ev ~and Chapter $ 15+

uP~OA m~om 2. ) l is cFR. So.~g 8 3.5-30

Qss Seal Injection Flow B 3.5.

B 3. EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.5 Seal Injection Flow BASES BACKGROUND This LCO is applicable only to those units at utilize the centrifugal charging pumps for safety inj tion (SI). The unction of the seal injection throttle ives during an a cident is similar to the function of e ECCS throttle va es in that each restricts flow fr the centrifugal cha ing pump header to the Reactor oolant System (RCS).

The re riction on reactor coolan pump (RCP) seal injection flow li 'ts the amount of ECCS ow that would be diverted from the 'njection path follow'ng an accident. This limit is based o safety analysis sumptions that are required because RCP eal injection ow is not isolated during SI.

APPLICABLE All ECCS subsyste s a taken credit for in the large SAFETY ANALYSES break loss of cool accident (LOCA) at full power (Ref. I). The LOC nalysis establishes the minimum flow for the ECCS pump . he centrifugal charging pumps are also credited in the mall reak LOCA analysis. This analysis establishes th flow an discharge head at the design point for the cent fugal char 'ng pumps. The steam generator tube ruptur and main stea line break event analyses also credit th centrifugal char ing pumps, but are not limiting in their design. Reference these analyses is made in assess' changes to the Seal njection System for eval tion of their effects in elation to the acceptance lim' in these analyses.

is LCO ensures that seal injectio flow of Z [40] gpm, ith centrifugal charging pump disch rge header pressure

> {2480] psig and charging flow contr valve full open, will be sufficient for RCP seal integri y but limited so that the ECCS trains will be capable of elivering sufficient water to match boiloff rates s n enough to minimize uncovering of the core following large LOCA. It also ensures that the centrifugal charging p mps will deliver sufficient water for a small LOCA and ufficient boron to maintain the core subcritical. For s lier LOCAs, the charging pumps alone deliver sufficient flui to overcome the loss and maintain RCS inventory. Se 1 (con inued)

B 3.5-31 R

Seal Injection Flo B 3. .5 BAS APPLICA E injection flow satisfies Criterion 2 of the NRC P icy SAFETY A LYSES Statement.

(continu d)

LCO The intent of the LCO limit on seal inject on flow is to make sure that flow through the RCP seal ater injection ine is low enough to ensure that suffi ent centrifugal c arging pump injection flow is direct d to the RCS via the in 'ection points (Ref. 2) .

The L 0 is not strictly a flow lim't, but rather a flow limit ased on a flow line resis nce. In order to establi the proper flow line sistance, a pressure and flow mus be known. The flow ine resistance is determined by assumin that the RCS pre ure is at normal operating pressure an that the centr'gal charging pump discharge pressure is eater than o equal to the value specified in this LCO. The centrifug charging pump discharge header pressure remain essent'ly constant through all the applicable MODES f th's LCO. A reduction in RCS pressure would result in m e ow being diverted to the RCP seal injection line than t normal operating pressure. The valve settings establish d at the prescribed centrifugal charging pump discharge he der ressure result in a conservative valve position ould S pressure decrease. The additional modifier of th's LCO, th control valve (charging flow for four loop uni s and air o rated seal injection for three loop units) eing full ope is required since the valve is designed t fail open for th accident condition. Mith the discharg pressure and contro valve position as specified by the 0, a flow limit is es blished. It is this flow limit at is used in the accid t analyses.

The imit on seal injection flow, ombined with the ce trifugal charging pump discharge eader pressure limit d an open wide condition of the cha ging flow control alve, must be met to render the ECCS PERABLE.

conditions are not met, the ECCS flow w 11 not be as assumed If these in the accident analyses.

APPLICABI TY In NODES I, 2, and 3, the seal injection flow imit is dictated by ECCS flow requirements, which are s ecified for

( ntinued)

B 3.5-32

~ss Seal Injection Flow '~

B 3.5 BASES APPLICABIL TY NODES I, 2, 3, and 4. The seal injecti on fl ow 1 im' s not (continue ) applicable for HODE 4 and lower, however, because igh seal injection flow is less critical as a result of e lower initial RCS pressure and decay heat removal re irements in these NODES. Therefore, RCP seal injection f ow must be limited in NODES I, 2, and 3 to ensure adeq te ECCS erformance.

