ML18024A679

Suppl Reload Licensing Submittal Presented as Justification for Proposed Amend Changes

ML18024A679
Person / Time
Site:	Browns Ferry
Issue date:	01/31/1979
From:	Ervin A GENERAL ELECTRIC CO.
To:
Shared Package
ML18024A680	List: ML18024A679 ML19263C290
References
78NED300, NEDO-24169, NUDOCS 7902130200
Download: ML18024A679 (47)

Text

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I'Pg M ENCLOSURE 1

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LIST OF TAOLESQCont'df Table 4.2.F Title Minimum Test and Calibration Frequency for Surveillance Instrumentation Page Ho.

4.2.G 4.2.H Survei 1 lance Requirements for Control Room Isolation Ins trimientat ion Hinimuin Test and Calibration Frequency for liood I'rotection Instrun~entation lO7 4.2.J.

3.5.J, 3.6.H 4.6.A Seismic llonitor'inrl Iristi<<ment S<<rve,illarice MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE.

Ssor.k buppre~srirs

(~r ubburs),

Reactor Coolant System Inservice Inspecl.iuri Schedule ir)A 171,172,172a 1 )t) 3.7.A 3.7.8 Primary Containment Isolation Valves Testable Penetratioris with Douhle 0-Ring Seal s 3.7.C Testable I'enetrations with Testdl)ie Oel low~

2~7 3.7.D 3.7.E 3,)).F Primary Contairinienl. Testable Isolation Valves Suppression Cliamber Infiuent Lines Stop-Clieck Globe Valve Leakage Rates Check Valves on Suppression Chamber Influent Lines 0

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3.7.H 4,0.A 4.".. O 3.11 F 3.A 6.0.A Testabl e Electrical Penetrations

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Radioactive Liquid llaste Sampling and An;,lysis Radioactive Gaseous Waste Sampling and Arralysis F>'e Prnrectiori System I!ydraulic Requirenients

!)rotection l.actors for Respirators Hi~)mum Shift Crew Requirements

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LIST OF ILLUSTRATIONS

~P3 ure 2.1.1 2.1-2 4.1-1

4. 2-1 3.4-1

3. 4-2 Title APRM Flow Reference Scram and APRM Rod Block Settings 0

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APRM Flow Bias Scram Vs. Reactor Core Flow Graphic Aid in the Selection of an Adequate Interval Between Tests System Unavailability Sodium Pentaborate Solution Volume Concentrati Requirements

. Sodium Penta bora te Solution Tempera ture Requirements on

~Pa e No.

13 26 49 119 138 139 3.5.2 3.6-1 3.6-2

6. 1-1 f Fac tor 0

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K 173 Minimum Temperature

'F Above Change in Transient Temperature

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0 188 Change in Charpy V Transition Temperature Vs.

Neutron Exposure

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e 189 TYA Office of Power Organization for Operation of Nuclear Power Plants 361 6.1-2 6.2-1

6. 3-1 Functional Organization 362 Review and Audit Function.............

363 In-Plant Fire Program Organi za tion........

364

LIHITIVC SAFETY SYSTEi~l S iYIH(i

'iiEL CLADDING INTEGRITY Z.l FUEL CLAOOENG IRTKCRXTY Applies to the interrelated vari-ables associated with fuel thermal behavior.

A l.icabilit Applies to trip settings of the instruments and devices uhich are prov'ded tc prevent the rczactor system safety limits fram being exceeded.

~Gt ective

'Io establish limits ~hich ensure the integrity of the fuel clad-d ing.

~Ot '

c ti ye To define the level of the process variables at uhich automatic pro-tective action i" initiated to pre-vent the fuel claCdlnc in agrity safety liai frau oeQ:g exceeded.

S ecificztions A. 'eactor pressure

> 800 psia and Core Flov > lOX of Rated.

S ecification The liroiting safety system settings shall be as sp cified beloM:

1fnen the reactor pressure is greater tl'.zn 800 psia, the existence of a ninicum criti-cal po;er ratio (NCPR) less t hac f.07 shzll cnnsti tcitc viola"ion of th fuel claddic~

integriq sa fe ty lt'ai A.

Neutron Flux Scr~~.

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APL'Ilux Scram Trip Setting (Run )lode) t When th Node SMitch i" the RUN position, the APP".

flux scram trip setting shall be:

S<(0.66V + 54K) c:here:

S ~ Setting in percent of rated thermal paver (3293 1%t)

W ~ Loop recirculation flower ra e in percent of rated (rated loop recirculation floci rate equals 34e2x106 lb/hr)

Amendment No.

