ML20126B503

Amend 59 to License DPR-33 Changing Tech Specs for Min Test & Calibr Frequency for Surveillance Instrumentation.Tech Specs Encl

ML20126B503
Person / Time
Site:	Browns Ferry
Issue date:	02/25/1980
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML20126B506	List: ML20126B503 ML20126B509 ML20126B512 ML20126B515
References
NUDOCS 8003130310
Download: ML20126B503 (98)

Text

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UNITED STATES

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g, NUCLEAR REGULATORY COMMISSION 7, g " 3 af WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO.60-259 BROWNS FERRY NUCLEAR PLANT, UNIT N0. 1 I

AMENDMENT TO FACILITY OPERATING LICENSE 4

Amendment No. 59 License No. DPR-33 1.

The Nuclear Regulatory Commission (the Comission) has found tMt:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated October 4,1979 as supplemented by letters dated j

January 15, 1980 and January 29, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility License No. DPR-33 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 59, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

80031303/Of

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION To Thomas Ippolito, Chief Operating Reactors Branch #3 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: February 25, 1980

ATTACHMENT TO LICENSE AMENDMENT NO. 59 FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Revise Appendix A as follows:

1.

Remove the following pages and replace with identically numbered pages:

vii/viii 25/26 133/134 159/160 181/182 9/10 29/30 145/146 167/168 218/219 11/12 92/98 147/148 169/170 220/221 15/16 111/112 149/150 171/172 254/255 17/18 131/132 157/158 172a 276/277 330/331 2.

The underlined pages are those being changed; marginal lines on these pages indicate the revised area. The overleaf pages are provided for convenience.

3.

Add the following page:

172b i

LIST 0F TABLES (Cont'd)

Table Title Page No.

4.2.F Hinimum Test and Calibration Frequency for Surveillance 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.I KAPLHCR vs Average Planar hposure

................ l'9D,172, 1T2-a, 17 2-b 1

3.6.H Shock Suppressors (Snubbers) 1 4.6.A Reactor Coolant System inservice Inspection Schedule 209 3.7.A Primary Containnent Isolation Valves 250 3.7.B Testable Penetrations with Double 0-Ring Seals......................

256 3.7.C Testable Penetrations with Testabic Cellows....

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 263 3.7.H Testable Electrical Penetrations 265 1

4.8.A Radioactive Liquid Waste Sampling and Analysis 287 4.8.B Radioactive Gaseous Wasta Sampling and Analysis..

288 3.11. A,

Fire Protection System Hydraulic Requirenients...

324 6.3.A Protection factors for Respirators 343 6.8.A Minimum Shift Crew Requirements..........

360 i

vii Amendment No. 59

LIST OF ItLUSTRATIONS P' age No.

Figure Title 2.1.1 APM Flow Reference Scram and APRM Rod Block 13 Settings....................

2.1 -2 APM Flow Bias Scram vs. Reactor Core Flow 26 4.1 -1 Graphic Aid in the Selection of an Adequate 49 Interval Between Tests.............

119 4.2-1 System Unavailability...............

3.4-1 Sodium Pentaborste Solution Volume Concentration i

138 Requirements..................

1 2

3.4-2 Sodium Pentaborate Solution Temperature 139 Requirements..................

173 3.5.2 Kf Factor.....................

3.6-1 Minimum Temperature 'F Above Change in Transient Temperature...................

188 3.6-2 Change in Charpy V Transition Temperature Ys.

189 Neutron Exposure................

6.1 -1 TVA Office of Power Organization for 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 Organization 364 i

VIII Amendment No. 35,47

4 4

S WrYI 1.tMIT I.inninc w.*Ys e S N C d

1.)

Flir:t Cl.r;DHING INT}EllWY P1 FUEL, Cl.ADDillC INTECl: TTY In the event of operation with the core maximum fraction of limiting power density (CHFLPD) greater than fraction of rated thernal power (FPf) the setting shall be modificd as follows:

SS (0.66W + 54%) FRP CMFLPD For no combination of loop recircu-lation flow rate and core thermal power shall the APRM flux scra= trip setting be allowed to exceed 120%

of 1ated thermal power.

(Note: These settings assume operation within the basic thermal hydraulic design criteria. These criteria arc IJICR y 18.5 kw/f t for 7X7 fuct ands l

13.4 kw/ft for 8X8, 8x8R, and P8x8R fliel, MCPR limits of Spec 3.5.k. If it is determined that either of theso design criteria is being violated during operation, actica bhall be initiated within 15 minutos to restore operation within prescribed lir:itt..

Surveillance requirements for Ali.:

scram setpoint are given in specification 4.1.B.

2.

APRM--When the reactor mode r.wi t ch is in the STARTUP POSITION, the APRM scram shall be set at Icss than or equal to 15% of rated power.

3.

IRM--The IRM scram shall be set at 1ess than or equal to 120/125 of full scale.

B.

APRM Rod Block Trio Settinn l

b.

C';re ';'hemal Powe* Limit (Peacto-Prersure 1 00 psia) 8 The APM Rod block trip setting r.hSI)

~

be:

D er, the reactor pressurc is less thsn or equal to 800 pala, 9

i Amendment No. 59

SAFETY I,IMIT LIMITItiG CAFETY SYS'19::/ CETTIt.M

+1. 1 FUEL CLADDIrlG INTEGRITY 2.1 FlFr:L CLADDIrlC litTECi:ITY or core coolant flow is les:

RB1 (0.66W + k2".)

than 10% of rated, the core thermal power shall not ex-where:

ceed 823 MWt (about 25% of i

U rated thermal power).

RB = R d block settir.c in farcent of rated thermal pwer (3?93 NWt.)

W

= Loop recirculation f'ov rate in percer.t of rated ' rated lunp recirculn'.ico f wu rate equala 3L.2 X 10 lb/ar)

In the event of coeration v!th the core maximum f* action of limiting power density (CMFLPD) greater than fraction of rated '

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

4 i

S4< (0.66W + h25 } FRP CMFLPD i

a C.

VThenever the reactor ic in C.

Scram and icoluation--; ', A i n.

above the shutdown condition with reactor low water ve:mel zero 1.cel irriair.ted fuel in the reac-ter vessel, the water level shall not be lese than 17.7

]

in, above the top of the D.

Scram--turbine stop < 10 net -ect norna.'. active fuel zone, valve closure valve cl.

e E.

Scram--turbine j

control valve ihmo trip of 1

1.

Fast clocure tN-nnt v i.

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aolt a l '..

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2.

Loso of cor. trol > D O pai;7 oil prescu.re F.

Scram--low con-

~ 23 incher.

densor vacuum ib, vacuu:a G.

Scram--main a tearn 1 In pert,2ut line isolation valv. clea re

+

H.

Main steam isolation L 8:S pci;,

l valve clocure--nuclear cyste.9 low pressure 10 Amendment No.

59

S AFETY L DitT LLMITING S A,7ETY SYSTE?4 SETTING A.1 Fuel Cladding Integrity 2.1 Fuel Claddine Integrity I.

Core spray and L?CI 3 378 in, actuation--reactor above vuonel low water level zero g

J.

HPCI and RCIC 1470 in, actuation--reactor above vessel law water level zero K.

Main steam isola-1470 in.

tion valve closure-- stove vesaci reactor low water zero level 1

4 l

11 Amendment No. 59

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1.1 R AS ES :

FUEL CLADDit:C INTECRIT*( SAFETY LIMIT l

The fuel cladding represents one of the physical barriers which separate radio-active materials from environo. The integrity of thic 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 cunalative and continuously seasurable.

Fuel cladding perf orations, however, can result f rom thermal stresses which occur f rom reactor operation significantly above design conditions and the protection system sotpoints. Uhile fission product nigration f rom cladding perf ormation is just as seasurable as that f rom use-related cracking, the thermally-caused cladding perf orations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deteriors-tion. Therefore, the fuel cladding safety limit is defined in terms of the reactor operating conditions which can result in cladding perforation.

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 Saf ety Limit is defined with margin

.co the conditions which would produce onset transition belling (NC?R of 1.0).

This establishes a Safety Limit such that the minimum critical power ratio (MCFR) is no less than 1.07.

MCFR >1.07 represents a conservative margin relative to the conditions required to maintain fuel cladding integrity, onsce of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possiblity of clad f ailure.

Since boiling transition is not a directly observable parameter, the sargin to boiling transition is calculated f rom plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution. The margin for each fuel assembly is characterised by the critical 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 minious value of this ratio for any bundle in the core is the minimum critical power ratio UCPR). It is assumed that the plant operation is controlled to the nominal protective setpoints via the instru-cented variables, i.e., normal plant operation presented on Tigure 2.1.1 by the ne inal exaeeced finw cantent lire. The Safetv Limit (WcPR of 1.07) han enf ficient conservatism to assure that in the event of an abnormal operational transtant iniciatef f rom a normal operating condition (MCPR > limits specified in specification 3 5.K)more than 99 9% of the fuel rods in the core are expected to avoid boiling transition. The margin between F.CF3 of 1.0 (onset of transition boiling) and the safety limit 1.07 ___ is d e rived f ron a detailed statistical analysis considering all of the uncertaLaties in moni-coring the core operating state including uncertainty in the boiling transition correlae. ion as described in Reference 1.

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

15 Amendment No. M. 47

l

.1.1 BASES Decause,the boiling transition correlation is based on a large quantity of rull scale data there is a very high confidence that operation of a fuel assembly at the condition of MCPR = 1.07, would not produce boiling tran-sition. Thus, although it is not required to establish the safety limit additional margin exists between the safety limit and the actual occurence of loss of cladding integrity.

