Text

Proposed Tech Specs Re Administrative Changes to License DPR-40
	ML20137Y180
Person / Time
Site:	Fort Calhoun
Issue date:	04/17/1997
From:	; OMAHA PUBLIC POWER DISTRICT
To:	;
Shared Package
ML20137Y145	List: LIC-97-0050, Forwards Application for Amend to License DPR-40,requesting Rev to Administrative Controls Consistent w/NUREG-1432 Standard TS; ML20137Y159; ML20137Y180;
References
	NUDOCS 9704220343
	Download: ML20137Y180 (75)
	v • d • e

.

( TECHNICAL SPECIFICATIONS TABLE OF CONTENTS PAGE DEFINITIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.0 SAFETY LIMITS AND LIMITING SAITTY SYSTEM SETTINGS . . . . . . . . . . . . . . . . 1-1 1.1 Safety Limits - Reactor Core .....................................11 1.2 Safety Limit. Reactor Coolant System Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.3 Limiting Safety System Settings, Reactor Protective System ... .. . . . . . . . ... .. . 1-6 2.0 LIMITING CONDITIONS FOR OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 0 2.0.1 General Requirements .. ............ ........ ..... .. .... ... 2-0 2.1 Reactor Coolant System . . .. . . . . . .... . ... .. . .. .. . .. . .. . . . .. .. ... 2-1 2.1.1 Operable Components ....................................21 2.1.2 Heatup and Cooldown Rate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3 i

2.1.3 Reactor Coolant Radioactivity ............ .... ............... 2-8 J

2.1.4 Reactor Coolant System Leakage Limits . . . . . . . . . . . . . . . . . . . . . . . . 2-11 2.1.5 Maximum Reactor Coolant Oxygen and Halogens Concentrations . . . . . . . . . 2-13 2.1.6 Pressurizer and Main Steam Safety Valves . . . . . . . . . . . . . . . . . . . . . . . 2-15 2.1.7 Pressurizer Operability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-16a 2.1.8 Reactor Coolant System Vents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-16b l

2.2 Chemical and Volume Control System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-17 2.3 Emergency Core Cooling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-20 2.4 Contammant Cooling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-24 1 2.5 Steam and Feedwater System ....................................2-28 2.6 Conta m mant System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-30 ;

2.7 Electrical Sy . ........................................2-32 i 2.8 Refuel' Cpeestions . . . . . .. . . .. . .. ... . .. . . .. .. . .. . . . . .. .. .. . 2-37 2.9 Radi 've pte Disposal System ................................2-40 2.10 Reactor Core . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-48 2.10.1 Minimum Conditions for Criticality . . . . ......................2-48 2.10.2 Reactivity Control Systems and Core Phys'es Parameter Limits . . . . . . . . . . . . . . . . .......................2-50 2.10.3 In-Core Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-54 2.10.4 Power Distribution Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-56 2.11 Deleted l e ,

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- , 169 9704220343 970417 PDR ADOCK 05000285 P PDR

TECHNICAL SPECIFICATIONS - FIGURES TABLE OF CONTENTS PAGE WHICH FIGURE DESCRIITION FIGURE FOLLOWS 11 TMLP Safety Limits 4 Pump Operations . . . . . . . . . . . . . . . . . . . . . . . . ..... 1-3 2-1A RCS Pressure-Temperature Limits for Heatup . . ...................... 2-6 2-1B RCS Pressure-Temperature Limits for Cooldown . . . . . . . ... ..... ... ... 2-6 2-3 Predicted Radiation Induced NDTT Shift . . . . . . . . . ......... .... ...... 2-6 2-11 MIN BAST Level vs Stored BAST Concentration . . . .......... ....... 2-19h 2-12 Boric Acid Solubility in Water ... ...... .................. .. '- 9h 2-10 Spent Fuel Poc,1 Region 2 Storage Criteria . . . . . . . . . . . ... .. .... 2 4 8396 2-8 Flux Peaking Augmentation Factors . . . . . . . . . . . . . . . . . . . . . . . . ....... 2-1 l

l I

i viii Arnendment No. 'le,126,131,14!,161,172

1 l

DEFINITIONS BEACTOR OPERATING CONDITIONS (Continued)

Cold Shutdown Condition (Operating Mode 4) !

mm C l The reactor coolant Tu,w is less than 210 F and the reactor coolant is ES11UTDOWN BORON CONCENTRATION at shutdown boron conecatratum bui3 REFUELING BORON, CON. CENTRATION.

l 1

Refueline , utdown Condition (Operating Mode 5) '

WO The reae:cr coc!an: is 2: refueling boren concentmtion-ami-Ta is lez than 210%

ThE:lrsaetofcoolant T , dis less~thaii210Vand thireactor coolant:isl>JREFUELING BORON CONCENTRATION; Any operation mvolvmg the shuffhng, removal, or replacement of nuc! car irradiated fuel- l CEA's, or star:up scumesToutside Ofithsisactor'pressureTvessell!Thisuspension of anp

^

REFUELING 10PERATION(shall'not preclude; completion of movement of a' component;to s safefdonserv. ative. .'po_sitiold.

The Refueline Baron Concentration A reactor coolant boron concentration of at least that specified in the CORE! OPERATING

~

1 LIMITSLREPORT Core Opera:Ing Li: nits Report which corresponds to'a dutdown margin of not less than 5% with all CEA's withdrawn.

Shutdown Boron Concentration The boron concentration required to make the reactor suberitical by the amount defined in Section 2.10.

Refueline Outace or Refueline Shutdown A plant outage or shutdown to perform refueling operations upon reaching the planned fuel depletion for a specific core.

Plant Oneratine Cycle The time period from a REFUELING,'SHEfDOWN Refueling Shutdown to the next REFUELING. SHUTDOWN Refueling Shu:dov/n.

2 Amendment No. 24,32,41,43,103,133,441

DEFINITIONS MISCELLANEOUS DEFINITIONS Operable - Ooerability A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of perfonning their related support function (s).

In Operation _ -

A system or component IN; OPERATION in operation if it is OPERABLE and is performing its design func ' n.

CEA's All full length shutdown and regulating control rods.

Non-trippable (NT) CEA's CEA's which are non-trippable.

Containment Intecrity Containment integrity is defined to exist when all of the following are met:

I (1) All nonautomatic containment isolation valves which are not required to be open during accident conditions and blind flanges, except for valves that are open under administrative control as permitted by Specification 2.6(1)a, are closed.

(2) The equipment hatch is properly closed and sealed.

(3) The personnel air lock satisfies Specification 2.6(1)b.

(4) All automatic containment isolation valves are operable, locked closed, or deactivated and secured in their closed position (or isolated by locked closed valves or blind flanges as permitted by a limiting condition for operation).

(5) The uncontrolled containment leakage satisfies Specification 3.5, and (6) The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is operable.

l 5 Amendment No. 52,109,151 i j

DEFINITIONS Core Alteration

% N_

The movement or manipulation of any componen: fuelfs6tifceslicactivitylchntrol cbmppnentQrpth6fcompoisnts affecting reactivity witliin trie" reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION Gere A!: era::en shall not preclude completion of movement of a component to a safe, conservative position. _

N - ~

Eauivalent Full Power Day (EFPD)

The time interval during power operation when the heat generated by the reactor is equivalent to reactor operation at 100% of rated power for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Shutdown Marcin Shutdown Margin shall be the amount of reactivity by which:

(1) the reactor is suberitical; or (2) the instantaneous amount of reactivity by which the reactor would be subcritical from its present condition assuming:

a. All known trippable full length control element assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn, and

b. No change in non-trippable control element assembly position.

Axial Shane Index -

The exte 1 AXIAL SHAPE INDEX Axia! Shape Index (Y s the power level detected by the lower e tu' rhnt detectors (Lyess 41 power level detected by the upper excor ticlear-instrumunrdurectors (U) divTdeifbyge sum of these power levels. The inte tal XXIAI{SHAPETINDEX Axial Shape Inde* (Y') used for the trip and pre-trip signals in the ctor protection system is the above value modified by the shape annealing factor, SAF, and a cTmstant, Br-tote ne t e true core axial power distribution for that channel.

Ye = L-U Yi = SAF x Ys + B L+U 6 Amendment No. 49,M,M9

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 R_eactor Coolant System (Continued) l j 2.1.1' Operable Comoonents (Continued)-

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(c) If fewer than the above required reactor coolant pumps] oops are cp;mbic OPERABLE, the required c...,,, l@s shall be restored to cpeab : OPERAB.12 status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the reactor shall be i Placed in ec!d du:dcwn COLDu. SHUTDOWN . -n-u within the next 12 i hours. '

(3) Cc!d Sh :dcwn er 210'F1 Ta A 300 F b6TgM210'FMitiffusFisths

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dnddf~ ' .c (a) .. At_1 cast-twal2) of the decay heat removal loops listed bebw shall be j

QPERABLE cp =b :: -.

f (i)% ant loop 1 and its associated steam generator and at j least one associated reactor coolant pump. l d

(ii) Reactor coolant loop 2 and its associated steam generator and at l

) least one associated reactor coolant pump.

(iii) One shutdown cooling pump, one shutdown cooling heat i exchanger, and associated shutdown cooling piping.

4 (iv) One shutdown cooling pump, in addition to that in (iii) above, '

one shutdown cooling heat exchanger, in addition to that in (iii) above,' and associated shutdown cooling piping.

(b Ocast'6iie%t)'of the_ decay heat removal loops listed above shall be i

IN m.O_PE_ RAT _ mION in c =:icn. i l (c) With no coolant 'p INkERATION, in ope =:!ca, suspend all operations involvin reduction in tion of the Reactor L Coolant System and initiate corrective action to return the required i coolant loop to operation in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . -

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2.0 LIMITING CONDITIONS FOR OPERATION '

2.1 Reactor Coolant System (Continued) 2.1.1 Operable Components (Continued) ,-

i (c) For de purpc=: of i::m: (a) and (b) above,1: centtimnent spray :

pump: can be censidered : :vailable shu:down-cce!!ng pumps only if '

bem of Me fe!!cwing conditicas are me:- !

(i) R=c:cr Cec! n: Sys::= tempera:ure-is le= 1:n 120*F (ii) The R=c cr Coc!an: Sys:cm :: ven: d .vii a vent == equal ic cr grea:c Man 17 in'.

. Excer:icns A!! decay h=: removal loops may be made inopemble for up :c 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provided (!) m cpem:ic= are per: nit:ed $2: would cau= dilutica of me reac:cr ecclan: system beren concen:ratica, (2) nc = fueling c;==:icnr := : !seg place, .

and (3) a!! containm=: pe=:=tica providing direct accez from $e een :imnen: ::m=;te= :o Se cu: !de atmesph:= are c!c=d wi$in i hou=.

(5) t At !=:: c= macter coc! ant pump cr c= !cw pr:=u= =fety injec:icn pu:np in Me du:down eccling mcde da!! be in operatica vtenever ;. change is being i med :n $e Scren ==en: ration of me reac:cr ecc!=: when fuel is in me reaeter- DELETED

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! (6) Both steam generators shall be filled above the low steam generator water level 4

trip set point and available to remove decay he:4 whenever the average

temperature of the reactor coolant is above 300 F. Each steam generator shall be demonstrated operable by performance of the inservice inspection program specified in Section 3.17 prior to exceeding a reactor coolant temperature of 300*F.

(7) Maximum reactor coolant system hydrostatic test pressure shall be 3125 psia.

A maximum of 10 cycles of 3125 psia hydrostatic tests are allowed.

(8) Reactor coolant system leak and hydrostatic test shall be conducted within the limitations of Figures 2-1A and 2-1B.

(9) Maximum secondary hydrostatic test pressure shall not exceed 1250 psia. A minimum measured temperature of 73*F is required. Only 10 cycles are permitted.

(10) Maximum steam generator steam side leak test pressure shall not exceed 1000 psia. A minimum measured temperature of 73 F is required.

(11) If no reactor coolant pumps are operating, a non-operating reactor coolant pump shall not be started while T, is below 385*F unless at least one of the following conditions is met:

2-2a Amendment No. 39,56,66,71,119,136, 461-

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l 2.0 L LIMITING CONDITIONS FOR OPERATION 2.1 Ecactor Coolant System (Continued) 2.1.1 Operable Components (Continued)

(a) A pressurizer steam space of 53% by volume or greater (50.6% or less actual level) exists, or l (b)- The steam generator secondary side temperature is less than 30*F above -

that of the reactor coolant system cold leg.

(12) Reactor Coolant System Pressure Isolation Valves I

(a) The integrity of all pressure isolation valves listed in Table 2.9 shall be demonstrated, except as specified in (b). Valve leakage shall not exceed the amounts indicated.

l (b) In the event that the integrity of any pressure isolation valve specified in Table 2-9 cannot be demonstrated, reactor operation may continue, provided that at least two valves in each high pressure line having a nonfunctional valve are in and remain in the raode corresponding to the isolated condition. Manual valves shall be locked in the closed position; motor operated valves shall be placed in the closed position and power i supply deenergized.

(c) If Specifications (a) and (b) abo've cannot be met, an orderly. shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . .

Rasis The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation and maintain DNBR above 1.18 during all normal operations and anticipated transients. ,

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WhenTSpebificiti6ti2;171.(2);Is]applicabliith6"isiEt6FE6blarit"pudijii3RCP5)lhrFiis6d to pr6 vide *foicsdiirculation heat' removal dsring heatup"and coulddwinTUndefithese 66nditi6nsfdscay;hea{femokAlfrequitunsnts are x 16w indigh'ihaiVIinsI$ reistdI CP]is]suffi.c_ishtlidiremossi.ydor,s?decsj!hend, bestanUsyste_m

-e .m .m (RCS)ilaopMth"one'R,d m m .mw P o be' OPERABLE

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RCs;cannot;be assured?

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mov;! ;;p;bility for removing decay h;;;; however, sing!: fail;= censidem:icn:

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reactor coolant loop or shutdoEh codlingToop provides'h(ufficient he I capability for removing decay heat, but single failure considerations require that at least l twc loops be operable. Thus, if the reactor coolant loops are not OPERABLE l cperabic, this specification requires two shutdown cooling pumps to be OPERABLE :

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2-2b Amendment No. 56, Order 4/20/81,70,

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2.0 LIMITING CONDITIONS FOR OPERATION I 2.1 Reactor Coolant System (Continued) '

2.1.1. Operable Components (Continued) _ ~

The = qui ==::: :c h:v: :nc du:dev n Oce!! g pump: cpe=b!: uten $:= i: !=: $2n 15 f::: cf ::= ;bov: $ 00= :===: $0: cing!: f !! = cf 6: cp:=:ing duiden'n ;

cec!! g Icep wi!! 20: ==l: in : cc=p!::: Ic= cf d:=y h=: ==cra! =pabi!!!y. "li$ i de ==:= v==1 b^d ==cved : d 15 f=: cf >::= beve $: 0c=, ; !=ge h=: sink i is av:1!:b!: f= ce= cccling; $=, in $c er::: cf a faile= cf de ep:=:ing de:denn

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The restrictions on availability of the containment spray pumps for shutdown cooling service ensure that the SI/CS pumps' suction header piping is not subjected to an l unanalyzed condition in this mode. Analysis has. determined that the minimum required i RCS vent area is 47 in2 This requirement may be met by removal of the pressurizer ;

manway which has a cross-sectional area greater than 47 in2 :

f I

When reactor coolant boron concentration is being changed, the process must be uniform throughout the reactor coolant system volume to prevent stratification of .

reactor coolant at lower boron concentration which could result in a teactivity insertion. -

Sufficient mixing of the reactor coolant is assured if one low pressure safety injection pump or one reactor coolant pump is in operation. The low pressure safety injection !

pump will circulate the reactor coolant system volume in less than 35 minutes when - -

operated at rated capacity. The pressurizer volume is relatively inactive; therefore, it ;

will tend to have a boron concentration higher than the rest of the reactor coolant :

system during a dilution operation. Administrative procedures will provide for use of pressurizer sprays to maintain a nominal spread between the boron concentration in the ~

pressurizer and the reactor coolant system during the addition of boron.m Both steam generators are required to be filled above the low steam generator water l level trip set point whenever the temperature of the reactor coolant is greater than the design temperature of the shutdown cooling system to assure a redundant heat removal system for the reactor.

