ML18283A897

From kanterella
Jump to navigation Jump to search
Enclosure 1: Proposed Changes to Browns Ferry Nuclear Plant Unit 1 Technical Specifications, Attachments 1, 2, & 3
ML18283A897
Person / Time
Site: Browns Ferry Tennessee Valley Authority icon.png
Issue date: 10/10/2018
From:
Tennessee Valley Authority
To:
Office of Nuclear Reactor Regulation
References
Download: ML18283A897 (162)


Text

ENCLOSURE 1-PROPOSED CHANGES TO BROWNS FERRY NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATIONS Revised Response to NRC Question 7 Concerning the Unit 1 Reload Amendment Reas'ons for Proposed Changes Proposed Changes

0 0

I~

ATTACHMENT 1 UESTION 7 Provide a schedule for completion of requested operational assurance tests as outlined in the appendices A and B (attached).

RESPONSE - Item 4.1 A endix A The tests listed below will be performed during startup following the Browns Ferry Unit 1 refueling outage:

(1) Control Rod Drive S stem Test All control rods will be functionally tested before fuel loading.

Friction and scram time tests will be performed during cold-open vessel testing following fuel loading. All control rods will be scram timed at rated pressure (> 950 psig) before exceeding 40 percent power. Technical specifications requirements will be met.

(2) Full Core Shutdown Mar in Test This test will verify the shutdown margin with the highest worth control rod withdrawn and compare the measured critical eigenvalue for a specified control pattern with the calculated value. Technical specifications requirements will be met.

,(3) Core Power Distribution Test The purposes of this test are to:

(a) Confirm the reproducibility of the Traversing Incore Probe (TIP) system readings.

(b) Determine the core power distribution in three dimensions and compare it to the predicted power distribution.

(c) Determine the core power symmetry. The results of TIP reproducibility and core power symmetry tests will be analyred promptly. Acceptance criteria are being formulated at the present time.

(4) Additional tests vill be conducted as necessary to verify proper operation of systems modified during the outage.

RESPONSE Item 4.2 A endix A A startup summary report vill be. submitted to the NRC in accordance with technical specification 6.7, to include summaries of fuel loading and startup tests as listed above.

For these reasons, TVA requests a reevaluation of the need, for the data~

requested.

RESPONSE - Item 4.3 A endix A The action requested is nov required by technical specifications for BFNP, RESPONSE Item 4.4 A endix A Details of the General Electric Fuel Surveillance Program were provided to the USNRC on a generic basis in Reference 3. In that submittal, that General Electric maintains an active program of surveillance of both it vas indicated production and prototypical BMR fuel, specifically intended to monitor per-formance in operating reactors and to identify and characterize unexpected phenomena which can influence fuel integrity and performance. Outage-oriented interim examinations on lead fuel of a particular design are performed contingent on fuel availability as influenced by plant operation.

As noted in Reference 3, the tvo most recent surveillance programs implemented by GE are the 8x8 Surveillance Program, initiated in 1974 with the introduction of the 8x8 fuel design, and the. Lead Test Assembly Program put in place in 1976 in support of the improved 8x8 design. The 8x8 Surveillance Program is directly applicable to the Browns.Ferry 1 Reload 1, since the fuel designs are the saOlee

41 0

The 8xO Surveillance Program features two 8x8 assemblies which reflect the design of approximately 6,000 such assanblies currently in operation. The two lead 8x8 assemblies were dimensionally and non-destructively characterized in 197'~, prior 'to irradiation, and subsequently placed in the Honticello and.quad Cities-1 reactor cores. To date there have been a total of four interim examinations performed on these bundles, with an additional examination planned for 1977.

Reference 3:~ Ronald Fngel (GE letter to Dr. D. F.. Ross.(USNRC),

Sub)act: General Electric Fuel Surveillance Program, Hay 18, 1977. ~

~

~

0 0

ATTACHMENT 2 REASONS FOR PROPOSED CHANGES TO BFNP UNIT 1 TECHNICAL SPECIFICATIONS In" general, proposed: changes which have been submitted to NRC by previous requests are marked in the margin by a single bar. However, any new change requested is marked in the, margin by a double bar. These new changes are as a result of several discussions with NRC staff members concerning resolution of varied. NRC questions concerning the unit 1 reload amendment.

C Page vii - Addition of Table 3.5.I by, TVA BFNP TS 94.

Page viii - Deletion of Figures 3.5.1-A and 3.5.1-B and addition of Figure 3.5.3 "by TVA BFNP TS 94.

Page 5 Addition of definition of CMFLPD per discussions with NRC staff and. deletion of Total Peaking Factor.

Page 8 Changing MCPR safety limit from 1.05 to 1.06 per TVA BFNP TS 86.

Page 9 - Substitution of CMFLPD and FRP for MTPF per discussions with NRC staff and. changing LHGR and MCPR limits per TVA BFNP TS 86.

Page 10 - Substitution of CMFLPD and FRP for MTPF per discussions with NRC staff.

Page 15 - Changes as a result of TVA BFNP TS 86.

Page 16 - Changes as a result of TVA BFNP TS 86 and substitution of CMFLPD for MTPF per discussion with NRC staff.

Page 17 Changes per TVA BFNP TS 86 and TVA BFNP TS 95.

Page 19 - Changes per TVA BFNP TS 86 and allowing steady-state operation without forced recirculation for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per discussions with NRC staff.

Page 21 Substituting CMFLPD and FRP for hiTPF per discussions with NRC staff. Also, changes per TVA BFNP TS 86.

Page 22 Changes per TVA BFNP TS 86.

0 0

Page 23 - Substituting CMFLPD a'nd FRP for MTPF per discussions with NRC staff. Also, changes per TVA BFNP TS 86.

Page 25 Changes per TVA BFNP TS 86 and TVA BFNP TS 95.

Page 27 Changes per TVA BFNP TS 95.

Page 28 Same as 27 a'bove.

Page 29 Changes per TVA BFNP TS 95.

Page 30 - .Same as 29 above.

Page 31 - Substitution of CMFLPD and FRP for MTPF per discussions with I

NRC staff.

Page 47 Same as 31 above.

Page 48 Same as 31 above.

Page 74 Same as 31 above.

Page 113 Changes per TVA BFNP TS 86.

Page 122 Same as 113 above.

Page 123 Changes per TVA BFNP TS 75, also clarification as to when this condition applies.

Page 124 Changes per TVA BFNP TS 75 and. TVA BFNP TS 86.

Page 125 Same as 113 above.

Page 129 - Same as 123 above.

Page 131 Changes per TVA BFNP TS 86.

Page 133 - Changes per TVA BFNP TS 75 ~

t Page 134 - Changes per TVA BFNP TS 86.

Page 157 - Changes per TVA BFNP TS 96.

Page 159 - Changes per TVA BFNP TS 94.

Page 160 Changes per TVA BFNP TS 86; also addition of power spiking penalty for Sx8 fuel and. LT for SxS fuel per NRC request.

Page 167 - Same as 157 above.

Page 168 Same as 159 above.

Page 169 Changes per TVA BFNP TS 86 and. TVA BFNP TS 95.

0 II

Pages 171,- Changes per TVA BFNP TS 94.

172 Page 173 No changes.

Page 173a Change per TVA BFNP TS 95.

Page 182 Changes allowing steady-state operation without forced recirculation for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per discussions with HRC staff.

pages 315, Changes per TVA BFNP TS 93.

327 Page 330 - Change per TVA BFNP TS 86.

IP

'ATTACHMENT 3

LIST OF TABLES Cont'd Table Title ~Pa e No.

4.2.F Minimum Test.and Calibration Frequency for Surveil lance Instrumentation 105 4.2. G Surveillance Requirements for Control Room Isolation Instrumentation 106 4.'2.H Minimum Test and Calibration Frequency for Flood Protection Instrumentation 107 4.2.J. Seismic Monitoring Instrument Surveillance 108 3.5.X MAPLHGR vs Average Planar Exposure 3.6.H Shock Suppressors (Snubbers) 4.6.A Reactor Coolant System Inservice Inspection-Schedule 209 3.7.A Primary Containment Isolation Valves 250 3.7.B Testable Penetrations with Double 0-Ring S eals . 256 3.7.C Testable Penetrations with Testable Bellows . 257 3.7.0 Primary Containment Testable Isolation Valves . 258 3.7.E Suppression Chamber Influent Lines Stop-Check Globe Valve Leakage Rates . 263 3.7.F Check Valves on Suppression Chamber Influent Lines ~ ~ 0 ~

  • ~ ~ ~ ~ ~ ~ ~ ~ 263 3.7.H Testable Electrical Penetrations 265 4.8.A Radioactive Liquid Waste Sampling and Analysis 287 4.8.8 Radioactive Gaseous Waste Sampling and Analysis . 288 3.11.A Fire Protection System Hydraulic Requirements . . . 324 6.3.A Protection Factors for Respirators 343 6.8.A Minimum Shift Crew Requirements . 360

LIST Ol I LLUSTRATIONS

~F1 ere Title P~ee Ne.

2.1.) APf'N Flow Reference Scram and APRH Rod Block Set tings t ~ ~ ~ ~ ~ ~ ~ ~, ~, ~, ~ e ~ ~ ~ 0 ~ ~ h 1-3 2.1-2 APRM i=low Bias Scram Vs. Reactor Core Flow;... 26 4.1-1 Graphic Aid in the Selection of an Adequate Interval Between Tests . . . . . . . . . ., . . . 49 4.2-1 System Unavailability .

3.4-1 Sodium Pentaborate Solution Volume'Oncentration Requirements 138 3.4-2 Sodium Pentaborate Solution Temperature, Requirements 139 3 5e2 Kf Factor e ~ '73 3'5.3 MCPR Vs Cyc1e Average Exposure l. l..l... ... ~ 173-a

3. 6-1 Minimum .Temperaturie 'F Above Change in Trans'ient Temperature . 188 3.6-2 Change in Char py V Tr ansi ti on Temperature Vs.

Neutron Exposure . . . . ~ .i . i. i ... . , . . . 189 6.1-1 TYA Office of Power Organization f'r Operation of Nuclear Power Plants...... 361 6.1-2 Functional Organization ~ 362 6.2-1 Review 'and Audit Function . 363 6.3-1 In-Plant Fire Program Organikatlion l . . . ~ . . . 364

1>>0 MPINITXONS (Cont 'd) 4 1 At least one door in each access opening is closed.

The standby gas treatment system is opezable.

3.,All Reactor BuQ.ding ventilation system automatic isolation valves are operable or deactivated in the isolated position.

for.a particular unit and the end: of the next subsequent refueling outage for the same unit.

Refuelin Outa e Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the staztup of the unit after that refueling. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled outage; however, where such outages occur within & months of the completion of the previous refueling outage, the required surveillance testing need not be performed until the next regularly scheduled outage.

Alteration of the Reactor Coze - The act of moving any component in the region above the core support plate, below the upper grid and vithi'n the shroud. Normal control rod movement with thc control rod drive hydraulic system is not defined as a core alteration. Normal movement of in-coze instrumentation and the traversing in-core probe is not defined as,a core alteration.

Reactor Vessel'ressure Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications .are those measured by the reactor vessel steam space detectors.

U>> Thermal Parameters

l. Minimum Critical Po~er Ratio (MCPR) - Minimum Critical Power Ratio MCPR is the value of the critical power ratio asso-ciated with the most limiting assembly in the reactoz core.

Critical Power Patio (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.

2. Transition Boilin - Transition boiling means the boiling regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intermit-tently with neither type being completely stable.
3. Core Maximum Fraction of Limiting Power Density (CMFLPD) - The highest ratio, for all fuel types in the core, of the maximum fuel rod power density (kW/ft) for a given fuel type to the limiting fuel rod. power density (kW/ft) for that fuel type.
4. Avera e Planar Linear Heat Generation Rate APLHGR The Average Planar, Heat Generation Rate is applicable 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'eight divided by the number of fuel rods in the fuel bundle.

h M 5~y

1~0 DEFINITIONS (Cont'd)

V. Xne C'nunentiL tion Xnstrument Calibration, -.. An~ ifistrument 0'alibration: means of .sn instrument', sign@1',output so that 'value'(s) it co&~sphnds> the,'d)ustment of Che.

~

+%thin,acceptable range, and kcclarhcy'," to a'.knoMn

.parameter,w>~lch the in'strum'ent monitors+

2~ Channel', - h channel is an arrangement ~Of,'a. sensor and;asao, ciated icomponents;.used to'kaluIat'e: plant.'variables and;pro duce diicrete outputs used. in logLc, A channel teisLinatde.

and~los!is,its. identity vhare &diVtdual'hannel outputs in logic., are'ombined tf 30 I'nstrument:Puncttionsl- Test -, Ari, instrument functional test a simulated s5gna1 into"the,instrument primary mealna'the'in)ection'of co-verify"tha proper* instalment channel reiponae, '<<lIIH 'ensor, and(or initiating 'acCion.

Xnstruient -Ch'eclk'- An instrument, chec'k" is, qualitative,deC'ermin'a-by ob'sarvatian; of instrument,~ 'tion.'of'!acceptable".opeiability behavior" dur8eg operation.: .Wi's .let'e&inIati'on,'h'all,. Seclude, i@are pos'sible, comparison of. the'in'strum'eat Wth other indepen-dent instruments measuring the eam ~ variable.

~Lo ic~Sstem 'Punctiofiil..Test '- A .logic system functiona1 Cist means"a test'.of all relays add coritacts ok a logic circuit go, insure all, components. ere.oplsrablk pIar 'design intent. 'Share practicable, ection..~rill go to 'ceeylkti'on i.e. ~ pumps and valvis operated.

vi11be'tarted 6, System -.A. tiip system means an.az'rangemenc of

'rip trip signaLs snd aux).3.ivory equipment required to inatrInasInt'Cannel trip

!Lnk':cilce'ction to= accompliih a protec.tijve. trkp function. h syitem may require one or 'more- instnuaent charine1 trip signals rklatect

,to one or more 'plant- parameters',ogder t,o ini,ciate trip system action. Initiation of protecICi~ve acL'ion may require the 4ri'ppIIn of .a single triji system ow thee kof'accede'st ,'tripping of eve'rip'ye to!ss ~ .

pr'coactive Ac'tion"- An; action initiated'y'he protection systcIm limit is. reached. A prot'.ec'tike ection can be at a'channel'r

'hen-a system l,evel.

8. Pr'otective Function;, A system protective action vhich results the',protective. action of'h'ie it:ha'nnels monitoring a palrt.L- 'ree cular plant condition.
9. Simulated..:Automatic.A< tuation; Simulated automatic actuation meass applying a simulated signal to the sensor to actuate'hs io 'q041st'.ton, 'iRNit

ia r ~to ie - S logic is sa'errsngeeent oi relsis, contacts,, sndocher components'h<<t, produces a" decision outDut ~

(a) ~tnicistia - s logic that receive signals iron ehsnasls aad produces decioion outputs to the actuation logic.

(1) Actuation - A logic that receives signals, (either fran initi<<tion logic or channels) and produces. decision output<<

to <<ccomplish,a protective action.

hxactional Tests -,A functional. test ia the manual oper<<tion or initi<<tion of a system, subayatemg o" component to verify that it functions within deaign tolerancea (aag., the manual start of <<

core <<pray peep to verify that it,runs and that it pump<< the required volueaa of vatar).

X. Shutdeva - Zhs reactor is in s shutdoea condition ches ths r ecr-or toads snitch is ia ths shut'dove aode position sed ae cora altaratioas are being parfoaaed.

Y. En in<<Brad S<<fs uard An snginasrad safeguard,is <<safety syatan the actions of which ara oaaenti<<1;to << a<<faty action raguired in x'eaponae to <<ccidenta, g Cumulative Downtime - The ca<~> tive downtime for those safety ooisponents end systems whose downtime is limited to j consecutive days prior to raquiring reaotoz shutdown shall, be limited to any 7 days in.a oonsecutive 30 day period.

SAFETY LIitXT LIHZTXNG SAFETY SYSTEM! SETTXNQ L .CLMDDlG INTEGRITY 2.1 HJEL CLADDING INTEGRITY h licabilit Apy]Licabf1 it~

hpplies to the inte~related vari- AppILies t:o trip settings of the ~

ables ass'ociated Mith fuel instruttents'nd devices trhich are theraal,behavior prov'ded to prevent the reactor safety limits from being 'ystem e'xceeded.,

Ob ectiv>>

To establish lixx&s which ensure To def~ine the level of the pr~acess the integrity of the fuel clad- variables at uhich aut:omati'c Pro'=

dingo tective action is i.nit:iated, to pre-vent t'e fuel claCding inte'gritty'.

saf..ty 3.ice.t from neing exceeded.

S ecifications Specification A. Reactor Pressure > 800 psia The ~lituiting safety'ystem settings shall and .Core FloIa > 107 of Rated,. be as specified belo~t Vnen the reactor p"essure is A. Neutron Flu@ Scr~~~

greater than 800 jisia, the e<istenre of a mi'.iimum c'riti.-

"al po"er ri tlo (YCPR) less

l. APRN Flute Scram Trip Setting (Run !lode) than 1.06 shal.l constitute viola"ion of the fuel claddic- R'hen th. Bode SMitch i" in integiigi safety limit. the RUN position, the APP~

flu)c scram trip setting shall bet S<(0 66K + 5'47)

+he..-e:

S ~ Setting in percent of rated. thermal po~er (3293 MVt)

W ~ Loop recirculation floe rate in percent of rated (rated loop re.circcLlation flot~ rate equals 34-2&,06 lb/hr)

SAFETY LTMIT LIMITING SAFETY SYSTVI SETTING

'I

~ FIIi.'L CI I'Dl)fNG TNTEGHT'I'Y

~

FVEIn CLADDING INTFGRTTY In, the event of operation with, the core maximum fraction of limiting power density (CMFLPD) greater than fraction of rated thermal power (FRP) the setting shall be modified as follows:

S~ (0.66W + 54K) FRP CMFLPD For no combination of loop recircu-lation flow rate and core thermal

'ower shall the APRM flux scram trip setting be allowed to exceed 120X

'of rated thermal power.

(Note: These settings assume operation within the basic thermal hydraulic design criteria. 'These c'riteria are LHGR K 18.5 kw/ft for 7X7 fuel and~

13.4 kw/ft for 8X8 fuel and MCPR within limits of Figure 3.5.3. If it is determined that either of these design criteria is being violated during operation, action shall be initiated within 15. minutes to restore operation within pr'escribed limits.

Surveillance requirements for APE!

scram setpoint are given in specification 4.1.B. C

2. APRM When the reactor mode switch is in the STARTUP POSITION, the APRM scr"m shall be set at less than or equal to 15K of rated power.

3~

IRM The IRM scram shall be set at less than or equal to 120/125: of full scale, B. APRM Rod Block Tri Settin B. Core 5.'hea~al Power Limit (Reactor Pressure <800 psia) The APRN Rod Mock trip setting shall be:

',"? e.- the reactor pressure is less than or equal to 800 psia,

! AFRTY .LIMIT . LIMTTTNG .SAFETY SYSTFM,SETTING t'.

1 FUEL CLADDING INTEG'RI'IY 2 1 I"UEL CLM)DING INTEGRITX o'r core coolant, flow 'is less ,RR< (0.66W + 421) than 10$ of'ated, .thc'ore thermal power shall not ex-., whar'c:

,c'eed,823'&t (a'bout. 25$

thermal power). of'ated

RB Bod block sett'ing is percent of rated thermal power. (3293 I'Mt)

Loop recirculation flow rate in percent of rated (rated loop recirculci)ion flow rate erluals, 34;2 X: 10 lb/hr)

In,the event of operation with the core maximum fraction cif limiting power densit~

(CIfFLPD) greater than fraction'f r'ate'.8 thermal power,(FRP) the s'etting shall be moclifiecI as follows:

s <(0.66m + 42Z)

Whenever the reactor .is vari 'C. 'l:ram.and isoluation-"<<'> '538 in. above the shutdown corrdition,~pith reac'tor low water vessel zer'o .'Level ir'radiated 'fuel in thie reac- f tor vessel, %he .water level shall'ot -be le!'s ths,n above the tcrp of thrr 17.7'n.*

D.- SCram turbine stop < 10,'ercent normal active fidel zon'e 'alve c'lo."'ure valve closure

)  ;,E. Scram--turbine control valvule Upon trip of 1., Fast closure the fast acting solenoid valves

(

2 ~ Loss of control > 550 psig oil pressure F. Scram-,-low c'on- 23'nches I

denser vacuum Hg vacuum G. Scram main steam < 10 'percent line isolation va3.ve closure H. Main stIeam i: olation > 825 psig valve closure nuclear systerI> low pressurie 10 IP

'.. 1 BAS F.S: HJFL CLADOItaC INTFCRITY S~Q'ETY LIMIT Tne fuel cladding represents one of the physical ca b arrxers which separate radio-active materiaLs from environs. The integrity of this c 1 a ddin g b arriez is related to its relative freedom from perforations a ons orr crac ki Al ing. Although some corrosion or use-related cracking may. occur during fissioa product.migration from this source is increme n th e 1 fieo f t h ecladding, ncrementat 11 y cumuLative 1 and continuously measurable. Fuel c1adding perforations ra ions, h owever, can result from.

thermaL stresses which occur from reactor operation significantly above design conditi'ons and the protection 'system setpoints. awhile fi i od c a ng per ormation is )ust as measurable as that from use-re1ated cracking, the thermally-caused cladding perforations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deteriora-tion. Therefore, the fuel .cladding safety limit is defined in. terms of the reactor operating conditions which'an result in cladding per'foration.

The fuel 'cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational transient. Because fuel damage is not, directly observable, the fuel:cladding Safety Limit is defined wi.th margin to t'e conditions which would produce onset transition boiling (MCPR of 1.0).

This establishes a Safety Limit such .that the, minimum critical power ratio (MCPR)

, is no less than 1.06 MCPR >1.06 represents a conservative margin relative to the conditions, required to maintain fuel cladding, integrity.

Onset of transition boiling results in a decrease in" heat transfer from the'clad an, therefore, elevated clad temperature and the, possiblity of clad failure.

Since boiling transition is not a directly observable parameter, the margin

,to ooiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution. The margin, for 'each, fuel assembly is characterired by the critical power ratio (CPR) which is the ratio of the bundle power which ~ould produce onset of transi.tion boiling divided by the actual bundle power. The minimum value of this zatio for any bundle in the core i's the minimum critical power ratio .(HCPR). Zt is assumed that the plait operation is controlled to the nominal protective setpoints via the instru-mented variables, i.e., normal plant operation presented on Figure Z.l.l by the.

no@ ingl exnor ted Fina control'ir o.. one Safetv Limit (MCP'0 nF L,4Q has suf~<cient co'nservatism to. assure that. in the event of an abnormal operational transient initiate~. from a normal operating condition (MCpR > value shown in Figure 3;5.3 for'x7 or,8x8 fuel assemblies) more than 99.9gtransiti'on. of the fuel rods in th core are expected to avoid boiling The marain between MCPR of 1.0 (onset of transition boiLing) and the safety limit, 1.0E - is derived from;,a detailed statistical, analysis considering all of the uncertainties in moni,-

toring the core operating; state including uncertainty in the boiling .transition correlation as described in Reference 1. The uncertainties employed in deriving the safety limit are provided at, the beginning of each fuel cycle.

15

1. 1 'BASES Because the boil.ing transition coxrelstion i.s 'based on a large quasi'n;~ty'of full scale Cata there is a very hi.gH c6nfiddnce that opex'ation of'I fIxeI, assembly at the condition of MCPR =1.06 would not produce boiling,tran; sition. Thus, although it 'is not required to establish the safety 'lMt additional maxgin exists betveen the'afety lu'nit', akd the actual'cc&exIce of loss of cladding integrity.

Hovever, if boiling be expected,.

transition vere Cladding temperatures to occur, clad perforation would not vould incrqase t;o approximately 1100 F whi.ch is belov the perforatio6 t'emgerILture of the cladding material. This hss been verified by testy ip (he, General Electric Teat Reactor (GETR} where fuel similar in design to BHYP operated above critical 'heat flux for a significant period, of time (30 minutes'} 'he without clarl per'foratXon.

If reactor pressure should ever exceed $ 400 IIsi's caring normal po'wex~

operating (the li'.mit of'pp1icability oi the boiling txsnsition has been itviolated.

would be assigned that the fu)l plpdpng integrity Safety corx.e-'ation}

Limit Xn addition to the boiling transitioxI l."Im/t,(MCPR ~ l.o6) operation ir constrained to a maximum LHGR of 18.5 kw/ft for 7x7 fuel d 13 4 k'r";.

or x8 fuel.. This limit is reached 'whf n 'the Core Maximum Fraction of Limiting Power Den'sity 'equals 1.0 (CMFLPD 1.0). For the case he' Co e, am lfaximum Free'tion of Limitint pacer Deneity exceed'ha Fraction cf Power, operation is permitted only at less than 100X of rated pow'er and only with reduced APRM scram m':etti'6gs's'equired b y spec:

At pressures belov, 800 psi'a, the core elevation pressure drop (0 QvAr,

'fi fated'hermal cat on 0 flov} is greatex'hsxi 4.56 psi. At, 1&v pove'rs and flovs this is nod.ntai'.ned in the. bypass're'ginn of the core. Since the pressure'ifferential pressure drop in the bypas's region is es<>entieLlly all elevation head" the core pressure drop st low powers and flow will alwsvs Se greater than 4.56 psi. An'al;y'seis show that vith eI. flow of 28XL03 1bs/hr bundle flov, bundle pres'sure drop is nearly independent. of bundle power and has a value of 3'5 ps'hus the buncQ:e flow M'th a 4.56 psi dri will .be greater than 28xlO lbs/hr. FiilI. scale ATLAS test 'data taken',.

i h st pressures from 14a7 pais to 800 psis i'ndi.ca'te that. the fuel assembly critical power at this flow is spproxiinatelIr 3.3,'i atm Pith the design peaking factors thi,s correispond's to a core thermal pover of more r t2xan 5 ~+. Thus, a core thermal power I.imit of 25/ for reactox pressures belov 800 psia. is consexvative; For the fuel in the;.'core during periods v4en t2>e reactor is shut'own s deration must also be given to vatex level requirements due to th'e off'ect of decay heat If Pater level should drop belo~ the top of the fuel during this time, the ability to remove decay )cert fs reduced, ,This reduc'tion in cooling capability could. lead to elevated,clading temperature and I. d p erforation. As long ss the fuel remains,covered vith water, suff'icient, cooling is available to prevent fuel clhdapeifo'ration.

1. 1 BASF.S The saf'ety limit has been established at 17.7 in. above the top of the irradiated fuel to provide a point which'can be, monitored and also pro-vide adequate margin. This, point .corresponds approximately to the top of the actual fuel assemblies and also to the lower reactor low water level trip (378" above vessel 'zero)".

REFERFNCE

1. General Electric 'BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application, NEDO 10958 and NEDE 10958.
2. General Electric BWR Reload' Licensing Amendment for BFNP Unit 1, NED0-24020, May 1977
3. General Electric BWR Increased. Relief Valve Simmer Margin Evaluation for Browns Ferry Nuclear Plant Unit 1, September. 27, 1977 17

PAGE DELETFD 2.1 BAS ': LXMITI'G SAPFTY SYSTFM SETTINGS RFLATFD TO HiFL CLADD NQ INTEGRITY Thelabno~al operational transients applicable to operation of the Br~s Ferry Nuc].ear Plant have been analyzed throughout the spectrum of planned operating con-d itipns up to the design thermal power condition of 3440 MWt. The analyses were based upon plant operation in accordance with the opezating map given in Reference 2. In addition, 3293 MWt is the licensed maximum power level of Browns Ferry Nuclear Plant and this represents the mavimum steady-state power which shall not kncjwingly be exceeded.

Conservatism is incorporated in the transient analyses in estimating the ~

'controlling factors such as void reactivity coefficient, control rod sczam

~

worth, scram d'clay time, peaking facctors, pnd axial power shapes. These factors are selected conservatively with respect to their ef fect on the applicable transient results as d'etcrmined by the current analysis model.

This transient model, evolved over many years, has been substantiated in opera-tion,ks a conservative 'tool or evaluating reactor dynamic performance.

Results obtained from a General Electric boiling water reactor. have been compared with predictions made by the 'model. The comparisions and results are 'summarized in Reference l.

The absolute value of the void reactivity coeEEicient used in thc analysis is conscrvativcly estimated to be about 257 greater than the nominal maximum value expec-ed to occur during thc core lifetime. The scram worth used has been dcrated to bc equivalent'to approx',.ately SC cf the total scram worth of the control rods. Thc scram delay time and rate of rod insertion allowed we. '+lib pn~lvee s arc conservatively sct equal to thc longest delay ard slov-st insertion rate acceptable by Technical Specifications.

The effect of scram worth, scram delay time and zod insertion rate, all conservatively applied, aze of greatest significance in the early portion of the negative reactivity insertion. The rapid insertion of negative reactivity is assured by the time requirements for 5X and 20> insertion.

By the time the rods are 60K inserted, approximately four dollars of negative reac-tivity has been inserted which strongly turns the transient, and accomplishes the desired effect. The times for 50K and 90X insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state condition.

For analyses of the thermal consequences of the transients a MCPR given,.by figure 3.5.3 is conservatively assumed to exist prior to initiation of the transients.

This choice of using conservative values of contzolling parameters and initiating transients at the design power level, produces more pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels.

Steady-state operation without forced recirculation will not be permitted for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

19

2 1 A E6 ln summary

l. The, licensed maximum poMer level is 3,293'l61t.

I

2. Analyses of, transients employ adequateIly conservative values of the controlling reactor para=';e < rs.

3 ~) The abnormal operational transients Mere analy",ed to a powei level of,'440 N:7.

The analytical procedures cov used result ic a: mare logical answer than the alternative method of assuming a higher starting paver in con)ucc-~

tion Mt,th the expected values for the parameters.:

The has<<es for indiyidlua.'L set points are diecussed below<

P, ~ Neutron Plux Sere~

l. 'PP~i High. Flux, .'scram Trip Setting ('Run a'lode')

The. average po~er range monito irg (APRH) system, vhic4 is calibrated using heat balance data taken Curiog steady-state conditions, reads

'percent of ra=ed pew'er (3293 HV't) . Because fission chambers pro>>

vide the bas!'.c input ei~wals the IZP~~ system responds neu ron flux,. During tra sients, thc instant'.aceous 'rate of dire'ctly'to'verage hist ti ann fej f rom the fuel (reactor the<<r.".al power) is less than the instantaneouiw ne'utron f1m. due to the time cor'stant of the fuel.