However, it boiling transition vere to occur, clad perforation would not be expected. Cladding temperatures vould increase to approximately 0

1100 F vhich is below the perforation temperature of the cladding material. This has been verified by tests in the General Electric Test Reactor (CETR) 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 abould ever exceed 1h00 psia during normal power operating (the limit of applicability of the boiling transition corre-a 1ation) it vould be assumed that the fuel cladding integrity Safety Limit has been violated.

In addition to the boiling transition limit (MCPR = 1.o6) operation is constrained to a maximum LIICR of 18.5 kv/ft for 7x7 fuel and 13.4 kv/f t for-g(p 8x8 fuels. This limit is reached when the Core Maximum Fraction of Limiting Power Density equals 1.0 (CHFLPD = 1.0).

For the case where Core Maximun Fraction of Liniting Power Density exceeds the Traction of Rated Thermal Power, operation is permitted only at less than 100% of rated power and only with reduced APRM scram settings as required by specification 2.1.A.1.

At pressures belov 800 psia, the core elevation pressure drop (0 pover, 0 flow) is greater than b.56 psi.

At low powers and flove this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low powers and flow vill always be greater than 4.56 psi. Analyses show that with a flow of 28X10J lbs/hr bundle 9

flov, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus,3the bundle flov vith a 4.56 psi driving head vill be greater than 28x10 lbs/hr. Full scale ATLAS test data taken at pressures from,14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors this corresponds to a core thermal power of more than 50%. Thus, a core thermal power Ilmit of 255 for reactor pressures belov 800 psia is conservative.

t For the fuel in the core during periods when the reactor is shut down, con-sideration must also be given to water IcVel requirements due to the effect of decay heat. If water level should drop below the top of the fuel during this time, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation. As long as the fuel remains covered with water, sufficient cooling is available to provent fuel clad perforation.

16 AmendmentNo.

59

l i

1,1 BASES The saf et) 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 lov vater level trip (378" above vessel sero).

Pf.lTRINCE 1.

Central Electric BVR Thermal Analysis Basis (CETAB) Dat a, Correlation i

and Design Application, NEDO 10958 and NEDE 10958.

i i

i

)

17 Amendment No.

59 i

l 1

i

E D

PAGE DELETED 4

1 4

18 1

Revised 1-17-79 1.1 k?[f

,t.

J. & K.

Practor low yater level set po t,n t f or init ia t ion of flFCI and RCIC, c l o r, a n g e-1 t n s t e.im 15o13 Jon valves, ed staffing LI(.I and co re.4p r.1y putsp t.

These systens riaintain adequate coolant inventory and provide core cooling with the objective of preventir.g e.scessive clad temperatures.

The devit.n of these systems to adequately perform the intended func-tion is based on the specified low level scram set point and initia=

tion het points. Transient analyses reported in Section it of the FS AJL desonst ra te that these conditions result in adequate salaty eargins for both the fuel and the systes pressure.

L.

R e f e r en,cis, 1.

Linford. P. !., "Ana ly tical Me thods o f Pl ant Transient Evatustions for the Gener al Elect r ic Bo!!!ng Water Reac t or," NIDo-10802, Feb.,19D.

j

2. Generic Reload Fuel Application, Licensing Topical Report.

hTDE-24011-P-A, and Addenda.

1 I

0 knendment No. 25, 59 i

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Revised 1-17-79 1,2 BAS"J Therefere,."ollowin ar.y trancient pressure monitor hi8ber in the vessel.

that is severe enough to cause cencern that this safety limit was viet.ated, a calcu!Ation will be perfer. d usins ait available in'crutier. to de:.er.

w eine if the safety limit was vicisted.

FET.73 CiS_

1, Finnt Sa.l tty Anslysis (E m TSAR Section ll."))

!.3:2 3eiler and Pressure Vessel Code Secti:n III 2,

3.

C.'.3 Pipir,4 Cede, Ocetien D31.1 (20'P TO)P.

<;,eter 7::sel enc Appurter.ances !!echenicel.%s101 h,

Sc: sect ica 'a,2) 5.

Generic Reload Fuel Application, Licensing Topical Report.

NEDE-24011-P-A and Addenda, 29 35,4[59 Amendment No.

w.,,es,-,

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2.2 BASES 1

f

(

REACTOR CCOLANT SYS 3 INTEGRITY 1

i To meet the safety design basis, thirteen relief valves have been

~

installed on the unit with a total capacity of 82.6% of nuclear boiler rated steam flow. The analysis of the worst overpressure transient, l

(3-second closure of all main steacline isolation. valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering

-12 valves operable, results in adequate nargin to the code allowable overpra.ssure limit of 1375 psig.

Ao meet the ope (acional design, the analysis of the plant isolation l

transient (generator load reject with bypass valve f ailure j

to open)shows that 12 of the 13 relief valves limit peak system j

pressure to a va1Ue which is well below the allowable vessel over-a

.pressura of 1375 psig.

j i

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i Amendment No. 59 30

TABLE 4.2.B (C.unt.aued)

Calibration In s truwnt Check _

Functional Test Function none ooce/3 months (1)

Instrsament Citannel Reactor low Pressure (FS-48-93 & 94) none once/ operating cycle (4)

Core Spray Auto Sequencing Timers (Norm.nl Power) none once/ operating cycle (4)

Core Spray Auto Sequencing Timers (Diesel Power) none once/ operating cycle (4)

LFCI Auto Sequencing Timers (Normal Fower) none once/ operating cycle (4)

LFCI Auto Sequencing Timers (Diesel Power) cone once/ operating cycle (4)

RERSV A3. 81. C3. D1 Timera (Bormal Power) none once/ operating cycle (4)

RERSV A3. Bl. c3. D1 Timers (Diesel Power) none once/ operating cycle (4)

ADS Timer Amendment No.

59

~.

TABLE 4.2.8 (Continued)

Function Functional Test Calibration Justrument Check Instrument Channel (1) once/3 e nths none RER Fump Discharge Pressure Instr w nt Channel (1) once/3 months none Core Spray Pump Discharge Pressure Core Spray Sparger to RFT d/p (1) once/3 months once/ day Trip System Bus Power Monitor oncefoperating cycle N/A sone Imotrument Channel Coodensate Storage Tank tow Level (1) once/3 unaths mese Instrument channel Seppression rhg W r High Level (1) once/3 months mene Instrument Channel Reactor Eigh Water Level (1) once/3 e nths once/ day Instrument Channel ECIC harbine Steam Line Righ Floer (1) once/3 months moes Inatrument Channel BCIC steam Line Space Righ Temperature (1) ence/3 months asee S

W 3.2 BAhJn i

instrumentation which initiates a in addition en reactor proterrion instrumentation has been provided which reactor scram, prot ec t ive initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator cr-T'nio se t of speci-in serious consequences.

rors before they resultthe limiting conditions of operation for the primary l

fications providen initiation of the cere cooling systems, con-system isolation function, The objectives of standby gas treatment systems.

trol rod block and to assure the effectiveness of the protec-the Specifications are (1) required by preserving its capability to tive instrumentat inn when of such ay9tems even during tolcrate a ningle tatture of any enmponent periods when pertiona of such syntomo are out of service for maintenaner, trip settings required to assure adequate pet-i and (11) to preneribe the When necessary, one channel may be made inoperable for brief formance.

intervais to ennduct required functional tests and calibrations.

instrumentation that initiate or control ente Some of the octtings on the the high eno ting have tolerances explicitly stated where and containment values are ooth critical and may have a substantial ef f ect en points of other instrueentation, where only the hith ot and low enfety. The set low end of the uctring hno a direct bearing on safety, are chosen at a

inadvertent actua-level away f rom the normal operating range to prevent tion of the saf ety system involved and exposure to abnormal attuations.

valves is initiated by protective instru-Actuation of primary containmentin Table 3.2.A which senses'the conditione for which iso-mentation shown Such instrumentation must be available whenever pri-lation is required.

mary containment integrity to required.

isolation in connec ted The instrumentattan which inittares primary system ja i dual how a r rais,emoo t.

The low watet level inatrumentation 9et to trip at 177.7" (518" tbovr vennel zero) abnva the top of thn active fuel closen isol at ion val. van in and the RilR Syntem, tirvveli and Suppre ntou Chamber exhausta and drains Reactor Vater Cletnup 1.inen (Grnop 7.tnd.1 laolarinn valves). The lowwater reactor water level instrumentatinn that in set to trip when reactor level i# 109.7" '(470"above vessel sero) above the top of the active fuel closes the Main Steam 1.ine taoistion Valves and Main Steam, RCIC, and itPCI Drain Valves (Group I and 7).

Details of vaive grouping and required times are given in Specification 3.7.

These trip settings are closing in the cane of a break in the largest adequate to prevent core uncovery line assuming the maximum closing time.

is set to trip when reactor The low reactor water level inuttumentation that the top of the active water level la 109.7" (470" above vessel zero) nbuve f uel (Table 3.2.8) also initiate the RCIC and HPCI, i

i ili Amendment No.

59

(

I

\\

l

4-k 3.2 BASES and trips the recirculation pumps.

The low reactor water level.inntrumentation that is act to trip when reactor water level te 17.7" (.378" above vensel zero) above the top of the active fuel (Tnble ).2.8) inttintes the 1.PCI, Core Spray Pumps, contributes to ADS init iat ion un i 9t art s the diesel generatora. These trip setting levels were chosen to be high ennuy,h to prevent npurtoun actuation but low enough to initiate CSCS operatLun uo that post accident cooling can be accomplished and the guidelinea of 10 CFR 100 vill not be violated.

For large breaks up to the completc circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation in initiated in time to meet the above criteria.

The hip.h drywell pregnure instrumentation in a diverw nLanal tn the water level inatrumentation and in addition to initiating CSCS, it causen innlation nf Cronp9 1 and 8 innlation valveg.

Fnr the brenke dincunned above, this instrumentation will initiate CSC5 operation at about the same time as the low water level instrumentation; thus the resultn given above are applicable here also.