The LTOP enable temperature has been established at T, = 385*F. The pressure transient analyses demonstrate that a single PORV is capable of mitigating overpressure events. Additional uncertainties have been applied to the Pressure-Temperature (P-T) limits to account for the case where a PORV is not available (T,>385*F) which is the reason for the discontinuity in the P-T Figures. The curves have been conservatively smoothed for operations use.

The design cyclic transients for the reactor system are given in USAR Section 4.2.2. l In addition, the steam generators are designed for additional conditions listed in USAR .

Section 4.3.4. Flooded and pressurized conditions on the steam side assure minimum j tube sheet temperature differential during leak testing. The minimum temperature for i pressurizing the steam generator steam side is 70"F; in measuring this temperature, the t instrument accuracy must be added to the 70 F; limit to determine the actu d measured limit. The measured temperature limit will be 73*F based upon use of an instrument i with a maximum inaccuracy ofi 2*F and an additional 1*F safety margin.

2-2c Amendment No. 5 , ./20/81/Ord= ,1,135,161

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I 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) '

2.1.1 Operable Components (Continued)

Formation of a 53% steam space ensures that the resulting pressure increase would not l

result in any overpressurization should the first reactor coolant pump be started when i the steam generator secondary side temperature is greater than that of the RCS cold leg. The steam space requirement is not applicable to the start of a reactor coolant pump if one or more pumps are in operation.

For the case in which the pressurizer steam space is less than 53%, limitation of the steam generator secondary side /RCS cold leg AT to 30 F ensures that a single low 1 setpoint PORV would prevent an overpressurization due to actuation of the first reactor ,

coolant pump. This requirement is not applicable to the start of a reactor coolant pump l if one or more pumps are operating.

m- %7_

H.e exceptien 10 Specifica:ica 2.1.1(A) requiring a!! containmen penetration; providing

! dirce: .ccess, f cm the contalmnen; ic the cuiside at=caphere bc ;!c;;d althin 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; require; that the equipment hatch be c!cced and held in phce by a minimum of four l belts:

References (1) USAR Section 4.3.7 1

l l

l i

l l

l l N 2-2d Amendment No. 5 4/20/81 Orde9+,136,161

2.0 LIMITING CONDITIONS FOR OPERATION 2.6 Containment System (Conti ued) -

c. Containmen integrity ch:11 not be vic!:ted "' hen ie rer;:cr ve::e! head is remcVed if 1: Scron concentration i: le:S 1:n refueling ce

~ ncentration. DELETED. -

d. Encep: for ::::ing one CED$3: time, per.itive recetivi:y change: chall not be md by CEA motion or Scron dilution un!::: 1: centainment in::grity i: in::::.DMFTED m ._

e. The containment purge isolation valves will be locked closed unless the reactor is in a cold or refueling shutdown condition.

(2) Internal Pressure The internal pressure shall not exceed 3 psig (except for containment leak rate tests).

(3) Hydrocen Purce System

a. Minimum Reauirements The reactor shall not be made critical unless all of the following requirements are met:

1. The containment isolation valves VA-280 and VA-289 shall be locked closed. Opening of these valves intermittently under administrative control is not allowed.

2. VA-80A and VA-80B with associated valves and piping to include VA-82 filters, are operable.

b. Modification of Minimum Reauirements After the reactor has been made critical, the minimum requirements may be modified to allow either or both of the following statements (i,ii) to be applicable at any one time. If the operability of the component (s) is not restored to meet the minimum requirements within the time specified below, the reactor shall be placed in a hot shutdown condition within six hours.

(i) One of the hydrogen purge fans, VA-80A or VA-80B, with associated valves and piping, may be inoperable provided the fan is restored to operable status within 30 days.

(ii) The hydrogen purge filter system, VA-82, may be inoperable provided the system is restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

2-31 Amendment No. 68,138,151

i 2.0 LIMITING CONDITIONS FOR OPERATION '

2.6 Containment System (Continued) l s-

- Basis The reactor coolant system conditions of cold shutdown assure that no steam will be

formed and,-hence, there would be no pressure buildup in the containment if the 1- reactor coolant system ruptures. The shut wy_n n gargmg_autselec_ted based _on tluyy of activiti ut. ...: refe:Fng Scron cencen::::!cn previd= ;

, 2. :dcwn margin which pr;;!ud= criti=lity under =y circum := =. Each CEDM

== h: :=::d and =m; have :wc CEA': :::::P ~

- ~ ~ - - -

! Regarding internal pressure limitations, the containment design pressure of 60 psig 2

would not be exceeded if the internal pressure before a major loss-of-coolant accident

. were as much as 3 psig.m The opening of locked or sealed closed containment

) isolation valves on an intermittent basis under administrative control includes the l following considerations: (1) stationing an operator, who is in constant communication l

with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not

, ' preclude access to close the valves and that this action will prevent the release of

radioactivity outside the containment. Operation of the purge isolation valves is +

3 prevented during normal operations due to the size of the valves (42 inches) and a i concern about their ability to close against the differential pressure that could result

from a LOCA or MSLB. Specification 2.6(1)a applies when both doors of the PAL 4 are declared inoperable, or the entire air lock assembly leakage exceeds the l

1 requirements of Specification 3.5(4). Specification 2.6(1)b(ii) applies when ;

i mechanisms other than a door, such as the inner door equalizing valve, are declared j inoperable. :

1 4

. The Hydrogen Purge System is required to be operable in order to control the quantity

[ of combustible gases in containment in a post-LOCA condition.m - The containment :

3 integrity will be protected by en3uring the penetration valves VA-280 and VA-289 are

" locked closed" while HCV-881 and HCV-882 are normally closed during power

) operation. The applicable surveillance testing requirements of Table 3-5 will ensure

[ that the system is capable of performing its design function. The blowers (VA-80A p and VA-80B), associated valves, and piping are single failure proof, have been designed as a Seismic Class I System, and are redundant to the VA-82 filter header.

, VA-80A or VA-80B is capable of providing sufficient hydrogen removal capabilities as !

4

' required by the USAR to prevent the hydrogen concentration inside of containment

from exceeding the 4% flammability limit.m Electrical Equipment qualification was _;

not required as the radiation doses in the area of the Hydrogen Purge System .

equipment were below the minimum requirements.*

, VA-80A or VA-80B with the associated valves and piping may be inoperable for 30 days. The redundancy of the blowers allows one blower with associated valves and ;

piping to be removed from operation while the other train has the capability to provide

100% hydrogen control.

! References

! (1) USAR, Section 14.16; Figure 14.16-2

[ (2) Regulatory Guide 1.7 (1971)

USAR, Section 14.17

j. (3)

! (4) Engineering Study 86-10, Calculation 53 2-31a Amendment No. 138,151 l

5 l 2.0 LIMITING CONDITIONS FOR OPERATION ,

i l \ 2.8 Refueling Ooerations i .,

lienbility

App to operating limitations dudng refueling operations.

i

j mi 1

To mini ~ the possibility of an accident occurring during ' g operations that

!j could affect p lic health and safety.

Specifications -

l The following conditi s shall be satisfied during any cling operations:

(1) The equipment ha and one door in the air ock shall be properly closed. In l addition, all automa containment isolatio valves shall be operable or at least

! one valve in each line hall be closed. >

! (2) One containment atmosp gaseo radiation monitor and one Auxiliuy

- Building Exhaust Stack g us ripfiation monitor that initiate closure of the

! containment pressure relief, ' ple, and purge system valves shall be tested

! and verified to be operable im intely prior to refueling operations. 'Ibe two j monitors shall employ one-o -o - o logic from separate contact outputs for i VIAS.

l (3) Radiation levels in containment spent fuel storage areas shall be l monitored continuous! .

l j (4) Whenever core g metry is being changed, tron flux shall be continuously j monitored by east two source range neutro ymonitors, with each monitor j . providing con

uous visualindication in the control m. When core geometry l is not being hanged, at least one source range n tron monitor shall be in i .

service.

l (5) At least /one shutdown cooling pump and heat exchanger 1 be in operation /

q How er, the pump and heat exchanger may be removed fro operation for up j to, e hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of core terations in the

~ ' *ty of the reactor coolant hot leg loops or dudng manipulati of a source.

l-l 1

i -

1 i

a

)

} 2-37 Amendment No. 25,5f,:22,152 4

4

_ _ __ _.___ _ _.. _ _ _ _ _ .~. _ _ . _ . _ _ _ . _ __

. 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 miinF Operations (Continued) i t

! (6) Direct communication between personnel in Ge control room and at the re g j machine shall be available whenever c~nanges in core geometry are taking .

i j

(7) y irraelintal fuel is being handled in the auxiliary building, exhaust y from the spent fuel pool area will be diverted through charcoal

mm.

i -

! (8) Deleted.

l (9) A minimum f 23 feet of water above the top of the shall be maintained l whenever ~ fuel is being handled.

4 f (10) Deleted. !

L.

(11) Storage in Region 2 o the spent fuel racks restricted to diose assemblies l whose parameters fall *

the " acceptable" ea of Figure 2-10. Storage in the j peripheral cells of Regi 2 shall be *cted to those assemblies whose j parameters fall within the area of Fi[gure 2-10.

l (12) A minimum boron concentrata of ppm shall be maintained in the Spent l l Fel Pool whenever storing u ' ' ted fuelin the Spent Fuel Pool.

i j If any of the above conditions are t , all refueling operations shall cease j immediately, work shall be initiated satisfy required conditions, and no operations

that may add positive reactivity to core be made, f i A spent fuel assembly may be sferred directly m the reactor core to the spent fuel

! pool Region 2 provided independent verificatig of assembly burnups has been

+

completed and the anemb humup meets the acceptance criteria identified in Technical i Specification Figure 2- .

! l i Movement ofirradi fuel from the reactor core shall not initiated before the reactor j core has been su ritkal for a minimum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if the tor has been operated at j power levels i cess of 2% rated power. - '

i i Bases 4

'Ihe whet and general procedures to be utilimi during refuelin operations are dir=M in the USAR. Detailed instructions and the above specifi tions provide assadance that no incident could occur during the refueling operations tha would

~

/

2-38 Amendment No. 5,24,25,43,75, 133,152,155,16^,174 i

! 2.0 IYMITING CONDITIONS FOR OPERATION I

2. Refuelina Ooerations (Continued) i

! t in a hazard to public health and safety. Whenever changes are not being in i edue geometry one flux monitor is sufficient. nis permits maintenance in the tation. Continuous monito vides flux j im

te indication of an unsafe co, ring of radiation levels and neutron ndition. De shutdown cooling pump ~ used to j

maintain uniform boron concentration.

j

. De shu margin as indicated will keep the core suberitical even if

! CEA's were withdrawn the core. During refueling operations, the reactor fueling cavity is filled with appro 'mately 250,000 gallons of borated water. The bo concentration of

] this water (of at apt the refueling boron concentration) is su 'ent to maintain the i reactor suberitical by more than 5%, including allowance for

ties, in the cold i

condition with all withdrawn.m Periodic checks o refueling water boron concentration ensures oper shutdown margin. Comm ication requirements allow

! the control room operator inforra the refueling mehi operator of any impending unsafe condition detected m the main control roo board indicators during fuel movement.

He restriction of not moving fuel the reactor fo a period of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after the power has been removed from the core vantage f the decay of the short half-life fission products and allows for any failed fuel pur itself of fission gases, thus reducing the

{

consequences of fuel handling accident.

J De ventilation air for lioth the contain t d the spent fuel pool area flows through absolute particulate filters and radia ~ {

mom rs before discharge at the ventilation discharge duct. In the event the s discharge hould indicate a release in excess of !

the limits in the technical specifi ons, the con ent ventilation flow paths will be closed automatically and the a 'liary building ven ' tion flow paths will be closed 1 manually. In addition, the ex ust vendlation ductwork om the spent fuel storage area is equipped with a charcoal dter which will be manuall l ut into operatior, whenever.

irradiated fuel is being led.

The basis for the 500 m boron concentration requirement with ral poisoned storage l racks is to maintal the k,, below 0.95 in the event a mispla unirradiated fuel assembly is loca next to a spent fuel assembly. A mispla unirradiated fuel assembly at 4.5 /o enrichment condition, in the absence of soluble ' son, may result in exceeding j design effective multiplication factor. Soluble boron in Spent Fuel.

Pool water, which credit is permitted under these conditions, would re that the effective ltiplication factor is maintained substantially less than the design ndition.

He bor concentration is periodically sampled in accordance with Specificatig 3.2.

Ref USAR, Section 9.5.1.2 2-39 Amendment No. 24,75,103,117,133, 141,155,16^,174

)

i NiE W 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueline 2.8.1 Refueline Shutdown 2.8.1(1) Boron Concentration Applicability Applies to Reactor Coolant System boron concentration when fuel is in the reactor and one or more reactor vessel head closure bolts are less than fully tensioned.

1 Obiective The limit on the boron concentration of the Reactor Coolant System ensures that the l reactor remains suberitical when the plant is in REFUELING SHUTDOWN (MODE 5).

Specification Boron concentration of the Reactor Coolant System shall be greater than or equal to ,

REFUELING BORON CONCENTRATION. l Reauired Actions (1) With the boron concentration not within limit, suspend CORE ALTERATIONS immediately, and (2) Suspend positive reactivity additions immediately, and (3) Initiate actions to restore boron concentration to within limits immediately.

2-37 Amendment No. 2-5,56,133,152

)

NEW 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueline 2.8.1 Refueline Shutdown 1

2.8.1(2) Nuclear Instrumentation l i

Annlicability l

Applies to the source range neutron monitors in MODE 5 with fuel in the reactor and with "

one or more reactor vessel head closure bolts less than fully tensioned.

l Obiective To monitor the core reactivity condition and to alert the operator to unexpected changes 1 in core reactivity when the plant is in REFUELING SHUTDOWN (MODE 5).

Specification Two source range neutron monitors shall be OPERADLE.

Reauired Actions (1) With only one source range neutron monitor OPERABLE, suspend CORE ALTERATIONS and positive reactivity additions immediately.

(2) With no source range neutron monitors OPERABLE, suspend CORE ALTERATIONS and positive reactivity additions immediately, and initiate actions to restore one source range neutron monitor to OPERABLE status immediately, and verify RCS boron concentration is greater than or equal to REFUELING BORON CONCENTRATION within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

2-38 Amendment No. 5,21,25,43,75, 133,152,155,169,174

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

l.

NW e

i 2.0' LIMITING CONDITIONS FOR OPERATION i

! 2.8 ' Refueline !

l '2.8.1- Refueline Shutdown

!. 2.8.1(3) Shutdown Cooling System - High Water Level 4

2 Acolicability i

Applies to shutdown cooling requirements in MODE 5 with fuel in the reactor and with one or more reactor vessel head closure bolts less than fully tensioned, and the refueling t

cavity water level 2 23 ft. above the top of the core,

! Obiective -

g To minimize the possibility of a loss of shutdown cooling accident occurring inside :

i containment that could affect public health and safety.

Soecification

! One OPERABLE Shutdown Cooling loop shall be IN OPERATION except as noted 3 below:

1

1. The required Shutdown Cooling loop may be removed from operation for s one hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . period, provided no operations are permitted that would cause dilution of the RCS boron concentration.

2. The required Shutdown Cooling loop may be inoperable for up to eight hours .

provided (1) no operations are permitted that would cause dilution of the RCS boron concentration, (2) no CORE ALTERATIONS er REFUELING -

OPERATIONS are taking place, (3) all containment penetratip d providing direct-access from' the containment atmosphere to the outside att:csphere are closed l within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and (4) at least one loop is available under administrative controls. !

