ML19309H334

Amends 73 & 14 to Licenses DPR-57 & NPF-5,respectively, Substituting Equivalent Terminology for Computation of Average Power Range Monitor Rod Block & Scram Setpoints & Revising Associated Surveillance Requirements

ML19309H334
Person / Time
Site:	Hatch
Issue date:	04/17/1980
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML19309H335	List: ML19309H334 ML19309H337 ML19309H340
References
TAC-13278, NUDOCS 8005130064
Download: ML19309H334 (46)

Text

.

pnsfCp UNITED STATES g

NUCLEAR REGULATORY COMMISSION n

{.'3 WASHINGTON, D. C. 20555 s-3005130 "s, -

jr GEORGIA POWER COMPANY f

OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321_

EDWIN I. HATCH NUCLEAR PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 73 License No. DPR-57 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The applications for amendment by Georgia Power Company, et al.,

(the licensee) dated October 18, 1978 and November 8, 1979, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regu-lations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be i

conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

l 2.

Accordingly, the license is amended by changes to the Technical 1

Specifications as indicated in the attachment to this license amend-j ment, and paragraphs 2, 2. A, 2.B and 2.C.(2) of Facility Operating i

License No. DPR-57 are hereby amended to read as follows:

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2-2.

Facility Operating License No. DPR-57 is hereby issued to the Georgia Power Company, the Oglethorpe Power Corporation, the Municipal Electric Authority of Georgia and the City of Dalton, Georgia to read as follows:

A.

This license applies to the Edwin I. Hatch Nuclear Plant Unit No.1, a direct cycle boiling water reactor and associated equipment (the facility), owned by the Georgia Power Company, the Oglethorpe Power Corporation, The Municipal Electric Authority of Georgia and the City of Dalton, Georgia. The facility is located eleven miles north of Baxley in Appling County, Georgia, and is described in the ' Final Safety Analysis Report' as supplemented and amended (Amendments 9 through 46) and the Environmental Report as supplemented and amended (Supplement I and Amendment 1).

B.

Subject to the conditions and requirements incorporated herein, the Commission hereby licenses.

(1) Pursuant to Section 104b of the Act and 10 CFR Part 50,

' Licensing of Production and Utilization Facilities,'

1 Georgia Power Company to possess, use, and operate the f

facility at the designated location in Appling County, Georgia, in accordance with the procedures and limitations set forth in this license; and the Georgia Power Company, the Oglethorpe Power Corporation, The Municipal Electric Authority of Georgia and the City of Dalton, Georgia to possess the facility in accordance with the procedures and limitations set forth in this license; C.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 73, are hereby incor-porated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUC' R REGULATORY COMMISSION je Thomas A. Ippolit, Chief Operating Reactors Branch #3 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 17, 1980 5

ATTACHMENT TO LICENSE AMENDMENT NO. 73 FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the Appendix "A" Technical Specifications The revised pages are identified by Amendment with the enclosed pages.

number and contain vertical lines indicating the area of change.

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i Insert Renove 1.0-7 1.0-7 1.1-2 1.1-2 1.1-3 1.1-3 1.1-13 1.1-13 3.1-1 3.1-1 3.1-2 3.1-2 3.1-17 3.1-17/18 3.1-1B*

3.11-4 3.11-4

• 0verleaf page provided for convenience only.

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HM.

Minimum Critical Power Ratio (MCPR) - Minimum Critical Power Ratio (MCPR) is the value of the critical power ratio associated with the most limiting assembly in the reactor core. Critical Power Ratio (CPR) is the ratio of that power in a fuel assembly, which is calculated to cause some point in the assembly to experience boiling transition, to the actual assembly operating power.

NN.

Trip System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action. Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.

00.

Cumulative Downtice - The cumulative downtime for those safety components and systems whose downtime is limited to 7 consecutive days prior to requiring reactor shutdown shall be limited to any 7 days in a consecutive 30 day period.

PP.

Fire Suppression Water System - A Fire Suppression Water System shall consist of: water storage tanks, pu=ps, and distribution piping with associated sectionalizing control or isolation valves. Such valves include yard hydrant curb valves, and the first valve ahead of the water flow alarm device on each sprinkler, hose stand pipe or spray system riser.

QQ.

Channel Calibration - A Channel Calibration is the adjust =ent, as necessary, of the 1

channel output such that it responds with the necessary range and accuracy to known values of the para =eter which the channel monitors. The Channel Calibration shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the Channel Functional Test.

The Channel Calibration may be per-f ormed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

RR.

Channel Functional Test - A Channel Functional Test shall be:

a.

Analog channels - the injection of a siculated signal into the channel as close to the primary sersor as practicable to verify operability including alarm and/or trip functions.

b.

Bistable Channels - the injection of a simulated signal into the channel sensor to verify operability including alar = and/or trip functions.

SS.

Fraction of Limiting Power Density (FLPD) - the ratio of the linear heat generation rate (LEGR) existing at a given location to the design LHGR for the bundle type.

Design LHGRs are 18.5 KW/ft for 7x7 bundles and 13.4 KW/ft for 8x8 bundles.

17.

Core Maximum Fraction of Limiting Power Density (CMFLPD) - the CMFLPD is the highest value existing in the core of the FLPD.

l Amendment No. I$, 50, 73 f

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LIMITING SAFETY SYSTEM SETTINGS SA E LIMITS 1.1.D. Reactor Water Level (Hot or Cold 2.1. A.l.c. APRM High High Flux Scram Trip Setting (Run Mode) (Continued)

Shutdown Condition)

S < 0.66 W + 54%

Whenever the reactor is in the Hot or Cold Shutdown Condition with irradiated fuel in the reactor vessel, where:

the water level shall be > 378 inches above vessel invert when fuel is S = Setting in percent of rated thermal power seated in the core, (2436 MWt)

W = Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate equals 6 lb/hr) 34.2 x 10 In the event of operation with a core taximum fraction of limiting power density (CMFLPD) Creater than the fraction of rated core thermal power (Core MW Thermal),

44Jb the APRM gain shall be adjusted up to 95% of rated thermal power as follows:

AFRM Reading > 100% x CMFIPD Provided that the adjusted APRM reading does not exceed 100% of rated thermal power and the required gain adjustment incre-ment does not exceed 10% of rated thermal power.

i For no combination of loop recir-culation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power.

Surveillance requirements for CMFLPD are given in Specifica-tion 4.1.B.

Amendment No. 27, 4Z, J2, 18, 67, 73 1.1-2 l

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LIMITING m m t Momt se,a itims

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SAFETY LIMITS 2.1.A.l.d.

APRM Rod Block Trip Setting The APRM rod block trip setting shall be:

S 1 0.66 W + 42%

RB where:

SR3 = R d block setting in

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percent of rated thermal power (2436 MWt)

W = Loop recirculation flow rate in percent of raced (rated loop recirculation flow rate equals 6 lb/hr) 34.2 x 10 In the event of operation with a core maximum fraction of limiting power density (CMFLPD) greater ths.n the fraction of rated core thermal power (Core MW Thermal), the APRM 24J6 gain shall be adjusted up to 95%

of rated thermal power as follows:

APR'i Reading 3,100% x CMFLPD Provided that the adjusted APRM reading does not exceed 100% of rated thermal power and the required gain adjustment incre-ment does not exceed 10% of rated thermal power.

2.

Reactor Water Low Level Scran Trip Setting (LLl)

Reactor water low level scram trip setting (LL1) shall be 3,12.5 inches (narrow range scale).

3.

Turbine Stop Valve Closure Scram Turbine stop valve closure scram trip setting shall be < 10 percen:

valve closure from full open. This scram is only effective when tur-bine steam flow is above 30% of rated, as measured by turbine first stage pressure.

y Amendment No. 19,42, J$. 73 1.1-3

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as:" Fiux Scre ie iet:ise's fh7fn Model'(C6htinued)

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T.e A.:RM .cw ref erenced scram trip settir.g at icil recir:ulatien flow is a: jus abia.: to 120% of rated :ower.

This re::ced f1:w referenced trip se ::in: atti result in an earlier stram durir; slew thermal ransients, suca es tne loss of 80*F feccwater nea-ing event, than.

w:;1d result wi n :ne 120% fixed hign neutron flux scram :rf o.