ACTIONS A.1 With t seal injection flow excee ng its limit, the amount of char 'ng flow available to the CS may be reduced. Under this Cond'tion, action must be t en to restore the flow to below its 'mit. The operator as 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from the time the flow is kno to be above the imit to correctly position the manual va ves and thus b in compliance with the accident analy is. The Co letion Time minimizes the otential ex os e of the lant to a LOCA with insufficient ingection ow an seal injection 1

flow~iMw imig< Thi s time i s conservative with r ct to the Completion Times of other ECCS LCOs; it istak'bas on operating experience and is sufficient for orrective actions by operations personnel.

B.l and B.2 When the R uired Actions can t be completed within the required ompletion Time, a co rolled shutdown must be f

initiat . The Completion Time 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for reaching NODE 3 rom NODE I is a reasonabl time for a controlled shutd n, based on operating exper nce and normal cooldown rat , and does not challenge plant afety systems or

'p ators. Continuing the plant shut own begun in Required A ion B. 1, an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is a reasonable time, ased on operating experience and norma cooldown rates, to reach NODE 4, where this LCO is no longe applicable.

B 3.5-33 R

Seal Injection Flow B 3.5 BASES (continued)

SURVE ILLA E SR 3.5.5.1 REQUIREHEN Verification every 31 days that the manual seal njection throttle valves are adjusted to give a flow wi in the limit ensures that proper manual seal injection thr ttle valve position, and hence, proper seal injection ow, is aintained. The Frequency of 31 days is b ed on f

gineering judgment and is consistent wi other ECCS valve Su veillance Frequencies. The Frequency has proven to be acc table through operating experienc .

As no d, the Surveillance is not re uired to be performed until 4 hours after the RCS pressur has stabilized within a

+ 20 psi range of normal operati pressure. The RCS pressure quirement is specific since this configuration will produc the required pres re conditions necessary to assure that e manual valves re set correctly. The exception is 'mited to 4 ho rs to ensure that the Surveillance is timely.

REFERENCES 1. FSAR, Chapter and Chapter [15].

2. 10 CFR 50.46.

B 3.5-34

BIT B 3.5.6 i

8 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.6 Boron Injection Tank (BIT)

BASES BACKGROUND The BIT is part of the Boron Injection System which is the rimary means of quickly introducing negativ reactivity to the Reactor Coolant System (RCS) on a afety injection

( ) signal.

The in flow path through the Boron I ection System is from e discharge of the centrifugal charging pumps through lines e uipped with a flow element a d two valves in parallel that open on an SI signal. The valves can be operated om the main control bo d. The valves and flow elements ha e main control board indications. Downstream of these valves, the flow enters e BIT (Ref. I).

The BIT is a st inless steel tank containing concentrated boric acid. Two trains of trip heaters are mounted on the tank to keep the mperat re of the boric acid solution above the precipita ion oint. The strip heaters are controlled by temper t e elements located near the bottom of the BIT. The temp ature elements also activate High and Low alarms on the m n ontrol board. In addition to the strip heaters on t e BIT, there is a recirculation system with a heat trac' syste including the piping section between the mot operated 'solation valves, which further ensures that t e boric acid tays in solution. The BIT is also equippe with a High Pre ure alarm on the main control board. The entire contents of he BIT are injected when required; hus, the contained an deliverable volumes are the same Durin normal oper ation, one of the wo BIT recirculation pump takes suction from the boron in 'tion surge tank (B T) and discharges to the BIT. The olution then returns t the BIST. Normally, one pump is runn'ng and one is shut ff. On receipt of an SI signal, the run ing pump shuts off and the air operated valves close. Flow t the BIT is then supplied from the centrifugal charging pumps The solution of the BIT is injected into the RCS through t RCS cold legs.

(contin ed)

B 3.5-35

BI B3.5 6 BA S (continued)

APPLICA LE During a main steam line break (MSLB) or loss of c olant SAFETY A LYSES accident (LOCA), the BIT provides an immediate s rce of concentrated boric acid that quickly introduces egative reactivity into the RCS.

The contents of the BIT are not credited fo core cooling or immediate boration in the LOCA analysis, b t for post LOCA recovery. The BIT maximum boron concent tion of 22,500] ppm is used to determine the m'nimum time for hot 1 recirculation switchover. The mi mum boron co entration of [20,000] ppm is use to determine the mini um mixed mean sump boron conce tration for post LOCA shutd n requirements.

