Text

Proposed Model Tech Specs for Installation of Permanent Scram Discharge Sys Mods
	ML17054D462
Person / Time
Site:	Dresden
Issue date:	06/24/1983
From:	; NRC
To:	;
Shared Package
ML17054D463	List: ML17054D462; ML20072G625;
References
	NUDOCS 8306280575
	Download: ML17054D462 (60)
	v • d • e

Enclosure.

2.

SAFETY LIMITS ANO LIMITING'AFETYSYSTEM'ETTINGS 2'.2.

LDlITING:SAFETY SYSTEM'ETTINGS REACTOR PROTECTION SYSTEM'INSTRUMENTATION SETPGINTS 2.2' The reactor-protection system instrumentation setpoints shall be set consisten< with. the: Trip. Setpoint. val'ues shown. i'n Table 2.2. 1"l.

APPLICABILITY:

As shown in; Table 3.3;1-1.

ACTION=

'Hith a= reactor protection'ystem instrumentati'on setjoint less'onservative than the. value-shown. in the All'owable VaTues" column-of Table 2'.2.1-1, declare.

the channel inoperabl'e and: apply the app1icab1'e.'ACTION statement requirement of'pecification 3.3.1 unti'1'he channel is restored: to OPERABLE status. with its setpoint, adj'usted consistent, with: the: Trip Setpoint. value.

GE"STS (BMR/4) 2;3

l

TRIP SETPOINT

< (lpga)/(12$ ) (jjyjsjpns of flit] sc(j]e

< (~S)g og RqgP Tl!ERMA'. Io~glj C)

IAOLE 2.2: l-l REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS l/l

~'

FUNCTIONAL UNIT

],

Iptermedjale Range poilipor, Neutron I-lug-IfigI1 2.

pverage I'owel Raggp lloni$or:

a.

Neutrons f:'faux-Upscale, Setdoqp ALLOQAOL)

VALUES

< (122)/(12$ ) sl)yisions!

of fqjl scale (20)X of RATFD THEPHAL POWER b.

C.

Flow Biase(j Sjitl(llate(j TIleimal'oWer=Upscale ~

1) l'low Biase(j

2) l{igl> Ploy P]apped Fixed Neutl'oq I'1 qx-Upscale (e.

inoperative Down.",r.a l p 7.

9.

10.

11.

12.

(Primary Containment)

(Drywell) Pi'essui'e - ijigIl Scram Oi;.cl);lr(l( polumrt Water Leyel

"- lltgh Turbjne Stop Valve - Closure Turbine Control Valve Fast C]osui e, Trip Oil Pressule

- I.ow Reactor Node Sqitch SI)qtdown Posjtion Hatiual Scram "See Bases Fjgure 0 3/4 3-1.

3.

Reactor Vnssst'l Sleam Oo!!Ie Piessqre

-. Iiiqh.

Reecton

!ie.sn)

Wpten Leyel -

LoW< Lpyel 3

5.

Main Steam l ine Isolation I)talye = C]osui'e 6.

flain Steam Line Radjatjon -

lIigh 0:g Wt(51)$; qjtI) a ma)ljmim oj'

($/3,5)X of RATPP T(EQUAL PO/Eg

~

< (118)X zf RATED Tll)RHAL POWER NA

> (5)X of lIhlI:.P T)ICRMAL t>0+it

< (10/I5) psiq

> (12.5) jDcIles al)yves inst.l'ljmeqt zei a"

< (PX closed

< (2.5) x fpll pqwer background

< (1,69) psjg

< (36)X of ful'I. peal(.

< (5)X cl'osed

> (500) psig N4 NA' 0.66 W+(54)X, q>

a it)aximum of (115;5)X of PATED TliERJAL POWER

< ()20)X of RATEO fllgPH/L I'OWER f

NA (3)X of RATED TIiERMAL POWER)

< (1065) psig

> (11.0) inches aboye instrumer>f zero

< (7%. closed

((3.0) ful'I P WP haci(ground

< (1,89) psjg I

< (39)X of fu)) scale

< (7)X closed

>(

) psig NA

I

LIMITING SAFETY SYSTEM SETTING BASES REACTOR PROTECTION SYSTEM'NSTRUMENTATION-SETPOINTS (Continued) 8 Scram Oischar e Volume-Mater Level-Hi h The. scram. discharge: volume receives-the water displaced by the motion of the control rod. drive pistons during, a. reactor scram.

Should this volume fill up to a point where there is insufficient volume to accept the displaced water at. pressures below 65 psig,, control rod" insertion would be.- hindered.

The. reac-tor is'herefore tripped when the water Tevel, has reached.

a point high enough.

to: fndicate. that it: is. indeed'. filling,up, but. the" volume is; still great enough to accommodate the water from the movement of the rods at. pressures below 65 ps'hen they are tripped; The; trip setpoint. for each scram discharge, volume is equivalent: to. a. contained: volume of (

) gallons of water.

9.

Turbine. Sto Valve.-Closure The-turbine stop; viva closure trip anticipates the. pressure,.

neutron flux, and" heat flux increases that would. result. from closure. of'he stop vaTves Mith a; trip: setting, of'5)X oP valve: c1osure. from full open, the resultant i'ncrease. fn heat'Tux. is'uch. that, adequate.

thermal margins. are maintained during the worst." case transient- (assuming the turbine bypass valves

'fail to) operate),

TO.. "Turbine: Control VaTve Fast. CTosure Tri OiT Pressure-Low The turbine control valve. fast closure. trip'nticipates the pressure, neut'ron flux, and heat, flux increas'e that'ould, result. from fast. closure of.

the: turbine-controT valves: due to. load; rejection coincident with failure of the turbine. bypass; valves.

The Reactor Protection System initiates a trip when fast cTosure. of the. control valves is initiated by the: fast. acting sole-noid valves-and: in less. than (30) miT1 seconds after the start of control valve fast closure.

This fs. achieved; by the action of the fast. acting solenoid valves in rapidly= reducinghydraulic trip oil pressure at the main turbine control val've: actuator, disc. dump. valves.

This loss of'pressure is sensed by pressure switches whose contacts form the: one-out.-of-two-twice logi'c input to the Reactor

-. Protec ion System.

This. trip setting,. a faster closure time, and a different valve characteristic.

from that of the: turbine. stop val.ve, combine to produce transients which are very similar to that for the stop valve=

Relevant tran-sient. analyses are-discussed in Section (15. 1'. 0) of the Final Safety Analysis Report 11.

Reactor Mode Switch Shutdown Position The reactor mode switch Shutdown position is a redundant channel to the automatic prot'ective: instrumentation channels and provides additional manual reactor trip capability.

12.

Manu'al Scram The Manual Scram is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

GE.-STS (SMR/4)

B 2.-9

ib ~

REACTIVITY CONTROL SYSTEMS 3/4. 1'. 3 CONTROL RODS CONTROL ROD OPERABILITY LIMITING.CONDITION FOR OPERATION 3..1:.3 1

A1T control 'rods. shall be OPERAB)E.

