Text

Proposed Tech Specs 1.1,1.2,2.1,2.2,3.3,4.3,3.5,4.5,3.6,4.6, 3.7,4.7,25.2 & Tables 3.2.F,3.5.I,3.5.K & 3.7.A to Accomodate Reload 4 Cycle 5 Operation at Facility. Explanation of Changes Encl
	ML20126J607
Person / Time
Site:	Browns Ferry
Issue date:	04/29/1981
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML20126J605	List: ML20126J600; ML20126J607; ML20126J611; ML20126J615;
References
	NUDOCS 8105010362
	Download: ML20126J607 (40)
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ENCLOSURE 1 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS BROWNS PERRY NUCLEAR PLANT UNIT 1 o

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8105010 % %

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D Section Page No.

C.

Coolant Leakage 180 D.

Safety and Relief Valves............

181 181 E.

Jet Pumps F.

Recirculation Pump Operation 182 G.

Structural Integrity.....

182 H.

Shock Suppressors (Snubbers)..........

185 227 3.7/4.7 Containment Systems A.

Primary Containiiient 227 8.

Standby Gas Treatment System.

236 i

C.

Secondary Containment 240 D.

Primary Containment isolation Valves......

242 E.

Control Room Emergency Ventilation.......

244 F.

Primary Containment Purge System........

246 G.

Containment Atmosphere Dilution System (CAD)..

248

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H.

Containment Atmosphere Monitoring (CAM) System H2 and 02 Analyzer..

249 3.8/4.8 Radioactive Materials 281 A.

Liqui d E f fl uents................

281 B.

Airborne Effluents...............

282 C.

Mechanical Vacuum Pump..

286 D.

Miscellaneous Radioactive Materials Sources 286 3.9/4.9 Auxiliary Electrical System 292 A.

Auxiliary Electrical Equipment..

292 i

B.

Operation with Inoperable Equipment 295 i

C.

Of>cration in Cold shutdown..

298 3.10/4.10 Core Alterations 302 f

A.

Refueling Interlocks..............

302 i

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LIST OF TA6LES (Cont'd) i l

Tab 1'e Title Page No.

g 4.2.F '

Minimum Test and Calibration Frequency for Surveillance Instrumentation 105' 4.2.G' Surveillance Requirements for Control 106 Room isolation Instrumentation.,........

4.2.H Minimuca Test and Calibration Frequency for Flood Proter. tion Instrianentation 107' Seismic Honitoring Instrument Surveillance 108 l4.2.J.

171,172,172-a 3 5.I KAPLHGR vs Average. Planar Exposure 3.6.H

. Shock Suppressors (Snubbers) 190 4.6.A Reactor Coolant System Inservice Inspection 209 Schedule 3.7.A Primary Containment Isolation Valves 250 3.7.B Testable Penetrations with Double 0 Ring seals......................

256 8

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

257

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3.7.0 Primary Containment Testable Isolation Valves...

250 3.7.E Suppression Chamber influent Lines Stop. Check (N

Globe Valve Leakage Rates............

263 3.7.F Check Valves on Suppression Chamber influent Lines 263 3.7.H Testable Electrical Penetrattoris 265 4.8.A Radioactive Liquid Waste SamplinD and Analysis 207 4.8.8 Radioactive Gaseous Waste Samplin'g and Analysis..

288 3.11.A,

Fire Protection System Hydraulic Requirernents...

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

360 l

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LIST OF ILLUSTRATIONS Figure Title _

PaerI!q.

l 2.1.1 APRM flow Reference Scram and APDi Rod Block 1

13

-Settings 2.1-2 APRM Flow Blas Scram Ys. Reactor Core flow 26 l

p 4.1-1 Craphic Aid in the Selection of an Adequate Interval Between Tests 49 4.2-1 System Unavailability...............

119 l

'3.4 1 Sedium Pentaborete Solution Volpe Concentration 138 l

Requirements 3.4 2 Sodium Pentaborate Solution Temperature i

Requirements 139 l 3.5.K-1 MCPR Limits 172b 3.5.2 Kf T ac to r..................... 1T3 I

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3.6-1 Hinimum Temperature 'F Above Change in Transient i

Temperature...................

188 l

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

i Neutron Exposure 109 6.1-1 TYA Office of Power Organtration for Operation of Huclear Power Plants.............

361 e

t 6.1-2 Functional Organization.........

362 l

r 6.2-1 Review'and Audit Function.............

363 f

6.3-1 In Plant fire Program Organization 364 l

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- 1,1111TitJC n/Jf'rY SYSTDI SETTIhc S AWY7 f.DtJT l

1. 3 Fttr:1. Ulf.nulHP. 1 HTK.H1'li P.1 FUEL, Cl.AnDillC 1NTi%)l: TTY In the event of operation with the core maximus f raction of limiting l

power density (CHT1,PD) r,rcater than

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fraction of rated thermal power (FRP) l the setting shall be piodificd as follows:

l Si (0.66W + 54%) TkP j

CMTLPD.

k For no combinacion of loop recircu-lation flow rate and core then::a1 power shall the APRM. flux scran trip setting be allowed to exceed 120 of rated thermal power.

(Note: These settings assume operation i

within the basic thermal hydraulic L

design criteria. These criteria arc l

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13.4 kw/f t ior 8X8, 8xBR, and P8x8R fuel, MCPR limits of Spec 3.5.k.. If 1

it is determined that either of these design criteria is being violcted l

during operation, action shall bc

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initiated within 15 v.inutoa to restore I'

operation within prescrit,rd li=J tu.

Surveillance requirements for,W.:

scram setpoint are given in specification 4.1.B.

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APRM--When the reactor mode switch is in the STARTUP POSITION, the i

APRM serem shall be set at 1 css i

than or equal to 15% of rated power.

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

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APRM Pod Block Trio Settine -

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e,re '.Se rnal Pove-1,hd t I

(Peneter Pretsure <803 psin)

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The APIV' Rod block trip setting r.h:1)

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P er. the reactor pressure is less thtn or equal to 800 psia, r

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Aeesuse the boiling transition correlation is based on a large quantity of full scale data there le a very high confidence that operation of a fuel assembly at the condition of MCpn =1.07.wou14 not produce. boiling tram-sition. ha, although it is not required to establish the safety limit l

additional margist exists between the safety limit and the actual securence of loss of elseding integrity.

j However, it boiling transition were to occur, else perforation would not l

be expected. Cladding temperatures would increase to appr18minately i

1100 7 wnich is below the perforation temperature of the cladding l

8 material. This has been verified by tests in the General Doctric Test l

Reactor (CETR) where fuel staller in design to BrWP operated shove j

the critical heat flus for a significest period of time (30 mientes)

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vithout clad perforation.

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If reactor pressure should ever esteed 1 BOO pela during seresi power l

operating (the limit of' applicability of the boiling transities sorre-j lation) it would be asetamed that the fuel eladding integrity Safety 1.iait has been violated.

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In addition to.the beiling tranettioe 11mit operation is I

constrained to a aaminus WCA of 13.4 kw/f t for all Sul fuels. This limit is f asched when the Core Hasiasm Fraction of Limiting Power Density equals 1.0 (CHFLPD = 1.0).

