ML18026B193

Amends 104 & 77 to Licenses DPR-52 & DPR-68,respectively, Revising Tech Specs to Change Neutron Flux Trip Setting Adjustment Factor from Limiting Safety Sys to Limiting Condition for Operation w/6 H Action Statement

ML18026B193
Person / Time
Site:	Browns Ferry
Issue date:	08/17/1984
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Tennessee Valley Authority
Shared Package
ML18026B194	List: ML18026B193 ML18026B195
References
DPR-52-A-104, DPR-68-A-077 NUDOCS 8408310431
Download: ML18026B193 (64)

Text

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t UNITED STATES t

NUCLEAR REGULATORY COMMlSSION WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING. LICENSE Amendment No. 104 License No.

DPR-52 1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated September 21, 1981 as supplemented June 3,

1982, complies wi-th the standards and requirements of the Atomic Eneroy Act of 1954, as amended (the Act), and, the Commission's rules and regulations set forth in 10 CFR, Chapter I; B.

The facility will operate in conformity with the application, 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 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 i'ssuance of this amendment is in accordance with 10 CFR Part 51 of the Commission'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 amendment and paragraph 2.C(2) of Facility Operating License No.

DPR-52 is hereby amended to read as follows:

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 104, are hereby incorporated in the license.

The licensee shall'operate the facility in accordance with the Technical Specifications.

B008310434 840827

,PDR,ADOCK 05000260 PDR

J

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION b">'ttachment:

Changes to the Technical Specifications Date of Issuance:

Augus,t 17,,

1984 Domenic B. Vassallo, Chief Operating Reactors Branch 82 Division of Licensing

'I

ATTACHMENT TO LICENSE 'AMENDYiENT NO. 104 FACILITY OPERATING LICENSE NO.

DPR-52 DOCKET NO. 50-260 Revise Appendix A as follows:

1.

Remove the following pages and replace wi'th identically numbered pages.

ii, 9, 10, 16, 21, 23, 31, 47, 48, 74,

160a, 169, 169a (new page) 2.

The marginal lines on these pages denote the area being changed.

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..Section D.

Reactivity Anomal;ies Paae Ho.

125 3.4/4.4 St A.

E.

Reactivity Control F.

Sc=c= Discharge Vol~e andby Liquid Control System Normal System Availability.

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126 126 135 135 B.

0peration with Inoperable Components...'.

Sodium Pentaborate Solution 136 137 3.5/4.5 Core and Con.ainment Cooling Systems A.

Core Spray System B.

Residual Heat Removal System (RHRS)

(LPCI and Containment Cooling)

C.

RHR Service Mater System and Emergency Equipment Cooling Mater System (EECMS)

D.

Equipment Area'oolers E.

High Pressure Coolant Injection System (HPCIS)

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F.

Reactor Core Isolation Cooling System

( RCICS)

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G.

Automatic Depressurization Sys em (ADS )

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H.

Maintenance of Filled Discharge Pipe I.

Average Planar Linear Heat Generation Bate J.

Linear Heat Gen ration Rate 143 145 151 154 156 157 158 159 V& ~

L.

H.

Minimum Cr',tical Power Ratio (HCPR)

APRN Setpoints 0

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Reporting Requirements 160

..160A 160 A

3. 6/4.6.

Primary System Boundary A.

Thermal and Pressurization Limitations

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1?4 B.

Coolant Chemis ry 176 Amendment No.

, 104 '~

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-GRITTY h.

c.

For no conbination of loop rec ircu-lation flou rotc and core thecal h

1 the APRH flux scram triF ZOX i c ting be allovcd to cxcced 3.

~of rated thermal pouer.

(Hote: These ser tings assume operat.ion within the basic the~1 hydraulic design erin. These criteria are

>QQR - 13 ~ 4 4~Ift for 8x8',

d P8x8R, Pnd HCPR uithin lioits of Spec fication 3 ~5. k. Ii is determined hat c Lthes of these n crfter.'a is being v'ol.-.cd des. tn during operation

~ action shal1 bc initia ed ufthin 15 rinu cs to rcs:ore oper=tion uithin prcscribnd ]i~its Surveillance require" ents for JCp~

scram setpoint are given in specification CD 3..8.

d.

The aPRN Bod block "-ip setting shal.l be:

Sp vhere:

(0 ~ 65M +<2)

Rod -b}.ock setting in percent of rated the r ma 1 poser (3293 MMt) l.oop recirculation f3.ov rate in percent of rated (rated 3.oop recirculation flow rate equals

34. 2 x

'lOa lb/hr)

Alsendaent i<o. >2, p$,

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=U-L CLABBER?'G I??TSGRITY

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FU-L CLADQTNG 74 GQ<>'g 2 ~

Reactor Pressure

~800 PSIh or Gore Flow -10X of rated.

e.

2.

Eixed High Neutron Flux Scram Tr9.p Set'ting When the mode switch is in the RUN position, the APRH fixed high flux scram trip setting shall

, be:

Sc120X power.

'PRM'nd IPA Trip ScttinFs (Startvp and Hot Standby Modes).

?'hen the reactor pressure is "800 PSLA or core flow is -10K of rated, the core thermal power shall not exceed 823 MWt ('25Z.of rated thermal power)

~

a.

APRM When the reactor mode sw'tch is in the STAR~i position, the APRM scram shall be se" at less than or equal to

'15% of, rated pare b.

XRM-"The ZRM scram shall be set at less.

than or equal to 120/125 of full scale.

n< ='1t No. gg, P,R,

]P4

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.3,. 1.

