Text

Order Confirming Licensee Commitments to Install Mods to Scram Discharge Sys by 831231
	ML20085B165
Person / Time
Site:	Hatch
Issue date:	06/24/1983
From:	Eisenhut D; Office of Nuclear Reactor Regulation
To:	; GEORGIA POWER CO.
Shared Package
ML20085B167	List: ML20085B165;
References
	TAC-42218, TAC-42219, NUDOCS 8307070557
	Download: ML20085B165 (35)
	v • d • e

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e 2:

7590-01 UNITED STATES OF AliERICA NUCLEAR REGULATORY COMMISSION In the Matter of

)

Dockets Nos. 50-321 and 50-366

)

GEORGIA POWER COMPAllY, ET AL

)

(Edwin I. Hatch Nuclear Plant,

)

Units 1 and 2)

)

ORDER C0tiFIRfilNG LICENSEE C0!HITMENTS Otl MODIFICATIONS TO THE SCRAM SYSTEM I.

The Georgia Power Company (GPC or the licensee) and three other co-owners are the holders of Facility Operating Licenses Nos. OPR-57 and NPF-5 which authorize operation of the Edwin I. Hatch Nuclear Plant, Units 1 and 2 (Hatch or the facilities) at steady state reactor power levels not in excess of 2436 megawatts thermal for each unit.

The facilities are boiling water reactors located at the licensee's site in Appling County, Georgia.

II.

During a routine shutdcwn of Browns Ferry Unit !!o. 3 on June 28, 1980, 75 of 185 control rods failed to fully insert in response to a manual scram from approximately 30% pcwer. All rods were subsecuently inserted within i

15 minutes and no reactor damage or hazard to the public occurred. Howeve r, the event did cause an in-depth review of the current BWR Control Rod Drive Systems which identified design deficiencies requiring both short and t

1cng-term corrective measures.

These measures were set forth in the

" Generic Safety Evaluation Report BUR Scram Discharge System", dated December 1,1980, prepared by the NRC staff.

8307070557 830624 PDR ADOCK 05000321 P

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, 7590-01 To provide reasonable assurance of safe operation pending implementation of long-term corrective measures, the short-term corrective measures have been implemented by lE Bulletin 80-17 (with supplements) and Orders issued on January 9,1981.

The Generic Safety Evaluation Report (SER) dated December 1,1980, endorsed the criteria and technical bases that were developed by a BWR Owners Subgroup for use in implementing permanent system modifications to correct identified deficiencies.

These criteria were designated as either functional, safety, operating, design, or surveillance, and when taken as a whole, comprise.an adequate set of criteria to resolve the issues raised during the Browns Ferry event investigation.

The SER further described an acceptable means of cocpliance with each criterion.

Pre-implementation approval of permanent modifications using the rethods described in the SER for compliance with the criteria will not be required.

Alternative methods of compliance will require specific NRC approval in advance of implementation.

In addition to the criteria proposed by the BWR Owners Subgroup, the SER added a criterion to address the potential for common cause failures of the scram level instrumentation. An acceptable means of complying with this criterion was the addition of diversity in.the design.

The addition of diverse instrumentation on the Scram Discharge Instrumented Volume will minimize recurrence of known common cause failures and thus improve system reliability.

t-

2.

_ 7590-01 Therefore, we.have concluded that diverse instrumentation should be pro-vided as required in the SER, with one exception: Alternative 2(d)(ii) has been deleted as a possible means of providing diversity, due to its reliance on prompt operator action.

The use of level sensors employing different operating principles, or the use of level sensors made by a different manufacturer, continues to be acceptabTe means of providing diverse instrumentation.

On October 1,1980, letters were sent to all BWR licensees requesting a commitment to reevaluate the present scram system and modify it as necessary to meet the design and performance criteria developed by the BUR Owners Subgroup.

The letter also requested a schedule for implementation.

III.

Because the implementation of modifications to meet the criteria proposed by +J1e BWR Owners Subgroup and endorsed by the NRC staff will restore the margins of safety in the BWR scram system, I have determined that these modifications should be completed on an expeditious schedule.

In response to our letter of October 1,1980, and additional discussions with the NRC staff, the licensee committed, by letters dated March 4,1981, and itay 13, 1982, to install the long-term modifications before Decemoer 31, 1983.

In view of the foregoing, I have determined that these comnitments are required in the interest of public health and safety and should, therefore, be confirmed by an immediately effective o; der.

7590-01 T*

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_4 IV.

Accordingly, pursuant to Sections 103,1611, and 182 of the Atomic Energy Act of 1954, as amended, and the Commission's regulations in 10 CFR Parts 2 and 50, IT IS HEREBY ORDERED EFFECTIVE IMMEDIATELY THAT:

1.

