ML19269C554

Amend 47 to License DPR-33 Changing Tech Specs to Permit Plant Operation in Cycle 3 Following Current Refueling Outage

ML19269C554
Person / Time
Site:	Browns Ferry
Issue date:	01/17/1979
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML19269C555	List: ML19269C554 ML19269C556 ML19269C557
References
NUDOCS 7902060016
Download: ML19269C554 (53)

Text

f a

[

4 UNITED STATES y*

t NUCLEAR REGULATORY COMMISSION I

E WASHINGTON, D. C. 20555 5

l

\\*****/

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 47 License No. DPR-33 1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated September 8, 1978, as supplemented by letters dated October 5, 1978, November 30, 1978, December 5, 1978, December 14, 1978, January 8,1979, and January 9,1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Cormission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

Accordin9 y, the license is amended by changes to the Technical Spec-1 2.

ifications 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 Amenc.nent No. 47, are hereby incorporated in the license.

The licensee shall operate the facility in L..-. uance with the Technical Specifications.

79020600lG

. 3.

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

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

Attachment:

Changes to the Technical Specifications Date of Issuance:

January 17, 1979

ATTACHMENT TO LICENSE AMENDMENT NO. 47 FACILITY OPERATING LICENSE NO. DPR-33 DOCKET NO. 50-259 Revise Appendix A as follows:

Remove the following pages and replacc with identically numbered pages:

1.

vii/viii 776 113/114 9/To 131/132 T5/16 133/134 T7/T8 139/140 T3/20 157/158 2T722 159/160 23/24 167/168 M/26 T69/170 77/28 173a M/TO 181/182 73/74 218/219 220/221 330'/331 The underlined pages are those being changed; marginal lines on these pages 2.

indicate the revised area. The overleaf page is provided for convenience.

3.

Add the following new page:

172a

LIST OF TABLES (Cont'd)

Page No.

Ti tle Table Hinimum Test and Calibration Frequency for 4.2.F 105 Surveillance Instrumentation Surveillance Requirements for Control 4.2.G 106 Room Isolation Instrumentation Hinimum Test and Calibration Frequency 4.2.H 107 for Flood Protection Instrumentation 108 Seismic Honitoring Instrement Surveillance 4.2.J.

............... 171,172, 172-a MAPLEGR vs Average (Planar Wsure 3.5.I 190 Shock Suppressors Snubbers) 3.6.H Reactor Coolant System Inservice Inspection 4.6.A 209 Schedule....................

250 3.7.A Primary Containment Isolation Valves 3.7.B Testable Penetrations with Double 0-Ring 256 Seals......................

257 Testable Penetrations with Testable Cellows....

3.7.C 258 3.7.D Priraary Containment Testable Isolation Valves...

3.7.E Suppression Chamber Influent Lines Stop-Check 263 Globe Valve Leakage Rates............

3.7.F Check Valves on Suppression Chamber influent 263 Lines 265 3.7.H Testable Electrical Penetrations Radioactive Liquid Waste Sampling and Analysis 287 4.8.A Radioactive Gaseous Waste Sampling and Analysis..

288 4.8.8 324 3.11. A Fire Protection System Hydraulic Requirements...

343 6.3.A Protection Factors for Respirators 360 6.8.A Minimum Shift Crew Requirements..........

vii Amendment No. 47

LIST OF ILLUSTRATIONS Title Page No.

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

26 2.1-2 APRM Flow Bias Scram Vs. Reactor Core Flow 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 Pentaborate Solution Volume Concentration 138 Requirements..................

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

1T3 3.5.2 Kf F ac to r.....................

3.6-1 Minimum Temperature *F Above Change in Transient 188 Temperature...................

3.6-2 Change in Charpy V Transition Temperature Vs.

189 Heutron Exposure................

6.1 -1 TVA Office of Power Organization for Operation 361 of Nuclear Power Plants.............

362 6.1-2 Functional Organization.........

363 6.2-1 Review'and Audit Function.............

364

6. 3-1 In-Plant Fire Program Organization VIII Amendment No. 33,47

1.0 DryINITIDW5 (Cont'd)

10. Logic - A logic is an arrangement of relays, contacts, and other components that produces a decision outout.

(a) Initiatir3 - A logic that receive signals from channels and produces decieion outputs to the actuation logic.

(b) Actuation - A logic that receives signals (either from initiation logic or channals) and produces decision outputs to accomplish a protective action.

W.

Functional Tests - A functional test is the manual operation or initiation of a system, subsysten, or component to verify that it functions within design tolerances (e.g., the nanual start of a cora spray pump to verify that it runs and that it pu:sps the required voluna of water).

X, Shutdown - The reactor is in a shutdown conditica when the reactor anode switch is in the shutdown nods position and no cc,ro alterations are being performed.

Y.

Engineered Saf eguard - An anginearad safeguard is a safety systan the actions of which are essential to a safety action raguited in reopense to accidents.

Z.

Cu:nulative Downtime - The cumlative downtine for those safaty ocmponents and systems whose downtine is limited to 7 consecutive days prior to requiring reactor shutdevn shall be limited to any 7 days in a ocusecutive 30 day pericd.

,- i 7

(

I

SAFETY LU!IT LU!ITING SAFETY SYSTEM SETTE C 1.

7UEL CLADDIt:0 INTEGR1TY 2.1 FUEL CLADDING INTECRITY, Applicability

_ Applicability Applies to the interrelated vari-Applies to trip raettings of the ables Associated with fuel instruments and devices which ace thermal behavior, provided to prevent the reactee systec safety limits f rom being exceeded.

Objective Objectise To establish linits which ensure To define the level of the process the integrity of the fuel clad-variables at which automatic pro-d in g.

tective action is initiated to pre-vcnt the fuel cladding integrity safety linit from ceing exceeded.

Specifications Specification A.

Reactor Pressure > 800 psia The limiting safety systen settings shall and Core Flow > 10% of Rated, be as specified below:

When the reactor pressure is A.

Neutron Flux Scram greater than 800 psia, the existence of a ninitum criti-1.

APRM Flux Scram Trip Setting cal po.er retio (MCPR) less (Run Ifode) thac 1.07 chall constitute violation of the fuel claddir.;

Uhen the Mode Switch it in integrity saf ety li=1t.

the RUN position, the APP.4 flux scram trip setting shall be:

S.1(0.66W + 54%)

where:

S = Setting in percent of rated charmal povar (3293 HWt)

W = Loop recirculatien flow rate in percent of rated (rated loop recirculatica flo s rate equals 34.2x106 lb/hr) 8 Amendment No. 3E, 47

I.131TT113C SAFFTY SYSTD1 SETTitic SAFFTY I.DlIT Pol n)El. C1.ADDTI;C INTml: TTY 1.)

yt t.:1. c a.rgi rir. 1 HTir.it T TY

. In the event of operation with the core maximum fraction of limiting power density (CHFLPD) greater than fraction of rated thermal power (FRP) the setting shall be modified as follows:

SS(0.66W + 54%) FRP CHFLPD For no combination of loop recircu-lation flow rate and core thermal power shall the APRM flux scram trip setting be alle,ed to exceed 120%

1of rated thermal power.

(Note: These settings assume operatLon within the basic thermal hydraulic design criteria. These criteria arc LIIGR g 18.5 kw/f t for 7K7 fuel and6 13.4 kw/ft for 8X8 and 8x8 R fuel, MCPR limits of Spec.3.5.K. Jr/

it is determined that either of these design criteria is being violcted during operation, action shall bc initiated within 15 ninutes to restore operation within prescribed linits.

Surveillance requirements for APE::

scram setpoint are given in specification 4.1.B.

2.

APRM--When the reactor mode switch is in the STARTUP POSITION, the APRM scrsm shall be set at less than or equal to 15% of rated power.

3.

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

B.

APRM Bod Block Trip Settinc

~~

b.

Core Sternal jowe" Limit (Peactqr Pee'sure <803 psia)

The APRI: Bod block trip setting, sh21]

~

be:

'..7 e r. t h 6 reactor pressure is Icss thsn or equal to 800 psia, 9

Amendment No. 3E, /J, 47

I.IHTTTNG SAFETY-SYSTEM SETTIN(;

tAFPfY LIMIT 1.1 FtIEI, CLADDING INTEGRITY 2.1 RTEL CLADDING IrrTEbHITY or core coolant flow is less RB.1 (0.66W + h2%)

~

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

eced 823 HWt (about 25% of U

rated thermal power).

RB = Bad block setting is percent of rated thermal power (3293 HWt) s W

= Loop recirculation flow rate in percent of rated (rateit loop recirculagion flow rate equals 34.2 X 10 lb/hr)

In the event of operation with the core maximum f raction of limiting power density (CHFLPD) greater than f raction of rated thernal pover (FRP) the setting shall be modified as follows:

FRP for two

]

RB -(0.66W +.42%,) O!FL1'D S

i recirculation loop operation.

I RB -(0.66W + 38.7%) FRP for one S

?

CMFLPD recircula 'on loop operation.

C.

Whenever the reactor is in C.

Scram and isolation--> 533 in. above the shutdown condition with reactor lov vater vessel zero le-irradiated fuel in the reac-tor vessel, the vater level shall not be less than 17.7 in. above the top of the D.

Scram--turbine stop f,10 percent normal active fuel zone.

valve elesure valve closure E.

Scran--turbine control valve Upon trip of 1.

Fast closure the fast actia solenoid valve 2.

Loss or control > 550 psig oil pressure F.

Scram--low con-

)

23 inches denser vacuum

, Hg vacuus G.

Scram--nain steam j,10 percent line isolation valve closure

'ain steam isolation > 825 psig H.

d valve closure--nuclear system low pressure 10 Amendment No. 35, 41 SLP 1 V SB y..

1.1 ft AS F.5 :

TUEl. CLADDit:C INTECRl'IY SAFETY LIMIT The fuel cladding represents one of the physical barriers which separate radio-active materials from environs. The integrity of this cladding barrier is related to its relativa freedom from perforations or cracking. Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.

Fuel cladding perforations, however, can result from thermal stresses which occur f rom reactor operation significantly above design conditions and the protection system setpoints. While fission product migration from cladding perf ormation is just as measurable as that f rom use-related cracking, the thermally-caused cladding perforations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deteriora-tion. Therefore, the fuel cladding safety limit is defined in terms of the reacter 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 Safety LLuit is defined with margin to the conditions which would produce onset transition boiling (MCPP of 1.0).

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

MCPR >l.07 tepresents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

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

Since boiling transition is not a directly observable parameter, the margin to ootling 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 characterized 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 oundle power. The minicum value of this ratio for any bundle in the core is the minimum critical power ratio 0 CPR).