35

SA~'r.s'Y ).D)JT

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J HTffi))71'Y r"rY qv Te P. l FUR)e ChhTlDJ))C JNTFG),'T'fY

~ In the event of operation with the core maximum fraction of limiting power density (CHFLPD) greater than fraction of rated thermal power (FPJ')

the setting shall be modified as follows:

S+ (O.66W + SCX)

FW CHFLPD For no combination of loop recircu-lation flow rate and core thermal

'ower shall the APRM flux scram trip setting be allowed to exceed 120X of tated thermal power.

(Note:

These settings assume operation within the basic thermal hydraulic design criteria.

The.e criteria sr~

Li)GR C 18. 5 kw/ft for 7Ã7 fuel and~

13.4 kw/Et for 8X8 and 8XSR fuel, MCPR within limits of Specification 3.5.k. If it is determined that either o! t)..cse design criteria is being viol ted during operation, action shall be initiated within 15 minutes to restore operation within piescribnd li"..:its.

Surveillance requirements for APF.:':

scram setpnint are given in specification 4.1.B.

2.

APR')

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

t B.

Core Pieaial Powe." Limit (Peart

". Pressure

< 00 psia) t

';.) e.". the reactor pressure is less tho..n or equal to 800 psia, 3,

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

B.

APRM Rod Block Tri n Settinr E1>e APN'o" block trip settin);

sh I) bc:

o 1

RPSF5:

."llFL Ct ADDIC 1}'TFCR}il SPEEDY L1MlT The fuel cladding represents one of the physical ba l } i h a arr ers which separate radio-active materials from environs. The integrity of thi cladding barrier is related to its relative fre6dom from perforations ra ons or cracking. Although some f corrosion or use.-related crac}:ing may occur duri th 1'f ission product migration from this source is inc ur ng t e i e of the claddin ncrementa13.y cumulative and gs continuously measurable. Fuel cladding perforati h r orations, ~ovever, can result from t ernal stresses'Mhich occur from reactor operatio i ifi conditions and the protection system setpoints. Phil fi i n s gn cantly above desi n g cladding performation is just as measurable a th t f s ~ e ission product g n rom thermally-caused cladding perforations signal a Ch h ld, as at rom use-related crackin the gn a t res old, beyond Mhich still Q ~ C e greater thermal stresses may cause gross rather than incr me t 1 1 8d Cion. Therefore, the fuel cladding safety limit is defi ed i f ncrementa c' ing deterior'a-operating conditions uhich can result in cladding perforation. s e n n terms o! the reactor The fuel cladding integrity limit is set such that no c 1 1 ed f 1 d a no ca cu at uel damage vould occur as'a result of an abnormal operational transient B en ecause ruel damage is not directly observable, the fuel cladding Safety Limit i d fi d s e ned uith margin Co the conditions Mhich uould produce onset transition b fli pfCP This establishes a Safety Limit such that the minimum c iti 1 on o ng tfCPR of 1.0). m a um cr C ca p~uer ratio (NCPR) is no less. than }g,07. HCPR > 1,07 represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. Onset of transition boiling resulrs in a decrease in heat t f f. n ea trans er f"om the clad

and, therefore, elevated clad temperature and the possiblit f

1 d f y o c a ailure. Since boiling transition is not a directly observable parameter, the margin co ooiling transition is calculated from plant operating p era ing parameters such as core pcrier, core floe, feedMater temperature and core pover di t ib ti e h I e s r ut on. Vne margin or eac fuel assembly is characterixed by the critical pove ti (CPR) h is the ratio of the bundle pover Mhich Mould produce onset of Cr i i b 1 divided by the actual bundle poMer. The minimum value of th i f e o rans t on boiling a ue o t is ratio for any bundle in the core is the minimum critical power ratio (HCPR). It is assumed that the plant operation is controlled to the nominal protective setpoi t e se po nts v a the instru-men ed variables,'.e., normal plant operation presented Fi 2,1,1 b on gure 1 by C5e n~}nal ex~ertt~ fine cnntrnl lin>>. <<e Safe tv Li~tt QgpgR of 1.07}ha~ <<<<<<cient conservatism to assure that in the event of an abnormal op ti 1 operat ona transient initfate~. from a normaL operating condition (NCPR > limits specified in specification 3.5.k) more than 99.9g of the fuel rods fn th>> core are expected to avoid boiling transition. The margin betveen MCpZ of 1.0 (onset of transition boiling) and the safety limit 1.07 is derived fron a detailed statisti "al analysis considering all of the uncertainties in moni-toring the core operating state including uncertainty in the boiling transition correlation as described in Reference l. The uncertainties e~ployed in deriving the'afety limit are provided at the beginning of each fuel cycle. 15