The

'.catr ficw referenced scram setpoint therefore decreases :he severity

(:CPR) of a sicw thermai transient anc allows lower Operating Limits if such a transient is the limiting abncrmal ccerational transient l

during a cer !in expcsure interval in the cycle.

The A??.M fixed high r.eutron flux signal does not incorporate the time constant, but responds directly to instantaneous neutron flux.

~ ' This scram se: point ' scrams the reactor during fast power increase

ransients if crecit is not taken for a direct (pesition) scram, and also serves to scram the reactor if credit is not taken for the flow referenced scram.

The APRM reading must be adjusted to ensure that the LHGR transient peak is not increased for any combination of CMFLPD and reactor core therral power. The liPRM reading is adjusted in accordance with the fornoia in Specification 2.1.A.l.c., when the CMFLPD is greater than the fraction of reted core thermal power.

Analyses of :ne limiting transients sn:w :nat ne scram a:f ustment is re:uirec te assure MC.R > 1.07 wnen the transien; is initiated from the c: era ing MC?R limit.

. F..' Red Bicek Tri Se

tinc c.

React:r cwsr ievel na.v be varied b.v mcVin,c. c:ntrol rods cr by varyina.

t.e recir:ulation flew rate. The APRM system Or:vides a control rod block e

r:d withcrawal beyonc a civen ;oin; a constant recirculaticn

: rev er.:

.cw ra:e, a*.d :nus :: protec: agains; the cor.ci*.icn of a MC?R iess than

'. 07.

This r:c biock trio se: ting, wr.ich is av::ma:ically variec with recirculation icco ficw rate, prevents an incrsase in -he reactor acwer levei o excessite values cue to control roc wi Sdrawal. Tne' flow -

variebie tric set:inc crovides substantial ma ;in frem fuel c'amage, assumino a s tacv.stit; coeration at the trip setting, ever the entire recircuia:icn ficw range.

The margin to the Safety Limit increases es

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the ficw decreases for the specified trip set:ir.; versus flow relation-sni;; therefore, the worst case MCPR which wcuid cccur during a steady-5: ate coeration is a: ICE: Of ra ed thermal : ewer because of the AFRM rod bicck trip settinc. The actual pcwer dis ributicn in the core is estab-lished by seecified controi red secuences and is menitcred centinuously.

by-the in-core LPRM system. The APRM reading is adjusted to compensate for a CMFLPD greater than the fraction of rated core thermal power.

2.

etc or 'Ja:er L w Level Scram Tri-Se::ine (LLl' The ; rip se::ing f;r low level scram is abcvs :r.e bc tem cf the se: ara ;r

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skirt.

This level is > 14 feet abcve the tre cf the active fuel.

This ievel has been us ed in transien; analyses dealir.; with ::ciant inver. orv

ecrease.

7,e rssul:s reecr:ed in c5AR Sec:i:n ic.3 sh:w -hat 3 scrar,i-

-is 'evel t:e:ua eiy :rc ec s :ne fue! inc :. e :ressure barrier.

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1::r:nicately 33 'nc es :e',ct. :(e ncrra :: era:ina anc.u.

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SURVEH, LANCE REQUIREMENTS IMITING CONDITICMS FOR OPERATION 3.1 REACTOR PROTECTION SYSTDI (RPS) 4.1 REACTOR PROTECTION SYSTEM (RPS)

Applicability Appifcabiliev The Surveillance Requirements asso-The Limiting Conditions for Operation associated with the Reactor Protection ciated with the Reactor Protection System apply to the instru=entation and System apply to the instrumentation and associated devices which initiate associated devices which initiate a a reactor scram.

reactor scran.

objective Obiective The objective of the Limiting Condi-The objective of the Surveillance Requirements is to specify the type tions for Operation is to assure and frequency of surveillance to be the operability cf the Reactor applied to the protection instrumen-Protection Systen.

tation to assure operability.

Specifications Specifications A. Sources of a TTio Signal Which A. Test and Calibration Requirements for the RPS Initiate a Pra: tor Scram RPS instrumentation systems and The instrunentation require enta associated wl:3 each source of a associated systems shall be func-scram signal shall be as given in tionally tested and calibrated as indicated in Table 4.1-1.

Table 3.1-1.

The action to be taken if the number When it is determined that a of operable channels is not met for channel has failed in the unsafe condition, the other RPS channels both trip syste=s is also given in that monitor the same variable Table 3.1-1.

shall be functionally tested B. Core Mariru: 7 action of' Limiting Power immediately before the trip system containing the failure is tripped.

Densi ty (C{FLT3)

The trip system containing the unsafe failure may be placed in If at any time during op'eration it is deterzined by nortal surveillance that the untripped condition during the the APRM readings are not in compliance period in which surveillance with sections 2.1. A. l. c a:d 2.1. A. l. d.

testing is being performed on the action vill be initiated within 15 oth'er RPS channels.

minutes to restore operation to within the prescribed linics.

If CMFLPD is B. Core Maximum Fraction of Limiting Power Densfry (CMFLPD) not reduced te co= ply with the above sectiens within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the APRM readings shall be adjusted to The CMFLPD shall be determined comply with the existing Cd7LPD according daily during reactor power ope-to Specifications 2.1.A.l.c and 2.1.A.l.d, ration > 25% and the APRM readings or reduce thernal power to < 25% within the shall be adjusted if necessary as required by Specifications 2.1.A.l.c next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

and 2.1.A.l.d.

Amendment No. 73 3.1-1

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t EURVIILLASCE REQUIN IS

.........., C...."..I0%5 70A OPERATION 3.1.C RPS Res ense Tice The syste= response time from the opening of the sensor contact up l

to and including the opening of the trip actuator centacts shall not exceed 50 ci111 seconds.

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Amendment No.

73 3.1-2

BASES FOR SURVEILLANCE REQUIREMENTS 4.1.A.

Test and Calibration Requirerents for the RPS (Continued)

Croup C devices are active only during a Fiven portion of the operational cycle. For example, the IRM is active during startup and inactive during full-power operation. Thus, the only test that is meaningful is the one performed just prior to shutdown or startup; i.e., the tests that are per-formed just prior to use of the instrument.

Calibration frequency of the instrument channel is divided into two cate-gories: These are as follows:

1.

Passive type indicating devices that can be compared with like units on a continuous reference.

ii.

Vacuum tube or semiconductor devides and detectors that drif t or lose sensitivity.

Experience with passive type instruments in generating stations and substa-tions indicates that the specified calibrations are adequate. For those devices which e= ploy amplifiers, etc., drif t specifications call for drift to be less than 0.4:/ month; 1.e.,

in the period of a month a drif t of

.4% could occur and still, provide for adequate margin. For the APRM system, drift of electronic apparatus is not the only consideration in determining a calibration frequency. Change in power distribution and loss of chamber sensitivity dictate a calibration every seven (7) days. Calibration on this frequency assures plant operation at or below thermal limits.

The sensitivity or LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate.

This is compensated for in the APRM system by calibrating twice a week using heat balante data and by calibrating individual LPRM's every 1000 effective full power hours using TIP traverse data.

'B.

Maxi ~.:n F* action of L1=iting Power Density OU'LPD)

Since change.s due to burnup are slow, and only a few control rods are moved daily, a daily check of the MFLPD is adequate. The determination of the MFLPD would establish whether or not adjustment of the APRM reading is required.

l Amendment No.

73 3,y_17

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., _2ATIS 707. SURVEILLANCE REQt'IREMENTS 1.0. 7efarences

1. I. M.'Jacobs, " Reliability of Engineered Safety Features as a Tucctien of 196,6, i

Tes;ing Frequency," Suclear Saf ety, Volu:ne 9, No. 4. July-August, pp 303-312.

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linear Hea; Cenerati:n Rate (iHCR)

This specification a ssures that the linear hes: geners:ien rate in any rod is less :han the design linear hea generation if fuel pellet densification is pcs:ula:ed.