BASES Because the bailfng transition carzclntfon fs based on n la"ge qunnt'ty az ful) scale data there fs n very high confidence that operation af a fuel assembly at the condition of. MCPR'~1.07 would not produce boilfng tran sition.

Thus, although, it is not required to estab1ish the safety ]f-it nddfcfannl zeargfn exists betvcen the safety lfnft and tho actual'ccurencc

.of lass of cladding integrity.

Hovever, ff boflfng tznnsition vere to occur, clad perforation would not bc expected.

Clodding temperatures would inc.ease to cppraxizrntely 1100 F vhfch is below*the perforation temperature of the cladding z4aterfal.

This has been veziffed by tests in thc General i3ectrfc Test Rcactoz (Gr~) vhcre fuel sinf)ar in design to ERiP operated above the critical heat flux for a significant period of tine (30 minutes) without clad perfaration.

t If reactor pressure should ever exceed 1400 psia during zza= D, paver aperntfng (the Unit of applicability of the boi3,ing trzLnsVion corre-lation). it vould be assed that the fuel c1adding intc~ty Safety ~t has been violated.

At pressures belov 800 psia, the core elevation pz'essure drop (0 paver, 0 flov) is greater than 4.56 psi.

At lov powers and flcrm this prcssure differential is maintained in the bypass region of the core.

Since the pressure drop in the bypass region is essentially all elevation head;.

the core prcssure drop at lov powers and flov vi11 alvztys be greater than 4.56 psi; Analyses shov thct with a flov of 28XlO~ 1bs/hr bundl'e flov, bund1e pressure drop is near3y independent of bundle power and has a value of 3.5,psf.

Thus, the 'aund1e flov vith a 4.56 psi d"iving head vill be greater than 28xl03 lbs/hr.

'Full scale ATLAS'est'ata taken at.presr<ucs f."om 14.7 psia to 800 psia indicate that thc fuc3. assembly crfticnl,pover nt this flov is.approx&ately 3.35 k%t.

Vith the design peeking factors thfs corresponds. to a core thermal pav r oz 4 occ than 50<..

Thus, n core therzzal power l~t of.25$ for reactor pressures below 000 psfn is conservative.

ic r thc fuel in the core. during periods vhcn the reactor is shut davn,,caa-oidcrntfon N4ust also bc given to vntcr level requfrcztcnts duc to the effect nf decay heat. If water level should drop belov the top of the fuel duz'fng thi" tftme, the nbflfty'to remove decay heat is reduced.

Thisercductfon in conlfng capahflf ty cauld lead to elevated cladding temperatures end clad fed faratfan.

As long as thc fuel remains covered with vatcre sufficient ccc ling is available to prevent fuel clad perforation.

Amendment Nn. ~ g, g Q

. 104

0

2.1.

BASES Analyses of the limiting transients show that no.scram ad)ustment is required to assure MCPR A 1

07 when the transient is initiated from MCPR limits specified in specification 3.5.k 2 ~

APRM Flux Scram Tri Settin

.(Refuel or Start

& Hot Standby Mode)

For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power.provides adequate thecal margin between the setpoint and the safety limit, 25 percert of rated.

The margin is adequate to accommodate anticipated maneuvers associated with power plant startup.

Effects of increasing pressure at zero or low void content are minor, 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 to be uniform by operating procedures backed up by the rod worth minimizer and the Rod Sequence Control System.

Thus, all of possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power r'se.

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks,'and because several rods must, be moved to change power by a significant percentage of rated

power, the rate of'ower rise is very slow.

Generally, the heat flux is in near equilibrium with the fission rate.

In an assumed uniform rod withdrawal approach to the scram leve, the rate of power r ise is no more than 5

percent of rated power per minute',

and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit.

,,The 15 percent APRM scram.remains active until the mode switch is placed in the RUN position.

This switch occurs when reactor pressurer is greater than 850 psig.

3, IRM'lux Scram Trio Setting The IRM System consists of 8 chambers, 0 in each of the reactor protection system logic channels.,

The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM.

The 5

decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size.

The IRM scram setting of 120 divisions is active in each range of the IRM.

For Amendment No. +P, P4, 1BA

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5 AS I.S

~'rnn (uel daasage

~ >1ssuckng a steady-stare orerat ion at the eriP setting> over the ent fre reef reufatf on C losd range.

Tl>r >Dartfn ro the Safety Lflafr increases

.ae 'the Clov decreases Cor the spl.c.f Cled trip setting versus (lov rc latfonsl>ip; therefOreo the SSOrSt enSe tfCPR Sdhfeh eOuld.OCCur duri>>R S eady-State OPerat iO>>

1$

a t }OAI oi rated chere>a l rove r because o C the APRON rod b lork c rip set t inc.

Tl>e acf Ua1 f>ol'cr disc ribvt io>>

1>>

1 >e core 1 s cs tab l i shed l>y specified eoo>trol rod seqvvnees and ii-Dsonfrored continuously by the fn-core LPRH sysren.

D. S<<cc r o c r So I

I Sc o

od I ol.ll

(~ro I >colo.>r II c)

The set point for the los> level serm is above the bor too o: the separaror sl.frt.

This level has been used fn transient analyses de>'.ling vi ch coolant. fnvenrory decrease.

The resul:s.reported in FSAR subseet1on 14:5 sbcsd that scraa>

and fsolat fan of all process 1fnes (except e>afn stem) at this level adequately pro(ects the Cue) and the pressure barrier, because MCPR is greater than 1.07.fn all cases,,

and systeo pressure does not reach the safety valve settings.

Tl>e scrav sett i>>g fs approxfesat<<ly 31 inches below the'ore>al operating range and is Lhuc

~ adequate ro avofd spurious ser~

D cori I ~V>l c 'r Scr 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 setting of 10",of valve closure fromm fu11 open, the resultant increase in heat flux is such that 'adequate thermal margins are, maintained even during tho worst cas'e trarsient that assumes the turbine bypass valves remain closed.