Venturts are provided in the main steam lines as a menna of meanuring steam flow and alno limiting the Inse of mass inventory f rom the vennel during a ateam line trenk accident. The primary function of the instra-mentation is to detect a break in the main occam line.

For the worst caec accident, main steam line break outside the dryvell, a trip setting of 140% of rated steam flov in conjunction with the flow limitera and main steam line valve closure, limits the masn inventory loss such that fuel is not uncovered, fuel cladding temperatures remain below 1000*F and release of radionceivity to the environo is well below 10 CFR 100 guidelines. Reference Svetton 14.6.5 FSAR.

i.tvintore monitoring instrumentation in prnvtiled in the ma in o t eam 1 1.u-enoneI to ilet.it L eo 9 in I he w.i s e n n. Tripn nie pinvtled nn thin i n u t r i. --

mentation nod when e xi erdmt. c.iunc e inmere of 190Lntion v a l v e 's.

. ie netting of 70n*f foi the main utenm itne tunnel detector 19 low ennngo (b detect icokn of the erder ot' l 's gpm; chus, it in capable of covering the entire spectrum of breaks. For larv.c breaks, t he h i t,h s t e act flow inutru-mentation is a backup to the temperature instrumentation.

liigh radiation monitnrn in the ma!n steam line tunnel have been prnviued to detect r.ross fuel failure as in the cont rol rod drop -accident, kitn the establinhed setting of 3 timeg normal background, and main occam line isolntton valvo cineure, t'14aton product release Lg limited r,n that

(

10 CFR 100 guidelines are not axcceded f or this accident. Reference l

l A n a ls m, vi t.n a n a kne l s e t. poi.n t, of 1 3 x norm 1[o 'uE'I c. 2 FSY. Wround, Secti l 6 a ornviced a so.

l ower c

Pressure instrumentation 14 provided to close the main steam isolattun valves in Run Mode when the mein steam line pressure dropa below d.b

-peig.

112 Amendment No.

59

~

rn

I

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/ 4. "1 nAS?S:

doce povide :he nperator utth a visus 1 indicatica of neu-tren 2 net, The conseque,cca u: re.c t ivity.iccid ento are functionn of the ir.itist neutron fl v..

The requireer.e.it of at leo 3t ) count.1 per necono aasurc2 that ar.f r.re nc ien t,

occur, n.t ns nt or above the initiil value of i

should it f r o.i 10* o f r a t e 1 para r i,. d i n t he en d y.ic., n t. ran i t e n t s Oac 09erabi 1.PM channel would he adequite cold randitions.

to rwnitor the approach in c rit!:slity usin.; hecoe.encou s A : sir. leer par t:rna of scat t e r ed cont rol rod v! thd r: val.

of two opersble SRai'a are provicied a s an cddeo conocevatin.

5.

The Rcd Block rionitor (RBM) is designed to auto:scically prevent fuel da. age in the even: of erroncous rod vi:hdraval from locstio.s of high power density durbt high power levci operacion. Two channels are provided, ina one of these may

'ce bypassed frco the console for etsintenance and/or testing.

Tripping of one of the channels vill block erroneous rod withdrasal soon enouuh to prevent f uci da:: age.

The spect-fled restric:Lons with one channel cut of re'vice conserva-j J

tively a)aure that fuel daaage vill not occur due to rod vithdraual cerors when this cond'. tion ex13:s.

A limiting control red pattern is a pattern which resultn in the core being on a thermal hydraulic linit, (fe, FtCPR given by Spec. 3. 6. K o r 1.H C R o f 18. 5 f o r 7x 7 e r 13.4 for 8x8,8x8R,& P8x8R). During use of such patterns, it is judged that testing of the RBt! system prior to w i t t. -

drawal of such rods to assure its operability will assure that improper withdrawal does not occur.

It is normally the responsibility of the If u c i c a r 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 suberitical at the rate fast enough to prevent fuci damage; ic, to prevent the t!CPR from becoming less than 1.07.

The limiting power transient is given in Reference 1.

Analysis l

of this trancient 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 protection, and t!CPR remains greater than 1, 0 7.

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

)

i Amendment No.

59

f 3.3/4.3 BASF}:

t particulate material (probably construction Jebrie) p?ui,ging an internal control rod drive filter. The design of the preatat control rod drive (Model 7RDB1443) is groasly improved by the j

relocation of the filter to a lo ation out 1

of the scrac drive path: 1.e., it ten no longer interfere with scram perfotmince, even if completely b1ceked.

4 4

The degraded perf ormance of the griginal drive (CP.D 7 RDL1 'ii A )

under dirty operating ronditions and the insensitivity of the rede=1aned drive (CKD7RDB1448) has been demon 9trated by a sert e of entineering test s under simulated reactor opc't. ting condttions.

i The successf ul perfornance of the n.u driv..

actual operating condit ions has also been demont erated by under conaistently good in-service test renutte for pinato uni..c. the l

Irive and may be inferred frem plante using t!.e olde r endel new driv with a modified Llarger screen size) inttiv.s1 filter which is 1 se prone to plugging. Da ta has been docus. nt ed by r e c. :11-lancJ i

reports in various operating pla tte.

Thorn teclude Oyster Creek, Monticello Dresden 2 an j Dreedor. 3.

App ri :tiuc t ely

{

$000 drive tests have been recorded to date.

i

. Tollowing identification of the " plugged filtet" problen, frequent scram testa were necessary to ensure proper perfermarce i

very However, the more frequent j

scram tests are nov ccnnideret* terrily unnecessary and envise for the following reasona:

1 1.

Errat'ic scran performance has been identified as due to ar obstructed drive filter in type "A" driver..

4 The driver. it.

BTRP are of the new "B" type design whose sertN perforr.cnt e to unaffected by filter condition.

2.

The dirt load is primarily released during startup of the reactor when the reactor and its systems are first l

to flaws and presoure and thermal stresses.

subjected Special atten-tion and sese. ores are now being taken to asnure cleaner i

systems.

Recetore with drives identical or stallar (shorter otroke, smaller piston stees) have operated through many i

refueling cycles with no sudden or erratic changes in scras serformance.

sufficient to detect anomalous drive perforwance.This preoperat d

i 3.

he 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> outage limit which inittsted the start of the frequent scrim testing is arbitrary, havinn no logical basis 4

bly be caused by an eventother than quantif ying a " major outage" wh j

i drive performance.

so severe as to possibly affect This requirement is unwise because it "on line" to avoid the additiont.1 testing due a 72-hou l

1 age.

i 132 i

3.3/4.3 AA_SE1:

for scran testing of all the The surveillance requirement control roJo after each refueling outane and 10% of'the, control 16-week intervals is adequate for detersiniog the opera-rods at is not so frequent as to bility of the control rod system yet cause exensive wear on the control rod system components.

The numerical values assigned to the predicted scram perfor-mance are based on the analysis of data from other BWR's with control red drives the same as those on Browns Ferry Nuclear Finnt.

The occurrence of scran times within the Ituits, but signifi-cantly lon.er than the average, should be viewed as an indica-tion of systematic problem with control rod drives especially if the number of drives exhibiting such scram times emeseds eight, the allowable number of inoperable rods.

390 milliseconds In the analytical creatment of the transients, are allowed between a neucron sensor reaching the scram point This is ade-and the start of negative reactivity insertion.

quote and conservative when compared to the typically observed time delay of abouc 270 milliseconds. Approximately 70 milli-seconda after neutroo flux reaches the trip point, the pilot scram valve colenoid power supply voltage goes to zero en approximately 200 milliseconds later, control rod action begins.

The 200 stiliseconds 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.1, 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.I.

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.1 will s11ow the subsequent withdrawal of any rod scrammed in the 100 percent to 50 percent rod density groups; however, it wfil maintain group notch control over all rods in the 50 percent density to preset power level range.

In addition, RSCG will prevent movement of rods in the 50 percent density to preset power level range until the scrammed rod has been withdrawn.

133 Amendment 35 i

Revised 1-17-79 l'

3.3/4.4 RASFS:

D.

Reactivity,, Anomalies During each fuel eyele excess operative reactivity as any burnable poison varten as fuel depletes a n,d j

in supplementary control is burned.

The magnitude of this exceng reactivity may be inferred from the 1

erttical rod configuration.

As fuel burnun pro-gresses, anomalous behavior in the eacess reactivity may be detected by comparison of the critical rod pattern at selected base states to the predicted 5

rod inventory at that state.

Power operating base j

conditlans provide the mest sensitive and directly

{

interpretable data relative to core reactivity,

{

Furthermore, using power operating base conditions i

1-permits frequent reactivity comparisons.

f Requirink a reactivity comparison at the specified frequency as<ures that a comparison will be made before the core reactivity change exceeds 1% o R a

Deviations in ento reactivity greater than I Li k a r e l

not expected and require thorough evaluation.

One i

percent reactivity into the core would not lead to i

transients exceeding design conditions of the reactor system.

1 References 1

1.

Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.

a 5

i i

f 134 Amendment No.

M.,47, 59 U

4 6

d' e o - --

..,.s

++

StJ'WE LLAtiCE PE0ln E tE:E httCTUfo CO'lDIT!O:fG F09 OPTPATIO'!

3.5.8 Realdunt Heat Removal Svetem 4.5.8 Residual Heat Recovsl Svar--

(RHRS) (LPCI and Containcent 1RHRS) (LPCI and Contain=ent cooling)

Cooling) 4 1.

The RitRS shall be operabic:

1.

a.

S1mulated once/

Automatic 03 rating (1) prior to a reactor Actuation Cycle startup from a Cold Test Condition; or (2) when there is irra-b.

Punp Cpara-Cace/

diated fuel in the bility

wnth reactor vessel and when the reactor vessel pres-c.