The p:ver spike penalty specified fc 7 x 7 fuel is based en the analysis presented in Set:ic

3. 2.1 of Ref erence 4 and Ref erences 5 and 6, and asse=es a linearly increasing variation in axial gaps betseen core botto: and cp, and assures with a 95 confidence, tha t no co:e than one fuel rod exceeds the design linear heat generation rate due to pcue: spiking.

The URGR as a functi n of ccre height shall be checked daily.during reacter operation at.>_ 25 pcVer to deternine if fuel burnup, or centrol rod ::verent has caused changes in power distribution.

For LHGR to be a limiting value below 25% rated thermal power, the ratio of peak LEGR to core average LHGR would have to be greater than 9.6, which is precluded by a considerable margin vhen e= ploying any persissible control rod pattern.

C.

Minitun Critical ?over Ez:f o (MC?R)

The re-uired cperating it:it MC7R as spec.ified in Specificatien 3.11.C is derived f a the established fuel cladding integrity Sef ety Li it MCPR cf 1.07 and an analysis of abnorcal cpera:ional :ransients presented in Refere..ce 7.

even:s vill reduce the. C?R belev :he cperating MC?R.

M "a:1ces ::ansien:

Te ass :e tha :he fuel cladding integri:y safe:y ~.ici: (MC?R of 1.07) is c:f a:ed during a.:icipated abnertal cperatienal ::ansients, the mos:

ac:

liniting ::ansien:s have been analyzed :o de:ernine which cne results in the larges: redue:ica 1: ::itical power ratic (a MC?R).

Addi:icn of the larges:

i MC?R := -he saf ety 12:1: MC?R gives the =inicu cpera:ing li:1: MCPR to av:id ticiatien ef :he saf ety limit should the,res: lixiting transient cecur.

The yge cf ::ansian:s evaluated were less of ficw, inc: case in pressure and p:ver, positive rea :ivity inser icn, and coolan: :e:peratu:e decrease.

2ne ava.ut:icn cf a given transient begins with the sys te: initial parkneters s h r.n in Tab *.e 6-2 :f Ref erence 9 that are inpe: :o a GE core dynamic behavi::

ans' = - -e:;ut e: p::gran described in Ref erence 3.

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.ac.e.<.

I Oce.,:ic;.ents :ha: vere input tc the transient cale.lational,recedure are based a new nethed cf :alculatien terned NEV which p cvides a better agreenen cbe veen the calcula:ed and plan: ins trenent power distributicas.

The outpu:s of this p :gra: alcag vi-h the initial MC?R fer: :he input for further analyses of the :her: ally lini:ing bundle vi:5 -he single channel transient ther:a1 hydraulic f C!.T ccie described in Ref erence 1.

The principal result of this evalua:1:n is the r edue:ica in MC?R caused by the ::ansient.

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he = cst li iting ::ansient fc: the 8 x SR fuel is the less 7:c= IOC4 to ICC4, *.es ting vi:5 a A C?R cf 0.14.

The = cst limiting event threurh-of ICC:7 feedva:e:

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cycle 4 f or S x 3 and 7 x 7 fuel is the Rod " Withdrawal E or (RWE) with a cu:

aC?R cf 0.17 fc: S x I and 0.19 for 7 x 7.

Ther ef::e, the MC?R's specified in 3.ll.C are based. cn 1:ss of 10C F f eedvater hasting and the Rod Withdrawal I :::.

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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 i

EDWIN I. HATCH NUCLEAR PLANT UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.14 License No. NPF-5 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The applications for amendment by Georgia Power Company, et al, (the licensee) dated June 5, 1979, November 8, 1979, and February 28, 1980, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chaptar I; B.

The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Comission; I

C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the i

health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and i

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all, applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amend-ment, and paragraphs 1.A,1.F, 2, 2.A, 2.B and 2.C.(2) of Facility Operating License No. DPR-57 are hereby amended to read as follows:

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1.

A.

The application for license filed by Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, And the City of Dalton, Georgia (the licensees) com-plies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I and all required notifications to other agencies or bodies have been duly made; F.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and the City of Dalton, Georgia are financially qualified to engage in the activities authorized by this operating license in accordance with the rules and regulations of the Commission; 2.

Facility Operating License No. NPF-5 is hereby issued to Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and the City of Dalton, Georgia to read as follows:

A.

The license applies to the Edwin I. Hatch Nuclear Plant, Unit No. 2, a boiling water reactor and associated equipment (the facility) owned by Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and the City of, Dalton, Georgia. The facility is located in Appling County, Georgia, and is described in the Final Safety Analysis Report as supplemented and amended (Amendments 18 through 45) and the Environmental Report as supplemented and amended (Supplements 1 and 2 and Amendment 1).

B.

Subject to the conditions and requirements incorporated herein, the Conmiscion hereby licenses Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and the City of Dalton, Georgia:

C.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.14, are hereby incor-porated in the license. The licensee shall operate the facility,, accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION l

c -- cc-Thomas A. Ip)oli o, hief Operating Re tctors Branch #3

- Division of 2perating Reactors

Attachment:

Changes to the Technical r

Specifications s

[hte of Issuance: April 17,'1980

ATTACHMENT TO LICENSE AMENDMENT N0.14 FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Remove Insert I

I II II 1 -1 1 -1 i

1-2 1 -2 1-3 1-3 1 -4 1 -4 1-5 1-6 2-3*

2-3*

2-4 2-4 2-5 2-5 2-6 (Deleted)

B 2-9 B 2-9 B 2-10 B 2-10 B 2-11 B 2-11 B 2-12*

B2-12*

3/4 2-5 3/4 2-5 3/4 2-6*

3/4 2-6*

3/4 3-1*

3/4 3-1*

3/4 3-2 3/43-2 3/4 3-5*

.3/4 3-5*

3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-8 3/4 3-37*

3/4 3-37*

3/4 3-38 3/4 3-38.

3/4 3-39*

3/4 3-39*

i 3/4 3-40 3/4 3-40 3/4 3-41 3/4 3-41 3/4 3-42*

3/4 3-42*

B 3/4 2-3 B 3/4 2-3 8 3/4 2-4*

B 3/4 2-4*

• 0verleaf page provided for convenience only.

s INDEX DEFINITIONS SECTION

+

f PAGE 1.0 DEFINITIONS 1-1 ACTI0N.....................................................

1 -1 AVERAGE PLANAR EXPOSURE....................................

1-1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.................

1-1 CHANNEL CALIBRATION........................................

1 -1 C H AN N E L C H EC K..............................................

1-1 CHANNEL FUNCTIONAL TEST....................................

1-2 CORE ALTERATION............................................

CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY............

12 l

12 CRITICAL POWER RATIO.................

1-2 DOSE EQUIVALENT I-131.......................

1-2 E-AV ERAGE DI S I NT EGRAT ION ENERGY............................

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSETIME.........

1-2 1-3 l

F RACT ION 0F L IM JT I NG POWER DENS ITY.........................

i 1-3 FREQUENCY N0TATION.........................................

1-3 I D E NT I F I E D L EA KAG E.........................................

1-3 ISOLAT ION SY ST EM RESPONSE TIME.............................

i

\\

LIMITING CONTROL R00 PATTERN...............................

1-3

)

)

LINEAR HEAT GENERATION RATE................................

1-3 1-3 LOGIC SYSTEM FUNCTI0tML TEST...............................

I 1-4 MINIMUM CRITICAL POWER RATI0...............................

i 1-4 OP ERAB L E - O P E RA B I L I T Y.....................................

1-4 OPERATIONAL C0tIDITION......................................

J 1-4 PHYSICS TESTS..............................................

HATCH-UNIT 2 I

' Amendment' No.14 i

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a s

-INDEX DEFINITIONS SECTION PAGE DEFINITIONS (Continued) 1-4 PRESSURE BOUNDARY LEAKAGE.........................'.........

1-5 PRIMARY CONTAINMENT INT EGRITY..............................

RATED THERMAL P0WER...........................'.............

1-5 r

1-5 REACTOR PROTECTION SYSTEM RESPONSE TIME....................

1-5 REPORTABLE 0CCURRENCE......................................

1-5 R00 DENSITY................................................

1-6 SECONDARY CONTAINMENT INTEGRITY............................

1-6 SHUTDOWN MARGIN............................................