For the sese where Core l

Hautaus Traction of LimitinC Power Density escoede the fraction of Rated i

Thermal Power, operstian is permitted only at less than 2001 of rated power and saly with reduced AFRM scree settings as required by specification 2.1.A.1.

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At pressures belev 800 peia, the sore elevation pressure dryp (0 power, O flow) is greater than b.56 poi.

At low powere and flows this pressure differential is maintained in the bypass region of'the sere. Simee the presevre drop la the bypass region is esseotially all elevation head, thecorepressuredropatlowpowersandflowwillalvagebegreater I

than b.56 psi. Analyses show that with a flow of 28K10 lbs/hr bundle

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flow, bundle pressure drop is nearly independent of bundle power and has i

a value of 3 5 psi. Thus,3the bundle flev vith a b.% psi driving head i

lbe/hr. pull scale ATLAS test data taken I

vill be greater than 28x10 at pressures from It.T pela to 800 pela ladicate that the feel assembly critical power at this flow is approminately 3 35 MWt. With tip desige peaking factors this corresponde to a sore theriani power of ante than 305. Thus, a core thermal power limit of 355 for reactor pressures' i

below 800 pela is senservative.

i For the fuel in the core during periods when the reactor is shut down, con.

sideration must also be given to water level requirements due to the effect l

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. yhis reduction in cooling capability could lead to elevated cladding temperatures and eled perforation., %s long as the fuel remains covered with water, suffielent l

cooling is available to prevent fuel clad perforation.

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2.1 B ASES :

LIMITING SAFETY SYSTEM SETTINGS 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 operating conditions up to the design thermal power condition of 3440 MWt.

The analyses 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 t

knowingly be exceeded.

I Conservatism is incorporated in the transient analyses in i

estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram

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delay time, peaking factors, and axial power shapes.

These factors are selected conservatively with respect to their effect on the applicable 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

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predictions made by the model.

The comparisions and i

results are summarized in References 1, 2, and 3.

i The absolute value of the void reactivity coefficient used in the analysis is conservatively estimated to be about i

25% greater than the nominr1 maximum value expected to occur during the core lifetime.

The scram worth used has l

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 as further described in reference 4.

The effect of scram i

worth, scram delay time and rod insertion rate, all l

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 i

time the rods are 60% inserted, approximately four dollars of negative reactivity has been inserted which strongly

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turns the transient, and accomplishes the desired effect.

The times for 50% and 90% 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.

r For analyses of the thermal consequences of the transients a MCpR > limits specified in specification 3.5.K 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 result by using expected values of control

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parameters and analyzing at higher power levels.

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Peactor low yater level set point f or ini t i at ion of tf f Cl and RCIC. clor. int _sita steam isolation valves, and starting LPtl and core spray puispa.

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These systen.s maintain adequate coolant inventony and provide core cooling with the objective of preventing utessive clad temperatures.

The dest;.n of these systems to adequately perform the intended func.

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tion is based on the specifled low level seras set point and initia.

tion set points. Transient analyses topoeted in Section 16 of the i

F5 AJt demonstra te that these tendittens result in adequate salsty i

sergins f or both the fuel and the systes pressure.

L.

s o f e r encis, 1.

Linford. P. t.

" Analytical Methods of Plank Trans tant tvatustions f or the Cencral Electric letting' Water teactor," ktDo 10602 Feb., 1973.

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2. Generic Reload Fuel Application, Licensing Topical Report.

NEDE-24011-P-A, and Addenda.

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"Qurilification of the One-Dimensional Core Transient Model for Boiling Wa ter Reac tor", NEDO-24154, NEDC-24154-P, Oc tober 1978.

4.

Let ter f rom R.11. Buchholz (CE) to P. S. Check (NRC), " Response to NRC request for information on OD,Y$ computer model," September 5, 1980.

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2.2 BASES REACTOR COOLANT SYSTEM INTEGRITY I

To meet the safe'y design basis, thirteen relief valves have been

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t installed on the unit with a total capacity of 83.9% of nuclear boiler rated steam flow. The analysis of the worst. overpressure transient, (3-second closure of all main steamline isolation valves) neglecting the direct scram (valve position scram) results in a maximus vessel l

pressure which, if a neutron flux scram is assumed considering~

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12 valves operable, results in adequate margin to the code allowable overpressure limit of 1375 pois.

- To meet 'the operational' design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open)shows that 12 of the 13 relief valves limit peak system pressure.to a valve which is well below the allowable vessel over,

pressure of 1375 psig.

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TABIE 3.2.F Surveillance Instrumentation

'A Minimum # or

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Onerable Instrument

-Type Indication Channels Instnament #

Instntment and Range Notes 2

lhH 94

-Dryvell and Torus

O.1, ' 20f, (1) liydrogen H M 104 Concentration 2

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PdI-64-137 Drywell to Suppression Indicator

-(1) (2) (5)

PdI-64-138 Chamber Differential O to 2 paid s

pressure -

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1 1.!MITINC CONnITtDNS FOR OPERATION SURVE21.1.ANCE REQUIRENENTS

.5' Control Rods 5.3.B Control Rods centrol rod directional control valves disarmed g,

electrically. This require-hf@% WMnh ment does not apply in the bserving a response in the refuel condition when the

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" **'h reactor is vented. Two con-time a rod is moved when trol rod drives may be removed

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as long as Specification 3.3.A.1 is met.

level of the RSCS.

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When the rod is fully with-drawn the first time after l

each refueling outage or af ter maintenance, observe that the drive does not go to the overtravel popition.

2.

The control rod drive 2.

The control red drive housing housing support system shall support system shall be inspected be in place during reactor after reassenbly and the resulta power operation or when the of the inspection recorded.

reactor coolant system ia pressurized above atmospheric 3.a.

Prior to the start'of control pressure with fuel in the reac.

rod withdrawal at startup the cap-ter vessel, unless all control ability of the Rod Sequence System rods are fully inserted and (RSCS) to properly fulfill its functions Specification 3.3.A.1 is met, shall be verified by the following check.s : -

3.

a.

Whenever the reactor is in the startup or run modes Cequence portion - Select a sequence below 20% rated power the and. attempt to withdraw a rod in the Rod Sequence control System remaining sequences.. Move one rod 3

(RSCS) shall be operable in a sequence and select the remain-except the RSCS constraints ing sequences and attempt to move may be suspended by means of a rod in each. Repeat for all the individual rod bypass sequences.

switches for Group notch portion - For each of the 1 - special criticality six comparator circuits go through i

tests, or test initiate comparator inhibit; I

2 - control rod scram timing verift: reset. N ~- M 5th arremnt per 4.3.C.1.

' test is allowed to continue until When RSCS is bypassed on completion is indicated by individual rods for these illumination of test complete light.

exceptions RWM must be oper-able per 3.3.B.3.c and a second licensed operator i

may not be usbd in lieu of l

RWM.

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i i.IMITINO CONDITIONS Fon OPERATTON' SimVEll.f.ANqh i:LotitHigig h

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4.3.B Cont rol Rodq_

JAN 101978 E3.B Control, Rods b.

Prior to attaining 20% rated power b.

Durinn the shutdown procedure no rod movement is permitted during rod insertion at shutdown the I

between the testing performed tests in 4.3.B.3.a shall be performed above 20% power and the rein-to verify RSCS capability.

statement of the RSCS re-straints at or above'207, c.