.l.l BASES Because the boiling transition correlation is based on a large quarrr,'ty ofs full scale data there is a very high confidence that operation of a fuel assembly at the condition of MCPH = 1.07 vould not produce boiling tran-sition. Thus, although it is not required to establish the safety limit additional margin exists between the safety limit and the actual occurence of los's of cladding integrity. Hovever, if boiling transition vere to occur, clad perforation would not be expected. Cladding temperatures vould increase to rrpproxirrrately 1100oF vhich is belov the perforation temperature of the cladding material. This has been verified by tests in thc General Electric Test Reactor (GETH) where fuel similar in design to BFNP operated above the critical heat flux for a significant period of thee (30 minuterr) vithout clod perforation. If,reactor pressure should ever exceed 1400 psia during normal pover operating (the limit of applicability of the boiling tranrrition corre-lation) it vould be assumed that the fuel cladding integrity Srrfety Limit has been violated. In addition to the boiling transition litoit (MCPR 1.07} operation is constrained to a maximum LHGR of 18.5 kw/ft for 7x7 fuel and 13.4 kw/ft for Sx8 and Sx8R fuel. This limit is reached when the Core Maximum Fraction of Limiting Power Density equals 1.0 (CMFLPD ~ 1.0). For the case where Core-Maximum Fraction of Limiting Power Density exceeds the Fraction of Hated Therrorrl Power, operation is permitted only at less than KOOK of rat.ed 2.1.A. 1. power snd only with reduced APRM scram settings as required by p ifi spec cat on At pressures belov 800 psia, the core elevation pressure drop (0 povcr, 0 flow) is greater than 4.56 psi. At lov powers and f1ovs this pressure differential is maintained in the bypass region of the core. Since the pressure, drop in the bypass region is essentially all elevation

head, the core prcssure drop at lov povers and flov vill alvays be greater than 4,56 psi.

Analyses shov that vith a flov of 283Q.03 lbs/hr bundle flov, bund1e pressure drop is nearly independent of bundle power and hpe a value o'f 3.5 psi. Thus, the bundle flov with a 4.56 psi driving head vill be greater than 28xl0 lbs/hr. Full scale ATLAS test data taken 3 at pressrucs from 14.7 psia to 800 psia indicate that the fuel assembly critical pover at this flov is approximately 3.35 HVt. With the derrign peaking factors this corresponds to a core thermal pov r of more than 50$.. Thus,'a core thermal pover limit of 251 for reactor presrrures belov 800 psia is conservative. For the Tuel in the core during periods when the reactor is shut dovn, con-sideration must also be given to vater level requirements duc to the effect of decay heat. If water level should drop belov the top of the fuel d i t vr ng his time, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation. As long as the fuel remains covered vith vater, sufficient conling is available to prevent fuel clad perforation. 16 Amendment No.,35

The safety limit has been established at 17. 7 in. above the top of the irradiated fuel to provide a point vhich can be monitored and also pro-vide adequate margin. This point corresponds approximately'to the top of the actual fuel assemblies and also to the lover reactor loM Mater level trip (378" above vessel sero). REFERENCE l. Cen'eral Electric BMR 'thermal Analysis Basis (CKTAB) Data, Correlation and Design Application, NEDO 10958 and NFDE 10958. p, General Electric Reload Licensing Amendment for BFNP Unit 2 Reload No. 2, NED0-24169, January 1979. n'and not'O. 35 l7

2.l BASKS: LI~ITINC SUETY SYSTE'.l SETTI.lGS PZLATED To FUEL cLADD;NG Il>TFCRITT abnon'al ope rat ion@I t r ns lent s appl I cable to op erat ion of the Br~a Ferry Nuclear Plant have been analyzed throughout the spectnn of planned operat jng ditions up to the design thermal parer condition of 3440 Nt. Tho anal)ses Mere based upon plant operation in accordance uith the operating map given in Fjaure 3 7 Of tne FSAE. ~ In addition, 3293 lNt is the licensed maximum pover level o f Brouns Ferry Nuclear P'ant, aod this represents the maximum steady-state pouer uhich shall not knovingly be exceeded. Qonservatism is incorporated in the transient analyses in estimating the contzolling factors, such as void reactivity coeffici nt, control rod scram ~or th ~ sc'ralo delay time, pea ~ing fau tars, hand axial pover shapes. These factors ace seiectcd conservatively uith respect to their effect on the applicable transient rcsulrs as determined by the current analysis model. This transient

model, evolved over many years, has been substantiated in opeza-tion as a conservative too'r evaluating reactor dynamic performance.