1-6 ST AGGERE D T EST BAS 15.......................................

t 1-6 THERMAL P0WER.............................................

I 1 U N I D E NT I F I ED L EA KAG E.......................................

1-7 TAE'_E 1.1, SUR'.'E ILLANCE FRE QUENCY NOTAT ION......................

1 -8 TAELE 1. 2, OPERAT IONAL CONDIT 10NS...............................

4 s

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HATCH-UNIT 2 II Amendment No.14 N

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1.0 DEFINITIONS The following terms are defined so that unifonn interpretation of these specifications may be achieved.

The defined tenns appear in capitalized type and shall be i alicable throughout.'these Technical Specifications.

ACTION ACTIONS shall be those additional requirements specified as corollary l

statements to each specification and shall be part of the specifications.

AVERAGE PLANAR EXPOSURE The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planer height l

and is equal to the sum of the exposure of all the fuel rods in the speci-fied bundle at the specified height divided by the number of fuel rods in the fuel bundle.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable l

to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL CAllBRATION A CHANNEL CAllBRATION shall be the adjustment, as necessary, of the channel l

output such that it responds with the necessary range and accur cy to known values of the parameter which the channel monitors.

The CHANNEL CALISRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

The CHANNEL CALIBRATION may be performed by any series of sequential, c<erlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior l

during operation by observation.

This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument chanr.els measuring the same parameter.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:

l a.

Analog channels - the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPEPAEILITY including alarm and/or trip functions and channel failure trips.

b.

Bistable channels - the injection of a simulated signal into the chan-nel sensor to verify OPERABILITY including alarm and/or trip functions.

HATCH-UNIT 2 1-1 Amendment No. 14 Wa

• DEFINITIONS CORE ALTERATION l

CORE ALTERATI0t; shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the Suspension of CORE ALTERATIONS shall not preclude completion of vessel.

the movement of a component to a safe conservative position.

CORE MAXIttVM FRACTION OF LI;11 TING POWER DENSITY _

The CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY (CMFLPD the largest FLPD which exists in the core for a given operating condition.

CRITICAL POWER RATIO The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the l

assembly which is calculated by application of the GEXL correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT l-131 DOSE EQUlVALENT l-131 shall be that concentration of I-131, uCi/ gram, l

which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, 1-133, 1-134, and I-135 actually The thyroid dose conversion factors used for this calculation present.

shall be those listed in Table 111 of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

T-AVERAGE DISINTEGRATION ENERGY

[ shall be the averace, weighted in proportion to the concentration of l

each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

EMERGENCY CORE C_00 LING SYSTEM (ECCS) RESPONSE TIME l

The EMERGENCY CORE C00LlHG SYSTEM (ECCS) RESPO!;5E Tl!1E shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. ).

Times shall include diesel generator starting and sequence loading delays where applicable.

Amendment t!o.14 1-2 HATCH-UNIT 2

DEFINITIONS FRACTION OF LIMITING POWER DENSITY The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the ratio of local, LHGR for any specific location on a fuel rod divided by the design LHGR.

FREQUENCY NOTATION The FREQUENCY NOTATION specified for the. performance of Surveillance

{

Requirements shall correspond to the intervals defined in Table 1.1.

IDENTIFIED LEAKAGE l

IDENTIFIED LEAKAGE shall be:

Leakage into collection systems, such as pump seal or valve a.

packing leaks, that is captured and conducted to a sump or collecting tank, or b.

Leakage into the contairinent atmosphere from sources that is both specifically located and known either not to interfere with the operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.

ISOLATION SYSTEM RESPONSE TIME The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from wh the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays where applicable.

LIMITING CONTROL R0D PATTERN A LIMITING CONTROL R0D PATTERN shall be a pattern which results in the l

core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE LINEAR HEAT GENERATION RATE (LHGR) shall be the power generation in an l

a aitrary length of fuel rod, usually one foot.

It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL TEST A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all relays and contacts l of a logic circuit, from sensor to activated device, to ensure that components are OPERABLE per design requirements.

HATCH - UNIT 2 1-3 Amendment No.

14

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DEFINITIONS MINIMUM CRITICAL POWER RATIO The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which l

exists in the core.

OPERABLE - OPEr. ABILITY A system, subsystem, train, component or device shall be OPERABLE or I

have OPERABILITY when it is capable of performing its specified function (s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL CONDITION An OPERATIONAL CONDITION shall be any one inclusive combination of mode

{

switch position and average reactor coolant tenperature as indicated in Table 1.2.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental l

nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault l

in a reactor toolant system component body, pipe wall or vessel wall.

5 HATCH - UNIT 2 1 -4 Amendment No.14 p.. e i

\\

r DEFINITIONS PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY shall exist when:

)

All penetrations required to be closed during accident con-l a.

ditions are c.ither:

1.

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2.

Closed by at least one manual valve, blind flange, l

or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification, 3.6.3.1.

b.

All equipment hatches are closed and sealed.

l Each containment air lock is OPERABLE pursuant to l

c.

Specification 3.6.1.3.

d.

The containment leakage rates are within the limits of

{

Specification 3.6.1.2.

The sealing mechanism associated with each penetration; e.g.

l e.

welds, bellows or 0-rings, is OPERABLE.

RATED THERMAL POWER RATED THERMAL POWER shall be a total reactor core heat transfer to the l reactor coolant of 2436 MWT.

REACTOR PROTECTION SYSTEM RESPONSE TIME REACTORPROTECTIONSYSTEMRESPONSETIMEshallbethetimeintervalfrom{

when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

REPORTA8LE OCCURRENCE A REPORTABLE OCCURRENCE shall be any of those conditions specified in l

Specification. 9.1.8 and 6.9.1.9.

ROD DENSITY R00 DENSITY shall be the number of control rod notches inserted as a l

f fraction of the total number of control rod notches.

All rods fully

~

inserted is equfvalent to 100% R0D DENSITY.

l

.d HATCH - UNIT 2

.1 - 5 Amendment No. 14

~..!.....

n.

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JEFINITIONS SEC'ONDARY CONTAltF.ENT INTEGRITY l

SECONDARY CONTAlfF.ENT INTEGRITY shall exist when:

All secondary containment ventilation system automatic isolation. [

a.

dampers are OPERABLE or secured in the isolated position, b.

The Stardby Gas Treatment System is OPERABLE pursuant to l

Specification 3.6.6.1.

At it:ast one door in each access to the secondary containment l

is closed.

The sealing mechanism associated with each penetration (e.g.,

l d.

welds, bellows or 0-rings) is OPERABLE.

SH'JTDOWN MARGIN l

SHUTDOWN MARGIt! shall be the amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68 F; and xenon free.

STAGGERED TEST BASIS STAGGERED TEST BASIS shall consist of:

l A test schedule for n systems, subsystems, trains or a.

other designated components obtained by dividing the specified test interval into n equal subintervals, The testing of one system, subsystem, train or other b.

designated components at the beginning of each subinterval.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate l

to the reactor coolant.

LINIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED l

LEAKAGE.

HATCH - UNIT 2 1-6 Amendment No. 14 4

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I SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMlTING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The reactor protection system instrumentation setpoints shall be 2.2.1 set consistent with the Trip Setpoint values shown in Table 2.2.1-1.

APPLICABILITY: As shown for each channel in Table 3.3.1 1 ACTION:

With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

[

(,

HATCH - UNIT 2 2-3

Y u,.;

.a TABLE 2.2.1-1 5'y REACTOR PROTECTI0tl SYSTEM lilSTRp1Fi!TATI0ti_S_ETP0lNTS, h

M FUNCTI0t!Al. IJillT TRIP SETPOINT ALLOWABLE VALUES 1.

Intermediate Range Monitor, fleutron Flux-liigh

- 120/125 divisions

< 120/125 divisions'

~

(2C51-K601 A,0,C,D,E,F,G.II) of full scale of full scale 2.

Average Power Range Monitor:

(7C51-K605 A B.C,0,E,r) a.

Neutron Flux-Upscale,15%

< 15/125 divisions

< 20/125 divisions l

of full scale of full scale b.