The licensee shall install the long term BWR scram discharge system modifications in conformance with the staff's Generic SER, which incorporates the BWR Owners Subgroup criteria, before December 31, 1983, or, in the alternative, the licensee shall place and maintain the facilities in a cold shutdown or refueling mode of operation until such modifications are made. Extensions of time for installation may be granted for good cause shown by the licensee. The modifications shall include diverse instrumentation as provided in the SER with the exception that Alternative 2(c)(ti) will not be accepted.

2.

For those cases in which a different method of ccmplying with the criteria than that described in the SER is chosen, the licensee shall suomit the design details and supporting analyses for aporoval to the Director, Division of Licensing, Washington, D.C.

20555 with a copy to the Regional l

Administrator of the appropriate NRC regional office, at least 3 months prior l

I to the required implementation date.

l 3.

Technical Specification changes required for operation witn the modified system shall be submitted at least 3 months prior to the recuired implementation date.

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. 7590-01 V.

The licensee may request a hearing on this Order wichin 25 days of the date of publication of this Order in the Federal Register. A request for hearing shall be submitted to the Director, Division of Licensing, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555.

A copy of the request shall also be sent to the Executive Legal Director at the U.S. Nuclear Regulatory Commission, Washington, D.C.

20555.

A REQUEST FOR HEARING SHALL NOT STAY THE IMMEDIATE EFFECTIVENESS OF THIS ORDER.

If a hearing is requested by the licensee, the Commission will issue I

an orde-designating the time and place of any such hearing.

If a hearing is held, the issue to be considered at such a hearing shall be whether the licensee should comply with the condition? set forth in Section IV of this Order.

The request for information nade in this Order was approved by OMB ander clearance number 3150-0083 which expires on December 31, 1983.

Comments on burden and duplication may be directed to the Office of Manage-ment and Budget, Reports Management, Room 3208, New Executive Office Buil ding, Washington, D.C.

This Order is effective upon issuance.

,FOR THE NUCLEAR REGULATORY CCMMISSION klLhIl

['N i, R

Carrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland this 24th day of June 1983.

4 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2. 2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setcoints shall be set consistant with the Trip Setpoint values shown in Tacle 2.2.1-1.

I APPLICABILITY: As shown in Taole 3.3.1-1.

ACTION:

With a reactor protaction system instrumentation set;cint less conservative than the value shown in :ne Allowaale Values column of Table 2.2.1-1, ceclare the channel inopersole and acply tne acplicable ACTICN statement requirement 4

of Specification 3.3.1 until the enannel is restored to OPERABLE status witn its satpoint adjusted consistant with the Trip Set;oint value.

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.s LIMITING SAFETY SYSTEM SETTING 5ASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 8.

Scram Discharte Volume Water Level-High The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor s: ram.

Should this volume fill up to a point where there is insufficient volume to accept the displaced water at pressures below 55 psig, control rod insertion would be hindered.

The reac-tor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to ace:mmodate the water from the movement of the rods at pressures below 55 psig when they are tripped.

The trip setooint for each scram discharge volume is equivalent to a contained volume of (

) gallons of water.

9.

Turbine Stoo Valve-Closure The turbine stop valve closure trip anti'cipates the pressure, neutron flux, and heat flux increases that would result from closure of the stoo valves. With a trip setting of (5)% of valve ci:sure from full Open, the l

resultan increase in heat flux is such that acequata thermal margins are maintained during the worst case transient (assuming the tur:ine cypass valves (f ail to) operate).

I 10.

Turcine Control Valve Fast Closure. Trio Gil Pressure-Lew The turbine centrol valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result feca fast closure of the tur ine controi valves due to load rejection coincident with failure of the tur ine by: ass valves.

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

This is achieved by the action of the fast acting solencic valves in ra: idly reducing hydraulic trip oil pressure a: :ne main tur:ine : ntrol valve actuator cisc dump valves.

This less of Oressure is sensed by pressure swit:nes wnese ::ntacts form tne One-cut-of-:wc-:wi:e 1 ;ic ir;u :o ne Reacter Or:te::icn System.

This trip setting, a f aster closure time, anc a dif ferent calve :Nar10:aris-i: from :nat Of :ne tur ine st:: valve, ::m:i9e :: ;r: duce ransients which are very similar to that f r the s:03 valve.

Relevan tran-sient analyses are discussed in Sectica (15.1.0) Of the Final Safety Analysis i

Repcrt.

11.

Rea::cr M:ce Saiten Shu:::wn ?csition The reactor mode switch Shutdown position is a reduncant channei to the aut:ma-ic protec-ive instrumentation channels and previces additional.anual reactor trip capability.

'12.

Facual Scram The Manual Scram is a redundant channel to the aut matic protective instrumentation channels and provide.s manual reacter tri: capability.

3E-STS (5WR/4) 5 2-9

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REACTIVITY CONTROL SYSTEMS 3/4.1.3 CONTROL RODS CONTROL R00 OPERABILITY LIMITING CONDITION FOR OPERATION 3.1.3.1 All centrol rods shall be OPERABLE.

APDLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

a.

With one control rod, inoperable due to being immovable, as a result of excessive friction or mechanical interference, or known to be untrippable:

1.

Within ene hour:

a)

Verify that the inoperable control red, if withdrawn, is separated frem al1~other incoeraole control rods by at least two control cells in all directions.

b)

Disar the associated directional control valves ** either:

l

1) Electrically, or

2) Hydraulically by closing the drive water and exhaust water isolation valves.

c)

C0 ply with Surveillance Recuirement 4.1.1.c.

Otherwise, be in at~least HOT SHUTCCWN within :ne next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , l

2.

Restore the inoperacle control red to CPERABLE status within 43 neurs l

or be in at least HOT 5 HUT 00WN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With one or more control rods trippable but inoperable for causes other l

than addressed in ACTION a, above:

1.

If the inocerable control roc (s) is withcrawn, wi:nin one hour:

a)

Verify that the inoperable withdrawn Ocntrol red (s) is separated frc all other inoperable control rods by at least two control cells in all directions, and b)

Cemenstrate the inserti:n capability of the 'n0; era:1e with:rawn l

centrol red (s) Oy inserting tne centrol rec (s) at least one noten by drive water pressure within the normal ::erating anga" Otherwise, insert the in::erable witndrawn control r:c(s) an: Oisar:

the associated directional control valves"* either:

l a)

Electrically, or I

b)

Hydraulically by closing the drive water and exhaust water l

isolation valves.

^Tne incperacie control red may then be withdrawn to a positicn no further withdrawn than its position when found to be inoperable.

"*May be rear:ed intermittently, under administrative control, to permit testing associated with restoring the control red to CPERABLE status.

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

REACTIVITY C0t4 TROL' SYSTEMS LIMITit:G C0tIDIT10ti FOR OPERATION (Continued)

ACTION (Continued) 2.

If the inoperable control rod (s) is inserted, within one hour disarm

the associated directional control valves " either:

a a)

Electrically, or b)

Hydraulically by closing the drive water and exhaust water l

isolation valves.

Otherwise, be in at least h0T SHUT 00bN within the next 12 hot:rs.

l With more than S control rods inoperable, be in at least HOT SHUT 00WN c.

within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVE!LLAtlCE REOUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by:

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At least once per 31 days verifying each valve to be open," and a.

5.

At least once per 32 days cycling each. valve -through ~at least one complete cycle of full travel.

4.1.3.1.2 When above the (preset power level) (low power setpoint) of the RkM l

and RSCS, all withdrawn control rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated CPERABLE by moving each control rod at least one notch:

At least once per 7 days, and a.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when any control rod is immovable as a result of excessive friction or mecnanical interference.

1.1.3.1.3 All control rocs shall be demonstratec CPERA3LE by performance of i

Surveillance Requirements 4.1.3.2, 4.1.3.4, 4.1.3.5, 4.'. 3.5 and 4.1.3.7.

^Tnese valves may be closed intermittently for testing under administrative controls.

"May se rearmed intermittently, uncer administrative control, to cermit testing associated with restoring the control rod to GPERA3LE status.

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

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REACTIVITY CONTROL SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) 4.1.3.1.4 The scram discharge volume shall be determined OPERABLE by demonstrating:

a.

The scram discharge volume drain and vent valves OPERABLE, when control rods are scram tested from a normal control rod configura-tion of less than or equal to (50)% R00 OENSITY at least once per 18 months, by verifying that the drain and vent valves:

1.

Close within (30) seconds after receipt of a signal for control rods to scram, and 2.

Open when the scram signal is reset.

l b.

Proper (ficat) (level sensor) response by performance of a CHANNEL l

FUNCTIONAL TEST of the scram discharge volume scram and control rod block level instrumentation (AT level measuring system) (after each scram from a pressurized condition) (at least once per 31 days).

1

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1 GE-STS (E'n'R/4) 3/4 1-5

.';s C.

REACTIVITI CONTROL SYSTEMS RCD ELCCK MONITOR LIMITING CONDITION FOR OPERATION 3.1.4.3 Soth red block monitor (REM) channels shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITION 1, wnen THERMAL COWER is greater than or equal to (30)% of RATED THERMAL POWER.

l ACTION:

a.

With one REM channel incperable, restore the incperable REM channel to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and verify that the reactor is not operating on a LIMITING CCNTROL R00 PATTERN; otherwise, place the incperbie red block monitor cnannel in the tripced condition within the next hour.

b.

With both REM cnannels inoperable, place at least One inoperable roc block monitor cnannel in tne tripped condition within one hour.

SURVEILLANCE REOUIREMENT5 a.1. 4. 3 Eacn of ne above requirec.EM cnannels snali :e cemenstrated C?ERAELE cy performance of 1:

a.

CHANNEL FUNCTIONAL TEST and CHANNEL CALIERATION at the frequencies and for the OPERATIONAL CCNDITIONS specified in Tacle 4.3.5-1.

b.