Motor Opera-Once/

ted valve mnch oure is greater than operability atmospheric, except sa apecified in specifica-tions 3.5.B.2, through d.

Pu:sp 71ow Race Once/3

con cha j

3. 5. 8. 7 and 3.9. 5. 3.

e.

Test Check Valve Onen/

2.

With the reactor vessel pres-Operati:n; sure lesa than 105 pair.. the C yc le AHRS may he removed trom ser-v

vice (except that two RHR pumps-containnent coolinr, mode and Each IJcI pe,p :he n. p inqt y

)

asacciated heat c:tchangers must 3,C00 gpeg ag e i n.,

remain operabic) for a period system prec:ura or 2 5 ra i. Two i

not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while LP I pu::ps in the :'W IM p r.h a l

deliver 13,C00 gp,c,aga;,,,:,. ;,

being drained of suppression indicated sys tem pruaure -

charter quality vater and 200 psig.

. filled with primary coolant I

quality water provided that 2.

An air test on the dryvell t.na during cooldovn two locps with torus heade o and no::len shall one pump per loop or one loop vith be conducted once/5 years.

e, two pumps, and associated diesel water test may be perfarned on j

generators, in the core spray syste:

the torus header in lieu of t%.

J are operable.

air test.

)

3.

Wh n it is determined that one RHR l

per h the a

r pu p (LPCI mode) is inoperabic at a may rrmain in operation for a time when operability is require..I, period not to exceed 7 days the remaining RHR pump,4 (LPCI mode) provided the rem ining RHR and active components in both access pumpa (LPCI mode) and both paths of the RHRS (LPCI mode).md accesa paths of the RHRS the CSS and the diesel r,enerators (LPCI mode) and the CSS and shall be demonst rated to be opera-the diesel generators r e:sain ble immediately and daily operable.

thereafter, 145 Amendment No. 59 s

4 M

PAGE DELETED n

d 146

'ilTING CumitTIOtn FtH< rii* Lit.A l l ON JU RV L lLL A;iC L 1(Li}il ! R L.'ilE T S J

'>.B Rdeldual itent R r* mo n i S y g '. e m

4. 5.B Residual liast Removal Systeq (RHRS) (LPCI and Containment (RKRS) (LPCI and Containment Cooiing)

CoolinA) 4 If any 2 RHR pucps (LPCI mode),

4.

No additional surveillance become inoperable, the reactor i required.

shall be placed in the cold l

Shutdown condition within 24 hodr3.

5.

tr one RHR pura (con:ain-

5. vnen it in deterniaed that one ment cooiinr. nide) o' ao-Rh*R pump (containment coolint nocinted heat exchanger is mode) or t.seocistad heat inoperable, the reactor exchanger is inoperabic at o may reruin in operation f or time when operability to re-a period not t>. exceed 30 quired, the reruining RitR days provided the ret:sininA pu=ps (containment cooling mode),

Rhp, pumps (containeent the associated heae exchangers cooling mode) and asso-and diesel generstors, and all ciated heet exchangers snd active components in the cecess diesel generators and all paths of the R11RS (concainm-nt access paths of the RHRS cooling node) shall be de-on-(containment cooling node) strated to bet operable irredia tely are operable.

an,i weekly thereafter until the inopersble RHR pump (contatnment cooling mode) and associa:ed hest exchanger is raturned to norral service.

6.

If two RHR pumps (containment

6. '4h e n i t is determined that wo coolin,t mode) or associatad RHA pumps (containment cooling hest exchangers are inopers-moje) or associated best exchangers ble, the reactor may remain are inoperable at J time when in operst ion f..r a pe riod coers'allity is required,

f. h e not to exceed 7 days pro-re=aining RMR pu:npa (containment vided the reu tning EHR pumps cooling teode), the associsted (containment cooling mode) heat exchangers, and diesel and associated heat exchan5ers annerators, and all active con-and all access paths of the ponents in the access pacha of RHR5 (c6ntainment cooling mode) ths RHR5 (contal:rment cooling 147 Amendment No. 59 i

L_..-,

k Li TINC CONDITIONS FOR OPERATION SURVEILLANCE REQUIP.EMENTS 3.5. 8 Residual llen t Removal System 4.5.5 Residual Heat Removal Svsten (RHRS)_ (LPCI and Containment (RKRS) (LPCI and Containment Cooling)

Cooling) f mode) shall be derenstrated are operable, to be operable immediately and daily thereafter until at least threo RHR pumps (containment cooling mode) and associated heat exchangers are returned to normal service.

7.

if two accean paths of the 7.

When it in determined that one RHRS (containment cooling or more access paths of the 3

mode) for each phase of the RHRS (containment coolint Tode) mode (drywell sprays, sup-are inoperable when accen9 to pression chamber sprays, required, all active componenta and suppression pool cooling) in the access paths of the RRR3 are not operable, the unit (containment ecoling mode) shs11 may remain in operation for a bo demonstrated to be operabia period not to exceed 7 days irmediately and all active ecm-provided at least one path ponento in the necess paths or each phase of the made which are'not backed by a eccend remains operablo.

operable access path for tha same phase of the mode (dryvell sprays, suppression chamber cprays i

and suppresoton pool cooling) shall be demonotrated to be cpera-ble daily thereafter untti the second path is returned to ner-mal service.

4 8.

If specifications 3.5.3.1 8.

No additional surveillance 4

through 3.5.B.7 are not met, required.

en orderly shutdown shall be initiated and the reactor shall be shutdown and placed in the cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

9.

When the rtactor vessel pres-9.

When the reactor vescal pressure eure.is atmospheric and irra-is stmospheric, the 73% pumra i

disted fuel is in the reactor and valves that art requirsd to vessel at least one RHR loop be operable shall be demonstrated with two pumps or two loops to be operable monthly.

with eno pump per loop shall 8

be operable. The pumps asso-cisted diesel generators must also be operable.

10.

No additional surveillance 10.

If the conditions of specifica-required.

tion 3.5.A.5 are met, LPCI and containment cooling are not 14g required.

Amendment No.

59 t.

I l

{

t i

I l

l 1

l l

LIMITENG CoffDIT10NS FOR UPERAT10tl SURVEILLA.$F. PEOUtdMi.y.

3.3.3 Residual Heat Removal Syster.

4.5.B Ra.tdual neat R.

<i '

s (M{RS ) (LPCI and Containment (MWJ ) (LPCI and Gi. t. ; c u t' Cooling)

Contingi

11. Wen there i t, trradiated fuel

11. The 'M punt"

.'si in the reactos and the reactor e

cent u n i t t. wht h le vessel pressure is greater titan I cross-conn, -

ic l '.

atmospheric, 2 RRR ptmps and l

snail be u m.

ti it !

aqsociated heat exchangers and I

operable = o.1 u

a-valves on an adjacent unit =ust c r os s-c oni.e c t. 9 ;. 5.. t y be operable and c apable of is required.

supplying cross-:ennect capabil-ity except as specified in speci i fication 3.5.B.lZ below.

(Notes Because cross-connect capability is not a short t e rm requirement, a component is not considered inoperable if cross-connect capability can be re-stored to service within

• hours.)

12.

If one RHR pump or associ-

12. When it is de te r:1 red ated heat exchanger located that one RhR puep or on the unit cross-connection associated heat exc har.ge r in the adjacent unit is Ln-located on the unit cross-operable for any reason (in-connection in the adja-ciuding valve innperability, cent unit la inopera'le o

pipe break, etc. ),

the at a tL=e when operabil-reactor may renaln in opera-icy is required, rte tion for a period not to remaining RrG cump and exceed 30 days provided the associated heat exenanger remaining RRA pump and on the unit cross-con-associated diesel generatot nection and the asacci-are operable, ated diesel generator shall be de.monstre. ed :n be operabic inmediately and every 15 davs charcat :e:

uatil the inoper aSle pur:-

a.1d associated heat ex-ctanger are returned to normal service.

f 1

149 1

Amendment No.

59 i

1 l

l

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRE ENTE.

13.

If RHR cross-connection flow or 13.

No additional surveillance heat removal capability is 3ost.

required.

the unit may remain in operation for a period not to exceed 10 days unless such capability is res tored.

14. All recirculation pump discharge valves shall be tested for operability 14.

All recirculation pump during any period of reactor cold shutdown discharge valves shall be operable prior to exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if reactor startup (or operability tests have closed if permitted not been performed elsewhere in these during the preceding speci fications).

31 days, i

150 Amendment No. 59

=

StJRVEII.734dE 2Qt!!RFMNTS LIMITIND CONDITIONS FOR OPERATION 3.5.F Reactor Core Isolation Cooling 4.5.F Reactor Core Isolation Cooling 2.

When it is determined that the 2.

If the RCICS is inoperable.

RCICS is inoperable, the llPCIS the reactor may remain in shall be demonstrated to be operat ten for a period not operable icnediately.

to exceed 7 days if the

}{PCIS is operabic during such time.

3.

If specifications 3.5.F.1 or 3.3.F.2 are not met, an orderly shutdown shall be initiated and the reactor shall bc depressurizerd to less than 122 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C.

Automatic Depressuritation C.

Automatic _Dqpressurization Syste.s (AUS)

System UWS)_

1.

Durir.g each operating cycle 1.

Four of the six valves of the following tests shall be the AutomnLic Deprc9 surf-performed on the ADS:

zation System shall be operabjc:

A si.ciulated aute:natic a.

actuation test shall N (1) prior to a startup performed prior to starton from a Cold Condition, after each refueling out-or, age. Manual surveillance of the relief valves is (2) whenever there is irra-covered, in 4.6.D.2.

diat.cd fuel in the reactor vessel and the teactor vessel pr m uro is greater than 105 puig, er. cept as specified in 3.5.C.2 and 2.