APPLICABILITY OPERATIONAL CONDITIONS' and':

ACTION:

r a

With one control rod: inoperable due to being; immovable, as

a. result of' excessive. friction or mechanical'nterfe'rence;.

or known to be untrippable.

l Within one hour:

b.

a).

Verify that'the inoperable control rod, if'ithdrawn; is.

separated.

from. all dther inoperable control rods. by at. least two control'el'ls i'n all directions.

b)'isarm. the: associated directionaT control valves"" either" 1)

Electrical lyoi-2)'ydraulica11y by closing, the drive. watewand. exhaust.

water iso Tatiorr valves c)

Comp Ty wi'th SurveiTl'ance. Requirement: 4;I.Lc; Otherwise',.

be: in at, least HOT SHUTDOWN Qithin the next 12'ours; 2

Restore: the inoperable control rod. to OPERABLE status. within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or.

be; in-; at least;.HOT'HUTDOWN withi'n the-next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months With one. or more control rods trippable-but. inoperable for causes other than addressed in ACTION a,, above:.

If the inoperable control rod(s) is withdrawn, within one hour:

a)'erify that the. inoperable withdrawn control rod(s) is sep'arated from al'1 other inoperable. control rods. by at least, two control

cells in. aTl'irections and.

b)

Demonstrate the insertion. capability of the inoperable withdrawn control rod(s) by inserting the control rod(s) at least. one notch by drive water pressure. within the normal operating range".

Otherwise, insert the inoperable withdrawn control rod(s) and disarm the. associated. directional control valves."" either:

a)

Electri'cally, or b)

. Hydrau'lically by closing the drive, water and exhaust. water isolation. valves'.

"The inoperable control rod may then be withdrawn to a position no further withdrawn than its. position when found to be inoperable.

""t1ay be rearmed iptermittently, under administrative control, to permit.

testing associated with restoring the: control rod to OPERABLE status.

GE-STS (BMR/4) 3/4. I.-3

I I

~ REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued ACTION (Continued) 2'.

If the: inoperable control rod(s) is inserted', within one hour disarm the'ssociated.

directional control valves"" ei.ther:

a)

Electrically, or s

b)

HydraulicalTy by closing, the drive water and exhaust water i'sol ation val ves.

Otherwise; be in, a< least HOT SHUTOQMN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

C.

Mfth more-than 8 controT rods: i'noperable,,

be. in at least HOT SHUTOOMN within 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s SURVEILLANCE RE UIREMENTS t

4.1.3".1 1 The scram discharge.

valume. drain and vent. valves. sha.ll be.

demonstrated:

OPERABLE by.

a AC Toast once. par Sl. days, ver ifyi'ng each valve to be. open,* snd

, b:. At least. once. per 9Z days'ycTing each,val've through at least. one complete-cycle.: of ful.l: travel,.

4'T.KI Z Nhen above: the (preset power level) (low power setpoint) of the RNN and RSCS', all'ithdrawn control, rods not. required to have their directional control valves: chsarmed" eTectrically or hydraul.ically shall be demonstrated.

OPERABLE by moving each control rod at, least one notch:

a At, least once per 7 days, and.

b.

. At least once-per 24'ours'hen any control rod; is immovable as a

result of excessive friction or mechanical interference.

4.1.3,1,3 Al'1'ontrol rods: shall be. demonstrated.

OPERABLE by performance of S'urveillance Requirements'. 4;1.3.,2, 4..1 3;.4', 4. 1.3.5, 4.. 1.3.6 and 4 1.3.7.

"These valves may be closed. intermittently for testing, under administrative control s'.

- May be rearmed intermittently, under administrative control, to permit.

testing associated with restoring the control rod to OPERABLE status.

GE-STS (BMR/4) 3/4 1-4.

\\ ~

Is m

I REACTIVITY CONTROL SYSTEMS.

SURVEILLANCE'E UIREMENTS.

Conti'nued 4.1 3'.l.4'he scram dischar ge volume. shall'e determined. OPERABLE'y demonstrating:.

a.

The scram. discharge-vol'ume dragon-and vent valves OPERABLE, when control rods are scram tested; from a normal control'od configura-tion: of. less'han or. equal to (50)X'.ROO OENSITY at least once per 18-'- months,, by: verifying, that. the drain. and-vent: valves L,

Close withi'n (30)'econds. after receipt of a signal for control.

rods. to. scram,.'and 2..

'pen; when: the-scram signal: is'eset; I

b.

Proper (fToat)'Tevef sensor);

response: by performance of a CHANNEL FUNCTIONAL TESTl'f,"the. scram, discharge volume scram, and control rod'ock level instrumentation; (LT 1'evel measuring-system) (after each

. scram from. a pressurized conditi'on). (at. least. once" per 31 days).

/

GE-.STS (BMR/4) 3/4" 1-5'

3 3

REACTIVITY CONTROL SYSTEMS ROO BLOCK MONITOR LIMITING,CONOITION FOR OPERATION 3'.1 4'=3'.

Both rod. bl'ock. monitor (RBM) channels shall be: OPERABLE.

APPLICABILITY:

OPERATIONAL CONOITION,1, when: THERMAL POMER is greater than

,Rqq'33)EH,T HATER THERHAE PEIIER.

ACTION."

a..

With one RBM'hannel inoperable, restore.. the inoperable RBM channel to. OPERABLE status. within. 24'ours'nd'erify that, the. reactor. is not operating. on a. LIMITING:CONTROL ROD PATTERN; otherwise, place the fnoperble. rod.block monitor channel in the: tripped'ondition.

within the next hour A

I b

Mith both. RBM'channels. inoperable,. place-at least one inoperable rod bl;ock. monitor channel in. the, tripped; condition'i'thin one hour.

SURVEILLANCE RE UIREMENTS'-

4..1.4.3 Each:, of the. above. requiretL RBM channels. shall'e demonstrated OPERABLE by performance. of. a=

CHANNEL FUNCTIONAL TEST'nd'HANNEL CALIBRATION at the frequencies and.for the OPERATIONAL CONOITIONS specified in Table 4.3'.6-1.

CHANNEL FUNCTIONALTEST prior*to control rod withdrawal when the reactor is-operating on a~ LIMITING:=CONTROL. ROO PATTERN; GE'-STS (BMR/4)

- 3/4'-18,

REACTIYITY CONTROL SYSTEMS BASES 3/4a l. 3 CONTROL RODS" The specification of'his section ensure that. (1) the minimum SHUTDOWN MARGIN is. maintained '2) the control: rod insertion times: are consistent with those: used, in. the. accident analysis,,

and; (3) limit the: potential effects of the rod drop accident.