Required Actions (1) .With no Shutdown Cooling loop IN OPERATION (except as allowed by notes 1 or 2 above),

a. Suspend operations involving a reduction in reactor coolant boron i concentration immediately, and

b. Suspend . loading ' of ' irradiated fuel assemblies into the reactor core immediately, and

c. Initiate actions to restore system to operation immediately, and

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

2-39 Amendment No. 24,75,103,117,133, 141,155,159,174

i NGW 2.0 LIMITING CONDITIONS FOR OPERATION i

2.8 Refueline 2.8.1 .Refueline Shutdown 2.8.1(4) Shutdown Cooline System - Low Water Level Applicability i

Applies to shutdown cooling requirements in MODE 5 with fuel in the reactor and with one or more reactor vessel head closure bolts less than fully tensioned, and the refueling a

cavity water level < 23 ft. above the top of the core.

Obiective j To minimize the possibility of a loss of shutdown cooling accident occurring inside of containment that could affect public health and safety.

Specification Two Shutdown Cooling loops shall be OPERABLE, and one Shutdown Cooling loop shall be IN OPERATION.

1

~ Reauired Actions <

(1) With one Shutdown Cooling loop inoperable either, l

a. Restore the inoperable Shutdown Cooling loop to OPERABLE status immediately, or ,

b. Initiate actions to establish at least 23 ft. of water above the top of the core immediately.

(2) With both Shutdown Cooling loops inoperable or one Shutdown Cooling loop not IN OPERATION,

a. Suspend operations involving a reduction in RCS boron concentration immediately, and

b. Initiate actions to restore at least one Shutdown Cooling loop to operation immediately, and

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

2-39a Amendment No.

N t5 %l

! 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueline 2.8.2 Refueline Onerations - Containment i

2.8.2(1) Containment Penetrations 1 i Anolicability i

! Applies to co,ntainment penetrations in MODE 5 during CORE ALTERATIONS and l REFUELING OPERATIONS inside containment.

Obiective To minimize the consequences of an accident occurring during CORE ALTERATIONS

and REFUELING OPERATIONS inside containment that could affect public health and

safety.

j Specification i

The containment penetrations shall be in the following status:

a. The equipment hatch closed and held in place by at least four bolts;

b. At least one door in the Personnel Air Lock closed; and .

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. capable of being closed by an OPERABLE Ventilation Isolation Actuation Signal.

Reauired Actions (1) With one or more containment penetrations not in required status, suspend CORE ALTERATIONS and REFUELING OPERATIONS within containment immediately.

2-39b Amendment No.

l NGW l 2.0 LIMITING CONDITIONS FOR OPERATION :

2.8 Refuelina 2.8.2 Refueline Operations - Containment l

l 2.8.2(2) Refueline Water level Apolicability i

Applies to the refueling water level during CORE ALTERATIONS, and during l REFUELING OPERATIONS inside of containment.

! Obiective

To minimize the consequences of a fuel handling accident durin; CORE ALTERATIONS

! and REFUELING OPERATIONS inside of the containment that could affect public health !

and safety.

Specification

! l l The refueling water level shall be 2 23 ft. above the top of the reactor vessel flange.

Reauired Actions l l

l (1) With the refueling water level not within limits, suspend CORE ALTERATIONS l l

immediately, and i

I (2) Suspend REFUELING OPERATIONS inside of containment immediately, and (3) Initiate actions to restore refueling water level to within limits immediately.

I l

1 2-39c Amendment No.

I

NGw I 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueline 2.8.2 Refueline Operations - Containment 2.8.2(3) Ventilation Isolation Actuation Siunal (VIAS) !

Annlicability Applies to operation of the Ventilation Isolation Actuation Signal (VIAS) during CORE ALTERATIONS and REFUELING OPERATIONS inside containment.

Obiective To minimize the consequences of an accident occurring during CORE ALTERATIONS or REFUELING OPERATIONS that could affect public health and safety.

Specification VIAS including manual actuation capability shall be OPERABLE with two gaseous radiation monitors OPERABLE and supplied by independent power supplies.

l Reauired Actions (1) With less than two radiation monitors OPERABLE, or VIAS manual actuation 1 capability inoperable, immediately suspend CORE ALTERATIONS and l REFUELING OPERATIONS.

2-39d Amendment No i

1 l

l 1

i i

NEW ,

l 2.0 LIMITING CONDITIONS FOR OPERATION l 2.8 Refueline ,

2.8.3 Refueline Onerations - Snent Fuel Pool I 1

2.8.3(1) Spent Fuel Assembly Storace l l '

l

Anolicability j i

l Applies to storage of spent fuel assemblies whenever any irradiated fuel assembly is stored l . in Region 2 (including peripheral cells) of the spent fuel pool. The provisions of !

) Specification 2.0.1 for Limiting Conditions for Operation are not applicable.

l Obiective i To minimize the possibility of an accident occurring during REFUELING. OPERATIONS l that could affect public health and safety.

Specification l

The combination of initial enrichment and burnup of each spent fuel assembly stored in l

Region 2 (including peripheral cells) of the spent fuel pool shall be within the acceptable burnup domain of Figure 2-10.

Reauired Actions )

1 l (1) With the requirements of the LCO not met, initiate action to move the i

noncomplying fuel assembly immediately.

l l

l 4

2-39e Amendment No.

i

1 NGW 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueline 2.8.3 Refueline Operations - Spent Fuel Pool 2.8.3(2) Snent Fuel Pool Water Level Annlicability Applies to the water level of the spent fuel pool during REFUELING OPERATIONS in the spent fuei pool. The provisions of Specification 2.0.1 for Limiting Conditions for '

Operation are not applicable.

Obiective To minimize the consequences of a fuel handling accident during REFUELING OPERATIONS in the spent fuel pool that could affect public health and safety. !

l Specification l

The spent fuel pool water level shall be 2: 23 ft. above the top of irradiated fuel l sssemblies seated in the storage racks.

Reauired Actions (1) With the spent fuel pool water level not within limits, suspend REFUELING OPERATIONS in the spent fuel pool immediately.

l 2-39f Amendment No.

NGW 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueline 2.8.3 Refueline Onerations - Spent Fuel Pool 2.8.3(3) Spent Fuel Pool Boron Concentration Annlicability I l

I Applies to the boron concentration of the spent fuel pool when unirradiated fuel assemblies are stored in the spent fuel pool. The provisions of Specification 2.0.1 for Limiting _

Conditions '. Operations are not applicable.

Obiective To minimize the possibility of an accident that could affect public health and safety from occurring when unirradiated fuel assemblies are stored in the spent fuel pool.

Specification The spent fuel pool boron concentration shall be 2 500 ppm.

&auired Actions (1) With the spent fuel pool boron concentration < 500 ppm, suspend REFUELING j OPERATIONS in the spent fuel pool immediately, and l

(2) Restore spent fuel pool boron concentration to 2 500 ppm immediately.

2-39g Amendment No.

. . - . . . - _ _ - - - - - . = -- . -- .. - .

Al 4

2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueline 2.8.3 Refueline Operations - Spent Fuel Pool 2.8.3(4) Spent Fuel Pool Area Ventilation Annlicability Applies to operation of the ventilation system in the spent fuel pool area during REFUELING OPERATIONS in the spent fuel pool. The provisions of Specification 2.0.1

for Limiting Condition for Operations are not applicable.

Obiective To minimize the consequences of an accident occurring during REFUELING OPERATIONS in the spent fuel pool that could affect public health and safety.

Snecification ,

The spent fuel pool area ventilation system shall be IN OPERATION.

l- Reouired Actions

(1) With the spent fuel pool area ventilation system not IN OPERATION, suspend 4

REFUELINO OPERATIONS in the spent fuel pool immediately.

l 4

4 l

i l

i 6

i 4

1 2-39h Amendment No.

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

4 dl b 1

l 2.0 LIMITING CONDITIONS FOR OPERATION 3 2.8 Befueline

2.8.3 Refueline Operations - Spent Fuel Pool 2.8.3(5) Ventilation Isolation Actuation Sicnal (VIAS)

Applicability 4

Applies to operation of the Ventilation Isolation Actuation Signal (VIAS) during REFUELING OPERATIONS in the spent fuel pool.

Obiective f

To minimize the consequences of an accident occurring during REFUELING

OPERATIONS in the spent fuel pool that could affect public health and safety.

Specification VIAS including manual actuation capability shall be OPERABLE with two gaseous radiation monitors on the auxiliary building exhaust stack OPERABLE, and supplied by independent power supplies.

Reauired Actions )

i (1). With less than two gaseous radiation monitors on the auxiliary building exhaust

)

stack OPERABLE or VIAS manual actuation capability inoperable, immediately i suspend REFUELING OPERATIONS.

I 2-39i Amendment No

l NSY t

i 2.0 LIMITING CONDITIONS FOR OPERATION

2.8 Refueline i l 3 Dases s

4 2.8.1 Refueline Shotdown i

l j 2.8.1(1) Boron Concentration '

The baron concentration of the water filling the reactor refueling cavity (of at least the j REFUELING BORON CONCENTRATION) is sufficient to maintain the reactor i subcritical by more than 5%, including allowance for uncertainties, in the cold condition

3. with all rods withdrawn. The REFUELING BORON CONCENTRATION is specified in !

!- the COLR. Periodic checks of the refueling water boron concentration ensure the proper i shutdown margin.

]

3 j When "immediately" is used as a completion time, the required action should be pursued ]

without delay and in a controlled manner. Suspension of CORE ALTERATIONS shall

} not preclude completion of movement of a component to a safe, conservative position.

t j 2.8.1(2) Nuclear Instrumentation I i

f Two OPERABLE source (wide) range neutron monitors are required to provide a signal-j to ensure that redundant monitoring capability is available to detect changes in core

reactivity. With only one source range neutron monitor OPERABLE, redundancy has been l

! lost. Since these instruments are the only direct means of monitoring core reactivity ;

{- conditions, CORE ALTERATIONS and positive reactivity additions must be suspended '

l immediately. When "immediately" is used as a completion time, the required action ;

should be pursued without delay and in a controlled manner.

l With no source range neutron monitor OPERABLE, there is no means of detecting i changes in core reactivity. However, since CORE ALTERATIONS and positive reactivity )

! additions are not being made, the core reactivity condition is stabilized until the monitors 8-are returned to OPERABLE status. This stabilized condition is determined by verifying that the required boron concentration exists. The completion time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is sufficient

, to obtain and analyze a reactor coolant sample for boron concentration. The frequency of l once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that unplanned changes in boron concentration would be i identified. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months frequency is reasonable, considering the low probability of a i change in core reactivity during this period.

),

i

} 2-39j Amendment No.

i f

4 i

A l

2 NBAl l

)

i l

2.0 LIMITING CONDITIONS FOR OPERATION 3 2.8 Refueline l i

1 Bases (Continued)

2.8.1(3) Shutdown Cooline System - Hinh Water Level The purposes of the SDC system in MODE 5 with fuel in the reactor and with one or.

more reactor vessel head closure bolt less than fully tensioned, are to remove decay heat i and sensible heat frcm the RCS, provide mixing of borated coolant, to provide' sufficient coolant circulation to minimize the effects of a boron dilution accident, and to prevent boron stratification.

i

~

Inadequate cooling of the reactor coolant could result in boiling of the reactor coolant j which could lead to a reduction in boron concentration in the coolant due to the boron plating out on components near the areas of boiling activity, and because of the possible

addition of water to the reactor vessel with a lower boron concentration than is required to keep the reactor suberitical. The loss of reactor coolant and the reduction of boron i

I concentration in the reactor coolant would eventually challenge the integrity of the fuel cladding, which is a fission product barrier.

1 An OPERABLE SDC loop consists of a SDC pump, a heat exchanger, valves, piping, l instruments and controls to ensure an OPERABLE flowpath and to determine the low end temperature. . The flowpath starts in one of the RCS hot legs and is returned to the RCS lj cold legs. The containment spray pumps can be considered as available shutdown cooling l pumps only if the RCS temperature is less than 120 F and the RCS is vented with a vent j area greater than or equal to 47 in2 The restrictions ensure that the SI/CS pumps' suction l header piping is not subjected to an unanalyzed condition in this mode. Analysis has j determined that the minimum required RCS vent area is 47 in2 This requirement may be 1 met by removal of the pressurizer manway which has a cross-sectional area greater than )

! 47 in2 i a.

Specification 2.8.1(3) is modified by an exception that allows the required operating SDC j j loop to be removed from service for up to I hour in each .8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period, provided no l

[ - operations are permitted that would cause a reduction of the RCS boron concentration. l

,' This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles, and RCS to SDC isolation valve testing. During this 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, ;

decay heat is removed by natural convection to the large mass of water in the refueling )

l cavity. Boron concentration reduction is prohibited because uniform concentration i

] distribution cannot be assured without forced circulation.

4 l ,

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2-39k Amendment No.

J 4

.rs , -. - , . . - - - . - - - , , - , - , . .. , . - . . . - ,

NN 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueline Bases (Continued) 2.8.1(3) Shutdown Cooline System - Ilieh Water Ixvel (Continued)

Specification 2.8.1(3) is modified by an exception to allow both trains of SDC out of service for up to eight hours provided, in part, that at least one SDC train is available under administrative controls. This allows evolutions such as Engineered Safety Feature testing to be completed when the SDC system is not fully OPERABLE but is considered available since only minor operator actions are required to restore the SDC system to OPERABLE status and place it IN OPERATION. A SDC loop is considered available under administrative controls if there are: (1) approved procedures, (2) a dedicated operator stationed at the controls if they are outside of the control room, and (3) direct communication between the dedicated operator and the control room. Similarly, the SDC system is considered available under administrative controls when an operator is not at the location of the controls provided: (1) procedural guidance is consulted prior to removing SDC from service to determine the time-to-boil, and (2) there is sufficient time for the operator to travel to the local controls and perform the required actions.

With the water level 2 23 feet above the top of the core, only one SDC loop is required for decay heat removal. Only one is required because the volume of water above the top of the core provides backup decay heat removal capability. The 23 ft level was selected because it ensures that adequate time is available to initiate emergency procedures to cool the core. For example, assuming the amount of decay heat generated one day after shutdown with an initial reactor coolant temperature of 200'F, this level provides approximately 25 minutes before the reactor coolant would boil. More time is available under conditions more representative of when this specification applies (i.e., when the reactor vessel closure head is removed). For example, five days after shutdown with the initial reactor coolant temperature of 130 F provides more than four hours before the reactor coolant would boil.

If the SDC loop requirements are not met, there will be no forced circulation to provide mixing to estabiish uniform boron concentrations. Therefore, actions that reduce boron concentration are required to be suspended immediately. Additionally, suspending any operation that would increase the decay heat load, such as loading a fuel assembly, is a prudent action under this condition. Closing the containment penetrations that provide direct access to the outside environment prevents fission products, if released from a loss of decay heat removal event, from escaping the containment. ~ A completioa time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable because most SDC problems can be repaired within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and because there is a low probability of the coolant boiling in that time.

When "immediately" is used as a completion time, the required action should be purmed without delay and in a controlled manner.

2-3 91 Amendment No

i /qGW 2.0 LIMITING CONDITIONS FOR OPERAT1QN 2.8 Refueline i

Bases (Continued) I

l 2.8.1(4) Shutdown Cooline System - Low Water Level l

With the water level < 23 feet above the tap of the core, two OPERABLE SDC loops are

required since the volume of water above the top of the core will not provide sufficient i backup decay heat removal capability.