obtained from a Cancral Electric boiling uater reactor have been compared uith predictions made by thc model. The comparisions and res~ its are summarized in Reference l. The absolute. va)uc of the void rcactiv.'ty coefficient used in thc analysis is ronscrvativcly estimated to be about 257 greater than thc o inal ~i value expec <<d tu occur during thc core lf.fctiac, Thc scram uorth used ha> been dcratcd to bc equivalent to approx"..ately 8(K o! the total scram.orth thc control rods. The scram delay time and rate of rod insertion alloucp %o ~ 1 s ant> vcpi 1rc consozva tive 1 y sc t equal to thc longest delay hard "lov-eat insertion rate acceptable by Tcchnical Specificativna. The effect of scram orth, scree delay time and rod insertion rate, all conservatively applied, ar>> of greatest significance in the early portion of the negative reactivity insertion. The rapid insertion of negative react fvity ia assured by the time requir emen ts for 5X and ZpZ inaert ion. By the time the rods are 60K inserted, approximately four dollars of negative reac-tivity has been inserted Which strongly turns the transient, and accomplishes the desired effect. The times for 50K and 90X insertion are given to assure proper completion of the expected performance in the earlier portion of the tiansicnt, and to establish the ultimate fully shutdoun steady-state condition. For aoalyses of the thermal consequences of ths transients a HCPRQ 1Mi,ts specified in specification 3.5.k is conservatively assumed to exist ~rior to initiation oi'he transients. I s chg J cu of using conserve tive values of controlling parameters and inlt 'ting the design pouer level, produces morc pessimm I ans>ers than uoul J resul by using expected values uf control parameters and analyzing ot higher pover levels. i Steady-state. oper ation Mithout forced recirculation uiii nnt be pe~itted or morc than 32 hnurs. and the star t of a rec)rculatson pumo from the nat circulation condition will not be permitted unless the temnerature difference between the loop to be stal ted and the core coolant temoeraturc is less than 7Jof. This reduces the oositive reactivity insertion to an acceotably low value. Amendment Ho. 35 19

2.X en'.-.s 'i The 'scram trip setting must be adjusted to ensure that the LHCR transient peak is not increased for any combination of CNFLPD and FRY The scram setting's adjusted in accordance with. the formula in specification 2.3..A.l when the CMFLPD exceeds FRP. analyses of the limcing transients shoM that no scram adjustment is required to assure HCPR il 07 Mhen the transient is initiated from HCPR x limits specified in specification 3.5.k. gyRii Flux Scram Tri Settin (Refuel or Start 6 Hot Standi 'Bode} For operation 'in ch &ca cup oode while che reactor ia at low pressure, c e APKM 'scram setting oi'5 percent of rated po-c= provid"s adequate thermal margin becv&en the secpoint aiid the teiecy limit, 2$ percent of rated. The margin is adequate Co accommodate anticipated maneuvers associated with power plant startup. Effects of increasing pressure at xero or loM void content are minor, cold Mater from sources avail'- able dur'ng acarctup 's not much colder Chan that already in the system, temperature coefficients are small, and control rod patterns sre con-strained to be uniio~ by operating procedures backed up by the rod iwrth miniaixer and the Rod Sequence Con ro'ystem. Morth of indivi-dusl rods is very loM in ~ uniform rod pattern. Thus, all of possible sources of reactivity input, unifo. control rod vithdrawal is the noae probable cause of s.'qnificant power rioe. Because the flux distribu ioa associa ed v]th uniform rod Mithdrawals does not involve hi h 1 1 J ~ ~ o ve g oca pea&a, and oecause several rods must bc moved to change pover b7 a oitpaificant percentage of raced

power, tnu rate of power r'se is 'very slow C

ll c a ov, enewllyg the..eac flux is in n.ar equilibrium with the 'fission race. In an oaau=ed unifor rod wichdraval approach to tho scram lev 1, Che rate of power rise is no more ~an 5 percent of rated power per minute, and, ne QRN srstcm, would be aore chan adequate to assure a scram before cne povur could exce&d Che sa ecy limit. The 15 perccn" APK+ sc am re"sins active until che <de sc i,cch is placed kn the RUii position. This switch occurs when reactor pressure is greater than 850 paig. 3. IlN Flux Serac Tri Setcin 1lle I R'l Sys ceo cons is c s of 8 eh&labe ~ u ln $ 3ch C I,>>i cyst &iii logic <<hannels. The IR't is a 5-decad& ih<<range of poM&r l&vi'l betui'&n that. cc v&r&d bv 5 decades are <<overed by the IR:I by means oi a rac~e are broken doMn into l0 ranges, each being one-hali o IIN scrao setting of 1.0 divis inn& is acciv& in eacli oi the reactor,"rnt"c-nstn.acct .-nich covers SRI.>>iV the AZit:!. T;-.e s~icch and the 5 d codes dec&dc in 5 i e ~ T~ c runic oi che IÃ~'i. For Amendment No., 35 2l

3 ~ IRpj Flux 5cram Trir Settin-(Cont inn') example, if the instrument were on range 1, the scram setting would be at 120 divisions for that range; likewise, if'he instrument was on range 5, the scram setting would be 120 divisions on that range..