It is assumed that the plant operation is controlled to the nor.inal protective setpoints via the instru-mented variables, i.e., no rmal plant operation presented on Figure 2.1.1 by the ne-f nal ex aee ted flew control lire.

rne Safetv L1= i t (MCPR of 1.07) h*= auf Ficient conservatism to assure that in the event of an abnormal operational transient initiatef f rom a normal operating condition (MCPR > limits specified in specification 3 5.K)more than 99 9% of the fuel rod s in t ha core are expected to avoid boiling transition. The margin between F.CP3 of 1.0 (onset of transition boiling) and the safety limit 1.07 ___ i s d e riv ed f ron a detailed statistical analysis considering all of the uncertainties in moni-toring the core operating state including uncertainty in the boiling transition correl 4*. 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

.1.1 BASES _

f Decause the boiling transition correlation is based on a large quant.ity o tion of a fuel (bli scale data there is a very high confidence that opera iling tran-assembly at the condition of MCPR =1.07, vould not produ additional margin exists between the safety limit and the actual occurence sition.

of loss of cladding integrity.

However, if boiling transition vere to occur, clad perforation vould not Cladding temperatures vould increase to approximately 1100 F vhich is below the perforation temperature of the cl be expected.

Reactor (CETR) where fuel similar in design to BFNP op material.

)

without clad perforation.

If reactor pressure should ever exceed 1400 psia during notsal power operating (the limit of applicability of the boiling transition has been violated.

In addition to the boiling transition limit (MCPR = 1.o6) operation is constrained to a maximum Ll!GR of 18.5 kv/ft for 7x7 fuel and 13.

This limit is reached when the Core Maximum Fraction of for 8x8 fuel.

For the case where Core Limiting Power Density equals 1.0 (CNFLPD = 1.0).

Maximun Fraction of Limiting Power Density exceeds the Fraction of Rated Thernal Power, operation is permitted only at less than 100% of rated power and only with reduced APRM scram settings as required by specificat At pressures belov 800 psia, the core elevation pressure drpp (0 power, 2.1.A.1.

O flow) is greater than 4.56 psi.

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

pressure drop in the bypass region is essentially all elevation head, the core pressure drop at lov powers and flow will alvays be gr than 4.56 psi.

flov, bundle pressure drop is nearly independent of bundle power and has the bundle flov vith a k.56 psi driving head a value of 3.5 psi. Thus,3 Full scale ATLAS test data taken lbs/hr.

will be greater than 28x10 at pressures from 14.7 psia to 800 psia indicate that the fuel assembly With the design critical power at this flow is approximately 3.35 MWt.

peaking factors this corresponds to a core thermal power of more than Thus, a core themal power limit of 25% for reactor pressures 50%.

belov 800 psia is conservative.

For the fuel in the core during periods when the reactor is shut down, con-sideration must also be given to vater level requirements due to the effect If water level should drop below the top of the fuel during This reduction in or decay heat.

this time, the ability to remove decay heat is reduced.

cooling capability could lead to elevated cladding temperatures and clad As long as the fuel remains covered with water, sufficient perforation.

cooling is available to prevent fuel clad perforation.

16 Amendment No. E, 47

1.1

,B A1F;S The saf ety limit has been established at 17.7 in. above the top of the irradiated fuel to provide a point which can be monitored and also pro-vide adequate margin. This point corresponds apprcximately to the top of the actual fuel assemblies and also to the lower reactor low water level trip (378" above vessel sero).

RJ.TF.RT.HCE General Electric BL*R Thermal Analysis Basis (CETA3) Data, Correlation 1.

and Design Application, NEDO 10958 and NT.DE 10958.

2 GE BWR Reload 2 Licensing Amendment for BFNP unit 1 reload 2, NEDo-2h136, August 1978 and Revision 1 dated November 1978.

Amendment No. 35, 47 17

9

(

PACE DELETED

/

\\.

q se JAN t o E' I_

LIMITING S AITTY SYSTEM sETTINCs RELATED TO T1/EL C1. ADDING INTT.CRITY 2.1 BASE 3:

The abnorual operational transients applicable to operation of the Browns Ferry Nuc1 car Plant have been analyzed throughout the spectrum of planned operating con-ditions up to the design thermal pover condition of 3440 FNt.

The analyses were based upon plant operation in accordance with the operat Lng map given in Figure 3.7-1 of tne FSAR.

In addition, 3293 int is the licensed maximum power level of Browns Ferry Nuclear Plant, and this represents the maximum steady-state power which shall not knowingly be e.xceeded.

is incorporated in the transient analyses in estimating the Conservatism control rod scram controlling f actors. such as void reactivity coef ficient, These peaking f actors, and axial power shapes.

worth, scram delay time, factors are selected conservatively with respect to their effect on the results as determined by the currant analysis model.

~

applicable transientmodel, evolved over many years, has been substantiated in opera-This transient tion as a conservative tool for evaluating reactor dynamic performance.

Results obtained f rom a Cancral Electric boiling vater reactor have been The comparisions and resulto compared with predictions made by the r.odel.

are summarizwd in Reference 1.

used in the analysis The absolute value of the void reactivity coefficient 25% greater than the no..inal maximus is conservatively estimated to be about The scras worth used haa value expec:ed tu occur during the core lifetine.to approximately 80t of the total scram been derated to be equivalentThe scram delay tiue and rate of rod insertion allowed the control rods.are conservatively act equal to the longest de, lay,ar.d slow-s.i., - Svses insertion rate acceptable by T4chnical Specifications.

est The eff ect of scrsm worth, scram delay time and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion. The rcpid insertion of negative reactivity is assured by the time requirements for 5 and 20Z insertion.

time the rods are 60% inserted, approximately four dollara of negative teac-By the tivity has been inserted which strongly turns the transient, and accomplishes'the desired effect. The ti=es for 507. and 90% insertion are given to assure proper cespletion of the expected performance in the earlice portion of the transient, and to establish the ultimate fully shutdown steady-state condition.

For aoslyses of the thermal consequences of the transients a HCPR) limits specified in specification 3.5.K is conservatively assumed to exist prior to initiation of the transients.

Th'.s chof re of using conservative values of controllLng parameters and initiscing the design power level. produces more pessimistic ansvers than transi nts at would result by using expected values of control par meters and analyzing at higher power levels.

Steady-stata operation without forced recirculation will not be permitted for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. and the start of a recirculation puno from the natural circulation condition will not be permitted unless the tennerature difference between the loop to be started and the core coolant temperaturc is less than 750F.

This reduces the positive reactivity insertion to an acceptably low value.

19 Amendment No. 3E, 47

2.1 EAlE" In suasary.

The licensed naximum power level is 3,293 Hut.

1.

Analyses of trancients employ adequately conservative values of the 2.

controlling reactor para eters.

f 3440 HWT.

The abnormal operational transients were analyzed to a power level o 3.

is a core logical answer than The analytical procedures cov used t esult j e the alternative cethod o' asguming a higher starting power is con us -

4.

f or the parameters.

tion with the expected values The bases for individual set points are discussed below:

Neutron Fluz Scren

/..

APRM P.igh ritx Stras Trip Setting (Run Mode)

The average pcuer range conitoring (app.M) systeo, which is calibrated 1.

balance date taken duriog stesdy-state conditions, reads f iss ion chambers pro-using heat Because of reted power (3,293 MWt).

responds directly to la percent s i gnals, the A?RM syste:

During transients the instantaneous rate of vide the basic input is less than the everage neutron flux.transf er f rom the f uel (recctor thermal power) of the fuel.

heat instantaneous neutron flux due to the tim

• conste.nt the the rmal

!=duced by dicturbancca, Therefore, during trenstents icdicated by the neutron flux than that power of the fuel vill be lessAnalyses reported in Sectica 14 of the Pisal demonst rated that vi:h a 120 perceo: scras trip at the serts setting.

l Safety Analysie Repor:

eetting, none of the absornsi opers:1onal traceicnts analyred vio ate limit a:4 there is a substan:is! cargiu f rca f uci Therefore, use of a flow-biased scras provides even addi:icasi the fuel safe:/

damage.

Figure 2.1.2 shows the flow bissed scr.a as a function of D4rKin-core flow.

the margia pre-An increase in the A?Rlt scra: setting would decresse is re. sche 1.

The sent before the fuel cicddin: iuregrity asic:y 11=1:

APRM scris set ting was deteruined by en analysia of sargins :: quired to provide e reasonable range for =aseeverict daring operation.

the frequency of spurious Reducing this operatin; =arcis sould increasecn reac:or saf ety because of the effec:

acrams, which have on adverseThus, the APRM ae::ing vos eclected resulting ther al stressee.

provides adequa:e margin for the fuel cleddin; in:egrity bacsuas it saf ety limit ye:

unnecessary acrasa.

20

\\

jhh

2 ;l C AE F.5 The scram trip setting must be adjusted 'to ensure that the LHCR transient The scram peak is not increased for any combination of CMFLPD and FRP.

setting is adjusted in accordance with,the formula in specification 2.1.A.1 when the CMFLPD exceeds FRP.

Analyses of the lloting transients show that no scras adjustment is required to assure MCPR > 1.07 when the transient is initiated f rom MCPR > limits specified in specification 3 5.K.

2.

APRM Flux Scram Trip Setting (Refuel or Start & Hot Standby Mode)

Tor cperation in the startup mode while the reactor is at low pressure, the APhM screm setting of'15 percent of ra:ed power provides adeuuste thernal cargin between the setpoint and the cafety linic, 25 percent of rated.

The margin is edeouste to accomr.odate anticipated maneuvera associated with power plant startup.

Effects of increasing pressure at zero or low void content are mir.or, cold water f ron sourcer avail-able durir.g startup is not much colder than that already in the system, temperature coefficients are scall, and centrol rod patterns are con-strained to be ur.iiorm by operating procedurce backed up by the ro'd vorr.h ninioteer ar.d the Rod Sequence Control Syste:. Vorth of ir.divt-dosi rods is very low in a uniform rod p ttern.

Thus, all of posai~ ole sources of reactivity input, unifor= control rod withdrawal is the noot probable cause of significant power rice.

Because the flux distribution associated with uaiform rod withdrawale does not involve high local peans, and because several rods nuet be r<oved to change power b7 a significant percentage of rated power. the rate of power rise is very slow. Generally, the hest flux is in r. tar equilibrius with the 'ficsion rate.

In an ceau=ed unifor= rod withdrawal approach to the scram level, the rate of power rise to no more t'.an 5 percent of rated pouer per minute, and the APRM sentes vould be core than adequate to assure a scras befere the power could exceed the safet, limit. The 15 percent AF?M ucram reszins active until the cade switca is placed in the RUN position. This evitch occurs when reactor pressure is greater than 850 peig.

3.

IRM Flux Sci an Trip Setting the IRM Sys tem consists of S chambers, 4 in each of the reactor ; rot.?c-t i.in system logie channels. Th e IMI is a 5-decade ins t ruoent which covers the range of powet 1-rul between that ec've red by t ne 51't and the Ar!ti.

T.m e 5 decades are revered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in siae.

The IRM scras setting of 120 divisions is active in each ran3e of the IRM.

For 21 Amendment No. 3E, 47

i

.1 BASES 3.

IRM Tlum Sern Trip Setting (Continued) if the instrument were on range 1, the scram setting would be at 120

example, divisions for that range; likewise, if the instrument was on range 5 the scram setting would be 120 divisions on that range. Thus, as the IRM is ranged up toA accommodate the increase in power level, the scram setting is also ranged up.

120 divisions on the IRM instruments remains in ef fece as long as the scran at reactor is in the startup mode.