When it is determined that ruir 3.5.C.3 below.

than two of the ADS valves are 2.

If three of the.alx ADS valves incapable of automatic opera t ion,

the In'CIS shall be demonstrat ed are known to be incapabic of to be operable imediatcly and automatic operation the daily thr.ri.3fter as lonf, as reactor inay remain in opera-Specification 3.5.C.2 applies.

tion for a period not to execed 7 days. Providt<l the firC1 system la operable.

(Note that the pre anre relter fune. tion of these valvos is assured by section 3.6.!) of thc.<c specifications and that this specification only apptjen to the Ans functlen.) If more than thtec of f.ne six ADS valves nic kncun te be Incap-abic nf ::nt omit te opctation, an Irmellat c order ly nhutdown shall, be init iat i d. "I t h the 157 reac t.n r in a het shmdwn con-j dition in 6 hourr and in a cold shutdown condition in the following 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.

J a

Amendment No. 59 I

e.

y iv

<y,, -.

.w w

9 y

yy y."

• Y d

t 1.jMITINC rn stil,l l0N' roe SPIP.ATION SURVEILLANCr. atryinNNTS I

3.5.C Autn=atte Deareaauritation 4.5.C Automatic,Depressurizarten Syste* (AUS)

System (ADS) 1 1

4 1

M 4

t l

3.

If specifications 3.5.G.1 and 3. 5.C. 2 c onnot be met, an orderly shutdown vill be initiated and the reactor weasel pregourt shall be reduced to 105 peig or less withir 24 houre.

II.

M.itntenance of rtiled Discharge H.

!4aintenance of Filled Discharge Ilfj; Pipe f

Whenever the core spray systems, The following surveillance require-LPCl,llPCI, or RCFC.are required ments shall be adhered to to assure to be operable, the discharge that the discharge pipine, of the 4

piping from the pump discharge core spray systems, LPCI, HPCl, and of these systems to the last RCIC are filled:

4 j

block valve shall be filled.

i 4

i 4

l 4

a 158 I

t

+

4 S t,"dVEII.LANCT. 64.'.H ty Mf NTS g,1)qtTING COHt)tTIONS Fna OPF_R.ATinN 4.$.H Hsintenance of Filled Discharge Piu 3,s.H Maintenan:e or Tilled Otscharge Pipe _

s

"=e auction of the RCIC and KPCI pumps shall be aligned to the condensate ggg g e gg,t.

storage tank, and the pressure suppres-Spray) and core spray systeos, the slon chamber head, tank shall norma.11y discharge piping of these systens be aligned to serve the discharge pipinr' sbil h ted i m the W h p & t of the RKR and CS pu=ps.

tie condensate and water flow determined.

head tard may be used to serve the P.KR an 1 CS discharge piping if the PSC head 2.

Folloving any period where the LPCI tank is unavailable. The pressure or core spray syste=s have cet been indicators on the dische.rge of the RKR required to be operable. the dis-and CS pumps shall indicate not less charge piping of the inoperatis sys-e,=. shall be vented f rom the high thse listed below, P1-75 20 L8 psig point prior to the return of the F1-75 L6 L8 psig eyacem to service.

F1 51 1.8 psis PI-7L-65 L8 psi 6 3.

Whenever the NPCI or RCIC syste:s is lined up to take auction from the

1. Average Planar Linear Heat Generation condensate storage tank, the dis-charge piping of the HPCI and RCIC Pa c e.

shall be vented from the high point During steady state power operation, the of the system and water flow observed liaximum Average Planar Heat Generation Rate (MAPLHCR) f or each type of f uel as on a monthly basis, a function of average planar exposure shall not exceed the limiting value 4.

Wen the PJJR$ and the CSS ara 're-lWstanytimeduringoperationitshown an Tables 3. 5.I 2;-2,-3,-g-(-6,f'Z quired to be operabla, the pressure indicators which monitor the dis-is

. determined by normal surveillance that charge lines shall be sonitored the limiting value for APLHCR is being daily and the pressure recorded.

escoedad, action shall be initiated with-in 15 minutes to restore operation to

.etthis the prescribed limits. If the apt.neCR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Z.

Maximum Avers:re Planer t.ieear Heat Ce aa n-Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

tion Rate MAPLHCR)

Surveillance and corresponding action shall continue until reactor operation The MAPLHCR f or each type of fuel a.; a lu.ic-is within the prescribed limits.

tion of average planar asposure sna11 be determined daily during reactor operation

f. Linest Hast Generation Rate (1.HCR) at > 25% rated charsal power.

During steady state power operation, the linear haag generation rate (LHCR) of gg any red in any fuel assembly at any asial location shall not e.scoed the The WR as a f unction of core hemt shM,.

mastens allovable 1.HCR as calculated by be checked daily during reactor cperation at the following equation:

3, 251 rated thermal power.

Amendment No. 59

-LIMITInc CONDZTIONS FOR OPERATION _

l SURVEILLANCE REQUZRE}1ENTS l

LHCR 4LHGR

- (dP/P)

(L/LT)

Max max LUGR = Design LHCR = 18.5 kW/ft for 7x7 fuel d

=13.4 kW/ for 8x8 fuel 8x8R and P8x8R fuel (d P/P)

= Maximum power spiking penalty l

= 0.026 for 7x7 fuel I

= 0.022~for 8x8,8x8R and P8x8R fuel LT = Total core length = 12.0 ft for 7x7 fuel and 8x8 I

= 12.5 ft for 8x8, 8x8R & P8x8R L = Axial position above bottom of core 3

If at any time during operation it is deter-j mined by normal surveillance that the limitin g value for LHGR is being exceeded, action shal l j

be initiated within 15 minutes to restore i

operation to within the prescribed limits.

If the LHGR 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 />.

Surveillance and 3.

corresponding action shall continue until reactor operation is within the prescribed j

limits.

j K.

Md.nimum Critical Power Ratio (MCPR)

K.

Minimum Critical Power Ratio From BOC to EOC-2000 dw0/T (MCPR) e the MCPR operating limit for BFNP 1 cycle 4 is 1.23 for 7x7 fuel, 1.24 for 8x8 fuel, MCPR shall be determined daily and 1.25 for 8x8R and P8x8R fuel. These during reactor power operation at 1

l limits apply to steady state power operation 25 rated thermal power and j

at rated power and flow. For core flows following any change in power level other than raced the MCPR shall be greater or distribution that would cause than the above limits times Kg.

Kg is the operation with a limiting control value shown in Figure 3.5.2.

From EOC red pattern as described in the i

-2000 to EOC the MCPR limits will be 1.23, bases for Specification 3.3.

l 1.27, and '1.28 for 7x7, 8x8 and 8x8R/P8x8R respectively.

If at any time during oper-ation it is determined by normal surveill-ance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the steady MCPR is not returned to within the pre-scribed limits within two (2) hours, the reactor shall be brought to the Cold j'

Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, surveillance and correspooding action shall g

continue until reactor operation is within the prescribed limits.

a 3e g'

L.

Recorcing Reo_uirements If any of the limiting values identified in Specifications 3.5.1, J. or K are exceeded and the specified action is taken, the event shall be logged and a

reported in a 30-day written report.

• 12.5 feet for 8x8R fuel 160 3

5

l WEl. '

3.S.C Automatic bepressurisettee system (ads)

This specification nasures the operability of the AD$ vnder all candi-tiene for which the depressurfsation of the nuclear systes is an essen-tial Yespenae to station abnormalities.

The nuclear system pressure relief system provides automatic nuclear system depressurination for small breaks in the nuelaar systes se that the low-pressure coolant injection (1.pc!) and the core spray evbeystone cas operate to protect the fuel barrier. Note that this speettiaatten appites only to the automatic feature of the pressure relief system.

I specification 3.6.D specifies the requiremente for the pressure relief (wactico of the valves.

It is possible for any nwnber of the valves asoir.neJ to the Aos to be incapable of perform 1Nt their ads functione l

because of inetrumentatinn failures yet be folly casable of performing their pressure relief function.

Because the automatic depressurisation system does not provide makeup to the reactor primary vessel, no credit is taken for the steam cooling of the sore caused by the system actuation to provide further conservatiaa to the C A.

With two ADS valves known to be incapable of automatic operation, four valves remain operable to perfonn their ADS function. The ECCS loss-i of-coolant accident analyses for small line breaks assumed that four of the six ADS valves were operable.

Reactor operation with ' three ADS valves inoperable is allowed to continue for seven days provided that the HPCI system is demonstrated to be operable, Operation with

) more than three of the six ADS valves inoperable is not acceptable.

Amendment No.

47

167

3.S nAsks Pi e 1,$.H

'4aintenance of Filled Discharge f

it the discharr,e piping of the core spray, LPC t. HPC15. and RCICS are not filled, a water hammer can develop in this pipine, when the pump and/or pumps are started. To minimisc damate to the discharge piping and to ensure added martin in the operation of these systems, this Technical Specific 4 tion requires the discharr,e lines to ba filled whenever the system is in an operable condition. If a discharge pipe t ai not filled the pumps that supply that itne must be assumed to be inoperable for Technical Specification pur-poses.

The core spray and RHR system discharge piping high point vent is visually checkcJ for water flow once a month prior to testing to ensure that the lines are fliied. The visual checking will avoid starting the core spray or RHA system with a discharge line not filled. In addition to the visual observstion and to ensu.re a filled d.ischamrge line other than prior to testing.

a pressure suppression chamber head tank is located approzir.ately 20 feet above j

the discharge line highpoint to supply makeup water for these systems. The condensate head tank located approximately 100 feet above the discharge high point serves as a backup charging system when the pressure suppression chamber head tank is net in service. System discharge pressure indicators are used to determine the water level above the discharge line high point. The indicators v111refleet approximately 30 peig for a vster level at the high point and L5 pels for a water level in the pressurosuppression chamber head tank and are ten-itored daily to ensure that the discharge lines are filled.