Flow Referenced Simulated Thennal

< (0.66 W + 51%),

< (0.66 W + 54%),

Power-Upsca le with a maximum with a maximum

< 113.5% of RATED

< 115.5% of RATED TPERMAL POWER

~ THERMAL POWER I.

c.

Fixed Neutron Flux-Upscale, 118%

--< 1187 of RATED

-< 120% of RATED m

TiiERt1AL POWER THERMAL POWER 3.

Reactor Vessel Steam Dome Pressure - High

< 1045 psig

< 1045 psig

( 2B21 -!!023 A,B,C,0) 4.

Reactor Vessel ifater Level - Low

-> 12.5 inches above

--> 12.5 inches above instrument zero*

instrument zero*

l (2C21-N017 A,B,C,0) 5.

"tain Steam Line Isolation Valve - Closure

< 107 closed

< 10% closed p

g (t!A) o.

?>

6.

Main Steam Line Radiation - High

-< 3 x full power

-< 3 x full power S

(2011-K603A,B,C,D) background background 7.

Drywell Pressure - liigh

< 2 psig

< 2 psig

( 2C71-fl002A,0,C,D)

^

• See Bases Figure B 3/4 3-1.

4

5 TAP.LE 2.2.1-1 (Continuedl N

7 REACTOR PROTECTIOff SYSTEM IflSTRUMENTATION SETPOINTS a

FUNCTI0t!AL UNIT IRIP SETP0f f!T ALL0HASLE VALUES n,

8.

Scram Discharge Volume Water I.evel - High

< 57.15 gallons

< 57.15 qallons (2C1'-N013A,B C.D)

~

9.

Turbine Stop Valve - Closure

-< 107 closed

-< 10% closed (NA) 10.

Turbine Control Valve 600 psig

, 600 psig l

Fast Closure, Trip Oil Pressure-Low (2C71-N005A,B,C,0) 11.

Reactor Mode Switch in Shutdown flA NA y

Posi tion (NA)

12. Manual Scram flA f!A (flA)

T 2

70 it a,

g

.. n

,'I

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1 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETP0ll;TS The Reactor Protection System Instrumentation Setpoints specified in Table 2.2.1-1 are the values at which the reacter trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are presented from exceeding their Safety Limits.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Alicwable Value is acceptable on the basis that each Allowable Value is equal to or less than the drif t allowance assumed for each ' trip in the safety analyses.

1.

Intennediate Range Monitor, Neutron Flux The IRM system consists of 8 chambers, 4 in ea:h of the reactor trip systems.

The IRM is a 5 decade 10 range instrJrent.

The trip setpoint of 120 divisions of scale is active in each of the 10 ranges.

Thus as the IRM is ranged up to accantodate the increase in power level, the trip setpoint is also ranged up.

The IRM instrJments provide for overlap with both the APRM and SRM systems.

The most sicnificant source of reactivity charges during the power increase are due to control rod withdrawal.

In orcer to ensure that the IRM provides the required protection, a range of red withdrawal acci-dents have been analyzed, Section 7.5 of the FSA?..

The most severe case involves an initial condition in which the reacter is just subtritical and the IRM's are not yet on scale.

Additional corservatism was taken in this analysis by assuming the IRM channel closest to the rod being withdrawn is bypassed.

The results of this analysis show that the reactor is shutdcan and peak power is limited to 11 of P,ATED TFERMAL P0'r.'ER, thus maintaining MCPR above 1.06.

Based on this analysis, the IRM provides protection against local control rod errors and centinuous withdrawal of control rods in sequence and provides backup protection for the APRM.

2.

Average Powcr Range Monitor for operation at low pressure and low flow during STARTUP, the APRM scram setting of 15/125 divisions of full scale neutron flux provides adequate thermal margin between the setpoint and tre Safety Lirits.

The margin accommodates the anticipated maneuvers asso:iated with power 4

plant startup.. Effects of increasing pressure at zero or low void content are minor,and cold water from sources available during startup is not much colder than that already in the system.

Temperature coefficients are small and control rod patterns are constrained by-the RSCS and RVM.

HATCH - UNIT 2 B 2-9 Amendment No. 14

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2.2 LIMITING SAFETY SYSTEM SETTIN35 BASES (Continued)

REACTOR PROTECTION S(STEM INSTRUMENTATI0tl SETPOINTS (Continued)

Average Power Range Monitor (Continued)

Of all the possible sources of reactivity input, uniform control rod witt.-

drawal is the most probable cause of significant power increase.

Be:ause the flux distribution associeted with uniform rod withdrawals does n:t involve high local peaks t nd because several rods must be moved to change power by a significant artount, the rate of power rise is very slow. Ge n-erally the heat flux is in near equilibrium with the fission rate.

In ar.

assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than S% of' RATED THERMAL POWER per minute anc the 3

i APRM system would be more than adequate to assure shutdown before tFe power could exceed the Safety Limit. The 1S% neutron flux trip remains active until the mode switch is placed in the Run position.

The APRM flux scram trip in the Run mode consists of a tiow referenced simulated themal power sciam setpoint and a fixed neutron flux scra, se -

point.

The APRM flow refert.nced neutron flux signal is passed throc;h a filtering network with a time constant which is representative of tre fuel dynamics.

This provides a flow referenced signal that approximates the average heat flux or thermal power that is developed in the core during transient or steady-state conditions.

The APRM flow referenced simulated thermal power scram trip setting at full recirculation flow is adjustable up to 113.5% of RATED THERMAL F04ER.

This reduced flow referenced trip setpoint will result in an earlier scram during slow thermal transients, such as the loss of 100 F feedwater heating event, than would result with the 118% fixed neutron flux scram trip.

Tre lower flow referenced scram setpoint therefore decreases the severi y, aCPR, of a slow thermal transient and allows lower operating limits if such a transient is the limiting abnormal operational transient during a certain exposure interval in the fuel cycle.

The APRM fixed neutron flux signal does not incorporate the time cor.stant, but responds directly to instantaneous neutron flux.

This_ scram se point i

scrams the reactor during fast power increase transients if credit is not taken for a dire:t (position) scram, and also serves to scram the reactor if credit is not taken for the flow referenced simulated thermal power scram.

1 The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility cf unnecessa y shutdown.

The flow referenced trip setpoint or APRM pain l

must be adjusted by the spccified fomula in Specification 3.2.2 in order to maintain these margins when the CHFLPD exceeds the FRTP.

l HATCH - UNIT 2 B 2-10 Amendment No. 14 9

y e

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pg

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 3.

Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to in-crease the power of the reactor by compressing voids thus adding reac tiv ity. The trip will quickly reduce the neutron flux, counte.-

acting the pressure increase by decreasing heat generation.

The trip setting is slightly higher than the operating pressure to permit normal 4

opera *fon without spurious trips. The setting provides for a wide margin to thm maximum allowab'e design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power / flow conditions when the turbine stop valve closure trip is bypassed.

For a turbine trip under these conditions, the transient analysis indicated a considerable margin to the thermal hydraulic limit.

4 Reactor Vessel Water Level-Low The reactor vessel water level trip setpoint was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection for the fuel and pressure barriers.

5.

Main Steam Line Isolation Vrive-Closure The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events.

The MSIVs are closed automatically from measured parameters such as high steam flow, high steam line radiation, low reactor water level, high steam tunnel temperature and low steam line pressure.

The MSIV closure scram anticipates the pressure and flux transients which could follow MSIV closure, and thereby protects reactor vessel pressure and fuel thermal / hydraulic Safety Limits.

6.

Main Steam Line Radiation-Hioh The main steam line radiation detectors are provided to detect a gross failure of the fuel cladding. When the high radiation is detected a trip is initiated to reduce the continued failure of fuel cladding.

At the same time the main steam line isolation valves are closed to limit the release of fission products.

The trip setting is high enough above background radiation levels to prevent spurious trips yet low enough to promptly detect gross failures in the fuel cladding.

HATCH - UNIT 2 B 2-11 Amendment No.

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LIMITING SAFETY SYSTEM SETTING BASES (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 7.

Drywell Pressure-High High pressure in the drywell could indicate a break in the nuclear process systems. The reactor is tripped in order to minimize the possi-bility of fuel damage and reduce the amount of energy being added to the coolant.