In addition, the APRM 15% scram prevents higher power operation without being in the RUN mode. The IRM scram provides The prctectinn for changes which occur both locally and over the entire core, significant sources of reactivity change during the power increase are most dua to control rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slou enough due to the physical limitation of withdrawing control rods, that heat flux is in equilibrium with the neutron flux and an IRM scram would result in a reactor shutdown well before any safety limit is exceeded. For the case of a single control rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analyris included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just suberitical and the IRM system is not yet on scale. This condition exists at quarter rod density. Quarter rod density is illustrated in paragraph 7.5.5 of the FS AR.

Additional conservatism was taken in this analysis by sssuming that the IRM channel closest to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power limited to one percent of rated power, thus maintaining MCPR above 1. 07.

Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence.

B. APRM Control Rod Block Reactor power level may be varied by coving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to brevent rod withdrawal beyond a given point at constant recir-cuclation flow rate, and thus to protect against the condition of a HCPR less than 1.07 This rod block trip set ting, which is automatically varried with recirculation loop flow rite, prevents an increase in the reactor power level to excess values due to control rod with-d r awal. The flow variable trip setting provides substantial margin 22 Amendment No. 33, 47

e 2.8 jgg, free feel damage, assuming a steady-state operation at the trip setting, over the eettre rectrentation flow range. The margin to the safety Limit increases se the flew decreases for the specified trip setting versus flow relationship; thovefore. the worst case McFR which could occur during steady-state operation is The at leat of wated thernal power because of the AFM rod block trip setting.

ectual power distribution in the core is established by specified control rod sequences As virh ths. APM scram and is vanitored continuously by the in-core 1.FM system.

arty setting, abe AFM rod block trip setting is adjusted downward if s'he

~

OFLPD exceeds FRP thus preserving the AFict rod block safety esargia.

C. Reactor Water 1.ov 1.avel scram and Isolation (Tateept Main steamlines)_

The set point for the low level scra:s is above the bottom of the separator skirt.

This level has been used in transient analyses dealing with coolant inventory The results reported in FsAR subsection 16.5 sbev that scram and isoistion decrease.

of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because McFR is greater than 1.07 in all esses, and The scram setting is systes pressuta does not reach the safety valve settings.

anprestsstely 31 inches below the normal operating range and is thus adequate to avoid spurf ous strass.

t

9. Turbine Ftop Valve Closure scram The turbine stop valve closure trip anticipates the press 9re, neutron flux and heat flux increases that would result from closure of the stop valves.

With a trip setting of 10% o' valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst case transie.it that assumes the turbine bypass valves y,emain closed. (Reference 2)

8. ?stbtme centrol falve scess 3.

rest closere scra.

Yhis turbine control valve fast' closure scram anticipates the pressure.

neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection coincident with failures of the turbine bypass valves. The Reactor Protection System initiates a scram when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than 30 milliseconds af ter the start of control valve fast closure. This is achieved by the action of the fast acting solenoid valves in rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc dump valves.

This loss of pressure is sensed by pressure switches whose contacts form the one-cut-of-two-twice logic input to the reactor protection system.

This trip settirg, a nominally 50% greater closure time and a different valve characteristic from that of the turbine stcp valve, combine to No signiff-produce transients very similar to that for the stop valve.

Relevant transient analyses are discussed cant change in HCPR occurs.

Thic scram in References 2 and 3 of the Final Safety Analysis Report.

is bypassed when turbine steam flow is below 30% of rated. cs sneasured by' turbine first stata pressure.

Amendment No.

35, 47 23

2,1 RASES 2.

Scram on loss of control oil pressure The turbius hydraulic control system operates using high pressure oil. There are several points in this oil system where a loss of oil pressure could result in a fast closure of the turbine control valves.

This fast closure of the turbine control valves is not protected by the generator load rejection scram, aines failure of the oil system would not result in the fast closure solenoid valves being actuated. For a turbine control valve fast closure, the core would be protected by the APRH and high reactor pressure scrams.

However, to provide the same margins as provided for the generator load rejection scram on fast closure of the turbine control valves. a scram has been added to the reactor protection system, which senses f ailure of control oil pressure to the tur-bine control system. This is an anticipatory scram and results in reactor shutdown before any arignificant increase in pressure or neutron flux occurs. The transient response is very similar to that resulting f rom the generator load rejection.

F.

Main Condenser Lov Vacuum Scram To protect the main condenser ab. inst overpressure, a loss of con-denser vacuum initiates automatic closure of the turbine stop valves and turbine bypass valves. To anticipate the transient and automatic scram resulting from the closure of the turbine stop valves, low con-denser vacuum inii:iates a scram. The lov vacuum scram set point is selected to initiate a scram befc..e the closure of the turbine stop f

valves is initiated.

(

l C. & H.

Main Stear. Line Isu etion on Low Pressure and Main Steam Line Isolation Scram l

The low pressure isolation of the main steam lines at 825 peig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage is taken of the scram feature that occurs when the main steam line isolation valves are closed, to provide for reactor shutdovu so that high power opera-tion at low reactor proosure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reac-ter at pressures lower than 825 poig requires that the reactor sede switch be in the STARTUP position, where protection of the fuel cladding integrity safety limit is provided by the IRM and APRM high neutron flux Thus, the combination of main steam if w low pressure isolation scrams.

and isolation valve closure scram assures the.

~ility of neutron flux scram protection over the entire range of ap,

. ability of the fuel cladding integrity safety limit.

In addition. the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure. With the scrams set at 10 percent of valvo closure, neucron flux does not increase.

21s

e 2.1 3A}Ey

,1.

J. 6 K.

Reactor 18:W Vater level set point f or initiat ion of HPCI and RCIC. closing main steam isolation valves. and starting LFCI and core apray putapn.

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

The design of these systems to adequately perform the intended func-tion is based on the specified low level scram set p o ir.t and initia-tion set points. Transient analyces reported in Section 14 of the FS AR demonstrete that these conditions result in adequate safety margins f or both tr.e fuel and the systes pressure.

L.

Re f er enc e_s_

Linford. R.

B., " Analytical Hethods of Plant Transient Evaluations f or l.

the General Electric Boiling Water Reactor," hTDO-10802 Feb.,1973, t

GE EWR Reload 2 Licensing Amendment for BFNP unit I reload 2, 2,.

NEDo-2L136, August 1978 and Revision 1 dated November 1978.

Amendment No. 35, 47

?.

'l -,

- - - - -- - r 8 r.[i...

'I,.,i.l'illl. l.

Ji.'.'i I :

.l."

3.,l.a t,l:a 8'

I

' l,. >

.4 i

i,

.,i.. i.';.

.. i a,.,, -,i ;,. i u

e.,.

i

,ie p.

i

-i

,d..

i,.

p.,

.l.

-.: l.i

,. I e

I,.

l.

a p'..

. a e e i

l

.i :

1.4,.

.a t

.... g g '

.,..-, ;..- -r

. a

._-.4 l,

I

.r, i.,

.g

. i.. o.

l i -

.n

. ' i...,

...., j.

l

,l,u,

..n.

e o

j

,i; p ;

ee '.

120.

-'."~'" *- I ' i '

-J-~ h qg

- T-'. q e,,g

~

~ ' ' '

i : l' l :

i

j

..l. _ :

2.

.c

..a

,i! ' ;e' I..:.[....* ;; l

,3.., I.-

- g.sAFRM TIUJ BIAS SCRAM de l?

, :i

.g.

y:;,b;, :.' ' " l ' * ' y. l "l-

.T' -- llot i

.:.I.

.it.lf:i.sci.

f,...:a :

. i

.. '., l,.

li'l,q 1:.

..g., o

',e

. j :}q;/, y ' ', ':..l:

a:

8r e

r

,l" 9,',i-p i, i

100'

~

,i b,l: 'B..

h o'9-T F[.;;lf.'. u.-

4 ' % ai

'j

.1;. ai.iii

.,..l

. 'i - l i:rs.,>;, l' ri(!!;l, p"

d..

~ i..

", '..'i. _I..j_.u.-

I

.up"r 'j;' :i..I,:

i t

ci #

ei l

nl!

I i.

i. ; -

I i

~

.c-<

o....

J.

i" i ; ",j.o ', ;l il ' ; ;, '.. :..;' ;'h.

I r.ip;9mj,_N:

.'..i,

..:u;~~h:M"y49 i

p, n:

i,

m 1.

.i. g3 c':.

"i,

_t.

t q; ;a ij ~ l.'7, b, !!.',, a :

. p,F -gfijf'.

~ i:.

d e,, p 4

.i g.

' :s pl. G 't i

'4./.li c.'Ly

.h-.

LD l

8 SIGN FIC/ CO?i" ?OL_Ty. E.

E_.I-] q3'_.][

m

. i.!

,,i j i.

u

. g l T" '.

a,..

_ :e.

ig, 9

. l,,'l.. _-4

,;' l'i.

pl I l, rp 7

d. ;--

70

..si.it,.I

, 'c a.+ r:,':

!I. T.. !v' l-~t.. :

T~

i i.,'

~

~l.li-

.'.;l :

l' _.1.,5

~

1-o' m

c

'ia

- ylo

.k,

.I i 1. a,. i,

i:

b p'

ai,.:

,1 r :r:,.

w 60

gi;

- :!@ i

p"r !,i,1. l.;;

P:iq t 4 t, l.'

a --l-,.;.:..... j :.. /-

q,.t.T.. [.

.i. i

'i

/,.

i: l: # - ) 4.1.d G W h y ; H

...q, H o. i., n

.n.. i h h- "

~,~

L,-

50,,

u.

- M-r--

,,,o,,

.k,,j,,o,

.,g

-- tl. u.' '. _ ' ;, g' ' ' r

'. ;.., o.

.ld..i:3. x,.w;;.fF

,4,'i..

i

"' E i

I j':.

i

I j [1

.i; i j. l

. 'c, ij.,.. "!j 3,n p Qj,...

lT i

i

' g '.

ir e.,

,.. -..L1;

-l a I NARTRAL CIRCUIA"' ION.:j.'

N.':, ll1 '.i ; @i.. ;7

". ! l.,,:

....i.

, i e

..i

1.

c. o,

.L;.

- lite h,p.c pl.jy j,' U.l.E i; ll..

i.

y'.;.

.., J a.

.,A,...

.. _t o4 ii u

i.:

-, p j..

, it.i i

_=a -. o f il p-

,..l O

>l 3 i,i.

.i i S.,

? eI p..

i: w g;4,n,n

,, 9' i i

i' i

l. Inl i ::;

FUMD...Sr: o LINE ' W. I:! :...

y-c I r ai,

~ " "

e.m.N i '.

.;H 6

u.

20 n::l'i.: I I

d.,"

" Il,'. ! ? ' ~

l-p

' d.

y'

'.;. o, yp

'~

..,c

... j h.

..j[ p...

i..

..I

,i p,-

2,

...e n I a -

.::r.

n

..a..,e 10

'oi

.ep

-g.

i*

i:.

i;ig', ;', hj i y;lir, l..h.;.i.3p:!- k';.

.L.i.