(Reference 2)

E.

Turbine Control Valve Fast Closure or Turbine Tri Scram Turbine control valve fast closuzc or turbine trip scram anticipates the

pressure, neutron flux, and heat fl>>x increase that cof>ld result from control val.ve fast closure due to load rejection or control valve closure due to turbine trip; each without bypass valve capahHily.

Thc ro>>cfor protection system initiates a scram in less tha>>

30 milliseconds atter the start of contzol valve fast closure due to load rejection or con ro) valve closure due to turbine trip.

This scram is achieved by rapidly reducinrf hydraulic control o! 1 pressure't the main turbine control valve actuator disc durp valves.

This loss of pressure iq srnsed by pressure switches whose contacts form the one-out-of-tw6-twice logic input to the reactor protection system.

This trip scttint>,

a no>ninally 50" qreatpr closure tine and a diffe> ant valve characteristic from that of the,turbine stop valve, combine t<>

produce transient:

very similar to that for the " top valve.

ho signi fi-cant cha>>ge in HCPR occurs.

Relevant transient analyses are discussed in Rtferences

? and 3 of the Final Safety Analysis Report.

This scram is bypassed wi>pn turbine steam flow is below 30,", of rated, as measurfd by turbine first state pressure.

23

II

LX.IYTII!CCO'.1DZTZO,IS FOR OPI'.RATION SUIVPEZ;I;LANCE Rl'.OUIREHENT!>

3.e 1 REACTOR PROTECTEON SYSTEH 4.1 REACTOR PJ(OTECTION SYSTEM

~

A oli.cabi.litv A olicabilitv Applies to the instrumentarion and associated devices which iniriate a reactor scram.

Applies to the survei.llance of the instrumentation and asso-ci.ated devices which initiate reactor scram.

~Ob eeeXve

'o assure the operability of the reactor protection system.

~Ob ective To specify the type and frequ ncy of surveillance to be applied to the protection instrumentation.

Soecif ication Soecif ication When there is,fuel in the vessel, the setpoints, minimum number of

'trip 'systems, and minimum number

'of instrument channels that must be, operable for each position of the reactor mode switch shall be as given in Table 3.1.A.

A.

Instrumentation systems shall be functionally tested and calibrated as indicated in Tables 4.1.A and 4.1.3 respec-tively.

C.

Vhen it is determined that a

channel is failed in the cc:safe condition, the othe'r RPS channe that monitor the same var'able shall be functionally tested immediately before.the trip sys tern containing the failure is tripped.

The trip system con-taining the unsafe failure may

'n tripped for shor t periods c f time to allow functional.te ti..

of the other trip system.

The trip system may be in the

~ untripped position for no morc.

than eight hours per functiorx~l test period for this testing.

31 Amendment No. P, 104

4l

The frequency o f calibration nf the A>RN F)ow Biasing Network has been established as each zefuelinw outage.

There are s'everal instruments w'hi'ch muse be calibrated and it will cake sevezal hours co perform che calibration of the entire network.

While the calibration is being per-

formed, a aero flow signal will be sent to half ot the APR'.f's resulting ln a half scram and rod block condition.

Thus, if the calibration were pezformcd duzing operation, flux shapini; would noc be possible.

Based on experience at other generating stations, drift of instruments, such as those in the Plow Biasing Network, is noc significant and therefore, to avoid. spurious

scrams, a calibration fzequency of each refueling out-age is established, Group (C) devices are.active only during a given portion of the opera'-

tional cycle.

Por example, the IRM is active duzing staztup and inactive during full-power operation.

Thus, the only test that is meaningful is the one performed )ust prior to shutdown or staztup: i.e., the tests that 'aze performed )use prior to use of the instrument.

Calibration fzequency of the instrument channel is divided iaeo t<<o

.groups.

These are as follows:,

1.

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

2.

Vacuum tube or semiconductor devices and deeectors that drift oz 1Dse sensitivity ~

Pxperience with passive type'nstruments in generating scaeions and sub-stations indicates that the specified calibrations aze adequace.

Por d

i' which em loy amplifiers, etc., drift specifications,ca11 f

d ift to be less than 0.4X/month; i.e., in the perio o

a mon a

drift of 4Z would occur and thus providing for adequate mazgi or z

o the hPRM system drift. of electronic apparatus is not the only considera-

.tion in determining a calibration frequency.

Change in power distrihu-t on an oss o

c a

i d 1 s of chamber sensitivity dictate a calibration every seven at or below days.

Calibzation on this frequency assures plant opezation a

thermal limits.

A comparison of Tables 4.1.A and 4.1.3 indicates that two instrument channels have been included in the latter table.

These are:

mode

'wicch in shutdown and, manual scram.

All of the devices or sensors d with these scram funceione are simple on-off switches

and, hence, calibration duzing operation is not applicable,

.c.,

is either on or off.

47 Amendment No. pf, ]pg

Ck

4. 1 'ASES

~ $

c The sensitivity of LPHM detectors decreases with exposure to neutron flux at a slower and approximately constant rate.

The APRM system, which uses the LPRM readings to detect a change in thermal power, will be calibrated every seven days using a heat balance to compensate for this change in sensitivity.

The HBM system uses the LPRM reading to detect a localized change in thermal power.

Zt applies a correction factor based on the APRM output signal to determine the percent thermal power and.therefore any change in LPHM sensitivity is compensated for by the APHM calibration.

The technical specification limits of CNFLPD,

CPR, and MAPLHGR are determined by the use of the process computer or other backup metnods.