The trip setting was selected as low as possible without causing spurious trips.

8.

Scram Discharge Volume Water Level-High 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, control rod insertion would be hindered.

The reactor 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 accommodate the water from the movement of the rods when they are tripped.

9.

Turbine Stop Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves. With a trip ratting of 10% of valve closure from fill open, the resultant increase in heat flux is such that adequate themal margins are maintained even during the worst case transient that assumes the turbine bypass valves remain closed.

This scram is bypassed when the turbine steam flow is below 30% of rated flow, as measured by turbine first stage pressure.

10. Turbine Control Valve Fast Closure, Trip Oil Pressure-Low The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux incre&se that could result from fast closure of the turbine control valves due to load rejection coincident with failures of the turbine bypass valves.

The Reactor Protection System initiates a trip when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than 30 milliseconds after the start of control valve fast closure.

This is achieved by the action of the fast actir.g solenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves.

This loss of pressure is sensed by HATCH - UNIT 2 B 2-12 i

L y

POWER DISTRIBUT10'i LIMITS 3/4.2.2 APRM SETPOINTS LIftITING CONDITIO:t FOR OPERATION I

3.2.2 The APRM flow referenced simulated then::a1 power scram trip set-point (S) and control rod block trip setpoint (SRB) shall be established

• according to the following relationships:

l S 1 (0.66W + 51%)

l S

1 (0.66W + 42',)

RB where:

S a nd S are in percent of RATED THERMAL POWER, and pg W = Loop recirculation flow in percent of rated flow.

APPLICABILITY:

C0l!DITION 1, when THERMAL POWER > 25% of RATED THERMAL POUER.

ACTION:

Eith 5 or S exceeding the allowable value, initiate corrective action gp are within 15 minutes and continue corrective action so that 5 and 5 within the required limits

• within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL P0td to l

less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4 SURVEILLANCE RE0)IREMENTS 4.2.2 The CMFLP3 shall be determined and the APRM flow referenced simulated thermal power scram and control. rod block trip setpoints or APRM readings adjusted, as required.

j a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 1

b.

Whenever THERMAL P0t!ER has been increased by at least 15! of RATED THERMAL POWER and steady state operating conditions have j

been established, and Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is l

c.

operatino with a CMFLPD > FRTP.

• l!ith CORE MAXIMJM FRACTION OF LIMITING POWER DENSITY (CMFLPD) greater L

W than the fraction of RATED THERMAL POWER (FRTP), R up to 95; 'of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or eoual to 100! times CHFLPD, provided that the adjusted APRM reading does not exceed 100t of RATED THERMAL POWER and the' required

. gain adjustment increment does not exceed 10% of RATED THERMAL POWER.

HATCH - UNIT 2 3/4 2-5 Amendment Ilo. 14 L

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core shown in flow, shall be equal to or greater than shown below x the Kf Figure 3.2.3-1.

1.30 up to 6,900 MWD / ton uranium, and a.

b.

1.34 from 6,900 MWD / ton uranium to the end of the first fuel cycle.

APPLICABILITY:

CONDITION 1, when THERMAL POWER > 25% RATED THERMAL POWER ACTION:

With MCPR less than the applicable limit determined from Figure 3.2.3-1, initiate corrective action within 15 minutes and continue corrective action so that MCPR is equal to or greater than the applicable limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less then 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.3 MCPR shall be determined to be equal to or greater than the applicable limit determined from Figure 3 2.3-1:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, Whenever THERMAL POWER has been increased by at least 15% of b.

RATED THERMAL' POWER and steady state operating conditions have been established, and Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is c.

operating with a LIMITING CONTROL R00 PATTERN for MCPR.

HATCH - UNIT 2 3/4 2-6

O 4

3/4.3 INSTRUMENTATION

,. 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 OPERAELE with the REACTOR PROTECTION SYST RESPONSE TIME as shown in Table 3.3.1-2.

Set points and interlocks are given in Table 2.2.1-1.

APPLICABILITY: As shown in Table 3.3.1-1.

ACTION:

With the requirements for the minimum number of OPERABLE channels not a.

satisfied for one trip system, place at least one inoperable channel in the tripped condition within one hour.

Ui+h the requirenents for the minimum n' mber of OPERABLE channels not u

b.

satisfied for both trip systems, place at least one inoperable channel in at least one trip system

• in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1.

The provisions of Specification 3.0.3 are not applicable in OPERA-c.

TIONAL CONDITION 5.

SURVEILLANCE REOUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTION TEST and CHANNEL CAllBRATION operations durine the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1-1.

4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months and shall include calibration of time delay relays and timers necessary for proper functioning of the trip system.

The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip 4.3.1.3 function of 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 logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N tines 18 months where N is the total number of redundar.t channels in a specific reactor trip function.

• 1f both channels are inoperable in one trip system, select at least one inorcrable channel in that trip system to place in the tripped conditions, except when this could cause the Trip Function to occur.

l HATCH - UNIT 2 3/4 3-1 Amendment No. 8 JUL 11979

TABLE 3.3.1-1 h.

REACTOR PROTECTION SYSTEtt INSTRUMEtiTATION "x

APPLICABLE MINIMUM NUMBER b

OPERATIONAL OPERABLE CHANNELS 5

FUNCTIO:.AL UtlIT CONDITI0tlS PER TRIP SYSTEft(a)

ACTION N

1.

Intermediate Range Monitors:

(2C51-K601 A, B, C, D, E, F, G, H) a.

tleutron Flux - High 2(c),5(b) 3 1

2 2

3,4(b) b.

Inopera tive 2, 5 3

1 3, 4 2

2 2.

Average Power Range Monitori (2C51-K605 A, B, C,.D, E, f) e a.

t!eutron Flux - Upscale,15%

2, 5 2

1 1-b.

Flow Referenced Simulated Thermal Power - Upscale 1

2 3

w 4-c.

Fixed Neutron Flux -

Upscale, 118%

1 2

3 d.

Inoperative 1, 2, 5 2

4 e.

Downscale 1

2 3

f.

LPRM 1, 2, 5 (d)

NA 3.

Reactor Vessel Steam Dome Pressure -

High (2021-N023 A, B, C, D) 1,2(e) 2(J,202 N045 5

0)

E 4.

Reactor Vessel Water Level -

2(j, fB2 4-5

]

8 Low (2B21-f!017 A, B, C, 0) 1, 2

s 2B21 -t!025-A,B)

E 5.

Main Steam Line Isolation Valve -

1{7) 4 3

m Closure (ilA) 1, 2 ')

2 6

I 5.

Main Steam Line Radiation - High (201 l-K603 A. II, C, 0) 7.

Drywell Pressure - High 1, 2 2

5 (2C71-N002A,B,C,0)

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION 9 -

In OPERATIONAL CONDITION 1 or 2, be in at least HOT SHUTDOWN l

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

In OPERATIONAL CONDITION 3 or 4, lock the reactor mode switch in the Shutdown position within one hour.

In OPERATIONAL CONDI' TION 5, suspend all operations involving CORE ALTERATIONS or positive reactivity changes'and fully insert all insertable control rods within one hour.

TABLE NOTATIONS 1

A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for a.

required surveillance without placing the trip system in the tripped condition proviced at least one OPERABLE channel in the same trip system is monitoring that parameter, b.

The " shorting links" shall be removed from the RPS circuitry during CORE ALTERATION 5 and shutdown marcin demonstrations performed in accordance with Specification 3.10.3.

I The IRM scrams are autor.atically byp'assed when the reactor vessel mode c.

switch is in ths Run position and all APRM channels are OPERABLE and on scale.

d.

An APM4 channel is inoperable if there are less than 2 LPRM inputs per level or less than eleven LPRM inputs to an APRM channel.

)

These' functions are not required to be OPERABLE when the reactor e.

cressure vessel head is unbolted or removed.

f.

This function' is automatically byoassed when the reactor mode switch is in other than the Run position, This function is not required to be OPERABLE when PRIMARY CONTAINMENT g.

INTEGRITY is not required.

h.

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2.

i.

These functions are bypassed when turbine first stage pressure is

<250* psig, equivalent to THERMAL POWER less than 30% of RATED THERMAL F0WER.