Thus, as the Md is ranged up to accommodate the increase in power level, the scram setting is also ranged up.

h scram at 120 divisions on the IRH instruments remains in effect as long as the reactor is in the startup mode. In addition, the APRM 15Z scram. prevents higher power operation without being in the RUN mode, The IPZf scram provides protection for changes which occur both locally and over the entire core. The most significant sources of reactivity change during the po~er increase are due to control rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods, that heat flux is tin equilibrium with the neutron flux and an IRM scram would result'n a reactor shutdown well before any safety limit is exceeded. For the case of a single control rod vithdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe ca~e involves an initial condition in which the reactor is just subcritical and the IRlf system is not yet on scale. This condition exist at quarter rod density. Quarter rod density is illustrated in paragraph 7.5.5 of the

FSAR, Additional conservatism was taken in this analysis by assuming that the IRN channel closest to the withdrawn

~ rod is bypass d. The results of this analysis show that the reactor is scrammed and peak power limited to one percent of rated po~er, thus maintaining HCPR, above,'1.0> Based on the above analysis, the IRN 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 moving control rods or by varying the recirculation flow rate, The APRH system provides a control rod block to prevent rod withdrawal beyond a given point at constant recir-cuclatjon flow rate, and thus to protect against the condition of a HCPR less than 1 07. This rod block trip setting, which is automatically varried with recirculation loop flow rate, prevents an increase in the reactor power level to excess values due to control rod with-

drawal, The flow variable trip setting provides substantial mnrp<Xn 22 Amendment No.

35

2'. l. bASES fron fuel damage, nssurcing a steady-state operation at the trip setting, over thc entire recirculation flow range. Thc margin to the Safety 'Limit increases the flou decreases for the sp~ c ificd r zip sett ing versus flou rt lationship; thetefore, the uorst case HCPR Mhich could occur durLIIR steady-state operation is 108K of rated thermal po~er because of the APRH rod block trip setting. Thc ~ actual poier, distributio<< I<< thc core i'stablished by specified co.astrol rod scquvnces and is monitored "cont inuously by the in-core LPRM system. As uith thlQ AFRIj scram trip s+I I ing, the App.'I zod block trip..et ting is adjusted doM<<Mard if CMFf FD exceed ~p thus preserving the APR.'I roJ block safety margin. Reactor Water LoM Level Scram and isolation (Fwce t Kvin SIcamlincs) point for the lo~ level scram Js above the bottom of thc separator skirt. ~is level has been <<sed in transient analyses dealing vi th coolant inventory zc cits reported in FSAR subsection 14.5 sl cu that scram and isolation of al 1 process lines (except main stean) at this level adequately protects thc fuel and the pressure barrier, because HCPR is greater than 1.07 in all cases, and system pressure does notreach the safety valve settings. The crau setting is approximately 31 inches beloM the normal operating range and is thus 'dequate to avoid spurious acre. The turbine stop valve closure trip anticipates the pressure, neutron flux 'nd heat flux increases that would result from closuIe of the stop valves. With a trip setting of 10'f valve closure from full open, the resultant increase in heat flux is such that adequate thermal maI gins are maintained even during the worst case transient that assumes the turbine bypass valves remain closed. (Reference 2) E. Turbine Control Valve Scrag I 1. Fast Closure Scram This turbine control valve fast closure scram anticipates the pressure, neutron flux, and heat flux increase that could result from fast closuIe of the turbine control valves due to load re)ection 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 after the start of control valve fast closure. This is achieved by the action of the fast acting solenoid valves in rapid1y reducing hydraulic contro) 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-out-of-two-twice logic input to the reactor protection

systenI, This trip settinq, a noRIinally 50" qreater closure time and a different va 1ve characteIist,ic from that of the turbine stop valve, combine tu produce transients very similar to that for the stop valve.

Ho signifi-cant change in t<CPR occurs. Relevant transient analyses are discussed in References 2 and 3 nf the Final Safety Analysis Report. This scralll is bypassed wlIen turbine steam flow is be'low 30>> of rated, as measured by" turbine first state

pressure, AmendIPqnt No.