These methods use LPRM read'ngs and TIP data to determine the power distribution.

Comoensat'n in the process compute" for changes in LPHM sensitivity will

.be made by performing a full core TIP traverse to uodate the computer calculated LPRM correction factors every 1000 effective full power hours.

As a.minimum the individual LPHM meter readings will be adjusted at the beginning o; each oper ating cycle before reaching 100 pere'ent power.

Amendment No,

, 104

'1p

noe e..m r vn enOA i'i

~ 4 ~ t The min~~~aumber

.of ope. able 'channels fo~ach trip detailed M>the startup and run positionsW-the reactor mode selector switch The SRH, IRM, and APRM (startup node),

blocks e~

need not be operable in "run" mode, and the APRM (flow b'ased) rod blocks need-not be operable in "star tup" mode.

Pith the number:of'PERABLE channels less than r quired by the minimum OPERABLE channels pa@ trip function requirement, place at least one inoperable channol in th'e tripped condition within one hour.

2 ~

8 ~ the recirculation loop flow in percent of'esign.

Trip level setting is in percent of rated power (3293 Mtlt).

3 0 e

IRM downscale is bypassed when it is on its lowest range.

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A m>eence r 8

e e v

t A 't, I m's sew SRM's A and C downscale functions are bypassed when IRM's A, C, E, and G are above range 2.

SRM's B and D downscale function is bypassed when XRM's B, D, F, and P are above range 2..'"."

SRM detector not m~

startup position is bypassed when the count rate is +)00 CPS or the above condition is satisfied.

5.. During repair or. calibration of equipment, not more than one SRM or RBM channel nor more than two APRM or IRM channels may be bypassed.

Bypassed channels are not counted as operable channels to meet the minimum operable charm 1 requirements.

Refer to section 3.10.B for SRM requirements during core alterations..

6 ZRH channels A, 'E, C,

G all in range "8 or above bypasses SRM channels A'nd C functions.

ZRM channels B, F, D,

H all in range 8 or above bypasses SRM channels B and D functions.

7.

The following operational restraints apply to the RSM only.

a.

Both RBM channels are bypassed when reactor power is ~ 30~

and when a peripheral control rod is selected.

b The RBM need not be operable in the "startup" position of the reactor mode selector switch.

Cm Two RBM channels are provided and only one of these may be bypassed from

.the con~le If the inoperable chuuxel cannot be restored with&

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the inoperable channel shaU. be placed in the tripped;condition within one hoar.

d. If minimum conditions for Table 3.2.C are not met, administrative controls shall be immediately imposed to prevent control.rod withdrswal Amendment No. $6, g,. g P

$4 ~

p4, ~l, l04

>0

)I

Limitinz Conditions for One'ration Surveillance Recuirements 3.5 Core and Containment Coolin Systems 4.5 Core and Containment Coolin S stems L.

APRM Setooints 1,

Whenever the core thermal power is> 25~~ of rated, the ration of FRP/CMFLPD shall be

> 1.0, -or the APRM scram and rod block setpoint equations listed in sections 2.1.A and 2.1.B shall, be multiplied by FRP/CiifFLPD as follows:

FRP/CMFLPD shall be determined daily when the reactor is > 25$ of rated thermal power.

Zc (0.66'W + 54Z)

QELPD (0.66W+ 42Z)

(

)

RP 'LP9 2.

When it is determined. that 3.5.L. 1 is not being met, hours is allowed to correot the condition.

3.

Zf 3.5.L. 1 and 3.5.L.2 cannot be met, the reactor power shall be reduced to < 25$ of rated thermal power within 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, M.

Reoortin Re uirements Xf any of the limiting values identified in Specifications 3.5.I, J, K, or L.3 are ex-ceeded and the specified remedial action is taken, the event shall be logged and reported in a 30-day written report.

160A Anendment No.

104

ig il

3.5.J; Linear Heat Generatio te (LHGR)

This specifica ion ass es that the linear heat generat n rate in any rod is less than the design Linear heat generation if fuel pellet densification is postulated.

The LHGR shall be checked daily during reactor operation at + 25$

,power to determine if fuel burnup, or control rod movement has caused changes in power distribution.

For LHGR to be a.Limiting value below 25+ rated thermaL power, the R factor would.have to be less than 0.241 which is precluded by a considerable margin. when employing any 'permissible control rod pattern.

3.5.K. Minimimum Cr'ScaL Power Ratio (MCPR)

At core thermal power levels less than or equal to 25$, the reactor will be oper ating at minimum recirculation pump speed and the moderator void content wiLL be very smaLL.

For all designated control rod patterns, which may be employed at th's po'nt, operating pLant experS.ence and thermal hydraulic anaLysis indicated that the result'ng MPCR value is in excess of requirements by a considerabLe mar gin.

With this low void content, any inadvertent core flow increase wouLd only place operati.on in a more conserative mode reLative to MCPR.

The daily requirement for caLculating MCPR above 25$ rated thermal power is suff'cient since power distribution shifts are very slow when there have not been significant power or controL rod changes.

The requirement for

. calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known foLLowing a change in power or power shape (regardless of magnitude) that couLd place operation at a thermal limit.

3.5.L Operation is constrained to a maximum LHGR of 18.5 kM/ft for 7x7 fuel and 13.4 kN/ft'or 8x8,

SxSR, and.PSxSR.

This limit is reached when core maximum fraction of lim'ting-power density (CMFLPD) equals 1.0.

For the case where Ci<ZLPD exceeds the fraction of rated thermal power, operation is permitted only at less than 100-percent rated power and only with APRM scram settings as required by specification 3.5.L. 1.