4

/'

j.

Also trips reacter coolant syster recirculation pump MG sets.

~

4 Also trips reactor coolant syster. recirculation pump motors.

L.

• In:cial setpoint.

Final setpoint to be determined during startup testing.

HATCH - UNIT 2 3/4 3-5 Amendment No. 8 gg( 3 $73

TABLE 3._3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES x3 9

RESPONSE TIME FUNCTIONAL UNIT (Seconds) i 5:

H 1.

Intemediate Range Monitors:

N-a.

I:eutron Flux - liigh*

NA f!A b.

Inopera tive

-2.

Average Power Range Monitor:*

Neutron Flux - Upscale,157 NA a.

_ 0.09**

b.

Flow Referenced Simulated Thermal Power - Upscalc 0.09 Fixed Neutron Flux - Upscale,118%

c.

d.

Inopera tive FIA

!!A d.

Inoperative e.

Downscale NA f.

LPRM NA R

3.

Reactor Vessel Stea-ae Pressure - liigh 1 0.SS

[

4.

Reactor Vessel Water Level - Low 1 1.05 E

5.

Main Steam Line Isolation Valve -Closure 1 0.06

~

' 6.

Main Steam Line Radiation - liigh i!A NA 7.

Drywell Pressure - High 8.

Scram Discharge Volume Water Level - Ifigh NA 9.

Turbine Stop Valve - Closure 1 0.06 k

10.

Turbine Control Valve Fast C'osure, 0.08, j

Trip Oil Pressure - Low 1

11. Reactor Ptode Switch in Shutdown Position f!A

.g NA

12. fianual Scram

?

• Neutron detectors are exempt from response time testing.

Response time shall t'e measured from detector output or input of first electronic component in channel.

• • flot including -simulated thermal power time constant.

1. easured from start of turbine control valve fast closure.

M

TABLE 4.3.1-1 REACTOR PROT _ECTION SYSTEM INSTRUMENTATION _ SURVEILLANCE REQUIREMENTS 5-M CPANNEL OPERATIONAL CllANf!EL.

FUNCTIONAL CPANNEL CONDITIONS IN WHICH

_F_U_N_CT_I.ON_A_L_._U.N..I_T CilLCK TES1 CAllBRATION ")

SURVEILLANCE REQUIRED I

czq 1.

Intermediate Range Monitors:

N a.

?!eutron Flux - liigh 0

S/U(b)(c)

R 2

i

~

D W

R 3,4,5 b.

Inoperative NA W

NA 2,3,4,5 2.

Average Power Range Monitor:

a.

Neutron flux - Upscale,157 S

S/U

,WId) S/,UIh)

Id)

Ih)(C)

,W 2

5 W(e)(f)' SA U

S b.

Flow Referenced Simulated S

S/'.'( g,W.

W 1

Thermal Power - Upscale S/U(b) W W(e), SA i

c.

Fixed Neutron Flux - Upscale, S

110" t'

'd.

Inoperative NA W

NA-1,2,5 e.

Downscale NA 9

NA 1

Y f.

LPRM 0

NA (g) 1, 2, 5 u

3.

Reactor Vessel Steam Dome Pressure - High NA.

M..

Q 1, 2 I"

4.

Reactor Vessel Water loud -

Low' D

M Q

1, 2 5.

Main Steam Line Isolation Valve -

R(h) 3 Closure NA M

Main Steam Line Radiation - High D

W(i)

R(d) 1, 2 k

6.

J s

7.

Drywell~ Pressure - High f'A M

Q 1, 2 2

8.

Scram Discharge Volume Water R(h) 1,2,5 Level - High NA M

Z 4

TABLE 4.3.1-l __(Continued)

_R E AC T 0,R PROTJC,T 10f t_1YJ,T FM_I NS_T RllHfgAT 10fl_Si!RV f i l.L AtlC E R EQU I RE,MEfq S, CHANNEL OPERATIONAL

'E CI'AN!!El IJNCTIONAL CHANNEL CONDITIO?!S IN WHICH FUNCTIONAL llNIT CHECK TEST CAllBRATION SURVEILLANCE REQUIRED e

g Ih) 1 9

Turbine Stop Vaive - Closure

!!A M

R

~

10. Turbine Control Valve Fast Closure, Trip Oii Pressure -

NA M

R 1

Low 11.

Reactor Mode Switch in Shutdown flA R

NA 1,2,3,4,5 Position

12. 51anual Scram flA M

t!A 1,2,3,4,5 Neutron detectors may be excluded from CHANNEL CALIBRATION.

a.

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

{

b.

The APRM, IRH and SRM channels shall be compared for overlap during each startup, if not w

c.

5 performed within the previous 7 days.

When changing from CONDIT10t! I to CONDITION 2, perform the required surveillance within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> d.

~

after entering CONDITION 2.

This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL P0t!ER e.

Any APRM channel gain adjustment made Adjust the APRM channel if the absolute difference > 27..

in compliance with' Specification 3.2.2 shall not be included in determining the absolute difference.

p This calibration shall consist of the adjustment of the APRM flow referenced simulated thermal E

f.

k power channel to conform to a calibrated flow signal.

LDRM's shall be calibrated at least once per 1000 effective full power hours (EFPH) using the A

g.

The y

TIP systen.

]

h.

Physical inspection and actuation of switches.

i.

Instrument alignment using a standard current sou' ce.

j.

Calibration using a standard radiation source.

9

INSTRUMENTATION 3/4.3.5 CONTROL R03 WITHDRAWAL BLOCK INSTRUMENTATION LIMITING CONDITIO'i FOR OPERATION 3.3.5 The control rod withdrawal block instrumentation shown in Table 3.3.5-1 shall be DPERABLE with their trip setpoints set consistent with the valuas shown in tne Trip Setpoint column of Table 3.3.5-2.

APPLICABILITY:

As shown in Table 3.3.5-1.

ACTION:

a.

With a control red withdrawal block instrumentation channel trip setpoint less conservative t, tan the value shown in the

]

Allowatie Values column of Table S.3.5-2, declare the channel inoperable until the channel is rest: red to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoir.t value.

b.

With tre requirements for the minimum number of OPERABLE channe".s not satisfied for any trip function, place that trip function in the tripped condition within one hour.

c.

The pr: visions of Specification 3.0.3 are not applicable in 0?EFATIONAL CONDITION 5.

S'JRVEILLANCE REQ'.'IF.EMENTS 4.3.5 Each of t e above required control rod withdrawal block instrumen-tation channels shall be demonstrated OPERABLE by the performance of ine CHANNEL CHEC?, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION cperations durin; the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.5-1.

4 t

6 HATCH - UNIT.2 3/4 3-37 o

. -. _ ~

TABLE 3.3.5-1 CONTROL R00 WITif0RAWAL BLOCK IN5TR11 MENTATION 5

.sM MIfil01M NUMBER OF APPLICABLE OPERAL: E CHANNELS OPERATI0t!AL TRIP FUt!CTION PER TRI? FUflCTION CONDITIONS C'

1.

APRM (2C51-K605 A, B, C, D, E, F) a.

Flow Referenced Simulated Thermal 4

1

~~

Power - Upscale b.

Inoperative 4

1, 2, 5 c.

Downscale 4

1 d.

Neutron Flux - High,127, 4

2, 5 2.

R0D BLOCK MONITOR (2C51-K605 RBM A and B) 1((a )

a.

Upscale 1

1 (a )

b.

Inopera tive 1

a) 1 c.

Downscale 1

m

^

r, 3.

SOURCE RANGE MONITORS (2C51-K600 A, B, C, D) w Detector not full in(b) k 3

2 a.

2 5

Upscale (c) h.

3 2

2 5

Inoperative (c) 3 2

c '.

2 5

Downscale(b) 3 2

d.

2 5

.F IrlTERMEDIATE RANGE MONITORS (d)

~

" " " " ~ ~ " " '

4.

(2C51-K601 A, B, C, D, E, F, G, H)

Detector not full in(*)

6-2, 5 a.

b.

Upscale 6

2, 5 6

2, 5 Innpera tiyg) 6 2

g c.

d.

Downscale 5.