For the SLB analysis, the BIT 's the primary mechanism for injecting oron into the core o counteract any positive increases i reactivity caus d by an RCS cooldown. The analysis use the minimum b ron concentration of the BIT; which also af cts both t departure from nucleate boiling and containment design a alyses. Reference to the LOCA and MSLB analyses is sed assess changes to the BIT to evaluate their ef ct n the acceptance limits contained in these analyses.

The minimum tempe tur limit of [l45]'F for the BIT ensures that the solutio does t reach the boric acid precipitation int. The temperature of the solution is monitored and alarmed on t e main control board.

The BIT bo n concentration 'mits are established to ensure that the ore remains subcriti al during post LOCA recovery.

The BIT ill counteract any pos ive increases in reactivity caused y an RCS cooldown.

The IT minimum water volume limit [1100] gallons is used to ensure that the appropriate quanti of highly borated w ter with sufficient negative reactiv y is injected into he RCS to shut down the core following n MSLB, to determine the hot leg recirculation switc over time, and to safeguard against boron precipitation.

The BIT satisfies Criteria 2 and 3 of the NRC olicy Statement.

(contin ed) 8 3.5-36

BIT B 3.5.

BA S (continued)

LCO This LCO establishes the minimum requirements for c tained volume, boron concentration, and temperature of th BIT inventory (Ref. 2). This ensures that an adequa supply of borated water is available in the event of a LO or HSLB to maintain the reactor subcritical following the e accidents.

To be considered OPERABLE, the limits estab shed in the SR or water volume, boron concentration, and temperature must b met.

If e equipment used to verify BIT pa ameters (temperature, volum and boron concentration) is termined to be inoper le, then the BIT is also in erable.

APPLICABILITY In MODES I, and 3, the BIT ERABILITY requirements are consistent wit those of LCO .5.2, "ECCS Operating."

In NODES 4, 5, an 6, the espective accidents are less severe, so the BI is no required in these lower MODES.

ACTIONS A,l If the required v lume is n t present in the BIT, both the hot leg recircu tion switch ver time analysis and the boron precipitation nalysis would n t be met. Under these conditions, ompt action must e taken to restore the volume to a ove its required lims to declare the tank OPERABLE, r the plant must be pla ed in a MODE in which the BIT is n required.

The BI boron concentration is conside d in the hot leg recir ulation switchover time analysis, he boron pre ipitation analysis, and the reactivit analysis for an MS B. If the concentration were not withi the required mits, these analyses could not be relied o . Under these conditions, prompt action must be taken to re ore the concentration to within its required limits, or the plant must be placed in a MODE in which the BIT is not equired.

The BIT temperature limit is established to ensure at the solution does not reach the bo} ic acid crystallizatio point. If the temperature of the solution drops below the (continue B 3.5-37 R

0 BIT B 3.5 BAS ACTIONS A. I (continued) minimum, prompt action must be taken to raise th temperature and declare the tank OPERABLE, or e plant must be placed in a HODE in which the BIT is not r uired.

The I hour Completion Time to restore the B to OPERABLE status is consistent with other Completio Times established or loss of a safety function and ensure that the plant w 1 not operate for long periods outsi e of the safety ana yses.

B. I B. and B.3 When Requ ed Action A. 1 cannot e completed within the required C pletion Time, a co trolled shutdown should be initiated. ix hours is a r sonable time, based on operating exp ience, to re h HODE 3 from full power conditions and o be borat d to the required SDH without challenging plan systems or operators. Bor ating to the required SDH assu ies thg the plant is in a safe condition, without need for ah, a 6itional boration.

After determining th the BIT is inoperable and the Required Actions of B. and B.2 have been completed, the tank must be retu ed t OPERABLE status within 7 days.

These actions en ure that the plant will not be operated with an inoper le BIT for a lengthy period of time. It should be not lI, however, t t changes to applicable HODES cannot be ma e until the BIT 's restored to OPERABLE status pursuant to the provisions of CO 3.0.4.

C.1 Even hough the RCS has been borated o a safe and stable con ition as a result of Required Acti n B.2, either the BIT m t be restored to OPERABLE status (Re uired Action C.l) or e plant must be placed in a condition s which the BIT is not required (HODE 4). The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Comple ion Time to reach HODE 4 is reasonable, based on operating ex erience and normal cooldown rates, and does not challeng plant safety systems or operators.