There~ore, du".ing tr~xnsients induced by disturbances, the the~>i pover of t'e. fuel vill be less thap that ii:dicated by 'the'eutron flux at, <<he sc='am setting. ~elyses repor ed in Section 14'f the P" nal

.Safety Analysis Report de"~onst:a ed rhat iCth a 120 percen. scram ,'trip',

.s'e t'ng, ';Ioae of the abno".~l operat:Loral t flcoients scaly=ed violate the fuel safe"y'!.mit acd there is a substantial margin fry fuel di~ge. Th=iefore, use of s flov-biased sera provides even addi=,iocal, m4,=,x'-.n.. Figure 2 ~,l.i~ shoes the floM biased scree as a function of" care flo~,.

An ircrease in the PZH.'f sera= set'tin'g wou"d 'decrease'he margia; pre-sent befoie tne fuel clcddin", 'L;>>err'ity safety ='"- is reach'..ne APRH scr<<m setting'ps de e:~ir. " by an analysis o- margins " quired

'to prov de a ".I:asonsble ran';;e for ~ne vatic,. du ing. operi ion.

Reducing this .operatin;; mar'gin iould iacr@as'e t';.. frequency of spurious ocra-a, vhich have an advc"ice ef;uct on reac or safety because of the result;Lng chir.~l itresses. Th s~, t'he'APPA sett'ng ~as selected bacause 't pion'des adeq ate margin for the fuel claddin" inte",=icy safaty limit'jl:." allcvs o'perat'ng ~rgin that reduces ha posslbilit'y of uI1QacslSsary ocrams

'st ~

20

BASES The scram trip setting must be ad)usted to ensure that the LHGR transient peak 'is not increased for any combination of CHPLPD and PRP. The scram 1 ~

setting is ad)usted in accordance with. the formula in specification 2.1."A.l when the CMFLPD exceeds FRP.

Analyses -of the linting transients show that no scram adjustment > is required to assure NCPR >l.06 +hen the transient is initiated from MCPR the values shown in Figure 3. 5. 3.

2. APR.'f Flux Scram Tri Settin (Refuel or Start & Hot Standb Node)

For operation in th" stertup node Mhile the reactor is at loM pressure, che APR.'! scram setting of: 15 percent of rated po.-cr provid s adequate

, chernal'argin betveen the setpoinc and the seiety limit, 25 percent of raced. The margin fs adequate to accommodate anticipated maneuvers

> associated Mi.ch povex plant startup. Effects of increasing pressure at xero or loM void content are r:inor, cold Matex from sources avail-able dur'r.g startup 's not much colder than that alxeady in the system, temperature coef ficients ere small, and control rod patterns ex'e con-strained co be ur.iform by ope'retina procedures backed uo by th rod

~birch xrinimi'zer end the Rod Sequence Con ro'ystem. Porch of indivi-dusl rods is very lov in a uniform rod p-ttern. Thus, all of posaiole sources of reactivity input, unifo. control rod vithdraval is the moat probable cause of s',,nificent poser rise. Because the ilux di.stribu ioa associa ed vxth uniform rod withdrawals does not involve high local peaks, and because several rods must bc moved to c'nenge pover by a aiyxificanc percentage of ratec power,,tnu rate oi pover r'se is very slow. Generally, the hest flux is in n nr equilibx'ium with the fission rate. In an asau=ed unifox= rod Michdrewal approach co the scram lev 1, the rate of pover rise is no more c"..en 5,percent of rated powex'er rrinuce, and cnc APR.'I system.

vould be aore chan acequace co asaux'e e scram before the power could exceed the sa ety limit. The 15 percent APK~ scram remains active until che ~de svitch is placed in the RUl) po"ition. This swicch occurs when x'eactor pressure is greater than 850 psig.

3. IRM Flux Scram Tri Settin The IR,'i System consists of 8 chambers, 4 in each oi the reaccor protec-tion system logic channels. The IB'i is a 5-decade instrument .;nich covers the range of poMer level between that covered by the SW and the APR.'!. Tne 5 decades are covered by the IL'! by means o: a range switch end che 5 decades are broken down into 10 ranges, each being one-hali of a decade in si e. The IRN scram setting of 120 divisions is active in each range oc the It&. For 21

.1 BASES

3. IRM .Flux Scram Trfo Sett fnn {Contin<<ed)'

example, if the instrument t'Iere on raflge 1, the scram setting would be at,129 if divisions for that range; likewise, the instrument was on range 5, the ~scxam setting would be 120 divisions on that range, Thus, as the IRM is ranged upj td accommodate the increase in povex level, the scram setting is a3Lso ranged,up, A, scram at 120 divisions on the IR6[ instruments remains in effect as long as the ~

reactor is in the start:up mode. In addition, the APRN 15K scram pre>rents, higher power operation without being in the MJN mode, The IRM: cram provides protection for changes which occur both:Locally and ovex the entire c:ore, The most significant sources of reactivLty change during the power increase are due"to control rod withdrawal. For insequence cont:rol rod withdrawal, the rate of change of power is slow enough due,to the physical limit:ation of withdrawing control rods, that he.at flux is in equilibrium with the neutron flux and an XRM scram would result in a reactor shutdown well before any safety limit is exceeded. 'Por'he case of a single Con'trol rod withdrawal error a range of rod vithdrawal accidents vas ana1'yzed. i This'naly.o~s included st:arting the>.accident at variou. power levels. The most severe ca~e involves, an initial con'dit'ion in which the react:or is )ust subcritical~and the IRM .ystem is not yet'n scale. This condition exist: at quarter;rod density. Quarter rod density is illustrated in paragraph 7.'5.5 of the FSAR. Additional conservatism was taken in this analysis'y assuming that the XRN channel. closest to .the withdrawn rod is 'bypas. ed,. The results of this ~analysis shov that the reactor is scrammed and peak, p'ower limited to one percent of rated povex, thus t>>tafntaining HCPR, above 1 06 Based on the above analysis>> t:,he IRK provides protection against local control rod w>i.thdraval errors and,continue>us, withdrawal of

.control rods in sequence, B. APRM Control Rod Block Reactor pover level may be varied by moving contxol rods or b'y varying the',recirculation flow rate. 'The APRM system provides a control 'r'od block to prevent rod wi,thdrawal beyond a given point at ~constant recir-cuclation flow rate, and thus to protect against the condition of a

. MCPR less than 1.06. 'Ibis rod block trip .setting, which is automatically

'arried with recirculation loop flow rate., prevents an increase in the'reactor power level, to excess values due to control rod with-dx'awal'.",. The flow vaxiable trip settfng provides,su,bstantfal margffl 22

2. 1 BASES from fuel damage, assuming, a stead'y-state operation at the trip setting, over the entire recirculati.on 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 HCPR which cou'd occur during steady-state'operation is at 108K of rated thermal power because of the APRH rod block trip setting. The actual power distribution in the core is escablished by speci.icd control rod sequences and is mani.cored continuously by the in-core LPRM system. As with the APRH scram trip setting, the APRH rod block trip setting is adjusted downward exceeds ~p if the thus preserving the APRH rod block safety margin.

CMFLPD C. Reactor Water Low Level Scram and Isolation (Exce t Hain Steamlines)

The set point for the low level scram is above the bottom of the separator skirt.

This level has been used in transient analyses dealing with coolant inventory decrease. The results reported in FSAR subsection 14.5 show that scram and isolacion of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because HCPR is greater than 1,06 in all cases, and t system pressure does not reach the safety valve settings. The scram setting is approximately 31 inches below the normal operating range and is thus adequate to avoid spurious scraps.

D. Turbine Sto Valve Closure Scram The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of < 10 percent of valve closure from ful.l open, the resultant increase in bundle power is limited suc'h that MCPR remains above 3.,06 even during che worst case transient that assumes the turbine bypass is

~ I closed. This scram is bypassed when turbine steam flow is below 30 percent of rated, as measured by turbine limits first stage pressure. Actuation of the pressure to well below the safety valve setting.

relief valves E. Turbine Control Valve Scram

1. Fast Closure Scram The reactor protection system initiates a scram within 30 Nsec after valves start to close. This setting and the fact that control valve che'ontrol closure time is approxiaately twice as long as that for the stop valves that resulcing transients, while similar, are less severe than for 'eans stop-valve closure. ho fuel damage occurs, and reactor system pressure does not exceed che relief valve sec point, which is approximately 280 psi below the safety limit.

23

2. l. BASES 2; Scram on lo',ss of control oil pressiira The turbine hydraulic coIatrol sy'stem opc'rates'sing high pressure; oil. The're,ar', several points i'n thiis oil system,vhere n'loss of, oil pressluri could result in a feet closure 'of the turbine control valves. This fast closure of the turbine control valves is by~ the generat<or load rejection scram, since faint!e 4f aot,'rotcctekl the oil aIyste<<<< vpuld'ot result in the fast closure solenoid valves being actuated. Por a turbine contr'ol 'valve fast closure,,

the core vould. be protecitcd by the APHH and high reactor. pressure scrams. Ho<iever to provide the same margins as provided for the ~

gcrieiitor 16ad rejection scram on fast closure of the turbine control vallves, a sera~ lpga been added to the reactor protectlioh.

system; .vhi<:h senses failure of control oi.l prcssure to the tury bine control system. This is 'an anticipatpry sera<a and reiuitsi eihuItdovn before any s'igniticant increase in pres>sure or in'eactor ncutr'own flux occurs. The transient response is very simil'ar generator load rejection.

,'to,'hat resulting from this P.. Mein Condenser lIov Vacuum Scram To protect the inain c:onidcnscr'gainst overprcssurie, a ]Loss of con-denser v'acuua< initiates au'tomatic cliosurc of the turbine stop versa and'urbine bypixss valves. To anticipate the transient and automatic scram resulting from the closure of the, turbine s'top valves, lov icon- i denser vacuu<<< initi'ates a scram. This lov vacuum ecr'am set point is o'eIe'cted to 'i+i<-'iate' iscram befc e the,.Losurc o:E thc turbine "is. in itop'alv'es C. 6 H. Main,Stage Line ls~ ation on Lov P'r<:ysurc,and Hain Steam Lin'e Isalation Scram The lov pres8iure isolationi of the main steam lines at 825 psig vss provided tio pirotect 'against rapid, reactor depressurization and the result'ing rapiid coolclov<o of the vessel. JLdvantagie is taken of the scram 'fcaturei .that oc:cure vhcn the msin stcam line isolation valves are closed, to, provic!lc for reactor shutdovn so that high pover opera-tion at lo'reiosur does not occur, thus providing protccti'on for th'e funnel cladding 1ntegrity safety limi.t. Operation iof the reac-tor it prcssurep lovcIr thin g?5 psig requires t'hat the reactor mode!

svitch be .fn'he STARTUP position,'hez'c pro'tection of the fuel cladding

'i'ntegrity isafety limit is provided by the IRH and APRH high neutron fliux scrams. Thus, the cc<mbination of 'main 'st&am~ line lo'v preissv're isola<tion and isolation vcIlve closure scram aihu'r<es 'this availability of neutron flux scram protectioai over the, entire range bf 'apgli'cability of the 'fuel c:ladding 'integrity safety lin<it. In'd'di0iori, the isolation valve closure scram arIt'iicipatoa the pressute ~and flux'. tWan'sisnt>> t'hat occur during nor<<<al or -inadvertent iiiolati*n Valve'losure. Mith the scrams iet at'10 percaeIt of .'valve closure, rieutr6a 11u'x doe'e not in'crease.

24

2. 1 BASES

,I. J. & K. Reactor lo~ ~ater level set oint for initiation of HPGI and RCIC closin main steam isolation valves and startin LPCI and core s ra um s.

These systems. maintain adequate coolant inventory and provide core cooling with the objective of preventing excessive clad temperatures.

The design of these systems to adequately perform the intended func-tion is based on the specifi'ed lov level scram set point and initia-tion set points. Transient analyses reported in Section 14 of the PSAR demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.

L.. References Linford', R. B., "Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Mater Reactor," NED0-10802, Feb., 1973.

2. General Electric NED0-24020, May 1977.

BUR Reload l Licensing Amendment for BFNP unit l, 0)

~ ~

30 General Electric BWR Increased Relief Valve Simmer Margin Evaluation for Browns Ferry Nuclear Plant Unit l, September 27, 1977

~g 25

~ l,:.

T ~

I',,"lsI o ~

I ~

s"'"s

~

~

I

~ s l ".II:El<

I: ~

EI il 1!I' I ~

ill( ~ ~

I I s.s<

~ 'I o

~ ~

'I ) ~ .': 's ~

I

"(.

I 'i

~ <<o 120:

I I 110

)".)

~ s I".';

ZR4 I

'A=-

"i.

~ ~ ~

~ Es SCiV'4 o, I.

~ ".",

'! o

":I:)I:!i ci s;',":. Ii ,.I, ik , I...I '

~

I Ill !ili i Iji( ls( I  :! Ici:! I

.I' ...

(:: ', I-..

.;El I i III

.i(csl Pi !ill :2:.E " I.'Ii I

i '! 90. ts ~ 'l

i i<l!

s

..I:!:!

I".,'Ls(ll Ii'I

.:I. I ..I !:.I.c ~

~ ~

I ki(

'll I'Ii "I I

~IIII I "Jt'~: I

((:l.ki I t.cE

. I'.  :

s

,s I ~ EI (lj o!Iss'll :Iio

~ o MSIGN FMV COE fTROT LDIE

'I  ::I.}(i I!II f<

I'.' sl: ik :..)'.':: TT:.

""t  ;;o 70 I o .. ...) I

,~ ~

EI

~ ~ o 60 )I i'I::  ! ~ 'Es i'.::::-'(il)llf I<i:I:".I I y(::-.'" go ciii .I Ill  !<<

'I:  !<lil'.! !I st ~ ~

- ,'::J.:.::::::,.:lii'l(l,

'tls il!I ~ ~

s ~

~

,;<i,

-~', i(()lll!I .;;I( ~', ".. I s EE'" III E Ic'!c; Il(kji ~

~ ~

I: I'

~

NMWRAL CZRCUKA+ZOII .:I i!<,'.! l I':!ll <(Is() I '. l" I:,'i<i

~

g I:..: :::.Eij'I'

'p.! 30' ~

(ji jvc:I !i' ii:i

lit o'

' n!

0 oii(jil 20

)'::

~ ~

(ii(IIIII:s'.!II:I Its! .!

'I'

': I I '.2( PU!6 S~ L~ I ~ ~ ~~

I

~ ~It

~

!~

ltl:

Isf Ill il(E.:-';

I

!I(i !::

~ II ~

s!IIII

(

Tlt '.jk i: I I')I!

ii'.I I,
;.';. (,I I

~

10; ..I ..

!': ~ ~ )! I.'its I I 'JI') )IE!

I ol

~ (

o I. <i .. ~

~

I'I.i0 I- "!I' OI"! IIOt (s

)

20 j[l 'I,.'!1o,,

"I'.::l.::"

~ ~

op f

'I:!. FCC RA'~ TEE DPSygf)~i,'!!"j"'g""(

.-L

~" R~ COOT'LFV FXGV OF

' s

'o APRI FXD'A':I'AS , SCEUQ4 Vs. RZ~R'O'RE '~~

.":-) 'I, FIG. 2. 1-2, s

~ s

'"'):i:.-

I

-L

I (I o),o

~

, o .

~ ti

.'ll

/-I s- o

~

J 26

SaFETY LIMIT LIMZTINO SAFETY SYSTEM SEITING L.2 REACTOR COOLANT SYSTEM INTEGRITY 2.2 REACTOR COOLANT SYSTEM INTEGRITY Aplies to limits on reactor coolant Applies to t rip se t ting's of the ay<stem pressure instruments and devices which are provided to prevent the reactor system safety limits from being exceeded.

O /ective O~b ecclve

'll Tv establish a limit below which To define tl<e level of the process the integrity of the reactor coolant variables et which automatic pro-system is not threatened due tn an tective action is initiated to overpc essur'e condition. prevent the pressure safety limit from being exceeded.

S ecification S ecification A. The pressure at the lowest point The limiting safety system settings of the reactor vessel shall not shall be as specified below; exceed 1,375 psig vhencver irradiated fuel is in the reac- Limiting Safety

'tor vessel e Protective Action S stem Sett<in A. Nuclear system 1250 psig safety valves + 13 psi (2 open-<<nuclear valves) system pressure

~ ~

B.. Nuclear system 1105 psig +

relief valves open--nuclear ll psi (4 valves) system pressure

'115 Psig +

ll psi (4 valves) 1125 psig +

ll psi (3 valves)

C. 'crimnuclear < 1,055 psig

, system high pressure 27

REACTOR.COOL'AHT SYSTEM<<t 1N EGRITY

~ 'ie safety liiai'ti 'fo'r the riiccor cobol'a4t syktc<m 1ires<iure hive beeh selecce<J

~ ich that: th<<y acre below pr'eisuxco ai: which it can be sho<~ that the integrity of the 'sy:<t m te noc endlsngered. H~'ever,'he pressure safety limits Are e st high eiiough i<uch ttiat no foi.-ese'eikblh dirkumstanccs can cauie 'che sist'em pre'saure t'o ri'i'e o'ver thieie limits; Thi'e pri ssure safety limits a'e 'arbicrax'ily a'e<l'ected to bii the 1oIiIeeit icra~nsierit c<veii;pr'ca!sures< illojied by the Ia<<pplikable ASIDE Btiiler'and P'res'sure Veisel C'o'de; Section III, and VSAS Piptng '"'odeiq

)ice, S'ec c ioh 831. li

't ie deitgn prie'siir!. (1,250 psig) of che reactor veiii.1 is established !iuch

'chat; Mhen th'e 10 p'errent allowance (123 psi) adlawed by the ASME Boiler and Pc'es'sbre t'e!isel Code .iecticin III for pressure tranifencs is iddcd 't<i'che ,

de!<ijn pr'esiux'ei s transi'ent preiisure limit of 1,375 prig is established.

C'orreijondin'Sly i the dei'ign<<preseure '(1,148 piig< f'r suet;ion aiid 1',326 pii<t for dischar je} of the r'eitctor relcix'culaciioi'i s'stem piping ax'e; such t hat',

~hen che 20 pe'icenc allowance (230 'in'65I pai} 'alloved by USAS.:Pt.ping Code, Se'ctiion B31.,1 foi pr'ee'suit'e tr'ansiciiiI:i ero added to the d'eaigtt pxeseuiI,'cs, craiis icnt" pi eeiiir'e lfmitii of I;37e 'and 1',59k" 'pily i'ire'iicablishi d. Thus, ihe',pr<<'essiir'e s<'afety limit a'pplicable ico p'ower o'jieration is established at 1,375'sig (t'hc lovcst ti;ancient over'p're'siiire<<'lloi~ed by the pex'tinent codes).

A'SME B<<oi'lei and Pressure VeQadil I o'de';Section III, it<nd USAS Piping Code, Secti'on B31'.1; The current c'ycle,'s si.fe'4y analysis conc'.erning the most severe abnormaiL oper'ational transient ressulting directl'y .Ln 'a r'eactor coola~t system pressure in'crease'is given in Reference '5 Ias 'suppl'emented by Referen'ce 6..

Theeactor'* veise1 pre!i's'ure code limit of 1,3i5 psig given in subsbcdioA 4 2

<'f th'e s'afe'ty ana1ys'is'e'po'rt is well'bove the peak pressure produced by '

the overp'i'essure transient described'bele.'i Thus, the pressure safety limit~'

appl'ic'abl'e'd powe'r',o'peration's well above thie peak pr'essure that can result:

'. du'e to reisonab'ly expected'v'er'p<ressuke tiraInsfents.

i tl'tgh'e'r ddaigii'r'iiiiiure<i have been eitabli'shed for ptping Mithin the reactor cool'srit sy'iiceia t'h;in' or the'eactor'e'seal'. 'hese iricreased desigri a<<coneiite*nc disign whijh assureii cha'c, if the pressure within <<th' r'eeet'or ve'as'e'I'oes< iiiot: exceed 1,375 p'sig', the pressures vithin the p'f ping p'resleu',res'xea'c'e" cannot e'rceedhe'i'.r'eepei':tive transient pressure limits due co static and puii'p"-'ea'ds'.

'I'I'ie'a'fe'c'y I'iiiit'f 1,375 p'sic accueil~a i'ppiie's co irony point in the reac'cox ve'seel; hov~"<<<er, bec<iuse of the static ua'ce'r head, che hf ghesc pressure point will o'ccvr a'c the bottom of. cha','essel., Becauie t: h>> pressure ie not monicox'ed ac chi's'oint, it cannot be diriccly'eteraiin<x'd if this safety limit h<is been vi'olated'; Alii'o;, because of che potentially va>..ying head 'le<eel and f lou pres-sure'r<ip'e", anquiv'iii'lerit'riia'sure~armlet'eriox;i determined for a 28

1.2 BASCS pressure monitor hirher in thc vessel. Therefore, following any transient

.that,is severe enough to cause concern that this nfety limit was violated, a calculation ~ill be performed using all available iniormation to deter-mine ii the saictv, limit was viol'ated.

REFERENCES

1. Plant Safety Analysis (BFNP FSAR Section 14.0)
2. ASME Boiler snd Pressure Vessel Code Section II I.
3. USAS Piping Code, Section B31.1
4. Reactor Vessel and Appurtenances Mechanical Design (BFNP 'FSAR Subsection 4.2)
5. GE BWR Reload 1 Licensing Amendment for BFNP unit l,NED0-24020.

May 3.977

6. TVA letter to NRC of September 27, 1977, Z. p, Darling to E. G. Case.

with Enclosure General Electric HHR Increased Relief Valve Simmer Margin Evaluation for Browns Ferry Nuclear Plant Unit..lg

2 ~ 2>> aAsss REACTOR: COOI,ANT: SYSTEM'kiTBGRITY."

The prcssure reli,ef system for, each;unit at the ILwrowns Ferry Nuclear Plant 'has been.si.zed to meet two de'sign bases. First,, the total safety/

re'ief, valve'apacity has been cst<<blishcd to, meet the oierpressure, pro-te" t ion'cri'teria of the: !Bilk: C'odc. Second; th<. distrib>~L'i<><r nf this requred'apacity between, safety valves and relief valves hu., been set to r'.eet cesi'gn basi: 4.>i.I,'-1'of sub.-ection. 4. wliich state" that 'th nuclear sy te-. reli'ef valves *shajLl. prevent op ning of the safety valves during norma'lan. isolati'o'ns and load reJections.

The d<'.tai s of the, analy.,is w:sich shops comnliah 'e .with tHe AS/2 C'odc reouire:..ento is presente<'. in subse"tion 4.4, of tii>e FS!'iR 'and the Reactor Vasss'el Overpressure Protection Summary Technical Repent submitted in re:-.jonse t'o: question': 4.1 datedl Dece;.bcr 3., 1971.

To';,ineet'he'e.co>wd design IIasis, the, total safety-relief caI'acitv,of lt I 8/i'7>> has beer< divide<I, intci,. 70X re'.lief. (11 viilves) and 142X safety I

(2'ilves),",. The,'n'alysis.c>f the. plant isolation ",transient is given in Reference 5.: st<pp'lemented'by Refer'ence. 6.;on p'age, 29.

30.

LIMITING CONDITIONS FOR, OPERATION SURVEILLANCE RE UIREMENTS

3. 1 REACTOR PROTECTION SYSTEM 4~1 REACTOR PROTECTION SYSTEM"-

A licabilit A licabilit Applies to the instrumentation Applies to the surveillance of and,associated devices, which the instrumentation and asso-

"- initiate .a reactor scram. ciated devices which initiate reactor scram.

~ob ective Ob ective To assure the operability of the To specify the type and frequency reactor protection system. of surveillance,to-be applied to the protectIon Instrumentation.

S ecification S ecification When there is fuel in the vessel, A. Instrumentation systems shall the setpoints, minimum number of be functionally tested and trip systems, and minimum number calibrated as indicated in of instrument channels that must Tables 4.1.A and 4.1.B respec-be operable for each position of tively.

the reactor mode switch shall be as given in Table 3.1.A. B. Daily during reactor power operation at greater than or equal to 254 ther-mal power, the ratio of Fraction of Rated. Power (FRP) to Core Maximum

'Fraction of Limiting Power Density (CMFLPD) shall be checked and the scram and APRM Rod Block settings given by equations in specifications 2.1.A.l and 2.1.B shall be calculated.

C. 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 sys-tem containing the failure is tripped. The trip system con-taining 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 t'esting.

PAGE DELETED 32

The frequency of calibration of the APRM Plow Biasing 'Network has been established < ~ f< natl v.

The sensitivity'f LPR'l detectors decreases vith exposure to neutron flux at a 'slow. and, approximately constant This is compensated for in APRM iystem by calibrating every 7 days using heat balance data and rate.,'he by calibrating, individual 1.PRM's, every 1000 ef factive full-pover hours usiug TIP traverse data.

48

TABM 3.1.C INSTRUHENTATION THA'f INITIATES ROD BLOCKS Kto ious No.

Operable Per '

~Trf ~Sc 5 Function Tri. Level Sct tin 2(1 j APRH Upscale (Plov Bias) < O,66w + 42X (2) 2(1) APRH Upscale (Startup Hode) (8) < 12Z 2 (1) APRH Dovnscale (9) > 3X 2 <1) APRH Inoperative (10 )

RW Upscale (Plov Bias) < 0.66w+ 41X (2)

RBH Dovaocale (9) > 32 RBH Inoperative (lo,)

3(1) IRH Upscale (8) <108/125 of full scale 3 (1) IRH Dovnocale (3) (8) 5/125 of full scale 3(1) IRH Detector not in Startup Position (8)

IRH Inoperative (8) (10 )

2(1) (6) 5 BU< Upscale (8) < 1 x 10 counts/oec.

2(1) (6) SRH Downscale (4)(8) > 3 counto/sec.

2(1) (6) SRH Detector not in Staitup Position (4)(8) 2(1) (6) SRH Inoperative (8) (108) 2 (1) Plov Bias CcLparator <10X difference in recirculation floors 2 (1) YloM Bias Upscale < 11,0I recirculation flov Rod Block Loafe M/A RQCS Restraint 147 psig turbine (PS-85-61A 5 first st~e pressure (approximately 30X pover)

PS-85-618)

!'~V~ FW! TAN.K 'y.2.C

l. For, the atartup and run, poaitioiin of the Reactor Mode Selector Svftch, there shall bc tvo operable or tripped trip system's for, each function.

The SN ~ IlVf, and APRH (Startup mode) ~ blocks need'ot be operable in "Run".mode, inCh the hPRPl (Plov biased) and RIN rod blocks need not be opirabie in "Startup" sjode. IE the first column cannot be e!et <<or one of the tvo trip systems, this condition may'x'iat for un to'seven days provided that during that time the oplerable stystem ia functionally teated ineediataly and daily thereafter; if thia condition last longer than seven days, the system vith the inoperable channel shall be tripped.

If the first cailuce cannot be met'fo>> both trip ayeteasq both trip eysteiaa shall be tripped.

2'. V is the recirculation loop few in percent, of, design. '1'rip level setting is in percent of rated power (3293 M~t). A ratio'of FRP/CHEAP'D <1.0 is permitted at reduced p'ower. See Specification 2.1 for APRH contxol rod black shtgoilnt.',

3 ~

4'.

IHH dovnaoale is'nJpasaed vhen it is on f.ts lovest, range~

This function is bypassed v4!n the count rate is > 100 cpa and IRM above range 2.

S. One instnuoent channel; i.e. one APAH or IR'Il or R'BH, per trip ayatecc sly be bypassed except only one of four SRM may be bypassed.

6 IRH channels A, E, CC all in range 8 bypasses HUC channe.La A 5 C functions.

IRH channels 8, P, DH all in range 8 bypasses SR!4 c'hannola ',I 5 D functions.

7~ The trip ie bypassed vhen the reactor pover is + 30X.

8. This function ia bypassed vhen the mode avitch is placed in Run.

9'.,This function is only'ctive vhen the node svitch is in Run. This function id automatically bypassed vhen th6 IRH ~inkt6sme'ntation 'is and not high. 'perable

10. The inoperative. trips are produc:ed by the folloving functionac
a. SIN a'nd IRM (1) Local "operate-wa.'Librate" avitch not'n'perate.

(2) Pover supply voltage lav.

(3) Circuit boards not in circuit.

(1) Local '"opera'ta-calibrate" avitch n'ot 'in'pera'ts,l (2) Lese than 14 LPRN inputs.

(3) Circuit boards,not in circuit.

74

,0 I~

3. 2 BASRS The HPCI high Elov and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation re-sults fn actuatfon of HPCI isol'ation valves. Trfppfng logic for the high flov is .a 1 out of 2 logic, and all sensors are required o be operable.

High temperature in the vicinity of the HPCI equipment fs sensed by 4 sets of 4 bimetallic temperature svitches. The 16 teraperature svitches are arranged in 2 trip systems with 8 temperatur svitche" in each trfp system.

~

The HPCI trip settings nf 90 psi for high flov and 200 F for high tem-perature are such 'that core uncovery is prevented and fission product release,is vithfn limits.

The RCIC high Elov and temperature fnstrumentatfon are arranged the sam ~

as that for'he HPCI. The trip setting of 450'-'20 for high flow and 200'F for, temperature are based on thc same criteria as the HPCI.

High temperature at the Reactor Cleanup System floor drain could indicate a break in tire cleanup system. When high temper" ture occurs, the cleanup system is isolated.

The instrumentation vhich initiates CSCS action is arranged in,a dual bus system. As for other vital instrumentation arranged in this fashion, the Specification preserves thc effectiveness of the ysten even during periods uhen nafntenance or testing is being perfumed. An exception to

.this is when logfc functional testing is being performed.

'he control,rod block functions are provided to prevent excessive control rod.withdrawal so that MCPR does not decrease to $ ,06. The trip logic for this function is 1 out of n: c.g., any trip on one of six APRH's, eight IRH's, or fou- SRH's will result fn a rod block.