The capability of the Rod Worth power. Alignment of rod Minimizer (RWM) shall be verified j

groups shall be accomplished by the following checks:

prior to performing the testn.

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Whenever the reactor is c.

in the startup or run modes 1.

The correctness of the helow 20% rated power the control rod withdrawal i

Rod Worth Minimizer shall be sequence input to the j

operable A second licensed RWM computer shall be operator stay verify that verified before reactor I

the operator at the reactor starttip or shutdown.

console is following the control rod program in lieu 2.

The RWM computer on line i

diagnostic test shall be i

of RWM except as specified in 3.3.B.3.a..

successfully performed.

3.

Prior to startup, proper annunciation of the selec-tion error of at least one out-of-sequence control rod shall be verified.

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4.

Prior to startup, the rod w

d.

If Specificationn 3.3.B.3.a throuch.c cannot be met the block function of the RMI f

reactor shall not be started.

shall be verified by movinn or if the reactor is in the an out-of-sequence control f

run or ntartup moden at Icn8 rod.

than 20% rated power, it 5.

Prior to obtainine, 20% rated f

shall be brought to a shut-down condition immediately.

power during rod insertion i

at shutdown, verify the j

latching of the proper rod l

group and proper annunciation i

5 after insert errors.

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When the RWM is not operable j

a seconci ikcr.::ad operator will j

verify that the correct rod program is followed except as j

specified in 3.3.B.3.a.

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LWqlTINf: Cn%ItT10N$ FOA OPERATION SURVf.!LLANCf. 6 4QUIRDEWTg 3.3.S' Centrol Reds 4.3.8 Centrol Rode 4.

Centrol rods shall not be l

i withdrawn for startup or refueling *unless at least f

two searce range channula have an observed count rate 4

Frier to control red withdraweL f

e o er sec n for startup or during refueling.

verify that at least two source l

S.

.During operation with range channels have se observed limiting control rod pat-count rate of at least three j

torns, as deternined by the designated qualified person.

counts per second.

j mel, eithers l

t a.

Beth REN channels shall S.

When a liettias contrel rod pattern exists, an insttweent be operabla:

functional test of the Eg4 shall be performed prior to withdraval of the designated b,

Contret rod withdrawal rod (s) and at least, ease per shall be blocked.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.

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Scram Insertion Times i

1.After each refueling outage all j

operable rods shall be scram time t

tested from the fully withdrivn i

position with the nuclear system l

pressure above 800 pais l

This j

l testing shall be completed prior to C. Beras fasertien Times exceeding 40% power.

Below 20%

l power', only. rods in those sequences i

1.

The average scram insertion (E12 and A

.o r 5 and B which were fullyyithdrNnin tN) region gime, based on the desnergi-sation of the scram pilot valve from 100% rod density to 502 rod oolenoids as time sero. of all density shall be scram time-tested.

f eperable control rods in the The sequence restraints imposed upon j

remeter power operation coa 1-the control rode in the 100-50 ties shall be ne greater thans percent rod density groupn to the

Insertad From Avg. Scram inser-preset power level may be removed Fully Withdrawn tien Times (see) by use of the individual bypass switches associated with those t

5 0.375 control rods which are fully or f

30 t

0.90 partially withdrawn and are not 50 2.0 within the 100-50 percent rod density to 3.500 groups.

In order to bypass a rod.

l the actual rod axial position must be i

knowns and the rod must be in the i

correct in-sequence position. As

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required by 3.3.B.3.a a second licensed l

i operator ma:r not be used in lieu of RWM for this testing.

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3.3/4.3 3Asts:

3.

The Rod Worth Minimizer (P.k'M) and the Rod Sequence Control i

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system (RSCS) re:itrict withdrava s and innertionn of rentrol rode to pre e.pcetff,d r.cquencen. All patterns annociated witt.

these ccquences have the characteria.t f c that, am,usinr, the l

vorst singic deviation f rom the sequence, the drop of any control rod f rom the fully inserted position to the position of the control rod drive vould not cause the reactor"to sustain a power excursion resulting in any pellet average enthalpy in excess of 280 calories per g am.

An enthalpy of 280 calories i

per gram is well below the icvel at which rapid fuel dispersal could occur (i.e., 425 calories per gram). Primary sy' stem 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 perfor ning; the fdnetien described above, the RL'M ar.d F3CS are:

i not required to irpose any restrictions at core power levels in excess of 20 percent of rated. Material in the cit:cd referer shows that it. is impossible to reach 280 calories per gre.m in the event of a control red drop occurring it power greater enan 2C i

. percent, regardless of the rod pattern. This is true for all notwal end abnormal patterns including those which maximize individual control red worth.

At power levels below 20 percent of rated, abnormal control rod patterns could preduce red vorths high enough to be of i

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concern rel.tive to the 260 caloric per gran red drop limit.

l In this ranr.e the R'.tM and the RSCS constrain the cc.ntrol rod seqvenees and patterns to those which involve only' acceptable rod worths.

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De Rod Vorth Minimizer and the -Pod Sequence Control Systen provide automatic supervision to assure that out of sequence control rods vill not be withdrawn or inserted; i.e., it linite operater devistions f rem planned withdrawal sequences. Ref.

l Section 7.16.5.3 of the TSAR. They serve as a backup to procedce.a control of control red sequences, which limit the maxinun reacts.

vity worth of control rods. Except during specified exceptions, when the Rod Worth Minimizer is out of service a second licensed operator can msnually fulfill the control red pattern con-formance function:: of this system. In this case, the RSCS is backed up by independent procedural controls to assure conformance.

Because it is allowable by bypass certain rods in the RSCS.during specified testing below 20 percent of r a t'e d p o we r in the startup or run modes, a second licensed operator is not an acceptable substitute for the RWH during this testing, N) i 129 f

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2ft.3 $AS?$t iloe pro ste!e the nparator with a visual indtratica of neu-tron level. The consvepiesten of re.:tivtry accider.ts art l

functionn of the ir.itial neutron fin. *The equiremut of at Irsst 3 tounta per netond easures that ar.f t.tcr.sient, should it occur, octins at ur above the inittil value nf 10'" of rntr.1 power i.. ?d in Lt.Unstyse., nf transient s f ruo cold'randitions. One oyerable !Jfi chan.it i voald he a,deluste to munitor the approach in crititality using hencoe.rneous patterm of scattered control tod v!!hdraval. A sir.1 :gr of two operable SRli's are proviLed as an edded cor.servation.

5.

The Rc4 Block Monitor (rem) is designed to ' auto =ati'r. ally prevent fuel de. age ih the event of errencous rod vithdrawal f ron locatio.s of high power density during high pover level operation. Two channels are provided, ir.d one of the se cuy be bypassed f roo the console f or ruintenance and/or testing.

Trippins of one of the channels vill block error.cous red withdrawat soon ensuuh to prevent fuel da. age.

The spect-fled restrictions with one chonnel cut of tervice conserva-tively assure that f ue'l damage vill not occur due to rod withdtsval,artors when this cond?. tion exists.

A limitinR contrp1 red pattern is a pattern which resulin

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in the core being on a thermal hydraulic linit, (ie.