.. : r y

,y, i _. M.D, I ~ l% j;i",,f,gi4;Inijd q'

c I' '

pTy;if6r.

9, i

w.

s~Ts~'

':r-'j l :.p ",t0'0$tiof 20h' 30fl' 40[

t 50l :j 60@ 7c,}d80".pso. 6100%llo} 120]"g 4

~

i d :iP t P !a.-

l-iw i.l.n.,.. q. I i

.i,o :, i..:., l a

. i. _.:

gi.,

.lJ CO.RI. C. 3,00IA. T,,:'..F. IC.,J R..A..'E @ OF I:ESIG l '!T.@.;..'

I i

e i l

,r.-

l

.;. ur 3 r-

.i....,,p i

.. i APRM FIM BIAS SCRAM Va. REACU)R CORE FIEJ l

~

~

FIG. 2.1 -2

- - -,l

.y l

~~. v -,.

4 -.e. -. -

p 8 il

-mt:. I l

.,j

=

.I~s-

.I 4al

...d."t.i I.;=

r l

.g 4.p.

-l -

t-i L,y.

i.

..: s b.'-

. I_

,da f*

} 1.., l g..._

1:.

+l e..

... '..z.

([_ '

. j g

_ _ _ _ _.)

...: 1o w:

s !PY LBuT

,, s.m T i ngq S A M.'f Y S Y!;T U 4 S M i d,,,,

r 4.2 REACTOR COOLANT SYSTD_4iNTECRIT_Y 2.2 REACTOR COOL. ANT SYSTD4 INTECRiTY Appiscability Applicability A, plies to limits on reactor coolant Applies to trip settings of the s;Jtem pressure instruments and devices which are provided to prevent the reactor system safety limits from being exceeded.

S.kjectIve Obge,t I,ve, fu establish a limit belev which To define the level of the process the integrity of the reactor coolant variables at which automatic pro-system is not threatened due tn en tective action is initiated to overpiessure condition.

prevent the pressure safety limit from being exceede'd.

Specifiestien Specification A.

The pressure at the lowest point The limiting safety syst.m settings of the reactor vessel shall not shall be as specified below:

exceed 1,375 psig whenever irradiated fuel is in the reac-Limitins Safety for vessel.

Protective Action System Settint i

l l

A.

Nucicar system 1105 pois +

relief valves 11 psi (4 open--nuclear valves )

system pressure 1115 Pais i 11 psi ( 4 valves) 1125 ps18 1 11 poi ( 5 valves) 6.

Scram--nuclear J 1.053 pois system high pressure Amendment No. 35, 47

4.3,,saj.g.

Eg ACTOR C00LtNy SYSTD1 INT &'CRI]

Jae safety limits for the reactor coolant system pressure have been selected

> scle ther they are below presource at which it can be shown that the integrity of the s y :. t.u is not endungered. Howe ver, t he pressure safety limits are r et high enough much that no foreseeable circumstantes can cause the systeu pressure to rise over these limits. The pressure safety limits are arbitrarily selected to be the lowest transient overpressures elleved by the applicable modes A5ME Roller and Pressure Yessel Code, Section 111. and USAS Firing wie. Section 331.1.

64e design pressure (1,250 pois) of the reactor vessel is established such that, when the 10 percent allowance (125 psi) allowed by the ASHE Boiler and Freneure Vessel Code Section !!! for preseure transients is added t.i the Jestga preer.re, a transient pressure limit of 1,375 peig is establishej.

CorresponJingly, the design pressure (1,1 8 psig for auction and 1,326 pois for discharge) of the reacter recirculation systec piping are s.:ch t hat, when the 20 percent allovence (230 and 265 p*') allowed by USAS Fiping Lode.

Section 531.1 for pressure transiente are added to the design pressures, t ransient yt esoure Itaire of 1,378 and 1.591 pain are established. Thus, the pressure esfety limit applicable to power operetten is established at 1.375 pois (the lowest transient overpressure allowed by the pertinent coJes),

ASHE Soller and Pressure Vessel Code, Section 111. sad USAS Piptog Code, Section B31.1.

The current cycle's safe 6y annlysis concerning the most severe abnermal operational transient ressulting directly in a reactor coolant system f

pressure increase is given in Refereaee 5.

The reactor vessel pressure code limit of 1.375 psig given in subsection 4.2 of the safety analysis report is well above the peak pressure produced by the averpresoure transient described above. Thus, the pressure safety limit applicable to power operation is well above the peak pressure that can result due to reasonably expected overpressure trengients, liigher design pressuree have been entablished for piping within the reactor coolant systers than f or the reactor vessel. These increased design pressures create a consistent design which savures that, if the pressure within the reactor vessel does not suceed 1,375 psig, the pressures within the piping annot escoed their respective transient pressure limits due to static and pump heads.

1he esfety limit of 1,375 pois secually applies to any point in the reactor vessel; however, because of the st atic cate'r heaJ, the highest pressure point vill occur at the bottom of the veneel.

Because the pressure is no; nonitorea at thie point. it cannot be directly determined if thte safety lit.t has been violated.

Also, becsuse of the potentially varying heaJ 1evel and flow pres-eure drope, an equivalent pressure cannot be a priori determined for a 28 Amendment No.

/E, 47

1.2 BASES pressure monitor higher in the vessel. Therefore,.~oiiovin; a..y trancient that is severe enough to cause concern that this safety limit was viointed, a calculation will be performed using all avaiiable in"cr:uttior. to 6ter-eine if the safety limit was violated.

RET 7.RINCES 1.

Plant Safety linslysis (B U P TSAR Section IL.'J) 2

!.S:E 3 oiler and Pressure Vessel Code Secti:n III 3.

USAS Pipir.c Code, Secticn D31.1 L.

itesctor '.' esel and Appurter.ances !!echenicol ;csica (F.7 F3R Subsectica L.2) 5.

GE BWR Reload 2 Licensing Amendment for BFNP unit 1 reload 2 NEDO-2hl36, August 1978 and Revision 1 dated November 1978.

)

I i

29 Amendment No.

35, 47

2.2 BASES REACTOR COOLA!.'T SYSTUT IfffwRITY To meet the safety design basis, thirteen relief valves have been installed on the unit with a total capacity of G2'.4% of nuclear boiler rated stean flow. The analysis of the vorst overpressure transient, (3-second closure cf all main stenn line isolation valves) neglecting the direct scran (valve position scram) results in a maximum vessel if a neutron flux scram is assumed considering pressure of 1273 psi 6 12 valves operable. This results in an 102 psig margin to the code allovable overpressure limit of 1375 psic.

To meet the operational design, the an (lysis of the plant isolation transient (turbine trip with bypass va've failure to open) assuming This a turbine trip scram is presented in reference 5 on page 29 analysis shows that 12 of the 13 relief valves limit pressure in the stean line to 1201 psig. This analysisshows that peak system pressure is limited to 1229 psig which is 1k6 psig below the allowed vessel over-pressure of /375 psig.

3 Amendment No. 3E, 47

TAtt.E 3.2.C INSTRUNDITATION THAT lt!ITIATES R3D BIACK$

a al.t x..

Operabis Per Trte 1.evel Settier, Trty sys (5)_

tonetton y

2(1)

AFRM Upscale (Flow Blas)

• 0.66u + 422 (2)

< 12I 2(1)

APEH Upecale (Startup Mode) (6) 2(1)

APRM Downscale (9) 1 31 (10 )

2(1)

APRM Inoperative b

1(7)

RBM Upscale (Plow Blas)

< 0.66W + @ (2) 1(7)

RBM Dovnocale (9) 1 31 (10 )

1(7)

RBM Inoperative

_ 108/125 of full scale 3(1)

IRM Upscale (8)

~w 3(1)

IRH oovascale (3)(8)

L 5/125 or rull ocale 3(1)

IRM Detector not in Startup Position (8)

(11)

(10*)

3(1)

IRH Inoperative (8) t 1 x 10' counts / cec.

2(1)(6)

ERM Upseate (8) 2(1)(4)

SRM Downscale (4)(8)

L 3 counts /sec.

2(1)(6)

S2M Detector not in Startup Position (4)(8)

(11) 2(1)(6)

SRM Inoperative (8)

(10s)

<10% difference in recirculation flows 2(1)

Flow Dies Cee::arator

< 1101 recirculation flov 2(1)

Flov Blas Upacale N/A 1(1)

Rod Block !.oete P(1 RSCS Hestraint 1187 psig turbine (PS-85-61A &

rirst stsee pressure (8PProx:-stely 301 power)

PS-85-618)

.T!?.!!.1. L^."1!.. hL L 1.

For the startup and run poettiona of the keector Mode Nelector Switch, there shall.

• tuo operable nr tripped trip systeae f or escle functinn.

The SRM, IM

.ad APM (Startup pode ),

blocks need not be operable in

• 'A un" oe d e, and the APR!t (rlow biased) and RBM rod blocks need not be operable in "Startup" code.

If the first column cannot be sost for one of the-tw trip sycta=s. this condition mey estet for up to seven days provided'that dur!nt, t ha t tt'io the operabic system is functionally tested i'wsediately and daily thereafter; if this cendition last longer then seven days, the system with the inoperable ch.erinal shall be tripped.

If the first column canno; be met,for both trip systems, both trip systerne sh.all be trippod.

2.

k'is the recirculation loop flow in ptrcent of desfr.n. Trip level setting is in percent of rated pn cr 02n1::Wt). A ratio of TRP/Curi.PD <l.0 in permitted at reducc3 power.

See Specifit'ation 2.1 for APPJi control rod block setpoint.

3 IM down s eale is trfps s s s<1 when it is en its icwe s t rang e.

4.

This function is bypassed vv.en the count rate le 1100 cps and IM abays ranAs 2.

5.

one instrureent channel:

1.e., one APM or IM or RFM, per trip systsu may be bypassed exc

,,c cmly one of four SP.M cay be bypassed.

6.

IRM channels A, E, C, C all in range 8 bypasses SM channele A & C functions.

tRM channels B, T D, H all in rango 8 bypasses SM channels 3 & D functions.

7.

The P.11cvinC cr+ rational rest rsint; apply to the RhM only:

&th RE'. : har.nels are bypassed when reactor power is 3 30%.

a.

b.

he FB" need nct be operable in the "startur" position of the reacter mode selector switch.

I Two RS4 channels are provided and only one of these may be c.

bypassed frce. the consc]e. An RE" channel may be out of service for testing and /cr maintenrance prc,vided this condition does not last longer than 2L hours in ar.y thirty day period.

d.

If minicum conditions for Tab 3c 3.2.c are not met, administrative controls shell be in=ediately imposed to prevent control rod vftkdrawal.

7L

3.2 BASES The HPCI hip.h flow and temperature instrumentattun are provided to detect a break in the HPCI steam piping. Tripping of this instrus.entation re-sults in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1 out of 2 logic, and all sensors are required to be operable, liigh temperature in the vicinity of the HPCI equipment is sensed by 4 sets of 4 bimetallic temperature switches. The 16 teaperature switches are arranged in 2 trip systems with 8 temperature switches in each trip system.