The minimum instr rment channel'equirements assure suEEfcient instrumenta-tion to assure the single failure crfterfa is met. The minimum instrument channel requirements for the RBH may be reduced oy one for maintenance, testing, or calibration. This tfrae period is only 3% of the operating time in a month and does not signiEicantly increase the risk of preventing an inadvertent control rod withdrawal.

The APRH rod. block function fs Elov biased and prevents a significant reduc-tion in HCPR, especially during operation at reduced Elov. The APRH pro-,

vides gross core protection; i. e., limits the gross core pover increase from vfthdraval of control rods fn the normal vithdraval sequence. The trips, are set so that HCPR is maintained greater than 1.06.

The RBH rod block function provides Local protection of the core; 'i.e.,

the prevention of critical power in a local region oE the core, for a single rod vithdraval error from a limiting control rod pattern:

113

If,.the, IRM.,channels are :Ln the, vorst condition of alloved bypass!, the sealj.'ng. arrang;em'ent Ps such .that for unbypa',ssed XRM'.channels, a rod block

signal is generated, before the,,detecte'd neutrons flIex has iI'!creased by more than a factor <of 10 A, dovns'cale .,fndicati~on. is!,.an,.indication the instrument has failed or the.

,instrument is: not, sensitive., enough. In ~it~her ca!se the instrument vill; not. respond to changes it!i control, rod m'otio~n and~ thus', c'ontrol rod motic!n i'i prevented.

f'r t'efueling:Lnterloc'ks The safety only vhen the also operate one logic !channel, and are required mode switch is:Ln the refueling position.

.For effect.ive i mergency core. cnoling for'm'all pipe breaks, the HPCI system aust.,'function, since,.r .actor piesyure does. not decrease rapid enough to allov; either core. spray br 1'.PCI'o operate ~in tihie The automatic pressure

~

relief function is provided as a backup to the. HPCI in the event the HPCI do'es,.not oprracr.. The arrangemen't of thr. C~rL1<<pit!g,'contacts is such as to pr'ov<idc. this fis!!c,t,ion...<~hen necesaary and~miniWLzle spurious;operation. The trip,setti>>ys given i>> the specification are.adequate to- assure the above criteria. are met. Thc specLfication p'reker0e6 the 'ef fee'tiv'enrss'f the, syatem,durLng pcrio<!ls <of ma:Lntenance, teatihg,'t calibration!, and also minimizi.s: the risk cif,inadvertent oper'ation; i.e'., 'onLy one inst'rument channel oo't of 'service.

Tvo;,post;,treatment. off-gas radiation monit'or0 ave prd!vided'zi!d,'wh'en their trip>'point. is. r'eachedl; cause, an isolation 'of'the,off gas line. Isolation i's jnitiated.when:.hot:h,:Lnstn!ments -reach, their, high-tri> p'oi!l!t or,'one tr'ip and t: he othei: a downscale 'trip or 'both have a downscaleh4s 'n,<upsc'al<e tiip.

Both instruments,are required for,.trip bait thsl ihstruments are set so that any insti'uments< are sel so t'hat the instantaneous stack release rale limit given Ln Specification 3.8 is not exceeded,;

Four ri'<tint Lon 1no>>itors are 'provt<l<<: I for each u>>it ~hich initiate Primary Contat>>ment,fsoLat ipin (Croup 6,isola'.ion-,valves) Reactor Building Isolation and operati'on c!f the:.Standby Gas "Tr<eat'me!!!t.,Sy's'tc!I!. 'hese i'nserueent channels monitor >tl!e,ridi'at ion:in the Reactor zone ventilation exhaust ducts and Refueling Zone. in'he Trip settinj:oE 100 mr/hr for the. monitors in the Refueling Zona are based Upon initiating.normal ventilatiain isolation andSCTS operatibn so th)>t none, of the, activity'i leased .,during the 're(ueling'accident leaves 'th!L Reactor <Building via -.'the normal venCLlation path'but rather all Chs activity is processed by'lhe SCTS.

Floe,-integ<rators,and dete<rmine,, leakage sump in 'the fill raCe and pump out,r'ate timers Are drj~ei.l.:A system vhereby the time interval to used to fill;a knovn volume vial be, utilized .Co 1IroIride 4 backup,',,An 'ait saM>lihg is<.also provided to detect leakage inside the primary containment 'ystem (See< Table 3.2.E) .

e LL4

~ .4ITIHG CONDITIONS FOR 0?ERATXON SURVKXLLVlCE RE UIR~"fi iTS

3. 3. A 'REACTXVXTY CONTROLS'.

i.3.A REACT"'/XTY CON:ROLS Control rod s with scram b. A second licensed operator t imes g res t er than those shall verify the confor-pezni t ted by S pec 1 f ica- mance to Specification tion 3.3.C'.3 aze inooez- 3.3.A.2.d before a rod may ab'le, but if they can be be .bypassed in,the Rod inserted with control .zod Sequence Control Syst a.

drive pressure they need not be disarmed electri- c hlien itthatis initio-lly dere=-

rol rc= is cally. mined a con incapable c: nor.-.al inscrtiom

d. Control zoda 'with a failed an attempt to fully insc =

"Full-in" or "Full-out" the control rod shal'e position switch nay be by- made. If thc contro1'c" canno be ful.ly inscr.c', z passed in the 'Rod Sequence Control System and consi-= shutdown+ margin test sh"Kl dered operable if the actual be made to dcmonstr"tc un '"

rod position is knotm. These this condition tha. hc "=:-.

rods must be moved in sequence can be radc suhc it'.ca: zor to theiz correct poaitions any reactivitl condition (full in on insertion or full during thc re.-.ainder oF t.".".

out on withdrawal). opera t ing c> clc r'i th the analytically dctcrrincd,

e. Control rods with inoperable highest worth control rod accumulators o" those whose capable of i'ithdzai;al, full position cannot be positively and ..11 othe . 'ithdrawn, determined shall be consi- control rods capable oF dered inoperable. insertion fully ins rtcd.

Inoperable control rods shall The control zod accumulators be positioned such that Speci- shall be determined 'operable fication 3.3.A.l is met. In at least once per 7 days by addition, during reactor power veri ying that the p'ressure operation, no more than one control rod in any 5 x 5 array and level detectors are.not the alarmed condition.

i .

may be inoperable (at least 4 operable control rods must separate any 2 inoperable ones). Xf this Specifica-tion cannot be act the reac-tor shall not ba started, oz if at power, the reactoz shall be bzought to a shut-down condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> B. Control Rods B. Control Rods

l. Each control rod shall be l'., The coupling integ~ ty'.shall be coupled to its .driv or verified for each Wthdravn con-completely inserted .and the trol rod as fol3.ass:

121

I.ZMITTNQ COND'KTEOHS FOR OPERATIC)N, SVRVEZI.I-ANCE ~

REQVXREHENTS'B Control'ods 4.3.B Control Rods control rod directional V~eri,fy,.that, the control. rod control valves disarcied: is following the drive by electrically. '.This require- obse:rving a response in the ment does,not apply in the nuclear instrumentatiovI each, refuel. conditionwhen the tQe a,rod is moved when reactor is vented. 'Iwo,con- t'e reactor is operatir'~g trol,r'od drives may be removed above the pre-'set power.

as lo'ng as, Specification level of the RSCS.

3. 3.A. l is met.
b. When the rod is fully with-dra~n the first. time afte each refueling outage or after maintenance, observe that the drive does not go to te overtravel position.
2. The control rod drive, '.'he'6nt'rol rod drive housing housing suppor't system shall support system s'hall be inspected be in place duri.ng reaclor after reassembly,,and 'the reIsuLts power operation or when the of'he: inspection, recorded.

reactor coolarit system is

'ressurized aboye at,'ospheric pressure with fardel in the reac-tor vessel, unless all contipl rods are fully insertted and Specification 3.3.A.l is met.

3. e. Mhenever the. reactor is in 3. I'rior to the start of control the startup or run iiiodes rod wi.th'drawel. at startup, and b'elow 2Q rated power the, prior to attaining 20(i rate,8 Rod Sequence Cont'rol System 'power during rod insertion at (RSCS) shall be operable. shutdown the capsb'lity of the Rod Sequence Control System (E<SCS) the Rod North Minimizer to pr'operly fulfilltheir func'tioas a~ha'.Ll be verified by the follow-ing chec):s:

122

Li'wITING CONDITIONS FOR OP RATION SURVEILLANCE RK L'IZ~~iS

~ .3.8 Control Rods '4.3.8 Control Rods During the shutdown procedure no rod movement is permitted The capability of the RSCS .to pzo-between the testing performed above 20'o power and the rein-perly fulf'ill its function shall be verified by the followirg tests:

statement of the RSCS re-straints at or above 20'o Sequence portion Select a sequence power. Alignment of rod and atte pt to withdr w a rod in the groups shall be accomplished remaining sequences. Hove one zod prior to performing the in a sequence and select. tne remain-tests'.

ing sequences ard attempt to move a rod in each. Repeat for aH.

Whenever the reactor is sequences.

in the staztup or run nodes Group notch portion Foz each of the below 20X rated pover the six comparator circuits go th ough Rod worth Hininizer shall be test initiare; conpazato inhibit; operable oz a second licensed verify; reset. On seventh attempt operator shall verify that test is allowed to continue until

'the operator at the reactor completion is indicated by console is following the illuntixmtion of test complete light.

control zod program.

A second licensed operator b. The capability of the Rod may not be used in leiu of North Hinini er (R'~N) shall ho van< 44 iA 1 eo s\,

the RWM during scram time cheeks.

iota LoIJVVlJJlp, testing in the startup or run modes below 20 percent 1 ofrated thermal po~er. The correctness of the control,rod withdzawal sequence input to the R'~N conputar shall be verified before reactor sta'tup oz shutdown.

If Specifications 3.3.3.3.a 2. The Rli~ computaz on line through .c cannot ba net the d'agnostic test, shall 'oe reactor shall not ba started, successfully perfozned.

oz if the reactor is in the 3. Pr'or to staztup, proper run or startup nodes at lass annunciation o" the selec-than 20K rated powez, it shall be brought to n shut- tion error of at least one down condition i~ediatsly. out-of-sequence control rod shall be verified.

4. Prior to staz up, the zod block "unction of the R~H shall ce verified by nov~eg an out-of-sequence cortzol rod.
5. Prior to o'otain'rg 20K rated power during zod inset"ion at shutdown, vez "y th latch'".g o" the proper zod l23 gzoup and pzcpez annunc'a ion after 'rsert errors.

T ter. CnVti t T rO4S FOR 0FER;n XOV SLJRVFILLhhtCg RtgU(RE>>, 4TS 3.3.B'ontrol Rods 4.3.8 Contro'L Rods

4. Control rods shall not be When required,, the prasenco withdrawn fox startup or. ipf,a second lgcetssed operator refueling unless at least )o verify the folle~ing ot two source range c:hannels <he cqrrect rod prograa sha'Ll have an obsexved c:ount rate Pe,ver ified.

equal to or greater than

~

three counts per second,. Pripr po,coqtrol,rod withdraua)L for startup or during refueling, 5..During operation with verify, that at least two,source limiting control rod pat- range channels have aa ob,sex'vecL terns, as determined by the count rate of at leattt three designated qualif led person- counts per second.

nel, either:

a. Both RBH channels shall ,5. When a limiting control rod be operabla: i petcicrn exists, an instnr ant or functionaL test of the RiVA shall be perfo wed prior to b Control rod withdrawal withdrawal of the designated shall be blocked.. rod(s) and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

C,. Scram Insertion Times

  • 1.After each refueling out'ag'e 'al,'1 operable rods shall becr'am'~

tested from the fully 'withdrawn~

position with the nucl'ea'r above 950 psig '(with

'sy,'stem'ressure saturation temperature).

shall be compl'et'ed prior to This';esting, C. Scram Xnsertion Times exc'ceding 40% power. Below 20%

occam esses powe'r',, only...rods in those sequences

l. The average scram insertion time, based on the deenergi-(A'12 a d A'34 'o B12 'd B were fully wit:hdrawn in thigh region

. ) which xation of the, scram pilot valve from 100% rod density to 5'0% rod solenoids as time zero, of all density shall be scx'am time t:ested.

operable control rods in the The sequence restraints;imjiosed upon reactor power operation condi-tion shell be no greater than: the control rods in the 100-50 percent rod density groups to t.he.

X Inserted From hvg. Scx'am Xnser- preset, power level may be removed, Full, ixhd W t ion T in e c) by use of'the 'in'dividual bypass switches associated wilth 'ftocie 5 0.375 control rods which are'ully 20 0,. 90 ti ally withdrawn and ard xtot 'ox'ar 50 2,.0 within t he 100-50 pex'cent xod, density I

90 3., 500 groups. In order to bypas] a rod, tlhe actual rod axial pcI s/ tion m'us't be known;, and the rod must'. 6e in'he.

correct in-sequence position.

LL'IITIHC CONDITIONS FOR I)PERATION SURVEILLAHCF. Rg UIR&fEHTS 3.3.C Scram Insertion T'mes 4.3.C Scram Insercion Times

2. The average of t.le scram inr r At 16~~rek intervals, 10X of t'av c tion times for the three fn! cesc operable contz'ol rod drives shall operablc control rods of aij be scron timed above 800 psig.

groups of four control rods in Whenever such scram 'time measure-a tuo-by-tuo arrcy shall be no ments are made, an evaluation greater chan: shall bc made to provide reason-able assurance that proper con-X Inserted From Avg. Scram Inser- trol rod drive performance is Full Wi chdr aun tion Times (sec) being maintained.

5 0.398 20 0.954 50 2.120 90 3. 800 The maximum a;ram Inscrtic i time for 90X Insertion, of !ny operable control rod shall not exceed 7.00 seconds.

D. React ivitv Anomalies D. Reac t ivit Anomalies The reactiv'ty eq divalent of During che scarcup test program and the difference br;:veen che actual startup folloving refueling outages, critical rod configuration and the the critical rod configurations vill expected configuration during pouer be compared to the expected confi-operation shall not exceed 1X 5k. gurations at selected operating con-If this limit is exceeded, the ditions. These comparisons uill be reactor vill be shuc dovn until the Used as base data for reactivity cause hss been determined and cor- monitoring during subsequent pouer rective actions have been taken as operation throughout the fuel cycle.

appropriate. At specific pouer operating condi-tioos, the critical rod canfigura-tion vill be compared to the confi-guration expected based upon appro-priately corrected past data. This comparison uill be made at least every full paver month.

12S

LIHITINC CONDITION.'i FOR .OP."",.RATION'tiRVVILLANCF,',REOU iKt~>HFNTS 3;3 4. 3 Rcai:tiv~it Control If Specificatioria -3. 3.'. and,D

'above 'cannot: .be, me,t, at> o'rderly ahutdo~ ihall be ,initiated arid, the reacto'r chill be 'iti the ihutd~ condition L~ithin 24 houra.

126

3.3/4 3 BASES>

30 Th" Rod Worth M5nimizer,(P~M) and the Rod Sequence Control System (RSCS) rc::trice withdrhva!s and 5nscrtions of control

.rods to pzc-spccifl<<d scqucncc... Ail patt~ms asrociatcu with these ccqu<<news have the claractcrl.".tlc that ~ as"uming the worst single deviation from the sequence, the drop of any control rod from the fully inserted position to the position of the control, rod drive would not cause, the reactor to sustain ~

a power excursion resulting in any pellet average enthalpy in excess of 280 calories per gram. An enthalpy of 280 calories per gram is well below the level at which rapid fuel dispersal could occur (i.e., 425 calories per gram). Primary system damage in this accident is not possible unless a significant amount of fuel is rapidly dispersed. Ref. Sections 3.6.6, 7.7.A, 7.16.5.3, and 14.6.2 of the 'ESAR and NFDO-10527 and supplements thereto.

In performing th fdncticn desczibed above, the K~M and RSCS az>>

not required to impose any restrictions at core power levels in excess of 20 percent of rated. Yaterial in the cited rcferc."c shows that it is impossible to reach 280 calories per gram in th event of a control rod drop occurring it power greater "than 20 percent, regardless of the rod pattern. This is true for all normal and abnormal patterns including those which maximize individual control rod worth.

ht power levels below 20 percent of rated, abnormal control rod patterns could prcduce rod worths high enough o be oi concern ral-tive to tl;e 280 calorie per gram zcd drop limit.

'Xn this range the kW and the RSCS constrain the control rod sequences aud patterns to those which involve only acceptable rod worths.

The Rod North Minimizer and the Rod Sequence Control System provide automatic supervision to assure that out of sequence control rods will not be withdzawn or inserted; i.e., i" limit.

operator deviations frcm planned withdrawal sequences. Rcf.

Section 7.16.5.3 of the CESAR. They serve as a backup to proccdui..~

control of control rod sequences, which 1'm't the maximum reacti-vity worth of control rods. In the event that the Rod cnorth Minimizer is out of service, when required, a second licensed operator can manually ful ill the control rod pattern con-formance functions of this system. In this case, the RSCS is back up by independent procedural controls to assure conformance.

  • Because it is allowable by bypass during scram time testing below certain rods in the 20 percent of RSCS rated. power 'in. the startup or run modes, a second licensed. operator is not an acceptable substitute-fir the RWM during this testing.

129

I The functions of thc R'WN and RSCS make Lt unnecessary~ to~

spccif y a 1 iccnsc limit on rod wor th to preclude unacccptablc consequences in thc event c f a control rod drop. At low powersbelow 20 percent, these devices force.adherence acceptable rod pattern . Above. 20 percent of rated power,

'o no constraint" on rod pattern is rcnui'.rc'd to assure that rod drop accident consequence. are acceptable. Cont'rol rod pattern constr-iints above 20 percent of rated power are imposed by powe.r distribution requirements, as def ined in Sect,ions 3.S;I, 3 .5.J, 4..'I.I and 4..'i.J of these technic~al ~

specifications. Power level for automatic bypass of fainction is sensed by first stage turbine pressure.',

the'SCS The Source Range !<onitor (SRM) syst.em performs no autoamtic safety system function; i.e; it has no scxam functioiw It

'I 30'

/C,3 llAS S:

provide he op ra or wi h a visual ind'c cion of neu-tron level. The consequences o: re'ctivi ty ac<<iderts are functions of the iritial nc~ tron fl~z. The rcquiren:ent of least 3 counts per second assures that a"..y transient, should it occur, begins at or above hc initial value of 10 of rated power u d in the analyses of rans'ants fron co'd con'ditions. Ona operabie SRH channel vould be adequate ~

to monitor th approach to. crit'celity using homogeneous patterns of scatter d control rod vi hdraval. A mini,. um of two operable SRH's are provided as an added corae latish.

Tha Rod,Block Monitor (RBH) is designed to auto at'cally prevent fuel damage in the event of erroneous rod wi hdraval, from locations oi high poMer density during 'high power level operation. Two channels are provided, and one of these may be bypassed froa the console for vain cnance and/or testing.

Tripping of one of the channels Mill block erroreous rod vithdraMal soon enough to prevent fuel damage. The speci-fied restrictions Mith one channe'ut of ervice conserva-tively assure that fuel Canape vill not occur due to rod vithdraval "rrors Mhen this conditfon exists.

A limiting control rod pattern is a pattern which iresults in the core being on a thermal hydraulic limit, (ie, MCPR given by figure 3.5.3 or LHGR of 18.5 for 7x7 or 13.4 for 8x8) During use of such patterns, it is judged that testing of the RBM system prior to with-drawal of such rods to assure its operability will assure that improper withdrawal does not occur.

,It is normally the responsibility of the Nuclear Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns. Other personnel qualified to per-form these functions may be designated by the plant superintendent to perform these functions.

Scram Insertion Times The control rod system is designated to bring the reactor subcritical at the rate fast enough to prevent fuel damage; ie, to prevent the MCPR from becoming less than 1.06. The limiting power transient is given in Reference 1. Analysis of this intransient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above specification provide the required pro'tection, and MCPR rema'ins""greater than 1.06.

On an early BWR, some degradation of control rod scram performance occured during plant startup and was determined to be caused by 131

3. 3/4. 3 BASK.'i:

part LCulate material (prehably COnatruCtiOn debriS) p."urging an

.internal ciontrol rod drive filter. The ideaign of the present i

control rod drive (Model 7RDB144B) 's'r'oesly improved by the relocatfon of the fil'ter to a location oiut 'of'the scram driVe f.c., it t:an no longer inrezferie vfth scram ,'ath:

if completely blocked.

performan'ce,'ven The degraded: pet formance oL'he 'original'rive (CRD7RD8144A) u'nder dirty operating; conditions and the fnsensitivity of the drive: (CRD7RDB144B) has been demonstratid by a 'edeeigned seri.:e of engineering teste unde'r simula'ted'eactor operating c'ondLtions. Thai eucccsefuJ. performance of the ncv drive conditions has oleo bein dempnetrated by under'ctual'perating conetstcnt'.y. good lLn-service test results For. plants using the

-ncv, Irive and may be inferred fr'om'pliant'e u'sfr'ig the, older aedel drfv vf'th.a modf fied (larger screen sire) int:ernal filter

~

1s prone to p]L'uggfng. Date hais- been documented by surveiil- vhi'ch'o:

i

)anc reports in various operating plants. These include Oyster Creak,, Monticallo, Dresden 2 and Dresden 3. Approxfmat'el

5000 drive tee'ts have been recorded to date.

Folloving ILdentiffcation of the "plugged filter" problem, very fre'quent: scram tests vere necessary to ensure proper the morc frequent'cram tests are nov considered totally',

performan'ce.'ovever, unnecessary and: unvise for tlie follovfng reasainst l'. Erratic scram performance has be'en identified as due to an o'bst'ructed drive filtet in type "A" drives. The drives in B'FNP ar'e of'he nev. "B'" type design vhose scram perfairmancc ie unaffected biy filter condition.,

2: Tlhe dirt load ie primarily released during startup of thc reactor vhen thee reactor and ite. systems are fir'et subjected to f love iand prceo>>rc and thermal strueeee. Special atten-tion, and mcaoureo are nov being taken tio assure cleaner system>i. Rei ctore vithi drives identical or similar (shorter at'rok'e smalJ et p Leton areas) have. operated through many refueling cyc les vfth no sudden or erratic changes in scram ierfor~mnce. This preoperational and startupi testing is wsfffcLent to detect anomaious drive perfonnance.

3. he 72-hour outage limit vhieh 'fnftiatcd'he start of the
requent'crew teoting ie arbit'rat'y, 'ha~zinIt n'o Jogfcal basis other than quintifying a 'majIor outage" vhfch might. reasona-bly be -cnuscd by an event eo eckcr'e ae to possibly affect drive pcrfori~nce., Thfe r'eqofrccment fs uni~fse because it provides',an- incentive for shbrtkut'c!tions'o hasten returtif sQ "on, line'l to avoid the additional 'testing due a'72-'-holur outagl .,

'32

The surveillance requirement for scram testing of all the control rods after each refueling outage and 10X of the control rods at 16-Meek intervals is adequate for determining the opera-bility of the control rod system yet is not so frequent as to cause excessive Mear on the control rod system components.

The numerical values assigned to the predicted scram perfor-nance are based on the analysis of data from other BWR's with control rod drives the same as those on Brogans Perry Nuclear Plant .

The occurrence of scram times within the limits, but signifi-cantly ion".,er than the average, should be vie~ed as an indica<<

tion of systematic problem Mith control rod drives especially

'if the number of drives exhibiting such scram times exceeds eight, the alloMable number of inoperable rods.

In the analytical treatment of the transients, 390 milliseconds are alloved betveen a neutron sensor reaching the scram point and thc start of negative reactivity insertion. This is ade-quate and conservative Mhen compared to the typically observed time delay of about 270 milliseconds. Approximately 70 milli-seconds af ter neutron flux reaches the trip point ~ the pilot scram valve solenoid pouer supply voltage goes to zero an approximately 200 milliseconds later, control rod motion begins.

The 200 milliseconds are included in the allowable scram inser>>

tion times specified in Specification 3.3.C.

  • In order to perform scram time testing as required by specification 4.3.C.l, the relaxation of certain restraints in the rod sequence control system is required. Individual rod bypass switches may be used as described in specification 4.3.C.1.

The position of any rod bypassed must be known to be in ac'cordance with rod withdrawal sequence.

Bypassing of rods in the manner described in specification 4.3.C.1 will allow the subsequent withdrawal of any rod scrammed in the 100 percent to 50 percent rod density groups; however, maintain group notch control over all rods in the it will 50 percent density to preset po~er level range. In addition, RSCS will prevent movement of rods in the 50 percent density to preset power level range unti'1 the scrammed rod has been withdrawn.

133

3.3/4.4 BASES:

D. Reactivit~Anomal:Les During each fuel cycle excess operative reactivity varies a.I fuel depletes and as any burnable poison in supplementary control is burned, The magnitude of this excess reactivity may be inferred from the crit:Leal rod'onfiguration. As .fuel burnup pro-gresse:, anomalous behavior in the excess reactiv:Lty may be detected by comparison of the critical rod pattern at- selected base states to the predicted rod inventory at that state. Power operating base cond:Ltions provide 'the most sensit:Lve and directly interpretable data relative to core reactivity; Furthermore., using power operating base conditions permit. frequent reactivity comparison..

Requ:Lring a. reactivLty comparison at the specified f requency a.ssures that a comparison will be made before th e core react:Lvity change exceeds 1% ~5 ~ ~

Devia tion s in core reactivity greater than 1/Ag are no t exp ec ted and requ:Lre thorough evaluation. One percent eactivity into the core wou.Ld not lead to transient s exceeding de.,ign conditions of the reactor sys t em,.

References 1., General Electric BWR Reload 1 Licensing Amendment for BFNP unit 1 NEDO 24020'ay 1977 134

I.IStYIt'C: Cn:tnt rtn>>>> Fctg opera tttR StlRVE) t.t.aictc rtf; tt IRFCF!tTS

3. 5. F R<<nc t nr Cnr<<

~ I so la t ion Cool in' 4.5.F Reactor Core Isolation Cooline lf thc RClcs i~ inopc rab'c, 2. When it is detcrrcincd that thc ttcc reactor may rcm.".i.n in RCICS is inoperablc, thc ltPCTS operation for a period not shall be demonstrated to be to exceed 7 days if thc operable i~~ed iateLy and ueekly HPC?S is operable durinR thereafter.

ccuch time.

3. If speci f iearions 3.5.F.l or 3.5.F.2 arr not mct, an orderly shutdo~ shall be initiated and thc. reacto.

.shall bc dcprcssurizcd to less than 122 pc:IR uithin 24 hours.

C. Automatic Drorcssccrization G. Automatic Ucnressurlaation

1. Four of t'h e six valves of 1. Durinp each operating cycle the Automatic Depressuri- the following teats s'hall be ration System shall be performed on the ADS:

operable:

a. A simulated automatic

'(1) prior to a itartccp actuation test shal.l be froccc a Cold Condition, perfoxwcd prior to startup or, after each rcfcceiinp out-age. manual surveillance (2) uhenevcr there is irra- of thc relief valves is di.ected 'fuel in the reac- covered, in 4.6.D.2.

tor vessel and the reactor vessel prcssccrc ia preatcr than 105 psiR, except ao specified in 3.5.C.2 and 3.5.C.3 below.

2. If more than two ADS valve axe 2, When it is determ',ned that more than known to be incapable of two of the ADS valves are automatic operation, the of automatic operation. theincapable Ht CIS reactor may remain in ooera- shall be demonstrated to be operable tion for a period not to ittmediately and daily thcrea ter as exceed 7 days, provided the long as Specification 3.5.6.2 HPCI system is operable. applies.

(Note that the pressure relief'unction of these valves is assured by section 3,6.0 of these specifications and that this specification only applies to the ADS function.)

157

't.lHLTLSC Cn'll>lTTONS FOR OPRPATLQH SllRVF ILL.ANCF. RV~JLRr.NFNTS 3.f.G Autn~ntic Dcnre~siiritat ion 4~5~C AutoNiit ic A~i~ ressuriaationi

~Ssti in (Al)S) Svstee ~AEiS)

(

3.. lf specifications 35.G.l and 3 5.G 2 cannot be acts an orderly ahuldoMn vill be initiated nnd thc reactor vessel prcssure shall be reduced to LO5 psLS or less ui tl>in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Malnr.n .. of iliad1 Disch~ar e H. 'aintenIinc'i: of 'F'illed Pisch~ar ~c

~PJ e P i~e Mhcncvcr thc core spray systems, ThIe t'ol loving surveillance rcc;uire-1.'VCI, llPCl, or RCK'.are requi',red ments shall bc adhered to to,issu e to l>c opcrnblc, thie discharRe 'ha t thc discharge pip inl; of the piping froiii the puep discharge 'ore'spray system'sp QPCQ'y /PCS and af these .systems to thc labt RCIC 'are .filled:

block valve shall bc fillccl.

15S

XHITINC CONDITIONS FOR< OPFRhTION SURVF.ILUNCE kl'. NInNHP.NTS of Filled Dischar Pine 4.5.H Maintenance of Filled Diachar Pie~

~ ~e Maintenance

.H suction of the RCZC and HPCZ pumps e

l. Every month prior to I

the e

t sting

~all be aligned to the conde..sate of the RHRS,(LPCI and Containmen storage tank, and the pressure suppres- Spray) and core spray systems, the sion chamber head tank shal1 normally discharge piping of these systems be aligned to serve the discharg piping shall be vented from the high point.

of %he RHR end CS pumps. The condensate'ead and water flow determined.

tank may be used. to serve the RHR ,

and CS discharge piping if the PSC head 2. Following any period where the LPCI tank is unavailable. The pressure or core spray systems have not been indicators on the discharge of the RHR required to be operable, the dis-and CS phnps sha11 indicate not less charge piping of the inoperabls sys-than 1isted below. tea shall be vented frcna the high P1-75-20 48 psig point prior to the return of the Pl-75-4S 48 psig system to service.