When in their normal standhv condition. the suction for the !!PCi and RCIC pesmpt are allr.ne.1 to the condensa te storsp,e tank, which is physically at a

hir.her elevatinn than the HpC15 and RCICS pintnr..

This assures that the HPCi and RCIC discharme pipina, remains filled. Further assurance is provided by observint votar flow from thene systoms hir.h points monthly.

tianimum AvarsJs Planar 1.inear Heat Ceneration R.ata (MAPGCR) 3,3,1, This specification assures that the peak cladding temperature f o11ovin5 the postulated design basis less-of-coolant accident vill not exceed the limit specified in the 10CFA50, Appendix K.

The peak cladding temperature following a postulated less-ef-coolnnt acci-dent is primarily a function of the average hcot generation rate of all the rode of a fuel assembly at any exial location and is only dependent second-

[

erily on the rod to rod power distribution within an assembly. Since ex-pected local variations in power dictribution within a fuel assembly affect the calculated pesk clad temperature by less than 1 200Y relative to the peak temperature for a typical fuel design, the limit on the averar,e linea r j

heat generation rate is sufficient to assure that calculated temperatures are within the 10CTR50 Appendix K limit. The limiting value f or MAPUCR is shown in Tables 3.5.I-1,

-2,

-3,

-4, -5,

-6, and -7.per reference 4 P

Amendment No.

59 160

3,$,g, Linear Heat Ceneration Rate (fHCR)

This opecification essures that the linear heat generation rate in any rod fa less than the des!Gn linear hent generation if fuel pollet denefficatiun is postulated. The power spike penalty specified is based on the anal-yste prr nnted in Section 3.2.1 of Ref erence 1 as modified in References lim nely incre. ting varlJtion in sktal gaps be-2 and 3, and assuams e tween core bottom and top, and assares with a 95% confidence, that no nore than one fuel rod catceds the design 11ocar heat generation rate due to power spikinc. Tnc LHCM as a function of cure height shall be checked daily dur-ing reactor oper. Lion at 2 75% power to determine if fuct burnup, or cun-trol rod moveient has caused chsnces in power distribution. For LHCR to be a limiting value below 25% rated thermal power, the MTrr would have to be greater than 10 which is precluded by a considerable cargin when employing eny permissibic control _re4 pattern.,

3.5.K.

hte tm z c r i t ic al Peue r Ra t io (MCPR)

At core t he rmal p ewe r lev el s le e s than or equal to 25:, the reactor will be oppreting at ninLaum recirculation pump epeed and the moderator void content will be very emall.

For all designated control rod patterns which may be ee-played at this poLat, operating plant esperience and thermal hydraulic anal-yele Lnd ic a t e d t ha t the resulting MCPR value le in excess of requiremente by a considerable narsio, Vith this low void content, any inadvertent core flew increase would only piece operation in a more conservative code rete-tive to MCPR. The daily requirement for calculating MCPP, above 25% rated thermal po,er le sufficient einee pover distribution ehlf te are very slow when there have not been signific ant pcwer or control rod changes. The requirement for calculating MCPR when a lindt Lng control rod pattern is approached esaures that HCFR will be kaovo following a change in power or power shape (regardless of ma gni t ude) that could place operatien at a the rm al l ta t t.

3,$,t, Report tog Requirement e The LCO's aseociated with monitoring the fuel rod operating conditinns are required to be eat at all tLaes, i.e., there is na allowable time in which the plant cas knowingly exceed the limiting valua a f or MAPLHCT4, LR C R, c.n d MC7R.

It to a requirement, se st at ed in Specif ic at ions 3 5.I..J end.K.

t ha t if at any eime during steady stete peser operet ten, it la d ete rmined that the limJting value s f or H>f LHCR, LHCR, or MCPR are exceedd actior, is then initiated to restore operation to within the prescribed limits. Thie actico te initiated as soon as normal surveillance indicates that an eperating L Lr-it has been reached.

Fach event involving steaoy state operation beyond a specified Itait shall be logged and reported quarterly, it must ce racc Antand thzt there is alwaye an action which would return any of the parametere (MAPLRCR, LBCR, or MCTR) to withLn prescribed ILaits, namely power reduction. Under cost circunst ances, thte vill not be the only siternative.

H.

E*f*f'D<'8 1.

"ruel Decei!! cation Ff f ects on General Electric Boir.ng Wa:J7 XJJJcor Puel," Suppleeent s 6, 7 and 8, KrtM-107 35, Au g u s t 19?).

2.

Su ppl emen t 1 to Technical Report on Densif ications of Cenera*.

Electric Rasetor Tuelo Deersber 14, 1974 (US A Ragulat or y St a f f).

j 3.

Communica t ion :

V. A Moo r e t o I. 5. Mi t c he ll, Mod i f ie d C T. hoo e l f or Puel Densification," Doc' set 50-J21 M4rch 27, 1974.

l 4.

Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and addenda.

169 knendment No. 59 N

J 4.$

Core and Contatument Cooling Systems T,ur,vellja..c Frequencies The testica interval for the core and containment e sling systems is based on industry practice, quantitative reliability analysis, judgement and i

practicality. The core cooling systems have not been designed to be fully testable during operatic,.

For example, in the case of the HPCI, automatic initiation duttnr powet operation vould result in pumping cold water into the reactor vessel which is not desirable. Complete ADS testing during power opera tion causes an undesirable loss-of-coolant inventory. To increase the availability of the core and containment cooling system, the components which make up the system; 1.e., instrumentation, pumps valves, etc., are tested frequently. The pumps and motor operated injection valves are also teste6 each month to assure their operability.

A simula ted automatic ac tua-tion test once each cycle combined with monthly tests of the pumps and injee-tion valven is deemed to be adequate testing of these systems.

l When components and subsystems are out-of-service, overall core and contain-t ment coolinr. reliability is maintained by demonstrating the operability of the remaining equipment. The derree of operability to be demonstrated depends on the nature of the reason f or the out-o f-service equip +ent.

For routine out-of-service periods caused by preventative maintenance, etc., the pump and valve operability checke vill be performed to der.onstrate operability of the remaining compenents. However, if a failure, design deficiency, cause the outage, then the demonstration of operability should he thorough enough to assure that a generic problem does not exist.

For example, if an out-of-service period van caused by f ailure of a pump to deliver rated capacity due to a design deficiency, the other pumps of this type might be subjected to a flow rate test in addition to the operability checks.

Whenever a CSCS system or loop is made inoperable because of a required tent or calibration, the other CSCS systems or loops thst are required to be operable shall be considered operable if they are within the required survell-j lance testing frequency and there is no reason to suspect they are inoperable.

~

If the function, ayatem, or loop under test or calibration is found inoperable l

i or exceeds the trip level setting, the LCO and the required surveillance testing for the system or loop chall apply.

Kedundant operable components are subjected to increased testing during equip-ment out-of-aervice times. This adds further conservatism and increases assurance that adequate cooling is available should the r.eed arise.

l Maximus Average Planar LHCR, LHCR, and MCPR The MAPLHCR, LHCR, and MCPR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribucien.

Since changes

'due to burnup are slow, and only a f ew control rods are moved daily

.a daily check of power distribution is adequate.

~

170 i

l f

I 1

a Table 3.5.I-1.

MAPLHCR VERSUS' AVERAGE PLANAR EXPOSURE Fuel Type:

Initial Core - Type 1&3 Average Planar Exposure HAPLHGR (Hud/t)

(ktJ/ f e) 200 15.0 1,000 15.1 5,000 16.0 10,000 16.3 15,000 16.1 20,000 15.4 25,000 14.2 30,000 13.1 Table 3.5.I-2 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: Initial Core - Type 2 Average Planar Exposure MAPLHGR (Mwd /t)

(kt1/f t) 200 15.6 1,000 15.5 5,000 16.2 10,000 16.5 15,000 16.5 20,000_

15.8 25,000 14.5 30,000 13.3 171 l

Amendment No. 59 l-

l Table 3.5.I-3 MAPLHCR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: 80274L j

Average Planar Exposure MAPLHCR (Mwd /t)

(kW/ft) 200 11.2 1,000 11.3 11.9 5,000 12.1 10,000 12.2 15,000 12.1 20,000 I

11.6 25,000 10.9 30,000 Table 3. 5.I-4 HAPLHCR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: 80274H Average Planar l

Exposure MAPLHCR Otwd / t)

(kW/ft) 200 11.1 1,000 11.2 5,000 11.8 10,000 12.1 15,000 12.2 20,000 12.0 25,000 11.5 30,000 10.9 172 Amendment No.

59

TABLE 3.5.I-5 MAPLHCR VERSUS AVERAGE PLAllAR EXPOSURE Fuel Type: 80R265H Average Planar MAPLHOR Exposure (Mwd /t)

(kv/rt))

200 11.5 11.6 1000 5000 11.9 12.1.

10,000 12.1 15,000 11.9 20,000 11.3 25,000 10.7 30,',00 TABLE 3 5.I-6 MAPPHOR VERSUS AVERAGE PLAflAR EXPOSURE Fuel Type: 8DB265L Average Planar MAPLHOR Exposure (mwd /t)

(kv/ft) 11.6 200 11.6 1000 12.1 5000 12.1 10,000 12.1 15,000 11.9 20,000 11.3 25,000 10.7 30,000 172-a i

Amendment NO. 59

TABLE 3.5.I-7 2

HAPLilGR VERSUS AVERAGE PLANAR EXPOSURE Fuci Type: P8DR284L Average Planar Exposure MAPLliCR (Mwd /t)

(kw/ft) 200 11.2 1000 11.3 5000 11'.8 10000 12.'O 15000 12.0 20000 11.8 25000 11.2 30000 10.8 l

l 172-b Amendment No. 59 i

4

1. I,N,i T 8 W*. rmfli T,lfi,t. 6,8]k, 0,?,1 A.A T,1Jw,8__,,,,,,,,,,,, J,UJvjit_l.td,Nr r, p QtI 1,14pJixT,,_,,,,

3.6.C Coel ma,t 8.c e k e t.