With one SDC loop inoperable, actions shall be immediately initiated and continued until l the SDC loop is restored to OPERABLE status, or until 2 23 feet of water is established above the top of the core. With the water level 2 23 feet above the top of the core, this 1

-Specification is no longer applicable, and Specification 2.8.1(3) is applicable. ;

An OPERABLE SDC loop consists of a SDC pump, a heat exchanger, valves, piping, instruments and controls to ensure an OPERABLE flowpath and to determine the low end !

temperature. The flowpath starts in one of the RCS hot legs and is returned to the RCS p cold legs. The containment spray pumps can be considered as available shutdown cooling l

. pumps only if the RCS temperature is less than 120*F and the RCS is vented with a vent '

area greater than or equal to 47 in2 l With both SDC loops inoperable, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Therefore, actions that reduce boron concentration are required to be suspended immediately. Closing the containment penetrations that provide direct access to the outside environment prevents fission products, if released from !

a loss of decay heat removal event, from escaping the containment. A completion time !

of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable because most SDC problems can be repaired within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and l because there is a low probability of the coolant boiling in that time. I When "immediately" is used as a completion time, the required action should be pursued without delay and in a controlled manner. l

{

l 2-39m Amendment No

_ _. _._. _ _ . __. ~ -- . _ - - . - . .._ .~ --

j N E*'

2.0' LIMITING CONDITIONS FOR OPERATION 2.8 Refueling Bases (Continued)

- 2.8.2(1) Containment Penetrations During CORE ALTERATIONS or REFUELING OPERATIONS inside of containment, a release of fission product radioactivity within the containment will be restricted from escaping to the environment when the LCO requirements are met. In MODE 5, the potential for contahunept pressurization as a result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere are less stringent than when the reactor is at power. The LCO does not require CONTAINMENT INTEGRITY.

Since there is no potential for containment pressurization as a result of a fuel handling accident, the Appendix J leakage criteria and tests are not required, The containment equipment hatch, which is part of the containment pressure boundary, provides a means of moving large equipment and components into and out of containment.

During CORE ALTERATIONS or REFUELING OPERATIONS inside of containment, the equipment hatch must be held in place by at least four bolts. Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced.

The Personnel Air I;x:k (PAL), which is also part of the containment pressure boundary, provides a means for personnel access into containment. The doors are normally l

imerlocked to prevent simultaneously opening when CONTAINMENT INTEGRITY is l

required. During periods of shutdown when containment closure is not required, the :

interlock may be disabled and both PAL doors allowed to remain open for extended l periods when frequent containment entry is necessary. During CORE ALTERATIONS I or REFUELING OPERATIONS inside of comainment, CONTAINMENT INTEGRITY is not required, therefore the door interlock mechanism may remain disabled, but one PAL door must always remain closed.

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. The specification is met when ;

one of the two automatic isolation valves per penetration is OPERABLE, or by closure of a manual isolation valve, blind flange, or equivalent. Equivalent isolation methods must be approved (through 10 CFR 50.59 safety euluation process) and may include use of a - l material that can provide a temporary, atmospheric pressure ventilation barrier for the !

other containment penetrations during fuel movements.

2-39n Amendment No.

,-n-

b 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueline !

Bases (Continued) ,

2.8.2(1) Containment Penetrations (Continued)

For automatic isolation valves with direct access to the outside atmosphere to be OPERABLE requires that the Ventilation Isolation Actuation Signal (VIAS) is OPERABLE in order to close the valves. This action prevents release of significant radionuclides from ;

the containment to atmosphere. During CORE ALTERATIONS and REFUELING -

OPERATIONS, the OPERABILITY of VIAS is addressed by Specification 2.8.2(3). ;

When "immediata!)" is used as a completion time, the required action should be pursued l without delay and in a controlled manner. ;

2.8.2(2) Refueline Water level Prior to REFT'ELING_ OPERATIONS inside containment, the reactor refueling cavity is filled with approximately 250,000 gallons of borated water. The minimum refueling water !

level meets the' assumption of iodine decontamination factors following a fuel handling ;

accident. When the water level is lower than the required level, CORE ALTERATIONS and REFUELING OPERATIONS inside of containment shall be suspended immediately.

This effectively precludes a fuel handling accident from occurring. When "immediately" -

is used as a completion time, the required action should be pursued without delay and in l a controlled manner. Suspension of REFUELING OPERATIONS and CORE -

ALTERATIONS shall not preclude completion of movement of a component to a safe, I conservative position. In addition to suspending REFUELING OPERATIONS and CORE !

ALTERATIONS, action to restore the refueling water level must be initiated immediately.

Movement of irradiated fuel from the reactor core is not initiated before the reactor core I has been suberitical for a minimum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if the reactor has been operated at power levels in excess of 2% rated power. The restriction of not moving fuel in the reactor for a period of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after the power has been removed from the core takes advantage of !

the decay of the short half-life fission products and allows for any failed fuel to purge itself of fission gases, thus reducing the consequences of a fuel handling accident.

2-39o Amendment No.

1 I

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! 2.0 LIMITING CONDITIONS FOR OPER.ATION 2.8 Refueline Bases (Continued) 2.8.2(3) Ventilation Isolation Actuation Sicnal (VIAS) l A Ventilation Isolation Actuation Signal (VIAS) is initiated by a Safety Injection Actuation

Signal (SIAS), a Containment Spray Actuation Signal (CSAS) or a Containment Radiation

High Signal (CRHS). During CORE ALTERATIONS and REFUELING OPERATIONS l- only the CRHS is required to respond to a fuel handling or reactivity accident. At least i two of the following three radiation monitors (Containment Monitor (RM-051),

Containment / Auxiliary Building Stack Swing Monitor (RM-052), Auxiliary Building Stack j Radiation Monitor (RM-062)) must be OPERABLE, powered from independent 480-VAC buses, and capable of actuating both the A and B trains of VIAS, to fulfill the i j requirements of this specification. The independent 480-VAC buses may be supplied by i a single 4160-VAC power source. In addition, one manual actuation channel is required

to be OPERABLE. (Note, the Offsite Dose Calculation Manual may have additional requirements / restrictions concerning operation of these monitors.)

l In the event that only one of the above radiation monitors is OPERABLE or VIAS manual actuation capability is inoperable, CORE ALTERATIONS and REFUELING OPERATIONS must be suspended thus precluding the possibility of a fuel

handling / reactivity accident. i

For the fuel handling accident in containment, the very conservative assumption that all I the rods in a single assembly fail with no credit taken for containment isolation or

} atmosphere filtration yields' doses at the exclusion area boundary (EAB) and low population zone (LPZ) that remain well within the limits of 10 CFR 100.

i

} VIAS initiates closure of the containment pressure relief, air sample, and purge system j valves, if open. This action prevents release of significant radionuclides from the l containment to the environment. The containment penetrations providing direct access to

the environment are required to be closed, or capable of being closed by an OPERABLE

[ VIAS in accordance with Specification 2.8.2(1). VIAS also initiates other actions, such

as opening of the air supply and exhaust dampers in the safety injection pump rooms in

preparation for safety injection pu'mp operation. These other functions are not required

to mitigate the consequences of a fuel handling accident, and therefore are not required to be OPERABLE.

[ When VIAS is inoperable, CORE ALTERATIONS and REFUELING OPERATIONS in

) containment- are immediately suspended. This effectively precludes a fuel handling accident from occurring. When "immediately" is used as a completion time, the required

4 action should be pursued without delay and in a controlled manner. Suspension of CORE j ALTERATIONS and REFUELING OPERATIONS shall not preclude completion of l movement of a component to a safe, conservative position.

i 2-39p Amendment No.

i

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

4 hl b i 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueling Bases (Continued)

, 2.8.3(1) Spent Fuel Assembly Storace i

The spent fuel pool is designed for noncriticality by use of neutron absorbing material.

The restrictions on the placement of fuel assemblies within the' spent fuel pool, according to Figure 2-10, and the accompanying LCO, ensures that the km of the spent fuel poo!

always remains < 0.95 assuming the pool to be flooded with unborated water.

A spent fuel assembly may be transferred directly from the reactor core to the spent fuel pool Region 2 provided an independent verification of assembly burnups has been completed and the assembly burnup meets the acceptance criteria identified in Figure 2-10.

When the configuration of fuel assemblies stored in Region 2 (including the peripheral cells) is not in accordance with Figure 2-10, immediate action must be taken to make the necessary fuel assembly movement (s) to bring the configuration into compliance with Figure 2-10. Acceptable fuel assembly burnup is not a prerequisite for Region 1 storage because Region 1 will maintain any type of fuel assembly that the plant is licensed for in a safe, coolable, subcritical geometry.

The provisions of Specification 2.0.1 for Limiting Conditions for Operations are not applicable. If moving fuel assemblies while in MODES 4 or 5, LCO 2.0.1 would not specify any actions. If moving fuel assemblies in MODES 1, 2, or 3, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown. When "immediately" is used as a completion time, the required action should be pursued without delay and in a controlled manner.

2.8.3(2) Spent Fuel Pool Water Level The minimum water level in the spent fuel pool meets the assumption of iodine decontamination factors following a fuel handling accident. When the water level is lower than the required level, the movement of irradiated fuel assemblies in the spent fuel pool is immediately suspended. This effectively precludes a fuel handling accident from occurring in the spent fuel pool. Suspension of REFUELING OPERATION shall not preclude completion of movement of a component to a safe, conservative position. The provisions of Specification 2.0.1 for Limiting Conditions for Operations are not applicable.

If moving fuel assemblies while in MODES 4 or 5, LCO 2.0.1 would not specify any actions. If moving fuel assemblies in MODES 1, 2, or 3, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown. When "immediately" is used as a completion time, the required action should be pursued without delay and in a controlled manner.

2-39q Amendment No.

L NGw' l

l 2.0 ' LIMITING CONDITIONS FOR OPERATION 4 2.8 - Refuelinn j Basy (Continued) 1 218.3(3) Spent Fuel Pool Boron Concentration I

The basis for the 500 ppm boron concentration requirement with Boral poisoned storage racks is to maintain the k, below 0.95 in the event a mistoaded unirradiated fuel assembly

! is located next to a spent fuel assembly. A misloaded unirradiated fuel assembly at i

' maximum enrichment condition, in the absence of soluble poison, may result in' exceeding the design effective multiplication factor. Soluble boron in the spent fuel pool water, for which credit is permitted under these conditions, would assure that the effective multiplication factor is maintained substantially less than the design condition, i

This LCO applies whenever unirradiated fuel assemblies are stored in the spent fuel pool.

i The boron concentration is periodically sampled in accordance with Specification 3.2.

! Sampling is performed prior to movement of unirradiated fuel to the spent fuel pool and

periodically when unirradiated fuel is stored in the spent fuel pool.

The provisiora of Specification 2.0.1 for Limiting Conditions for Operations are not

l. applicable. If moving fuel assemblies while in MODES 4 or 5, LCO 2.0.1 would not j specify any actions. If moving fuel assemblies in MODES 1, 2, or 3, the fuel movement

-is independent of reactor operation. Therefore, inability to suspend movement of fuel

[ assemblies is not sufficient reason to require a reactor shutdown.

f When "immediately" is used as a completion time, the required action should be pursued l without delay and in a controlled manner. Suspension of refueling operations shall not I preclude completion of movement of a component to a safe, conservative position.

2.8.3(4) Spent Fuel Pool Area Ventilation i The spent fuel pool area ventilation system contains a charcoal filter to prevent release of i significant radionuclides to the outside atmosphere. The system does not automatically realign and therefore must be IN OPERATION prior to REFUELING OPERATIONS in the spent fuel pool. When the. spent fuel pool area ventilation system is not IN OPERATION, the movement of irradiated fuel assemblies in the spent fuel pool is j immediately suspended. This' effectively precludes a fuel handling accident from occurring in the spent fuel pool. When "immediately" is used as a completion time, the required

. action should be pursued without delay and in a controlled manner. Suspension of REFUELING OPERATIONS shall not preclude completion of movement of a component to a safe, conservative position.

I-2-39r Amendment No.

__ _ _ _.. _..-_ _- _ _ _ _ ~ _ _ _ _ . _ _ . . .- . - . _ . _ .

NW

! 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Befueline Bases (Continued) 2.8.3(4) Spent Fuel Pool Area Ventilation (Continued) i The provisions of Specification 2.0.1 for Limiting Conditions for Operations are not i

applicable. If moving fuel assemblies while in MODES 4 or 5, LCO 2.0.1 would not specify any actions. If moving fuel assemblies in MODES 1, 2, or 3, the fuel movement l is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

2.8.3(5) Ventilation Isolation Actuation Sienal (VIAS) l A Ventilation Isolation Actuation Signal (VIAS) is initiated by a Safety Injection Actuation l Signal (SIAS), a Containment Spray Actuation Signal (CSAS) or a Containment Radiation l High Signal (CRHS). During REFUELING OPERATIONS, only the CRHS is required to respond to a fuel handling or reactivity accident. The requirements of this specification l are met when the Containment / Auxiliary Building Stack Swing Monitor (RM-052) and the L Auxiliary Building Stack Radiation Monitor (RM-062) are OPERABLE, monitoring the l Auxiliary Building exhaust stack, powered from independent 480-VAC buses and capable

! of actuating both the A and B trains of VIAS When the RCS is below 300*F, the 480-l VAC buses may be supplied by a single 4160-VAC power source. Above 300 F,

{

Specification 2.7 requires both 4160-VAC buses to be operable. In addition, one manual )

actuation channel is required to be OPERABLE. (Note, the Offsite Dose Calculation )

Manual' may have additional requirements / restrictions concerning operation of these l monitors.)

l In the event that one of the above radiation monitors becomes inoperable, or both are OPERABLE but RM-052 is not monitoring the exhaust stack, or VIAS manual actuation capability is inoperable, REFUELING OPERATIONS must be suspended thus precluding the possibility of a fuel handling accident The doses calculated at the exclusion area boundary (EAB) and low population zone (LPZ) for a fuel handling accident in'the spent fuel pool are well within 10 CFR 100 limits using conservative assumptions i.e., all rods in a single assembly fail with no credit taken for iodine filtration by VA-66.

VIAS aligns the control room air filtration system to the filtered air makeup mode, which !

prevents significant radionuclides from entering the contro! room. VIAS also initiates other actions, such as opening of the air supply and exhaust dampers in the safety injection pump rooms in preparation for safety injection pump operation. These other functions are L not required to mitigate the consequences of a fuel handling accident, and therefore are not .

~ required to be OPERABLE.

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! 2-39s Amendment No.

l

[ M 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueline j Bases (Cantinued)

! - 2.8.3(5) Ventilation Isolation Actuation Signal (VIAS) (continued)

^

When conducting REFUELING OPERATIONS in the spent fuel pool during MODES 1 and 2, LCO 2.15 is also applicable to VIAS. The allowable bypass condition for inoperable CRHS during MODES 1 and 2 is to close the containment pressure relief, air

sample, and purge system valves. This is justified because a SIAS or CSAS will still i initiate a VIAS. Since SIAS and CSAS would not initiate in response to a fuel handling i accident, both the actions of this specification and 2.15 must be followed when the CRHS

j. is inoperable in MODES 1 or 2 and REFUELING OPERATIONS are being conducted in i the spent fuel pool.

i i When VIAS is inoperable REFUELING OPERATIONS in the spent fuel pool are

immediately suspended. This effectively precludes a fuel handling accident from

! occurring. When "immediately" is used as a completion time, the required action should j' be pursued without delay- and in a controlled manner. Suspension of REFUELING 1 OPERATIONS shall not prec'ude completion of movement of a component to a safe, j

conservative position.

References l j (1) USAR Section 9.5 l

(2) USAR Section 14.18 1 .

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l 2-39t Amendment No.

4 i 2.0 LfM?TTNG CON 0 fit 0NS FOR OPE 8ATTON l 2.12 Centrel Doom systems Acelicabi14tv A

lies to the control room air conditioning and filtering systems Obie tive To limi and post the environmental conditions in the control room, upder normal j A conditions.

/

j Soecification

1. If the cont room air temperature reaches 105'F immediate action i shall be tak

! temperature wi anreduce this temperature and tp monitor the the control cabinets. If the temperature within the control. room 'r the control cabinets excpfds 120'F and cannot be i reduced below 120' in fcur hours, the rea ter will be placed in hot shutdown within the allowing six (6) ho s.

i

2. A thermometer must be the control om at all times.

3. All the areas of the plan which h a safety related instrumentation

} will be observed during hotKune onal testing to determine local l temperatures and monitored dur operation if normal plant j ventilation is not available.

l 4. With one control room air f ltration system inoperable, restore the 4

inoperable system to oper ble status %ithin 7 days or be in at least

{

cold shutdown condition ithin the foRowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

11111 The reactor protectiv system panels and the eng eered safety features panels were designed for, and the instrumentation s tested at, 120'F.