The. ACTION statements:

permit. variations from the basic.

requirements. but at'he.

same time impose. more restrictive criteria for continued.

operation A 1'imitation on inoperable. rods is'et. such that the resultant effect on total rod. worth and: scram shape wil'1'e kept to a minimum The requirements for the various. scrams time measurements'nsure that any indication of systematic.

prob'lems with rod. drives. will.be: investigated on a timely basis Damage within the control,rod drive mechanism. coul'd. be a: generic problem,,

therefore-with a. controT rod; immovable because of. excessive friction or.

mechanical interference. operation of'he reactor is limited to a. time period which is reasonabl'e-to determine. the cause of the inoperability. and at, the same, time. prevent operation with a large number of inoperable control rods.

'ontrol. rods that. are inoperable for other reasons are permitted to be taken out of'ervice.- provided. that those. in the nonfully.-inserted position. are consistent with the. SHUTDOWN MARGINS requirements; The number of'ontrol rods permitted; to be inoperabTe: could be more: than the eight allowed: by the.. specification, but. the occurrence. of eight inoperable rods.: could. be: indicative: of a, generic problem and. the: reactor must be shutdown for investigation and resolution of. the probl'em.

The control rod. system is designed.'o. bring the: reactor subcritical at a rate. fast. enough to prevent the MCPR from becoming less than (1.06) during the limiting power transient analyzed, in Section (15.

) of the, FSAR.

This analysis shows that the negative reactivity rates resulting from. the scram with the average: response of all the drives as. given in the-specifications, provide the required protection and MCPR'emains greater than (1.06).

The occurrence of scram times longer then. those sperified'. should'e viewed as an. indication of a systemic problem with the rod drives and therefore the-hurveillance interval is. reduced i'n order to. prevent. operation of the reactor for long periods of time with a. potentialTy serious problem The scram discharge-volume. is required to be OPERABLE so that it will be available: when needed to accept discharge, water from the control rods during a reactor scram and: will'solate the" reactor coolant system from the containment when required; Control rods with inoperable accumulators are declared inoperable and Specification 3; 1.3. 1'hen applies This prevents a pattern of inoperable accumulators. that. would result, in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive. water pressure.

Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.

GE-STS (BWR/4)

B 3/4 1-2

~

REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued)

ControT rod: coupling-integrity is. required to ensure compliance with the analysis of the'rod=drop accident. in the FSAR The overtravel position feature provides-the only positive means of determining. that a rod is properly coupled and. therefore'this check must. be-. performed. prior-to achieving criticality after compTeting CORE ALTERATIONS that could, have: affected the control rod coupling.

i'ntegrity.

The subsequent check, is performed as a backup to the initial demon-stration..

In. order te ensurt that the.-control'od patterns.

can be followed and there-fore:- that other parameters.

are. within their-limits, the. control rod position indication system must be. OPERABLE.

The-control rod'ousing. support restricts the. outward movement of a control rod'. t0, less than. (3) inches fn the event. of a. housing failure-.

The amount of rod. reactivity which could. be added by this. smaIl amount of rod withdrawal is.

less than' normal'ithdrawal increment and wiIT not. contribute. to any damage to. the. primary cooTant: system.

The support= is not required when there is no pressure; to act: as': driving force. to rapidTy. eject a drive housing.

The, required. surveillance intervals're. adequate to determine that: the.

rods are OPERABLE and: not= so frequent; as: to. cause excessive. wear on.-the-system<<,.

components.

3'/4; T.4-CONTROL ROD PROGRAM CONTROLS.

Control rod withdrawal'nce insertion sequences are established, ta assure.

that: the. maximum insequence individual'ontrol rod or control, rod. segments, which are withdrawn at. any time. during the fueT cycTe coul'd not'e worth enough to result; in a peak fuel enthalpy greater than 28Q cal/gm in the event of a control rod. drop accident The specified: sequences, are characterized by homogeneous,,

scattered patterns. of control rod. withdrawal.

When THERMAL POWER is. greater than (20)~'f RATED, THERMAL POWER, there is no possible rod. worth which, if.

dropped, at. the. design rate.- of the veTocity limiter, could result in a peak enthalpy of 280 cal/gm; Thus'equiring, the RSCS'nd RWM to be OPERABLE when THERMAL POWER's less than or equal to (20)X: of RATED THERMAL POWER'rovides adequate control.

The. RSCS and. RWM'provide automatic supervision to assure. that out-of-sequence.

rods will not: be withdrawn or inserted II The analysis of. the rod, drop accident is presented in Section (15.

) of the FSAR and. the. techniques of the. analysis are-presented in a topical report, Reference I, and two-supplements,.

References

2. and 3.

The RBM is designed. to automatically prevent fuel damage in the event of erroneous rod, withdrawal from locations of high power density during high power operation.

Two channels are<< provided.

Tripping one-of the channels wi 11 block erroneous rod withdrawal soon enough to. prevent uel damage.

This system backs up the written sequence used by the operator for withdrawal of control rods.

GE-STS (BWR/4)

B'/4 1"3

3/4. 3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM'NSTRUMENTATION LIMITING CONOITION* FOR OPERATION 3.3.T As a. minimum, the. reactor protection system instrumentation channels shown in. Table. 3.3 l-l shall be OPERABLE with the REACTOR. PROTECTION SYSTEM RESPONSE TIME as. shown in Table 3.3; 1-Z; APPLICABILITY'.

As shown in Table '3'3.1-1 ACTION:

a.

with the-number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip, System requirement for one'rip system, place the: inoperable channel(s) and/or that trip system in the tripped condi-tion~ within one. hour The: provisions of Specification 3.0.4 are not.

app Ti'cab 1 e b

Mith the number of'OPERABLE channels 1'ess than required. by the Minimum OPERABLE ChanneTs per Trip System, requirement for both trip systems.,

place at: least. one trip. system"~ in: the. tripped: condition within one hour

and,

~

take the: ACTION'equired by Table 3.3.1-1.

SURVEILLANCE RE UIREMENTS 4.3.1 1'ach reactor protection system instrumentation. channel shall be.

demonstrated" OPERABLE by the performance. of'he. CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations. for the. OPERATIONAL CONDITIONS and at the. frequencies.

shown in Table 4;3.1.1-1.

4.3.1.2.

LOGIC. SYSTEM FUNCTIONAL TESTS'nd simul'ated automatic operation of all channels shall'e.- performed: at. least. once per

18. months.

4.3.1.3 The. REACTOR'ROTECTION SYSTEM'ESPONSE,'IME of each reactor trip functional unit'. shown in Table. 3'.3.1-2. shall be demonstrated to be-within its limit at least. once per 18 months.

Each test shall include at least. one channel per trip system such that'lT channels are tested at least onc'e every N times 18 months where N is the" total number of redundant. channels in a specific. reactor trip system; "An inoperable. channel need not'e placed i'n the tripped condition where this would cause the Trip Function to occur.