4.6.C Coolant _tga h l

3.

If the condition in 1 or 2 iove eennot be met. an orderly j

shutdovn shall be int (1sted and the reactor shall be shut.

D.

pelief valves a

down in the Cold Condition within 74 hours8.564815e-4 days <br />0.0206 hours <br />1.223545e-4 weeks <br />2.8157e-5 months <br />.

j, Approrteately one-half of all D.

B,elicf valves relief valves shall be bench-i checked er replaced with a 1.

When more than one valve.

bench-checked valve each epera-is known to ting cycle. All 1} valves be failed, an ordery shut-vill have dovu sh.s11 he initiated and been checked or replaced upor, the reactor depressurised to the cooolecion of every second teos chara 10$ psig within 24 cycle.

bewts.

2.

Once during each operating l

cycle, each relief valve sha!!

be manually opened until therr.o-couples dovnstream of the valve indicate steam is flovir,g f roa the valve.

3.

The integrity of the relief valve bellows shall be continuously =enitored.

4.

At least one relief valve shall be disassembicJ and inspected each operstic.g cycle.

E.

Jet Pumps E.

Jet Pumps 1.

Whenever the resetor is in the 2.

Whenever there is recirculation startup or run modes. all jet flow with the reactor in the pumps shall be operable. If startup or run nodes with both it le determined that a jer r3 circulation pumps running, pwep la inopetable, or if two jet pwsp operability shall be er more jet pump flow instru-checked daily by verifyLag that ment failures occur and can-the following condi:1ons do not not be correctcJ within 12 occur simultaneously:

hours, se orderly shutdo.m eka!! be int:tsted and the a.

The two recirculation loops resetor shall be shutdown in have a flov imbalar.ce of the Cold Condition within 24 15% or more when the pumps bevre.

are operated at the some speed.

181 i

l i

l 1

l Amendment No. 47

OPEAAT!c9 Su mv r.1 LLA ut t a t qu t RLMt 6 Lly_IT,tyfr.,[mhi f M FOR 4.6.E Je t Punpe J. 6 l.

'J e t Pu'Pa g,, 'The indicated value et core 3 6.7 Jet P.mp now Missatch flow rate varie s f rom the walue derived f rom loop flov measurements by more than lo!.

The dif f user to leker plenum c.

dif f erential pre ssure read-3

$ng en an individual jet puarp varies f rom the mean of all jet pu:rp dif f eren-tial pressures by more than 102.

2.

Whenever there is recirculation II'" "ith *h* ***CE** 1" Th' 1.

The reactor shall not be Startup or F.6m Mode and one re-operated with one recirculation drealati n pump d's operating

~

loop out of service for more erith the equalize r valve closed, than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With the reactor

'the dif f user to lover plen'un operating, if one recirculation 41f f erential pressure shall be loop is out of service, the checked daily and the differen-plant shall be placed in a hot tial pressure of an individual shutdown condition within jet ptrup in a loop shallinot 4

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the loop is vary f rom the mean of 'all jet sooner returned to service, pump dif f erential pressures in that loop by more than ICI.

'2, Following one pu:.? opa. ration, 4

the discharge valve of the lowg.

i F.

Jet Pump Flow Mie,s.steh speed ptcp may not be opened j

unless the speed of the faster Recirculstion pump speeds shall 1.

yump is less than Af, of its be checked and logged at least C

rated speed.

once 1per day, f

3.

Steady state operation with both recirculation pumps out of ser-Vice for up to 12 hrs is per-niitted. During such interval J

restart of the recirculation pumps is permitted, provided the loop discharge temperature is Within 75 F of the saturation 0

temperature of the reactor vessel water as detemined by dome pressure.

C.

Struccural Intecrity 1.

Table 4.6.A together vith sup-C.

structural Inteerity picmentar7 notes, specifies the 1.

The structural integrity of the primary systen shall be 1 82 Amendment No.

59 i

3.6/6.6

. BASES:

The basis for the equilibrium coolant iodine activity limit is a computed dose to the thyroid of 36 rem at the exclusion distance dur,ing the 2-hour period following a steam line break. This dose is computed with the conservative assumption of a relcase oi 140,000 lbs of coolant prior to closure of the isolation valves, and a X/Q value of 3.4 x 10~' Sec/m3 The maximum activity limit during a short term transient is established from consideration of a maximum iodine inhalation dose less than 300 rem.

The probability of a stean,line break accident coincident with an iodine concentration transient is significantly lower than that of the accident alone, since operation of the reactor with iodine levels above the equilibrium value is limited to 5 percent of total operation.

The san-pling f requenc ies arc established in order to detect the occurrence of an iodine transient which 0.ay exceed the equilibrtum concentration limit, and to assure that the maximum coolant iodine concentrations are not exceeded. Additional sampling is r equired following power changes and off-gas transients, since present data indicate that the iodine peaking phenomenon is related to these e"ents.

3. 6. C / 4. fi. C Coolant 1.eakare A11ovahic leekate rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipco and on the ability to makeup coolant system Icakage in the event of loas of offsite a-c power. The normally expected background leakage due to equipment design and the detection capability for determining coolant sys-tem leakage were 91oo considered in ectab11shing the limits. The behavior of cracks in pipinr, nyatema has been experitsentally and analytically inves-tigated 49 part of the USAEC sponsored Reactor Primary Coolant System Rupturc Study (t he Pipe Rupture Study). Vork utilizing the data obtained in this study indicates that leakage from a crack can be detected before the crack grows to n danr,erous or critical site by meenanically or thermally induced cyclic leading, or stress corrosion cracking or some other mechanism charcettrir.ed by gradual crack growth. This evidence sugr,ests that for leak-age somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or cracks associated with such leak-age would grow rapidly, llovever, the establichment of allowable unidentified leakage greater than that given in 3.6.C on the basis of the data presently availabic vould he premature because of uncertaintics associated with the data.

For leakage of the order of 5 spn, as specified in 3.6.C. the experi-mental and analytical data sugrent a reasonable margin of safety that such leakane magnitude vould not result from a crack approaching the critical site for rapid propagation. Leakage less than the magnitude specified can be 218

i 3.6/h.6 TACE 3 detected rensenably in a catter of feu hours utilizing the available leaksce detection sche:cs, and if the crigin cennot be determined in a reesenably short time the unit sheuld be shut down to allev further investigation and corrective action, i

The total leakege rate censists of all leaksca, ident tried and unidenti-fled, which flows to the dryvell flocr drain and equipeent drain sumps.

The capacity of the dryuell fleer surp pucp is 50 spm and.the capacity cf the dryvell ec.uip,ent su:p puer is el.se 50 rpn.

Re.90 val of 25 sp:

from either of these cueps can be accc:plished with censiderable narsin.

RETE E CES 1.

Nuclear System Leakage Rate Limits (BTU? pSAR Subsectien k.10) i 3.6.D/4.6.D Relief Valves To meet the safety basis thirteen relief valves have been installed on the enic with a total capacity of 92.6%of q

nuclear boiler rated steam flow. The analysis of the j

worst overpressure transient, (3-second closure of all main i

steam line isolation valves) neglecting the direct scram

&alve position scram) results in a maximum vessel pressure 4.

which, if a neutron flux scram is assumed considering 12-3 valves inoperable, results in adequate margin to the code allowable overpressure limit of 1375 psig.

To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to j

open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel over-pressure of 1375 psig.

a 1

219

' Amendment No. 59

Revised 1-17-79 3.6/4.6 B AS ES :

Experience in relief valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations. The relief valves are benchtested every second operating cycle to ensure that their set points are vithin the f 1 percent tolerance. The relief valves are tested in place once per operating cycle to establish that they vill open and pass steam.

The requirements established above apply when the nuclear system cac be pressurized above ambient conditions. These requiraments are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventus1 overpressure relief would be needed. However, these transients are much less severe, in terms of pressure, than those starting at rated conditions. The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurised.

RITERENCES

'J 1.

Nuclear System Pressure Relief Systes (BTNP TSAR Subsection 4.4) s Amendment 22 in response to AIC Question 4.2 of December 6, 1971.

3.

" Protection Against overpressure" (ASME Boiler and Fressure Vessel Cods,Section III, Article 9) 4.

Browns Terry Nuclear Plant Design Deficiency Report--Target Rock Safety-Relief Valves, transmitted by J. E. C111 eland to T. E. Kruest, August 29, 1973.

5.

Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.

3.6.E/4.6.E Jee Pumps Teilure of a jet pusep nossle assembly holddown mechsoism, nogale assembly and/or riser, would increase the cross-sectional flow area for blowtown following the design basis double-ended line break. Also, failure cf the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break. Therefore, if a failure occurred, repairs must be made.

The detection technique is as followe. With the two recirculation Fumps balanced in speed to within

$ percent, the flow races in both recircula-tion loops will be verified by control room monitoring instruments. If the evo flow rate values do not dif f er by more than 10 percent, riser and nozzle assembly integrity has been verified.

220

/

knendment No.,47, 59

s.6/4.6

!LAsrg:

If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) dif fuser measurements

)

will be taken to define the location within the vessel of failed jet puno nottle (or rlser) and the unit shut down for repairs. If the potential blowdovn flow area is increased, the system resistance to the recirculation pump is also reduced; hence, the af fected drtve pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a nine.le nossle' failure). If th-two loops are balanced in flew at the same pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leakar,e path past the core thus reducing the core flow rate. The reverse flow thtour.h the inactive jet pump would still be indicated by a positive differential pressure but the not effect would be a slight decrease (3 per-cent to 6 percent) in the total core flow measured. This decrease, to6 ether with the loop flow increase, would result in a lack of correlation between meanured and derived core flow rate.