SCRAM DISCHARGE VOLUME (2Cll-fl013E)

.a.

vater iavel uin" 1

1, 2, 5(f)

i TABLE 3.3.5-1 (Continued)

CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION s

NOTE a.

When THERMAL POWER exceeds the preset power level of the RWM and RSCS.

b.

This fanction is bypassed if detector is reading > 100 cps or the IRM channels are on range 3 or higher.

c.

This function is bypassed when the associated IRM channels are on range 8 or higher.

d.

A total of 6 IRM instruments must be OPERABLE.

e.

This function is bypassed when the IRM channels are on range 1.

f.

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2.

HATCH - UNIT.2 3/4 3-39'

TA'3tE 3.3.5-2 C.O.N.i.l.t.0L..1100.W.i.l.ll.h.RA.W.A.L _ltL.O.C. _K_

t!S.i.k.U.ML.il.l.A.l.l.0. N_.S_I:l_PO.I._N.T._S_

2M n

TRIP FU_f.!CTIO_N

_T_R.I.P__S_E_T P.u..l.N.T A_L_LOWABLE VALUE 1.

APRM m

_a.

Flow Referenced Simulated Thermal Power - Upscale

< (0.66 W + 4?7)*

< (0.66 W + 42%)*

b.

Inopera tive fiA NA c.

Downscale

> 3/125 of full scale

> 3/125 of full scale d.

tieutron Flux - High,121.

< 12/125 of full scale

< 12/125 of full scale 2.

R0D BLOCK MONITOR (0.66 H + 41%)

a.

. Upscale

< (0.66W + 41%)

b.

Inoperative liA IIA '

c.

Downscale

> 3/125 of full scale

> 3/125 of full scale

'3.

SOURCE RANGE MONITORS

-w I,

a.

Detector not full in NA NA 5

5 b.

Upscale

- 1 x 10 cps

< 1 x 10 cps c.

Inoperative NA NA d.

Downscale

> 3 ces

> 3 cps 4.

INTERMEDIATE RANGE MONITORS a.

Detector not full in NA flA y

108/125 of full scale e

b.

Upscale

< 108/125 of full scale E

c.

Inoperative fiA fiA d.

Downscale

_ 5/125 of full scale

> 6/125 of full scale i

A 5.

SCRAM DISCHARGE VOLUME F

a.

Wa ter Level-High y 36.2 gallons

< 36.2 gallons

• The Average Power Range Monitor rod. block function is varied as a function of recirculation loop flow (W).

The trip setting of this function must be maintained in accordance with Specification 3.2.2.

O TABLE 4.3.5-1 CONTROL R0D WITHDRAWA_L_pl0CK !!isTR,UMENTATI0fLSURVEILLANCE REQUIREMEf1TS x

li Cl!AriNEL OPERATIONAL E

CHANNEL FUNCTIONAL CHAf!NEL CONDIT10flS IN WHICH Id )

SURVEILLANCE REQUIRED TRIP FuriCTION CHECK TEST CAllBRAT10fl h

1.

APRM N

a.

Flow Referenced Simulated Thermal Power-Upscale NA S/UbI,M R

1 b.

Inopera tive NA S/U

,M NA 1, 2, c M

N I

c.

Downscale NA S/U(h)',M d.

fleutron I lox - liigh, 12~

NA S/i!

R 2, 5 2.

ROD BLOCK MONITOR S/U(b)

R 1(d) a.

Upscale flA S/U((b),M 1((d) b),M NA b.

Inopera tive NA d) j c.

Downscale NA S/U

,g p

3.

SOURCE RANGE f10NITORS a.

Detector not full in f;A S/U

,W NA 2, 5

^-

S/U(b),W W

R 2, 5 h.

Upscale NA u2 1

c.

Inopera tive f!A S/U(b),W f!A 2, 5

~

S/U R

2, 5 d.

Downscale flA 4.

.INTERf1EDIATE RANGE MONITORS S/ti((b)

(c)

NA 2, 5 a.

Detector not full in NA b)

(c)

R 2, 5 b.

Upscale flA S/U

),(c) c.

Inopera t ive f!A S/U

?!A 2, 5 d.

Downscale f!A S/U(b), (c)

R 2, 5 g

{

5.

SCRAf1 DISCHARGE VOLUME 1, 2, 5 ')

I

{

a.

Water Level-High NA Q

R s

l

o 4

+

TABLE 4.3.5-1 (Continued)

CONTROL R00 WITHDRAWAL BLOCK INSTRUMENTATION SURVEILLANCE i

NOTES:

Neutron detectors may be excluded from CHANNEL CALIBRATION.

a.

i b.

Within 24 houro prior to startup, if not performed within the previous 7 days.

I When changing from CONDITION 1 to CONDITION 2, perform the required surveillance within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering CONDITION 2.

i c.

When THERMAL POWER exceeds the preset power level of the RWM d.

and RSCS.

e.

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2.

.i i

i w

1 n-i 4

HATCH - UNIT 2 3/4.3-42 i

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 APRM SETPOINTS 3

The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER.

The scram setting and rod block functions of the APRM i

instruments or APRM readings must be adjusted to ensure that the MCPR does not become less than 1.0 in the degraded situation.

The scram settings and rod block settings or APRM readings are adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and CMFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.06, and an analysis of abnormal operational transients.

For any abnormal operating transient analysis evaluation with the initial condition of the re' actor being at the steady state operating litit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting as given in Specification 2.2.1.

To assure that the fuel cladding integrity Safety Limits is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR)

.The type of transients evaluated were loss of flow, increase -in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient which determines the required steady state MCPR limit is the load rejection trip with failure of the turbine bypass.

This transienteyfelds the largest a MCPR.

When added to the Safety Limit MCPR of 1.06 the required minimum operating limit MCPR of Specification 3.2.3 is obtained.

HATCH - UNIT 2

. B 3/4 2-3

. Amendment No. 14 l

4

,8

t P0MER DISTRIBUTIO*! LIMITS BASES MINIM.!M CRITICAL POWER RATIO (Continued)

The evaluation of a given transient begins with the system ini-f al parameters shown in FSAR Table 15.1-6 that are input to a GE-core dynamic behavior transient computer program described in NED0-10802(3)

Also, the void reactivity coefficients that were input to the trans'ent calculational procedure are based on a new method of calculation ter ed flEV which provides a better agreement between the calculated and plant instrument power distributions.

The outputs of this pro, ram along with the initial MCPR form the input for further analyses of

..ie thermally limiting bundle with the sing channel transient thermal hydraulic SCAT The principal result of this evalt.ation code described in NE00-20566 is the reductier,in MCPR caused by the transient.

The purpose of the K factor is to define operating limits at cther g

than rated flow conditions.

At less than 100% of rated flow the re:; ired MCPR is the product of the operating limit MCPR and the K, factor.

Specif-ically, the K factor provides the required thermal margih to prote:t agair.st a flow increase transient. The most limiting transient ini:iated from less than rated flow conditions is the recirculation pump speed up caused by a motor-cenerator speed control failure.

factors For operation in the automatic flow control mode, the K7 assure that the operating limit MCPR of Specification 3.2.3 will no: be violated should the r.ost limiting transient occur at less than rate: flow.

factors assure that the Safety In the manual flow control mode, the Kf Limit MCPR will not be violated should the most limiting transient cccrr at less than rated flow.

The K factor values shown in Figure 3.2.3-1 were developed generically f

and are applicable to all BWR/2, BWR/3 and BWR/4 reactors. The Kf facters were derived using the flow control line corresponding to R'TED THER!GL POWER at rated core flow.

factors were calculated such For the manual flow control mode, the Kf that for the maximum flow rate, as limited by the pump scoop tube sst point,.

and.the corresponding THERMAL POWER along the rated flow control li e, the limiting _ bundle's relative power, was adjusted until the MCPR was s'.icntly above the Safety Limit. Using this relative bundle power, tne MCPP.s were calculated at different points along the rated flow control line co respond-ing to different_ core flows.

The ratio of the MCPR calculated at a civen point of core flow, divided by the operating limit MCPR, determines the K.

f HATCH - UNIT 2 B 3/4 2-4

.c i

.a " * :

.