,35 23

l ~ J, 4 j(. Reactor lou voters level set oint for ink tint ion o( i)pCl and an4 core s in ue s. These systems maintain adequate coolant inventory and provide core cooling Mith the objective nf preventing excessive clad temperatures. The design of these systems to adequately perform the intended func-tion is based on the specified lou level scram set point and initia-tion set points. Transient analyses reported in Section 14 of the FSAR demonstrate that these cond'itions,result in adequate safety margins for both the fuel and the system pressure. L. References Linford, R. 8., "Analytical Hethods of Plant Transient Evaluations for the Ceneral Electric Boiling Water Reactor," HED0-10802, Feb., 197>, t 2. General Electric Reload Licensing Amendment for BFNP Unit 2 Reload No. 2, NED0-24169, January 1979 25

higgs pressure moni'or higher in the vessel. Therero".e, ."ollovin, e."." trene'ert that is severe enough to cause concern that this safety limit was violated, a calculation wiii be performed usinp all available 1n. or.",ation to d ',:er-mine if the safety limit was violated, REi'.rGlfCZS 1, Plant Safety Analysis (3:.P.SAD Section 1'..3) 2, AS!Z Boiler and. Pressure Vessel Code Sec 1"n iiX USAS Pi-..ing Code, Secticn D3lel ~ 'v ecto" V~=se 1 and Aoaurtenances l!eche n ice 1 srcsi) n ( r l.P Subsecticn 4.2) 5, General Electric Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Power Station Unit 2 Reload No. 2, NED0-24169, January 1979.

2.2 BASES r Tne pressu-c reli f s"stem f." each unit at t'.".e 3:.";ns ."-err.. iiuclear Plant has b en sized to mee two desi".n Cases. r ir"t, "he total safet J/ relief valve capacity nas been esteolfsi:ed to ~ t.h overpr ssurc pro-tection criteria of thc AS!P. "ode, econd, ti: distribution of this rcouircd capacity between safety valves and relief,valves i!as been et t ~cat desi~n C"sis 4,4,4-1 of subsectin 1:,L '.".ici! sItptes tnn th nuclea>> system relief valves shall "."event openinz of ":",e sa.'c'y valves durin.- nor:",.el plan is"lations ana l'ad rc'ecti"..ns. T;!c details '[ the an" 1"sis wi!ich si:ows c=:.".'e.".c"::i;i'. recui"emcnts is presented in subs cti"n ~." =. the J"S;".i'esscl Gve>>.rrcssure P[ otecticn Su=.;.:n"" 7ec'.-.n'c-".1."-.e".;rt res>>~nsn a o <u sa i ~ 1~ ~ 1 dated 1 r ce ". e ] ? 'Y< 1 and ti!e >Qac tcl w>uo>+'iv d n To neet the oafetv design basis thirteen sa "e-.::-re! ief velves have been installed on unit 2 with a total ce,.ac";." of 84.2% ca nuclear oo'l r rated stca i fl w, The analysis o[ thc worst ":erpressur ran icn, {~- seccnd closure of all rain steam 1'ne isolation val'cs) neglectin-ti!e direct scram (valve pc iticn scram~ resul in a ":axi:um vessel prcssure o" 1285 psi[, if o neutrcn flue. screw is assu...'ed. This results in an 9O psiq margin o tne coda all wnble verpressure li=i f 1~7'j psig. To meet thc o,crntionnl des'-n basis, tile t tp'! Sa'e 'elic." capacii,y o. 84.2% of nuclear boil r rated '!as bc n divided into 7"'> relic'll vnlv s) and L<<.2% safety (2 valves), The analysis of the plant isclaticn transient (turbine trip with o~ass volvo failure to open) assu=in." a tur'"l[.e trip sera-is ".r~ sented in Reference 5 on cape 29, Vnis ana>;sis sh ws that tnc 11 relief valves lie;it pressure at toe safety valves o 12ll psi", we11 b low the settinp of, the safety valves, Theref r, tne safety val'es will not p n. This analysis sh ws tha. p ak system pressur is limi ed to 1236 psi-which is 139 ps ig below th allowed vess 1 ov rpressu"e cf 13 (j p" iG. ~ 30

en'pa ThBLE 3.2.B (Continued) 10. Only oae trip system for esch cooler fan. ll. In only tvo of the four 4160 V shutdown boards, Sea aoto 13. oaly onc of the four 4160 V shutdovn boards. See nota 13. 13 ~ hn eeHkrgency 4160 V shutdova board is considered a trip oyster. 14, RHRSM pump Mould be inoperable. Refer to section C.S.C for tc;s roquiroments of a RHRSV puap being inoperable. ]5. The <<ccident signal ia the satisfactory completion of n e::~~-out-of-tvo taken tvice logic of the dryvell high preooure plus low reactor pres-sure or the vosoel Icw water level (~ 376" above vessel sero) oH,giaatiag ia the core spray oyster trip system 16, The ADS circuitry is capeble ot accomplishing its protacttvo action . with oae operable trip oysters. Therefore ono trip system may ba taken out of service for functional testing aad calibretiaa for a period not to exceed 8 hours. 17. Two RPT systems exist, either of which will trip both recirculation pumps. The systems will be individually functionally tested monthly. If the test period for one RPT system exceeds 2 consecutive

hours, the system will be declared inoperable.