Finally, the affected jet pump diffuser dif f erential pressure signal would be reduced because the backflow would be less then the normal forward flow.

A nnrzte-riger system failure could also Renerate the coincident failure of a )ct pump diffuser body; however, the converse is not true.

The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.

3.6.F/4.6.F Jet Pump Flow MiamAcch 221 Amendment No.

59

l 1

NOTES FOR TABLE 3.7 A i

Keys O = Open

)

C = Closed i

SC - Stay's Closed GC a coes Closed l

Note: Isolation groupings are as follows:

The valveo in Group 1 are actuated by any one of the folleving Croup 1:

conditions:

1.

Reactor Vessel Low Water Level (470")

2.

Hain Steam 11ne High Radiation 3.

Kain Steamline High Flow 4.

Main Steaaline Space High Temperature 5.

Hsin Sa.eaaline Low Pressure Group 2: The valves in Group 2 are actuated by any of the following conditione:

1.

Reactor Vessel Low Vater Level (538")

3 2.

High Drywell Prsesure Group 3': The valves in Group 3 are actuated by any of the following conditions:

1.

Reactor Low Water Level (538")

2.

Reactor Vater Cleanup Systes High Temperature 3.

Reactor Water Cleanup System High Drain Temperatura Group 4: The valves in Group 4 are actuated by any of the following conditions:

1.

,HPC1 Steamline Space High Temperature 2.

RPCI Steamline High Flow 3.

HPCI Steamline Low Pressure Group 5: The valves in Group 5 are actuated by any of the following condition 1.

RCIC Steamline Space High Temperature 2.

RCIC Steamline High Flow 3.

RCIC Steseline Low Pressure Group 6: The valves in Group 6 are actuated by any of the following condit ions :

1.

Reactor Vessel Low Water Level (538")-

2.

High Drywell Pressure 3.

Reactor Building Ventilation High Radiation 254 Amendment No. 59

Group 7: The valves in Group 7 are automatically actuated by only the following condition:

1.

Reactor vessel lov water level (470")

Group 8: The velves in Group 8 are automatically actuated by only the following condition:

2.

High Dryvell pressure 1

f l

J 255 Amendment No.

59

BASES in the rn ged ohipbu rd environnent on the..J Savannah lopjl 3726).

Pres-sure alrnp across the conhined llEPA fil ter n and charcon1 adsorbers of less than 6 inches of wate r at "the sys tem den t rn flow rate will indicate that the filters ant 1 berhe m nrn not clnered by exec <.t.tvc amounts of foreign natter.

Hea t e r c apab il i t y, pr m.ura drop and a i r d t :.t r i bo t ton nhould be de t e rmined at least once per operating cycle to show system pe r f onnance capah tlit y.

Th e frequency of tests and sample analysis are necessary to show that the i

I HEPA filters and charcoal adsorbers can perf orm as evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon ref rigerant shall be per-forced in accordance with US ATC r.eport DP-1982.

Iodine re oval ef ficiency tests shall follow EDT Standard M-16-lT The charcoal adsorber efficiency I

test procedures should allow for the removal of_oite adiod ar_t m emptying of one bed f rom the tray, mixing the adsorbent thoroughly and obtaining at least two samples. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed.

If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified according to Table 1 of Regulatory Guide 1.52.

The replacement tray for the edsorber tray removed for the test should neet the sare adsorbent quality. Tests of the HEPA filters with DOP aerosol shall be performed in accordance to ANSI N510-1975.

Any HEPA filters found defective shall be replaced with filters qualified pursuant to Ragulatory Position C 3.d of #cgulatory Guide 1.52.

All elements of the heater should be demonstrated to be functional and operable during the test of heater capacity.

Operati5n of each filter train for a minimum of 10 hrs each month will prevent moisture buildup in the filters and adsorber systen.

Vith doors closed and f an in operation, DOP aerosol shall be sprayed externally nlong the full linear periphery of each respective door to check the gasket seal. Any detection of DOP in the fan exhaust shall be considered an I

unacceptable test result and the gaskets repairs and test repeated.

If significant painting, fire or chemical release occurs such that the llEPA filter or charcoal adsorber could become contaminated from the funes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for operational use.

The determination of significant shall be made by the operator on duty at the time of the incident. Knowledgeable staff members should be consulted prior to making this determination.

Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability. If one standby gas treatment system is inoperable, the ot her s'ystems must be tested daily. This substantiates the availability of the operable systers and thus reactor operation and refueling operation can continue for a limited period of tire.

3.7.D/4.7.D P r ima ry Containment Isolation Valves Double isolation valves arc provided on lincs penetrating the prir.ary con-tainment end open to the free space of the containment. Clocure of one of the valveo in each line vould be sufficient to maintain the integrity of the pressure suppression systen. Automatic initiation 10 required to mini-mise the potential leaksgo patne fecm tha contain=ent in ths event of a loss of coolant accident.

276

I,

't 4

i a

DASES Croup _1 - procesa lines are isolated by reactor vessel low veter level (490") in order to allow for removal of decay heat subsequent to a isolate in time for proper operation of the core standby scrav, yet coclina systemo. The valves in group 1 are also closed when proceso inotrumentation detects excencive main steam line flow, high radiation, low proonure, or main steam space high temperature.

2 - isolation valves are closed by reactor veosel low water level Group

($3S") or high dryveL1 proosure. The g(oup 2 isolation signal also " iso-latco" the reactor building and starts the standby gas treatment system.

It is not desirable to actuate the group 2 isolation signal by a tran-sient or spurious oignal.

Group 3 - process lines are normally in use and it is therefore not desirable to cause spurioun isolation due to high dryvell pressure resulting from non-safety related causes. To protect the reactor from a possible pipe break in the systeia, isolation is provided by high temperature in the I

cleanup nystem area or high flow throu$h the inlet to the cleanup system.

Also, since the vesaci could potentially be drained through the cleanup system, a low level isolation is provided.

C r o,uj_4 and 5 - process lines are designed to remain operable and mitigate the conocquencen of an accident which resulto in the isolation of other process lines. The oignals which initiate isolation of Croup 4 and 5 proccos lines are theref ore indicative of a condition which would render them inoperable.

6 - lines are connected to the primary containment but not directly Croup to the reactor vessel. These valves are isolated en reactor low water level (538"), high dryvell pressure, or reactor building ventilation high rodiecion which would indicate a possible accident and necessitato 1

ptimary containment isolation.

a Croup 7 - process lines are closed only on reactor low water icvel (470'$.

These close on the came signal that initiates HPCIS and RCICS to ensure that the valves are not open when HPCIS or RCICS action is required.

Croup 8 - ILne (traveling in-core probe) is isolated on high dryvell pres-sUre. This is to assure that this line does not provide a leakage path when, containment pressure indicates a possible accident condition.

The maximum closure time for the automatic isolation valves of the prirary containment and reactor veasel isolation control system have been selected in consideration of the dr.ign intent to prevent core uncovering followine, pipe breaks outsiac the primary containment and the need to contain released fisolon products following pipe breaks inside the prieary containment.

In satisfying this deoign intent an additional margin has been included ir.

specif y ing maximum closure times. This margin permits identification of degraded valve performance, prier to exceeding the design closure times, 277 i

l l.<

l Amendment No. 59

5.0.

HA.10R nr. SIC:i y ATURES 5.1

$1TL JLAttikES Browns Ferry unit J is located at Browns Ferry Nuclear Plant site on property ovned by the United States and in custody of the TVA.

The site shall consist of approximately 840 acres on the north shore of Llheelcr Lake at Tennessee River Mile 294 in Limestone County, Alabama. The minimum distance from the outside of the secondary containment building to the boundary of the exclusion area as defined in 10 CFR 100.3 shall be 4,000 feet.

5.2 REACTOR A.

The reactor core may contain 764 fuel assemblies censisting of 7x7 assemblies having 49 fuel rods each, 8xS assemblies

)

having 63 fuel rods each, and 8x8R (and P8x8R) assem'olics having 62 fuel rods each.

j B.

The reactor core shall contain 185 cruciform-shaped control rods. The control material shall be boron carbide powder (B C) compacted to approximately 70 percent of thearctical 4

density.

5.3 REACTOR VESSEL The reactor vessel shall be as described in Table 4.2-2 of the FSAR. The applicable design codes shall be as described in Table 4.2-1 of the FSAR.

$.4 CONTAINMENT A.

The principal design parameters for the primary containment shall be 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 an described in Section 5.3 of the FSAR, C.

Penetr.1tions to the primary containment and piping passing through such penetrations shall be designed in accordance with the standards set forth in Section 5.2.3.4 of the FSAR.

5.5 rUEL STORA,CE A.

The arrangement of fuel in the new-fuel storage facility shall be such that k for dry conditions, is less than 0.90andfloodedisless, than 0.95 (Section 10.2 of FSAR).

gg 330 knendment NO. 59 I

~. --

. - -. _ ~

50 MAJOR DESIGN TEATURES (Continued)

It.

The h r,r the opent rm:1 ott,rege Iml uhell be lene than k,, equal to 0.95 Fuel stored in the pool shall not contain more than 15.2 grams of uranium-235 per axial centimeter of fuel assembly.

Loads greater than 1000 pounds shall not be carried over spent.

C.

f te.cl assemblies stored in the spent fuel pool.

5.6 SEISMIC DESIGN The station class I structures and systems have been designed to withstand a design basis earthquake with ground acceleration The operational basis earthquake used in the plant of 0,2g.

design assu=ed a ground acceleration of 0.lg (see Section 2.5 of

'the FSAR).

i.

I 331 Amendment No. 42 I