In these

cases, the inoperable channel shall be restored to OPERAOLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACT!..'l required by Table 3.3.1-1 for that Trip Function shall be taken.

""Ifmore channels are inoperable in one. trip system than-in the o her, place the-trip system with more inoperable channels in the tripped condition, except when this would cause the Trip Function to occur.

GE-STS (BWR/4) 3/4 3-1

'0

Cl fi1 I-I Vl TABLE 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION

( l7 FUNCTIONAL UNIT l.

Iqterqipdjate Pgllgp fjotljtof'g a.

(eqttop F]ux = )jjgt>

b.

I<~operatjve b.

C.

ploi] Bjased Sjpq)ate) Tllerqa]

Power Upscale Fixed Neutron F1ux Upsca)y Inoperative (e.

Poqnscale 2.

Ayerage Power Rqpge Mopjfop-a.

neutron Fluq

.- Upscale>

Setdown APP )lC/ABLE OPERATIONAL CONDITIONS 2

~~ ~~(c) 2 3

lf 5

2 g(g) 1 1] 2

~(c) 1(g)

MlNIMUf'j OPERABLE CIIANNELS PER TRIP SYSTEM a

3 3(d) 3

. 3(d) 2 2(0) 2 2

2(d)

ACTION 1

2 3

1 2

3 1

2 3

1 2')t 3.

Reactor Vessel Steam Oome Pressure - ttjgtl Reactor Vessel pater Level - Low; Level 3

ttain Steam Line Isolation Valve-Closure 1,

2

>(g)

0

TABLE 3.3.1-1 (Coqtinued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION

. FUNCTIONAL UNIT 6.

Majq Steam Line Radjatioq

=

Iiiqg 7.

(Prjpary Coqtajnyept.) (Or~q)])

pt'essjt'e

-. gjgh 8.

Scrag Discharge Po)mme

'gatet'eyel

= gigI1 9.

Turgine Stop Va)ve - Closut'e

)0.

Turbine Contlol Yar)ve Fast C)osprey, Va)ye Trip System OiliI'res~pry T. Lqg

)1.

Reactor-Mode Syitch Shutdown posjtiop

) 2.

Manual Scram API'LlC(B! E OPEgATlggAI-CONDITIONS

)

<(r) 2(h) 2~i ~

)(j)

)((q))(i) l, 2

3) g 5

1, 2, 3

5 MINION OI!ERQBLP CQ(tiPLS PER TRIP SYSTEM a'

2

~(k) 2(k) 2 2

2 ACTMN 5

)

3

~ri

~ r II~

A ~

~ ~

~

. TABLE 3.3.1'-I (Continued)

REACTORPROTECTION'YSTEM INSTRUMENTATION ACTION ACTION 1 ACTION 2'CTION:

3'CTION 4'CTION5 ACTION 6'-

ACTION 7 ACTION: 8i ACTION 9 E

Be in at. least HOT SHUTDOWN'ithin 12. hours..

Yerify all'nsertable. control'ods toi be inserted. in the core-and.'Tock the reactor mode switch in the Shutdown position within. one. hour Suspend'll. operations'nvolving CORE ALTERAT/ONS~. and insert.

all insertable control rods within one hour,.

Be: in at least STARTUP wi'thin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Be; in. STARTUP" withe the. main steam. 1'ine isolation. valves closed within: 6: hours.. or in at. least HOT SHUTDOWN-within'2, hours Initiate a: reduction i'- THERMAL POWER'ithin-15. minutes and reduce turbine. ffrst; stage pressure to < (250) psig, equivaTent to. THERMAL POWER" Tess than (30)X of'ATED. THERMAL POWER', withi'n 2". hours; Yerify.: al.l fnsertable: control. rods'o be inserted. within one hour Lock, the reactor.

mode switch in.the Shutdown position within one hour,.

N

.Suspend. all',operations involvi'ng CORE ALTERATIONS", and insert alT insertable control'ods. and.lock the. reactor mode switch in the" SHUTDOWN'osition within one hour.

"Except'ovement. of IRM, SRM" or special'movable: detectors, or replacement of LPRM strings provided. SRM'nstrumen ation is OPERABL per Specification 3.9.2.

GE"STS (BWR/4) 3/4: 3-4

I

TABLE. 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM'NSTRUMENTATION.

, TABLE NOTATIONS (a)

A channel'ay be" placed'n an inoperable status for up to 2'ours for required: surveill'ance without pl'acing; the trip system in the tripped condition. provided't: least. one OPERABLE channel in the same trip system i's monitoring that parameter.".

(b)

This'unction'hall be automatical,ly bypassed when the reactor. mode switch is in the Runi position (c)

The "shorti'ng; links-" shall'e.

removed f'rom-the RPS circuitry prior to.

and. during.. the-ti'me. any control rod. is withdrawn" and; shutdown margin demonstrati'ons performed per Specification 3 10..3.

1 (d)

The non-.coincident.

NMS-reactor trip, function-logic is. such. that all'hannels.

go to. both: trip-systems'herefore,.

when: the "shorting links." are. removed,

'the Minimum;- OPERABLE Channels Per." Trip-System. is. O'PRMS'nd 6

IRMS'.

(e)

An; APRM'. channels.fs-inoperable.- if there. are.. Tess. than 2.'PRM'nputs; per 1'evel'r.

1'ess, than'(11)y LPRM~ inputs. to. an APRM'hannel.,

(f)

This;.function. fs'ot 'requi'red. to be; OPERABLE when the reactor.,pressure vessel'.

head. i's unbolted'r removed per Speci.fi'cation 3; 10. 1 (g)

This function shal,l be. automatically, bypassed when the reactor mode switch is: not'n" the. Rurr~ position.

(h).

This function. is'ot: required" to: be OPERABLE'hen 'PRIMARY CONTAINMENT INTEGRITY's: no< requi'red.

(i)

With any controT rod withdrawn.

Not appTicabTe to control rods. removed per Specification 3.9" 10.,1: or 3;9'.10.2:

(j)

This. function'hall'e. automati'cally bypassed: when turbine first stage pressure is

< (250) psi'g; equivaTent: to THERMAL POWER less than (30)X.

of'ATED.'HERMALPOWER

'k)

Also. actuates: the. EOC'-,RPT'ystem; "Not required for control'ods removed per Specification 3.9.10.1 or 3.9.10.2.

GE"STS (BWR/4)

~

3/4 3-5

\\

0

TABLE 3.3.1=2 REACTOR PROTECTION SYSTEN

RESPONSE

TINES FUNCTIONAL UHl I RES)'PHSE TtHE 2.

Intermedi a te Range Noni tops:

a.

Nputror~ Flux - bligh b.

]no))era tiye Average Powep Range Honitor".'.

Neutrons F]>ix - Upscale, Setgottln-b.

Flog Biased Simulated Theppql Powep - Upsga]e c.

$ jyed Hpqtron F)ug - Upsca]e d.

Inopepqtjve (e.