(con inued) 8 3.5-38

B B 3. .6 BAS (continued)

SURVEILL NCE SR 3.5.6.1 RE(UI REM TS Verification every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the BIT water emperature is at or above the specified minimum temperat re is frequent enough to identify a temperature change tha would approach the acceptable limit. The solution temper ure is also monitored by an alarm that provides furt r assurance of protection against low temperature. Th'requency has been own to be acceptable through operati g experience.

SR .5.6.2 Verifi tion every 7 days that e BIT contained volume is above th required limit is fr quent enough to assure that this volu will be availabl for quick injection into the RCS. If t volume is too ow, the BIT would not provide enough borat d water to en ure subcriticality during recirculation r to shut own the core following an HSLB.

Since the BIT lume is ormally stable, a 7 day Frequency is appropriate a d has been shown to be acceptable through operating experie ce SR 3.5.6.3 Ver ification ery 7 da that the boron concentration of the BIT is w'in the req ired band ensures that the reactor remains sub ritical follow g a LOCA; it limits return to power fol wing an HSLB, an maintains the resulting sump pH in an ac eptable range so tha boron precipitation will not occur i the core. In additio the effect of chloride and caust'c stress corrosion on mec nical systems and comp nents will be minimized.

T e BIT is in a recirculation loop at provides continuous irculation of the boric acid solutio through the BIT and the boric acid tank (BAT). There are number of points along the recirculation loop where loca samples can be taken. The actual location used to take sample of the solution is specified in the plant Surveil ance procedures.

Sampling from the BAT to verify the concent tion of the BIT is not recommended, since this sample may not be homogenous and the boron concentration of the two tanks m differ.

(co inued}

8 3.5-39

BIT B 3.5.

BASES SURVEILLA E SR 3.5.6.3 (continued)

REQUIREMEN The sample should be taken from the BIT or from point in the flow path of the BIT recirculation loop.

REFERENCES 1. FSAR, Chapter [6] and Chapter [15].

2. 10 CFR 50.46.

B 3.5-40

ATTACHMENT E Cross Reference Between Ginna Station Technical Specifications and NUREG-1431 ONLY RELEVANT SECTIONS ARE PROVIDED AT THIS TINE

GINNA STATION TS CROSS REFERENCE TO NUREG-1431 GINNA STATION TS 8 NUREG-1431 ATTACH. A NOTES 3.3.1.1.A LCO 3.5.4 13.II 3.3.1.1.A SR 3.5.4.1 3.3.1.1.A SR 3.5.4.2 3.3.1.1.6 13.VII 3.3.1.1.B 13. VI 3.3.1.1.B LCO 3.5.1 13.I 3.3.1.1.C LCO 3.5.2 13.III 3.3.1.1.D LCO 3.5.2 3.3.1.1. E LCO 3.5.2 3.3.1.1. F LCO 3.5.1 3.3.1.1.F LCO 3.5.2 3.3.1.1.F LCO 3.5.4 3.3.1.1.G LCO 3.5.2 13.VIII 3.3.1.1.I LCO 3.5.1 3.3.1.1.I SR 3.5.1.5 3.3.1.1.J LCO 3.5.2 3.3.1.2 LCO 3.5.4 13.II 3.3.1.3 LCO 3.5.1 13.I 3.3.1.4 LCO 3.5.2 3.3.1.5 LCO 3.5.2 3.3.1.5.A LCO 3.5.2 3.3.1.5.B LCO 3.5.2 3.3.1.5.C LCO 3.5.2 3.3.1.5.D LCO 3.5.2 13. IV 4.5 32. IV 4.5.1.1.A SR 3.5.2.5 32.I 4.5.1.1.B SR 3.5.2.5 4.5.2.1.A SR 3.5.2.4 32.II 4.5.2.1.B SR 3.5.2.4 32.II 4.5.2.2.C 32.III TABLE 4. 1-1, 15 SR 3.5.4.1 TABLE 4. 1-1, 24 SR 3.5.1.2 TABLE 4.1-1, 24 SR 3.5.1.3 28.I.A TABLE 4.1-2, 14 28.II.C TABLE 4.1-2, 14 28.II.A TABLE 4. 1-2, 14 SR 3.5.1.4 28.II.B TABLE 4.1-2, 3 SR 3.5 '.2 28.II.D TABLE 4.5-1 SR 3.5,2.3 32.II

• Attachment A, Section C.2 note number which discusses and justifies all changes to the Ginna TS section.

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