MCPR given by Sacc.'3.5.K or LHCR of 13.4 kw/ft.

During use of such patterns, it is judged that testing of the R Bll system prior to viti.-

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

It is normally the responsibility of the llu c i c a r Engineer to identify these limiting patterub 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 syste9 is designated to bring the reactor suberitical at the rate fast enough tn prevent fuel d a r.o r. c s ic. to prevent tho MCPR from becoming less than 1.07.

The

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limitinC power transient is given in Reference 1. Analy.i>

of this trancient shows that the neCative reactivity rates resulting frns the scram with the average response of all the drives as given in the above specifseation provide the r e q u i r,e d protection, and HCPR remains greater than 2.07.

On an early BWR, some degradation of control rod scram petfornnnre occured durIng plant startup and was determined **

to be c a n a...' by s

I l

l

_., - - _ _ ~ _.,

l i

Lf_MITINC CON 9tT10HS FOR CPERATION SURVEILLANCE REQUIREMENTS l

l4.5 CORE AND CONTAINKINT COOLING 3.5 CORE AND CONTAINMENT COOLING gYSTf.MS SYSTDt3 l

l Applicability Applicability _

j i

Applies to the operational Applies to the surveillance statue of the core and contain-requiremente of the core and containment cooling systone when ment coolins systems.

the corresponding limiting condi-l tion for operatio4 is in effect.

l Objective Objective i

To assure the operability of To verify the operability of the the core and containment cooling core and containment cooling systems under all conditions for systems under all conditions for which this cooling capability is which this cooling capability is j

en essential r9eponse to plant an essential response to plant i

abnormalities.

Specification Specification j

A.

Core Spray System (CSS)

A. Core Spray System (Css) 1.

The CSS shall be opera-1.

Core Spray Systou 'festing.

i ble i

Item Frequency (1) prior to reactor startup from a a.

Simulated Once/

l cold condition, or Automatic Operat.ing Actuation Cycle (2) when there is irra-test disted fuel in the vessel and when the b.

Pump Opera-Once/

l reactor vessel pres-bility wanth i

eure is greater than atmospheric pressure.

c.

Motor Once/

f

^

except as specified Operated month in specification Valve 3.5.A.2.

Operability l

l d.

System flow Once/3 rate Each months i

loop shall

{

deliver at a

least 6250

{

spm against I

a system

{

head corree-I ponding to a l

l

{

143

}

f f

i h

t

,..n.

.-.a--

-~

--n,

.. s

,.,,,w

,,,-,-,v,

,,_nn,

i 4

I,ItGTUr0 CO'fDITIO:ts P3R 0?r?ATIO?!

SURVEILLAftCE BE0tf!R' E:CCITT 3.5.8 Meeldunt Heat Removal Svatem 4.5.8 Realdual Heat Removal 5 *re,n 2

I (RHRS) (LFCI and Containment (RHRS) (LPCI and Containment Coeling)

--Coeling) f 1.

The AllRS.shall be operable:

1.

a.

Simuisted Once/

l Automatic Operating (1) prior to a reactor Actuation Cycle startup from a Cold Test Condition; or

.(2) when there is irra-b.

Fump opera-Once/

disted fuel in the bility month reactor vessel and when the reactor vessel pres-c.

Motor Opera--

Once/

sure is Areater than ted valve month stmoapheric except.as operability l

I specified in specifica-tions 3.5.3.2, through d.

Pump Flov Rate Once/3 l

3. 5. r, 7 montho f

e.

Test Check Valve Once/

2.

With the. reactor vensel pres-oure lesa thon 105 p a i P.. the Operating l

AHRS may he rea.oved f rom ser-Cycle j

vice (except that tun RHR pumps-l containment coolin;;. mode and Each LPCI pump shall deliver 9000 associated heat exchangers must gpm against an indicated system remain operabic) for a period pressure of 125 psig. Two LPCI pumps I

not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while in the same loop shall deliver I

being drained of suppression 15,000 gpm against an indicated chamber quality vater and system pressure of 200 psig.

filled with prieary coolant quality vater previded that 2.

An air test on the dryvell e,r.a-during cooldoen two loeps with torus heade s and no:: lea shall

?

Q one pump per loop or one loop vith be conducted once/5 years. i.

g two pumps, and associated diesel vater test may be perf. rned on generators, in the core spray syster are operable.

g 3.

If one ftllM pump (LPCI code) 3.

When it is determined that one RHH ta innperable. the reactor pump (LPCI mode) is innperable at a may remain in nperation f or a time when operability is re9utred*

period not to exceed 7 days the remaining RHR pumps (LPC) mode) _

I provided the rem ining RHR and active components in both access pumpa (LPCI mode) and both paths of the RHRS (LPCI modo) and

{

necesa patha of the RHR$

(LPCI mode) and the CSS and the CSS and the diesel r.cncrators the diesel sentrators remain shall he demonstrated to be opera-

operable, bic immediately and daily

{

(

thereafter.

=

145 3

f I

i i

i E

~,

LIMITING' CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

')

i 3.5.I Average Planar Linear Heat Generation 4.5.I Maximum Average Planar Linear Heat

(

Rate Generation Rate (MAPLHGR)

During steady state power operation, the The MAPLHCR for each type of fuel as a Maximum Average Planar Heat Generation function of average planar exposure l

Rate-(MAPHGR);for each type of fuel as shall be determined daily during a function of average planar exposure reactor operation at 2 25% rated shall not exceed the limiting value thermal power.

shown in Tables 3.5.I-1 through 3.5.I-5.

-If at any time during steady state j

operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall bei initiated within 15 minutes to restore l

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 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

I J.

Linear Heat Generation Rate (LHCR) t J,

Linear Heat Generation Kate (LHGR)

During steady state power operation, the The LHGR for 8x8, 8x8R, and P8x8R fuel linear heat generation rate (LHCR) of shall be cheeked daily during reactor any rod in any fuel assembly at any operation at 225% rated thermal power.

axial location shall not exceed 13.4 Kw/ft, If at any time during steady state a

operation it is determined by normal I

surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If the LHCR is not returned to within I

the prescribed limits within two (2)

{

hours, the reactor shall be brought to j

the' Cold Shutdown condition within 36 t

hours.

Surveillance and corresponding action shall continue until reactor oper-ation is within the prescribed limits.

i l

f e

l l

s i

d i

i h

l 159 l

i 5

i r

e LIMITING CONDITIONS FOR OPEPATION SURVEILLANCE REQUIREMENTS 3.5.G Automatic Depressurization 4.5.G Automatic Depressurization System system 3.

If specifications 3.5.G.1 and 3.5.G.2 cannot be met, an orderly shutdown will be initiated and the

{

reactor vessel I

pressure'shall be reduced to 105 psig or less within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I H.

Maintenance of Filled i

Discharge Pipe H.

Maintenance of F'illed

(

Discharge Pipe Whenever the core spray t

systems, LPCI, HPCI, or The following surveillance RCIC are required to be requirements shall be

[

operable, the discharge adhered to assure that the piping from the pump discharge piping of the discharge of these systems core spray systems, LPCI, to the last block valve HPCI, and RCIC are filled:

I shall be f111ed.