Pownscale (0 09)(aa)

< (0,09)

N$

Hg) 3.

Reactor Vessel Steam Ooze Pressure

- Pjgh 4.,

Reactor Vessel Mater Level - Low, Pevel'3 5.

Hajn Steam Ljne Isolation Palye - t:losgpe 6.

Hain Steam gine radiation - High 7.

(Ppimary Contajnment) (Orpull) Pryssupt. - gjgh 8.

Scpam Ojscpqpge Volupe )(ate}

Leypl - pigh 9.

Turbine Stpp Valve - Closure

]0.

Turbine Control Palve Fast Closure, frip Oil Pressupe

- Log ll.

Reactor Hode Swjtch Shutdown Positjon 12.

Manual Scratch

< (O.55)

< (l.05)

(0.06)

PA

< (0.09)

< (0.08)ll HA HA Neutron deLectors are exempt fpom response tjme testing.

Response

tjme shall be measured from the del.ectop oui.put op from the input of thy first electronic component in the chal>nel, (This provision is not applicable to Construction f'ermjts docketed after January 1, f978, See Regulatory Gujdp 1.18, November l977.)

"*(Hot) Including simulateg thermal power time constant, 6.a 1 seconds.

IIHeasured ft'om start of turbine control valve fast cloqury.

TABLE 4. 3. l.l-l REACTOR PROTECTION SYSTEH INSTRUt'lENTATION SURVEILLANCE RE UIRB!ENTS.

FUHCl IOHAL UNIT 1

~

I!>tet mediate Rapge

!!ai>jt'ot's;-

a.

Peutt'on flux'- !!jgh C! IRNNE)-

CHECK C!!ANHEL FUNCTIONS!

TEST S/u,g,(b)

Sgu('~, P S

OPERATIONAL

'HANNEL CONDITIO(S FOR WHICH CALIBRATION(-)

SURVEILLANCE RE UIREO 2

3) g,'5 b.

Inoperatjye 2.

Average Power Ra!lge llonjtor a.

Neutron Flux-

Upscale, Setdoyn S/U,S,(b)

S/U S

SA SA 2

3, 5

b.

Fjow Oiaseg Sjpq]atgd T!

e

)

I'P. et

= <P...)

S S((g~)

S/!j('-), g g("~('-);SA,(R("))

c.

Fixed 'Neutrons F]ux-Upscale d.

Iqoper ati ve (e.

Do>vnsca] e 3.

Reactor Vessel Steaqi Pome Pressure

- !!igh HA S

(S)

S/g(')

W W

SA (R)

'1

>, 2, 3, 5

I I) 1, 2

6.

Reactot Vessel Matet Level-Loq, Level 3

Hain Steam Line Isolation Valve - Closure

!lain Steam.Line Radiation-fligh (S) 1, 2 2( )

7.

(Primary Contajnment)

(Dtywell)

Pressure - Higll (S)

(R)

), 2

TABLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREt1ENTS FUNCTIONAL UNIT CIIANNEL CflECK CllA)i(EL OP ETIO JACAL FUJCTlO(A)

CHANNEL CONOI7IOliS P'OR MHIC(

TEST CALIBRATION SURVEILLANCE RE UIREO 8.

10.

12.

's)

Reactor Mode Switch Shutdown Positjon Manual Scram

=

Scram Discharge Volume pater Level - High (S) 7urbine Stop Va]ye - Clos)re (S)

Turbjne Control Valve-pap)

Closure palve Tl'jp SyStetq Oi] Pressure

- pow (8)

(5)

W NA 2 g(j) 1,2,3,4,5 (a)

Neural on detectors may be egclqded frog Cll/N(pp CALfpRA7/ON.

b)

The IRq any SRM channels shall be detail.mjned to ovel]ap fqr at ]gast (Q) deca)Ps dtlljPg eqcll startup after entering OPERATIONAL CONDlTION 2 and the I)g alod APRM chaqnels jhOll he determjped to overlap or at least (<) decades goring each coptrolleg shutdown, if not performed gifhjp the Previoqs g days.

(c) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months priOr to startup, if not perfolged yjthjl> tile previous 7 days, (d)

This calibration sl>all consist of the adjustment of the APRQ channel to conform to the power values calculated by a heat balance durjng OPERATIONAL CONDI7ION 1 when THERMAL POWER > 25K of RATED THERMAL POMER.

Adjust the APRM channe] if the a)solute difference is greater than 2X of RAT)0 7HERMAL POMER.

Any APRM channel gain adjustment made jq compliance with Specification 3.2.2 shall not be included in determjning the abso)ute djfferencp.

(e)

This calibration shal] consist of the adjustmppt of the APRM flow biased channe]

to conform to a caljbrated flow sjgnal.

(f)

The LPRMs shall be calibrated at. least once per 1000 pffectjve frill power hours (EFPH) using.the TIP system.

(g)

Verify measured core flow to be greater than or equal to established core flow at the exjsting pump speed.

((h) This calibration'shall consjst of (the adjustment, as requjred, of) (verjfying) the 6 t l second simulated thermal power time constant.)

. (i) 7his functjon is not required to be OPERABLE alien the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.

(j)

Pith any contlol rod withdrawn.

loot applicable to control gods removed per Specification 3.9.10.1 or 3.9.10.2.

INSTRUMENTATION 3/4..3.6 CONTROL R00 BLOCK INSTRUMENTATION LIMITING:CONDITION. FOR OPERATION 3;3.6.

The. control'od, block. instrumentation, channels shown in Table 3.3.6-1 shal'1'e.

OPERABLE with their trip setpoints. set. consistent. with the values shown in. the: Trip Setpoint column of'TabTe 3.3.6;2; APPLICABILITY."As shown-i'n Tabl'e 3.3'.6'-1'.

ACTION!

a.

With a control'od. block instrumentation channel trip setpoint less

- conservative. than the.: value shown in the Allowabl'e= Values column of Table 3'.3 6.-2'declare-the channel'noperable.

until the channel is restored, to OPERABLE status with its'rip setpoint. adjusted. consistent, wi'th the Trip'etpoint, val'ue.

~

b With. the. number, of OPERABLE channels. less than required'y the Minimum OPERABLE Channel's, per Trip. Function requirement,, take the ACTION~

requi,red by Table 3 3 6;-1 SURVEILLANCE RE UIREMENTS 4;3;6'ach of the above: required: control rod: block trfp, systems and.

instrumentation channels. shal.l-be demonstrated OPERABLE by the performance of the. CHANNEL CHECK, CHANNEL'. FUNCTIONALTEST and. CHANNEL-CALIBRATION operations for the OPERATIONAL CONDITIONS and. at the frequencies shown i'n Table 4.3.6-1.

GE-STS (BWR/4) 3/4 3-47

I

TABLE 3.3.6"1 CONTROL ROD BLOCK INSTRUMENTATION GJ I

CO TRIP FUNCTION l.