CHANNEL FUNCTIONAL TEST prior to control roc witacrawal when tne reactor is cperating on a LIMITING CONTROL RCD PATTERN.

GE-STS (EWR/4) 3/4 1-18

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REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 CONTROL RODS The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertior; 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 the same time impose mo e festrictive criteria for continued

' operation.

A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be ktot to a minimum.

The requirements for the various scram time measurements ensure tact any indication of systematic problems with rod drives will be investigated on a timely basis.

Damage within the cor, trol rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is r:asonable to determine the cause of the inoperability and at the same time prevent ooeration with a large number of inoperacle control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are consistent with the SHUTDOWN MARGIN requiremer.ts.

The number of control rods permitted to be incperable c:uld be m:re than the eignt allowed by the specification, but the occurrence of eight incperable rods could be indicative of a generic problem and the reacter must be shutdown for invest'gation and resolution of the problem.

The control red system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than (1.05) curing the limiting power transient analyzed in Section (15.__) of the FSAR.

This analysis shews that tne negative reactivity rates resulting from the scram with the ace age rescense of a!' tne drives as given in tne specifications, Orovide :ne required protection anc MCPR remains greatar than (1.05).

The occurrence of scram times loncer nen tnose scecified shoulo :e viewed as an incicati:n of a systemic proble5 with tne red crives and therefore the surveillance interval is recuced in Orcer to prevent Oceration of :ne reactor f:r long cariccs of time with a potentially serious preolen.

The scram ciscnarge volume is recuired to be CPERABLE so that it will be available wnen needed to accect ciscnarge water from tne control rocs during a l

reactor scram and will isolate the react r c0olant system fr0m the ::ntainment when required.

Centrol rods with inoperable accumulaters are declared incperable and Specification 3.1.3.1 then applies.

This prevents a pattern of inceerable accumulators that would result in less reactivity insertion on a scran than has been analyted even though control rods with inoperable accumulators may still be inserted witn normal. drive water pressure.

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

l GE-STS (SWR /4)

B 3/4 1-2

,=

'Ed REACTIVITY CONTROL SYSTEMS BASES CONTROL R005 (Continued)

Control red coupling integrity is required to ensure ccmpliance with the analysis of the red drop accident in the FSAR.

The overtravel position feature provices the only positive means of determining that a red is properly coupled and therefore this check must be performed prior to acnieving criticality after completing CORE ALTERATIONS that could have affected the centrol red coupling integrity.

The subsequent check is performed as a backuo to the initial demen-stration.

In order to ensure that the control roc patterns can be followed and tnere-fore that other parameters are within their limits, the control red position indication system must be OPERABLE.

The control red housing succort restricts the outward movement sf a :ontrol rod to less than (3) inches in the event of a housing failure.

The amount of red reactivity which could be added by this small amount of red withdrawal is less 'than a normal withdrawal increment and will not contribute to any camage to the primary coolant system.

The sucport is not required wnen there is no pressure to act as a driving force to rapicly eject a drive housing.

The required surveillance intervals are adetuate to cetermine that the rods are OPERABLE and not so frecuent as to cause excessive wear en the system ccmcenents.

3/a.1.1 CONTROL RCO ?c0 GRAM CONTROLS Centrol red withdrawal and insertion secuences are estamiisnec :: assure that one maximum insequence individual control red or c:ntrol red segments which are withdrawn at any time during the fuel cycle could not ce wortn enougn to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod creo accicent.

Thc icecifiec secuences are cnaracterized by homogeneous, scatterec patterns of centrol roc withdrawal. When THERMAL PCWER is greatar than (20)% of RATED THERMAL PCWER, there is no possible red worta which, if drec=ed at the design rate of the velocity limiter, c uld result in a peak enthaley of 280 cal /gm. Thus retuiring tne RSCS ar.d RhM to :e OPERA 3LE.nen THERMAL PCWER is less than or ecual to (20)% of RATED THERMAL ?ChER provices acecuate ::ntr:1.

The RSCS anc IWM Orevide aut:matic sucervision : assure that cut-f-secuence recs will not ce witnerawn Or insertec.

The analysis of the r0d croc accident is presentec in Section (15.

) of tne FSAR anc the tacanicues of tne analysis are cresentec in a taaical Ficort, Reference i, acc two su:clements, References 2 anc 3.

The REM is cesignec to autcoatically ;revent fuel camage in :ne event of erroneous roc witnerawal from Iccations of nign ocwer censity curing hign :ower c;eration.

Two enannels are provided.

Tricping one of tne cnannels will :lecx erroneous rod withcrawal soon encugn to prevent fuel damage.

This system cacks up tne written secuence used by tne operator for witnerawal of c:ntrol recs.

GE-STS (SWR /a) 3 3/4 1-3

.?