The scram trip setting and rod block trip

.setting are adjusted to ensure that no combination of CMFLPD and FRP will increas the LHGR transient peak beyond that allowed by the 1-percent plas ic strain limit.

A 6-hour time period to achieve this condition is justified since he additional margin gained by the setdown adjustment is above and beyond that ensured by the safety analysis.

169 Amendment No 104

Q

BASES 3 ~

Reoortinz Reouirements The LCO's assoc ated with mon'toring the fuel rod operating conditions ar required to be met at all times, i.e., there is no allowable tim in which the plant can knowingly exceed the "limiting values for

MAPLHGR,

'HGR, and MCPR.

Xt is a requirement, as stated in ~pecification 3.5.i, J,

and K, that if at any time during steady state power operation it is determined that the limiting values for MAPLHGR, LHGR, or MCPR are

exceeded, action is then initiated to restore operation to within the prescribed limits.

Th's action is initiated as soon as normal surveillance indicates that an. operating limit has been reached.

Each event involving steady state operation beyond a specified limit shall be reported

<<t&n 30 days.

It must be recognized that there is always an action which would return any of the parameters (MAPLHGR, LHGR, or MCPR) to within orescr'bed limits, namely power reduction.

Under most circumstances, this will not be the only alternative.

"II~

1.

"Fuel Densification Effects on General Electric Boiling 'later Reactor 8~el,"

S polements 6, 7, and 8, NElM-10735, August 1973.

2.

Suplement 1 to Technical Report on Densification of General Electric Reactor Fue's, December 14, 1974 (USA Regulatory Staff).

3.

Communication:

V.

A'. Moore to T-.S. Mitch'ell, "Modified GE Model for Fuel Densif'cation,"

Doc!<et 50-321, Mar ch 27~,1974 Gen r'c Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.

Letter fro=

.=

Such:-.olz (G"-) to P.

S Check (NRC),

"R

),

. esponse to NRC reauest fc-.

=-.=c....etior-c" C:"".Z computer model," Se~t~oer 5,

1980.

169A Amendment No.

104

%AS REGui

+>>*<<+

t UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON," O. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BRO'I"NS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 77 Li,cense No. DPR-68 1.

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

A.

The application for amendment by Tennessee-Valley Authority (the licensee) dated September 21, 1981 as supplemented June 3, 1982, complies 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; B.

The faci,lity will operate in conformity with the application, 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 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'.

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.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph

.2.C(2) of Facility Operating License No. DPR-68 is hereby amended to read as follows:

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

77, are hereby incorporated in the license.

The 1'icensee shall.operate the facility in accordance with the Technical Specifications.

lpga 0

J

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes. to the Technical Specifications Date of Issuance:

August l7, 198'4 Domenic B. Vassallo, Chief Operating Reactors Branch b2 Division of Licensing

<Qi

ATTACHMENT TO LICENSE AMENDMENT NO.

77 FACILITY OPERATING LICENSE NO.

DPR-68 DOCKET NO. 50-296 Revise Appendix A as follows:

1.

Remove the following pages and replace with identically numbered pages.

i,i, 10, 11, 16, 20, 22, 31, 46, 47, 77,

167a, 177, 178 2.

The marginal lines on these pages denote the area being changed.

i

Section

3. 514.5 C.

Scram Tnsertion Times D.

Reactivity Anomalie E.

Reactivity Contr~i F..

Scram Discharge Volume Standby Liquid'Control System A.

Normal System Availability B.

Operation with Inoperable Components C.

Sodium Pentaborate Solution Core and Containment Cooling Systems A.

Core Spray Syst: em B.

Residual Heat Removal system (RHRS)

(LPCX and Containment Cooling)

C.

RHR Service Water System and Emergency Equipment Cooling Water S ys tern (EECWS )

D.

Equipment Area Cooler =

E.

High Pressu e Coolant Injection System (HPCXS)

F.

Reactor Core Isolation Cooling System (RCICS)

Pave No.

128 129 1 2.9 I29 137 137 139 139 146 146 149 155 158 159

'60 G.

Automatic Depressurization System (ADS) 1 6'1 H.

Maintenance of Filled Discharge Pipe T..

Averaqe Planar Linear Heat Generation Rate 163 165'.

Linear Heat Generation Rate 166

3. 6z4. 6 K.

Minimum Critical Po~er Ratio (HCPR)

L.

APR'4 Setpoints Report nq Requirements Prima ry Sy s tern Boundary A.

Thermal and Pressurization Limitations 167 167>

167'84 184 amendment Ho.

iO

SAFETY LIMIT LIHITING SAFETY SYSTEH SETTING 1.1 FUEL CLADDING INTEGRITY 2e 1 FUEL CLADDING NTEGR TY Q

> Loop recircu-

. lation flov rate in per-cent of rated (rated loop recirculation floe rate equals 3oi2xl0 ~ 1hlhr) 10 For no combination of loop recirculation flo~ rate and core thermal po~er shall the APRH flux scram trip setting be allotted to exceed 120% of rated thermal pover.

(NOTE:

These settings assume

'peration vithin the basic thermal hydraulic.c design criteria.

These criteria are LHGR S

13.QkM/ft and MCPR vithin 1&its of specification 3.5.K.

Amendment No.g, g1

, PS, 77

il

0

-ihVl'"II LIMIT LIMITING. hFETY SYSTFM ETTIHQ 1

1 FUEL CLADDING INTEGRITY 2 1

. FUEL CLADDING INTEGRITY Xf 'it is determined that. either 'of these design criteria is being violated during"'c'tion shall be initiated within 15 minutes to restore operation within the prescribed

'imits.