ROD BLOCK MONITOR ~

a.

Upsca]e b.

Ipopet ati ve c.

Downscale 2.

APRM a.

F]ow Biased Peutt'oq Plux-Upscale b,

Inoperatiyt.

c.

Dovinscale d.

Peutt'on Flux "-'pscale, Startup 3.

SOURCE RANGE MONITORS a.

Detector not fp]1 tp--

Upscale c.

Inoperati ye d.

Downs cal e INTERMEDIATE RANGE MONITORS a.

Detector not fu)] iq ---

b.

Upscale c.

Inopet ati~g)

. d.

Downscale QIQIMUM OPERABLE CQQQNELS PER TRIP FUNCTION 2

2

/

?

3

?

3

/

2 2

API!I /CABLE OPERATIONAL CONDITIONS

]

A' A'".

1l)gr5 2, 5 2

5 2

25.

2 5

2, 5

2r 5

2, 5

ACTION 60

$0

'60 61 61 61 61 61 61 61 61 6]

6]

61 61 61 61 61 61 5.

SCRAM DISCHARGE VOLUME a.

Water Level-Hjgh b.

Scram Trip.Bypass 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscal e b.

Inoperative c.

(Comparatot)

(Downscale)

(2)

(2) 2 2

2 1

2 5A'A (1

2 ) 5*$

1 1

1 62 62 62 62

TABLE 3.3.6-1 (Continued)

CONTROL'OD'LOCK'NSTRUMENTATION ACTION'0.

ACTTON Declare: the RBM'noperable. and. take. the ACTION. required by Speci fication: 3'.,1'.4.3'..

ACTION: 61

With the number of OPERABLE Channels:

a4 One less than required. by the Minimum OPERABLE Channels per Trip. Function requirement, restore the. inoperable channel'to OPERABLE status'ithin 7 days or place the i'noperahTe.. channel in the. tripped condition within.the next hour; b.

Two. or more. less. than required'y the Minimum OPERABLE Channel's per Trip: Function: requirement, place at. least one.- inoperab1'e: channel in: the.- tripped'ondition within.

one hour.=

AGTION 62

With. the number of OPERABLE, channel's, less than required:by the Minimum. OPERABLE. Channels:

per. Trip Function requi'rement,, place the. inoperab1'e. channel i'- the tripped'ondition within one hour..

I NOTES With.. THERMAL POWER > (30)X. of-RATED THERMAL POWER.

A With, more than one control'od:-withdrawn; 'Not applicable to control rods removed per Specification-3.9 10'.1 or 3.9;10.'2'he:

RBM'hall be automatically. bypassed.

when a peripheral control rod. is selected. (or the reference.-APRM" channel indicates less than (30)X of RATED THERMAL POWER)..

b'his function shall be. automatically bypassed ifdetector count, rate is

> 100. cps or. the IRM'hannels; are on range. (3) or. higher.

c; This function shal'1 be automatically channels: are: on range 8'r* higher;.

bypassed.

when the associated IRM d.

I Thi* function shall be automatically bypassed'when the IRM channels are on range 3 or higher.

e.

This; function shall be automatically bypassed when the IRM 'channels are on range l.

GE-STS (BWR/4) 3/4 3"49

TRIP FUNCTION TABLE 3.3.6"2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS-

. TRIP SETPOINT ALLOWBLE VALUE I

CJl CD l.

ROD BLOCK MONITOR a.

Ups cal e b.

Inoper atj ve c.

Downscale 2.

APRH a.

Flog Biased Neutpoq Flux-Upscale b.

Inoperative c.

Doynsca)e d.

Neutron Flux - Upsca)e, Stat'tup

.3.

SOURCE RANGE MONITORS a.

Detector not fq)) iq b.

Upscale c.

- Inoperative d.

Downscale'.

INTERHEDIATE RANGE HONITORS a.

Detector pop fu)) iq b.

Upscale c.

inoperative d.

Downscale 5.

SCRAM DISCHARGE VOLUME a.

pater Leve]-pigli b.

Scram Trip Bypass 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscal e b.

Inoperative c.

(Comparator)

(Downscale)

< O,t 6 M 'g (4a)$

HA

> (5)X of RAT)0 T(E(HAL PO+P r

< 0,66 g + (42)X"

> (5)X gf ggED TtIEP+P PPPER

< ($2)X of PAtPP 'fg)RMAL I'OlfEg

<,(2 x 10 ) cps.

Ng

> (3) gps W

< (108/j.p5) gjyjsjons of fu)] spa]e gA

> (5/125) piyjsjons of fu]l scale

< (

) inches Ng

< (108/$ 25) giyisions of fg]'f scale NA

< (10)X flog deyiatioD

< 0,66 g W (43)g W

> ())X of PATED 'fHE+$P POMMER

< 0.66 W t (45)X"-

> (3)X pf ECTED THECAL EO'EEIL

< (14+ of RgPD TNERHAI PO)'IER NA

< (5xlO) cps NA

> (2) cps PA

< (110/125) djvjsions of fu)l scale PA

> (3/125) givisjoqs of full scale

< (

) jncPes NA

< (ill/125) djVisions of full scale gA

< (11)X f)ow geviation "The Average Poyer Range ponitor pod b]ocfc functioD is yatied as a function of recjrculation loop flow (8).

/he trip setting of this function must be qaiptaiped ir, accordance qjt4 Specjfication 3.2.2.

TABLE 4.3.6-l CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE RE UIREMENTS TRIP FUNCTION C(A(NEL CHECK CHANEL OPgRATIOQL FU(CJlO@L CHANNEL

( )

CO)DITIONS FOR 'ERICH TEST CALIBRATION S RVEILLANCE RE UIRED 1.

ROD BLOCK MONITOR a.

Ups ca] e, b.

Inopepatjye c.

Pownscale 2.

APRM a.

Flow Biased Neutron Flux-Upscale b.

Inoperative c.

Downscale d.

Neutron.Flux = Upscale, Startpp 3.

SOURCE RANGE MONITORS a.

Detector got fu)) jn b.

Upsca]e c.

Inoperative d.

Downscale 4.

INTERMEDIATE RANGE MONITORS a.

Detector not fu)) jn b.

Upscale c.

Inoperative d.

Downscale 5..

SCRAM DISCHARGE VOLUME NA NA gA (Ng)

NA (Nq)

(NP)

Ng NA

~ NA S/U("-)('-

(

S g(h)( )'gM(q)

S/~(..)(c)'~ c

""(b) "

(>)

S/U s/u(>>',g s/u(>

'-'"(i) "

(h)

S/U(b)

S/U(b)'g S/U-S/U(p) ~'g (b)

S/U

'/u(b-)',q (Q)

NA (Q)

(Q)

R NA NA W

Q

],2)5

]

2, 5

2) 5 2) 5 2,

5 2) 5 2p f5 2,

5 2,

5 a.

Water Level-High.