{

l t

The suction of the RCIC and HPCI pumps 1.

Every' month prior to the testing shall be aligned to the condensate of the RllRS (LPCI and Containment j

storage tank, and the pressure suppres-S ray) and core spray systen, the P

sion chamber head tank shall normally discharge piping of these systems be aligned to serve the discharge piping shall be vented from the high point i

of the RHR and CS pumps. The condensate' and water flow determined.

head tank may be used to serve the RllR i

i and CS discharge piping if the PSC head 2.

Following any period where the LPCI i

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 pumps shall indicate not less charge piping of the inoperable sys-I than listed below, tem shall be vented from the high I

Pl-75-20 48 psig peint prior to the return of the P1-75 48 Pl-74-51

~48 psig system to service.

45 psig i

Pl-74-65 ~ 48 psig 3.

Whenever the HPCI or RCIC system is lived up to take suction from the i

condensate storage tank, the dis-l charge piping of the HPCI and RCIC shall be vented from the high point j

of the system and water flow observed on a monthly basis.

4.

When the RHRS and the CSS are re-quired to be operable, the pressure

(

indicators which monitor the dis-

)

charge lines shall be monitored I

daily and the pressure recorded.

4 158 N

i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUTREFfENTS 3.5.K Minimum Critical Power Ratio 4.5.K. Minimum Critical Power (MCPR)

Ratio (MCPR)

The minimum critical power ratio (MCPR)

1. MCPR shall be determined daily during i

as a function of scram time and core reactor power operation at>25% rated flow, shall be equal to or greater than thermal power and following any shown in Figure 3.5.K-1 multiplied by change in power level or distribution i

the Kg shown in Figure 3 5.2, where:

that would cause operation with a

[

limiting control rod pattern as e

T = 0 or Tave TB

, whichever is described in the bases for T A-TB greater Specification 3.3

2. The MCPR limit shall be determined i

,IA=0.90 sec (Specification 3 3.C.1 scram for each fuel type 8X8, 8X8R, P8X8R, time limit to 20% insertion from from figure 3.5.K-1 respectively fully withdrawn) using.

y F

g IB=0 710+1.65 N

(0.053) [Ref5_

a. T= 0.0 prior to initial scram l

e

~

j~

n time measurements for the cycle,

~

performed in accordance with T(

specification 4.3.C.1.

I Tave= I'l r)

b. Tas defined in specification n = number of surveillance rod tests 3.5.K following the conclusion of performed to date in cycle (including each scram time surveillance test BOC test).

required by specifications 4.3.C.1 and 4.3.C.2.

T=

Scram time to 20% insertion from i

fully withdrawn of the ith rod.

The determination of the limit must be completed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of each scram time surveillance N=

total number of active rods measured required by specification 4.3.C.

in specification 4.3.C.1 at BOC k

If at any time during steady state operation it is determined by normal surveilance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed l

limits. If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown ccV.ition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , surveillance and corresponding action,shall continue until reactor operation is within the prescribed limits.

L. Reporting Requirements If any of the limiting values identified in Specification 3 5.I J, or K are l

l exceeded during steady state operation l

and specified action is taken;the event shall be logged and reported in a 30-day report.

t i

160 l

3.5.J Linear 11 eat Generation Rate (LilGR)

Thin specification assures that the linear heat generation rate in any rod is less than the design linear heat generation if fuel I

pellet densification is postulated.

B The LilCR for 8x8, 8x8R, and P8x8R fuel shall be checked da,ily dur-

{

ing reactor opetution at 125% power to detet,eine if fuel bu'rnup, or con-trol rod movenient has caused chsntes in power distribution. For LHCR to bu a limiting value below 25% rated therwal power, the HTrl' would have to be greater than 10 which is preclu,ted by a considerable martin when employi6g sny permissible control _ rod pattern.,

r 3,3,g.

Min taus Critied Pwer R.atio (MCP A)

At core thermal power levele less than or equal to 23:, the reactor will be aparettes at minimum recirculation purup speed and the nuderator vold content l

vill be t e ry small. For all dest 6 cat ed control rod pat terns shich imay be ein-y played et t hi s re tt.t. operating plant eEperience and thermal hydraulic enel-yele edicated that the reeutting NCPA value to in taceee of' requiremente i

by a considerable martia. With this lov void content, any inadvertent e. ore t

flow lacreses would ortly p1*ece operatf on in a more conservative code rels-i give to NCPR. The delly requirement for calculating NCPr. above 25I rated thereat i

power is suf ficient eince power distribution shif te are very etcw when there have not been significant. pcver or control red changes. The requirement for calculatted PCP A when a 11alting control rod pattern to approached ensures that

[

HCPA vill be Leovn following a change'1n power or power shape (tegardless of magnitude) that could place operation at a thetiest limit.

5. 5.1.. Reporttet Re qu i r em ent e j

The LCo's seeociated with monitoring the f uel rod operatinn condit inese are required to be sat at all t imes, i.e., t here is na allowable time in which 1

the plant cam bovingly exceed the liniting value s f or MAPUICR, IJICR. cnd f

MCTA.

le le e requirement. se et e r ed in S pecif ic at ions 3 5.1..J.

risd.K.

t ha t if at an y t ime du r l og s t e a d y s t a t e powe r ope r e t ien, it le deter.alned that the liniting vduce for N/1DICR, t.HC R, or NC P R a r e e ic t ed ed a c t iot. le

?

then inittsted to restore operation to within the prescribed limite. Thie actico is toitiated as soon as normal surveillance indicates that an cperating ltr.

)

it h4s been reached. Fach event involvine eteady state operation beyond a specified I

11xit shall b' reported within 30 days. It must ce reccania==1 ttu there is alveys an act ion v>!rh would return soy of the paramstere DtAPU4CK, int, er McFR) to within prescribed limits, namely power reduction. 16 der oost circumstances, thte vill not be the only alternettve.

H.

8'I"'84'8 i

1.

TV.1 Decei!!cetico Tf f ec t e co General tiectric Boir.ng Waar kaastor Puel." Su pp t eee nt e 6, 7, and 3. ktIP'-10 7 3 5 Au gus t 19 M.

s t

2.

Su p pl emen t 1 to fechnics! Report on Densif ications of Ceneral r

tiectric Practor Puole. Decrzber 14, 1974 (L'5 A R.sgulat or y $t a f f ).

l J.

Commu nic.a t i on :

V. A. Moore t o 1. 5. Hi t c he ll, Moo t f ie d CT. P.ao e l f or Puel Dens if ic a t ico, Doc'se t 50-321, March 27, 197.

4.

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

I 5.

Letter f rom R.11. Buchholz (C.E.') to p. S. Check (NRC),

" Response to NRC request forinformationonODYNcomputermodel/'

September 5, 1980.