3/4.3 INSTRL' MENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

APoLICA2ILITY:

As shcwn in Table 3.3.1-1.

ACTION:

With the number of OPERABLE channels less than requirec by the Minimum a.

OPERABLE Channels per Trip System requirement for one tric system, place the inoperable channel (s) and/or that trip system in tne tripped :enci-tion

within one hour.

The provisions of Specification 3.0.4 are not applicable.

b.

With the number of OPERABLE channels less than recuired by the Minimum OPERABLE Channels per Trip System requirement for both trio systems, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1.

SURVEILLANCE REOUIREMENTS 4.3.1.1 Eacn reactor protaction system instrumentation cnannel shall be demonstrated OPERABLE by tne performance of the 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 and simulated automatic operation of all cnannels shall be performea at least once per 13 mentns.

4. 3.1. 3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Tacle 3.3.1-2 snall te cemonstratec :: be within its I

limit at least once per la months.

Each test shall include at least one l

Channel per trip sys.em such.ha. all channels are testec at least once every i

N times 18 months where N is the total number of redundant :nannels in a s ecific reactor trip system.

l "An :ncperacie :nannel neec not ce piacec in :ne tri :ec conci-ion wnere tnis l

woulc cause the Trip Function to 0 :ur In : ness :ases, tne i : erable cha-,e; snail be restorec to GPER* ELE status wi.hir 2 n:urs r ne CT'.."

e:uired 53 Table 3.3.1-1 for tnat T-i: Function snail te.a3.en.

l "If more channels are incpera:le in one tri: syster. tnan in the 0:ner, place the trio system wita mcre inc ersole channels in :ne trippec con:ition, excect wnen this would cause the Trip Function to oc:ur.

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

li; TAllil 3.3.1-1 Iti ACIOlt Pit 0lECilull SYSilH !!1 Silt 0HENTATION M

i*f Al'Pl.lCAllLE HINIMUM c

OPEllAllollAI.

OPEllAllLE CllANNELS l ullCllutlAL lillI~l Coll 0lliolls Pelt lillP SYSlLH (a)

ACT10tl 1.

Intenneiliate Itange lionitorsII'I:

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3 I

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1 S.

liain Steam Iine Isolation Valve -

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Closure I

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Rf ACIOR PROTEclloil SYSTEM lilSIRUMEllTATION u

APPLiCAllt E HINIHUH R

OPERAI10NAl.

OPERABLE CllANNELS S

fullC110NAL UtilI CONDillotiS PER TRIP SYSTLH (a1 ACTION 6.

Main Steam Line Itadiation -

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1 8.

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I) 4 6

9.

Turlaine Stop Valve - Closure I

{

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11.

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1 3, 4 1

7 S

I 3

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12.

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1 8

3, 4 2

l 5

2 9

1 J

A

..s TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION ACTION 1 Se in at least HOT SHUT 00WN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 Verify all insertable control rods to be insertec in the core and lock the reactor moce switch in the Shutdown position within one hour.

ACTION 3 Suspend all operations involving CORE ALTERATIONS" and insert all insertaole control rocs witnin one hour.

ACTION 4 Se in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 5 Se in STARTUP with the main steam line isolation valves closed within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months or in at least HOT SHUT 00WN witnin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 5 Initiate a reduction in THERMAL PCWER within 15 minutes and reduce turbine first stage pressure to < (250) psig, ecuivalent to THERMAL POVER less than (30)% of RATED THERMAL PCWER, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

ACTION 7 Verify all insertsole control rods to be inserted within one-hour.

ACTION 8 Lock the reactor mode switen in the Shut:0wn position within one hour.

ACTION 9 Suspend all operations involving CORE ALTERATIONS", and insert all insertaole control rods and lock the reactor mode switch in the SHUTOCWN position within one hour.

I "Ex:ept ::.ement of IR!' SR" or spe:ial ::.aole de e:::rs, or rea'.a:ement of L?? strings crevi:ec SR" instrumentation :s CPERAELE cer 3:ecifi:st Or I. i 2.

GE-STS (EWPd4) 3/a 3-4

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trio system is monitoring that parameter.

(c) This function shall be automatically bypassed when the reacter mode switcn is in the Run position.

(c) The " shorting links" shall be removed from the RPS circ.uitry prior to and curing the time any control r0d is withcrawn" and snutdown margin demonstrations performed per Specification 3.10.3.

(d) The non-coincident NMS reactor trip function logic is such that all channels go to both trip systems.

Therefore, when the " shorting links" are removec, tne Minimum GPERABLE Channels Per Trip System is 4 APRMS anc 5 IRMS.

(e) An APRM channel is inoperable if there are less than 2 L?RM inputs per level or less than (11) LPRM inputs to an APRM channel.

(f) This. function _is not.recuired.to be.0PERAELE when.the reactor-pressure vessel head is unbolted or removed per Scecification 3.10.1.