If both RPT systems are inoperable or if 1 RPT system is inoperable for more than 72 hours, an orderly power reduction shall be initiated and reactor power shall be less than 85X within 4 hours. 72

'.B (Continued) function RHR Area Cooler fan Logic Functional Test Teated during funcfional test of inetrBnaent

Channels, RHR sBotor start and thermostat (RHR area caoler fan).

No other. test required. Calibration N/h Instrument Check Core Spray Area Cooler Fan Logic Tested during logic systen functional test of instruaent

channels, core spray aotor start and thcaaa stat (core spray area coaler fan).

Ho ather test required. Nlh Inatrunent Channel-Core Spray Motors h or D Start Tested during functional N/h test of core spray puap (refer to section 4.5.h). N/A Inatruaent Chsnnel-Core Spray Hotors 5 or C Start Tested during functional N/k test of core spray puep (refer to section 4.5.h). N/h Instruncnt Channel-Core Spray Loop l hccident Signal Tested during logic systen functional test of core spray systea. N/h N/A Inatreacnt Channel-Core Spray Loop 2 Accident Signal RHRSQ Initiate Logic HPT initiate logic l RPT breaker ~ 1 Tested durins logic eyatesL functional test of core spray p Syetene once/6 months once/month once/operating cycle N/h N/A N/A N/h '.f/h N/A N/A

, BASRS The HPCI hfph i'ou snd Ccroprrnturc fn"trumenta<fun nrr provided to detect reak in thc HPCl steam pfpfng. Tr lppfnp of thl". Instrru<<ntation re-sul ts fn actuation of HPCf f sul at fon valves.

Tt Ippl<

>rc at tlic Proctor Clcanuo System . lnor drain could fndfcatc a break fn tlic clear!:ip system. t'hcn tifgh temperature

occurs, the cleanup system fs isolated.

The fnstruncnt atIon uhfth fnf t fates CSCS action is arranzrd fn a dual bus system. hs fcr other vital tnslrumencatfon arm>>gcd fn this fashfon, Chc Specification preserves thc effectiveness of th>> systen cvrn during periods uhen malntcnance or testing fs befnp per(orncd. An exception to this fs uhcn lug!c functiona.'cstinr, fs bcfnp pcr(orne-.. l 'he control !orl block (unc.Uo>>s are provided tn prevent excessive control rod vfthdraual so th; t tlCPP. does not dccrcaso to 1.07. Tire trfp logic for Chfs (unct fon fs 1 out of n:

c. g..

nny tr 1 p on one of sion 4pR<'s, eight TL'1's. or four SR.'1's vill result fn a rod block. Tlie !ofnfmum li sr 'ment channri r~qulrc.".cnts assure cion to assur c tire sfr:pic failure triter!a ! o net. channel requirements !nr the PB.'l n y bc rcduccd ny

testing, or calibration.