The HPCI trip settings of 90 psi for high flow er.d 20C*T f o-high tem-perature are such that core uncovery is prevented and fission product release is within limits.

The RCIC high flow and temperature instrumentation are arranged the samt as that for the HPC1. The trip setting of 450" H 0 for high flow and y

200*T for temperature are based on the same criteris as the HPCI.

Itigh tenperature at the Reactor Cleanuo Systee floor drain could indicate a break in the cleanup system. When high temperature occurs, the cleanup system is isolated.

The inswrumentation which initiates CSCS action is arranced in a dual bus system.

As f or other vital inst rumentation arranged in this f ashion, the Specification preserves the ef f ectiveness of the systen even during periods when natntenance or testing is beinr. perforned. An execption to this is when logic functional testine, is being performed.

The control rod block functlons are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to 1.07. The trip logic for this function is 1 out of n:

e.g., any t rip on one of six APRH's, eight IRM's, or four SRM's will result in a rod block.

The minimum instr, ment channel requirenents assure sufficient instrumenta-tion to assure the cinP.le failure criteria is met.

The minimum instrument channel requirements for the RBM may be reduced ~oy one for maintenance, testing, or es11bration. This time period is only 3% of the operating time fu o nonth and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.

The APRM rod block function is flow b? sed and prevents a significant reduc-tion in HCPR, especially during oper tion at reduced flow. The APAM pro-vides gross core protection; i.e.,

1 aits the gross core power increase f rom withdrawal of control rods in the normal witherawal sequence. The trips are set so that MCPR is maintained greater than 1.0 7.

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

the prevention of critical power in a local region of the core, for a single r6' withdrawal error from a limiting control rod pattern.

113 Amendment No. AA, 47 I

e

no-w 3.2 sAsrs If the IRH channels are in the vorst condition of allowed bypn=n, the sealing arrangenent is such that for unbypassed IF.M channels, n rod blni.L signal is generated before the detected neutrons flux has incicaced by more than a factor of 10.

A downncale indication is an indication the instrenent has failed or the instrument is nnt sensitive enough.

In either case the inctnment will not respond to changes in control rod uotion and thus, contrcl rod motion is prevented.

The refueling interlocks also operate one logic channel, and r.re requir.d for safety only when the mode switch is in the refueling po.-ition.

For effective energency core ecoling for sna11 pipe breaks, the HPCI r.ytt. s must function since reactor pressure does not decreace rapi?. enough to allow either core spray or LPCI to operate in tine.

The autu. e. tic pr.ra. ire relief function is provided as a backup to ine lipC I in the event.the ll:J' does not operate. The arr.ingenent of the tripping contacts in euch t-t, provide this function when necessary and mininize spurious operation.

The trip settings given in the specification are ader,uate to e s.:ve t he cbr.. c criteria are net.

1h: specification preserver the ef f ec tive.cc.c of the:

systen during periods of naintenance, testing, or celibration, and elco minimizes the risk of inadvertent operation; i.e.,

only one instrur. cat channel out of service.

Two post treatment of f-gas radiecion monitors are provided and, when their trip point is reached, cause en isolation of the eff-gas line.

Inolatien is initiated when both instrumer ts.each their high trip point or one han an upscale trip and the other a dovuscale trip or both have a dounscalc I

trip..

Both instruments are required for trip but the instruments are set so that any instruments are set so that the instantaaeous stack rcledae rete limit given in Specification 3.8 is not exceeded.

Tour radiation iaonitors are proviile.1 for each. unit shich init1cte prionry Containment Isolation (Group 6 isolation valves) Reactor Building Isolation and operation of the Standby Gas Treatment System.

These inctrument chn t:n e l s monitor the radiation in the Reactor zone ventilation exhaust ducto end in the Refueling Zone.

Trip setting of 100 mr/hr for the monitors in the Refueling Zcne are ba:cd initiating normal ventilation isolation and SCTS operstien so that upor none of the activity released during the refueling accident 1 caves the Reactor Building via the normal ventilation path but rather all ths activity is processed by the SCTS.

Flev integrators and sump fill rate and punp out rate tieces are used to determine leakage it ne dryvell. A systen whereby the ties interval to fill a known volume will be utilized to provide a backup.

An air sanpling systen is alco provided to detect Icakane incid: the prisnry cent:in::nc (See Table 3.2.E).

!!4

1/f. 3 RASPS:

does proride :he opera:or with a visual indiestioa of neu-tron Inet.

The consequencen u: re.:tiv!cy.;ccidents are functionn of the initial neveron flur..

'*he requirerr.-n c of at least 3 counti per necond aa.:arci that r.f trensient, a

should it occur, twins at or above the inittil value of 10" of rated power i.. ?d in t!.: ensty.e. of.ransients froo cold c: nidi tions. One 09erabi, '.Pti ch annel vould he adequste to raonitor the appronch tr' criticality using hczov,encous patterna of scattered con:rol rod v.*:hd:: val.

A nir.1 :e>

of two opersble SR?t's are provided as an edded conoerv.ati.4m.

5.

The Rcd Block Monitor (RBM) is desi;:ned to autoaati'cally prevent fuel da : age in the event of errencous rod withdrawal f rori locatio.s of high power den:aity during h13h power Icvel operacion. Two channels are provided. ind one of these may be bypassed froo the console f or eu'n=enance and/or testing.

Tripping of one of the channels vill block erroneous rod withdrawal soon enough to prevent fuci da. age.

The spect-fled restric:fons vich one channel cut of re. vice conserva-tively asaure that fuel danage vill not occur due to rod withdraual errors when this cond.ition exis:s.

A limiting control rod pattern is a pattern which results in the core being on a thermal hydraulic licit, (ie, MCPR given by Spe c. 3 5. K or LHCR of 18.5 for 7x7 or

13. 4 for 8x8 and 8x8R). During use of such patterns, it is judged that testing of the RBli sys t em prior to with-drawal of such rods to assure its operability will assure that improper withdrawal does not occur.

It is normally the responsibility of the Nuclear Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to tne 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 fuel damage; ic, v prevent the MCPR from becoming less than 1.07.

The limi,.ng power transient is given in Reference 1.

Analysis of thic transient shows that the negative reactivity rates resuS*ing from the scram with the average response of all the drives as given in the above specificction provide the required protection, and MCPR remains greater than 1. 0 7.

On an early BWR, some degradation of control rod scram performance occured during. plant startup and was deurmined io be c a u m' by 131 Amendment No. AA, 47

3.3/4.3 BASFi:

part iculate caterial (probably construction debris) p?.e,ging an internal control rod drive filter. The design of the prcatat control rod drive (Model 7RDB1443) is gro sly !mproved by the relocation of the filter to a lo ation out of the scrac drive path: 1.e., it tan no longer interfere with scram perfor nce, even if completely blecked.

The degraded perf ormance of the original drive (CKD 7RDL1 '.I. A) under dirty operating ronditions and the insensitivity of the rede=1aned drive (CRD 7RDB1445) has been demonotreted by a seri es of engineering tents under cimulated reactor opeer. ting condttions.

The successf ul perfornance of the ntu driv.. cnder actu21 operating conditions has also been demonstrated by cona'stently 2o0:1 in-service test renutte f or pleato usef..c. the Irive and may be inferred from plants using t!.o old. r rode l new driv with a modified (larger sc reen size) into:T.s1 filtc r which to 1 se prune to plugging. Data han been docus. nted by s ece:11-lanca reports in various operating pla tte.

There. include Oyster Creek, Monticello, Dresden 2 ani Dreedtr. 3.

5000 drive tests have been recorded to date.

Appr niectely yollowing Identification of the " plugged filtet" problen, vet y frequent serem testa were necessary to ensure proper perfermarce.

However, the more frequent scram testa are nce considerec' tettlly unnecessary and cnwise for the following reasona:

1.

Errat'ic scraci perforoance has been identifie! as due to ar obstructed drive filter in type "A" driver..

The driver. is BTRP are of the new "B" type design whose screa perforncnc e j

is unaffected by filter condition.

2.

The dirt load is primarily re* eased during startup of the reactor when the reactor and its ayatens rre first subjected to flows and preso. ire end thermal stresses.

Special atten-tion and near ureo este now being taken to asnure cleaner systems.

Recctore with drives identical or similar (shorter etroke, smaller piston areas) have operated through many refueling cycles with no sudden or erratic changes in scram se rfortanc e.

This preoperational and startup testing ic 4uf ficient to detect anomalous drive perforvance.

3.

hs 72-hour outage limit which initisted the start of the frequent scra m testing ie arbitrary, havin,, no logical basis other than quantifying a " major outage" uhich eight bly be caused by an event reasons-so severe as to poccibly affect drive performance. This requirement is unwise because it provides an incentive for shortcut actions to hasten returni g "on line" to avoid the additione.1 testing due a 72-hour outage 132 i

3.3/4.3 BASES:

The surveillance requirement for scram testing of all the control rods af ter each refueling outage and 10I of the control

(

rods at 16-veek intervals in adequate for determining the opera-is not so frequent as to bility of the control rod system yet cause excessive wear on the control rod system components.

The numerical valuce assitted to the predicted scram perfor-sance are based on the ana ysis of data from other BWR's with control rod drives the saae as those on Browns Ferry Nuclear Finnt.

The occurrence of scrap ticies within the limits, but signif i-cantly lone.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 exceeds eight, the allowable number of inoperable rods.

In the analytical treatment of the transients, 390 milliseconds are allowed between a neutron sensor reaching the scram point and the start of negative reactivity insertion. This is ade-quate and conservative when compared to the typically observed time delay of about 270 milliseconds. Approximately 70 milli-seconda after neutron flux reaches the trip point, the pilot scram valve colenoid power supply voltage goes to zero an approximately 200 mil 11seconda later, control rod motion begins.

The 200 milliseconds are included in the allovsble scram inser-tion timen 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.1.

The position of any rod bypassed must be known to be in accordance with rod withdrawal sequence.

Bypassing of rods in the manner described in specification 4.3.C.1 will allow the subsequent withdrawal of any rod scrammed in the 100 percent to 50 percent rod density groups; however, it will maintain group notch control over all rods in the 50 percent density to preret power level range.

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

133 Amendment 35 JAN i s W

3.3/4.4 BASFS:

D.

Reactivity Anomalics During each fuel cycle excess operative reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned.

The magnitude of this excess reactivity may be inferred from the critical rod configuration.

As fuel burnup pro-gresses, anomalous behavior in the excess resetivity may be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state.

Power operating base conditions provide the most sensitive and directly interpretable data relative to core reactivity.

Furthermore, using power operating base conditions permits frequent reactivity comparisons.

Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1 % o ff Deviations in core reactivity greater than I Lo k a r e not expected and require thorough evaluation.

One percent reactivity into the core would not lead to transients exceeding design conditions of the reactor system.

References 1.

General Electric BWR Reload 2 Licensing Amendment for BFNP uni t 1 reloe.d 2, NEDO-24136, August 1978 and Revision 1 dated November 1978.