(g) This function shall be automatical.ly bypassed when the reactor moce switen is not in the Run position.

(h) This function is not required to be OPERAELE when PRIMARY CCNTAINMENT INTEGRITY is not required.

(i) With any control rod withdrawn.

Not tiplicable to control rces removed cer Speci fication 3. 9.10.1 or 3. 9.10. 2.

(j) This function shall be automatically bypassed hen turbine first stage pressure is 5 (250) psig, equivaient to THERMAL PCwER isss tnan (30)%

of RATED THERMAL POWER.

(k) Also actuates tne ECC-RPT system.

i t

i

' :: re;Jirec for control rods remces: :e-5pecifica :n 2.9.10.1 r 3.9.10.2.

GE-STS (EWR/c) 3/c 3-5

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INSTRUMENTATION 3/4.3.6 CONTROL ROD 3 LOCK INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.6.

The control rod block instrumentation channels shown in Table 3.3.5-1 shall be OPERABLE with their trip setpoints set consistent with the values

,shown in the Trip Setpoint column of Table 3.3.5-2.

APPLICABILITY: As shown in Table 3.3.6-1.

ACTION:

a.

With a control rod block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjustad consistent with the Trip Setpoint value.

b.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, take the ACTION required by Table 3.3.5-1.

SURVEILLANCE REQUIREMENTS 4.3.6 Each of the above required control rod block trip systems and instrumentation channels shall be demonstrated OPERABLE by the cerformance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATICNAL CONDITIONS and at the frequencies shown in Table 4.3.5-1.

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

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CONTROL RCD BLOCK INSTRUMENTATION ACTION Declare the RSM inoperable and take the ACT:3N required by ACTION 60 Specification 3.1.4.3.

With the number of OPERABLE Channels:

ACTION 61 a.

One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore tne inoperable channelto OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour.

b.

Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, clace at least one inoperable cnannel in the tripoed condition within o..e hour.

ACTION 62 -

With the number of OPERABLE cnannels less tnan required by tne Minimum OPERABLE Channels per Trip Function requirement, place the innperaole channel. in_ he trippec. condition. witnin one hou.

NOTES With THERMAL POWER > (30)% of RATED THERMAL PCWER.

With more than one control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

a.

The RSM shall be automatically bypassed when a peripheral control rod is selected (or the reference APRM channel indicates less tnan (30)% of RATED THERMAL POWER).

5.

This function shall be automatically byoa: sed if detector count rata is-

> 100 cps or the IRM cnannels are on range (3) or higner.

c.

This function snall b'e automatically bypassea wnen the associated IRM

hannels are on range S or higher.

d.

This function snail be automatically bypassed when the IRM channels are on range 3 or higher, This function shall be automatically bypassed when the IRM channels are e.

on range 1.

GE-STS (SWR /4) 3/4 3-49

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[3 CllAllNEL OPERAll0NAL CilANNEL lutlCIIONAL CllANilEL CONDITIONS FOR WillCll

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SURVEILLANCE REQUIRED I

1.

ROD lil0CK M0lllIOR a.

Upscale NA S/U(b)((1 (c) gx q

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S/U(b)(c), (<-)

NA l^

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Downscale NA 2.

APRM a.

Flow Biasettlieutron " lux -

Upscale (llA)

S/U

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Inoperative NA S/Ug,),M NA 1, 2, 5 c.

Downscale (NA)

S/Ug,),M (Q)

I d.

Neutron flux - Upscale, Startup (NA)

, S/U

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3.

SOURCE RANGE MONIIORS -

a.

Detector not full in NA S/il,,W NA 2, S y

(n 12.

Upscale NA S/U(b),W Q

2, S 1

c.

I nopera t.ive NA S/Ug,),W NA 2, S d.

Downscale NA S/U

,W Q

2, S 4.

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Detector not tull in NA S/U b.

lipscale NA S/U b,W Q

2, S S/Il,,)),W NA 2, 5 c.

Inoperative NA d.

Downscale NA S/U

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TABLE 4.3.6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REOUIREMENTS NOTES:

a.

Neutron detectors may be excluded from CHANNEL CALIBRATICN.

b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7 days.

c.

Includes reactor manual control multiplexing system input.

With THERMAL POWER > (30)% of RATED THERMAL POWER.

With more than one control rod withdrawn.

Not aoplicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

GE-ST5 (5WR/4) 3/4 3-52

REFUELING OPERATIONS BASES 3/4.9.6 REFUELING PLATFORM The OPERASILITY requirements ensure that (1) toe refueling platform will be used for 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 rods, and (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

t 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL The restriction on movement of loaos in excess of tne nominal weight of a fuel assembly over other fuel assemblies in the storage pool ensures that in the event this load is dropped 1) the activity release will be limited to that contained in a single fuel assembly, and 2) any possible distortion of fuel in the storage racks will not result in a critical array.