NA b.

Scram Trip Bypass NA 6.

REACTOR COOLANT SYSTEM RECIRCULATION. FLOW a.

Ups cal e b.

Inoperative c.

(Comparator) (DowOscale)'O)

(Q) 5/U(b)

(b) "

S/U(b))M S/U R

NA Q

NA 2

5)hk (l, 2,) 5<<

1

TABLE 4; 3..6-1 (Continued)

CONTROL ROO BLOCK INSTRUMENTATION~ SURVEILLANCE RE UIREMENTS NOTES'..

a Neutron detectors may be excluded. from CHANNEL'ALIBRATION.

b.

Within. 24-hours; prior to. startup; i'f no't= performed, within the previ'ous. T days;,

c; Includes'eactor, manual control* muTtfpTexing. system; input.,

With THERMAL. POWER'" (30)X-of'ATED: THERMAL POWER II With more than, one;.control'od.withdrawn..

Not appli'cable to control'qds.

removed;.per. Specification 3'.,9..10.1 or 3'.9;10'.2'E-STS (BWR/4) 3/4 3-52;

)

~

'EFUELING OPERATIONS BASES 3/4; 9. 6 REFUELING PLATFORM The OPERABILITY requirements ensure that (1) the refueling platform will be used'or handling, control rods and: fuel assemblies. within the reactor pressure vessel (2) each crane and hoist has sufficient load capacity for handling fuel assemblies and control'ods,. and'3) the core internals and, pressure vessel are. protected. from excessive li'fting force-in the event they are inadvertently engaged during lifting operations..

3'/4.9.7 CRANE TRAVEL SPENT FUEL STORAGE POOL The restr iction on movement of loads in excess of the nominal weight. of a.

fuel assembly over other fuel assemtrlies in the storage pool ensures that in" the event. this. Toad is dropped. 1) the activity release will.be limited to that contained'n, a single fuel assembly, and, 2) any possible distortion of fuel in the storage racks wil'1 not result in a: critical array.

This assumption is.

consistent with the-activity rel'ease assumed in the: safety analyses.

3/4'.9.'

and: 3/4'.9;9 WATER'EVEL. REACTOR. VESSEL and MATER LEVEL -SPENT FUEL STORAGE POOL The restrictions on minimum'ater. level ensure that. sufficient water depth is avail'able to remove. (99)X of the. assumed (10)X iodine: gap activity released from: the rupture: of an irradiated, fuel assembly:

This; minimum water depth is, consistent. with: the assumptions of'he. accident; analysis.

3'/4.9'. 1'O'ONTROL ROD REMOVAL These.'pecifications ensure.- tha< mai'ntenance'or repair of control rods or controT r od",drives-. will be.- performed. under conditions that. limit the-probability of inadvertent criticality:.,

The. requirements for simultaneous removal of more-than one control rod are, more. stringent since the. SHUTDOWN MARGIN specification prqvides: for the core. to, remain-subcritical with only one control rod fully withdrawn.

3/4.9.'ll RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement. that. at..least one residual heat removal loop be OPERABLE or that. an alternate method. capable of decay heat removal be demonstrated.

and that an alternate method. of coolant'ixing be in operation ensures that 1) suf-ficient cooling capacity is avail'able to remove decay heat and maintain the water in the.. reactor pressure vessel below 140 F as, required during REFUELING, and 2) sufficient. coolant, circulation would" be. available through the reactor core to assure

accurate, temperature: indication and to distribute and prevent stratification of the poison in the event. it becomes necessary to actuate the standby liquid control system.

The requirement to have.two shutdown cooling mode loops OPERABLE when there is less than (23) feet of water above the reactor vessel flange ensures that a

single failure of.the operating loop will not result in a complete loss of resid-ual heat'emoval capability.

Mith the reactor vessel head removed and (23) feet of water above the. reactor vessel

flange, a large heat sink is available for core cooling.

Thus, in the event. a failure of the operating RHR loop, adequate time is provided to initiate alternate methods capable of decay heat removal or. emergency procedures to cool the core.

GE-STS (BWR/4)

B 3/4 9-2'

I

REFUELIHG OPERATIONS 3/4.9. 10 CONTROL ROD REMOVAL SINGLE CONTROL ROD REt10VAL LIHITIHG COHDITION FOR OPERATION 3'.9.10. T One control rod'nd/or the associated control rod drive mechanism r.:ay be removed from the core. and/or reactor pressure vessel provided that at.

least the following'equirements're satisfied until a control rod and associ-ated. control rod drive mechanism are reinstalled.

and the control rod is fully inserted in the core.

a.

The reactor.

mode switch is OPERABLE and locked in the Shutdown position or in the Refuel position per Table 1.2. and Specification 3.9.1.

b.

The source range monitors (SRM) are OPERABLE per Specification 3.9.2 c;

The SHUTDOMN MARGIN'. requirements: of'pecification 3.1.1 are satisfied,,

except that. the control 'rod; selected, to be removed;,

1 l/ay'e assumed. to be the-highest worth control rod'equired. to be. assumed. to be: fully withdrawn by the SHUTDOMH MARGIN test, and..

2.

Heed not be assumed.'o be immovable or untrippable.

d.

All other control" rods in a five-by-five array centered on the control rod, being'emoved: are inserted. and. electrically. or hydraulically disarmed'r the four, fuel assemblies surrounding the control rod or control rod drive mechanism to be. removed, from the core and/or reactor vessel are removed from; the: core cell.

e.

All'ther control rods are inserted.

APPl ICABILITY:

OPERATIONAL CONDITIONSnd'.

ACTTON:

Mith the requirements of the.- above specification not satisfied, suspend removal

.of the control rod and/or associated control rod drive mechanism from the. core and/or reactor pressure vessel and: initiate action to satisfy the above

'requirements.

I GE-STS (BMR/4) 3/4 9"12

~

I>>

)

l g

~ I

~ ~

~

REFUELING OPERATIONS

~ I, Y J SURVEILLANCE RE UIREHENTS:

4-;9'.10;1 Mithin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to. the start. of removal of a control rod. and/or the associated control'od dri've mechani'sm from the core and/or reactor pressure vessel and, at least once. per. 24'ours thereafter until a control rod and associ-ated control rod'rive mechani'sm~ are" reinstalled.

and. the control rod, is. inserted.

in the: core, verify. that:

I a

The reactor'ode switch is OPERABLE,and locked in the Shutdown position, or. in the. Ref'uel'osition< with. the "one. rod out." Refuel position interlock. OPERABLE: per, Specification 3'.9..1.

The SRM'hannel's are: OPERABLE per Specification 3.9.2.