134 Amendment No.

35, 47

l

.g 90 CAL" TION TEMPERAT'JRE MUST BE EQUAL TO OR CREATER TRAN

-~~

THAT INDICATED BY THE CIAtVr;

~~

I g

/

i p

70

/

m I

i p

7 4

$ 60 ei l

H I

/

l I

l

_L I

50

/

l

/

j i

I/

r J

\\

l

\\

/!

.=0 15 10

's SODi H F ENTABORATE 501ArrION (as w/o Na B2 10 16*"2 }

BROwHS F ERRY HUCLE AR PL ANT FINAL SAF ETY AN ALY115 REPORT SODIUM PENTABORATE SOLUTION TDOEPMJRE REQUIRDGES TIGURE 3.la-2 139 i

4 i

6 gASE]:

STANDBY LtQUID CONTROL SYSTEM A,

If no more than one operable control rod is withdrawn, the basic shutdown reactivity requireisent for the core is satisfied and the Standby L! quid Control System is not required. Thus, the basic reactivity requirement for the core is the primary determinant of when the 11guid control sys-tem is required.

The purpose of the liquid control system is to provide the capability of bringing the resetor from full power to a cold, zenon-free shutdown condi-tion assuming that none of the withdrawn control rods can be inserted.

To meet this objectiva, the liquid control system is designed to inject a quantity of boron that produces a concentration greater than 600 ppe of boron in the reactor core in 1ses than 125 minutes. The 600 ppa con-centration in the reactor core is required to bring the reactor from full power to a soberitical condition, considering the hot to cold reactivity difference, menon poisoning, etc. The time requirement for inserting the boron solution was selected to overridt the rate of reactivity insertion caused by cooldown of the rasetor fol-

.oving the menon poison peak.

T e minimum limitation on the relief valve setting is intended to prevent the loss of liquid control solution via the lifting of a relief valve at too low a pressure. The upper limit on ths relief valve settings provides system protection from overpressure.

3.

Only one of the two standby liquid control pumping loops is needed for operhting the system. One inoperable pumping circuit does not immed-tately threaten shutdown espability, and reactor operation can continue while the circuit is being repaired. Assurance that the remaining system will perform its intended function and that the long-tern average availability of the system is not reduced is obtained f ro a one-out-of-two system bf an allowable equipment out-of-service time of one-third of the normal surveillance frequency. This method determines an equip-ment out-of-service time of ten days. Additional conservatism is introduced by reducing the allowsble out-of-service time to seven days, and by increased testing of the operable redundant component.

C.

I.evel indication and alarm indicate whether the solution volume has changed, which might indicate a possible solution concentration change.

The test interval has been established in consideration of these factors.

Temperature and liquid level alarus for the system are annunciated in the control room.

The solution is kept at leset 10*y above the saturat1Jn temperature to guard against boron precipitation. The margin is included in Figure 3.4.2.

The volase concentration requirement of the solution are such that should evaporation occur from any point within the curve, a lov level alsru will annunciate before the temperature-conesutration requirements are exceeded.

140 knendment No. 47

~

LIM 5TIhD CONDITIONS FOR OPERATION SURVEII.I ANCE 2EQUIREMENTS 3.5.F Reactor Core Isolation Cooling 4.5.F Reactor Core Isolation Cooling 2.

If the RCICS is inoperable, 2.

When it is determined that the the reactor may remain in RCICS in inoperable, the HPCIS operatten for a period not shall be demonstrated to be to exceed 7 days if the operable icnediately and weekly RPCIS is operabic during thereafter.

such time.

3.

If specificationn 3.5.F.1 or 3.5.r.2 are not met, an orderly shut.down shall be initiated and the reactor shall be depressurizced 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 />.

G.

Automat ic Depressurization G.

Automatic Depressurization System (ADS)

System (ADS) l 1.

Four of the six valves of 1.

During each operating cycle the Autonatic Deprec uri-the following tests shall be zation System shall be performed on the ADS:

operable:

a.

A simulated automatic (1) prior to a startup actuation test shall be from a Cold Condition, performed prior to startup or, after each refueling out-age. Manual surveillance (2) whenever there is irra-of the relief valves is diat.ed fuel in the reactor covered, in 4.6.D.2.

vessel and the reactor vessel pr+:Mure is greater than 105 puig, except as speelficd in 3.5.G.2 and 3.5.G.3 below.

2.

When it is determined that ria r e than two of the ADS valves are 2.

If thtee of the six ADS valves incapable of automatic operation,

are known to be incapable of the IIPCIS shall be demonstrated autcmatic operation, the to be operable immediately and reactor :nay remain in opera-daily thereafter as long as tion for.i perind not to Specification 3.5.G.2 applies.

exceed 7 d.nys, pro.id?d the lirC1 ny t-n I: 'i c r A l +.

(Note that the pie suic rellef function of tbesc l

valves in assured bv section 3.6.D of t hs.:o spectticetinns a that this speci8frarion en alpties to the A% f unct ion.)

If more than thr ee of r.ne six ADS l

valves ni c kni 'm te be Incap-a51c ot :. n t on i t i c opc: a t ion, an iruw-itat e urat ly ?.hutdo m sha ll. be initia:6 it.

ith the 157 reactnr in a het sh"

• ib en con-dition in 6 hour: 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 />.

Amendment No. 25, 28, 47 i

i 1 tstlTISC Cyn(ll lone ros ')P I PA7_ ION SUKVEILLANCF. RE7J1RNENTS 3.i.C Aut nan t ic Denreemortration 4.5.C Automatic Depressurization System (ADS)

System (ADS) 3.

If specifications 3.5.G.1 and 3.5.C. 2 cannot be met, an orderly shutdovn vill be initiated and the reactor vessel preSeure shall be reduced to 105 peig or less withir. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

11. Hasntenanec of Filled Discharge H.

Maintenance of Filled Discharge Flpe Pipe Whenever the core spray oystems, The following surveillance require-LPC1. IIPCI, or RCIC.are required ments shall be adhered to to assure to lee operable, the discharge that the discharge pipinr, of the piping from the pump discharge core spray systems, LPCI, HPC1, and of these systems to the last RCIC are filled:

block valve shall be filled.

158

LIMtflWC CONDITIONS FOR OPFMT[0N S WVEIL(ANCE ea.itf(f MTNTS 4.5.H Maintensace of Tilled Discharge F13 3.5.M Maintenance or rilled Discharge Piee We suction of the RCIC and HPCI pumps d all be alis;ned to the conder. sate of the RNRS (LPCI and Containment storage tank, and the pressure suppres-Spray) and core spray systems, the sion chamber head tank shall nomally discharge piping of these systems be aligned to serve the discharge pipint' shall be vented from the high point of the RNR and CS pumps. Tre condensate and water flow determined.

head tank may be used to serve the RHR and CS discharge piping if the PSC head 2.

Following any period where the LPCI tank is unavailable. The pressure or core stray systems have not been indicators on the discharge or the RNR required to be operable, the dis-and CS pumps shall indicate not less charge piping of the inoperable sys-than listed below.

te:n shall be vented from the high Pl-75-20 L8 psig point prior to the retura of the F1-75 L6 L8 psig

,7,g,, g,,,,vge,,

F1-7k-51 L8 psis 71-7k-65 La ps18 3.

Whenever the HPCI or RCIC systes is lined up to take suction from the

1. Average Planar Linear Heat Generation condensate storage tank, the dis-charge piping of the HPCI and RCIC Rate During steady state power operation, the shall be vented from the high point liaximum Average Planar Heat Generation of the system and water flow observed Rate (MAPLHCR) f or each type of f uel as on a monthly basis, a function of average planar exposure f

shall not exceed the limiting value 4.

When the RHR$ and the CSS ara re-lTfatanytimeduringoperationitshown in Tables 3. 5.I-6-2,-3,-4 -5,/-6.

quired to be operabla, the pressure j

is indicators which nonitor the dis-

.Jetermined by normal surveillance that charge lines shall be sanitored the liatting value for APLHGR is being daily and the pressure recorded.

exceeded, action shall be initiated with-in 15 minutes to restore operation to withis the prescribed limits. If the APtJfCR 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 />.

I.

Maximum Averste Planar Liccar Heat Oenern-Surveillance and corresponding action tion Rate DiAPLHCR) shall continue until reactor operation The MAPLHGR f or each type of fuel a; a fu.ic-is within the prescribed lunits.

tion of average planar exposura shall be determined daily during reactor operation

g. Linear Hast Caneration Rate (LHCR) at > 25% rated therusi power.

During steady state power operation, the linear heat generation rate (LHCR) of any rod in any fuel assembly at any axial location shall not e.tceed the The IJICR as a function of core hatzht shal.

maximum allowable LHCR as calculated by be checked daily during reactor cperation at the following equation:

125% rated thermal power.

159 Amendment No. 35, 47 4

LIMITING CONJITIONS FOR OPr.oM :ON 5tJRVT ILLELF. REQUIRDENTS LHCR

< LHCR [1 - O P/P)

(L/LT))

max -

d m a2 De91gn LitGR = 13. 5 kW/tt. for 7x7 fuel LitGR

=

J

. g_5 kulft for HnRfuel OP/P) ximg y y g iking penalty

=

. o.o22 ior ass fuel LT"- Total core length a llo. f eet for 7X7 fuel 12.2 feet for 8X8 fuel

=

I, = Axial position above bottom of core If'at any time during operation it is det er-mined by normal sut veillance that the limiting value f or LHCR '.s being exceeded, act ion shall bi initiated vrshin 15 minutes to restore operation to within the prescribed limits.

If the LHGR is not returned to within the prescribed ILmits 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 corresponding action shall continue until reactor operation is within the prescribed Itzits.

E' Hir.inum Critical Power Ratio K.

Minimum Critical Power Ratio (MCPR)

(MCPR)

The MCPR operating limit for cycle 3 is MCPR shall be determined daily 1.34 for 7x7 fuel and 1.43 for 8x8 and during reacter power L7eration at 8x8R fuel. These limits apply to

> 25% rated thernal power and fol-ste.idy state power operation at rated power Ihwir chanbe in power level or m

and flow. For core flows other than rated, dist ribution th-t would cause opers theMCPR shall be greater than the above limits tioa with a limiting control rod times Ef.

1:r is the value shown in Figure pat tern as described in the bassa f.

3 5 2-If'at any t ime during operat ion Specification 3.3.

it in detenair.ed by normal surveillance that the limit ing value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If the steady r, tate MCPR is not returned t o within the presc r ibed limit s withir two (2) hours, the reactor shall be brought to the Cold Shutdown condit ion within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Surveillance and corresponding action shall c en t i n ue-unt il r e.w t er operatton is wtthin the prescribed limitg.

L.

yeportin3 Re ilu i_r em eilty if any of the limiting itue% identitled ia Spec i f ic at ions 3. $.1, J. or K are exceeded.ind the specified remedial action is taken, the event shall be logged and reported in a 10-day written report.

160 Amendment No. 35, 47

1 _"A8M 3.5.C Automatic betrassurination system (ADS)

This opsc111 cation ensures the operability of the AD$ under all condi-tions for which the depressurization of the nuclear systas is sa essen-tial response to station abnormalities.