This assumption is consistent with the activity release assumed in the safety analyses.

l 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL -SPENT FUEL STORAGE PCOL The restrictions on minimum water level ensure that sufficient water depth is available to remove (99)% of the assumec (10)% iodine gao act vity released from the rupture of an irradiated fuel assembly.

This minimum water depth is consistent with the assumptions of the accident analysis.

3/4.9.10 CONTROL RCD REMOVAL These specifications ensure that maintenance or repair of control rods or control rod drives will be performed under conditions that limit the orobability of inadvertent criticality.

The requirements for simultaneous removal of more than one control rod are more stringent since the SHUTDCWN MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn.

3/4.9.11 RESICUAL HEAT REMOVAL AND CCOLANT CIRCULATICN The requirement that at least one residual heat removal loop be OPERASLE or that an alternate method cacable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation ensures that 1) su'-

ficient cooling capacity is available to remove decay heat and maintain tne water in the reactor pressure vessel below 100 F as recuirac during REFUELING, and 2) sufficient coolant circulation would te available tnrougn tne reactor core to assure accurate temcerature incication anc to cistribute and o event i'

stratificatica of the poison in tne event it becomes necessary to actuate tne stancby liquic control system.

The reouirement to nave two snutcown cooling moce locos CPERABLE onen there is less nan (23) feet of water above the reactor vessel flange ensures nat a single failure of the coerating icop will not result in a com lete loss of resic-ual neat removal capability.

Witn the reactor vessel head removed and (23) feet of water aoove the reactor vessel flange, a large neat sink is available for core cooling.

Thus, in the event a failure of tne coerating RHR loop, adequate time is provided to initiate alternate methods capable of decay heat removal l

or emergency procedures to cool the core.

4 l

GE-STS (BWR/4) 3 3/c 9-2

o REFUELING OPERATIONS 3/4.9.10 CONTROL R00 REMOVAL SINGLE CONTROL RCD REMOVAL LIMITING CONDITION FOR OPERATION 3.9.10.1 One control rod and/or the associated control rod drive mechanism may be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until a control red and associ-ated control rod drive mechanism are reinstalled and the control red is fully insertec in the core.

a.

The reactor mcde 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 SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied, except that the control rod selected to be removed; 1.

May be assumed to be the highest worth centrol rod required to be assumed to be fully withdrawn by the SHUIDCWN MARGIN test, and 2.

Need not be assumed to be imacvable er untrippable.

d.

All other control rods in a five-by-five array centered on the control rod being removed are inserted and electrically or hydraulically disarmed or the four fuel assemolies surrounding the control red or control rod drive mechanism to be removed frem the core and/or reactor vessel are removed from the core cell.

a.

All other centrol rods are inserted.

APPLICABILITY:

OPERATIONAL CONDITIONS 4 and 5.

ACTION:

With the requirements of the above specification not satisfied, suspend removal of the control rod and/or associated centrol rod drive mecnanism from tne core and/or reactor pressure vessei and initiate action to satisfy the above requirements.

I i

1 GE-STS (SWR /4) 3/4 9-12

REFUELING OPERATIONS SURVEILLANCE REOUIREMENTS 4.9.10.1 Within 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 rod drive mechanis.t. from the core and/or reactor pressure vessel and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter until a control rod and associ-ated control rod drive mechanism are reinstalled and the control rod is inserted in the core, verify that:

The reactor mode switch is OPERABLE and locked in the Shutdown position a.

or in the Refuel position with the "one rod out" Refuel position interlock OPERABLE per Specification 3.9.1.

b.

The SRM channels are OPERABLE per Specification 3.9.2.

l The SHUT 00WN MARGIN requirements of Specification 3.1.1 are satisfied c.

per Speci fication 3. 9.10.1. c.

l d.-

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

e.

All other control rods are insertad.

l l

GE-STS (EWR/4) 3/4 9-13 L

o 3

./

m

~

REFUELING OPERATIONS MULTIPLE CONTROL ROD REMOVAL LIMITING CONDITION FOR OPERATION 3.9.10.2 Any number 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 satisfied until all control rods and control rod drive mechanisms are reinstallec and all control rods are inserted in the Core.

The reactor mode switch is OPERABLE and locked in the Shutdown position a.

or in the Refuel Refuel position " position per Specification 3.3.1, except that the one rod-out" interlock may be bypassed, as required, for those control rods and/or control rod drive mechanisms to be removed, after the fuel assemblies have been removed as specified below.

b.

The source range monitors (SRM) are CPERABLE per Specification 3.9.2.

The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.

c.

d.

All other control rods are either inserted or have the surrounding four fuel assemblies removed from the care ceil.

The four fuel assemblies surrounding each control rod or control rod e.

crive mechanism to be removed from the core and/or reactor vessel l

are removed from the core cell.

APPLICABILITY:

OPERATIONAL CONDITION 5.