169 l

i i

r,,

. ~, -

=

J Table 3.5.I-1 i

MAPLilGR VERSUS AVERACE Pl.ANAR EXPOSURE e

Fuel Type: 8DB274L

{

Average Planar l

Exposure MAPLHGR (Mwd /t)

(kW/ft) l i

200 11.2 1,000 11.3 l

5,000 11.9

~

10,000 12.1 15,000-12.2 20,000 12.1 25,000 11.6 30,000 10.9 l

1 35,000 9.9 40,000 9.3

{

5 Table 3.5.I-2 MAPLHCR VERSUS AVERACE PLANAR EXPOSURE Fuel Type: 8DB274H l

i i

Average Planar t

Exposure MAPLHGR f

(Hwd/t)

(kW/ft) i 200 11.1 i

1,000 11.2 l

l 5,000 11.8 i

12.1 l

10,000 i

15,000 12.2 j

20,000 12.0 l

t 25,000 11.5

{

30,000 10.9 i

35,000 10.0 l

9*3 I

40,000 I

d 1

171 l

6 i

.--,.,,,e m,.,.-e,

., r r -,

,,,.,m,,,-

..,,....-.y.,-,,,,,,,,,,,--,,,-y-,,v,,,,-..-~v-~,-,n~

- t Table 3.5.I-3 MAPLHCR VERSUS AVERAGE PLANAR EXPOSURE t

Fuel Type: 8DRB26511 Average Planar Exposure MAPLHGR f

-(mwd /t)

(kW/ft) i 200 11.5 1,000 11.6 3

5,000 11.9 10,000 12.1 15,000 12.1 i

e 20,000' 11.9 l

?

25,000 11.3 I

30,000 10.7 35,000 10.2 i

40,000 9.6 l

l Table 3.5.I-4 l

MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE f

Fuel Type: 8DRB265L and P8DRB265L i

Average Planar Exposure MAPLHGR f

(mwd /t)

(kW/ft) i

[

200 11.6 1,000 11.6 I

5,000 12.1 i

i 10,000 12.1 15,000 12.1 i

20,000 11.9

)

?

I

-25,000 11.3 J

l 30,000 10.7 35,000 10.2 40,000 9,6 l

i 4

I 172 f

i

-q g

r s

w,,

-,+n-

,.,-,-e--,-,,y,-,,.w,,

y

--,.e-..,v,-wv,,,,-n aww w, v n,, -

,w-a-

Table 3. 5.I-5 MAPLilGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: P8DRB284L, CLTA-1, GLTA-2 Exposure MAPLilGR (mwd /t)

(kW/ft) 200 11.2 1000 11.3 5000 11.8 10,000 12.0 15,000 12.0 20,000 11.8 25,000 11.2 30,000 10.8 35,000 10.2 40,000 9.5 t

+

i f

172a kw

I e

4 e

.w.-

1 h..

8g'

' m gm e....

...e..e.ee...g..e

.e.

.g.

g.

el g,

I

.6 g

.g

,g

P

.3

..*..p

+0 '1 l.

...=...

.gl

- l8 l.

..'.I..

I-6 48...

1g 9 9

e g

... g *.

'. e l.'

g a

..q 9

.d.

9

.=.

l2

,'**+..

...og.,..

y g

g g.

g k

g

..6.

g

.,...[....

,6.

.,l,.-

1. 31 _.

. t...

1..

.l..-

.l.