Surveillance requirements 'for

.APRN scram setpoints'

. are. given in Specification 4-1-B)-

3 The APRM Rod block trip setting sha'll. be:.

SRB( (0-668

+42%)

where:

S RB Rod block'etti'ng in'ercent-of rated thermal power (3293 MWt)

Loop recirculation flow rate in percent of rated (rated loop recircu-lation flow rate equals.

34 2 x 10~ lb/hr)

Amendment No. gg, 77

ll

uncertainties employed in deriving.the safety limit are provided at the beqinninq of each fuel cycle.

Because the boiling transition corrc),ation is based on a large

<<zpQiQy of ful1 seal e data there i

~ a very high confidence that operation of a fuel assembly at the condition of MCPR =

> ~ 07 would not produce boiling transition.

Thus, although it is not required to establish the safety limit additional margin exists between the safety limit and the actual occurrence of loss of cladding integrity.

However, if boiling transition were to occur, clad perforation would not be expected.

Cladding temperatures would increase to approximately 1100~F which is below the perforation temperature of the cladding material.

This has been verified by tests in the General Flectric Test Reactor (GETR) where fuel similar in design to BFNP operated above the critical heat flux for a significant period of time (30 minutes) without clad perfoiation.

Zf reactor pressure should ever exceed 1400 psia during normal power operating (the limit applicability of the boiling transition correlation) it would be assumed that. the fuel cladding integrity Safety Limit has been violated.

At pressures below 800 psia, the core elevation pressure drop (0

po~er, 0 flow) is greater than 4.56 psi.

At low powers and flows this pressure differential is maintained in the bypass region of the core.

Since the pressure drop in the bypass region is essentially all elevation

head, the core pressure drop at low powers and flows will always be greater than 4.56 psi.

Analyses show that with a flow of 28x10~ lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a

value of 3. 5 psi.

Thus, the bundle flow with a 4.56 psi driving head will be greater than 28x10~ lbs/hr.

Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.

With the design peaking factors this corresponds to a core thermal power of more than 50%.

Thus, a core thermal power limit of 25% for reactor pressures below 800 psia is con ser vative.

For the fuel in the core during periods when the reactor is

shutdown, consideration must also be given to water level requirements due to the effect of decay heat.

Ef water level 16 Amendment No

, 77

)~

i 0

Analyses of the limiting transients sho~ that no sera~

adjustment ia required to assure tCPR ) l.07 when the transient is initiated from HCPR ) ~~ ~.

ApRM F'lux Scram Tzi Setting Refuel or Sta t S Rot Standv Bode For operation in the startup mode ~hile he reac"oz is lcm pressure, the APRN. scram settinq o!

15 percent of rated povez provid s'ad'equate thermal marchen betMeen the setpoint and the safety limi', 25 percent of rated.

The marqin is adequate to accomodate anticxpated maneuvers associated wit.h po ez plant startuo.

Ff fecto of increasing pressure at'xezo or lov vozd content aze minor, cold water from sources available during staztup is no:

much coldez than that alzeady id the system, temperature coefficients are small, and control zod paztcrns are constrained to be unif ozm by operating procedures backed up by the red cnorth mznimxzer and the Rod Sequence Control System.

orth of individual zods is very lol" in a uniform rod pattern.

Thus, all of possible sources of reactivity input, un'form control rob vithdzaMal is the most probable cause of significant powr rise.

Because the flux, distribution associated

~ith unif orm rod acithdravals does not involve high loc l pe ks, and because several rods must be moved to change a

oca, pover by d.signif icant percent age of rated povez f rate of power rise is very sly.

Cenerally, the heat flux is in near equilibrium arith the fission ra e

In an assumed uni.form od vi phd anal approacn to the scram level, the rate of parer rise i.'s no more than 5 percent of rated po~er p. rinute, and the APiW system auld be more tnan adequate to assure a scram before the po ez could exceed the safety limit.

The 15 percent APRN scram remains active until the mode svitch is placed in the RUH position.

This ski'tch occurs

~hen reactor z3ressuze is creater than 850 psig 3.

3>>>>u-F>>ux sera.

Tr aettiu The'R'8 he IR'N System consists of 8 chamber',

33 i.n each of the

.reactor protect>on system logic channels.

The >> i iii See Section 3.5'K.

20 Amendment No;

it

, A%

a given'oint at constant recirculation flow.rate, and thus to proto ct against the condition of a HcpR less than 1.05.

This ro<l block trip setting, which is autdmatically varied with;recirculation. loop flow rate, prevents an increase in the reactor power level to excess values due to control rod withdrawal.

The.flow variable wrip setting provides substantial margin from fuel. da..age, assuming a steady-state operation at the trip setting, over the entire recirculation flow range..

The margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur duiinq the steady-state operation is. at 108% of rated thermal power because of the APRM rod block trip setting.

The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system.

Reactor Water Low.Level-Scram and Xsolation Exce t Main'heamlines 3

D The set point for the low level sc.-.am is above the bottom of'he separator skirt.

This level has been used in transient analyses dealing with coolant'inventory decrease.

The result's reported in FSAR subsection N14.5 show.th'at scram and isolation of. all process lines (except main steam) at this level adequately protects the fuel and the pressure 'barrier, because MCPR is greater than 1.05 in all cases, and system pressure does not reach the safety valve settinqs The scram setting is approximately 31 inches below the normal operatinq ranqe and is thus adequate'to avoid spurious scrams.

Turbine.Sto Valve Closure Scram The turbine stop valve closure trip~anticipates the pressur& neutron flux and heat flux increases that would result from closure of the stop valves.

With a trip setting of 10X of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintai>>ed even during the worst case transient that assumes the turbine.bypass valves

. remain closed.