There is a maxim 5'F temperature difference bet n the inside and outside of the c trol cabinets. Therefore, if the tisperature of the control room e eeds IC5'F and if the temperature diffQential between the inside and ou ide of the control cabinet cannot be reduced to obtain an inside temp ature no greater than 120*F within four hourh the reactor will be sh down and the conditions corrected to preclude o eration of cpmponen in an untested environment.

If th control room air treatment system is found to be inopera e, there is n immediate threat to the control room and reactor operation y co inue for a limited period of time while repairs are being made.

If the s tem cannot be repaired within seven (7) days, the reactor is shutdown nd brought to cold shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2-59 Amendment No. 15,128,130 e

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! 2.0 - LIMITING CONDITIONS FOR OPERATION !

2.12 Control Room Systems l 2.12.1 Control Room Air Filtration System - Operatina l Applicability Applies to the operational status of the control room air filtration system when the reactor l l coolant temperature Ty u _>_ 210"F. '

l Obiective To assure operability of equipment required to filter control room air following a Design Basis Accident.

l L Specification Two control room air filtration trains shall be OPERABLE.

Reauired Actions (1) With one control room air filtration train inoperable, restore the inoperable train ,

to OPERABLE status within 7 days. '

(2) With the required actions of (1) not met, be in IIOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

(3) With two control room air filtration trains inoperable, enter LCO 2.0.1 immediately.

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i 2-59 Amendment No. 15, 128, 130 !

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N2W' d

2.0 LIMITING CONDITIONS FOR OPERATION 3 2.12 Control Room Systems j

2.12.2 Control Room Air Conditioning System

)_ Anolicability 4 Applies to the operational status of the control room air conditioning system when the

reactor coolant temperature T,oia 1 210 F.

i Obiective i

j To assure operability of equipment required to maintain air temperature within the control room following a Design Basis Accident.

l 1' Specification 4 Two control room air conditioning trains shall be OPERABLE.

Required Actions l'

j (1) With one control room air conditioning train inoperable, restore the inoperable

, train to OPERABLE status within 30 days.

4 l (2) With the required actions of (1) not met, be in HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , j and COLD SHUTDOWN within the following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

i.

(3) With two control room air conditioning trains inoperable, enter LCO 2.0.1 immediately, i

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2-59a Amendment No.

N ELU 2.0 LIMITING CONDITIONS FOR OPERATION 2.12 Control Room Systems Bases 2.12.1 Control Room Air Filtration System - Operatine The control room air filtration system is designed to maintain radiation doses to control room personnel within the limits of General Design Criterion (GDC) 19. When the control room ventilation system is placed in the filtered air makeup mode either manually or after receiving a VIAS, the unfiltered outside air duct is isolated to prevent significant radionuclides from entering the control room.

A control room air filtration train is OPERABLE when the associated train level components and the system level components are OPERABLE and the train can provide filtered outside air and recirculation air to the control room. Train level components consist of the outside air filter unit isolation dampers (PCV-6680A-1, PCV-6680B-1), the outside air filter unit fan (VA-63A, VA-63B), the outside air filter unit (VA-64A, VA-64B), and the outside air filter unit isolation damper (PCV-6680A-2, PCV-6680B-2) and associated ductwork.

i l

System level components consist of the unfiltered outside air duct isolation dampers (PCV-6681A and PCV-6681B), the recirculation duct isolation damper (PCV-6682) and associated ductwork. IF either or both unfiltered outside air duct isolation dampers (PCV-6881A, PCV-6681B) are inoperable, the control room air filtration system is considered OPERABLE if the unfiltered outside air duct is isolated. If only a single l unfiltered outside air duct isolation damper is OPERABLE and the unfiltered outside air )

duct is not isolated, then the 7 day LCO applies. If both unfiltered outside air duct ;

isolation dampers are inoperable concurrently with an unisolated flowpath through the l unfiltered outside air ductwork to the control room, then both trains are inoperable and LCO 2.0.1 applies. I h cecirculation duct does not require redundant dampers to meet single failure proof criteaa. Damper PCV-6682 meets the acceptance criteria for the damper repair option described in Standard Review Plan 6.4, Appendix A. A radioactivity release requires PCV-6682 to open, should PCV-6682 fail to open, it can be repaired or repositioned open before control room doses exceed the allowable limits of GDC 19.

With the reactor coolant temperature Tm .>_ 210"F, two trains of the control room air filtration system are required to be OPERABLE. If one train .s inoperable it shall be restored to OPERABLE status within 7 days. In this condition the remaining train is adequate to perform the control room radiation protection function. The 7 day completion time is based on the low probability of an accident occurring during this time period, and the ability of the remaining train to provide the required function.

2-59b Amendment No.

NEW 2.0 LIMITING CONDITIONS FOR OPERATION ;

2.12 Control Room Systems Bases (Continued) 2.12.1 Control Room Air Filtration System - Operatine (Continued)

If the inoperable train cannot be restored to OPERABLE within the allowed completion l time, the plant must be placed in a MODE where the specification is no longer applicable. l With two trains inoperable, the control room air filtration system may not be capable of l performing its design function and the plant nust be placed in a MODE where the specification is no longer applicable.

2.12.2 Control Room Air Conditionine System I l

The control room air conditioning system is required to ensure the control room '

temperature will not exceed equipment OPERABILITY requirements. The reactor protective system panels and the engineered safety features panels were designed for, and the instrumentation was tested at,120"F. The temperature inside the control cabinets is at most 15 F warmer than the temperature of the control room due to heat produced by the electronic circuitry. Therefore, the temperature of the control room will not affect OPERABILITY of the control cabinets as long as it doesn't exceed 105 F.

During non-emergency operation, the control room temperature may be maintained by using Component Cooling Water (CCW). During design basis accident conditions, the CCW isolation valves to air conditioning units (VA-45A and VA-46B) are automatically closed on a VIAS. This prevents CCW that has been heated by components following a design basis accident from adding heat to the control room. When VIAS is in override, closing these valves maintains the OPERABILITY of the associated air conditioning unit.

With the reactor coolant temperature Ta ;>_. 210 F, two trains of the control room air conditioning system are required to be OPERABLE. If one train is inoperable it shall be restored to OPERA'BLE status within 30 days. In this condition the remaining train is acequate to maintain the control room temperature. With both trains inoperable, the control room air conditioning system may not be capable of performing its intended function and LCO 2.0.1 must be entered immediately.

References (1) USAR Section 9.10 2-59c Amendment No.

TABLE 3-3 (Continued)

MINIMUM FREOUENCIES FOR CHECKS. CALIBRATIONS AND TESTING OF MISCELLANEOUS INSTRUMENTATION AND CONTROL _S Surveillance Channel DescrisWion Function Frequency Surveillance Method

8. Dropped CEA Indication a. Test R a. Insert a negative rate of change power signal to all four Power Range Safety Channels to test alarm.

b. Test R b. Insert CEA's below lower electrical limit to test dropped CEA alarm.

9 Calorimetric Instrumen- a. Calibrate R a. Apply known d/p to feedwater flow tation sensors.

10. Control Room Ventilation a. Test R a. Check damper operation for DBA mode.

b. Test R b. Check centrol room for positive pressure.

I1. Containment llumidity a. Test R a. Place sensor in a known high humidity Detector atmosphere.

12. Interlocks-Isolation Valves a. Test R a. Known pressure of 265 psia applied to Shutdown Cooling Line both pressure transmitters.

-u w

13. Control Room nemme a. Test R a. Cempare reading ' "5 c 'ibr :M ter Air ConditioninpSfstefris"" the cre e- !r net "% l-2%

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i

-l TABLE 3-4 (Continued)

MINIMUM FREOUENCIES FOR SAMPLING TESTS 1

Type of Measurement Sample and Analysis and Analysis Frequencies !

I l

1. Reactor Coolant i

! ~'

(Continued)

(c) Cold Shutdown (1) Chloride 1 per 3 days j (Operating Mode 4)

$ (d) Refueling Shutdown (1) Chloride 1 per 3 days m ,

3 (Operating Mode 5) (2) Boron Concentration 1 per 3 days

I

j j (e) Refueling Operation (1) Chloride 1 3 da

I

(2) Boron Concentration 1 per [ day [Wf4*

2. SIRW Tank Boron Concentration ' p: : 21 & y: M

3. Concentrated Boric Boron Concentration ! per' tys W Acid Tanks 1 4. St Tanks Boron Concentration ' per 3! &y(M

5. Spent Fuel Pool Boron Concentration ! per 21 ty: Se5]F66fn6fs!4 beW

6. Steam Generator Blowdown Isotopic Analysis for Dose ! per ty:* h

, (Operating Modes 1 and 2) Equivalent I-131 1

.i (1) Until the radioactivity of the reactor coolant is restored to .< I pCilgm DOSE EQUIVALENT I-131. ,

] l (2) Sample to be taken after a minimum of 2 EFPD and 20 days of power operation have elapsed since reactor was j subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

(3) Boron and chloride sampl' g/ analyses are not r uired when the core has been off-loaded. Reinitiate boron and chloride sampling /analy s =: d": prior to r loading fuel into the cavity to assure adequate shutdown margin ;

and allowab ofide vels_arg _

${Pj Pridfid}1ayny{mirMiated;fuelWmbliesiin

@)oresMth spen (fuelhlj thejpen[fue[p60[im4Mskighe(Mrriul@eO (4)(5)]*% len Steam Generator Dose Equivalent 1-131 exceeds 50 percent the limits in Specif on 2.20, the sampling

] an analysis frequency shall be increased to a minimum of 5 times per week. When Steam Generator Dose l uivalent I-131 exceeds 75 percent of this limit, the sampling and analysis frequency shall be increased to a minimum of once per day.

1 1

3 19 Amendment No. 28,67,86,124d31rl#2rl-7-2

TABLE 3-5 (Continued)

Test Freauency

17. liydrogen Purge 1. Verify all manual valves are operable by R System completing at least one cycle.

2. Cycle each automatic valve through at R least one complete cycle of full travel from the control room. Verification of ;

the valve cycling may be determined by the ,

observation of position indicating lights.

3. Initiate flow through the VA-80A and VA-80B blowers, IIEPA filter, and charcoal adsorbers !

and verify that the system operates for at least !

(a) 30 minutes with suction from the a) M ,

auxiliary building (Room 59)

(b) 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months with suction from the b) R ,

containment t

4. Verify the pressure drop across the R l VA-82 HEPAs and charcoal filter to be less t than 6 inches of water. Verify a system flow j rate of greater than 80 scfm and less than 230 scfm during system operation when tested ,

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U.S. Nuclear Regulatory Commission LIC-97-0050 f

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ATTACHMENT B 1,

4 4

4 3

4 0

5

l DISCUSSION, JUSTIFICATION AND NO SIGNIFICANT HAZARDS C0KSIDERATION DISCUSSION AND JUSTIFICATION The Omaha Public Power District (0 PPD) proposes to revise the Fort Calhoun Station (FCS) Unit No.1 Technical Specifications (TS) 2.12, " Control Room :

Systems," to delete the Limiting Condition for Operation (LCO) and associated surveillance for the control room temperature and replace it with an LC0 and surveillance on the control room air conditioning (A/C) system. The remainder of TS 2.12 is being rewritten consistent with the requirements of the Combustion Engineering Standard TS (NUREG-1432 Rev.1). In reviewing 1 requirements for refueling and shutdown operations, additional TS improvements I were identified. Therefore, the definition section, TS 2.1 " Reactor Coolant I System," 2.6 " Containment System," 2.8 " Refueling Operations," and associated surveillance requirements are proposed for revision to incorporate the design ,

basis requirements for refueling operations and to correspond to NUREG-1432. l l

BACKGROUND l l

0 PPD identified a potential accident scenario that could have resulted in the '

inoperability of both control room A/C units. The scenarios involved either a large break Loss of Coolant Accident (LOCA) or a Main Steam Line Break (MSLB) l inside containment. An evaluation showed that under certain conditions, post-accident Component Cooling Water (CCW) temperature could have risen to a point at which the control room air A/C unit's compressor would have shutdown. This potential accident scenario was reported in Licensee Event Report (LER) 50-285/94-10, dated December 14, 1994 and LER 94-10 Revision 1, dated March 31, 1995. As a corrective action, a modification was completed that installed two air-cooled condenser units and provided for isolation of the CCW system from the control room A/C system upon receipt of a Ventilation Isolation Actuation Signal (VIAS).

The current control room A/C system consists of two redundant A/C units, VA-46A and VA-468. Each unit has sufficient capacity to meet the cooling requirements for personnel and equipment inside the control room envelope.

Each A/C unit is equipped with an air-cooled condenser located inside a protective enclosure outdoors on the roof of the Auxiliary Building. Each A/C unit's refrigerant compressor, air cooling coils, fans, and dampers are located inside the control room envelope. Each unit has a waterside economizer coil that allows air cooling with CCW. When cooling water temperature is sufficiently low, a temperature-activated valve at each A/C unit allows cooling water flow through the waterside economizer.

This valve also diverts flow away from the waterside economizer if cooling water temperature is too high. The air-operated CCW isolation valves to the A/C units fail closed and are automatically closed on a VIAS to prevent CCW flow through the waterside economizers in a post-accident situation.

1

DISCUSSION AND JUSTIFICATION (Continued):

LIMITING CONDITIONS FOR OPERATION CONTROL ROOM SDecification 2.12 " Control Room Systems" Soecification 2.12(1) 1 TS 2.12(1) requires that the temperature within the control room and control cabinets be maintained below 120 F. This value does not meet any of the four I criteria contained in 10 CFR 50.36 for inclusion in TS. However, the ;

equipment required to maintain this temperature, the control room air i conditioning system, meets Criterion 3 of 10 CFR 50.36 in that the system functions to mitigate a design basis accident by maintaining the control room in a habitable environment. Therefore, it is proposed that this TS be revised i

to delete the control room temperature as an LC0 and require that two control room air conditioning trains be operable when the reactor coolant temperature !

I is above 210 F. The design temperature limits of instrumentation and controls I inside the control room are discussed in the Basis Section of TS 2.12.

The allowed outage time for one train of control room air conditioning is proposed as 30 days. This is consistent with Combustion Engineering Standard 3 TS 3.7.12 (NUREG-1432 Rev. 1). In addition, the FCS Probabilistic Risk i Assessment model was reviewed and validated a 30-day outage time as non-risk significant. The impact on Core Damage Frequency (CDF) from a 30-day LC0 was based on the assumption that one cooling unit was always inoperable and thus under the LC0 for the entire year. This allows the analysis to consider unlimited entries into the LCO and a full LC0 duration for each entry. Using l this assumption, the baseline (annually) CDF of 1.53E-5 would increase by

21.6% to a frequency of 1.86E-5. By EPRI's "PSA Applications Guide," this

,' small increase in CDF can be classified as "non-risk significant."

Specification 2.12(2)

TS 2.12(2) requires that a thermometer be in the control room at all times.

l This instrumentation does not meet any of the four criteria contained in 10 CFR 50.36 for inclusion in the TS. Therefore, OPPD proposes to relocate this requirement to the FCS Updated Safety Analysis Report (USAR).

Specification 2.12(3)

TS 2.12(3) requires that all areas of the plant containing safety related instrumentation be observed during hot functional testing to determine local temperatures and monitored during operation if normal plant ventilation is not available. It is proposed to delete this TS. The requirement to monitor and determine local temperatures during hot functional testing was met during the initial startup phase of FCS and is no longer applicable.