P1-74-51 48 psig Pl-74-65 48 ps g 3~ Whenever the HPCI or'CIC system is up to take suction from the 'ined I. Avera e Planar Linear Heat Generation condensate storage tank, the dis-Rate chazge piping of the HPCI and RCIC During steady state power operation, the shall be vented from the high point Maximum Avezage Planar Heat Generation of the system and water flow observed Rate (MAPLHGR) for each type of fuel as on a monthly basis.

a,function of average planar exposure shall not exceed the limiting value 4. When the RHRS and the CSS aza r,e-Tables 3.5.I-1, -2i -3 6 -4 ' quized to be operable, th'e pressure Tf at any time during opezation it is indicators which monitor the dis-

'~etermined by normal surveillance that charge lines shall be monitored the limiting value for APLHGR is being daily and the pressure recorded.

exceeded, action shall be initiated with-in 15 minutes to restore operation to within the:prescribed limits. If the APLHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown, condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Maximum Average Planar Linear Heat Qenara-Surveillance and corresponding action Cion Rate (MAPLHGR) shall continue until reactor operation The MPLHGR'or each type of f el a< a func-is within the prescribed limits. tion of average planar exposur shall be detemnined daily during reactor operation J'. Linear Heat Generation Rata LHGR) at > 25Z rated thermal power.

During steady state power operation, the linear heat genezation rate (LHGR) of J, Linear Heat Generation Rate.

any rod in any fuel assembly at any (LHGR) axial location shall not exceed the The LHGR as a function of core height shal.

maximum allowable LHGR as calculated by the following equation: be checked daily during reactor operation at

> 25X rated thermal power.

159

T.INITlhG COEt JLTIOHS FOR OPP!tAT ON S VRVF. ILLANCF. RFAAUIRFMFNTS

< LHGR. (1 - @p/p)

LHGRi max d 'ax (L/LT,'ij LHGRd Design LHGR - 18.5 kV/ft. for for Sx8fuel 7x7fuiel'3.4

%W/ft 0 P/P) ~ Maxpnpm pqqpr spiking p'enalty

= 0. 022 for SXS .fuel Total core lengt:h 12,0'. feet for 7X7 1'uel

=:, 12.2 feet for 8XS fuel, L Axial position a'bove bottom of core If att any time during ojeration it is deter-mined: by normal surveil.l.ance that the for LHGR is 1being exceeded, 1'imiting'alue, initiated'ithin 15 'minutes to action restore shIll'e operation to within the prescribecl limits.

If the LHGR is not'etui.ned t:o wit:hin the pr'escribed limi.ts within two (2) ?{ours, th' reactor shall b'e brought: to the Cold Shutdow'n condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresp'onding action shall continue until reactor operation is wit'hin the pres< ribed limits.

K. Minimum Critical Power. Ratio (MCPR) Minimum Critical Pc~e Ratio The MCPR operatin'g limit for BFNP 1 is now .{BCPR) dependendent upon t?te aver'ajge exposure for The MCPR operating limit for the 7x7 th{'ore.

MCPR s'hall be det e ie 11.

fuel asseinblies is in general different from during reactor p r P ra ion at,

'that for the SxS aseiemlbliesFor steady state > 25X rated theim B i P r f power, operation the value, for 'MCPR will be as lpwgng nny change 0 r ol-'evek o,r shown in Figure 3.5.3. These v'alu'es of MCPR distribution th" t w u d 4 441 oper4 are'. for operation at, rited pow'er and f'low. tion w:Lth 4 limitin on r 1 rod For .core flows other than rat'ed the MCPR shall pattern as descri b n: h b)oes fc be greater thin MCPRimes Kf, where MCPR's Specification 3.3 the appr'opriate value'rom Figure 3,.5.3 and Kf Xs as shown in Figure 3;5.2.

If'at any tom'e during op'eration it is.determined by norm'al surveillance that lgmiting value for MCPR i. being exceeded', 'hi ection shall be init:iated within 15 minutes

restore operation tc~ within the prescribed to limits. If the stea{dy state ?'lCPR is 'not ret'urned to within t: he prescribed lim:Lts within two (2) hours, the r'eactor shall be brought to the Co'ld Shutdown cond:l.tio'n with.'Ln 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />,.

Surveillance and c'orresponding action shall Continue until reactor operatiori is wi{thin the 'prescribed limits.

L. Re ortin Re uirements If any of the 'limiting vaLues identified in Specifications 3.5.I, J', or K are exceeded and the specified remedial action:Ls taken, the event shall be logged and ii'eported in a 30-day written report.

~ i 160

3.5 BASKs 3.$ 'C'utomatic De ressurisation .S stem (Ans This specification ensures the operability of the ADS under all condi-tions for vhich the depressurisatlon of the nuclear system io an esoen tisl response to station abnormalities.

The nuclear system pressure relief system provides automatic nuclear system depressuriration for small breaks in the nuclear system so that the lou-pressure coolant in)ection (LPCI) and the core spray subsystems can operate to protect the fuel barrier. Note that this specification applies only to the automatic feature of the pressure relief system.

Speci&cation 3.6,b specifies the requirements for the pressure relief function of the vnlves. It is possible for any numbor of the valves assipncd to thc hOS to be incapable of performing their 'ADs functions because of instrumentatinn failures yet ba fully cipable of performing their prcssure reiief function.

Because the automatic dcpressuriaation system does not provide makeup to the reactor primary vessel, no credit ie taken tor the steam cooling of the core caused by the system actuation to provide further conservatism to the CSCS ~

With two ADS valve known to be incapable of automatic opera'tfott, <<<<

valves remain operable to perform their ADS function, The ECCS loss-of-coolant ace)dent, analyses for small leone breaks assumed that <<<<

of the s1x ADS valves were operable. Reactor operat)on w)th more than

~s Valves )noperable 1s only allowed to continue for seven days provided that the HPCl system )s demonstrated to be operable.

y. 5 I ASV.S 3.5.II Natntcnancc of FII led I'lisc~har>> Pi~e, IE the dtscliarge piping of the core Spray, LPCI. IIPCILS, and RCICS arc not filled, a voter hammer can develop in this piping ~hen the pump, a@4/q!r pumps azc started. To minimisc damage to the discharge pipirig and to ensure added siargin in t'e operation of these systems, this Tcchnical Specification requires the discharge lines to b>> filled vhenevcr the system is 'in sn poses'~

operable.cor'idition. If a discharge pipe is not filled,;the pumps that supply that line must be assumed to be inoperable for,Technical Specification pur-The core spray and RHR system illscharge piping high point vent is vlz~ua'.Lly checked for untcz floM once a month prior to testing to ensure that the lines azc filled. The visual checking Mill avoid starting; the core spray or RIIR system vlth a discharge line not filled. In addition to the visual observzltion and tio ensure a filled dlschargei line otl~er than prior to testing,

. a pressure suppression chamber head taztk $ s located approximately 20 Sect above the discharge line highpoint to supply !malt>>up iizLter for these syst!ems. The condensate head tank located apprqx+agely 100 i'eet above, the Qsqhpge high point serves as a backup charging!system yhein the pressure suppression chamber head tank is not;in service. System! di!scourge pressure irqdicators are used to determine the vater 1evel above t4e !dizlchtLrge line high point. The indicators villre.'H.ect approximately 30 psig, 'for a water leve1 st the high point and 45 psig i'or a vater level in the pressureztuppression chamber head tank and ar n-itored dai1y to ensure that the df.sch~gcilittes at.e filled.

Mhen in their normal standby condition, the siiction For the llPCI ind RCIC Izumps arc allgricd to thc conitrnsatc storigc tank, Mhlch is phystcally at a higher elevation titan thc IIPCIS and RClCS pipinI.. This asisures theat the IIPCI and RCIC discharge pipini, rci,sins filled. Further assurance is provided by observing Miitnr flier fzom chase systems high points monthly.

3.$ .Z. thiximum /overage Planar Linear Heat Cisneratioiz Rata ()9LPLHCR)

This specification assures that the pclak~clpdding temperature folloMing tha postulated design basis loss-of-coo'Lant accident Mill not. exceed the lilsit SpeCified in the lQCFR50hppCndi'X K.

The peak cladding tiemperature following a postulated lois-of-coolant acci-dent is primarily a funct:ioia of the, average, heat generation rate of all the rods of a fuel assembly at any axial location and is. only'i pendent second-arily on thc rod to rod power djstributicin Mithin an assembly. Sing e "ex-pected local variations in poMcr diptribution uithin a fuel asseml>Iy affect the calculated peak clad temperature by gcsti tihan 4 20 P,relative to the pealz temperature Eoz' typical fuel, design, the limi.t on thc nvcragii. linear heat generation rate is suEficicnt to,assure that: calculated tcmpciatures are Mithin the lCCHI50 Appendix K limit. Thc limiting value for NAPiUIQR t

ia 0hz'n Tables 3,5.$ -1, -2, 3 r 4

i l68

l. Q Ot;NINL'ftONS (l;ont'dp l ht least one door in each access opening is closed.

2~ The standby gas treatment system is operable.

3 hll Reactor Building ventilation system automatic isolation valves are operable or deactivated in the isolated'osition.

for.a particular unit and the end of the next subsequent refueling outage for the same unit. I Refuelin Outa e - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and t'e staztup of~

the unit after that refueling. Foz the purpose of designating frequency of testing and surveillance,,a refueling outage shalL .

I mean a regulazly scheduled outage; howeve., where such outrages occur. within 8 months of, the completion of the previous refueling

~

outage, the required surveillance testing need not be performed .

until the next regular'ly scheduled outage.

Alteration of the Reactor Core The act of moving any component in the region above the core support p1.ate, below the upper gxid and within the shroud. Normal control rod movement with the control. rod drive hydraulic system is not defined as a core altexation. Normal movement of in-core instrumentation and the:traversing in-core probe is not defined as a core alteration.

T Reactor Vessel Pressure -'Unless.otherwiee indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.

Thermal Parameters

l. Minimum Critical Power Ratio (NCPR) Min&mum Critical Power Ratio HCPR) is the value of, the criticalpower ratio asso>>

ciated with the most limiting assembly in the reactor core.

Critical Power Putio (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.,

~ ~ ~

2 Transition Boilin - Transition boiling means the boiling, regime between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intexmit-tently with neither type being completely stable.

3. Core Maximum Fraction of Limiting Power Density (QPLPD) The highest ratio, for all fuel types in the core, of the maximum fuel rod power density (kV/ft) for a given fuel type to the limiting fuel rod power density (kN/ft) for that fuel type.
4. Avera e Planar Linear Heat Generation Rate APLHGR The Average Planaz Heat Generation Rate is applicable to a specific planar height,and is equal to .the sum of the lineaz 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.

DEFINITIONS . (Conit ')

V yXnstrumentatiori

~c means t ei Instrument Calibration Ax< iI>stzu:<<ent calibration ad)ust<<<'enact 'of an'nstrument signal output s'o tha't it corr'lhspQnds, vith'in cacceptabl.<x range, and accuracy, to a knoMn va1ue(s) o f he, parameter Mh'ich the instrument monitors.

2~ Cfiarinel. - A channel is an atragg<<me4t 'of'a sensor snd used to "ekaliuaxie plant'ariables'nd asso-'iated'o<apon'eiits,

'diicrete outpu<::st used,in .'lo jic'. iA channel .terminates p'o-',d<<cce

'ay'ncaa'1~oiacst.'its identity vh'ere individual 'channel output's'ar<<<

cc<mbina'ed'!Ln'log i'c'.'g 3~ Instrument Functional Test -,'iQ instrument functional telst mehnsi tl<e';in)ection of a 'simu'la'ted 'siig'Wl into 'h'e ins4ru'ment'rimary sinasoar 't'o verify thai'roper instkmu'int channel iaspcinse, i a1Ianh e'nd(or initiati'ng'c t'ion'.

.Instrumint Check An <Lns'tr<imeht!chick i6 qualitati've dete6aiha>

.tion"',ofcceptable opirability by'bsexvation of duri'ng'peration. 'This" 'defer'nination el>all includ'e, instrument'eihavior ciimpcir'i<son df'. She instrument vith'ther.jLndepln- 'vs<sr<s'osi>ibl."e",

.'diilntt init~iw'ms'nts 'iIiaaiur'in'elhi',sa'me'<yva'riable.

,5. Lo ic S stem Punctional.Tee<< -, A logic')st@m functional test

,asians'a'est 'of" all relays alnd contacts of a logic circuit 'to

'".insure'll'ioiponenti'ire o'pler<iblia Per'e!sign intent. 'her'e

'px'alcticabli, ."scti<ia'ilail'1 go 'to I'c&pl'ation; i.e., pumps vill be (started'<and valves operated.

,6. Tr~i ~S item - .A trip system, means, an arringement of instrum'ent

'cCiannel'rip signals and auxiilisr'y equipment requixed to inLtiatII<

.ac't'ion to a'cc'ot<<pl'iih a protective'rip function. x'ip system

'may requ'ire one or'o'e inst'ru<i'~en't c'ha<'inel trip signals x'elILt~d

'to oite or 'more plant paraineters in order to initiate t'riP l'nit'iatibn of protective section majr xequire the trIip)inI<', sist,am'action.

of' singl'e.It'r'ip i<ysteji< ow the coincident trip'ping of tee~ t'x'ip sys tt<ms ~i it 7a Protective .A<.tion -'n aetio& init:iated b'y the protection system vhan 'aLait is'eacIlied. A pxotective action can be at a channel or,ski stem lavel.

.8. Protective Function - A system,protective action vh4ch prote'ctive sctiont o'f the ch'annelid monitoring a parti" results'-from"the.

cul'ar'plant c:ondition.

9. Simulated Auta<Latic Acxuatiod ~ SILmulat".ed'u'tomas'ic 'actus'tibn ayplyttict,~ ~ ttailattaataaal cc ttie aacact ta <<ctcata I.ta cir'ccuit int question.

'aaaaa 0 tap, t

FOE). CL(il)DTWG JHTEGllT'fY In the event of operation with the core maximum fraction of limiting power density (APLPD) greater than fraction of rated thermal power (FPJ.'j the setting shall be modified as follows:

6~ (0.66W + 54X) FRP CMFLPD For no combination of loop recircu-lati'on flow rate and core thermal

'ower shall the APRN flux scram trip setting be allowed 'to exceed. 120%

of rated thermal power'.

(Note: These settings assume operariF~.

within the basic thermal hydraulic design LHGR criteria. These MWD/T; 1.29 otherwiseT.

it is determined design

'f criteria 18.5 1;w/ft and MCPR > (1.2$

are that either of these criteria is being v'ol ted ii'8000 during operation, action shall be initiated within 15 minutes to restore operation within pr'escribnd limit: .

Surveillance requirements for APR::.

scram setpnint are given in specification ~6.1.B.

2. APRN linen the reactor mode switch is in the STARTUP POSITION, the APRM scr"m shall be set at less than or equal to 15/ of rated power.

3~

IRM The IRM scram shall be set at less than or equal to 120/125 of full'cale, E B. APRM Rod Block Trio Settin Core The~~al Power Limit (Reacto" Pressure <800 psia) The APRM Rod'block trip setting shall be:

';:? e.", the reactor pressu.e is less 1 th:.ni or ecual to 800 psia,

SAFFTY LItlIT I,'KHTTTNG . AFPfY'Y. TFPI SETTING 1 1 FUr>L CLADDING INTEGRITY 2. 1 I'UEL CLADDING INTEGRITY or core coolant flow is Less Rg< (oi66W + 42")

than 10@ of rated, the core thermal powe shall not ex- uhesl c:

ceed 823 ?51t (about 255 of rated thermal power). RB

= Rod block setting is percent of rated thermal power (3293 Mlt)

W = Loop recirculation fiow rate in percent of rated (rated loop recircula).ion flow rate equals 34.2 X 10 'b/hr)

Xn the event of operation with the core maximum fraction of limiting!pointed d>jns'it~'p (CHFLPD) greater than fraction of,ra'tec?

thermal power (FRP) the setting shall be modified as follows:

SRB 0.66W + 42K)

C. Whenever the reactor is in C. Scram and isoluation ) 538 in. above the shutdown condit:ion with Ireactor low water vesse]L zerc> level irradiated fuel in the reac-tor vessel, the water level shall not be 1 ss t?Ian 17.7 in'. above the top of the D. Scram turbin, stop < 10 percent normal active fuel zone. valve cd'osiire' valve closure E. Scram turbine control valve Upon trip o 1: Fast closure the fast actin.-

.olenc>id valves

2. Loss of control. > '550 ps.ig o:il pressu e F. Scram--low con- . 23 inches denser vacuum Hg va< uum G,. Scram--main steam < 10 percen't

,line isolation valve closure H,. iMain steam isolation -" 825 ps4g valve closure--nuclear system low pressure l>0

BASF.S: FlJEL CLADDIllG INTF.GRITY SAFETY LIHTT Toe'uel c1 add ing represents one nt" the physical barr l e zs which separate radio-active materials Ezon environs. The integrity of this cladding barrier is

,related to its relative freedom from perforations or cracking. Although some corrosion or use>>related'cracking may occur during the life of .the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel" cladding perforations, however, can zesult from thermal stresses which occur from reactor operation significantly above design conditi'ons and the pzotection system setpoints. While fission product migration from cladding perfozmation is just as measurable asthat from use-related'racking, the thermally-caused cladding perforations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deteriora-tion. Therefore, the fuel cladding safety limit is defined in terms of the reactor operating conditions which can result in cladding peiforation.

1 The'. fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational transient. Because fuel damage is not directly observable, the fuel cladding Safety Limit is defined with margin to th'e conditions which would produce onset transition boiling (MCPR of 1.0).

This establishes a Safety Limit such that the minimum critical power ratio (MCPR) is no less than 1.05.'CPR )1.05 represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

Onset oE transition boiling results in a decrease in heat transfer from the clad and,'here'fore, elevated clad temperature and the possiblity of clad failure.

Since boiling transi'tion is not a directly observable parameter, the margin to boiling transition is calculated fzom plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution. The mazgin for each Euel assembly is characterized by 'the crgtical power ratio (CPR) which is the ratio of the bundle power which would produce onset of transition 'boiling divided by the actual bundle power. The minimum value of this ratio for any bundle in the core is the, minimum critical power ratio (MCPR). It is assumed that the plant operation is controlled to the nominal protective setpoints via the instru-mented vari'ables, i.e., normal plant operation presented on Figure 2.1.1 by the non inal ex~ecto~ Fl(w control lir o.. <ne Safetv Limit (HGP5l nF 1,AS) 4as sufficient con'sezvatism to assure that in the event of an abnormal operational transient ini<tiate~. Ezom a normal operating condition (MCPR )> 25) more than 99.9X of the fuel rods in t¹ ~

coze are expected to avoid boiling transition. The margin between MCPR of 1.0 (onset of transition boiling) and the safety limit 1.05 is derived from a detailed statistical analysis considering all of the uncertainties in moni-toring the core operating state including uncertainty in the boiling transition correlation as described in Reference 1. The uncertainties employed in deriving the saEety limit are provided at the beginning of each fuel cycle.

The MCPR value used in the ECCS performance evaluation '(1.18) is less limiting than the MCPR for operation (1.25).

l 15

.1. 1 BASES Because the boili'.ng transition correllation iS based on a large quantity of foll scale data there is a very high confidence that operation of a fuel assembly at the condit;ion of MCPR = 3J.05 vbuld not produce boiling tran-it sition. Thus, although is not req'uired'o establish the safety limit additional margin,exiets between, the safety limit and the actual ocAurlen5e of loss of clsddi.ng integrity.

However, be expected.

if boiling transition vere to occur, clad perforation would not Cleidding temperatures vould increase to approximately 1100oF which is below the perfora ion temperature of the cladding material. This has been verifie<i ny tests in the General El'ectric Test Reactor (GETR) whiere fue.L similar ii.n de.ign to SFi'6'perated, above the critical heat flux for a significant period of'",'ime (30 minutes) vithout clad perf'oration,.

If reactor pressure shou.'Ld ever exceed 1400 psis during, normal pover operating (the limit of applicability of the boiling transition it vou.'Ld be assumed that the fuel cladding integrity, Safety Limit corre-'ation) has been violated..

In addition to the boiling transition limit OlCPR <<:L.05 ) operation is constrained to s maximum LHGR of 1815 limit is reached when the Core ~1faximum Fraction of kw/ft,,'his Limiting Pover Dens:Lty equals 1.0 (C~iFLPD <<1.0).. For the ease cohere Core YAaximum Fraction of Limiting Power Density'xceeds the Fraction of Rated Thermal Pover, operation is permitted only at 1ess thin 100/ of rated power and only with reduced APRN .scram settings as required by specifi'cation 2.1.A.l.

At pressures below 800 psia, the core elevation pressure drop (0 power, 0 flow) is greater than 4.5i6 psi. At 1'ow povers snd flovs this pressu're differential is maintained in th<e bypass region of the core. Since the pressure drop iin 'the bypass region is~ es~sentially all elevatio'n head, the core pressure drop at; lov povers snd f3.ow vi.ll always be greater than 4.56 psi. Analyses show that vith s flow of'8210'> lbs/hr bundle f'low, bundle pressure drop:Ls nearly independent of bundle pover and has a value of 3.5 psi. Thus, the bundle flow with a 4.56 psi. driving head will,be greater tihan 28xl0 lbs/hr. Full scale.ATlLAS test data taken at pressures from 14.7 psia to 800 psi'.a indicate that. the fuel assembly.

critical power, at thiis flow is spproxiimately ,'3.35 iNt. With th'e factors this corresponds to a core thermal power of more than des'i'eaking 50$ . Thus, a core thermal power limit of 25@ for reactor press'ures belov 800 psia is conservative.

For the fuel in the core during periods vhen the reactor is shut down, cot<-,

sideration must also be given to vater level requirements due tO the'ffeht decay heat. Xf,viater level should drop belov'he top of the fuel'uring

'f this time, the ability to rcmove decay heat's r'educed. Tlhis r0ddction. in cooling capability could lead to elevated c'ladding temperatures snd clad perforation. As long as the fuel rmain: covered witlh water, s>xfficient cooling is available to prevent fuel cilad'perforatiion;

~ I aa))<a $ $$ $ I / I )I)

I I II j,ag" I ' 3 a a"

~ > ~ ~ a ~

I I

".'" .' i " ." ."'

', ' a1 ~ 'I

<<, ~ '1. ',$ I

';~ ~

' ' ~ ~ ~ '. !. '!

~ ~ 3'iV,.i'l . Tn>> . iai'u. war u; on plant. operatio i In accozdancI'ith the operating map given in Figure j."(-1 '.:.d p I'he FEAR. In addition, 3293 !Ãt is the licensed maximum power level o f BroMns Ferry Nuclear Plant, and this represents the aavimum steady-state po e'r which shall not- knqwingly be exceeded.

(',onsezyat sm is incbrporated in the transient ana'lyses in estimating the-cpn rolling factors ~ such as void reactivity coef fici nt, control'od scram scram delay time, peaking factors, hand axial po~er shapes. These factors are selected conservatively with respect to their effect on the analysis applicable transient ra suits as d'etcrmined by the current model This, transient model, evolved over many years, has been substantiated in opera-tion 6s a conservative 'too'r evaluating 'reactor dynamic perfonnance.

Results obtained from a General Electric boiling water reactor have been compazed with predictions made by the'model. The comparisions and results are summarized in Reference 1.

The absolute value Qf the void reactivity coefficient used in thc analysis is conservat'ivcly cstimatcd to be about 25K grcatcz 'than the nominal m3ximum value expected to occur during thc core lifetime. The scram worth used has been dcrated to 'bc equivalent'o approximately 80'f the total scram worth of thc control rods. Thc scram delay time and rate of rod insertion allowed w- 'i)>> ~~. >vcr ~ arc consoavativcly ct equal to thc longest delay and low-st insertion rate acceptable by Technical Specificatisns ~

The e'ffe'ct of scr~ worth, sczam delay time and rod inseztion rate, all conservatively applied', are of greatest significance in the early, portion of the negative reactivity insertion. 'The rapid insertion of negative reactivity is assured by the time requirements ror 5X and 20<'nsertion.

By the time the rods are 60K'.inserted, approximately four dollazq of negative zeac-tIvity has been inserted which strongly turns the transient, and accomplishes the desi"ed effect. The times foz 507.'nd 90X insertion are given to assure proper completion of the expected performance in the earlier portion of the tiansient, and to establish the ultimate fully. shutdown steady-state condition.

For analyses of the thermal consequences of the transients a HCPR of 1.25 (1.29 iS core average exposure is P 8000 MMD/T) is conservatively assumed to exist prior to initiation of the transients.

This" choice of using conservative values of controlling parameters and initiating transients at the design power level, produces more pessimistic answers than would.resul by usIng expected values of control parameters and analyzing at higher pover levels.

Steady-state operation without forced recirculation will, not be permitt d i'or more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

a 19

The licensed ma:<<'zimuii power, level is 3;293 MWt.

2. Analyses of t:ratiaients employ adequate!Ly conservative values of the co'ntiolling reactor parameters.

The; abnozmal operational t:ransients wex'e analyzed to a'ower level. af 3440'MT~

II 4,. The analytical .procedures aow used x,'esult ia, a more logical answer 'than alteinative method of assuming d. high~ei~ starting power in icon)uac~ 'he t'ion Mth the exp'ected va!Lues fox thle gaxIamkte'rs.'I The b'oses for individual set pointA ate <<did'cussed below'.

Neiitron Plux, Scram

,l. APRH High Flux. Sera~ Tr'ip Setting (Run kfode)

The avei age jiower'a<<age, m<i'ni,toring (APiM)~ shyster<<, which is calib'rated, using .heat b'glance data, ti'iken duriIriffsteady~state cond'.tions,',reads ia, perce'nt of, rate'd power. (3,293.HWt<<). B'e'crush fisSion rhambers px'o-',

v'ide'hj, basic gap<<ut', signals, t: he PERMyste'ci 'respo<<ids directl'y to neutron flux. Duiing trarisiIents, the i stan'ta'aeous rate, of 'aver'age

'There.ore;, during tianaients induced, by d~isturbances, the thetIx<<a

'f heat'ransfer fro<<ti'he fu'el (reactoz<<.thermal po'ver) is less than',the initantaneousi'-neutioia flux.due 'o tHe tMe tI.on'stant the fuel. i pmier of the fuel w'i;l.l be less than '-hat indicated the neutroa flic i*t the sciam set;ting. Analyses rhpoited. ~in Section by14 of the ~P~~al Safety, Analysis Rep'ort de'moastrat6d that with 'a 120 percha .,scia.i of the abao<<~l operational transients analyzed 'violat trip'ettingno'ne tlie',fue.'L safety lim'it and there is a substantial mirgin frcn flue).

damage. Th ref ore', use of' fl.ow-biased scram provides even addittioaal Mr@i<'igure, 2.l.i+ shows the flow'. biased. sczam as a function of care flow.

.<<l An incieaie 'in'the: A!~Rf scram setting would'de'creas<<<< tl'.e'margin before t: he fuel c1cddin" i:<<<<egiity s'aiety liMt is,reach'<<f pre-,'ent

~

scram setting was determined'y'n a'nalysis of margi s'.".qu'Lre'd The'APRON to provi'de a reasonable range for'sae<<'ave'ring during operation.

Reducing th's opez'sting margin would, increase tha, frequency of .spuxious scr'ax<<s, which have: an adverse e feet on reactor safety because of 'the

.re'sultirfg thermal stiesses. Th'us',, the'Ap'iN'setting was selected bac'ause "it'iovide's sda'quate margin. foI.. the fuel 'clrdding integrity

'sa!ety- limit yeast all<<~s operating'a'rgILn 'that reduces hn possibili; of unnocosllar y acr,~i ~

Ef&'r.5 The scram trip setting must be combination adjusted to ensure that the LHGR transient k 'is not increased for any of CHFLPD and FRP. The scram 2.1.A. 1 adjusted in accordance with. the formula in specification

~

'etting isC),'tFLPD 0

! when the exceeds FRP'.

Analyses, of the limei.!g, transients:show that no scram adjustment is required ii

,eo assure HCPR >l'05 when ehe transient Ss initiated from MCPR>> ~5 ('1 29 ~

core sversge exposure .is > 8000 WDlT).

,,ApR9 Flux Scram Trio Settin (Refuel or Start &'ot Scandb Node)

For operation in eh" stercup noCe while ehe reactor is at loM pressure, che APRM scram setting of 15 percent of raced po"e" provid"s adequsce thermal margin between the setpoint and the afety limit, 25 percent of raced. The margin is adequate co accommodate anticioated maneuvers associated with power plant startup. Effects of increasing pressure

at zero or low void content are minor, cold wacer from sources avail-abl e during -star up 's noc much colder thon that already in the system, t temperature coe ficients are small, and control rod patterns are con-strained to be uniform by operating procedures backed up by the rod

~>reh.miniuizer snd the Rod Sequence Con ro'ystems. Porch of indivi-dusl rods is very low in a uniform rod p" ccern. Thus, all of possible sources of reactivity input, unifo. control rod vf chdraval is the moot probable cause of s',",nificant power rioe. Because t'e flux distribu ion associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a sires.ficanc percentage of rated power, th>> race of power rise is very slow. Cenerally, the heat, flux is in n.nr quilibrium with the fission rate. Zn an csau=ed unifor rod uichdrawal approach co ch scram 'ev 1, ehe rate of power rise is no*more c"..an 5 percent of rated power per minute, and tne APRN system.

wou}d be more chen adequate eo assure a scram before the pover could exceed che sa"ecy limit. The 15 percent APK". scram re..eins active until',the ~de swicch is placed. in che RUl) po-icion. This switch occurs when reaccor pressure is greater tl:an 850 psig.