ACTION:

With the requirements of the above specification not satisfied, suspend removal of control rods and/or control rod drive mechanisms from the core anc/or reactor pressure vessel and initiate action to satisfy the acove recuirements.

l l

i l

i GE-STS (BWR/4) 3/4 9-14 n -

e REFUELING OPERATIONS SURVEILLANCE REGUIREMENTS 4.9.10.2.1 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the start of removal of control reds and/or control rod drive mechanisms from the core and/or reactor pressure vessel and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter until all control rods and control red drive mechanisms are reinstalled and all control rods are inserted in the core, verify that:

a.

The reactor mode switch is OPERABLE and locked in the Shutdcwn position or in the Refuel position per Specification 3.9.1.

b.

The Sai channels are OPERABLE per Specification 3.9.2.

c.

The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.

d.

All other control rods are either inserted or have the surrounding four fuel assemblies removed frem the core cell.

e.

The four fuel assemblies surrounding each control rod and/or control rod drive mechanism to be removed frca the care and/0r reactor zessel are removed from the core cell.

4.9.10.2.2 Following replacement of all. control rods and/or control red drive mechanisms removed in accordance with this specification, perform a functional test of the "One-red-cut" Refuel position interlock, if this function had been bypassed.

GE-STS (BWR/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 provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater ficw to the reactor vessel without providing actuation of any of the emergency core cooling equipment.

Operation with a trip set less conservative than its Trip Set::oint but within its specified Allowable Value is acceptacle on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumec for each trip in the safety analyses.

3/4.3.5 CONTROL ROD BLOCK INSTRUMENTATION l

The control rod block functions are provided consistent with the require-ments of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/A.2 Power Distribution Limits.

The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.

Cperation with a trip set less conservative than its Trip Setpoint but within its specified Allowacle Value is acceptable on the basis that the difference Detween each Trip Setooint and the Allowable Value is ecual to or less than the drift allowance assumed for each trip in the safety analyses.

3/4.3.7 MONITORING INSTR 9ENTATION

~

3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERASILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the individual channelsi (2) the alarm or automatic action is initiated when the radi-ation level trip set::oint is exceeded; and (3) sufficient information is avail-aole on selected plant parameters to monitor anc assess these variables follow-ing an accicent.

This capability is consistent with the recom:::encations of (NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980).

3.3.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring instrumentation ensures tnat suf-ficient cacability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features im::ortant to safety.

This ca::acility is required to permit comcarison of the measured resconse to that usec in the design basis for the unit.

(This instrumentation is consistent with tne recommendations of Regulatory Guice 1.12 " Instrumentation for Earthquakes",

Acril 1974. )

3/J.3.7.3 METEOR 0LCGICAL MCNITCRING INSTRUMENTATION The OPERASILITY of ne meteorological monitoring instrumentation ensures that sufficient meteorological data is availaole for estimating potential radia-tion cases to tne public as a result of routine or e.ccidental 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 tne public.

(This instrumentation is consistent with the recommenda-tions of Regulatory Guide 1.23 "Onsite Meteorological Programs," February 1972. )

GE-STS (E',iR/4) 3 3/4 3-4

M 3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a.

Preserve the integrity of the fuel cladding.

b.

Preserve the integrity of the reactor coolant system.

c.

Minimize the energy which must be adsorbed following a loss-of-coolant accident, and d.

Prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to creserve the ability of the system to perform its intended function even during ceriods wnen instrument enannels may be out of service because Of main-tenance. When necessary, one channel may be made inoperable for brief intarvals to conduct required surveillance.

The reactor protection systam i's made up of two independent trip systems.

There are usually four channels to monitor each parameter witn two channels in each trip system.

The outputs of tne channels in a trip system are comoined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram.

The system meets the intent of IEEE-279 for nuclear power plant protection systems.

The bases for the trip settings of tne RPS are ciscussed in the cases for Scecification 2.2.1.

The measurement of resconse time at the scecified frecuencies cr0vides assurance that the crotective functions associated witn esca channel are ccm-pleted within the t'ime limit assumed in the accident analysis.

No creci: was taken for these channels with resecnse times indicated as nct s:plicable.

Response time may be demcastrated by any series of secuential, Overl'apping or total channel tast measurement, Orovided such tests demonstrate tne total channel response time as defined.

Sensor rescense time verification may be demonstra:ec by either (i) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensers with certified respense times.

GE-STS (5WR/4) 3 3/1 3-1

v t e Letter Sequence Other
TAC:42218, (Approved, Closed) TAC:42219, (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance... Administration Questions 1 September 1981 Answers Results Approval... Other: ML19327A527, ML19327A533, ML19330B521, ML19330B523, ML19331B934, ML19336A759, ML19340B179, ML19344B508, ML19347C984, ML19347C997, ML19350A630, ML20053A188, ML20080J137, ML20080P638, ML20085B165, ML20244B663, ML20244B666
MONTHYEAR1983-10-03 [Table View]

ML20085B165

Text

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