'(Reference 2)-

Turbine control valve fast closure or turbine "trip scram anti.cipates the

pressure, neutron flux, and heat flux fncrcas<<

that

<<c>>>1>l r<<>>ulc 'fzo>>>

cont.roi v:>lvc'. fasr clos>>rc due tn load r<'jerti>>>> ot'>>>>t rol valv<<cl<>s>>t> ~

duc t>>.t>>rl>i>><< trip; each witl>out'ypass vaiv<<cap:>1>il ity.

Ti>>~

) >>':<<tor

~

protccti on syst<<>>> i>>it latcs a scram 1>> lc ss tl>>>> 30 mi I I isc>>>><<ls:>I I <<r

.the st'art of control valve f~st closure due to 1>>a>i, r<<j<<c tib>> or >>>>>tr>>l valve closur<<due I:o turbine trip.

'I'his scr:lm is <<cl>iuvc>l by t>>l>i>lly reducinct hvdrau'I ic contre> I 22 Amendment Ho. 9, Jg,,

77

IO i4

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3

1 REACTOR PROTECTION.'SYSTEM A

licabi lit Applies to the instrumentation and associated d'evices which initiate a reactor scram.

4 1

REACTOR PROTECTION SYSTEM A licabilit Applies to the surveillance of the in'strumentation and associated devices which initiate reactor scram.

Ob 'ective

~Ob 'ective To assure the operability of the reactor protection system.

To specify the type and frequency of surveillance to be applied to the protection instrumentation.

S cification Suecifica&dn When"there is fuel in the vessel, the setpoints, minimum number of trip systems, and minimum number of instrument channels that must be operable for each po'sition of the reactor mode switch shall be as given in Table 3.1.A.

A.

Instrumentation systems shall be functionally tested and calibrated as indicated in Tables 4.1.A and 4.1.B respectively.

31 Ci When it is determined that a channel is failed in the unsafe condition, the other RPS channels that monitor the same variable shall be functionally tested immediately before the trip system containing the failure is tripped.

The trip system containing the unsafe failure may be untripped for short periods of time to allow functional testing of the other trip system.

The trip system may be in the

~

untripped position for no more than eight hours per functional test period for this. testing ~

Amendment No 1

77

<~

Thy frequency of calibration of the APRM Flow Biasing Network has been established as each refueling outage.

There are several instruments which must be calibxated and it will take several hours. to perform the calibration of the entire network.

While the calibration is being performed, 'a zero flow signal will be sent to.half of the APRH's resultinq in a half scram and rod block condition.

Thus,'f the cali oration were performed during operation, flux shaping would not be possible.

Based on experience at other generating stations, drift of instruments, such as those in the Flow Biasing Netwoxk, is not, significant and therefore, to avoid spurious

scrams, a calibration frequency of each refuelinq outage is, established.

Group (C) devices are active only during a given portion of the operational cycle.

For example, the ZRM 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 ox startup; i.e., the tests that are performed just prior to use of the instrument.

Calibra'tion frequency of 'the instrument channel is divided into two groups.

These are as follows:

1.

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

2.

Vacuum tube or semiconductor devices and detectors that drift or lose sensitivity.

Experience with passive type instruments in generating stations and substations indicates that the specified calibrations are adequate.

For those devices which employ amplifiers etc., drift specifications call for drift to be less than 0.4%/month: i.e.,

in the.period of a month a drift of.4% would occur and thus providinq for 'adequate margin.

For the APRM system dxift 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 days.

Calibration on this frequency assures plant operation at or below thermal limits.

A comparison of Table 4.1.A and 4.1.B indicates that two instrument channels have not been included, in the latter table.

These are:

mode switch in shutdown and manual scram.

All of the devices or sensors associated with these scram functions are simple on-off switches

and, hence, calibration during operation is not applicable, i.e., the switch is either on or off.

46 Amendment No.

77

i 0

N. 1

. BASES

~go The sensitiv'ty of LPRN detectors decreases with exposure to neutron flux at a slow and, approximately constant rate.

The APRH system, which uses the LPRN readings to detect a change in.thermal power, will"'e calibrated every seven days using a heat balance to compensate for this change in sens'tivity.

The RBN system uses the LPRM reading to detect a localized change in thermal powe Tt applies a correction-factor based on the APRN output signal to determine the percent thermal power and therefore any change in: LPRH sensitivity is compensated for by the APRM calibration.

he technical specification limits of CNFLPD,

CPR, and MAPLHGR are determined by the use of the process computer or other backup methods.

These methods use LPRN,readings and TIP data to determine the power distribution.

Compensation in the process compute." for changes in LPRN sensitivity will be made'y performing a full core TIP traverse'to update the computer calculated LPRH correction factors every 1000 effective full power hours.

As a minimum the individual LPRM meter readings will be adjusted at the beginning of each operating cycle before reaching 100 percent power.

47 Amendment Ho.

77

0

1 ~

The minimum number of operable: channels for each trip Eunction is detailed for the star tup and run positions of t ~ reactor mohe selector meit~ The

SRM, IRM, and AFRM (star mode),

blocks need not be opeMable in "run" mohe, and the AFRt (flow biased) rod blocks need not be operable in "startup" mode.

Pith the number of OPERABLE channels less than required by the minimum OPERABLE channels per'rip function requit ement, place at least one inoperable channel in the tripped condition within one hour.

2.

H is the recirculation loop flow in percent of design.

Trip level setting is in percent of rated power (3293 NHt).

See Specification 2 1 for APRM control rod block setpoint.

E 3.

IRM downscale is bypassed when it is on its lowest mneme.

SRH's A and C do;rnscale functions are bypassed when lRM's A, C, E, and G are above range 2.