4 II% Fl,ux Scram Tri Settin The TR') 'System consists of 8 chambers, 4 in each o,:" ehe rr actor protec-t illn sy st em logic channels . The EB't is a 5-decade ins e remen c,h ich covers the~ range of power 1"vel between that covered by ehe SR)f and che Q'R.'!. Tne covered the o: range switch and the '5 decades 5 decades are by ZRN by means a are broken Cown into 10 ranges, each being on -hali of a decade in si"e. The IR't scram seeting of 120. divisions is active in each range oi the It%. For

~

~ f~

'*'f P

v IH a'

2.1 BASES 3~ IRM Flux Scram Tri~Settin~(Cont,inured) example, if the inst:rument vere on range 1E the scram sett,ing vould ~be~ at 120 divisions for t:hat range; 1:Lkevise, if the instrument: vas on range 5, the scram netting would bc 120 dLvisionn on that range, Thus, as thc XRN is rnngcd up to accommodate the increase in power level, the scram setting is also ranged up, A scram at 120 divisions on the XKC instruments remains in effect as 1'on~', a's t:he re'actor is in t: he startup mode. Xn addition, the APPA 15%%d'scram pover operation without being in the RUN mode, The. XRN scram provides pre'ver'>ts'igher protection for changes which occur bot:h loc'.ally 'and over the enti.re 'core, The most significant sources. of react:ivity change during the power incretesd aire due to control rod withdrawal, Por insequence control. rod withdraval, 'thh rate of change of power is slow enough dUe 'to'h'e Phy'sidal'imi'tat:ion of withdraving control rods, that heat flux's in equilibrium with the heutron and an XRN scram would resul.t in a reactor shutdown well. before any safety 'lux limit is exceeded. For the case of a singl'e control rod withdrawal error, a.

range of rod withdraval accidents was analyzed. This analysis:Lncluded't arting the accident at variious popover .'Levels. The most severe case involves'an it'd.tial condition in which the reactor is gust subcrit:ical and the XRH system is not yet on scale. 'This condition exists at quarter rod density. Quarter rod density is illustrated in paragraph 7.5.5 of the PSAR, Additional conservatism was taken in th:Ls analysis by assuadng that'he XRN channel closest t:o <the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power limited t:o ott6 phrden4 of rated poqeg, tttc(s Mi,ntaining lfCPR above 1,05Based on the above aIzalysLs> the 4@) provMeo protection against local control rod withdrawal. errors'n~'d continuous 'withdrawal of control rods in sequence. )I B. APRN Control Rod Block Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides 'a c'ontrol rod block ko prevent rod withdrawal beyond a given poi'nt'at'dnstan't r'ec:ir-cuclation flow rate, and thus to prot:ect against t'he'condition of a MCPR less than 1.05. This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevhnt's dn,fncrease in the reactor power level to excess value: due to control rod with-drawal. The flow variable trip setting provides substantial margin 22

2. 1 BASES from fuel damage, assuming a steady-state operation at the trip setting, oyeer 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 cou'd occur during s eady-state operation is at 108/ of rated thermal power because of the APRN rod block trip setting. The actual power distribution in the core is established by speci ied control rod sequences and is monitored continuously by the in-core LPP'I system. As, with the APRM scram rod block trip setting is adjusted downward trip setting, the APR~lexceeds if the CMFLPD +p thus preserving the APB.'I 'rod block safety margin.

C. Reactor Water Low 'Level Scram and Isolation (Fwce t ~guin SteamIines)

The set point for the low level scram is above the bottom o: the separator skirt.

This level has been used in transient analyses dealing with coolant inventory decrease. The results reported in FSAR subsection 14.5 show that scram and isolation of all process lines (except main stean) a" this level adequately protects the fuel and the pressuze barrier, because MCPR is greater than 1.05 in all cases, and system pressure does not reach the safety valve settings. The scram setting is approximately 31 inches below the normal operating range and is thus adequate to avoid spuzious scr~.

D. Turbine Stc Valve Closure Scram C

The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result frnn rapid closure of the turbine stop valves. With a scram trip setting of < 10 percent of valve clos<<rc from full open, the resultant increase in bundle power is limited such that HCPR remains

'1.0%even during the "orst case transient that assumes the turbine bypass is closed, This scram is bypassed when turbine steam flow is below 30 percent of rated, as measured by turbine first stage pressure. Actuation of the relief valves limits pressure to well below the safety valve 'setting.

E. Turbine Control Valve ScracL

1. Fast Closure Scram 1

The reactor protection system initiates a, scram within 30 Nsec after the control valves start to close. This setting and the fact that control valve closure time is approximatelv twice as long as that for the stop valves means that resulting transients, while similar, are less evere than for stop-valve closure. Xo fuel damage occurs, and reactor system pressure does not exceed the relief valve set point, which is approximateIy 280 psi below the safety limit.

23

2. ll .BASES
2. Scram.cia loIis of co<ntrol oil pressure The turbine hydraulic control system operates using, high preisure oil. Th'ere.,ar<e several points i'n this bil', s<j<'stem where a lbss c<iuld re'suit in a 8ss<t cIloiur<e 'bf the turbine kontrbl of'il'reIsi<ere valve'i. This fas't c*1'osure of the turbine control valves 'is not protected by the generator load 'rej<ec'ti6n 'scram, since falIQre,'f the oil,'ystem would not result in the fast closure solenoid valves be:Lng, actuated. For a tu'ibine'*nt'rol valVe fast 'cldsuIre) the core would be protected by the'APISH'and high reactor pressure.

scram'. ,However, to provide the s~Lme margins as provided for the<

~ 'ener'ator load ii)ection scram o'n )as't ctlosure of the turbine valves's scram has been a(lded to the reactor protection, ontiol.

system, which iIens'es failure of corLtrol oil pressure to the ~tu0-controlsjrstem. This is an anticipatory scram and resuIltls in 'inc reactor shut<down Ibefore any significant increase in pressure or rieutron fluxoc:curi. TheI transient rlesponse is very similar t~)

that resulting ftom the ginerator load re)ection.

e Hain Condenser Lov Vacuum. Scram l

To protict ehi main condenser against overprcssures a loss of vacuum initiates automatic el&sure of the 'turbine stop valves con-'enser aiid turbine. bypass valvei. To anticILpiIte'h'e tra'nsicnt and automitic scram, resulting from the closure of the turbine stop valves, lov cc~n-denaer va'cuum initiates a scrame Thk 1&'va'culm 'scram'et point is slelectee'd': to i>>I<<<<<<a scr~iii befc e the c]Iosure of the turbine stop valves is, i' G. S H. <<Sin Stean Line,'i<a al:ien ennLeS pretan're anl< <<ain Stean Line Saolat!Lan : Sera'n.

l l

The low pressure isolation of the main'Istejsm lines at 825 psig vas to. pro'tact igiinst rapid reactor ldepressurization and the 'royid'ed ra<pid cooldpvn o'f the vessel. Advantage is taken of the 'esulting scram, feature that <occurs when the. main'team l.ine isolation valves are,cloied; t'o. provide for reiictor shutdovn so that high power at. low re<accor pr'ea<sur doei not occv'r,'hus providing pr'otc'.ction oper'a-'ion."

for the fuel claddi'ng'ntegrity safety 1'imit. Operation of the keaIc-tor, it pressur<es: lover than,. 82> psig 'requi'res that the reactor in the. STi&TUP position, where prot;ection of the fuel <<:laddin mode'witch..be, integrity safety 1'imit is. provided'y the IRH and APRM high neutron'lu scram@. Thus's thencombicsation .of mai'n <t<<te'am line low pressure iIIolk,ti~on and'. isolation, valve cloiurse. scram assure}'s thl aVa9.1ability .cif cistron ove'r'he entire'ange of applicability of Ihe fidel 'lux,sc'ram';protection claddingl integrfty,safety limit. Xn additions the isolation vsl4e closure scram'.anticipitaa .the pressurle sInd~ fl'ux transients that 6cci~ir '

during:,.normal or'. inadv'ertent isolation valve closure. With the s'crim at 10, percent of valve- cl'os'ures neutron flux does not increas'e. '

'et.

'f

~ a" I

LIHITING CONDITIONS FOR OPERATION SURVEILLANCE RE UIREHENTS 3.'1 REACTOR PROTECTION SXSTEH 4e 1 REACTOR PROTECTION SXSTEH;

~ A licabilit A licabilit Applies to the instrumentation Applies to the surveillance of and associated devices which the instrumentation.and asso-initiate a reactor scram. ciated'evices which initiate reactor scram. e

~Ob ective ~eb ective To assure the operability of the To specify the type and frequency reactor protection system. of surveillance to be applied to the protection instrumentation.

S ecification S ecification When there is fuel in the vessel, A. Instrumentation systems shall the setpoints, minimum number of be functionally tested and trip systems, and minimum number calibrated as indicated in of instrument channels that must Tables 4.1.A and 4.1eB respec-be operable for each position of tively.

the reactor mode switch shall be B. Daily during reactor power operation as given in Table 3.1.A. at greater than or equal to 25'~ ther-mal power, the ratio of Fraction of Rated Power (FRP) to,Core Maximum Fraction of Limiting Power Density (CMFLPD) shall be checked and the scram and.APRM Rod Block settings given by equations in specifications 2.1.A.1 and 2.1.B shall be calculated, C. When it is '

determined .that a channel is failed in the unsafe condition, the other RPS channel.

that monitor the same variable shall be functionally tested immediately before the trip sys-tem containing the failure is tripped. The trip system con-taining the unsafe failure may b:

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 pex functional test period. for ithis testing.

h ~

PAINE DELETFD 32'.

4.1 The frequency of calibracLon of the 4'gH Flow Btasing .'network has been established <<s each refueling outage. There nrc several instruments which must be c.-.librated and ic vill'Aj}e cake several hours to perform the calibration of the entire network. rhe calibration is being per-formed, a zero Elow signal will be senr. to 'half of the APRH's resulting in a half scram and "rod block cond Lc Lon. Thus, if the calibration were performed during operation, flux shaping would noc be possible. Based on experience ac other generating stations, drif oE instruments, such as those in the Flow Biasing Network, Ls not sL".nificanc and therefore, co. avoid spurious scrams, a calibr cion frequency of each refueling out-age is established.

Croup (C) devices are active only during a riven portion of the opera-tional cycle. For example, ch" 1RH is active during stertup and inactive during full-power operation. Thus, chc only test chat is meaningful 's thc one performed )ust prior to shutdown or startup; i.c. ~ the tests that are performed gust prior to use of the instrument.

Calibration frcqu<<ncy of thc instrument cha icl Ls divided into two groups. These are as follows:

l. Passive type indicating devices that can be compared with like units on a continuous basis.
2. Vacuum tube or semiconductor devices and detectors that dri.ft or lose sensitivity.

Experience with passiv" type instruments in generating stations and sub-stations indicates that the specified calibrations are adequa c. For those devices which employ a"plifiers, etc., driEt specifications call for drift to be less chan 0.4Xlmonth; i.e., in the period of a month a d~ ift of .4X would occur acd thus providing Eor adequate margin. For the APR.'l syscem drift of clcctronic apparatus Ls noc 'he only considera-tion Ln determining a calibration frequency. Change .n power discribu-

,tion and loss of chamber sensitivity dicta e a calib ation every seven days. CalLbration on this frequency assures plant operation at or below thermal li..its.

A comparison of Tables 4.I.A and 4.l.B indicates that two instru ent channels have not been included in the latter cable. Those are: mode switch Ln shutdown and manual scram. All of che devices or sensors associated with these scram functions are simple on-o'.f switches and, hence, calibration during operation is not oppl'cable, i.e., the switch is either on or off.

The ratio of Core bfaximum Fraction of Limiting Power Density (JPLPD) to Fraction of Rated Power (FRP) shall be checked, out once per day to determine if the APRM scram requires ad)ustment. This will normally be done by checking the LPRH readings. Only a small number of .control rods are moved daily 47

4. 1 BASES

during steady-st at~ operation and thus the y.agip is )

not expected chaugs~ s'1 PYI 1 y 1<'lint 1 v.

The sensitivity of LPR'1 detectors decreases ~rith.exposure to neutron flux at a sloM and appro"...imately con.stant ra'te. Thid i0 compen"ated for in the APM system by calibrating every 7 days using 'heat balance data and by calibrating, individual LPRM's every 1000 effective. full-po~er hours using TIP traverse data.

TABLE 3.2.~

INSTRUHENTATIOM THAT INIT~ES ROD BLOCKS Minimum No.

Operable Per Tri S s 5 Function Tri Level Settin 2(1) APRH Upscale (Flov Bias) + 0 66g + 42X (2)'-

2(1) APRH Upscale (Star tup Mode) (8) c 12X 2 (1) APRH Downscale (9) > 3X 2(1) APRH Inoperative (10 )

1(7) MW Upscale (Plov Bias) < 0.66W + 41X (2) 1(7) RBH DoMnscale (9) p 3X 1(7) RSH Inoperative (10 )

3(1) IRH Ups~le (8) <108/125 of full scale 3(1) IRH Doynscale (3) (8) 5/125 of full scale 3(1) IRH Detector not in Startup Position (8) 3(1) IRH Inoperative (8) (io')

5 2(I) (6) SRH Upscale (8) < 1 x 10 counts/sec 2(1) (6) SRH DosTAscale (4) (8) > 3 counts/sec.

2(1) (6) SRH Detector not in Startup Position (4)(8) 2(l) (6) SRH Inoperative (8) (loa) 2(l) Flmr Bi"s Ccmp&latof < 10X diEEerence in recirculation flea 2(1} YloM Bias Upscale < 110X recirculation flov Rod Block Loqic N/A RSCS Restraint 147 psig turbine (ps-g p, s first stage pressure (approximately 30X pove

Yor thc stortup ind run positkons of the Non<<ter !'Inde Selector Svitch ~

there shall bc two:operatable or tripp<<d trip syatcrne for each function.

The SRH, IIN, 'and hPRH (Startup mod<<) ~ blocks n<<ad not 'be operable in (

"Run".~de ~ and the hPRH (Flow bitascd) and RiBH rod b:Loc'ks need not be operable in "Startup".mode. If the first ~colmen cannot be net for one of thc tvo trip systems, this condition nay exist for,us to seven days provide'd that during that thee the operabl<<cyst<<m is functionally~

tested immediately and 'daily thereafter; if this condition last longer:

than seven day>i, 'the system vith the inoperable channel sC'~ll be tripped.

If the first colu>~n cann'ot be <<at;for both tkipi spstens'> both trip eyste~ shall be trippi>d.

2. V -is the recirculation loop few in percent of design. 'Trip level settinR is in percent of rated power (3293

'"'l N~t). A ratio of FRPfC.'LFLPD <1..0 iis permitted at reduced power. See Specification 2~1 fox APRN cont:rol rod bloc'k setpoint.

3. ~ downscale is 1~assed vhen it is cn its lowest range A. This function is 'b'ypasi>ed.when the count cate is + 100 'cps and IRN above range 2.
5. One in'strueent channel;, i.e., one hPR'{ or IR.'L or RM, per trip system sLay be bypassed except only one of four SRH may be bypassed.
6. IRH channels AE, C, G all in range 8 brasses SPA channels A 6 C functions.

IRH channels 8 F, D, H all in range 8 bypasses SAN channels B 6 D functions.

7 The trip is bypassed vhen the reactor power is '< 30X.

8. Thi's 'function is bypassed vhan the rods svitch is placed in Run.
9. This function 'is only active vhen the node svit'ch is in Run. This functio'n i8 autoeati'cally bypassed vhen the IRK inst~'entation's, operable and not high.

10, The inoperative trips are produced by the folloving functions:

a. SRH and ZRM (1) Loc'al "operate~aiibrate" svitch not in operate.

(2) Pere'er supply vali:age lov.,

(3) Circuit boards not in circu't.

(1) Local "operate-calibrate'" switch not i~a ops@at~!.

'(2) Less than 14 LB'nputs.

'(3) Circuit boards rxit'in circuit, 74

[BITING 'CO:rDl.r Lucis vv?t 'v? ~i~i>> . ~

.3.8 Control Rodr? 4.3.B.Control Rods b~ During the shutdown procedure no rod movement is permitted The capability of the RSCS to pro-between the testing per ormed above 20: power and the rein-perly ful verified ill its function shall be by the fol.lowing tests:

statement of the RSCS re- Sequence pc tion Select n sequence straints at or above 20: and at e pt to withdr w a rod in the power. Alignment of rod remaining sequences. Move one rod groups shall bc accomplished in a seq ence and select tne remain-prior to perforrrring the tests. ing sequences and atte ot to move a rod in each. Repeat for a11 Whenever the reactor is sequences ~

c.

in the stnrtup or run mades Croup notch portion For each of the bcloM 20." rated power the six cor.par tor circuits go th"ough Rod worth Hininizer shal.l be test initiate; corparator inhibit; operable or n second licensad verify; reset. On seventh att pt operator shal.l verify that test is a'lowed to con inue until the ope" .or at the reactor corapletion is indicated by console is folloving the illumine icn of test complete light.

con rol rod program.

The capabi'ity of the Rod A second- licensed operator Mor th Mini+i er (R~N) shall may not be used in leiu of ?t>> v>>e < +4~4 t ~ >>t fed AVWAQS the RH'M during scram time checks:

testing in the startup or run modes below 20 percent 1.

(

0 of rated thermal power.

The correctness of the contro rod vithdrawa1 sequence '..put to the P.'~.f cc"putsr shall be verified oefore reactor s a up or shu'tdc If Specifications 3.3.3.3.a 2.

d he R!".". computer cr.

agnostic test shall ce line through .c cannot ba rret ths renctor shall not be started, successfully performed.

or if the reactor is in the 3. Pr'or to sta"tup, proper run or startup modes at Iaas annunc'at'on o" the sr?lec-thnn 20X rated power, shall bc brought to n shut-it tion error of at least one out-of-sequence contre? rod down condition i.ediatsly. sha'e ver'ed.

4. Prior to =

startup, the nct on of the Z~~

rod'lock shall ce ver'fied by =oving an out-of-sequence cor. "ol rod.

5. Pr'or to ootz'..'rg 20K rated power dur'".g rod inser:ion at shutdcrw, ver'"y he latch ".g o" the proper rcd

'group and prcper an"unc"a"'cn 123 after 'rsert e ors.

4 T trlG CONOIT IOUS FOR .OP ER i T ION SURVF ILLeQv'CF, RE UIRE)KNTS 3 3 B'ontrol Rods 4.3.B Control Rods

,4 Control rods shall,not be when required the preIenca withdra~ for stai;cup ior Ce of a second, Licensed ops'rator

~ refueling un'.Less it least 'to- verify the follcviu3 iof

- two soux'ce range channels 'theI carr'act rdd program~shalL

'have an observed c:ount hach race 4d vdrifidd e equaL to or I~reateIr .chan

three counts pdr second., P'rior,co,control rod Mithdri'xvsL for't'art'.up or during xcfud]Ling, 5 ..Duiing o pc ra c ion wr veiKfy that at least, two source limiting conlcrol rod pait- range channels have an observed.

tarns, as dcic'er61ned by the coupe xate of at least threie designated qualified'e.rson- eau>its per second.

. nel either:

~

a. Both RB'i channel.s shall ,5., Q'heic a liniting contro'l xod bc operable.: pattern exists, an instru..ent or functional test of the 'RB'A aha]LL 'be perfo'~ed prior to b Control rod, w/Lchdreival vkcbdraaaL of cne desigaaccc) shaL1 be bloc)ccd. rod(sl and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

G ~ Scram Xnsex'tion Times

  • 1.After each refueling outage all operabjie rods. shall be scram te:sti d f rain. the fully withdxawR~<

t~

po:sition wi,th the nuclear system pressure above 950 psig (wit'h satuxatiain temperature); This test:Lng sha,ll be comp,'l.eted pxi.or to C. Scran Insertion Times exceedi'.ng 40% power. Below'20%

powe'r', only. rods in tho6'e 'sequences 1 The average scram insertion and'34 or B12:and B, wh'ich (A12 tine, baaed'n iat'ion ofne thd deenergi-scrisn pilot veilve Iire1ie ftilly withdx'awn 5.n'h3 ) xc.'.gian

~

.f xi'im '00% rod density to 50% rod soL'esioid's. as ti.rec zeroof; all density shall be scxaia tiAe 'e'.st.ed.

opcrablc, contrpl rods >Ln the.

reactor poMer operation The sequence, restraints imposed upon shall be no.greater then:-

condi.-'i'on the contxol 'rods i.n the 100-50 lpercent: xod'ensity g&o6p4 t!o 'the X Znqcrtnd; Prom hvg., Scram"Inser- pr'eset power level maf be removed tio'ri Ti'aes (seel use of the individIxak 6y$ ass

'y

'C :sw".'itches associated with t!hose partially withdxawn ahd full/

.5 0.375 control rods which ax& dr 20'0;" 0'.902;Oi a&e not the 100-50 percent'raid 'densit 'ithin 90 5 0 groups. Xn order to bypass 'a rod, the actual rod axial )Iokit.ialn 'mu'st' known; and the rod mus t be iin the.

correct i.n-sequence positi'.oa.

124

3 3/4 3 SJJfi~E 30 Th"-Rod worth Nnimi.zer (PUN) and the Rod Sc'quencc ControI System (RSCS) rc::trier withdrawa'.s and fnscrtions of con<<o>

rods to pre-spccif f<<J scqucnccs. Ail patrrrns associated with these ccqu<<nccs have thc character I.".r fc thar, as'uming th<

worst single deviation from the scqu<<ace; the drop-of any control rod from thc fully inserted position to the position of the control rod drive would, not cause the reactor to sustain a power excursion resulting in any .pe1let average enthalpy in excess of 280 calories per g am. An enthalpy,of 280 calories

,per gram is well below t'e level at, which rapid fidel disp rsal "

could occur (i.e., 425 calories per gram). primary system damage in this accident 's not possible unless a significant amount of fuel is rapidly dispersed. Ref. Sections 3.6.6, 7.7;A, 7.16.5.3, and 14.6.2 of the CESAR and NFDO-10527 and supplements thereto.

Zn performing th fdncticn descr.bed above, the R~M ard RSCS ar>>

not required to irpose an> restrictions at core power levels in excess of 20 percent of rated. Y~terial in the cited rcferc."c shows that it is impossible to reach 280 calories per gram in th ~

  • ~ event of a control rod drop occurring xt poser greater than 2C

.percent, regardless of the rod patte~. This is true for all normal and abnormal patterns including those which maximize individual control rod worth.

ht power levels below 20 percent of rated, abnormal control rod patterns could prcduce rod worths high enough to be of concern relative to the 2o0 calorie per gram rcd drop limit.

TrL this range the AN and the RSCS constrain the control rod sequences and patterns to those whi=h involve only acceptable rod worths.

The Rod Worth Hinimizer and the Pod Sequence Control System provide automatic supervision to assure that out of sequence control rods vill not b w'thdrawn or inserted; i.e., i lim't=.

operator deviations from planned withdrawal secuences. Rcf.

Section 7.16.5.3 of the CESAR. They serve as a backup to procedure control of control rod sequences, wh ch limit the maximum reacti-vity worth of control rods. In the event that the Rod Worth Minimizer is out of service, when required, a second licensed operator can manually ful.ill the control rod pattern con-formance function'f this system. In this case, the RSCS is back up by independent procedural controls to assure conformance.

  • Because RSCS it is allowable by bypass certain rods in the during scram time testing below 20 percent of rated power in the startup or run modes, a second licensed operator is not an acceptable substitute fear the RWM during this testing.

V 6

1.'he functions ~if the RM and .RSCS risks. l t unnecessary, tc specify-f- a license limit on roc) .wor:h.to preclude unacccp~tablci consequeiaccs in the event oL' control rod drop. At low giovieis, below. 20 percent, the.;e. devices .force adherence to a'ccep1:able. rod,patterns. Above '20'ercent of rated power, rio icoiiistraint on rod" patte'rn is required to assure chit rod Clio'acc:Ldent consequences are acceptable. Ccmtroli red ~

pattern i:onst'raints above 20 percent of rated power are ~

imposed l>y powei,distr'ibu ion requirements, as defiined zn I Sectioris 3.5 I, 3.5.J, 4.5.I-, and 4.5.,J- of these techzsical specif ications. Power level for automatic bypass of thei DISCS,function is sensed lby first st'age turbine preslsuke.

4. 'I'he Source "Range Monit'or'O'RM) system .performs no autc)metic swfcty. system fuhctiovi; i.e., it has rio,scram function. It

3,3/4,3 BASES:

The s4urvcillance requirement for scram testing of all the cnntrol rods after each refueling outage and lOX of'the control rods at 16-Meek intervals is3 adequate for deter433ining the opera-bility of the control rod system yet is not, so frequent as to cause excessive Mear on the control rod system .components.

The numerical values assigned to the predicted scram perfor-Rance are based on the analysis of data 'from other BWR' with control rod drives the aaae as those on BroMns Ferry Nuclear plant.

The occurrence of scrar4 times Mithin the limits, but signifi-cantly longer than the average, should be vieMed as an indica-tion of systematic problem Mith control rod drives especially if the number of .drives exhibiting, such scram tines exceeds eight, the allowable number of inoperable rods.

In'he analytical treatment of the transients, 390 milliseconds are alloMed betveen a neutron sensor reaching the scram point and thc start of negative reactivity insertion. This is ade-quate and cons3ervative Mhen compared to the typically observed time delay of about 270 ailliseconds. Approx<ately 70 milli-seconds after neutron flux reaches the trip point ~ the pilot scram valve solenoid pover supply voltage goes to zero an approximately 200 milliseconds later, contro1. rod motion begins.

The 200 c3tlliseconds are included in the allovable scram inser-tion tines specified in Specification 3.3.C.

  • In order to perform scram time testing as required. I by specification 4.3.C.l, the relaxation of certain restraints in the rod sequence control system is required. Individual rod bypass switches may be used as described in specification 4.3.C.1.

The position of any rod bypassed must be known to

,be in accordance with rod withdrawal sequence.

Bypassing of rods in the manner described in specification 4.3.C.l will allow the subsequent withdrawal of any rod scrammed in the 100 percent to 50 percent rod density groups; however, it will maintain group notch control over all rods in the 50 percent density to preset power level range. In addition, RSCS will prevent movement of rods in the 50 percent density to preset power level range until the scrammed rod has been withdrawn.

133

3.3/4.4 BASES:

D. Reactivit~ Anomalies During each fuel cycle excess operative reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned. The magnitude of this .excess reactivity may be inferred from the critical rod configuration. As fuel burnup pro-gresses, anomalous behavior in the excess reactivity may be detected by compar Lson of the critical rod, pattern at selected base : tates to the pzedicted inventory at that state. Power operating base 'od conditior<s provide thi'ost sensitive and directly interpretable data relative to core react iwrity.

Furthermore, using power operating base condition's permits frequent reactivity comparisons.

Requiring, a react;ivity comparison at the specified

'frequency, assures 'that a. compa'ri'so'n )~ill'be made befor'e the core reactivity'ha'nge exceeds 1X A ~ .

Devia,tions in core reactivity greater t:hen 1/d$ are not expected and require tho'rough evaluation. One percent reactivity into the core wou2.d not lead to transients exceeding design cohd,itiods of the reactor system.

134

I, IYs IT I Nrs r~)NI) IT I<)H s FOR 0?I RATION SURVI'.II.LANCI'. KI," II I RFHENT 3.6. C Coolant Leaks)Pe 4.6.C Coolant Lcaka e

3. If the condition in, 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the .reactor shall be shut-down in the Cold Condition vithin 24 hours. 1. ht least one safety valve and approximately one-half of all Safe't and Relief Yalves relief valves shall be bench-checked or replaced vith a
l. When morc than one valve, bench-checked valve each opera-ting cycle. All 13 valves (2 safety or relief, is to be failed, an ordery shut-known safety and llorrelief) will have down shall bc. initiated and been checked replaced upon thc reactor depressurixcd .to the comolction of every second less than 105 psig vithin 24 cycle.

hours.

2. Once during each operating

.cycle, each relief valve shall

. be manually opened until thcrmo-

. couples downstream of the valve indicate steam is flowing from the valve.

3; The integrity of the relief/

safety valve bellovs shall be continuously monitored; .

4. At least one relief valve shall be disassembled and inspected each operating cycle.

E. ~Jet Pum s E. ~Jet Pum s

l. Whenever the 'reactor is in the I. Whenever there is. recirculation scartup or run modes, all'get flov vith the reac tor in the pumps shall be operable. If startup or run modes with both it is decermincd chat a get recirculation pumps running, pump is inoperable, or if tvo )et pump operability shall be or mor e $ e t pump f lov ins tru- checked'aily by veri'fying chat ment failures occur and can- the folloving conditions do not not be corrected vithin 12 occur simultaneously:

hours, an orderly shutdown shall bc initiated and the a. The tvo recirculation loops reactor shall'e shutdovn in have a flov imbalance of thc Cold Condition vichin 24 15X or more vhcn cne pumps hours ~ are operated at the same speed.

'81

I'IHIT I HCi COHf) l'I IOUS FOR OPERAT IO.'I SURVEILLAN(',K R~El V IRhtEHT

3. 6. E J~63 3'63. 3366 ~4.6.E Jet Pours a 3:6. b. The indlceted value of core
1. @hen both recirculation pumps flou rate varies frock the ar'e in. steady state Ioperation, value derived from loop the speed of'he faster pump flov measurements by more sha11 be maintained. within then 10'l..

122$ the speed o the slower pump when core power is RF/o or c. The diffuser to lower plenum more of rated power or 13'j$ the,. differential pressure reed-speed of the slower ynmrp ~rhen ing on en individual get core power is below pump varies from the mean power. of alL )et differen-80'f'ated, pump

2. H'pecif'ication 3.6.F.1 tial pressures by more t'han cannot be met, one recirculation lOX.

pump sha11 be tripped. 2. Nienever there is recirculation flow with the reactor in the The reactor sna11 not be Stertup or. Run Mode and one re-operated. with one recirculation circulation pump is operating loop out of service for more vith the equalizer valve closed, than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Mith the reactor, the diffueer to lover plenum operating, if one recirculation differential pressure shell b checked daily and the differen-loop is out of'ervice, the tial'press'ure of an indfvfdusl plant shall be placed in a hot i)et pump in e loop shall not.

shutdown condition within . 3va'ry lfrom the mean of all )et 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the loop is sooner returned to sex vice. pump 'differential pressures in-that loop .by more than 10X.

Fo11owing one pump operation, F. J t Pu Flov Mismatch the discharge valve of'he low speed pump may not be opened.

iirQ.ess the speed of the faster

l. Recircu1etion'alp'peeds shell be checked and logged et leaait pump is less than $ 0~ of its once per dey.

rated. "peed.

5~ Steady state operation with both recirculation pumps out of serv'ice for up to 12 hrs. is permitted. During . uc'h interva restart of the recirculation pumps is permitted. The plant shall be placed in a hot shut-down condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> unless one loop is returned to service. The total e,lapsed time in natural circulation and one pump operation must be no greater than 2i hours. Structural In ter ri t~

Stru'ctural Xnte rit C.

G.