~~0 psid)~~

~~-7.25" (0 psid dif feren-I tial pressure control) b.~~

~~Maximum water level =~~

~~-1" i~~

~~227~~

~~i LIMITING CONDITIONS FOR OPERATION ~~~

~~SURVEILLANCE REQUIREMENTS E9_N_3INMENT SYSTEMS l~~

~~3. ~1 CONTAINMENT SYSTEMS b 'I T~~
i L
6.
Drywell-Suppression Chamber 6.
Drywell-Suppression Chamber l
D1fferential Pressure Differential Pressure a.
Dif f erential pressure a.
The pressure diff er-between the drywell and ential between the suppression chamber shall drywell and suppression be maintained at equal-chamber shall be recorded to or greater thsn 1.1 at least once each shif t.
paid except as specified in (1) and (2) below:
l (1) This differential shall be established within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of achieving operating temperature and pressure. The differential pressure may be reduced to i
less than 1.1 paid 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to a scheduled shutdown.
i (2) This differential may be decreased to less than 1.1 paid for a maximum of four i
t hours during required j
operability testing of the RFCI system, RCIC system and the drywell-pressure suppression chamber vacuu,,m breakers, b.
If the differential pressure of specifica-tion 3.7.A.6.a cannot be I
maintained and the differential pressure I
cannot be restored within the subsequent six (6) l hour period, an orderly shutdown shall be init-iated and the reactor I
shall be in the Cold 1
Shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
235a
' l LIMITING CONDITIONS FOR OPERATION 1 SURVEILLANCE REQUIREMENTS I
3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS i
i H.
Containment Atmosphere H.
Containment Atmosphere
~
Monitoring (CAM) System -
Monitoring (CAM) System -
j H Analyzer
~
H Analyzer 2
-2 1.
Whenever the reactor is 1.
Each hydrogen analyzer.
not in cold shutdown, two system shall be demon-independent gas analyzer strated.0PERABLE at systems shall be operable least once per quarter l
for monitoring the drywell by performing a CHANNEL j
and the torus.
CALIBRATION using standard
~
gas samples containing 2.
With one hydrogen analyzer a cominal eight volume inoperable, restore at percent hydrogen balance least two hydrogen nitrogen.
i
^
analyzers to OPERABLE status within 30 days or 2.
Each hydrogen analyzer be in at least HOT system shall be demonstrated i
SHUIDOWN within the next OPERABLE by performing 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 's.
a CHANNEL FUNCTIONAL TEST monthly.
l 3.
With no hydrogen analyzer j
OPERABLE the reactor t
shall'i. in HOT SHUTDOWN f
within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
E l
I I
f a
1 1
e d
i i
I
\\
1 249 i
s
.-.m
TABLE 3.7.A (Continued)
Action os Number of Power Maxt:sms Operated Valves Operating Normal Initiating Grous valve Identification Inboard Outboard Time (sec.)
Position
- 511taal 3
Reactor water cleanup system supply isolation valves FCV-69-1, & 2 1
1 30 0
GC J
~
3 Reactor water cleanup system return isolation valves FCV-69-12 1
60 0
GC 4
HPCIS steamline isolation valves 1
1 20 0
GC FCV-73-2 & 3 5
RCICS steamline isolation valves 1
1 15 0
CC,
g FCV-71-2 & 3 3
6 Dryvell nitrogen purge inlet isola-1 5
C
)
SC tien valves (FCV-76-18) 1 6
Suppression chamber nitrogen purge 1
5 C
SC inlet isolation valves (FCV-76-19) 1 6
Dryvell Main Exhaust isolation C
SC 2
2.5 valves (FCV-64-29 and 30) 6 Suppreeston chamber main exhaust 2
2.5 C
SC isolation valver (FCV-64-32 and 33) 6 Dryw11/ Suppression Chamber perse C
SC 1
2.5 inlet (PCV-64-17) 6 Dryeell Atwephere purge inlet C
SC DCV-44-la) 1 2.5 4
m.- - --- m.
.--_,.----.w.,_
m... -,.,-
w...e.
r-c
..-.......,r.
..,,--.-...~.-..,.--.,,-~....,w,
TABLE 3.7. A (Continued) -
Nurber of Power-Maximum Action on Operated Valves Operating Monnal Initiating Group Valve Identification -
Inboard Outboard Time (sec.) Position Signal 6
Suppression Chanber purge inlet (FCV-64-19) 1 2.5 C
SC a
6 Drywell/ Suppression Chamber nitro-
~~gen purge inlet (FCV-76-17) 1 S-C SC 6~~
Dgwell Exhaust Valve Bypass to Standby Gas Treatment System
( FCV-64-31) 1 5
C SC 6
Suppression Chanber Exhaust Valve Bypass to Standby Gas Treatment System (FCV-64-34) 1 5
C SC,
i 6
Drywell/ Suppression Chamber Nitrogen Purge Inlet (FCV-76-24) 1 5
C SC i
I, 7
RCIC Steamline Drain (FCV-71-6A, 68) 2 5
0 GC j
7 RCIC Condensate Pump Drain (FCV-71-7A, 78) 2 5
0 GC 4
7 HPCI Hotwell pump discharge isola-j tion valves (FCV-73-17A,178) 2 5
C SC i
i 7
HPCI steamline drain (FCV-75-57, 58) 2 5
0 GC I
8 TIP Suide Tubes (5)
I per guide NA C
GC tube l
i 4
i
.-.-.~.--..m-.~...
.---.-.e..
.rw--e...,e..---e-...-..--mm,..-.,.r-tr w--
.-...-.e..-.
-.-,.-.--.-s--
..-..m.m-
- - - -. - + -
BASE 5_
3.7.A & 4.7.A Primary Containment The integrity of the primary containment and operation of the core standby cooling system in combination, limit the off-site doses to values less than those suggested in 10 CFR 100 in the event of a break tr, the primary system piping., Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists. Concern about such a violation exists wh.en-An ever the reactor is critical and above atmospheric pressure.
exception is made to.this requirement during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required. There will be no pressure on the system at this time, thus greatly reducing the chances of a The reactor may be taken critical during this period; pipe break.
however, restrictive operating procedures will be in effect again to minimize the probability of an accident occur ring. Procedures and the Rod Worth Minimizer would limit control worth such that a rod drop would not result in any fuel damage. In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shall be operational during this time, offer a sufficient barrier to keep offsite doses well below 10 CFR 100 limits..
The pressure sucpression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of The pressure suppression chamber water volume must absorb the system.
the associated decay and structural sensible heat released during primary system blowdown from 1.035 psig. Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss of coolant accident, the pressure resulting from isothermal compression plus the vapor oressure of the liquid must not exceed 62 psig, the suppression chamoer maximum pressure. The design volume of the.
suppression chamoer (water and air) was ottained by considering that the total volume of reactor coolant to be concensed is discharsed to the suppression chambea and that the drywell volume is ourged to the suppression chamber.
Using the mir.imum or maximum water levels given in the specifications, con-tainment pressure during the design basis accident is approximately 49 psig, vnich is below the maximum of 62 psig. The maximum water ' level indi-cation of -1 inch corresponds to a devneomer submergence of 3 feet 3
7 inches and a water volume of 127,800 cubic feet with or'128.700 ft without the drywell-suppression chamber differential pressure control. The minimum water level indication of -6.25 inches with dif ferential pressure cont rol and
-7.25 inches without differential pressure control corresponds to a downcomer submergence of approximately 3 feet and a water volume of approximately 123,000 cubic feet. Maintaining the water level between these levels will assure that the torus water volume and down-comer submergence are within the aforementioned limits during normal
~
plant operation. Alarms, adjusted for instrument error, will notify the operator when the limits of the torus water level are approached.
The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation. Thus, with respect to down-comer submergence, this specification is adequate. The maximum temperature at the end of blowdown tested during the Humboldt Bay and Bodega Bay tests was 170*F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperaturee above 170*F.
267 u
BASES Should it be necessary to drain the suppression chamber, this should only be done hen there is no requirement for core standby cooling systems operatibilit w
Under full power operation conditions, blowdown from an initial suppression chamber water temperature of 95'F results in a peak long term water temperature of 170*F which is sufficient for complete condensation. At this temperature and atmospheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus there is not dependency on containment overpressure.
Experimental data indicate that excessive steam condensing loads can be I
avoided if the peak temperature of the' suppression pool is maintained below 200*F local.
l Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressuirzed in a timely manner to avoid the regime of potentially high suppression chamber l
loadings.
Limiting suppression pool temperature to 105 F during RCIC, HpCI, or I
relief valve operation when decay heat and stored energy is removed from the primary system by discharging reactor steam directly to the suppression chamber assures adequate margin for controlled blowdown anytime during i
RCIC operation and assures margin for complete condensation of steam from the design basis loss-of-coolant accident.
In addition to the limits on temperature of the suppression chamber pool l
water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open. This action would include:
(1) use of all available means to close the valve, (2) initiate, suppression pool water cooling heat exchangers (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the. pool.
If a loss-of-coolant accident were to occur when the reactor water temperature is below approximately 330 F, the containment pressure will
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not exceed the 62 psig code permissible pressures even if no condensation i
were to occur. The maximum allowable pool temperature, whenaver the reactor is above 212*F, shall be governed by this specification. Thus, specifying water volume-temperature requirements applicable for reactor-water temperature above 212*F provides additional margin above that available at 330*F.
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268
1 In conjunction with the bbrk I Containment Short Term Program, a plant unique analysis was performed (" Torus Support System and Attached Piping Analysis for
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the Browns Ferry Nuclear Plant Units 1, 2, and 3," dated September 9,1976 and l
r supplemented October 12, 1976) which demonstrated a factor of safety of at l
1 east two for the weakest element in the. suppression chamber support system L