2

k l DISCUSSION AND JUSTIFICATION (Continued):

The requirement to monitor temperatures within the plant during normal operation does not meet any of the four criteria contained in 10 CFR 50.36 for inclusion in TS and therefore is being deleted. The requirement to control 4

temperatures for safety related instrumentation and controls, and initiate supplementary cooling if required, is currently described in USAR Section 9.10. These USAR requirements are controlled by plant procedures. Any changes to these requirements would require that an evaluation be conducted in accordance with 10 CFR 50.59.

i Specification 2.12(4)

TS 2.12(4) allows one control room air filtration system to be inoperable for 7 days or a plant shutdown must be commenced. This specification does not state which modes of operation it applies to. Therefore, it is proposed to i revise this specification to require two trains of control room air filtration j systems to be operable when the reactor coolant temperature is above 210 F.

The allowed outage time will be maintained at 7 days and a total of 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months will be allowed to take the plant to cold shutdown. The 42 hour4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months time period i is consistent with TS 2.0.1 which addresses equipment outages in excess of

) what is specifically allowed by individual specifications.

REFUELING OPERATIONS The design bases of the fuel handling accident and refueling operations were j

^

reviewed and several TS inadequacies were identified related to refueling I operations. Actual plant operation in accordance with plant procedures has !

been within the plant design bases. Due to these inadequacies, revisions are !

proposed for the TS Definitions section, TS 2.6 on containment integrity, and '

TS 2.8 on refueling operations to reflect NUREG-1432.

Definitions Cold Shutdown Condition & Refuelina Shutdown Condition i

Clarifications are proposed for the definitions of Cold Shutdown Condition (Mode 4), and Refueling Shutdown Condition (Mode 5) to better clarify when these definitions apply concerning boron concentration. As currently stated, j the definitions apply when boron concentration is g shutdown boron

. concentration and d refueling boron concentration, respectively. Thus, a boron concentration greater than these values is not specifically addressed by the current definitions, yet the definitions are still applicable.

l 3

3 3

J DISCUSSION AND JUSTIFICATION (Continued):

I Refuelino Operations l

l The definition of Refueling Operations is being revised to delete control l element assemblies (CEA) or startup sources from the definition as these are items that are included in the definition of Core Alterations. Additionally, it is being revised to specify that the definition is limited to movement of irradiated fuel outside of the reactor pressure vessel as fuel movement inside the reactor vessel is included in the definition of Core Alteration. Finally, i

a clarification is being added to state that suspension of refueling l operations shall not preclude completion of movement of irradiated fuel to a ,

safe, conservative position. !

Core Alteration l 1

f The definition for Core Alterations is being revised to reflect the i requirements of NUREG-1432. This revision deletes "any component" from the definition and clarifies that the components considered by this definition are 4

those that could affect reactivity. In addition, the revision adds nuclear i

fuel to the definition such that movement of fuel within the reactor vessel '

will be defined as a core alteration and not a refueling operation.

In Operation 4

The definition of In Operation is being revised to include the definition of i operable.

Specification 2.1 " Reactor Coolant System" ,

I It is proposed to revise TS 2.1.1(3) to include shutdown cooling (SDC) i l requirements when the reactor coolant system (RCS) temperature is below 210 F l with fuel in the reactor and the reactor vessel head fully tensioned. The definitions of Cold Shutdown (Mode 4) and Refueling Shutdown (Mode 5) 4 contained in the TS make no distinction as to the status of the reactor vessel head or RCS temperature. The only difference between the two defined modes is

boron concentration. Higher or lower boron concentration affects shutdown margin but does not affect decay heat load, which is the basis for this

specification.

Technical Specification 2.1.1(4) was intended to address shutdown cooling requirements during Mode 5 with the vessel head removed, consistent with the

definition of the refueling mode contained in Standard TS. However, additional requirements for shutdown cooling are addressed by TS 2.8. )

I Therefore, it is proposed to delete TS 2.1.1(4) and the exception. New specifications addressing shutdown cooling loop requirements during MODE 5 i

with fuel in the reactor and with one or more reactor vessel head closure bolts less than fully tensioned, are proposed for TS 2.8 (Refueling). ,

f 4

4

.. - - - - ~ - . - - - - . - . . . - - -. .-

DISCUSSION AND JUSTIFICATION (Continued):

The associated statements supporting these items in the Basis section are also proposed for deletion. Prior to any reactor vessel head closure bolts being j loosened, TS 2.1.1 will be applicable which will require two shutdown cooling !

loops. As soon as a closure bolt is loosened, the new proposed TS 2.8 would i be applicable, which also requires two shutdown cooling loops whenever there is a level of less than 23 feet of water above fuel in the reactor core.

jipfcification for Shutdown Coolina I

'It is proposed to incorporate the requirements for shutdown cooling into two new Specifications: 2.8.1(3) and 2.8.1(4). The proposed specifications are similar. to NUREG-1432 Specifications except for reauired water level and the exception to allow shutdown cooling to be inoperable far a limited period of time as currently allowed by TS 2.1.1.

Water Level NUREG-1432 requires that two trains of SDC be operable when there is less than 23 feet of water above the reactor vessel flar.ge. The basis for this requirement is that this level is consistent with the level required during fuel movements. It is proposed that this level be 23 feet above the fuel in the core instead of 23 feet above the reactor vessel flange. Requiring 23 feet of water above fuel in the reactor core is much more conservative than the 15 feet currently' required as considerably more water is provided for a heat sink. This value is also consistent with other Combustion Engineering plants, including ANO Unit 2, Calvert Cliffs Unit 2, San Onofre Units 2 and 3, Saint Lucie Unit 1, and a proposed change submitted by Waterford 3. A separate TS will require 23 feet of water above the vessel flange for fuel movements, as this level is not based on decay heat requirements.

Exceptions for Inoperable SDC Trains Currently, TS 2.1.1(3) has an exception to allow both trains of SDC out of service for eight hours provided (1) no operations are permitted that would cause dilution of the reactor coolant boron concentration, (2) no refueling

. operations are taking place, and (3) all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere are closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . It is proposed that this exception be maintained in the new TS 2.8.1(3) but additional restrictions and guidance be incorporated into the Basis Section to allow the SDC loops to be inoperable for up to eight hours under administrative controls provided that the SDC system is available.

5

DISCUSSION AND JUSTIFICATION (Continued):

Currently, this exception is used during engineered safety feature testing l which requires that the system be put in alignments such that minor operator i actions would be required for the system to be considered fully operable. l Allowing the equipment to be inoperable but available for a period up ta eight l hours is more conservative than the current TS.

4 A SDC loop is considered available under administrative controls if there are:

(1) approved procedures, (2) a dedicated operator stationed at the controls if they are outside of the control room, and (3) direct communication between the

dedicated operator and the control room. Similarly, the SDC system is i considered available under administrative controls when an operator is not at the location of the controls provided: (1) procedural guidance is consulted

, prior to removing SDC from service to determine the time-to-boil, and (2)

! there is sufficient time for the operator to travel to the local controls and j perform the required actions.

These proposed requirements are consistent with the definition and requirements for available equipment contained in the draft rule on shutdown operations, dated April 5,1996. Additional changes to shutdown cooling requirements may be necessary when the shutdown rule is issued; however, the proposed TS is more conservative than current requirements.

This exception will be limited to periods when the water level is 23 feet or more above the fuel in the reactor vessel.

The requirements of proposed TS 2.1.1(3) which will address shutdown cooling requirements when the vessel head is installed, are similar to NUREG-1432 Specifications 3.4.7 and 3.4.8. The difference is that the Standard TS has divided the requirements into two specifications, one when RCS loops are filled (steam generators are operable) and one when RCS loops are not filled.

Specification 2.6 " Containment System" Specification 2.6(1)c is proposed for deletion. Specification 2.6(1)c requires containment integrity when the reactor vessel head is removed if the boron concentration is less than refueling boron concentration. However, TS 2.6(1)c has no required actions and therefore, TS 2.0.1 must be entered when the LC0 is not met. In this situation, (reactor vecsel head removed), TS 2.0.1 is ineffective because the plant would alret- be in Refueling Shutdown.

Specification 2.6(1)d is also proposed for deletion as it is overly restrictive and unnecessary. Specification 2.6(1)a does not require containment integrity when the reactor is in Cold Shutdown or Refueling Shutdown. However, TS 2.6(1)d requires containment integrity regardless of operating mode for positive reactivity changes made by CEA motion or boron dilution unless limited to testing one control element drive mechanism (CEDM) at a time.

6

3

~

DISCUSSION AND JUSTIFICATION (Continued):

4 Specification 2.8.1(1) as proposed eliminates the need for containment I

integrity when the reactor is in Refueling Shutdown. Specification 2.8.1(1) requires the reactor coolant to be at refueling boron concentration (a shutdown margin of not less than 5% with all CEAs withdrawn) when-the reactor is in Refueling Shutdown. In the event this limit is exceeded (i.e., shutdown margin of less than 5%), TS 2.8.1(1) requires that positive reactivity '

i additions be immediately suspended and immediate actions be taken to restore the boron concentration to within limits.

Small positive reactivity increases whether by CEA motion or boron dilution i will not cause a criticality event due to the need to maintain at least a 5%

shutdown margin. Therefore, the requirement to maintain containment integrity i is unnecessarily restrictive since a criticality event cannot occur when a 1

shutdown margin of at least 5% exists. Specification 2.8.1(1) as proposed is i

consistent with the requirements of NUREG-1432, Specification 3.9.1.

{ Specification 2.8.2(1) is proposed to provide requirements for containment closure during core alterations and refueling operations inside containment.

As stated in USAR Section 14.18, the fuel handling accident does not take 4

credit for containment isolation. Therefore, requiring containment integrity

. for a fuel handling accident is also unnecessary and requirements for l containment closure are proposed for addition to TS 2.8 consistent with NUREG-1432, Specification 3.9.2.

Specification 2.10.2 governs operation of CEAs and monitoring of selected core l parameters. Specification 2.10.2 ensures (1) adequate shutdown margin j following a reactor trip, (2) that the moderator temperature coefficient (MTC) 4 is within the limits of the safety analysis, and (3) control element assembly operation is within the limits of the setpoint and safety analysis.

l Specification 2.10.2 ensures that the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality and provides actions (i.e.,

boration) to be taken to ensure that the required shutdown margin is available. Thus, TS 2.10.2 precludes the need for containment integrity when

the plant is in Cold Shutdown.

Specification 2.8 " Refueling Operations"

It is proposed that TS 2.8 be rewritten to reflect NUREG-1432. Currently, this specification applies to any refueling operation. However, no distinction is made between refueling operations within containment and refueling operations within the spent fuel pool. In addition, several initial

assumptions of a fuel handling accident are not addressed by the current TS t

2.8.

7

. - -- - -.- . .- - . - - - . _ _ - - . - - . - - =_ -.

l 1

l !

DISCUSSION AND JUSTIFICATION (Continued):

Specification 2.8(1)

The current TS 2.8(1) is inadequate. This specification requires that the j equipment hatch and one door in the Personnel Air Lock be properly closed, and l

. all automatic containment isolation valves be operable or at least one valve 1 i closed. However, TS 2.8(1) does not define what is meant by a properly closed i equipment hatch; that information is contained in the Basis of TS 2.1.1.

4

.i

In addition, inclusion of all automatic containment isolation valves instead '

of those providing direct access to the outside atmosphere is incorrect. The

containment isolation system is defined in USAR Section 5.9.5 as those devices actuated by a Containment Isolation Actuation Signal (CIAS) or a Steam Generator Isolation Signal (SGIS). This includes many valves such as the j containment cooling and filtering unit's component cooling water supply and

. return valves that have no design basis function during a fuel handling

! accident. A CIAS is initiated by a Containment Pressure High Signal or a Pressurizer Pressure Low Signal. Neither of these signals are required to be

, operable during refueling operations as these signals would/could not respond i

to a fuel handling accident.

. The correct requirements are specified in TS 2.8(2) which only requires that j closure be initiated by the Ventilation Isolation Actuation Signal (VIAS) for j -the containment pressure relief, air sample, and purge system ~ valves. Due to j these inadequacies, it is proposed to delete TS 2.8(1) and replace it with a 4

new specification (2.8.2(1)). The new specification is consistent with NUREG-j 1432, Specification 3.9.3.

Soecification 2.8(2)

It is proposed that TS 2.8(2) be deleted and replaced by new Specifications 2.8.2(3) and 2.8.3(5). The requirement to maintain an operable Ventilation .

Isolation Actuation Signal with input from the containment atmosphere gaseous I and auxiliary building exhaust stack gaseous radiation monitors is consistent ;

with current requirements and required actions are consistent with NUREG-1432, Specification 3.3.8. Radiation Monitor RM-052 functions as a " swing" monitor i.e., it can be aligned to monitor either containment or the auxiliary building exhaust ventilation stack. Radiai. ion Monitor RM-052 is powered by either MCC-3B1/AI-40C (like RM-051) or MCC-4C2/AI-40D (like RM-062).

Technical Specification 2.7, Electrical System, is not required to be applied when the RCS is below 300 F. Above 300 F, TS 2.7 requires both 4160-VAC buses to be operable. Thus, above 300 F the reqeired radiation monitors must be powered from independent 480-VAC buses supplied by independent 4160-VAC buses.

During refueling outages, bus alignments other than those used during power operation'are used to permit electrical system maintenance and modifications.

8

i DISCUSSION AND JUSTIFICATION (Continued):

In the loss of offsite power event, the radiation monitor sample pumps and control room HVAC units stop and will not restart until the emergency diesel generators (EDGs) reenergize the system. The fuel handling equipment also stops and does not restart when the EDGs reenergize the system, thus minimizing the potential of a fuel handling accident.

When the EDGs reenergize the buses, VIAS will operate as designed. Therefore, j when the RCS is below 300*F, the required monitors need only be powered from independent 480-VAC buses supplied by a single 4160-VAC bus.

There is_no need to assume that a fuel handling accident occurs immediately followed by a loss of offsite power. However, in the unlikely event that this should occur, there would be no effect on the site boundary dose since VIAS is not credited in USAR Section 14.18 (Fuel Handling Accident). In this situation, when the EDGs reenergize the buses, the control room HVAC units will restart in the filtered air makeup mode and the stack radiation monitor sample pump will restart. However, the containment radiation monitor sample lines remain isolated preventing the restart of the monitor sample pump after receipt of a VIAS. j Soecification 2.8(3)

It is proposed that TS 2.8(3) be deleted. This requirement does not meet any of the four criteria contained in 10 CFR 50.36 for inclusion in the TS. The requirements that radiation levels in containment and the spent fuel pool shall be monitored during refueling operations will be incorporated into the USAR.

Specification 2.8(4)

It is proposed that TS 2.8(4) be deleted and replaced with a new Specification 2.8.1(2). The requirement to maintain two source range monitors operable to monitor core alterations is consistent with the current TS and NUREG-1432, Specification 3.9.2.

Specification 2.8(5)

It is proposed that TS 2.8(5) be deleted and replaced with new Specifications 2.8.1(3) and 2.8.1(4). The requirement to maintain at least one shutdown cooling loop in operation with the exception that the loop may be removed from operation for one hour per eight hour period during activities in the vicinity of the reactor coolant hot leg is consistent with the current TS and NUREG-1432, Specification 3.9.4. As explained above, TS 2.1.1(3) currently has an exception allowing both trains of SDC to be taken out of service for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provided several conditions are met. This exception is being deleted from TS 2.1.1 and will be relocated to TS 2.8.1(3) along with additional restrictions and guidance resulting in a more conservative specification than currently exists.

9

4

, DISCUSSION AND JUSTIFICATION (Continued):

Specification 2.8(6)

It is proposed that TS 2.8(6) be deleted. This requirement does not meet any of the four criteria contained in 10 CFR 50.36 for inclusion in the TS. The requirements that direct communication between personnel in the control room and at the refueling machine shall be available whenever core alterations are taking place will be incorporated into the USAR.