1. The structural integrity l. Table 4.6.A 'together vich sup primary system shall of'he i

be,

~ pl emen tsry notes, s pecif ies 'he 182

LIMITING COiNDXTXONS FOR OPERATION SURVEILLANCE RE UXREMENTS

,3 11 PIRE PROTECTION SYSTEMS C.ll FIRE PROTECTION SYSTEMS h) licabilit A licabil'it Applies go the operating status oi- thc Applies to the surveillance require-high pressure water, ments of the high pressure water, and C02 fire protec-tion:systems for the reactor building, and C02 fire protection systems for diese'1 generator buildings, control, the reactor building, diesel generator bay, intake pumping station, cable 'buildings, control bay, intake pumping tunnel to the intake pumping station", station, cable tunnel to the intake and the fixed spray system for cable pumping station, and the fixed spray trays along the south wall of the system for cable trays along the south turbine building, elevation 586. wa11 of the turbine building, eleva-tion 586 when the corresponding limit-ing conditions for operation are in Ob ective: effect.

To assure availabilit of'ire ~Ob ective:

Protection S stems.

e ""

To verify the operability of the Fire Protection Systems.

Hi h Pressure Fire Protection S stem A. Hi h Pressure Fire

,The High Pressure Protection S stem

,Fire Protection System shall have: High Pressure Fire Protection,System an Two (2) high Tes ting:

pressure fire pumps operable Item Precrpuenc r and aligned to the high a. Simulated Once/year pressure..fire automatic header. and manual actuation of

b. Automatic high pressure initiation logic 'umps and auto-operable. matic valve operability',

b- ~Pum Once/month Operability

c. Deleted
d. Pump Once/3 yen~'apability 315

0 LIMITING CONDITIONS FOR OPERATION S(SURVEILLANCE REQUIREMENTS;,

3..11 FIRE PROTECTION SY. TEM 4. 1 1 FIRE PROTFCTION SYSTEMS ~

checked to be 2664 gpm at 250 feet he'ad

e. Spray Once/year header ance nozzle inspection for blockage
f. System Twice/year flush in con3unctzon with s'emi-annual addition .of biocide to, the Raw Cooling Water System
g. B uilding On ce/3 hydraulic year,s per formance veri ficati.on h., Yard loop Once/y ear and cool-ing tower loop hydraulic performance vexifica,tion 316

'L'fle l.*s ys vu< v ivy <

flow and pressure to an individual load. listed on Table 3.11.A while maintaining a design raw service water load or 1132 gpm.

4. 1 1 BASES testing of both the High pressure Fire System and the CO

'eriodic Fire protection System, will provide positive indication of their operability. Xf only one of the pumps supplying the High Pressure Fire System is operable, the pump that is operable will be checked immediately and: daily tho'reafter to demonstrate operability. Xf the CO~. Fire Protection System be'come -.

inoperable in the cable spreading*room, one 125-pound (or larger) fire extinguishere will be placed at each entra'nce to the cable spreading room.

Annual testing of automatic valves and control devices is in accordance with NFPA code Vol. XX, 1975, section 15, paragraph 6015. More frequent testing would require excessive automatic system inoperability, since there are a large number of automa-tic valves installed and. various portions of the system must be isolated during an extended period of time during this test.

wet fire header flushing, spray header inspection for blockage, and nozzle inspection for blockage will preven", detect, and remove buildup of sludge or other material to ensure continued.

operabi;lity. System flushes in conjunction with the semiannual addition of biocide to the Raw Cooling orater System will help t prevent the growth of crustaceans which could reduce nozzle discharge.

Semiannual tests of heat and smoke detectors are in accordance with the NFPA code.

>lith the exception of continuous strip h at detectors panels, all non-class A supervised detector circuits which provide alarm only are hardwired through conduits and/or cable trays from the detector to the main control room alarm panels with no active components between. Non-class A circuits also actuate the HPCX water-fog system, the CO< system in the di sel generator buildings, and isolate ventilation in shutdown board rooms The test frequency and method's specified are justified for the following reasons:

An analysis was made of worst-case fire detection circuits at Browns Ferry to. determine the probability of no undetected failure of the circuits occurring between system test times as specified in th surveillance requ'.rements. A circuit is defined as the wire connections and c)mponents 'that affect transmission of an alarm signal between "he =ire detectors and t he control room annunciator. Three c'cuitsere anal yzed wt<lch welf'epres".nt'1tii'~ of an. ':I. ~rm-only'ci rcuit,

~

a water-fog circuit, and a CO, circuit. The spreading room B smoke detector was selected as the worst-case alarm-only circuit because it had the largest number of wi:res and connections in a single circuit. The HPCX water-fog circuit was selected for analysis because it is the onl'y water-fog circuit in the area of applicability for technical specifications. The Standby Diesel Generator Room A CO<

327

1 gt c j<rcu'j."t ~o'ra's s~l:ectecl .bhcaiisi.. '.it contained 2 6ut'f 3 det:ector.

'logic, tt>e .rno" t compel;i.cated 4Oz circUit 3.ogic:Calculations Ver'e 'b*ased on 'fail'ure 'rcxtes -foi 'wi'res; 'c'inrxections,

'c:oiipo'nen'ts as shown i'n App'endix 3;-II; rof ';ogASH-140(i.

aiid'i:icu'it f

~Failui'e rat es ',w',o"re consxi:dered fo'r the ollciwing circuit component:s':

1 0 '.Copen ci.rc"ui:t t 2-: S'hoit to grouixd 3~ :Sho'rt 'tio 'po~~er

'4- '.'X'imin'g mo'toi" 'fai:lur'e 'to,start

'5-. Relay fa'ilu>.e "tc~ energ'izle

'N'o'r'mally 'open 'coritact'fa'ilure t;o clo'se 7~ 'Normally jp6'n or n'o'rmally clc)sed contact she)rt

'8.. Normall'I'. "I:lI)sed; contact IopI~n'ii.ncj

'.9.'i'mi;ng;:sw;i;tch f'ailure:to 'trar'rsf er The calc'ul'a'ted.,p'r'obabii.l.ities:(!P'f) do& no undetected fai3,"ure

'of the ciicuits 'occuXrirx'g, were 'a's follows based on the" sp'e,ci.f jed, 'tes't f'reque"r'r'c'j~.

Spreading ARF.A Room B TE'.ST FREQUENCY One Month "0

0. 9'75287 HPCI Water Fog Si.x Months 0. 977175 Sta'n'dby, Diesel '(~en Etoom A CO2 Si.x Months 0 9 57 5 9r)

The ',worst case.;nf the th'ree ar'i as 'co'nside*red x.s Spreading Poom O'. The ]pr'obabi.lity of;und'ected failure .is approxirn'ateely 1/40;- which mr ans that. 'one undetec't:ed failure will occur on

'the aver'a'g~ 'eI.".y 4'Ci months over a6 6xt'en'ded period of time and tha't,t: he fax.lxxre could exist up to one month. The

. fre'quency,

'of testi.'ncI~ is thus mixch cgeaater tlhan the frequency of 'fa'ilare, and y'r'odxxces ciicuits 'w'i'.th adequate- reliability,.

2; 'ci'rcui'ts checlcs by initi:atior'x 6f "end 'of the, line or end of the 'br'an'clIx detectciis ~iill iI'nore thoroughly test the par'allel cu'ic'ui.'ts 'than tes'tirxg on a -rotate.'hg de'tector ba. is. This

test i.'s nIot a d'citector test; bi'it 'i5 a tes't to'simulate the effe'ct. of electrical 'supervision. as defined iri t:h'e NFPA

.;co,de'. +

3; Testi rig oE,circuits w'hich actuate CO2,- water;: or vent!ilation sys,t'ems r'cqrii-'x.:.e~ di'sabli:ng the automatic feature of thi P.ice sy.t:enr for'the arr a; A suiveillance piogram ~'hich 'p'ro'tection

di'sabled th'ese ci'rcuit.s nnoiittily'ould signific
antly ability of these circuits to pi'ovid@ fire suppression.

reduke-'he

+Ref:. NiPA Code 7'0p-".9'; paragraph 1'11'1,'od~ I20-15, paragraph 1312

'fo'r definition'f C'lass A s'ystem~, and Cbde 72A-18, Article 240.

ENCLOSURE 3

.PROPOSED CHANGES TO BROWNS FERRX NUCLEAR PLANT UNIT 3. TECHNICAL SPECIFICATIONS Reasons for Proposed. Changes Proposed Changes

0 k

ATTACHMENT 1 REASONS FOR PROPOSED CHANGES TO BFNP UNIT 3 TECHNICAL SPECIFICATIONS H'

e As stated in Attachment 2 to Enclosure 1, any proposed changes which have been submitted. to NRC by previous requests are marked in the margin by a single bsr. Any new changes requested are signified by a double bar in the margin. Most of the new changes are being proposed ini order to provide continuity in the terminology used throughout the plant as far as thermal parameters are concerned,. These changes are as a result of the unit 1 reload amendment.

t" Page 6 - Addition of definition of CMFLPD and deletion of Total Peaking Factor.

Page 10 - Substitution of CMFLPD and FRP for MTPF.

Page 12 - Substitution of CMFLPD and FRP for MTPF.

Page 16 - Substitution of CMFLPD and. FRP for MTPF.

Page 19 Changes allowing steady-state operation without forced recirculation for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Page 20 - Substitution of CMFLPD and FRP for MTPF.

Page 22 - Substitution of CMFLPD and FRP for MTPF.

Page 31 - Substitution of CMFLPD and FRP for MTPF.

Page 46 - Same as 31 above.

Page 77 - Same ss, 31 above.

Page 124 - Changes per TVA BFNP TS 75, also clarification as to when this condition applies.

Page 128 - Changes per TVA BFNP TS 75.

Page 133 - Same as 124 above.

I Page 135 Rearrangement of several paragraphs to accomodate changes on page 13 6.

Page 136 - Changes per TVA BFNP TS 75.

Page 196 - Changes a1lowing steady-state operation without forced

~

~ recirculation for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Pages 347 - Changes per TVA BFNP TS 93.

357

~

~

gl t

JI 4~'4 ( 'I II

ATTACHMENT 2 Il 3~ Core Maximum Fraction of Limiting Power Density (CMFLPD) The highest ratio, for all fuel types in the core, of the maximum fuel rod power density (kW/ft) for a given fuel type to the limiting fuel rod power density (kW/ft) for that fuel type.

avera e planar Linea.- Peat Generation Rate ~LYHGR

- The Average Planar '.Jeat Generation Rate is applicable to a specific planar height and is equal to the sum of the all the fuel rods linear heat generation rates for in the specified'undle at the specified height divided by the number of fuel rods in the fuel bundle.

Ve nstrumentation le Instrument Calibration, - An instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range, and accuracy, to a known value(s) of the parameter which the instrument monitors.

2~ channel A channel is an arrangement of a sensor and associated components used to evaluate plant variables and produce discrete outputs used in logic. A channel terminates and loses its where individual channel outputs are combined in identity logic.

3~ Instrument Functional Test - An instrument functional test means the injection of a simulated signal into the instrument primary sensor to verify the proper instrument channel response, alarm',

and/or initiating act'ion.

Instrument check - An instrument check is qualitative determination of acceptable operability by observation of instrument behavior during operation. This determination shall include, where possible, comparison of the instrument with other, independent instruments measuring the same variable.

5~ ic S stem Functional Test - A'ogic system functional test means a test of all relays and contacts of a logic circuit to insure all components are operable per design intent. Mhere practicable, action will go to completion; i.e.g pumps will be started and valves operated.

6~

instrument channel trip signals and auxiliary equipment required to initiate action to accomplish 1

I a, prot'ec;tiiie .tr.ip functipng, $ grip system or more .instrument channeel trip signals may'ieqid:re,one rIelaited. to .oiie or moie plarit paracnsetieris in order to i

ti ji'-:system "actiorr, -In'itiation of fpIeo'tcectiiir'em'.ac<'ion 'ma'require the. txipping. of a

'ii'nltiate i'irncjl'e"'triji aiy'itcexa or. thie-. clinic'ident t'ripping of',

'C140 'tz'i,p 83(st'.ems 7~ 'Protective Ac:tion - An, act(on, initiated by the

protection 'system when a limit is reached. A protec'tive ac
tion can be at a channel or system

.'1 iiv,iml;

8 'Pjotec~tve~inc~iin -, A system protective action

'v'ihip>'eau]Ltc'ii'fro>n the protcecMye, action of the

,.c'ha'iiiieIls.mian'itoiiiiig

\

a particular plant condition.

caculcation'"'.means a]pplying"'a" simulated signa3L to the senior 'to,actuate the circui't in question.

c 10'. 'L~Lc 'A "lo'cjic iis can arirangementi of relays, "cceIta'c:ts',.:cmcl .cd er,compiaiients that produces a cdec'ision 'oiitjiu't.

(a)":1ni~t'jt'3LncjI - A .1'jgic that Ireceives signals "f:rom":ichanaIels and':produces diecision outputs to

':t:he,'acccti'ia't;ion ..'Logicc.

(b) 'dict'uat'i'ci'n -' logic that r'eceives signals

'"(eithIerr."ifrom",iiiitia'tion logic or channels) and i

gircaduceis
elec'isicon outIputs to accoimplish a

.'proi'tee'ULvei ac'tioni

Puriotion'a3 .Teiits - ac funotional te'st is the manua3.

Operat i'OniC>r 'initi'atX'On cCaf a,eyatem, Suheyatem, Or c'oinponent to veiify 'tha't. it 'functions within design

.toleianc'es '(':e.'g.' t:he ma'nual start',of' core spray pump to verify that .i't vo1ume,.o'f i~at,er)'-.

nines awd 'thiat it pu'mps, the required

'Xe ~g~utdcjiwn -, '.Tbie,reactsor i's 9;n' shutdown Condition when

'.th'*'-',rreactoi:;-iiiode switch is 'in the "shutdown mode position

.an'd 'naI coze; ail'terations are bIeing perforrnecIl.

-Ye  !~Ri" ii'eiered:Raifefuctrcl -. An ehciineered safec3uard is a

".safety syst'en'the. action's of'which 'aire essential to a

sarfety 'a'ctia'os ireq3mired in response t
o accidents.

,Z ~ 'annul'a~ve," Dciwiitiin'e - 'ThIe cumulat4,ve dpwntime for those safety" compone'nts"'and -syestemci~ whose downtiae is limited

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 141 FUEL CLADDING INTEGRITY 2~ 1 FUEL CL'ADDING INTEGRITY W = Loop recircu-lation flow rate in of rated per-'ent

{rated loop recirculation flow rate equals 34;2x104 lb/hr)

In the event of operation with the core maximum fraction of limiting power density (CMFLPD) greater than fraction of rated thermal pover (FRP) the setting shall he modified as follovs:

FRP SS(0 66M + 54%)

0MFZFS For no combination of loop recirculation flow rate and corp thermal power shall the APRM flux scram trip setting be allowed to of rated thermal exceed'20%

power' (NOTE: These settings assume operation within the basic thermal hYdraulic design criteria. These criteria are LHGR 5 13.4kW/ft and MCPR 2 1 '7 10

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING F 1 FUEL CLADDING INTEGRITY 2~ 1 FUEL, CLADDING INTEGRITY If it xs determined that either, of these design criteria is being violated <Turing operat Lon action shall be initiated within 15 minutes to restore operation within the prescribed limits,. See sjecification 3.,5.J and 3.5.K.

surveillance requirements for maximum total peaking factor are given in Specification 4.,1.B) .

2i APRM"-Mhen the reactor mode switch is in the STAlRTtJP position the APRM scram shall, be set, at less than or ecpxal to 15% of rated power.

3~ IRM The IRM scram shall be set at less than or casual to 120/125 of seal.e.

full

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTINGS'

'F LC N T GR TY' ' FUEL CLADDING'NTEGRITY Bi Core Thermal Pover mit B APRM Rod Block Tri Settin Reactor Pressure 5800 s'a The APRM Rod, setting, shall block be:

trip When the reactor pressure is less than or equal to. 8 800 psia, or core coolant (0 ',66W +02%)

flee is less. than 10% of vhere:

rated, the core thermal poster shall not exceed 823 MWt (about 25% of rated S~ = Rod block setting theraal'oer)- in .percent of rated ther'mal poorer (3293'Wt)

W ~ Loop recirculation floe rate in percent of rated (rated loop recirculation fl,ov rate equals 30 2 x 104 lh/hr)

In the event of operation w3.th the core maximum fraction of limiting power density

{CMFLPD) greater than fraction of rated thermal power {FRP) the setting shall be modified as follows:

C. Foyer Transient S ~ (0.66W +q2% }

that the Safety Limit CMFLPD To ensure establ'ished in Specification 1.1.A and 1.1.8 is not exceeded, each required scram: shall be initiated by its expected scram signal. The Safety Limit shall be assumed to. be exceeded'hen scram is accaaplished by means other than the expected scram signal.

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1 FUEL GLADDING INTEGRXTY'UELCLADDXNG XNTEGRITY D. t Shu down'. Condit ion C Scram and isola- h 538 in.

tion reactor above Whenever the reactor is in low wat.er vessel the shutdo~m condit ion level zero

'with irradi.ated fue3. in the reactor vessel, the D Scram--turbine 6 )0 per-water leve]L 'hall 'not be

~

stop valve e',en't valve less than 'l7.7 in. above clc sure c:losure the. top of t?ie normal active fuel ..one. E,. Scram -turbine control valve Fa st closure-.-Upon

'trip of the fast acting solenoid valves

2. ass of con- h. 550 p tro3. oil p:;.es sure Scram--low con-,,h g3'nches denser vacuum Hg vacuum G~ Scram- -main 5 I'0 per-st,earn line cent valve isolation closure Main steam iso3.a <,825,psig tion valve closure

.nuclear system low pressure Core spray and 378 in.

L1?CI actuation-- above.

'xeactor low water vessel level zero HPCI and RCX,C 090 in.

actuation--reac- above tor low water vessel level zero K. Main:steam isola-tion valve '.~ bove closure--reactor vessel low water level zero

unc~:rtainties employ~'d in derivinq the safety limit are provided thr. he@i nn ing ot each fuel cycle.

il'lio M(:I'I< value u>>cil i>> tho I'((:,'I pcrloniruiic <<vilu>>t:io>> (I.:IH) i>> lc>>>>.

limit iiq! thm the NI'lt lor operation (J.Z7).

C'ecduse the boiling transition corri l.~tion is based on a large (Iucln ti ty of 3 o f ull scale data there i a very high confidence that ope'ration of a fuel assembly at the condition of MCPR = 1.05 wou'ld not produce boiling transition. Thus, although it is. not requireed too eestablish the safety limit additional margin exists between the safety limit and the actual occurrence of oss of cladding integrity.

How'ever, ~ if boiling transition would not be expected.

were to occur, Cladding temperatures clad perforation would increase to approximately 11000F which is below the perforation temperature of the cladding material ~ This has been. verified by tests in the General Electric 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 perforation.

If 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.

ln addition to the boiling transition limit (MCPR=1.05) operation is constrained to a maximum LHGR of 13.4 kV/ft. This limit is reached when the Core Maximum Fraction of Limiting Pover Density equals 1.0 (CMFLPD=1.0). For the case where CMFLPD exceeds the Fraction of Rated Thermal Pover, operation is permitted only at less than 100$ of rated pover and on1y vith reduced APRM scram settings as required by specification 2.1.A.l.

At pressures below 800 psia, the core elevation pressure drop (0 power, 0 flow) is greater than 4.56 psi. .At low powers and flows this pressure differential is maintained in the bypass regi e ion 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 28x103 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 28x103 lbs/hr. Full scale ATLAS tes data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.3S 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 conservative.

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,. Xf water level 16

should drop below the top of the fuel during this time, the ability to remove decay heat is reduced.'his xeduction in cooling. capability could lead to elevated cladding temperatures~

and. clad perfoiatioii. As long as the ~fuel~remains, coveri d with

'wate'i, sufficient ccioling is available to prevent 'fuel clad perforation.

Thet safety limit ha!~ hieen established ,'at, '17.7 in. 'bove the top of the irradiated fuel, to provide a point which can, be monitored and also provide ad'equate'argin. This point corresponds approximately to the top of the actual, fuel. assemblies and the lowei r'eactor low water level trip (378< above vessel al'so'o iero) .

REFERENCE

1. General Electiic ZiWR Thei..mal Analy'sis SaSis (GETAB) Data.,

Corielation and"Design Application, 10958 ~

NEDO 10958 'nd NEDE.

17

2. 1 BASES: LIMITING SAFETY SYSTEM SETT N RELATED TO FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the Browns Ferry Nuclear Plant have been analyzed throughout the spectrum of planned operatinq condia.'ons up to the design thermal" power condition of 3440 l%t. The a.x>lyses were based upon plant operation in accordance with the operating map given in Figure 3.7-1 of the FSAR. In addition, 3293 MWt is the licensed maximum power level of Browns Ferry Nuclear Plant, and this represents the maximum steady-state power which shall not knowingly be exceeded.

Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and ax'ial power. shapes. These .factors, are selected conservatively with respect to their effect on the applicalbe transient results as determined by the current analysis model.

This transient model, evolved over many years, has been substantiated in operation as a conservative tool for evaluating reactor dynamic performance. Results obtained from,a General Electric boiling water reactor have been compared with predictions made by the model. The comparisions and results are summarized in Reference 1.

The absolute value of the void reactivity coefficient used in the analysis is conservatively estimated to he about 25% greater than the'nominal maximum value expected to occur during the core lifetime. The scram worth used has been derated to be equivalent to approximately 80% of the total scram worth of the control rods. The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by Technical .,

Specifications. The effect of scram worth, scram delay time and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative, reactivity insertion. The rapid insertion of negative reactivity. is assured by 'the time requirements for 5% and 20% insertion. By the time the rods are 60% inserted, approximately four dollars of:negative reactivity has been inserted which strongly turns the transient, and accomplishes the desired effect. .The times for 50% and 90%

ins'ertion are given to assure proper completion of the expected performance in the earlier .portion of the transient, and to establish the ultimate fully shutdown .steady-state condition.

For'nalyses of the thermal consequences of the transients a MCPR of 1.27 is conservatively assumed to exist prior to initiation of the "transients. This choice of using conservative values of controllinq parameters and initiating,transients at the design power level, produces more pessimistic answers: than would result by using expected'alues of control parameters and analyzing .at higher power levels.

Steady.-state operation without, forced r'ec'irc'.ulation will not be

. permitted for more;than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,.

In summary:

1 ~ The licensed maxi.mum power leve~ is 3293 MMt.

20 Analyses of transi:ents employ adequately conservative the controllinq reactor paramete!rs!.

values'f i

3 ~ The abnormal operational transients were analyzed, to a power level of 3440 )4Wt.

The. analytical procedures now used !redulIt !in a more logical answer than the, aLlternative method, of assuming a higher starting power in conjunction with the expected values for the parameters.,

The bases for individual ~set points are discussed below:

A~ Neution Flux Sera!m 1 ~ APRM High Flux Scram Tr:Lp Setting (Run Mode)

The,. average power. range monitoring (APRM) system, which is calijbrated using heat balance data taken during steady-~state cIonditions, z;eads in per!ceil of rated'ower.

(3,293 lSft) .:Because f;Lssion chcLmh!ers.jirovide the basic input s:lgrials, the APRM system responds direct:Ly to average neutron flux. During transients~ the instantaneous rate of heat transfer from the fuel.

(reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel.

Therefore during transients induced by disturban!ces, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting.

Analyse~s reported in Section N14 of the Final Safety Analysis Report demonstrated that with a 120 percent scram trip setting, nione of the abnormal operational transients analyzed, violate thi fuel ~safety limit and

.there is ai substantial margin fiom fuel damage.

Therefore, use of a flow-biased 8cr'am px'.ov'ides eVen additional margin. Figure 2.1,.2 shows the flow. biased scram as a functi.on of coze flow.

An increase iri the ApRM scram setting'would decrease the margin present, before the fuel cladding integiity safety limit is zeached. The APRM scram setting was determined by an analysis of margins required to provide a, reasonable ringe for rnaneuverir>g during operation.

Reducinq this <<operating margin would increase the frequency of spurious scrams, which have an adverse effect on, reactor safety because 'of'he stresses. Thus, the AP& setting was selected

're'sulti'ng'hermal 19

becaus>>

cladding itintegrity provides adequate margin safety'limit yet for the fuel allows operating margin that reduces the possibility of unnecessary scrams.

The scram trip setting musi be adjusted to ensure that the LHGR transient peak i~ riot increased for any combination of CMFLPD andFRP. The scram setting i,s adjusted in accordance with the formula in Specification 2. 1.A.1, when the'CMFLPD exceeds FRP.

Analyses of the limiting transients show that no scram adjustment is required to assure MCPR > 1.05 when the tiansient is initiated from MCPR >1 APRM Flux Scram Tri Settin Refuel or Start 6 Hot For operation i;n the startup mode while the reactor is at. low pressure, the APRM scram setting of 15 percent of rated power provides adequate thermal margin between the setpoi;nt and the safety limit, 25 percent of rated. The margin is adequate to accomodate anticipated maneuvers associ;ated 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 pat'terns are constrained to be uniform by operating piocedures backed up by the rod worth minimizer and the Rod 'Sequence Control System. Worth of individual rods is very low in a uniform rod pattern. Thus, all of possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant

,power rise. 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 power 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 level, the rate of power rise 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 pressure is greater than 850 psig.

IRM-Flux Scram Tri Settin The IRM System consists of 8 chambers, 0 in each of the.'a reactor protection system logic channels. The IRM is 20

5-decade instrument wIhic:h covers the range of Knower level. bIet~ieeri It:hat coIyei.ed by the SRM and the 'APRM., The 5 decades are', c'overed b]p the IRM by miearLs of a range switch ancl the 5 decades are broken down into 10

.be:i.'ncaa one-half of decade in 'size. The IRN scram ranges,'ach setting of 120 divi<<siIoni ca i active in each range of'.the For examjple;- if the :.nstrument were on range 1, 'RN.

the scram set'tinct would b'e at 120 div'isions for .that range; ii) ewise, if the iristrument was on r'aIage 5, the scram setting'oisldl be 120 divisions on that range.

Thus, as the IRM is ranged.*up. to accommodate t1>e

.incr'ease,,ii.n power <<level'he scram setting is also ranged .i~p. A,scram at 120 divis'ioris ion the IR)C.

instruments reInIaiLns iIneffect as long air the r'ea'car's in the startup m<)de. The APRM 15 percent scram will p'ievent hiLghIer p'ower operation without being in the. run, mode a The IRN s'c:ram provides Ipiotect:ion for charges iihi.ch occur both local]Ly and over thee entire core.

The Inost ciignificanit sourres of reactivLtj chahge during the powio'i increase, a'e due to control r'~xi withdrawal.

For, i:nsieqiience 'c()ntrol rocl wit,hdrawal; the. rate cif change, Iof power:is slow, enough due to the physicail limitatiori aif 'withdlrawing contro]L rods, that he'at; flux is in eIjuiili.briuin iiith the neutron flux and an IRM scram would result in a react@i shutdown w'ell'efoie linsit 'is exc:ceded. For. tlute case of a .sincjle any'a'fety c<<'ontiol rc)d withdraiwal terror this trans. Lent has been

'analyze'd .ii.n paiagraiph '7;5.5.4 iof the,FSAR. In order to ensure thclt the I(&t provides. adequate prot, ection against the si'rigle ro'd w:itl',Ldrawal erior, a range ci'f re with'diawa]L aiccidI~nt:s was analy'zed. This analysis included'tartinIj t:he accident at var'ious power levels.

which.

mo'st severe t'e case invo]Lves an initial condition in ie.actor i.s just subcritical and the IRM system

'he is'ot ye1 ann scale. TIhis condition exist.s at quarter rod den'sity; Quarter rood densit]j i.s illustratied in paragiaIph 7.5,.5 of the FSAR. Additional conservatism

.,was takien irIi this analysis by. assuming that the IRM

.channel c]LoiIest Ito the .withdrawn red is bypassed'. The results of,t.his analysis show that the >i:eactor is.

scr'ammed andi', p'eak K>ower 1'irnite6 Q one percent of rated power,. thi'is'maintairiing MCPR above 1 05'ased on the above an'a]Lysis, the IRN provi.des- protection against local'onticil'od jiithdrawal errors and continuous w'ithdrawa]L of; control, rods in 'seCjuenc'e.

Control Rc)d- Block 'PRM R'eactor power level inay be. varied,,by moVing i ontrol rods or by, var'ying th'e recirculation flow rate. ~

The~ APRM system provides a control rod-,block to prevent ~rod withdrawa'I b'eyond 2l

L5nenr Hest Generation Rate LHCR This specification assures that the linear heat generation rate in any rod is less than thc design linear bent generation if fuel pellet densification is postulated. Thc power spike penalty specified is based on thc. anal-ysis presnntad in Section 3.2.1 of Reference 1's modified in References 2 and 3, and assumes a linenrly increasing variation in axial. gaps bc-tvecn core bottom and top, and assures with a 95% confidence, that no morc thon one fuel rod exceeds the design linear heat gencrnt5on rate duc to poc;cr spiking. Thc L))GR as a function of core height shall bc checked daily dur'-

ing reactor operation at > 25% po~er to dctcrminc if fuel burnup, or'on-trol'od movement has cau cd changes in power distribution. For Lt)GR to bc i limiting value below 25% rate:d thermal power, the MTPF would greater than 10 which is prccluclcd by a considerable margin when employing have to be snv permissible control rod pattern.

Minimum Critical Power Ratio MCPR At core thermal power levels lese than or equal to 25Ã, the reactor will be operating at minimum recirculation pump speed and the moderator void content vill ba very small. Por all designated control rod patterns which may be em-ployed at this point, operating plant experience and thermal hydraulic anal yoio indicated that the resulting MCPR value ia in excess of requirements by i considerable margin. Pith this low void content, any inadvertent core flow increase cwould only place operaticn in a more conservative mode rela-tive to MCPR. The daily requirement for calculating MCPR above 25% rated thermal power is sufficient 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 limf.ting control rod pattern is approached ensures that MCPR vill be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.

Re ortin Re uirements The LCO~s associated with monitoring the fuel rod operating conditions are required to be met at all times, i.e., there is no allowable time in which.

the plant can tcnowingly exceed the limiting values for MAPLHGR, ZBGR, and MCPR. It io a requirement, as stated in Specifications 3.5.1 .J s~Q that if at any time .during steady state power operation, it is determined

.that the limiting values for ?QZLHGR, LHGR, or MCPR are exceeded action is then initiated to restore operation ro within the prescribed limits. This action is initiated as soon as normal surveillance indicates that an operating 3.im-it has been reached. Fach event involving steady state operation beyond a specified limit shall be logged and reported quarterly. It must be recognised that there is always an action vhich would return any of the parameters (';)APLHGR, LKGR, or MCPR) to within prescribed limits, namely power reduction. Under most circumstances, this will not be the only alternative.