SRM's.B and D downscale function is bypassed when ERM's B, D, F, and H are above range 2.

SRM detector not in startup position is bypassed when the count rate is >100 CPS or the above condition is satisfied.

5.

During repair ot calibration of equipment, not more than one SRM or RBM channel nor more than two APRM or XBH channels may be bypassed.

Bypassed channels are not counted as operable channels to meet the minimum operable channel requirements.

Refer to section 3.10.B=for SRM requirements during core alterations.

6.

IRM channels A, E, C,

G all in range 8 or above bypasses SRM channels A and C functions.

ERM channels B,

P', D, H all in range 8 or above bypasse~FN channels B and D functions.

7.

The folloving operational restraints apply to the RBM only.

,a.

Both RE4 channels are bypassed vhen reactor pover is < 30$

and when a peripheral control rod is selected.

b.

The RBH need not be operable in the -startup" posltlon of the reactor mode selector svitch.

c.

Tvo RL4 channels are provided nnd only one of these mny be byp:i~scd from thc console. If the inoperable channel cannot be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the inoperable channel shall be placed in the tripped condition within one hour.

If minimum condition" for Table 3.2.C arc not met, administrative controls, shall be immediately imposed to prevent control rod vithd rnvnl.

AlABndment /pe., g, 77

i'

Limitin);.Conditions for Ooeration Surveillance Reouirements 3.5 Core and Containment Coolin S stems 4.5 Core and,Containmen" Coolin S stems L.

APRM Setooints L.

APRM Setooints 1.

Whenever the core thermal power is> 25% of rated, the ration of FHP/CMFLPD shall.be

> 1.0, or the APRM scram and rod block setpoint equations listed in sections 2.,1.A and 2.1.B shall be multiplied by FRP/CMFLPD as follows:

FRP/CMFLPD shall be determined daily when the reactor is > 25$ of rated thermal power.

6<<

(0.66W + 54Z} QPDPD 6'

(0 ~ 66W+ 42X)

(

)

RB-

~LPD

'2 ~

When it is determined. that I'.5.L.1 is not being met, hours is allowed to correct the condition.

3 ~ If 3.5.L. 1 and 3.5.L.2 cannot be met, the reactor power shall be reduced to< 25$ of rated thermal power within 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

M.

Reoortin Recuirements If any of the 'limiting values identified in Specifications 3.5.I, J, K, or L.3 are ex-ceeded and the specified remedial action is taken, the event shall be logged and reported in a 30-day written report.

167A Anendment No.

77

0

gene ra tion if fue1 pellet densif ication is postulated.

The LHGR shall be checked daily during reactor operation at 25% power to determine if fuel burnup, or control rod movement has caused changes in. powex distxibution.

For LHGR to be a limiting value belo~

25% rated thermal power, the MZPF would have to be greater than 10 which is precluded by a considerable margin when employing any permissible control rod pattern.

K ~

Minimum Critical Power'atio (MCPR At core thermal power levels less than or equal to '25%, the reactor will be operating at minimum recirculation pump speed and the mocerator voi'd content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hyd aulic analysis indicated that the resulting MCPR value i's in excess oz requirements by a considerable margin.

With this low void

conten, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR.

The daily requirement for calculating MCPR above 25% rated thermal power is sufficient since power distribution shifts are very slow when there'ave not been significant power or control rod changes.

The requirement for calculating MCPR

~hen a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power snape (regardl'ess of magnitude) that could place operation at thermal limit.

3.5.L APRH Setooints Operation is constrained to a maximum LHGR of 18.5 kV/ft for 7x7 fuel and 13.0 kN/ft for SxS, Sx8R,. and PSxSR.

This limit is reached when core maximum fraction of limiting power density (CMFLPD) equals 1.0.

For the case where CHFLPD exceeds the fraction of rated thermal power, oper ation is permitted only at less than 100-percent rated power and only with APRH scram'ettings as required by specification 3.5.L.

1

~

The scram trip setting and, rod block trip setting are adjusted to ensure that no combination of CMFLPD and FRP will increase tne LHGR transient peak 'beyond that allowed by the 1-percent plastic strain limit.

A 6-hour time period to achieve this condition 's justified s'nce the additional margin gained by the setdown adjustment is above and beyond that-ensured by the safety analysis.

177

'mendment No.

77

. 3.5 BASES.

3.5.M Reoortin Reauirements The LCO's assoc ated with monitoring the fuel rod operating conditions are required to be met at, all times,, i.e., there is no alloyable time in which the plant can knowingly exceed the limiting values for MAPLHGR, LHGR, and MCPR. It is a requirement, as stated in Specification 3.5.X~

J, and K, that if at.any time during steady state power operation it is determined that the limi ing values for MAPLHGR, LHGR, or MCPR are

exceeded, action is then initiated to restore operation to within the prescribed limits.

This action is initiated as soon as normal surveillance indicates that an oper ating limit has been reached.

Each event involving steady state operation beyond a specified limit shall be reported within 30 days.

It must be recognized that there is always an action which would return any of the parameters (MAPLHGR, LHGR, or MCPR) to within prescribed limits, namely power reduction.

Under most circumtances, this will not be the only alternative.

~:-,

s or towns Ferry 1.

'Loss-of-Coolant Accident Analvsi f

B

'- nuclear Plant Unit 3, NEDO-24194A and Addenda n

L

~

IIBVR Transient Analysis.".odcl Utili.in the Program " TVA-TR81-01-A.

i i. ing the RETURN 3.

Generic Reload Fuel Appl'icatca ion, icensing Top ica l Repor t, NEDE-24011-P-A and Addenda.

17.8 Amendment Vo