The: SHUTOOVN I'!ARGINirequirements of Specification 3'.l are.- satisfied, per, Specifi'cation; 3'..10.1 c;,

A'll other. controT'ods'- fn a; five-by-five array centered: on the control rod.'eing, removed: are i'nserted and erectri'cally or hydraulically disarmed: or the* four fuel assembTfes; surrounding'he-control rod. or

-control'od'rfve mechanism. to'e. removed. from the: core and/or, reactor vessel are removed'rom. the core cel'1.,

e Al.l'ther, control, rods, are inserted.

GE-STS (BMR/4) 3/4. 9-13

~

~t,g

~

~ I v

~

jg.

REFUELING OPERATIONS MULTIPLE CONTROL ROD REMOVAL

~

'IMITING CONDITION FOR OPERATION:

I 3.9.10.2:

Any'umber of. control rods, and/or control rod drive mechanisms may be: removed; from the. core and/or, reactor pressure vessel provided that, at. least.

the following requirements are satisffed until all control rods and control'od drive. mechanisms are reinstalled'nd'. all control rods at e inserted'n the core"..

av The reactor-mode switch is OPERABLE and locked in the Shutdown position or fn the. Refuel position per Specification,3 9.1, except that the Refuel position "one-rod-. out"'nterlock may be bypassed, as. required, for those control rods and/or control'od. drive mechanisms to be removed, after-the: fuel assembl.ies'ave been removed. as specified..

beTow b.

The. source range'monitors.

(SRM), are. OPERABLF per Specification 3.9'.2.'.

The: SHUTDOMN MARGIN requirements of Specification 3.1..1 are. satjsfied d',

AlT other control'ods; are: either inserted ot have the surrounding four fueT assemblies.

removed from the core ceIT e;.

The four fuel assemblies-surrounchng each control-rod or control rod drfve: mechanism to. be: removed; from. the: core: and/or reactor vessel are: removed from the core cell.

APPLICABILITY:

OPERATIONAL CONDITION 5 ACTION:

Mith the requirements'f the.. above specification not satisfied, suspend removal of control. rods'nd/or control rod. drive, mechanisms from the core and/or reactor pressure vessel and, initiate action to. satisfy the above. requirements.

GE"STS (BMR/4) 3/4 9"14

~

~ ~

I

~ I v

~

'EFUELIHG; OPERATIONS SURVEILLANCE'E UIREMENTS 4 9;10'.2.1'fthi'n 4'ours prior to. the start'. of removal of control rods and/or.

control rod-drive. mechanisms from the core and/or reactor-pressure vessel and:

at least once per 24'ours; thereafter until'll control rods and control rod drive mechanisms are reinstalTed.

and. all control rods are inserted. in the core,,

verify. that:

a.

The reactor mode. switch's OPERABlE and locked. in the Shutdown position.

or in the-Refuel position-per Specification 3.9 1..

b; The SRM" channel's. are OPERABLE: per Specification 3;9.2; The.. SHUTOOMN-; MARGIN.'requirements: of'Specification 3'.l.l are satisfied..

\\

d; AlT other control" rods are either inserted or have-the surrounding.

four fuel'ssemblies removed;from. the core cell.,

e=

The four fuel-'ssembHes-.

surrounding; each control rod: and/or control' rod.'rive. mechanism: to. be. removed. from the. core and/or reactor'gessel're.-

removed: from'he. cora cel'7 4'.9'.10'.2..2.;

FoTYowi'ng: repTacement of.'le'ontroT." rods-and/or control rod drive:

mechanisms; removed'n. accordance; with. this, specificatfon, perform. a. functional test. of the. "one-. rod.-out."'efuel position; interlock,, ifthis function had', been bypassed'.

GE-STS (BMR/4)

'3/4 9-15 '

o

~

INSTRUMENTATION BASES 3/4.3'.5 REACTOR CORE ISOLATION COOLING SYSTEM'ACTUATION INSTRUMENTATION The. reactor core isolation cooling system actuation instrumentation is'rovided to. initiate actions to assure adequate core. cooling in the event of reactor isolation from. its primary heat. sink. and the loss of feedwater flow to the reactor vessel without providing actuation of any. of the emergency core cooling equipment.

'Operation. with a trip. set 1'ess conservative: than its Trip Setpoint.but within" its specified: Allowable Val'ue is acceptable.

on the basis that the.

difference between each Trip. Setpoint and: the Allowable Value is equal to or less than the drift. allowance. assumed. for each trip in the safety analyses.

3/4.,3..6 CONTROL R00 BLOCK'NSTRUMENTATION The. control rod block functions are. provided consistent with the require-ments of the speciffcations in S'ection 3/4 1..4, Gontrol Rod. Program Controls and Section: 3/4.2'ower Oistribution Limits.

The. trip logic is arranged.

so that.

a trip in. any: one.. of the inputs will result in a: control rod. block Operation with a, trip. set: less'onservative'than its Trip Setpoint. but within its specified A1Towable Value is. acceptable.

on the. basis that. the differencebetween, each, Trip'etpoint and'he All'owable VaTue is, equal to or Tess than the-drift allowance. assumed for each trip in the safety analyses; 3/4:.

3;-7'ONITORING'NSTRUMENTATION'/4'.

3'.7;1 RADIATION>MONITORING. INSTRUMENTATION..

The OPERABIL'ITY',of'he radiation monitoring instrumentation ensures. that;.

(1) the radiation levaTs are continually measured'n the. areas served. by the indivJdual'hannel's;,

(2) the alarm or automatic action is initiated when the radi" ation level trip setpoint. is'xceeded; and. (3) sufficient. information is. avail-able on selected. plant. parameters to monitor and assess these variables follow-ing an accident.

This capability is consistent witfrthe recommendations of (NUREG'-073T, "Clarification of TMI Action Plan R'equirements,"

November 1980).

3.4.3.T. 2 SEISMIC'ONITORING. INSTRUMENTATION The OPERABILITY of'he seismic: monitoring; instrumentation ensures that suf-ficient capability is available: to promptly determine the magnitude of a. seismic event. and, evaluate the response: of'hose features important to safety.

This capability is required to permit. comparison of the measured response to that used in the: design basis for the. unit.

(This instrumentation is consistent with the. recommendations of Regulatory Guide 1. 12 "Instrumentation for Earthquakes",

April 1974.)

3/4'. 3. 7; 3' METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the. meteorological monitoring instrumentation ensures that. sufficient meteorological data is available for. estimating potential radia-tion doses to the public as a.result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective. measures to protect the health and safety of the public.

(This instrumentation is consistent with the recommenda-tions of Regulatory Guide 1.23 "Onsite Meteorological Programs,"

February 1972.)

GE"STS (BWR/4)

B 3/4 3-4

~ ~

~ I p

g~

~

g

~

I

~

I

~

I

~

~ L

~

~ ~

~

I

~

~ ~

~

I I

~

I I

~

I

~

I

~

I

~

~ ~

~

~ ~

I

~

~ I

~

I

~

~ ~

~

~ ~

~

j' 0

g ~

~

I J

IE ~

ML17054D462

Text

RESPONSE

Response

Navigation menu

Search