M. References

l. "Puel Densification Effects on General Electric Boiling Va'c" Reactor.

Puel," Supplements 6, 7, and 8, NEW-10735, August 1973.

4

2. Supplement 1 to Technical Report on Densifications of General Electric Reactor Fuels, December 14, 1974 (USA Regulatory Staff) .
3. Communication: V. A. Moore to I. S. Mitchell, "Modif'ed GE Model for Puel Densification," Docket 50-321, March 27, 1974.
4. General Electric BWR Reload 1 Licensing Amendment for BFNP unit 1,

.NED0-24020, May 1977.

5. General Electric BWR Increased. Relief Valve Simmer Margig Evaluation for Browns Perry Nuclear Plant Unit 1, Sept:ember 27, 197/

169

Core and Contaiumcnt Cooling Systems SurveIlla.xcc Fre<tuencies The testing interva!L for the cor<. and containment "a~ling systems is,based, on industry practice, quantitative reliability analysis, )<)dgement and practicality. Thie core cooling systems have not'een designed to be fully teotablc during operation. Fair example, in the case oE the HPCI:, automatic initiation during po~er operation would result in pumping <'ld water ~into ~

the reactor vcsseil which is not desirable. Complete ADS testing during power operation causes an undesirable loss-of-coolant inventory. To increase thc availability of the core and- containment cooling system, the component@

which make up the s'stem;,'.e., instrumentation, pumps, valves, etc.,i are i tested frequently. The pumps and motor operated in)ection valves are a]Lso, tested each month to assure their operability. A simulated automatic actua-tion test once each cycle= combined with monthly tests of the p<xmps and in)ecz tion valves is deemed to be adequate testing of these systems.

Nicn components and subsystems are out-of-service, overall core and contain-,

5 mrnt cooling rcliahility is maintained by <llcmonstrating the operabiliIty'I of thc remaining cquipmcnt. Thc degree of operability to bc demonstrated depends on the nature of the reason for the. out-of-service equipment. For routine out~f-service periods caused by pxcventative maintenance, etc. the,pu<<<p and valve operability checks will be performed to demonstrate operability of thc remaining components. However, .if a Eailurc, design deficiencycause the outage, than t'e demonstration oE opcx'ability should be thorough enough, to assure that a generic problem does not exisit. For example, if an out-of-service period was caused by failure of a pump to deliver rated capacity due to a design deficiency, the other pumps oF. this type might be sub)ected to a flow rate test in addition to the opexability checks.

Mhencvcr a CSCS system or loop is made inoperable because of a required:

test or calibration, the othex; CSCS systems or loops that are required to be

'operable shall be considered operable if they are within the required surveil-

'lance testing frequency and there is no reason to suspect 'they are inoperable.

'If the function, system, or loop undcx'est or calibration is found inoperable or exceeds the trip level setting, the LCO and the required surveillance testing for the system or loop shall apply.

'Redundant operable components are subjected to increa ed testing during eguiP-ment out of-service times. This adds fu'rther'onservatism'nd inc'rcases assurance that adequate cooling is available should the need arise.

The HAPLHGR, LHGR, and MCPR shall, be checked dail~y to determine if fuel burnup, or control rod move<ident has caused changes in power distribution. SinIce changes to burnup are, sjLow, and oaly a Eew control rods are moved daily, a: daily 'due check of power distribution is adiequate.

170

'able 3.5.l-l MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type- initial Core - T e 1 6 3 Average Planar Exposure MAPLHGR PCT

.(Mwd/t ~k't4/f t) ~F) 200 15.0 1926 1,000 15.1 1902 5,000 16.0 1975 10,000 16.3 2047 15,000 16.1 2151 20,'000 15.4 2136 25,000 14.2 2035 30,000 13.1 1922 Table 3.5.X-2 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: initial Core T e 2 Average Planar Exposure MAPLHGR PCT (Mwd/t't) (IGt//8 I ) ('F) 200 15.6 1973 1,000 15.5 1956 5,000 16.2 1973 10,000 16.5 2063 15,000 16.5 2143 20,000 15.8 2119 25,000 14.5 2005

.30,000 13.3 1886 171

Table 3.5.X-3 ifAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: 80278L Average Planar PCT Exposure MAPLHGR Hwd/t ~F

11. 2 1652 200 11.3 1645 1,000 11.9 1648 5,000 12.1 1626 10,000 12.2 1642 15,000
12. 1 1642 20,000 11.6 1603 25',000 10..9 1537 30, 000 Table 3.5.X<<'4 l~fAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: 892788 Average Planar PCT Exposure l'fAPLHGR
Nwd/t). (F) 11.1 1646 200 11.2 1635

. 1,000 11.8 1640 5,000 12.1 1630 10,000 12.2 1647 15,000 12.0 1648 20,000 11.5 1608 25,000 10.9 15I 7 30,000 172

0 BRONNS FERRY NUCLEAR PLANT F lGURK 3.5.2 K) FACTOR AUTOMATlC FLOW CONTROL ANUAL F OW CONTROL Scoop- Tube Set-.Point Calibr'ation pokition such that flowmax 102;5 4 101.0 112.0%

117.0 lo

@O M . FO CORE FLtRf,'L

0 CP 0

X 8

1.44 F

U E

L Q 1.32 I

I I

O 7 7

I F

U I E

L

1. 34 BOC-2 3440 CYCLE AVERAGE EXPOSURE (MWD/TONNE)

MCPR VS CYCLE AVERAGE EXPOSURE -FIGURE 3.5.3

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE~UXRDKNT I

3.'6 PRIMARY SYSTEM BOUNJQARY, 4.6 PRIMARY SYSTEM BO'UNDARY A~21icability k Applies to the'perating, status Appl:Les tc> the periodic exapd.rjat$ 'or) of'the reactor coolant sy'tem. and testir~g reguirerrrents fop tkhe, reactor coolant system.

~ob ective Ot~>e cti.ve 5

To assure the integri'ty and safe To determine the condition of the operation of the reactor coolant reactor, coolant system and the system. operation ol.'he safety devices related to it.

S 'ecification S ecification r

\

A. Thermal and Pressurization A. Thermal and Pressurizat;ion Limitations Limitations l

1. The average rate of reactor 3.. Zkuring heatups and, cooldowns,,

coolant temperature change the, following parameter."..'hall during 'norrkkal heatup .or be recorded, and, z'eactor coolee cooldown slzal3. not exceed, s,nt, tempera't'ai..e determined at; 3.00 2'/hr when averaged oyer 3.5-minute interva3.s 'until 3 a one-hour period. suc'cessi've readings at each giv'en location are wikthin 5o

a. Steam Dome Pressure (Convert to upper vessel region temperature)
b. Reactor bott;om drain temperature
c. Recirculation loops A and 3
d. Reactor vessel. bottom head temperate:rre
e. Reactor vessel she33. adjacent; to shell flange I 2e During all opezations wit;h a. 2~ Rea,ctor vessel met;al temoerature critical core, other than ,at, the outside surface of the for low level physics testsp bottom head. in the vici.nity of tria

. except when the vessel is ,cojrrtrol rod drive housi.ng rLnd vented, the reactor vessel, reactor vessel shell adjacent shell and fluid temperatures to shell flange shall 'kZe recorded sha11 be at or above the at least every 15 minutes duri+

temperature of cue g3 of inservice hydrostatic or le+

figure 3.6-1. testing when the vesse3, pres is ~ 312 psige

Js, I

h

>.r.. t t t t;r. rn:tDr Trna.; FOR O, t RATtO<t SVRVRrt.aAVCr; RVnura<>i Nj"

3. 6. C Coolant l.eakave 4.6.C Coolant t.enka c
3. If the cond ition in 1 or 2 r above canno t be met, an orderly shutdown shall be initiaced $rr, and thc reaccor shall be ehuc- D. Safety and Relief Valves down in thc Cold Condic.ion within 24 hours. l. Ac least one s-fcty valve 3r.d approximately one-half of all D. relief .valy'es'shall bc bench-checkcd o= replaced with a 3.. Mhen morc than onc valve, bench-checked valve each ope-.'a-safety or relief, ie known to ting cycle. All 13 valves (2 be failed, an ordery shut- safety and ll relief) v'll have down shall be initiated and been checked or replaced upm:

the reactor depressuriacd to the connie([on.of every'econd less chart 105 psig within 24 cycle.

hours.

2. Once during each operating cycle, each relief valve sha':

be manually oper.ed until thcrmo-couples do~stream of the valve indicate stcam sis flowing from the valve s.eggs, ~

h

3. The integrity of the relief.i 8& f cty valve bellows shall be continuousLy 'onitored.
4. At least one relief valve sh:,11 be disas cmblcd and in"pec:ed each operacir.g cycle.

J~es Pua s Kr. ~Jee puu s

1. Mhenevcr the reactor is in the 1. Mhenever there is recirculation srartup or run modes, all get flow vich the reactor in the pumps shell be operable. If startup or run nodes vith both it is determined that a )et: recirculation punps runn'ng, pump is inoperable, or if two ]et pump operability shall be or more )et pump flow instru- checked da'y by verifying chat toent failures occur and can- the fol lowing cond c ion s i c o no t not be corrcctcd within 12 ~

occur s inul taneoualy:

hours, an orda r ly shutdown h shall bc initiated and the a. The tvo recirculation loops reactor shall be shutdovn in have a flow imbalance of thc Cold Condition vithin 24 15K or nore when the pumps hours ~ are operated at the same speed.

181

r

l. lN I 7 1 Nt"r (:AHf)lT I i)NS FOR OPKRAT loft,, . SURVEKt.LA.'tdtl I~tt'.< U1RL~tENT

~see 'v . s 4.6.E Jet P~uia~e 3.6.F Jet Flow Mismatch b. The indicated value of,cox'e 1; When both recirculation pumps flov rate varie.s 'from t'e are in, steady state operation, value der'ived'rom loop the speed of the,f'aster pump flov aeasurei)ents by more shall be maintained.t~ithin ,th'an 1QX.

12@ the speed.o the sltover pump when core porter .is 8t@ or c. The diffuser to loiter plentmt more of rated. pover or 13'.if~ the ,diLfferential,pressure, read-speed. of'he slower. pump when 'ng on an individual 5el:

core powex is below 80~ 'pu555p varies from the mean power. of'ated.

of all 5et pump, differen-2.

5 Ix, specif'ication 3.6.,F.1 tial pri.siuraa by. more than

,cannot be met, one recirculation ilOX.

pump sha11 be trippecL.

2~ Whenever there is recirculation flow.with the reactor'n the!

The reactor sha33. not be I Starltut) o'r Run'Hede 'and one re-operatecL with one recircu'tation circulation pump is operating loop out of'er55>ce f'r more

" with'he equaliier valve clcised, ttte dif fuser t'o lo5rer plenu555 thari 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With the. reactor ,

differential pressux'e shall b operating, if- one recirculation ,

daily a'nd the differen-loop is out of'ervice, the ~

checked ti,nl pressur'e of an individual p3.ant shaU. be placed izl a hc)t ', 5et pump in,a loop shall not shutdown condition rrithin vary from the mean of all 5et 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> un1ess the loop is pump, differential pressures in sooner returned to service. ,

that loop by moie th,an lOX.

4., Fo11owing one pump operation, >et Pum FloM Hismatch the discharge valve cif t;he low

'speed. p)Lmp may not be opened, Recirculetion,pump speeds shall unless the speed, of the faster be checIceg at)d* logged at !Least pump is 3.ess than 5P~ of its once per day

'ra,ted speed.

5. Steady state operat:ion with both recirculation pumps out of service for up to 12 hrs. is 1 ~

.permitted., During such in'terva restart of the recirculation, pumps. is permitted. 'Thc.'lant shall be placed in -a hot: shut-

.down condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> unless one loop- is. returned to service. The total elapsed time, in~ natural: circulation: and one pump operation must: be no greater than 2ri hours. Stru'cti.)ral Entg~ri~t G. Structural Inte ritg of.

G.

I., The structural integrity 1be Tiible'4.'6.A together vt.th sup-the primary system shall 1'.',

plementary notes, specifi'es the 182

A. Hefuelinr, Interlocks Complete functional testing of all refueling interlocks before any refueling outage will provide positive indication that the interlocks operate in the situations for which they were designed. By )nadinp each hoist with a weight equal to the fuel assembly, positioninI; the reI'ueJing platform, and withdrawing control rods, the interlocks can bc

,sub,)ected to valid operational tests. Where redundancy is provided in the logic circuitry, tests can be performed to assure that each rerlun-dant logic element can independently perform its function.

B; Co> e Wonitorin Requiring the SRM's to be functionally tested prior to any core altera-tion assures, that the SR'!,s will be operable at the start o. that altera-tion. The daily response check of the SRM's ensures their continued operability.

BI FI"hl 3C .'S

l. Fuel Pool Cooling and Cleanup Svstem (BFNP FSAR Subsection 10.5)
2. Spent Fuel Storage (BFNP FSAR Subsection 10.3) 314

LIMITING.CONDITIONS FOR OPERATION SURVEILIJDCE RE UIPZHENTS 3.'ll FXRE PROTECTION SYSTEMS ll FXRE, P 0 'C 'XON SYSTEMS

~A~licab ility:

Applies to the operating stat:us of the Applies to the surveillance require-high pressure water, ments of the. high pressure water, and C02 fire pxotec-tion systems for the reactor building, and C02 fire pxotection systeras for diesel generator buildings, contxol the reactor building, diesel generator bay, intake pumping station, cable buildings, control bay, intake pumping tunnel t'o the intake pumping st:ation, station, cable tunnel to the intake and the fixed spray system for cable pumping stat:ion, and the fixed spray trays along the south wall of the system for cable trays along the: south turbine building, elevation 586. wall of the tuxbine building, eleva-tion 586 when the corresponding 1imit>>

ing conditions for operation are in Ob ective: effect.

To assure availabilit of'Fixe ~Ob 'ective:

Protection S stems.

To verify the operability of the Fire Protection System.

Hi h Pressure Fixe S e 'if cation:

A' High Pressure Fire The High Pressure Protection System Fire Protection System shall have: High Pressure Fire Proteotion System a~ 5>o (2!) high Tes ting:

px'essure fire pumps opexable Xtem Frequency and aligned to the high a. Simulat:ed Once/year pressure fire automat:ic header. and manua.l actuation of

b. Automatic high pressure initiation logic ptImps and auto-operable. matic valve operability Pu~m l)nce/month Operability
c. Deleted d+ Pulllp Once/.'3 yea i capability 315

3 11 BA~ES The High Pressure Fire and COq Fixe Protection specifications are provided in order to meet the preestablished levels of operability during a fire in either or all of the three units.

Requiring a patrolling fire watch with portable fire equipment the automatic initiation is lost will provide (as does the if automatic system) for eax3.y xeporting and immediate fire fighting capability in the event of a fire occurxence.

The Hi gh pressure Fire P otection System is supplied by three pumps aligned to the high pressure fire header. The reactors may remain in ooeration for a period not to exceed 7 days pumps are out of service. If if two at least two pumps are not made ope able in seven days ox if all pumps are lost during his seven

,day period, the reactors will be placed in the cold shutdown condition wit?'in 2rr hours.

For the ar. as of applicability, the fire protection water distribution system minimum capacity of 266rr gpn ai 250'ead at the fire pump discnarge consists of the following design loads:

1 0 Sprinkler System (0.30 gpm/f" z/Qorr0 tz axea) 1332 gpm 2 ~ 1 1/2" Hand Hose Lines 200 @pm

3. Raw Service Water Load 1132 corn TOTAL 266rr gpm The CG Fixe Protection System is consid red operable with a minimum of 8 1/2 tons (0. 5 tank) COz in storage or units 1 and 2; and a minimum of 3 tons (0.5 tank) COz in storage for unit 3 An immediate and continuous fir watch in the cable spreading room or any diesel generator building area cO, fire will be established protection is lost in this roon and will continue until i

CO, firo protection is restored.

To assure close supervision of fire protection sys"errr activities, the removal f rom service of any component in eit..er the Hicn Pressure Fire System or the CO> Fire Protection Svstem "or any reason otn r than testing or emergency operations will require Plant Superintendent approval.

Early reporting and immediate fire fighting capability in the event of a fire occurrence will b orovid d (as ~'ith the autonatic system) oy requiring a patrolling f i "e watch than .one detector for a given protected zone"=is inooerable if more roving .ire watch fo areas in which automatic =ire suppression systems are to be installed will provice additional i.nterjm fj.re protection for areas tha" have be n determined to need additional protection.

326

The fire protection system is designed',to supply the required flow and pressure to <<n indivi dual load l jsted on Table 3. 11. A while maintaining a design raw service water load of 1132 gpm.

4.'11 - BASES periodic testing of, both the High pressure Fire System and the CO Fi're protection Sy tern will provide popit',ive j.ndieation of their operability. If only one of t:he pumps 'supplying the High Pressure Fire System is operable, the Jurnp Chest is operable will be checked immedi.ately and daily thereafter to demonstrate operability. If the CO~. Fire ProtectiOn 'Sy't'm becomes inoperable" in the cable spreading room, one 125-pound'or larger) fire,extinguishere will be placed at each entrance to the cable spreading room.

Annual testing of automatic valves and control devices is in accordance 'withe MFPA code Vol. IX, 1975, section 15, paragraph 6015., More frequent testing would reqqire excessive automatic, system inoperability, since there are a large number of automa-tic valves installed and various portions of'he system must be isolated during an extended period of time during tLis test.

Wet f'ire header flushing, spray header 'nsp'ec'tion f or blockage,,

and n'ozzie inspection. for b)Lockage will preven,, detect, and remove buildup of sludge or ot:her material to ensure continued ~

operability. System flushes in conjunction with the semiannual, addit:ion of b'oc!Lde to the Raw C'ooling Water System will help preve'nt the growth of crustacean. which could reduce nozzle discharge.

Semiannual tests of: heat and . moke detectors are in accordance with the NF&A code.

with the exception of continuous strip heat detectors panels, a3.1, non-class A supervi.sed detector circuits which provide alarm only are hardwired through conduits and/or cable trays from the detector to the maiLn control room alarm panels >Iith no active components between., Non-class A circuits also act:uate the HPCZ water-fog system, the CO> system in the di sel generator buildings, and isolat;e ventilation in shutdown board rooms. The test'frequency and methods speci. fied are justified for the following reasons:

An analysis was made of worst-case fire detect:ion circuits at Browns Ferry tc) determine the probability of no undetected failure of the circuits occurring bet.ween syst.em test timesi' as soecified in t.h surve.'Lllance requ'.remen<.:s. A circuit.

defined as the wire connections and c>mponent - that affect o!3 an alarm .-ignal between t'e fire det>ctors ,'ransmission an 3 t h ~ control room annunri ntor. Three cxrcuits were analyzed which were repre;'. nt ativ-..'f, a'n il:arm-only circuit a water-fog circuit, and a CO~ cirruit. The:oreading room B smok detector was selected as the worst-ease alarm-only

.circuit because it had the 3,.-argest number of wires and connections in a single c Lrcuit. The HPCZ wat:er-fog circuit:

'was selecteD ror analysis because:Lt is the only water-fog circuit in the area of applicability for technical specifications,, The Standby Diese.'L Generator Room A CO<

327 IIc

5. 0 HA.IOR IIFS IO:I FI:.ATURFS
5. I S ITI'. Fl ATII!<l8 GroMns Ferry unit 1 is located at Brogans Forry Nuclear Plant site on property oLEncd by the United States and in custody of thc TVA., The site shall consist of approximately 840 acres on the north shore of <wheeler Lake at Tennessee River file 294 in Limestone County, Alabama. The minimum distance from thc outside of the secondary containment building to the boundary of the exclusion area as defined in 10 CFR 100.3 shall bc 4,000 feet.

5.2 REACTOR A. The core shall consist of l68'uel assemblies of 64 fuel rods each and 596 fuel assemblies of 49 fuel rods each, B. The reactor core shall contain 185 cruciform-shaped control rods. The control material shall be boron carbide povder (84C) compacted to approximately 70 percent of theoretical density.

5. 3 REACIOR VESSEL The reactor vessel shall be as described in Table 4.2-2 of the FSAR. The applicable design codes shall be as describeo in Table 4.2-1 of the FSAR. I
5. 4 CONTA lNNESIT A. The principal design parameters for thc primary containment shall bc as given in Table 5.2-1 of the FSAR. The applicable design codes shall be as described in Section 5.2 of the FSAR.

B. The secondary containment shall be as described in Section

5. 3 of the FSAR, C. Pcnetrations to the primary containment and piping passing through such penetrations shall be designed in accordance vith the standards set forth in Section 5.2.3.4 of the FSAR.
5. 5 FUFL STORA(:E A. Thc arrangement of iucl in the no~-fuel storage facility shall be such that k ., for dry condit fons, is less than 0.90 and flooded is iefss than 0.95 (Section 10.2 of FSAR).

S ff'30

g.O MAJOR OFS1Cl> FRATORI;S'ContinUcId)

B. The k ~f f of the spen't fuel storaye ponI shall be less than or equal to O.9(I for normal conditions and 0.95 for abnormal conditi9ns (Sections LO.i3 of the, FSAR),.

5.6 SV.IS.IIC OF.SIGiil The station cia s I structures and systems have been desi';ned to wt thst;phd a desiRn basis earthquake Mi:h. pround accelera-t ion of .iD. 2R. The operational balsik ebrthqdak0 used in the plant deaipn assumed a ground acceleration of O.lp (see Section 2.5 vf the FSAR).

331

ENCLOSURE 2 PROPOSED CHANGES TO BROWNS FERRY NUCLEAR PLANT UNIT 2 TECHNICAL SPECIFICATIONS - Reasons f'r Proposed Changes - Proposed Changes

ATTACHMENT' REASONS FOR PROPOSED CHANGES TO BFNP'NIT' TECHNICAL SPECIFICATIONS As stated in Attachment 2 to Enclosure 1, any proposed changes which have been submitted to NRC by previous requests are marked in the margin by a single bar. Any new changes requested are signified by a double bar in the margin. Most of the new changes are being .proposed in order to provide continuity in the terminology used. throughout the plant as far as thermal parametersare concerned. These changes are as a result unit reload amendment.

of'he 1

Page 5 Addition of definition of CMFLPD and .deletion of 'Total Peaking Factor.

Page .9 Substitution of CMFLPD and. FRP for MTPF.

Page 10 Substitution of CMFLPD and FRP for MTPF.

Page 16 Substitution of CMFLPD and FRP for, MTPF.

Page 19 Changes per TVA BFNP TS 85 and. allowing steady-state operation without forced recirculation for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Page 21 - Substitution of CMFLPD and'RP for MTPF.

Page 23 - Same as 21 above.

Page 31 Same as 21 a'bove.

Page 47 Same as 21 above.

Page 48 Same as 21 above.

Page 74 Same as 21 above.

Page 123 Changes per TVA BFNP TS 75, also clarification as, to when this condition applies.

Page 124 Changes per TVA BFNP TS 75.

Page 129- Same as 123 above; Page 133 Same as 124 above.

Page 182 Changes allowing steady-state operation without forced recirculation for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Pages 315,- Changes per TVA BFNP TS 93.

327

ATTACHMENT 2

a given point at constant recirculation flow rate, and1.05. thus to protect against the condition of a MCPR less than This rod block trip setting, which is automatically 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 izip setting provides substantial margin from fuel da sage, 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 during 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. As with the APRM scram trip setting< the APRM rod block trip setting is adjusted downward if the CMFLPD exceeds FRP preserving the APRM rod block safety margin.

thus Reactor Water Low Leve Scram a d Iso ation Exce t Main Steam ines The set point for the low level sc"am is above the bottom of the separator skirt. This level has been used in transient analyses dealing with .coolant inventory decrease. The results reported in FSAR subsection N14.5 show that 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 settings. The scram setting is. approximately 31 inches below the normal operating range and is thus adequate to avoid spurious scrams.

Turbi e Sto Valve C osure Scram The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of .the turbine stop valves. With a scram trip setting of 510 percent of valve closure from full open, the resultant increase in bundle power is limited such that MCPR remains above 1.05 even during the worst case transient that assumes the turbine bypass is closed. This scram is bypassed when turbine steam flow is below 30 percent of rated, as measured by turbine first stage pressure.

Actuation of the relief valves limits pressure to well below the safety valve setting.

Turbine Control Valve Scram

1. Fast Closure Scram The reactor protection system initiates a scram within 30 Msec after the control valves start to close, This 22

setting and the fact that contrpl,valve~ r;leisure time is approximately twice as long as that for the stop valves means that resulting transientswhile similar, axe less, severe than for stop-valve closure,. No fuel damage occurs, and reactor system pressure does not exceed the relief valve set point, wh ch is approximately 280 psi, below the safety limit.

2. Scram on loss of control oil pressure

.The turbine hydraulic contxoli sysgerr> operates using high pressure oil. There are several points in this oil system where a loss of oi.l pressure,could xesult in a, fast closure of the turbine control valves. This fast closure of the turbine contxol va.Lves is not protected by the generator load rejection scram, since failure of the oil system would not result in the fast closure solenoid va.'Lves being actuated., For a turbine contro'.L valve f ast closure, the cox e would be protected by the APRH and high reactor pressure scrams. Ho~ever, to provide the same margins as provided for the generator load re.jection scram on fast closure of the turbine control valves, a scram has been added to the .xeactor protection system,, which senses failure of control oil pressure to the turbine control system. This is ar>

anticipatory scram and results in reactor shutdown befoxe any significant increase in pressure or neutron flux occurs. The transient Xeqporrse j.s very similar,to, that resulting from the generator load rejection.

F. Main Condenser Low Vacuum Scram To protect the main concenser agai.nst overpressure, a loss of condenser vacuum initiates automatic closure of the turbine stop valves and turbine bypass valves. To anticipate the transient and automatic scram resulting from the closure of the turbine stop valves, low condensex'acuum i.nitiates a scram. The low vaccum scram set point is se3.ected to initiate a scram before the closure of the turbine stop valves is initiated.

G. 6 H. Hain Steam Tine Iso'.Lation on L( w Pressuxe and Main Steam Line Isolation Scram The low pressure i.solation of the, maire steam, lines at 825 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantag>> is taken of the <cram feature,that occurS when the main steam line isolation valves are closed, to provide for 'reactor shutdown so that high power operation at low reactor pxessure does not occur, thurs p oviding protection for the fuel cladding integrity safety limit.

Operation of the reactor at pressures lower than 825 psig requires that the reactor mode .".witch be in the STARTUP Amendment No 4

Revised 5-19-77 2.2 BASES REACTOR COOLANT SYSTEM INTEGRITY The pressure relief system for each unit at the Browns Ferry Nuclear Plant has been sized to meet two design bases. First, the total safety/relief valve capacity has been established to meet the over-pressure protection criteria of the ASME Code. Second, the distribution of this required capacity between safety valves and relief valves has been set .to meet design basis 4.4.4-1 of sub-section 4.4 which states that the nuclear system relief valves shall prevent opening of the safety valves during normal plant isolations and load rejections.

Thirteen safety/relief valves have been installed on each unit with a total capacity of 78.7X of rated steam flow. The analysis of the worst overpressure transient, (closure of all main steam line isolation valves using ll Target Rock Safety Relief Valves and 2 Dresser Safety Valves the most limiting configuration) neglecting the direct scram (isolation valve position scram), results in a peak nuclear system pressure at the bottom of the vessel of 1304 psig if a high neutron flux scram is assumed. The resulting 71, psig margin to the ASME Code limit of 1375 psig assures adequate protection against overpressurization.

(Referenced from Sections 3.1.3 and 3.2.3 of NEDO-21165 of January 1976, entitled "GETAB Analysis Including the Effects of Neutron-Effective Voids and Substitution of Crosby Valves - BFNP Unit 3," modified by TVA letter to NRC dated April 21, 1977, J. E. Gilleland to A. Schwencer).

To meet the second design basis, the total safety/reliefcapacity of 78.7% has been divided into 64.5X relief (ll valves) and 14.2X safety (2 valves). The analysis of the most severe abnormal operational transient resulting in a nuclear system pressure increase (turbine trip from high power without bypass) assuming that the scram is initiated by the position switches on the turbine stop valves is presented in Section 3.2.1 of NEDO-21165 of January 1966, entitled, "GETAB Analysis Including the Effects of Neutron-Effective Voids and Substitution of Crosby Valves-BFNP Unit 3," modified by TVA letter to NRC dated April 21, 1977, J. E.

Gilleland to A. Schwencer.) This analysis shows that the relief valves limit pressure at the safety valves to 1178 psig which is 72 psig below the spring safety valve setpoint. This analysis also shows that the peak nuclear system pressure is 1219 psig at the bottom of the vessel which is 156 psig below the allowed ASME Code limit of 1375 psig.

30 Amendment No. 5

LXMITING CONDITIONS FOR OP]ERATXON SURVRLL1MCE RFNUI REAGENTS 3 1 REACTOR PROTE]CTXObi SYS~E&l slane- " 'unauna A licabilit ~j~~tg Applies to the surveillance Applies to the instrumentation instrumentation and of'he and associated devices whiich associated devices which initiate a react:or scram. fLnitiatis reactor scram-()+b']ective To spec:i.fy the type and Ob ective frequency of surveillance ito b0 applied to the protectiOn To assure the operability of inetrumentwtion the reactor<prot.ection system.,

S~icificatXon A Inistmaentation systems shall be,Eunctionally tested and calibrat:ed ae

.indicated in Tables 4 1.A The setpoints minimum number and 4.1.B respective1y.

of trip systems, and minimum number of instrument channels Dewily during reactor pougr 6pdration that must be operable for each at] +edter t'han or equal to 25$

position of the reactor mode switch shall be as given in Table 3.1.A.

'f t2Ierma3, po~er, the ratio of Fraction R,ate,d Power (FRP) to Core Maximum Fraction of Limiting Power Density (CMFLPD) shall be checked. and the scrmm snd APRM Rod Block setting:

given by equations in specifications 2.1.A.l and 2.1.3 shall be calculated.