The nuclear system pressure relief system provides automatic auclear system depressurization for small breaks in the nuclear system so that the low-pressure coolant injection (LFCI) and the core spray sobrystema can operste to protect the fuel berrier. Note that this specification applies only to the automatic feature of the pressure relief system.

i Specification 3.6.n specifies the requirements for the pressure relief function of the valves.

It la possible for any number of the valves aeoir,ncJ to the ADS to be incapable of performing their ADS functions because of instrumentatinn failures yet be fully capable of performing their pressure reitef function.

Because the automatic depressurization system does not provide makeup to r.h4 reactor primary vessel, no credit is taken for the steam cooling of the core caused by the system actuation to provide further censervatism to the C3C3 With two ADS valves known to be incapable of automatic operation, four valves remain operable to perform their ADS function. The ECCS loss-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. 35, 47 3 67

35 _f65{5 1.$.H Maintenance of Filled Ditcharp,e Pipe 11 the discharge piping of the core spray. LPCI. HPCIS, and RCICS are not filled, a water hemmer can develop in this pipine, when the pump and/or pumps are started. To minimize damata to the discharge pipina and to ensure added martin in the operation of these systems, this Technical Specification requirca the discharp.e lines to be filled whenever the system is in an operable condition. If a discharge pipe in not filled, the pumps that supply that line must be assumed to be inoperable for Technical Specification pur-posse.

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 filled. The visual checking will avoid starting the core spray or RHR system with a discharge line not (111ed. In addition to the visual observation and to ensure a filled discharge line other than prior to testing, a pressure suppression chamber head tank is located approximately 20 feet above the discharge line highpoint to supply makeup wuter for these systems. The condensate head tank located approximately 100 feet above the df scharge high point serves as a backup charging system when the pressure suppression chamber head tank is not in service. System discharge pressure indicators are used to determine the water level above the discharge line high point. The indicators v111refleet approximately 30 psig for a vnter level at the high point and h5 psig for a water level in the pressuresuppression chamber head tank and are men-itored daily to ensure that the discharge lines are filled.

When in their normal standhv condition, the suction for the llPCI and RCIC pumps are allene,1 to the condensa te storar,e tank, which is physically at a

hir.her elevation than the HPC15 and RCICS pipinr..

This assures that the HPCI and RCIC discharac pipina. remains filled. Further assurance is provided by abnerving water flow from theme systems hirh points monthly.

3.5.I.

Maximum t.vsrage Flanar I.inear Heat,Ceneration Rats (MAP 1JiCR)

This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident vill not exceed the limit specified in the 10CTR50, Appendix K.

The posk cladding temperature following a postulated loss-of-coolant acci-dont is primarily a function of the average hcot generation rate of all the rode of a fuel assembly at any axial location and is only dependent second-arily on the rod to rod power distribution within an assembly. Since ex-pected local vartations in power distribution within a fuel assenhly affect the calculated peak clad temperature by less than 1 200r relative to the peak temperature for a typical fuel design, the limit on the average linaar heat generation rate is sufficient to assure that calculated temperatures are within the 10CTR50 Appendix K limit. The limiting value for HAPLHCR is shown in Tables 3.5.I-1,-2 -3. 4,-5 4-6.The analyses supporting these limiting values is presented in NEDO-24056 and NEDO-24136.

Amendment No.

M, 47 168

4.5 Core and Contatument CoolinR Systems Survet11a..c Frequencies The testing interval for the core and containment costing systems is based on industry practice, quantitative reliability analysis, judgement and practicality. The core cooling systems have not been designed to be fully testable during operation.

For example, in the case of the HPC1, automatic initiation during power operation would resalt in pumping cold water into the reactor vessel which is not desirable. Complete ADS testing during power operation causes an undesirable losa-of-coolant inventory. To increase the availability of the core and containment cooling system, the components which make up the system; i.e.,

instrumentation, pumps, valves, etc., are teeted frequently. The pumps and motor operated injection valves are also tested each month to assure their operability.

A sinulated automatic actua-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.

When componeuts and subsystems are out-of-service, overall core and contain-ment cooline, reliability is maintained by demoestrating the operability of the remaining equipment. The der.ree of operability to be demonstrated depends on the nature of the reason for the out-of-service equipment.

For routine out-of-nervice 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 be thorough enough to assure that a generic problem does not exist.

For example, if an out-of-service period was caused by failure of a pump to deliver rated capacity due to a design deficiency, the other pumps of this type might be 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 that are required to be

(

operable shall be considered operable if they are within the required surveil-lance testing frequency and there is no reason to suspect they are inoperabic.

If the function, oyntem, or loop under test or calibration is found inoperable or exceeds the trip level setting, the LCO and the required surveillance testing for the system or loop shall apply.

Redundant ope.rable components are subjected to increased testing during equip-ment out-of-s ervice times. This adds further conservatism and increases assurance that adequate cooling is available should the r.eed arise.

Maximum Average Planar LHCR, LHCR, and MCPR The MAPLRCR, LHCR, and MCPR shall be checked daily to determine if fuel burnup, or control rod movement has caused changes in power distribution. Since changes

'due to burnup are slow, and only a few control rods are moved daily, a daily check of power distribution is adequate.

170 JkN 10 WJ8

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

This opecification assures that the linear heat generation rate in any rod is less than the desten linear hent generation if fuel pellet dengification is postulated. The power spike penalty specit' icd is haed on the anal-ysis prc=ented in Sec tion 3.2.1 of Reference 1 as modified in References 2 a nd 3, a nd a s s un.c s a linently incren ing variation in sa,lal gaps be-tween core bottum and top, and as u res with a '/R confidence, that no note than one f uel rod cuecds the de.igo 11ocar heat generation rate due to power e

sp1Linc. Tnc LHCK a:. a f unction of Cot e hcil;ht shall be checked daily dur-ing reactor operation at 2 25% power to determine if fuel burnup, or con-trol rod movement hw caused changes in power distribution. For LHCR to be a li: siting value below 2 M rated thermal power, the MTPF vould have to be greater than 10 which is precluded by a considerabic cargin when cmploying sny permissibic control rod pattern.

f 3.5.K.

Minimus Critical Power Ratio (MCPR)

At core thermal pcver levels lese than or equal to 25%, the reactor will be operatios at nintaus recirculation pump speed and the moderator void content will be very eull.

For all designated control rod patterns which uay be em-ployed at this point,

l operatins Plant experience and thetual hydraulic anal-yele indicated that the by a considerable margin.resulting MCPR value le in excess of requirestante

• tith this low void content, any inadvertent core flow locrease vould only place operation in a more conservative code rels-tive to MCTP. The delly requirement powr to suf ficient for calculating MCPR above 25% rated thermal have not been significantsince power distribution shifts are very slow when there pcver or control rod changes. The requirement for calculating PCPR when a liniting control rod pattern to approached ensures that HCPA vill be known following a change in power or power shape (regardless of siagnitude) that could place operation at a thermal limit.

3. 5.1.. Reporting Requirenents The LCO's associated with monitoring the fuel red operating cor.ditinns are required to be met at all t imes, i.e., there is na allevable time in which the plant can knowingly exceed the liciiting value9 f or MAPUICR, IRCR, cad NCPR.

It is a requirement. se stated in Specifications 3.5.1.J. end.K.

t hat if at any t!::e durlog steady state poser operatien, it la determined that the limiting values for MAPUICR, LHCR, or MCPR are exceeded action is then initiated to restore operation to within the prescribed limits.

This action is initiated as soon as normal surveillance indicates that it W been reached.

an eperating lir-Fach event involving steady state operstion beyond a limit shall be legged and reported quarterly.

It must ec reccgnismi thu specified there is always an action which would return any of the parameters (MAPIRC R,

IRCR, or MCPR) to within prescribed limits, ne=ely power reduction.

Under moet circums t anc e s, this vill not be the only alternative.

H.

References

" Fuel Densification Ef fects on General Electric Boi..ng thier us tor 1.

Puel," Supple: ent s 6, 7, and 8, hTIP-107 35, Augu s t 19 5.

2.

Supplement 1 to Technical Report on Densifications of General Electric Reactor Wels, December 14, 1974 (USA Regulatory Staff).

3.

Con:nnica t ion :

V. A. Moore to I. S. Hitchell, "Moolfied CT. hacel f or Puel Denaltication," Docket 50-321, March 27, 1974 4

General Elect ric BWK Reload 2 Liqensing A:nendment for BFNP unitl reload 2, NEDO-2!,136, Au6ust 1970 and Revision 1 dated November'1978.

169 Amendment No. 3E, 47

TABLE 3.5.I-5 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: 8DR265H Average Planar Exposure MAPLHGR PCT

(?Gd /t )

(kv/ft))

( F) 200 11.5 1707 1000 11.6 1698 5000 11.9 1681 10,000 12.1 1666 15,000 12.1 1688 20,000 11.9 1687 25,000 11.3 1639 30,000 10.7 1580 TABLE 3.5.I-5 MAPPHGR VERSUS AVERAGE PLANAR EXPOSUPI Fuel Type: 8DR265L Average Flanar Exposure MAPLEGR PCT

(!Gd /t )

(kv/ft)

( F) 200 11.6 1711 1000 11.6 1700 5000 12.1 1692 10,000 12.1 1663 15,000 12.1 1683 20,000 11.9 1683 25,000 J1.3 1637 30,000 10.7 1579 Amendment No. 47 172-a

DELETE IT3-a Amendment No. 35, 47

l.8H,8 T 8 8.. ('rpi,s[IT,ltw. 6,rpt, o,?l;RATI,fw8

,_Suyv1.tl.l. Aurr y,0 gild.3NT,,_, _ _

7 3.6.C Coelant 1.c a k a ta 4.6.C Coolant _ tin k a r,e 3.

If the condition in 1 or 2 above cannot be met, an orderly shutduvn shall be initiat-d and the reactor shall be shut-D.

Retter valves down in the Cold Cundition i

witlein 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

J.

A pproximat ely oric-half of all D.

Relief valves relief valves shall be bench-checked er replaced with a 1.

When more than one valve.

bench-checked valve each cpera-la known to ting cycle. All 13 valves be failed, an ordery shut-

~

will have down shall be int:1ated and been checked or replaced upor.

the reactor depressurized to the completion of every second teos thara 105 psig within 24 cycle.

bours.

2.

Once during each operating cycle, each relief valve shall be manually opened until therr.o-couples downstream of the valve indicate steam is flovtr,; froa the valve.

3.

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

4.

At least one relief valve shall be disassembled and inapec:<d each operating cycle.

E.

Jet Fumps E.

Jet Pumps 8.

Whenever the teactor is in the 1.

Whenever there is recirculatica startup or run modes, all jet flow with the reactor in the pumps shall be operable.

If startup or run nodes with both it is determincd that a jer recircula' ion pumps running, pu=p in inoperable, or if two jet pump operability shall be er more jet pump flow instru-checked daily by verifying that ment fattures occur and can-the following conditions do nat not be correcteJ within 12 occur sinaltaneously:

hours an ord.rly shutdo.m shall be Int:tated and the a.

The two recirculation leaps reactor shall be shutdown in have a flow imbalar.co of the Cold Condition within 24 15% or core when the pumps bours.

are operated at the same speed.