and attached piping. The maintenance of a drywell-suppression chamber differen-tial pressure of 1.1 psid and a suppressior. chamber water level corresponding to a downcomer submergence range of 3.06 feet to 3.58 feet will assure the i
integrity of the suppression chamber when subjected to post-LO'CA suppression pool hydredynamic forces.
Inertinn
?
The. relatively small containment volume inherent in the CE-BWR pressure suppres-tion containment and the large amount of zirconium in the core are such that
(
the occurrence of a very limited (a percent or so) reaction of the zirconium and steam during a loss-of-coolant accident could lead to the liberation of hydrogen combined with an air atmosphere to result in a flammable concentration in the containment.' If a sufficient amount of hydrogen is generated and oxygen is available in stoichiometric quantities the subsequent ignition of the hydrogen in rapid recembination rate could' lead to failure of the containment to maintain a low leakage integrity. The <4% hydrogen concentration minimizes the possibility of hydrogen combustion following a loss-of-coolant accident.
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i 269 1
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BASES The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is much more probable then the occurrence _
of the loss-of-coolant accident upon which the specified oxygen concentration limit is based. Permitting access to the drywell for leak inspections during a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of safety. Thus, to preclude the possibility of starting the reactor and operating for extended periods of time with significant leaks in the primary system, leak inspections are scheduled during startup periods, when the primary system is at or near rated operating temperature and pressure. The 24-hour period to provide inerting is judged to be sufficient to ~ perform the leak inspection and establish the required oxygen concentration.
To ensure that the hydrogen concentration is mainegined less than 4% following an accident, liquid nitrogen is male!Ained on-site for containment atmosphere dilution. About 2260 gallons would be sufficient as a 7-day supply,.and replenishment facilities can deliver liquid nitrogen to the site within one day; therefore, a requirement of 2500 gallons is conservative. Following a loss of coolant accident the Containment Air Monitoring (CAM) System continuously monitors the hydrogen concentration of the containment j
volume. Two independent systems ( a system consists of one hydrogen sensing circuit) are installed in the drywell and the torus. Each sensor j
and associated circuit is periodically checked by a calibration gas,to verify operation. Failure of one system does not reduce the ability to monitor system atmosphere as a second independent and redundant system vill still l
be operable.
In terms of separability, redundancy for a failure of the torus system is based upon at least one operable drywell system. The drywell hydrogen concentration can be used to limit the torus hydrogen concentration during post LOCA conditions. Post LOCA calculations show that the CAD system initiated within two hours at a flow rate of 100 scfm will limit the peak drywell and wetwell hydrogen con-centration to 3.6% (at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) and 3.8% (at 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months ), respectively.
i This is based upon purge iniciation after 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months at a flow rate of 100 scfm to maintain containment pressure below 30 psig. Thus, peak torus hydrogen concentration can be controlled below 4.0 percent using either the direct torus hydrogen monitoring system or the drywell hydrogen monitoring syste'm with appropriate conservatism (s 3.8%),
as a guide for CAD / Purge operations.
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1 270 1
'4 g,0 KAJOR nr. SIGN FLATURES t
5.1 51TL JgunEs j
Browns Terry unit J is located at Browns Terry Nuclear Plant site on property ovned by the United States and in custody of the TVA. The alte shall consist of approximately 840 acrks f
on the north shore of L/ heeler Lake at Tennessee River Mile 294 in Limestone County, Alabama. The minimum distance troe l
the outside of the secondary containment building to the boundary of the exclusion area as defined in 10 CFR 100.3 1
shall be 4,000 feet.
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l 5.2 REACTOR
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A.
The reactor core may contain 764 fuel assemblies consisting '
I of 8x8 assemblics having 63 fuel rods.each, and 8x8R (and P8x8R) asser.blies having 62 fuel rods each.
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B.
The reactor core shall contain 185 cruciform-shaped control i
rods. The control material shall be boron carbido powdct (B C) compacted to approximately 70 percent of theoretical j
6 1
l density.
5.3 REACTOR VV.SSEL
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The reactor vessel shall be as described in Table 4.2-2 of the F5AR. Thi applicable design codes shall be as described in Table 4.2-1 of the FSAA.
5,4
,tONTAIRMLNT A.
The principal design parameters for the primary containment shall be as given in Table 5.2-1 of the FSAR. The appitcable desitn codes shall be as described in Section 5.2 of the TSAA.
B.
The secondary containment shall be an described in Section 5.3 of the r$AR.
C.
Penetrations to the primary containment and pipine, passing through such penetrations shall be designed in accordance with the standards set fotth in Section 5.2.3.4 of the r5AR.
i 5.5 rutt STntAr,t A.
The arrangement of fuel in the new-fuel storate f acill:y tshall be such that k for dry conditions, is less than 0.90andfloodedisTesa, than 0.95 (Section 10.2 of TSAR.).
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ENCLOSURE 2 f
EXPLANATION OF CHANGES i
1.
Incorporate changes associated with reload analysis, implementation of ODYN option B surveillance requirements, removal of all 7 x 7 fuel types, and elimination of 8 x 8 fuel power spiking penalty -
pages vii, viii, 9, 16, 19, 25, 131, 159, 160, 169, 171, 172, 172a, 172b,219, and 330, 2.
Relax rod sequence control system requirements to allow low power physics testing - pages 122, 123, 124, and 129.
t 3.
Clarification of requirements on charging ECCS systems - page 158 F
4.
Mark I torus modifications; shorten downcomers and add T-quenchers -
pages 227, 235a, 267, 268, and 269.
5.
Containment purge modification to decrease valve closure time -
4 pages 251 and 252.
. 6.
Move natural circulation bases to appropriate section and rename section 3.6.F - pages 19, 182, and 221.
7.
Miscellaneous corrections - pages 143, 145, and 173a.
8.
Hydrogen Monitoring - pages 79, 249, and 270 JUSTIFICATION The justification provided consists of the GE report " Supplemental f
Reload Licensing Submittal for Browns Ferry Nuclear Plant Unit 1 Reload No. 4 (Cycle 5)" (Y1003J01A19) and Errata and Addenda to GE I
topical report " Loss-of-Coolant Accident Analysis for Browns Ferry Nucicar Plant Unit 1 (NEDO-24056).
Justification for the new Hydrogen Monitoring T.ystem is provided.
This justification is identical to that included with the unit 3 license amendment request TVA BFNP TS 148 (reference TVA letter f rom J.L. Cross to H.R.Denton dated September 5,1980).
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BROWNS FERRY NUCLEAR PLANT PRIMARY CONTAINMENT HYDROGEN MONITORING SYSTEM Each reactor is equipped with two totally independent systems for monitoring hydrogen concentrations in the drywell and the torus. Each system includes a Hays Republic Division, Milton Roy Company, Model 643D Condu-therm thermal conductivity type gas analyzer, sample pumps, sample
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moisture removal equipment, and associated valves and controls, all mounted in a cabinet external to the primary containment. Gas samples are withdrawn from either the drywell or the torus for analysis.
The hydrogen analyzers use the principle of thermal conductivity to analyze the concentration of hydrogen in a gas mixture containing primarily i
nitrogen. The thermal conductivity of the sample changes linearly with the change in hydrogen concentration. The analyzer assembly consists of two identical electrically self-heated, glass-covered, temperature-sensitive resistors which are mounted in separate chambers in the analyzer cell block. These resistors form two legs of a Wheatstone bridge. A l
reference gas with known thermal conductivity dif fuses into one of the cell chambers (the reference cell), and the drywell or torus sample diffuses into the other chamber (the measuring cell). The reference gas absorbs heat from the reference resistor in direct ratio to its thermal conductivity. The amount of heat aasorbed remains constant since the reference gas has a constant thermal conductivily, me.intaining the temperature and the resistance of the reference realstor,s constant. The sample gas absorbs heat from the measuring resistor in direct ratio to its thermal conductivity. As the composition of the sample gas changes, j
its thermal conductivity changes. This causes the amount of heat absorbed by the measuring resistor to change, changing the resistance of the measuring resistor. The Wheatstone bridge is calibrated so that t perev.t hydrogen balances the bridge, producing no electrical output. As the l
resistance of the measuring resistor changes, the bridge becomes unbalanced, producing an electrical output that is projortional to the hydrogen concentration in the gas sample. This signal is transmitted to a recorder in the control room.
i Monitoring is continuous with an accuracy of 2 percent.
No special operating procedures are require.
f Each system is qualified for sampics at 340 F, 45 psig, 100 percent RH, and post-LOCA fission product activity. All piping,, cabling, the
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equipment cabinets, valves, readouts, and recorders are seismic Class I.
Each system ia powered from separate electrical fuses. All sample lines which penetrate the primary containment a're equipped with one inboard and one outboard isolation valve per line.
Each system is designed to fully comply with the requirements of NUREG 0578 and Regulatory Guide 1.7 for primary containment hydrogen monitoring.
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ML20126J607

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