Ce When it, iis determined that a i9xannel ie failed in the unsafe condition, the other RPS channels thiat

,monitor the same variable shall be:EunctionalLly tested iauaediately before the txip system containing the failure ie tripped.

The trip csystem containing the resafe failure may ]be untripped for short periods of time to alloie functional testing of tihe other trip system. The trip system may,'be in. the .

untripped position for no more than eight hours per fmmthanal t:oot period, for this testing 31

The frequency of calibration of the APRM Flow Biasing Network has been established as each refueling outage. There are several instruments which must be calibrated and it vill take several hours to perform the calibration of the entire network. While the calihration is being performed, a zero flow signal will be sent to half of the APRM's resultinq in a half scram and rod block condition. Thus, if the ca19 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 Network, is not significant and therefore, to avoid spurious scrams, a calibration frequency of each refueling 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-gower operation. Thus, the test that is meaningful is the one performed just prior to only shutdown or startup; i.e. ~ the tests that are performed just prior to use of the instrument.

Calibration 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 ~

providing 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 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.

The ratio of Core Maximum Fraction of Limiting Power Density (CMFLPD) to Fraction of Rated Power (FRP) shall be checked out once per day to determine scram requires ad)ustment.

if the APRM This will normally be done by checking the LPRM readings.

Only a small number of control rods are moved daily during steady-state operation and thus the ratio is not expected to change significantly.

The sensitivity .of LPRM detectors decreases'ith exposure to neutron flux at a slow and approximately. constant x'ate This is compe'nsated for in the APRM system by calibrating every 7 days using heat balance data and by calibrating ~individual LPBM~s e'very 1000 effective full-powei hours c'asing TIP 'traverse data 47

TABLB 3e2 ~ C IHSTRVNENTATZON THAT INZTIATES ROD BLOCKS Nininue No.

Operable Per Punction Tri Level Set n 2 (1) APRN Upscale (Plow Bias) ~0 ~ 668442% (2) 2 (1) APRN Upscale (Startup Node) (8) ~ 12%

2 (1) APRN Downscale (9) 3%

2 (1) APRN Inoperative (10b) 1(7) RBN Upscale (Plow Bias) < 0.66Si40% (2) 1(7) RBN Downscale (9) + 3%

1 (7) RBN Inoperative (10c) 3 (1) IRN Upscale (8) ~ 108/125 of full scale 3 (1) ZRN Downscale (3) (8) + 5/125 of full scale 3 (1) ZRN Detector not in Startup Position (8) 3 (1) IRN Inoperative (8) (10a) 2(1) (6) SRN Upscale (8) ~ 1 x ltf counts/sec.

2(1) (6) SRN Downscale (4) (8) ~3 counts/sec.

2(1) (6) SRN Detector not in Startup Position (4) (8) (11) 2(1) (6) SRN Inoperative (8) (10a) 2 (1) Plow Bias Comparator ~10% difference ia recirculation flows 2 (1) Plow Bias Upscale ~110% recirculation flow Rod Block Logic R/h 2(1) RSCS Restraint 147 psig turbine (PS-85-61h and first-stage pressure PS-85-61B)

NOTES For< TABLE 3.2.C F'r the. startup and'ru'n Ritions'f the Reactor Nodle selector Switch, there shall be two operable or tripped trip

~~

systems for each function. The SlRH. IRM, andt APRN (Startup mode) ~ blocks. need not be operable i.n "Run~'ode, and the APRM (Flow biased) and, RBN rod blocks need not be operable in

<>startup>> mode. If the .first column cannot ke met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested, immediately and, daily thereafter; this condition last longer than sIeven days the system with if the inoperable channel sha.'Ll be tripped. Xf the first column cannot be met for both trip systemsboth trip systems shall be tripped.

24 w is the recirculation loop flow .in percent of design. Tri.p level setting is in percent of rated power (3I293 NMt;).

A ratfi,o of FRP/CHFLPD 4,1.0 is permitted at reduced. power, See Speci.fication 2. 11 for APRN control rod k>lock setpoint.

3. IRM downscale is, bypassed when it is on its lowest range
4. This function is',bypassed when the count rate is P 100 cps and IRM above range 2.
5. One instrument channe3,;,i.i . ~ ane APRN or IRN or RBM, per trip system may be bypassed except only one of four SRM may be bypassed.
6. IRM channels A, E, C, G all i.n range 8 bypasses SRM channels A 6 C functions.

IRN channels B, F, D, H all i.n range 8 bypasses SRM channels B 6 D functions.

The trip is bypassed when the reactor poWer is '+ 30%.

8. This function is, bypassed when the mode Switoh ~is~ placed in Run.
9. This function is on.'Ly active when the mode~ sWit'ch'iS in Run.

This function is automatically bypassed when the IRN instrumentation is operable and, not high.

10. The inoperative trips are produced by the following functions:
a. SRN and IRN (1) Local "operate-cali.brate" switch'not in operate.

(2) Power supply voltage,low.

77

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3 3 REACTIVITY CONTROL 0 ' REACTIVITY CONTROL

b. During the a~ The capability shutdown of the RSCS to procedure no rod pxoperly fulfill movement is its function permitted shall be between the verified by the testing following tests:

performed above 20% power and Sequence portion the Select a reinstatement of sequence and the RSCS attempt to restraints at or withdraw a rod above 20'X power. in the remaining Alignment of rod sequences. Move groups shall be one rod in a accomplished sequence and prior to select the performing the remaining tests. sequences and attempt to move co Whenever the a rod in each.

reactor is in Repeat fox all the startup or sequences.

run modes belo~

20% xated power Group notch the Rod Worth portion - For Minimizer shall each of the six be operable or a comparator second licensed circuits go, operator shall through test verify that the initiate;.

operator at the comparator reactor console inhibit; verify; is following the reset. On control rod seventh attempt program. A second test is allowed licensed operator to continue may not be used in until completion leiu of the BWM dur- is indicated by ing scram time test- illumination of ing in the startup o test complete run modes below 20$ light.

of rated thermal b. The capability power. of the Rod Worth Minimizer (RWM) shall be verified the l24 followingby checks:

LZHITING CONDITIONS FOR OPERATION SURVECLIJLNCE REQU~IREMFNTS il

3. 3 REACTIVITY CONTROL 4 ' 'REACTIVITY CONTROX dO If The correctness Specif icaitions 3+3 B 3 ~ aL , of the through .c contre)l roid cainncit 'be met vithch.awal the reactor sequence, shall not, be input started, or the reactor is if ciom]put RtR to'he er in the run or shall be startup modes at verified less than 20% biefori!

rated power, shall be brought it reactor start6p or to a s?>utdem s hutd can.

condition immedia~tely. 2 The RthC ciom]puter on line diagnostic test sha,ll be successf ully per.f os~ d.

30 Prior tci startup proper a,nn'uncia,ti on aif the sel ectiain

,error of', at least on' out-of-sequence'ontrol rod shall be verified.

4~ Prior to startup, the re block function of t: he RWC shall be verif:i.ecl'y moving ann

'125 out-of-'equence

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3 ACTIVITY CONTROL 4 3 REACTIVITY CONTROL c Scram Insertion Times Co Scram Insertion Times The average scram After each refueling outage all insertion time, based operable rods sha11 be scram time on the deenergization tested from the fully withdrawn of the scram pilot position with the nuclear system valve solenoids as pressure above 950 psig (with time zero,,of all saturation temperature.) This operable control rods testing shall be completed prior in the reactor power to exceeding 40$ power. Below operation condition 20fu power, only rods in those shall be no greater sequences (A12 and Ap4 or B12 than: and Bg4) which were fully with-

% Inserted From Avg. Scram Inser- drawn in the region from 100$

rod density to 50$ rod. density shall be scram time tested.. The 5 0. 375 sequence restraints imposed 20 0. 90 upon the control rods in the 50 2 0 100-50 percent rod density groups 90 3.5 to the preset power level may be removed. by use of the indi-

2. The average of the vidua1 bypass switches associated scram insertion times with those control rods which for the three fastest are fully ol."partially withdrawn operable control rods of all groups of four and are not within the 100-50 control rods in a percent rod density groups. In two-by-two array order to bypass a rod, the shall 'be no greater actual rod axial position must than: be known; and the rod must be in

% Inserted From Avg. Scram Inser- the correct in-sequence position.

Full Withdrawn tio Times

2. At 16 week intervals, 10$ of the 5 0 ~ 39 operab'e control tod drives 20 8'.954 2 120 shall be scram timed above 50 800'sig. Whenever such scram 90 3 800 time measurements are made, an
3. The maximum scram evaluation shall be made to insertion time for provide reasonable assurance 90% insertion of any that proper control rod. drive operable control rod performance is being shall not exceed 7.00 maintained.

seconds.

128

LIMITING CONDITIONS FOR OPERM'ION SURVEILLANCE REQUIREMENTS 3 3 REACTIVITY CONTROL 0~3 8 'IVITY CONTROL D Reactivit Anomalies p., keactiv t Anomalies The reactivity equivad.ent During the Startup test of the difference between program and startup the actual critical rod following refueling configuration and the outages, the critical rod expected configuration configurations will be during power operation compared to the expected shall not exceed 1% 6 k. configurations at selected If this limit is exceeded, the reactor will be shut operating conditions.

These comparisons will be down until the c:ause hais 'used as base data for been determined and react;ivity monitoring corrective actions have during subsequent power been taken as appropriate,. ppig aition throughout th e fuel cycle. A',specific power operatinq conditions, the critical rod configuration will,'be.

compared to the configuration expected based upon appropr iat;ely correct ed pas t; data. This comparison will be made at least every month.

full power E

If S peci ficat;ions 3.3. C and D above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the shutdown condit:ion wit:hin 24 hours.

129

integrity is maintained, the possibility of a rod dropout accident is eliminated. The overtravel position feature provides a positive check as only uncoupled drives may reach this position. Neutron instrumentation response to rod movement provides a verification that the rod is following its drive. Absence of such response to drive movement could indicate an uncoupled condition. Rod position indication is required for proper function of the rod sequence control system and the rod worth minimizer.

2~ The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage to the primary coolant system. The design basis is given in subsection 3. 5.2 of the FSAR and the safety evaluation is given in subsection 3.5.4. This support is not required if the reactor coolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a drive housing.

Additionally, the support is not required if if all control rods are fully inserted and an adequate shutdown margin with one control rod withdrawn has been demonstrated, since the reactor would remain subcritical even in the event of complete ejection of the strongest control rod.

3~ The Rod Worth Minimizer (RWM) and the Rod Sequence Control System (RSCS) restrict withdrawals and insertions of control rods to pxe-specified sequences All patterns associated with these sequences have the characteristic that, assuming the worst single deviation from the sequence, the drop of any contxol rod from the fully inserted position to the reactor position of the control to sustain a power rod drive would not cause the excursion resulting in any pellet average enthalpy in excess of 280 calories per gram. An enthalpy of 280 calories per gram is well below the level at which rapid.

fuel dispersal could occur (i.e., 425 calories per gram). Primary system damage in this accident is not possible unless a significant amount of fuel is rapidly dispersed. Ref. Sections 3.6. 6, 7. 7.A, 7. 16.5.3, and 14.6.2 of the FSAR and NEDO-'10527 and supplements thereto.

In performing the function described above, the RWM and RSCS are not required to impose any restrictions at core power levels in excess of 20 percent of rated. Material in the cited reference shows that it is impossible to reach 280 calories per gram in the event of a control rod drop occurring at power greater than 20 percent, 132

regardless of the rod pattern. Thxs xs t:rue ror ai.z normal and abnormal pattterns incl.uding those which maximize individual control rod worth.

At power levels below 20 percent of rated, abnormal control rod patterns could produce rod worths high enough to be of cioncern relative to t]he 280 calorie per gram rod drop limit,. In, this range the RWM and the RSCS constrain the control rod sequences and patterns to those wh:ich involve only acceptable rod worths.

The Rod Worthy Minimizer and the Rod Sequence Control System provide automatic supervision to assure that sequence control rods will not be withdrawn ox out'f inserted; i.e., it limits operator deviati.ons from planned withdlrawal sequences. Ref. Section 7.16.5.3 Thiey serve as a backup to procedure cont:rol of'he FSAR,.

of control rod sequences, which limit the maximum reactivity worth of control xods. In the event that the Rod Worth Minimizer is out of service, when required, licensed operator can manually fulfillthe a'econd control rod pattern conformance functions of this system. In this case, the RSCS is backed up by independent: procedural. controls to assure conformance.

The functions of the RWM and RSCS make specify a license limit on rod worth to pr'eclude it unnecessary to unacceptable consequences in the event of a contxol rod drop. At low powers, below 20 percent, tliese devices force adherence to acceptable rod patterns. Above 20 percent of rated power, no constraint on x'od pattern to assure that rod drop accident consequences is'equired are acceptable. Control rod pattern Constraints above 20" percent of rated power are imposed by power distribution requirements, as defined in Section 3.5. I, 3.5.J, 4,.5.,I, and 4.,5.J of these technical specifications. '.Power level for automatic'. bypas. of the RSCS function is:sense.d by firSt stagi> turbine pressure.

Because the instrument. has an instrument error of +10 percent of full jower the nominal instrument setting is 30 percent of rated power.

Because it, is allowable to bypass cer4ain rods in the RSCS during scram time testing below 201~ of rated power in the startup or run modes, a second licensed operator is not an acceptable'x'<bs'tit',use f'r the RWN during thLs testing.

The Source Range Monitor (SRM) system performs no automatic: safety system function. It it funct;ions; i.e., has no scram does provide the operator with a vi,sual indication of neut:ron level. The cbn<iequences of reactivity accidents are functions of the initi.al neutron flux. The requirement of a6 )l.eastounts per second assures that any transient, should, begins at: or above the initial yalu<~ of 10-s of rated it occur, power used in the analyses of transients from cold conditions. One operable SRM channel would be adequate to monitor the approach to criticali.ty using homogeneous patterns of scattered control. rod withdrawal. A minimum 133

~ .

,of two operable conservatism.

SRM ~ s are provided as an added

5. The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation. Two channels are provided, and one of these may be bypassed from the console for maintenance and/or testing. Tripping of one of the channels will block erroneous rod withdrawal soon enough

,to prevent fuel damage. The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists.

A limiting control rod pattern is a pattern which results in the core being on a thermal hydraulic limit (i.e., MCPR -1.27 or LHGR = 13.4) . During use of such patterns, it is judged that testing of the RBM system prior to withdrawal of such rods to assure its operability will assure that improper withdrawal does not occur. It is normall'y the responsibility of the Nuclear Engineer to identify these limting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns. Other personnel qualified to perform these functions may be designated by the plant superintendent to perform these functions.

C. Scram Insertion Times The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than 1.05. The limiting power transient is that resulting from that of Rod Withdrawal Error {RWZ) .

Analysis of this transient shows that the negative reactivity rates resulting from the scram (FSAR Figure N3.6-9) with the average response of all the drives as given in the above specification, provide the required protection, and MCPR remains greater than 1.05.

On an early BWR, some degradation of control rod scram, performance occurred during plant startup and was determined to be caused by particulate material (probably construction debris) plugging an internal control rod drive filter. The design of the present control rod drive (Model 7RDB144B) is grossly improved by the relocation of the filter to a location out of the scram drive path; i.e.',

interfere with scram performance, even if it can no longer completely blocked.

l34

't'h~ ~ <J<.<rr'~<l<<l per L<)rrnar<c>> of the orig'Lnal drive (Cl<urtCUer<<<<a,l uri J~;r

  • > ty Oper.<tiri<l conditions and the insensitivity of: the r<<le~~ i<in~.'d <lr ive (CRD7RDB1rlOB) has been demonstrated, by a sex i< ri of er<gi ne>>ri.ng tests under simulated reactor operating, ror litions. Th>> successful performance otf thi~ new drive un<ler actual operating conditions has also been demonstrated by consistently good in-service test results for. plants using the new drive and may be:inferred from plants using the older model drive with a modified (larger screen size) internal filter which is less prone to pl,ugging. Data has been documented by surveillance reports in various, operating plants. These include Oyster Creek, MonticelloDresden 2 and Dresden 3,. Approximate.'Ly 5000 drive tests have been recorded to date. Following identi.Eication of the "plugged filter'< problem, very frequent scram tests were necessary to ensure proper performance. However, the more frequent sdram tests. are now considered totally unnecessary and unwise for the following reasons: Frratic scram performance has been identified as due to an obstructed drive filter in type <<A<<, drives. The, drives in BFNjP are, of the new <<B<< type des.ign whose scram performance is unaffected by filte'r condition'. 2 The dirt load is primarily released during startup of the reactor when the reactor and its systems are first 0 subjected to:Elows and pressure and thermal stresses. Special attention and measures are now being taken to assure cleaner systems. Reactors with drives identical or similar (shorter stroke,, smaller piston areas) have operated through many reiEueling cycles with no sudden oX erratic changes in scram performance. This preoperational and staxtup test9.ng is 'sufficient to detect an,omalous drive perfox'mance. 3 The 72-hour outage limit which initiated the start of the frequent, scram testing is arbitrary, having no ~ logical basis other than quantifying a <<major outage<< which might reasonably be caused by an event so severe as to possibly affect drive performance. This requirement is unwise because it provides an incentive for shortcut, actions to hasten retuxning <<on line" to avoid the addit:Lonal testing due a 72-hour outage. The surveilla,nce requirement foIr scram testing of all the control rods after each refueling 'outage and 10% of, the control rods at 16-week intervals is adequate for determining the operability of the control rod system yet is not so,frequent as to cause excessive wear on the control rod system compceents. The numerical values ass:igned t'o the predicted scram performance are based on the analysis of data from other BWR ~ s with control rod drives the same as those on Br owns Ferry Nuclear Plant., The occurrence of scram times within the limits, but significantly longer than the average, should be viewed as an inclication of systematic problem with contxol rodi drives especially if the number of drives exhibiting such scram times exceeds eight,'he'llowable number of inoperable rods,. l35 In the analytical treatment of the transients, 390 milliseconds are allowed between a neutron sensor reaching the scram point and the start of negative reactivity insertion. This is adequate and conservative when compared to the typically observed time delay of about 270 milliseconds. Approximately 70 milliseconds after neutron flux reaches the trip point, the pilot scram valve solenoid power supply voltage goes to zero an approximately 200 milliseconds later, control rod motion begins. The 200 milliseconds are included in the allowable scram insertion times specified in Specification 3.3.C. 'n order to perform scram time testing as required by specification, 4.3.C.l, the relaxation of certain restraints in the rod sequencebe control system is required. Individual rod. bypass switches may used as described in specification 4.3.C.l. The position of'ny rod bypassed. must be known to be in accordance with rod withdrawal sequence. Bypassing of rods in the manner described in specification 4.3.C.1 will allow the subsequent withdrawal of any rod. however," scrammed in the 100 percent to 50 percent rod density groups; it willlevel maintain group notch control over all rods in the 50 percent range. In addition, RSCS will prevent movement of rods in power the 50 percent density to a preset power level range until the scrammed rod has been withdrawn. During each fuel cycle excess operative reactivity varies as fuel depletes. and as any burnable poison in supplementary control is burned. The magnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup proqresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state. Power operating base conditions provide the most .sensitive and directly interpretable data relative to core reactivity. Furthermore, using power operating base conditions permits frequent reactivity comparisons. Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% hK. Deviations in core reactivity greater than 1% hK are not expected and require thorough evaluation. One percent reactivity limit is considered safe since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the reactor system. 136 LIMITING CONDITIONS FOR, OPERATION SURVEILLMC.'E REQ~UIREMENTS 3~4 STANDBY LI UID CONTROL SYST'gM 4 ','TANDBY LI UXD CONTROL' S EM A licabilit ~A~licabi li~t Applies to the operating status Applies to the surveillapcq of the Standby Liquid Control requirements of the Standby System. Liquid, Control System. Ob'ective Object.ive To assure the availabil.'ity of a Tb trerifiy the opegability of, system with the capability to the Standby Liquid Control shut down the reactor and System. maintain the shutalown condition without the use of: control rods. S ecification Steicification Ao Normal S stem The operabilit'y. of the
    1. The standby liquid Standby Liquid Control control system shiall System shall be verified be operable at all by the perfor'mance of the times when there is fo3.lowing tests:
    fuel in the react;or vessel and the l1. At least once pei reactor is not in a. month each pump loop shutdown condition shal3. be funct.ionally with al:L operable tested. control rods fully inserted except as 2. At least once during specified in 3.4.,B.1. each operating cycle: a0 Check that thie isetting of the system relief valves is 1425 + 75 psig. b Manually initiate the system, except explosive valves. Pump boron solution 137 through the' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS s 6 gag~ s , 4~6 P Y S ST ND Y
    5. Steady state operation with G. Structural Inte rit both recircu1ation pumps out of service for up to Table 4,.6.A together 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is. permitted. with,suppl'ementary During such interva1, notes, specifies the restart of the recirculation inservice inspection pumps is permitted. The surveillance plant shall 'be placed in a requirements of the hot shutdown condition reactor coolant within 12 .hours un1ess one system as follows:
    loop is returned to service. a. areas to be The total'lapsed time in inspected natural circulation percent, of areas and'ne pump operation must b. no greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. to be inspected during the inspection G 'tructural Inte rit interval
    1. The structural c. inspection integrity of the frequency primary system shall be maintained at the d. methods used for level required by the inspection original acceptance standards throughout 2~ Evaluation of the life of the inservice inspections plant. The reactor will be made to the shall be maintained acceptance standards in a cold shutdown specified for the condition until each original equipment.
    indication of a defect has been 3 ~ The inspection investigated and interval shall be 10 evaluated. years. 4 ~ Additional inspections shall be performed on certain circumferential pipe welds as listed to provide additional protection against pipe whip, which could damage auxiliary and control systems Feedwater- GFW-9 ~ KFW-13 ~ GFW 12r GFW 26 g GFW- 29' 'FW-31 KFW 39 e GFW- 15 ~ 196 KFW-38, and GFW-32 LIMITING CONDITIONS POR OPEKLTION SURVEILLANCE RRQUXREMENTS ' ~ 6 PRIMARY SYST 'OUNQQWY 4 ~ '6 PRX Main steam-GMS-6, KMS-;?4, GMS-32>> KMS- 10 4>> GMS-15>> and GMS-,24 RHR - DSRHR- 6>> DSRHR-7 ~ and DSRHR-4 Core Spray-DSCS.-12 ~ 'SCS-11, DS~CS-5>> and DSCS-4 Reactor Cleanup -DSRWC-4 ~ DSRWC"3, DSRWC-6, and DSRWC-5 HPCX - THPCX-70 THPCX-70A THPCX-71, and THPCX-72 5., System hydrostatic tests in accordance with Article IS-590 of .Section XI of the ASME Code at, or near the end of each inspection interval and prior to startup folloving each refueling outage. The pressure-temperature iLimits for these tests be vill in accordance with specification 3 ~ 6 ~ A,.3 ~ REFERENCE 1 ~ Plant Safety Analysis i[BFNP PSM'. subsection 4 ~ 12) 197
    n. 'Retuc ging In~ter ccke Complete functional testing of all refueling interlocks before any refueling" outage will provide positive indication that the interlocks operate in the situations for which they were designed. By loading each hoist with a weight equal to the fuel assembly, positioning. the refueling platform, and withdrawing control rods, the interlocks can be subjected to valid operational tests. Where redundancy is provided in the logic circuitry, tests can be performed to assure that each redundant logic element can independently perform its function.
    B. Core Monitorin Requiring the SR' to be functionally tested prior to any core alteration assures that the SRM's will be operable at the start of that alteration. The daily response check of the SRM~s ensures their continued operability. t REFERENCES
    1. Fuel Pool Cooling and Cleanup System
    '1O. 5> (BFNP FSAR Subsection
    2. Spent, Fuel Storage (BFNP FSAR Subsection 10.3)
    LXNITINQ CONDITIOSS ',FOR OPilRLTXQN SUEARELU~E REQI'JIRENKC'S OP 3~1 P.IRE PR 1. 11 FIRE PRC73.'ECXKIQ~~ST $ 8 ~1 ica~bi lit P Applies to operat:Lncg states of the high pressure water and CO~ Appl,ies to the surveillance fire protection systens for thee requirement,s of t,he high reactor building, diesel pres, sure water anci C,'O~ f.ix'e generator buildings;, control protection systems for the bay~ intake pus!pi!kg station, 'reactor building, diesel cable tunnel to the intake generator'buil<Lngs, contrdL pumping station, and the fixed bag, i&take pumping static+'i' spray. system for cable trays cable tunnel to the intake ~ along the south wall of the pu~ing qtation. and the fixed turbine buildng,. 'elevation 586. system fox cable txays 'pxay alone~ the south wall of th6 turbine building, elevatiocI! the corresponding 1$ 586'hen Ob ective: for operation ade ~in galIting'onditions To assure availabi1'it@ of Pire effe t Protect'ion Systeras. ghat've: To verify t!he operability of the E'ire Protection Systews ~ A Hi h Pressure Fire Protection S stem
    1. The High Pressure, A+ Px'e > we P. re Fire Protection System shall have 1~ High Pressure Fire
    a. Two (2) 'igh Protection System pressure fire Testing pumps operabLe and a.ligned to Item Frequen~cr the 'higlx preeeare fire Simulated Onc: e/;year hea'der. automatic and manual
    b. Automatic. actuation of initiation logic high pressure operable.
    pumps and auto-matic valve operability b,. Pusan~ Once~montlL! Operability c Deleted
    d. :Pump Once/3 ye::!
    capability 3i7 3 1 'I BA~~~~ Th~ I'.icth Pressure Fire and CO> Fire Protection specifications are provided in order to meet the preestablished'evels of operability during a fire in either or all of the three units. Requiring a patrolling fire watch with portable.'fire equipment the automatic initiation is lost vill provide (as does the if automatic system) for early reporting and immediate fire fighting capability in the event of a fire occurrence. The High pressure Fire Protection System is supplied by three pumps aligned to the high pressure fire header. The reactors may remain in operation for a period not to exceed 7 days pumps are out of service. if Xt at least'wo pumps are not made two operable in seven days or if all pumps" are lost during this seven day period, the reactors will be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For the areas of appLicability, the fire protection water distribution system minimum capacity of 266¹ gpm at 250 ~ head at the fire pump Ridcharge consists of the following design loads: 1 Sprinkler System (0.30 gpm/ft~/4440 ft* area) 1332 gpm 2~ 1 1/2< Hand Hose Lines 200 gpm 3~ Faw Serv'ice Water Load 2~m 'IOTAL 266¹ cpm The CG, Fire Protection System is considered operable with a minimum of 8 1/2 tons (0.5 tank) CO< in storage for units 1 and 2; and a minimum o'f 3 tons (0 ~ 5 tank) COq in storage for unit 3. An immediate and continuous fire watch in the cable spreading room or any diesel generator building area Co> fire protection is lost in this room and will be established vill continue until if CO~ fire protection is restored. To assure close supervision of fire protection system activities, the removal from service of any component in either the High Pressure Fire System or the CO~ Fire Protection System for any reason other than testing or emergency operations will require Plant Superintendent approval. Early referting'nd immediate fire fighting capability in the event of.. a fire. occurrence vill be provi'ded (as with the automatic system) i>y requiring a patrolling fire watch than one. detector for a given protected zone is inoperable if more A roving fire watch for areas in which automatic fire suppression systems are tn b~ installed will provide additional interim fire protection for. areas that have been determined to need additional protection. 356 Thr t'X'ro nrot~ction system is designed to .supply the required flow-.ind pr~...sure. t'o <<n'ndividual load 1:isted on Table 3.11.A wh'il, maintaininq a design raw service water load. of 1132 gpm I~
    4. 11, BASES Periodic t~sting of both t:he High Pressure Fire System and the CO~i, Fiie Protection System will provide positjvq gndication of their operability. X'f only one of the pumps supplying the High Pressuie Fire Sys;tern is. ojmrable,, the pump that is operable wilg be. checked immediately. and, dlaily, thereafter to dewenstrate operability. If the CO., Fire Pr!otection 8yhtW ibecoees inoperable in the, .cable spreading room, one 125-pound i[or larger) i fire extinguishere wil1 be placeIB at each entrance to the cable spreadinq room.
    Annual testing of automatic valves,and control devices is in accordance with NPPA code Vol. XI, 1975 sect:ion 15',paragraph 6O15,. More frequent testing would require excessive automatic syst: em inoperability, since there are a large number of automa-tic, valves installed and'arious portions of the system must be isolated during an extended period of thne during 'this test. wet fire header flushing, spray header in6~tion for block'age, an9 nozzle inspection for blockage vill piWerit, detect~ and remove buildup of: sludqe or other material to ensure continued operability. System flushes in congunc'tibn with the semiannual~ addition of bioc'id'e. to the- Raw, Coolirig Water System will help pri'.vent the growth of crustaceans, which could reduce nozzle discharge. Semiannual tests of hIeat a<ndI smo]ke dete'ctors are in accordance wit,h the NFPA codle. with the exception of contanuous strip heat detectors panels, all non-class A supervi'sed det:ector circuits which provide, alarm only are t.ardwired,through conduits ajnd/or cable trays. from the detector to the mai.n control room alarm. panels vith no active components between., 1Moii-class A circuits.a3Lso actuate the HPCI vater-fog system,,the CO~, s'stem in the diesel generator huxlRings, and'isolate ventilation in shutdown board rooms. The test frequency. and'ethods specified: are )ustified for the fol.lowing reasons: 1., An analysis. vas made of worst:-case .~fire'detection circu$ ts at Browns Ferry to deterniine the. probability',of no undetected failure. of .the circuits occurring between system test specifi~d;'in t!h~ surveillance requirements. A circuit is times's defined as the wire connections and components that affect. tr<<nsmission of, an;,ctlalrmi: signal between tHe fire detectors and the control roo~n a~rinunciator., Three c.ircuits were analvzed which vere repiesentative of an alarm-only circuit, a water-foq'ircuitaind a. CO> circuit. We~ spreading roen IB smoke detector was selected as the. worst-case alarm-only circuit because it 'hadl the largest number of wires and connections in a single circuit;. The HPcx w'ater-'-fog circuit: was ..-elected'or <<nally!sis'because circuit in the ari.a of'pplicability it is the only, vater-fog for technical specification's. Th'e Standby Diesel Generator Room A CO> 357