Soecification 2.8(7)

It is proposed that TS 2.8(7) be deleted and replaced with a new Specification 2.8.3(4). The requirement to place the spent fuel pool ventilation system in 1 operation prior to refueling operations is consistent with the current TS. It

, is being clarified that this specification only applies to refueling e operations in the spent fuel pool, and not when conducting refueling

operations inside of containment. Additionally, it is being clarified that TS l 2.0.1 is not applicable to this activity, as reactor operation is independent 4

of fuel movement in the spent fuel pool.

Specification 2.8(9)

The current specification 2.8(9) is inadequate. This specification requires a

, minimum of 23 feet of water above the top of the core. This does not meet the initial conditions assumed in the fuel handling accident as documented in USAR i

Section 14.18. USAR Section 14.18 assumes 23 feet of vster above where the fuel could land if dropped. In order to meet this initial condition, a minimum of 23 feet of water above the reactor vessel flange is required, as l this is the highest point where a fuel bundle could land if dropped.

Procedures reflect the requirement to maintain 23 feet of water above the reactor vessel flange during refueling operations. The proposed Specification 3

2.8.2(2) is consistent with NUREG-1432, Specification 3.7.16.

Specification 2.8(10) i Specification 2.8(10) was deleted in Amendment 174 and the requirements relocated to the Design Features Section. Therefore, no new specification is proposed.

A Specification 2.8(11)

The current specification is inadequate. The specification provides restrictions on storage of fuel in the spent fuel pool; however there are no 4 required actions to address situations when the specification is not met. It I is proposed that TS 2.8(11) be deleted and replaced with a new Specification

2.8.3(1) that requires that a misloaded fuel assembly be moved immediately.

Additionally, it is being clarified that TS 2.0.1 is not applicable to this activity, as reacter operation is independent of fuel movements in the spent fuel pool .

' 10

DISCUSSION AND JUSTIFICATION (Continued):

l Specification 2.8(12)

It is proposed that TS 2.8(12) be deleted and replaced with Specification 2.8.3(3). The requirement to maintain 500 ppm boron concentration in the spent fuel pool whenever unirradiated fuel is stored there is consistent with the current TS and the required actions are consistent with NUREG-1432, Specification 3.7.17.

Restriction on Movement of Irradiated Fuel from the Reactor Core l The restriction on irradiated fuel movement unless the core has been subcritical for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if the reactor has been operated at power .

levels above 2% is proposed for relocation to the Bases of TS 2.8.2(2) for l consistency with NUREG-1432, B 3.9.6. 1 i

Reactor Coolant System Boron Concentration Currently, there is no specification for boron concentration; refueling boron concentration is included in the definition of Mode 5. However, there are no required actions to be taken if the boron concentration should be below the

, refueling concentration. Therefore, it is proposed that a new Specification 2.8.1(1) be incorporated consistent with NUREG-1432, Specification 3.9.1. l I

Spent Fuel pool Water Level l

Currently, there is no specification for spent fuel pool water level. The I

! water level of the spent fuel pool is an initial condition assumed in USAR Section 14.18. It is proposed that a new Specification 2.8.3(2) be l incorporated into TS 2.8 which is consistent with NUREG-1432, Specification j 3.7.16. ,

l SURVEILLANCE TEST REQUIREMENTS CONTROL ROOM Specification 3.1. Table 3-3. Item 13.

Technical Specification 3.1, Table 3-3, Item 13 requires that the thermometer in the control room be compared with a calibrated thermometer and replaced if out of tolerance on a refueling frequency. It is proposed that this surveillance be deleted to be consistent with deletion of the LCO requirement j to maintain a thermometer in the control room. A new surveillance is proposed

! to verify that the control room air conditioning system has the capability to remove the assumed heat load. This surveillance will ensure the operability

, requirements for TS 2.12 are met. The test and frequency are consistent with NUREG-1432.

i 11

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

3

! DISCUSSION AND JUSTIFICATION (Continued):

The air-operated CCW isolation valves to the A/C units fail closed and are automatically closed on a VIAS to prevent CCW flow through the waterside economizers in a post-accident situation. These valves are currently tested in accordance with TS 3.3 (FCS Inservice Testing Program). Prior to the modification, the valves were tested as fail-open valves. No TS changes are necessary.

l The control room air filtration system is currently tested on a refueling frequency in accordance with TS 3.2, Table 3-5, Item 10a. No TS changes are necessary.

i l REFUELING OPERATIONS Reactor Coolant Boron Concentration Durina Refuelino Operations The Reactor Coolant System boron concentration is currently sampled in accordance with TS 3.2, Table 3-4, Item 1(e). It is proposed to revise the frequency from once per shift during refueling operatie.S to once per 3 days, !

, which is consistent with NUREG-1432. As stated in the Nst a ci TS 2.8 and i L USAR Section 14.18, the reactor cavity is filled with o er 200,000 gallons of borated water prior to the start of refueling operations. The requirements j for sampling the reactor coolant during the remainder of Mode 5 (when fuel is 1

in the reactor) is once per 3 days in accordance with Table 3-4, Item 1(d).

This proposed change will make the sampling consistent with the requirements ,

of Item 1(d) and NUREG-1432. l i

i A change to TS 3.2, Table 3-4, Footnote 3 is also proposed. The change will

! require that boron concentration / chloride analyses be performed prior to reloading fuel into the cavity if not completed within the previous 3 days. I Currently, Footnote 3 states "one shift prior to reloading fuel." This change :

) is consistent with the guidance of NUREG-1432.

j Soent Fuel Pool Boron Concentration The spent fuel pool boron concentration is currently sampled in accordance

- with TS 3.2, Table 3-4, Item 5. It is proposed to revise the frequency of the sampling to prior to movement of unirradiated fuel in the spent fuel pool and once per week whenever unirradiated fuel is stored in the spent fuel pool to be consistent with the requirements of the LCO.

Source Ranae Neutron Monitors

Currently, a channel check and calibration of the wide range neutron monitors is performed in accordance with TS 3.1, Table 3-1, Item 2. No TS changes are i necessary.

4 12

DISCUSSION AND JUSTIFICATION (Continued):

SURVEILLANCE TEST REQUIREMENTS (Continued):

Containment Penetrations Currently, there is no surveillance to determine the status of containment penetrations during refueling operations. Therefore, a new surveillance is proposed for incorporation into Table 3-5 to verify the status of required containment penetrations once per 7 days consistent with NUREG-1432. The requirements of NUREG-1432 to verify that the containment purge and exhaust valves actuate to the isolation position on a refueling frequency are part of the Containment Radiation High Signal test required by TS 3.1, Table 3-2. Item 4.

Shutdown Coolina loops Currently, there is no surveillance requirement to verify that the required shutdown cooling loops are operable and in operation or to verify correct breaker lineup for the shutdown cooling loop that is not in operation.

Therefore, a new surveillance is proposed for incorporation into TS 3.2, Table 3-5 consistent with NUREG-1432.

Refuelina Water Level Currently, there is no surveillance requirement to verify the refueling water level during refueling operations. Therefore, a new surveillance is proposed for incorporation into TS 3.2, Table 3-5 consistent with NUREG-1432.

Soent Fuel Pool Water Level Currently, there is no surveillance requirement to verify the spent fuel pool water level during refueling operations. Therefore, a new surveillance is proposed for incorporation into TS 3.2, Table 3-5 consistent with NUREG-1432.

Spent Fuel Initial Enrichment /Burnuo Verification Currently, the requirement to conduct a verification of initial enrichment and burnup of spent fuel that will be stored in Region 2 is included as a general requirement of TS 2.8. It is proposed to relocate this requirement into a surveillance in TS 3.2, Table 3-5, consistent with NUREG-1432.

ADMINISTRATIVE CHANGES It is proposed to capitalize terms that are defined in the definition section of the TS to be consistent with Standard TS. It is proposed that sampling frequencies of items contained on page 3-19 be revised to incorporate frequencies defined in TS 3.0.2. Therefore frequencies stated as once per 31 days will be noted as "M," and frequencies stated as once per 7 days will be noted as "W." The Amendment numbers on pages 2-2c and 2-2d are being revised to state the pages were revised by "4/20/81 Order" versus 4/81 Order.

13

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION:

The proposed changes do not involve significant hazards consideration because operation of Fort Calhoun Station (FCS) Unit No.1 in accordance with these changes would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change will incorporate new requirements for the control room air conditioning system, control room filtration system, and refueling operations. In addition, the proposed change will ensure that the Limiting Condition for Operations and surveillance requirements are

_ consistent with the design basis of a fuel handling accident as documented in the FCS Updated Safety Analysis Report (USAR).

CONTROL ROOM SYSTEMS The control room air conditioning (A/C) system consists of two redundant A/C units, VA-46A and VA-46B. Each unit has sufficient capacity to meet the cooling requirements for personnel and equipment inside the control l room envelope. Each A/C unit is equipped with an air-cooled condenser ;

located inside a protective enclosure outdoors on the roof of the Auxiliary Building. Each A/C unit's refrigerant compressor, air cooling i coils, fans, and dampers are located inside of the control room envelope. Each unit has a waterside economizer coil that allows air cooling with Component Cooling Water (CCW). When cooling water temperature is sufficiently low, a temperature-activated valve at each A/C unit allows cooling water flow through the waterside economizer.

This valve also diverts flow away from the waterside economizer if cooling water temperature is too high. The air-operated CCW isolation valves to the A/C units fail closed and are automatically closed on a Ventili. tion Isolation Actuation Signal (VIAS) to prevent CCW flow '

through the waterside economizers in a post-accident situation.

Technical Specification (TS) 2.12(1) requires that the temperature within the control room and control cabinets be maintained below 120 F.

This value does not meet any of the four criteria contained in 10 CFR 50.36 for inclusion in TS. However, the equipment required to maintain this temperature, the control room air conditioning system, meets Criterion 3 of 10 CFR 50.36 in that the system functions to mitigate a design basis accident by maintaining the control room in a habitable environment.

Therefore, it is proposed that this TS be revised to delete the control room temperature as a LC0 and require that two control room air conditioning trains be operable when the reactor coolant temperature is above 210 F. The design temperature limits of instrumentation and controls inside of the control room will be maintained in the Basis Section of TS 2.12.

14

l BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

The allowed outage time for one train of control room air conditioning

.is proposed as 30 days. This is consistent with Combustion Engineering Standard TS 3.7.12 (NUREG-1432 Rev. 1). In addition, the FCS Probabilistic Risk Assessment model was reviewed and validated a 30 day outage time as being non-risk significant. The impact on Core Damage 3

t Frequency (CDF) from a 30 day LC0 was based on the assumption that one cooling unit was always inoperable and thus under the LC0 for an entire i year. This allows the analysis to consider unlimited entries into the LC0 and a full LC0 duration for each entry. Using this assumption, the j baseline (annually) CDF of 1.53E-5 would increase by 21.6% to a frequency of 1.86E-5. In accordance with EPRI's "PSA Applications

~

Guide," this small increase in CDF can be classified as "non-risk 1 significant."

Specification 2.12(2)

Specification 2.12(2) requires that a thermometer be in the control room

at all times. This instrumentation does not meet any of the four

criteria contained in 10 CFR 50.36 for inclusion in the FCS TS.

l Therefore, the requirement is proposed for relocation to the FCS USAR.

i Specification 2.12(3)

Specification 2.12(3) requires that all areas of the plant containing safety related instrumentation be observed during hot functional testing to determine local temperatures and monitored during operation if normal c plant ventilation is not available. It is proposed to delete this TS.

i The requirement to monitor and determine local temperatures during hot l functional testing was met during the initial startup phase of FCS and is no longer applicable. The requirement to monitor temperatures within the plant during normal operation does not meet any of the four criteria contained in 10 CFR 50.36 for inclusion in TS and therefore is being deleted.

! The requirement to control temperatures for safety related i instrumentation and controls, and initiate supplementary cooling if required, is currently described in USAR Section 9.10. These USAR

{ requirements are controlled by plant procedures. Any changes to these requirements would require an evaluation be conducted in accordance with g 10 CFR 50.59.

Soecification 2.12(4)

Specification 2.12(4) allows one control room air filtration system to be inoperable for 7 days or a plant shutdown be commencad. This 4

specification does not state which modes of operation it applies to.

i 4

15

- = . - - . . .. -. . - . . - - - . --. . .

1 1

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

Therefore, it is proposed to revise this specification to require two trains of control room air filtration systems to be operable when the reactor coolant temperature is above 210*F. The allowed outage time will be maintained at 7 days and a total of 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months will be allowed to i take the plant to cold shutdown. The 42 hour4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months time period is consistent i with TS 2.0.1 which addresses equipment outages in excess of what is specifically allowed by individual specifications.

' The proposed changes for the control room systems consist of providing additional restrictions on operation of the control room air filtration

systems and control room air conditioning system. These changes ensure that equipment required to mitigate the consequences of an accident are i

! operable. Therefore, the proposed changes do-not increase the

probability or consequences of an accident previously evaluated.

REFUELING OPERATIONS 1

The design bases of the fuel handling accident and refueling operations l were reviewed and several inadequacies were identified related to refueling operations. Therefore, revisions are proposed for the TS i Definition section, TS 2.6 on containment integrity, and TS 2.8 on l

refueling operations to reflect NUREG-1432. i

\

Definitions Cold Shutdown Condition & Refueh na Shutdown Condition The changes. proposed for the definitions of Cold Shutdown Condition, and Refueling Shutdown Condition clarify these definitions. The plant is in Cold Shutdown when T,g, is less than 210 F, and the reactor coolant is at least Shutdown Boron Concentration but less than Refueling Boron Concentration. Similarly, the definition for Refueling Shutdown is clarified to apply when T,g, is less than 210 F and the reactor coolant is at least Refueling Boron Concentration. This change does not propose ,

any new operating modes but merely clarifies when the definitions are 1 applicable. !

Core Alterations The definition for Core Alterations is being revised to reflect the requirements of NUREG-1432. This revision deletes "any component" from the definition and clarifies that the components considered by this definition are those that could affect reactivity. In addition, the revision' adds nuclear fuel to the definition such that movement of fuel within the reactor vessel will be defined as a core alteration and not a refueling operation.

16

i f

l BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued): <

Refuelina Operations

. The definition of Refueling Operations is being revised to delete i control element assemblies (CEA) or startup sources from the definition l since these are items that are included in the definition of Core Alterations. Additionally, it is being revised to specify that the '

definition'is limited to movement of irradiated fuel outside of the reactor pressure vessel since fuel movement inside the reactor vessel'is included.in the definition of Core Alteration. Finally, a clarification is being added to state that suspension of refueling operations shall not preclude completion of movement of irradiated fuel to a safe, conservative position.

In Operation The definition of In Operation is being revised to include the definition of operable. This is a more conservative interpretation than currently exists.

Specificat % 2.1 " Reactor Coolant System" It is proposed to revise TS 2.1.1(3) to include shutdown cooling requirements when the reactor coola.nt system (RCS) temperature is below 210*F with fuel in the reactor and the reactor vessel head fully tensioned. The definitions of Cold Shutdown (Mode 4) and Refueling Shutdown (Mode 5) contained in the TS make no distinction as to the status of the reactor vessel head or RCS temperature. The only difference between the two defined modes is boron concentration. Higher or lower boron concentration affects shutdown margin but does not affect decay heat load, which is the basis for this specification.

Technical Specification 2.1.1(4) was intended to address shutdown cooling requirements during refueling operations. However, this is already addressed in TS 2.8. Therefore, it is proposed to delete TS 2.1.1(4) and the exception since new specifications addressing shutdown cooling loop requirements during Mode 5 with fuel in the reactor and with one or more reactor vessel head closure bolts less than fully tensioned are proposed ~ for-inclusion in TS 2.8 (Refueling Operations).

The associated statements supporting these items in the Basis section are also proposed for deletion. Prior to any reactor vessel head closure bolts being loosened, TS 2.1.1 will be applicable which will require two shutdown cooling loops. As soon as a closure bolt is loosened, the new proposed TS 2.8 would be applicable which also requires two shutdown cooling loops whenever there is less than 23 feet of water above the core. The requirements of TS 2.1.1(3) are similar to NUREG-1432, Specifications 3.4.7 and 3.4.8.