181 Amendment No. 47

4.tMIT,tNr. CfMf8tf3"NS FOR OPERAT109 suavEttt.awcr arquist!MDd

.~

J.6.E

'J e t f*uspa 4.6.E Jet Purope 3 6.T Jet PL?m Flow Missatch g -The indiosted value of core 1.

When both recircule. tion pu=ps flow rate varies from the are in steady state operation, walue M 1 M i m toop the speed of the faster pu=p flow measurements h mee shall be maintained within than 10%.

122% the speed of the slower pump when core power is 8$ or c.

The dif fuser to leker plenum siore of rated power or 135% the differential pressure read-speed of the clover p=p when ing on.sn individual jet core power is below B5 of pump varies Irom the mean rated power.

of all jet pu:rp differen-tisi pressures by amore than 2.

If specification 3.6.F.1 20%.

cannot be met, one recirculation puc:p shall be tripped.

2.

4thenever there 1,s recirculation flov vith the reactor in the

3. The reactor shall not be Startup or Run Mode and one re-operated with one recirculation circalation pump is operating loop out of service for more urith the equalizer valve closed, than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With the reactor

'the dif f user to lower plen~ure operating, if one recirculation 41fferential pressure shall be loop is out of service, the checked daily and the dif feren-plant shall be placed in a hot tial p.ressure of an individual shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the loop is jet pump in a loop sha,11 tiot sooner returned to service.

vary from the sem of all jet pu=p differential pressures in I

that loop by more than lot.

4.

Following one pu=p operation, i

the discharge valve of the low,-

speed pep may not be opened F.

Jet Fump Flou Miematch unless the speed of the faster pu:xp is less than SC(, of its 1.

Recirculation. pu:sp speeds shall be checked and logged at least' rated speed.

ence per day.

~

5. Steady state operation with both recirculation pumps out of ser-vice for up to 12 hrs is per-2ni tted.

During such interval restart of the recirculation pumps is permitted, provided tne loop discharge temperature is within 75cf of the saturation temperature of the reactor vessel water as determined by dome pressure.

C.

Structural Integrity 1.

Table 4.6.A to6 ether vich sup-C.

Structural Integrity 1.

The structural integrity of plementen notes, specifies the the primary system shall be 1 82

[ k, 47 Amendnent No.

3.6/4.6 BASES:

The basis for the equilibrium coolant iodine activity limit is a computed dose to the thyroid of 36 rem at the exclusion distance during the 2-hour period following a steam line break. This dose is computed with the conservative assumption of a release of 140,000 lbs of coolant prior to closure of the isola tion 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 lowcr 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.

c established in order to detect the The sampling frequencies occurrence of an iodine trausient which n.ay exceed the equilibrium concentration limit, and to assure tha t the maximum coolant iodine concentrations are not exceeded. Additional sampling is required following power changes and off-gas transicnts, since present data indicate that the iodine peaking phenonenon is related to these events.

3.6.C/4.6 C Coolant 1.eakace Allo"ahle lecksne rates of coolant from the reactor coolant system have been haar d on the predicted and experinentally observed behavior of cracks in pipcn and on the ability to makeup coolint systen leakage in the event of loss of offelte a-e power.

The normally expected background leakage due to

(

equipment design and the detection capability for determining coolant sys-ter leakage were n]oo considered in cotablishing the limits. The behavior of cracka in piping nyate,s hno been experinentally and analytically inves-tigated a's part of the tfSAEC cponsored Reactor Primary Coolant System Rupture Study (t he Pipe Rupture Study). Work utilizing the data obtained in this otudy indicates that leakage from a crack can be detected before the crack grows to n dangerous or critical oize by meenanically or thernally induced cyclic leading, or stress corrosion cracking or some other mechanism characterized by gradual crack growth. This evidence suggests that for leak-age somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or cracks asooeiated with such leak-age would grow rapidly. llovever, the establichment of allowable unidentified leakage greater than ttat given in 3.6.C on the basis of the data presently available would he prenature because of uncertainties associated with the data.

For leakage of tie order of 5 gpn, ao specified in 3.6.C. the experi-mental and annlytical dita suggent a reasonable margin of rafety that such loakage msgnitude would not rcuult f rom a crack approaching the critical site for rapid propagation.

Leakage less than the magnitude specified can be 218

3.6/4.6 casts detected rensenably in a aatter of feu hours utilizing the evnilable leakage detection schemes, and if the crigin ecanot be determined in a ressenably short time the unit should be shut dosn to allow further investigation and corrective action.

The total leskege rate consists of all leakage, identified and unidenti-fied, which flows to the drywell flocr drain and eouipment drain sucps.

The capacity of the dryuell floor sue:p pucp is 50 gpm and the capacity of the dryuell ec.uir,ent surp pump is elso 50 gpn.

Renovel of 25 spo from either of these sumps can be accomplished with considerable nargin.

RETERE;'CES 1

Nuclear System Leakage Rate Limits (BFNP pSAR Subsection h.10) 3.6.D/h.6.D Relier Valves To neet the safety design basis, thirteen relief valves have been installed on the unit with a total enracity of PF IJ. of nuclear boiler rnted stean flow.

The analysis of the vorst overpressure transient, (3-secen1 closure of all main stean line isolation valves) neglecting the direct scran (valve position scran) results in a naxinum vessel pressure of 1273 psig if a neutron flux neran is assuned considering 12 valves opernble. This results in an 102 psig margin to the code allowable overpressure limit of 1375 psig.

To meet the operational desi n, the analysis of the plant isolation 6

transient (turbinc trip with bypass valve failure to open) assuming a turbine trip scran is presented in Reference 5.

This analysis shows that 12 of the 13 relief valves limit pressure in the stean line to 1201 psic. This analysin-hows that penk system pressure is linited to 1229 psic which is 146 psig below the allowed vessel over-pressure of /375 psig.

219 Amendment No. 47

3.6/4.6 B AS ES :

valve operation shows that a testing of Experience in relief failures or 50 percent of the valves per year is adequate to detect deteriorations. The rollef valves are benchtested every second operating cycle to ensure that their set points are within the The relief valves are tested in place once per

+ 1 percent tolerance.

operating cycle to establish that they will open and pass steam.

system can be The requirements established above apply when the nucientThese requirements are a conditions.

pressurized above ambient at nuclear system pressures below normal operating pressures because at these conditions abnormal operational transients could possibly start eventual overpressure relief would be needed. However, these such that than those starting transients are much less severe, in terms of pressure, The valves need not be functional when the vessel at rated conditions.

head is removed, since the ntelear system cannot be pressurized.

REFERENCES _

Nuclear System Pressure Relief System (BFNP FSAR Subsection 4.4) 1.

Amendment 22 in response to AEC Question 4.2 of December 6,1971.

2.

3.

" Protection Against Overpressure" (ASME Boiler and Pressure Vessel Code,Section III, Article 9)

Browns Ferry Nuclear Plant Design Deficiency Report--Target Rock 4.

Safety-Relief Valves, transmitted by J. E. Gilleland to F. E. Krussi, August 29, 1973.

j 5.

GE BVR Reload 2 Licensing Amendment for BFNP unit 1 reload 2, NEDO-2L136, August 1978 a6d Revision 1 dated November 1978.

3. 6. E/4. 6. E Jet Pumps Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break. Also, failure of 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 :echnique is as follows. With the two recirculation pumps balanced in speed to within + 5 percent, the flow rates in both recircula-If the tion loops will be verified by control room monitoring instruments.

riser and nozzle two flow rate values do not differ by more than 10 percent, assembly integrity has been verified.

220 Amendment No. 47

4 6/4.6 qAff3 If they do differ by 10 percent or r. ore, the core flow rate measured by the jet pump diffuser differential pressure system P.ust be checked against the core flow rate derived from the measured values of loop flow to core flow If the difference between measured and derived core flow rate correlation.

(with the derived value higher) diffuser r.easurements is 10 percent or more will be taken to define the lecation within the vessel of failed jet pump nozzle tot riser) ar.d the uri.

shut down for repairs. If the potential blowdown flow area la incrt seed, the svetem resistance to the recirculation pump is also tcduced; hence, the af fected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a aine.le nozzle failure).

If om two loops are balanced in flow at the name 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 le a k a r.e path past the core thus reducing the core flow rate.

The revarse would still be indi:ated by a positive flow throur.h the inactive jet pump differential pressure but the net effect would be a slight decrease (3 per-cent to 6 percent) in the total core flow measured. Inis decrease, together with the loop flew increase, vould result in a lack of correlation between mea =ored and derived core flow rate.

Finally, the affected jet pump diffuser differential pressure signal vould be reduced because the backflov vould be ben the normal forward flow.

less t A nnrzle-rtaer system failure could also generate the coincident failure of a je: ro-p diffumer body: hevever, the converse is net true.

The lack of subatantial stress in the jet pump diffuser bedy enkes failure impousible any initial nozzle-riser syster. failure.

without an i

3.6.F/4.6.F Jet Pump Flow Mismatch The irr! laon selection log.: has been previcusly cescribed in the EFNP FSAR.

Fer nome limited low probability accidente with the re:irculation loop opera-ting with lorr.e speed differences, it is possible for the logic to select the vr on e. loop for injection. For these limited conditions the core spray itself is adequate to prevent fuel temperatures from exceeding allevable limits.

How-ever, to limit tne probability even further, a procedural limitation has been placed on the allevable variation in speed between the recirculation pumps.

Analyses indicate that above 80% power the loep select logic could be expected speed differential up to 14% ef their averase speed. Belov to functsen at a 801 power the loop select logic would be exoected to function at a speed differential up to 2 0 ?.

of their average speed. This specificatien provides

+ 10% and + 15*4 of the average speed for margin occause the 11.its are set at the abova and belov 80% power cases, respectively. If the reactor is opera-tina on one pump, the loop select lotte trips that pump before making the loop selection.

221

5.0 MAJOR DESIGN FFATURES 5.1 SITE,FLATURES Browns Ferry unit J is located at Browns Ferry Nuclear Plant site on property owned by the United States and in custody of the TVA.

The site shall consist of approximately 840 acres on the north shore of Ilheeler 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 bc 4,000 feet.

5.2 REACTOR A.

The core shall consist of 442 fuel assemblies of 49 fuel rods each, 166 fuel assemblies of 63 fuel rods each, and 156 fuel assemblies of 62 fuel rods each.

B.

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

density.

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

5.4 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 FUEL STORAGE A.

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

gg 330 Amendment NO. 35, 47

50 MAJOR DESIGN FEATURES (Continued) l chull be lenc than Er'. ref the opent Twil otr,rve poo The h. equal to 0 95 Fuel stored in the pool shall not it.

contain more than 15.2 grams of uranium-2'S per axial centimeter of fuel assembly.

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

fuel assemblies stored in the spent fuel pool.

5.6 SEISMIC DESIGN The station class I structures and syste=s have been designed to withstand a design basis earthquake with ground acceleration The operational basis earthquake used in the plant of 0.2g.

design assuned a ground acceleration of 0.lg (see Gection 2.5 of the FSAR).

\\

Amendment No. 42 331 h ~ $2 /~ ((

l