17

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

Specification 2.6 " Containment System" Currently,'TS 2.6(1)c states that containment integrity shall not be

, violated when the reactor vessel head is removed if the boron concentration is less than refueling concentration. However, Specification 2.6(1)c has no required actions and therefore, TS 2.0.1 must be entered when the LC0 is not met. In this situation, (reactor vessel head removed), TS 2.0.1 is ineffective because the plant would

,. already be in Refueling Shutdown. Thus, TS 2.6(1)c is proposed for i deletion.

i i Currently, Specification 2.6(1)d requires that except for testing one control element drive mechanism at a time, positive reactivity changes shall not be made by CEA motion or boron dilution unless containment integrity is intact. Soecification 2.6(1)d is proposed for deletion as it is unnecessarily restrictive.

Specification 2.8.1(1) as proposed eliminates the need for containment integrity when the reactor is in Refueling Shutdown. Specification 2.8.1(1) requires sufficient shutdown margin to preclude a criticality event and also prescribes actions to restore the shutdown margin if necessary. Small positive reactivity increases whether by CEA motion or boron dilution will not cause a criticality event due to the need to maintain at least a 5% shutdown margin. Therefore, the requirement to maintain containment integrity is unnecessarily restrictive since a criticality event cannot occur when a shutdown margin of at least 5%

exists. Specification 2.8.1(1) is consistent with the requirements of NUREG-1432, Specification 3.9.1.

A new specification (TS 2.8.2(1)) is proposed that provides requirements for containment closure during core alterations and refueling operations inside of containment. The design basis of the Fort Calhoun Station does not require full containment integrity during a fuel handling accident. As stated in USAR Section 14.1? the fuel handling accident does not take credit for containment isolation. Therefore, requiring full containment integrity is inappropriate and requirements for containment closure are proposed for addition to TS 2.8 consistent with NUREG-1432 Specification 3.9.2.

Specification 2.10.2 governs operation of CEAs and monitoring of selected core parameters. . Specification 2.10.2 ensures (1) adequate shutdown margin following a reactor trip, (2) that the moderator temperature coefficient (MTC) is within the limits of the safety analysis, and (3) CEA operation is within the limits of the setpoint and safety analysis. Specification 2.10.2 ensures that the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality and provides actions (i.e., boration) to be taken to ensure that the required shutdown margin is available. Thus, TS 2.10.2 precludes the need for containment integrity when the plant is in cold shutdown.

18

BASIS FOR'NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

Specification 2.8 " Refueling Operations" It is proposed that TS 2.8 be rewritten to reflect NUREG-1432.

Currently, this specification applies to any refueling operation.

However, no distinction is made between refueling operations within containment and refueling operations within the spent fuel pool. In addition, several initial assumptions of a fuel handling accident are not addressed by the current TS 2.8.

Soecification 2.8(1)

The current TS 2.8(1) is inadequate. This specification requires that the equipment hatch and one door in the Personnel Air Lock be properly closed, and all automatic containment isolation valves be operable or at least one valve closed. The specification does not define what is meant by a properly closed equipment hatch; that information is currently contained in the Basis of TS 2.1.1. In addition, inclusion of all automatic containment isolation valves instead of those providing direct access to the outside atmospnere is incorrect.

The containment isolation system is defined in USAR Section 5.9.5 as those devices actuated by a Containment Isolation Actuation Signal (CIAS) or a Steam Generator Isolation Signal (SGIS). This includes many valves that have no design basis function during a fuel handling accident. A CIAS is initiated by a Containment Pressure High Signal or a Pressurizer Pressure Low Signal. Neither of these signals are I required to be_ operable during refueling operations as these signals I would/could not respond to a fuel handling accident.

The correct requirements are specified in TS 2.8(2) which only requires that closure be initiated by the Ventilation Isolation Actuation Signal (VIAS) for the containment pressure relief, air sample, and purge system valves. Due to these inadequacies, it is proposed to delete TS 2.8(1) and replace it with a new Specification 2.8.2(1) which is consistent with NUREG-1432 Specification 3.9.3.

Specification 2.8(2)

It is proposed that TS 2.8(2) be deleted and . replaced by new Specifications 2.8.2(3) and 2.8.3(5). The requirement to maintain an operable Ventilation Isolation Actuation Signal with input from the :

containment atmosphere gaseous and auxiliary building exhaust stack gaseous radiation monitors is consistent with current requirements and required actions are consistent with NUREG-1432, Specification 3.3.8.

Radiation Monitor RM-052 functions as a " swing" monitor, i.e., it can be aligned to monitor either containment or the auxiliary building exhaust ventilation stack. Radiation Monitor RM-052 is powered by either MCC-3B1/Al-40C (like RM-051) or MCC-4C2/AI-40D (like RM-062).

19

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

1 Technical Specification 2.7, Electrical System is not required to be I applied when the RCS is below 300 F. Above 300*F, TS 2.7 requires both 4160-VAC buses to be operable. Thus, above 300*F the required radiation monitors must be powered from independent 480-VAC buses supplied by independent 4160-VAC buses. During refueling outsges, bus alignments other than those used during power operation are used to permit i electrical system maintenance and modifications.

In the loss of offsite power event, the radiation monitor sample pumps

, and control room HVAC units stop and will not restart until the emergency diesel generators (EDGs) reenergize the system. The fuel handling equipment also stops and does not restart when the EDGs reenergize the system, thus minimizing the potential of a fuel handling l accident. When the EDGs reenergize the buses, VIAS will operate as designed. Therefore, when the RCS is below 300 F, the required monitors i

need only be powered from independent 480-VAC buses supplied by a single )

4160-VAC bus. l There is no need to assume that a fuel handling accident occurs ;

immediately followed by a loss of offsite power. However, in the I unlikely event that this should occur, there would be no effect on the site boundary dose since VIAS is not credited in USAR Section 14.18 )

(Fuel Handling Accident). In this situation, when the EDGs reenergize the buses, the control room HVAC units will restart in the filtered air makeup mode and the stack radiation monitor sample pump will restart.

However, the containment radiation monitor sample lines remain isolated preventing the restart of the monitor sample pump after receipt of a VIAS.

Soecification 2.8(3)

It is proposed that TS 2.8(3) be deleted. This requirement does not meet any of the four criteria contained in 10 CFR 50.36 for inclusion in the TS. The requirement that radiation levels in containment and the spent fuel pool shall be monitored during refueling operations will be incorporated into the FCS USAR.

Specification 2.8(6)

It is preposed that TS 2.8(6) be deleted. This requirement does not meet any of the four criteria contained in 10 CFR 50.36 for inclusion in the TS. The requirements that direct communication between personnel in the control room and at the refueling machine shall be available whenever core alterations are taking place will be incorporated into the FCS USAR.

j 20

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

-Soecification 2.8(7)

It is proposed that TS 2.8(7) be deleted and replaced with a new Specification 2.8.3(4). The requirement to place the spent fuel pool ventilation system in operation prior to refueling operations is consistent with the current TS. It is being clarified that this specification only applies to refueling operations in the spent fuel I pool, and not when conducting refueling operations inside of containment. Additionally, it is being clarified that TS 2.0.1 is not applicable to this activity, as reactor operation is independent of fuel movements in the spent f al pool.

Specification 2 N91 l The current Specif4 cation 2.8(9) is inadequate. This specification requires a minimum of 23 feet of water above the top of the core. This {

does not meet the initial conditions assumed in the fuel handling i accident as documented in USAR Section 14.18. USAR Section 14.18 l assumes 23 feet of water above where the fuel could land if dropped. In order to meet this initial condition, a minimum of 23 feet of water above the reactor vessel flange is required, as this is the highest point where a fuel bundle could land if dropped. Procedures reflect the requirement to maintain 23 feet of water above the reactor vessel flange during refueling operations. The proposed revision is consistent with NUREG-1432, Specification 3.7.16.

Specification 2.8(11)

The current specification is inadequate. The specification provides restrictions on storage of fuel in the spent fuel pool; however. there 1 are no required actions to address situations when the specification is not met. It is proposed that TS 2.8(11) be deleted and replaced with a new Specification 2.8.3(1) that requires that a misloaded fuel assembly ,

be moved immediately. Additionally, it is being clarified that TS 2.0.1 i

is not applicable to this activity, as reactor operation is independent of fuel movements in the spent fuel pool.

Soecification 2.8(12) l It is proposed that TS 2.8(12) be deleted and replaced with a new !

Specification 2.8.3(3). The requirement to maintain 500 ppm boron J concentration in the spent fuel pool whenever unirradiated fuel is stored there is consistent with the current TS and the required actions are consistent with NUREG-1432, Specification 3.7.17.

21

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

Restriction on Movement of Irradiated Fuel from the Reactor Core The restriction on irradiated fuel movement unless the core has been subcritical for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if the reactor has been operated at power levels above 2% is proposed for relocation to the Bases of TS 2.8.2(2). This requirement does not meet any of the four criteria contained in 10 CFR 50.36 for inclusion in the TS. This is consistent with NUREG-1432, B 3.9.6.

Reactor Coolant System Boron Concentration Currently, there is no specification for boron concentration. Refueling boron concentration is included in the definition of Mode 5. However, there are no required actions to be taken if the boron concentration should be below refueling concentration. Therefore, it is proposed that a new Specification 2.8.1(1) be incorporated consistent with NUREG-1432, Specification 3.9.1.

Soent Fuel Pool Water Level Currently, there is no specification for spent fuel pool water level.

The water level of the spent fuel pool is an initial condition assumed in USAR Section 14.18. It is proposed that a new Specification 2.8.3(2) be incorporated into TS 2.8, which is consistent with NUREG-1432, Specification 3.7.16.

The proposed changes for the RCS and containment during shutdown, and requirements for refueling operations, consist of providing additional restrictions on operation, and changes to make the requirements of the TS Limiting Conditions for Operation consistent with the initial conditions and assumptions of the fuel handling accident as documented in USAR Section 14.18. Therefore, the proposed changes do not increase the probability or consequences of an accident previously. evaluated.

SURVEILLANCE REQUIREMENTS CONTROL ROOM Specification 3.1. Table 3-3. Item 13.

Specification 3.1, Table 3-3, Item 13 requires that the thermometer in the control room be compared with a calibrated thermometer and replaced if out of tolerance on a refueling frequency. It is proposed that this surveillance be deleted to be consistent with deletion of the LC0 requirement to maintain a thermometer in the control room.

A new surveillance is proposed to verify that the control room air conditioning system has the capability to remove the assumed heat load.

This surveillance will ensure the operability requirements for TS 2.12 are met. The test and frequency is consistent with NUREG-1432.

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l l

. l BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

i

! The air-operated CCW isolation valves to the A/C units fail closed and l are automatically closed on a VIAS to prevent CCW flow through the I waterside economizers in a post-accident situation. These valves are currently tested in accordance with TS 3.3 (FCS Inservice Testing i Program). Prior to the modification, the valves were tested as fail-open valves.

No TS changes are necessary.

i The control room air filtration system is currently tested on a refueling frequency in accordance with TS 3.2, Table 3-5, Item 10a. No TS changes are necessary.

l REFUELING OPERATIONS Reactor Coolant Boron Concentration Durina Refuelina Operations The Reactor Coolant System boron concentration is currently sampled in 1 accordance with TS 3.2, Table 3-4, Item 1(e). It is proposed to revise the frequency from once per shift during refueling operations to once j l par 3 days which is consistent with NUREG-1432. As stated in the basis :

of TS 2.8 and USAR Section 14.18, the reactur cavity is filled with over j

200,000 gallons of borated water prior to the start of refueling '

operations. The requirements for sampling the reactor coolant during the remainder of Mode 5 is performed once per 3 days in accordance with Table 3-4, Item 1(d). This proposed change will make the sampling I consistent with the requirements of Item 1(d) and NUREG-1432. l Spent Fuel Pool Boron Concentration The spent fuel pool boron concentration is currently sampled in accordance with TS'3.2, Table 3-4, Item 5. It is proposed to revise the frequency of the sampling to prior to movement of unirradiated fuel in

the spent fuel pool and once per week whenever unirradiated fuel is stored there to be consistent with the requirements of the LCO.

Source Ranoe Neg ron Monitors Currently, a channel check and calibration of the wide range neutron monitors is performed in accordance with TS 3.1, Table 3-1, Item 2.

Containment Penetrations Currently, there is no surveillance to determine the status of containment penetrations during refueling operations. Therefore, a new surveillance is proposed for TS 3.2, Table 3-5 to verify the status of required containment penetrations once per 7 days consistent with NUREG-1432.

23 m _ -.

i 1

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

The requirement of NUREG-1432 to verify that the containment purge and exhaust valves actuate to the isolation position on a refueling frequency is currently tested as part of the Containment Radiation High Signal test required by TS 3.1, Table 3-2. Item 4.

Shutdown Coolina Loops 1' '

Currently, there is no surveillance requirement to verify that the required shutdown cooling loops are operable and in operation or to verify correct breaker lineup for the shutdown cooling loop that is not in operation. Therefore a new surveillance is proposed to be incorporated into TS 3.2, Table 3-5 consistant with NUREG-1432.

Refuelina Water Level Currently, there is no surveillance requirement to verify the refueling water level during refueling operations. Therefore, a new surveillance is proposed for incorporation into TS 3.2, Table 3-5 consistent with NUREG-1432.

Spent Fuel Pool Water Level Currently, there is no surveillance requirement to verify the spent fuel pool water'1evel during refueling operations. Therefore, a new surveillance is proposed for incorporation into TS 3.2, Table 3-5 consistent with NUREG-1432.

Spent Fuel Initial Enrichment /Burnup Verification ,

1 Currently, the requirement to conduct a verification of initial enrichment and burnup of spent fuel that will be stored in Region 2 is included as a general requirement of TS 2.8. It is proposed to relocate :

this requirement into a surveillance in TS 3.2, Table 3-5, consistent 3 with NUREG-1432. -l The proposed changes for the surveillance requirements consist of providing additional testing requirements to ensure that the Limiting Condition for Operations will be met. One surveillance frequency related to the sampling of the reactor coolant system boron concentration during refueling operations is being reduced from a frequency of once per shift to once every 3 days. However, this frec 7ency is consistent with the frequency of sampling during the remainder of Mode 5 when fuel is in the reactor and is more than adequate due to the large volume (over 200,000 gallons) of borated water required during refueling operations. Therefore, the proposed changes do not increase the probability or consequences of an accident previously evaluated.

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BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

ADMINISTRATIVE CHANGES The remainder of TS 2.8 requirements of refueling operations are proposed to be reformatted into individual TS LCOs. It is also proposed )

that sampling frequencies of items contained in TS 3.2, Table 3-4, (page 3-19), be revised to incorporate frequencies defined in TS 3.0.2.

Therefore, frequencies stated as once per 31 days will be noted as "M,"

and frequencies stated as once per 7 days will be noted as "W." These proposed changes have no effect on the probability or consequences of an ;

accident previously evaluated.

(2) Create the possibility of a new or different kind of accident from any accident previously evaluated. l l

There will be no physical alterations to the plant configuration. No I changes in operating modes are proposed although minor changes to the l definitions of Cold Shutdown Condition and Refueling Shutdown Condition !

are proposed for clarification purposes. The proposed changes incorporate additional restrictions on the operation and testing of i equipment required to mitigate an accident and to ensure the initial conditions and assumptions of the design basis accidents are maintained and controlled by the Technical Specifications.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

(3) Involve a significant reduction in a margin of safety.

The proposed changes ensure that assumptions of the fuel handling accident are maintained by Technical Specification Limiting Condition for Operation and surveillance requirements. The assumptions of the fuel handling accident that may affect a margin of safety are not being changed. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Therefore, based on the above considerations, it is OPPD's position that this proposed amendment does not involve significant hazards considerations as defined by 10 CFR 50.92 and the proposed changes will not result in a condition which significantly alters the impact of the Station on the environment. Thus, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and pursuant to 10 CFR 51.22(b) no environmental assessment need be prepared.

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ML20137Y180

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