Th!s tfroe period fs only fn a month, and does noC sign!,ffcantly fncrcnse the friadvcrtent control rod vfthdraual. sufffcfcnt fnstruroenta-Thc Din!mum fnstruxent one for maintenance, 3:: of the operating tfac rfrk of preventing an The APRN r& block funct fon 1 s flou b'.ascd and pt events a sfgnf ffcant reduc-tion fn tlCF!l, especially Cur!ng operant ton at reduced (lou. The APR.'( pro-vides grosp core prctcccfori; f.e., lf'rftri che gross ccrc poucr increase fror. vfChdraual of crntrol rods ln thc normal u!Chcraual sequence. The i trips are set rio t'rat HCPR fs ronfn afncd greater than 1.07. a Thc RBM rod hlnck (ur': t fnn prov!i!cs Ir ca'rot rc fi n n( the core,'. e., Chr prevent ion of c r f c fc! 1 p< cr fn a local rcg fon of Che core, foc a s ingle rod uf thdravii1 error from n 1 frofcfrg control rod pat tern. )13 Jn< <i<n<',!c c<pera ~ nr vl '.h << viova'1 1<<die' ton of <<cu- ! <v< l. T!>e cn<<sect<<>>~< r~ <<.'e >eel<<! r < ~c< td "rta f<<i<( t 1o<<n c< I:!<. 1 n! . a I <<c" t << ~ I l ~.".. ':h" r",qc.'e<<.,",.". t o[ <<t lena! 3 cnu<ir i ti>>r nec<<<<i! aaaurra that '<:./ tr<'asiunt, occur, r<'t<<~ <<t <<r zliovc 1!<<< I<<'.t! <l v<<lu<<nf lo nf f<<t<<1 I <<v r < ~. <I I:< t <<<<11/s>>s <<I ar< ent'< I'<ru < r><i<.!1'.1 ~nn. 0:<;; o <>>rnbi>> '. ~ <<t1 (I>>na I vn <'d be a<lcquqtc ~tn <<<<nl <nr the approach tc ~ t: I t!t>llty i.'s.'n:, hcco!te<<cou> Dtlcrn < of scat t c red co <'all md w.:,.d: va I. I -< "..1..<v o'wo operable S.".'I'a are provided r<s an cddeo coracrjatiaz. The !ted lrlc<c~.'lonitor (Ra.,) is des'.Rned to auto at'.call) preve<<L fuel d".<i<<<". 1<< tn.'ve<<t c<l e::nneous rod vi hdrawal f my 'ne" t lo.".a c1 1/ dut t:<Z hi!th power le iel ope-,a<, ton. Two RBM channels are provided, and one of these may be bypassed from the console for maintenance and/or testing. hutomatic rod withdrawal blocks from,one of the channels will blcck erroneous rod withdrawal soon enough to prevent fuel damage. The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists, h ] irni t 1 ng.ann trol rod pat tern is a pat tern whi el< re su] l... in the: core being on a thermal hydraulic lir<it, (ie, MCPR given by Specification 3.5.k or LHGR of 18.5 for 7x7 or 13./< fc)r 8QQ a nd 8x8+ 0 u r 1 n g u s c' o f s u c h p a t t e r n s, 1 t I s rhat t t st I ng nf t lie RBII system pr ior to, w j (I<- drawn] nf such rods to assure its operability wil] asrure tliat improper withdrawal does not occur. It is normally the responsibility of the Huclear Fngjneer to identify these limiting patterns and r'n desjgnated rods either when thc patterns are initially established or as tliey develop duc to the occurrence o f inoperable control rods in o ther tlian 1imi t ing patterns. Otlier persnnnel qual ificd to per-form these functions may be designated by the plant superintendent to perform these functions. Scram Insrrl.ion Times The control rod system is designated to bring the reactor subcritical at the rate fast enough to prevent fuel damage; ie, to prevent the >ICFR from becoming less than 1 07. The ) imiting po<"c r transient is given in Reference l. Analysis of this tran 1ent shnws that the ncgatlve reactivity ral:es resulting frnm the ..cram with the average respnnse of all the drives as g1ven in the above specification provide the required proc<<et Inn, and IICPR remains greater than 1.07. On,a n c or'" I l l', snr<<; degradat Inn of control rod serac< pci (nri'ri<r"r ""cure d c!ui'Jng nl ant st art up and was determjnc d r cl hc r.<< ~;

3. 3/4. 4 BASES:

I D. Reactivit Anomalies During each fuel cycle excess operative reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned. The magnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup pro-

gresses, anomalous behavior in the excess reactivity may be detected by 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 lX A~ Devia tions in core reactivity greater than lyQ+ are not expected and require thorough evaluations One percent reactivity into the core would not lead to transients exceeding design conditions of the reactor system. References l. General Electric Supplemental Reload Licensing Submittal for Browns Perry Nucleyr Power Station Unit' Reload No. 2, NED0-24169, January 1979. 134 At>>raiment No. 35 OASFS: STANOBY hL~Ul0 CONTROl SYSTEM A. 1( no more than onc operable control rod Ls MithdraMn, the basic shutdown reactivity requirement for the core is satisfied'nd the Standby Liquid Control System is not required.

Thus, the basic reactivity requirement for the core is the primary determinant of Mhen the liquid control sys-tem is, requir'ed.

The purpose of the liquid control system is to provide the capability of , bringing thc reactor from full pouer to a cold, xenon-tree shutdoMn condi-tion assuming that none of the MLthdrsvn control rods can be inserted. To meet this objectiv"., the liquid control system is designed to inject s quantity of boron that produces a concentration greater than 600 ppm of boron Ln the renctor core in lese.than 125 minutes. The 600 ppm con-centrat ton in the reactor core is required to bring the reactor from full pnver to o subcritical condition, considering the hot to cold reactivity dLEference, xenon poisoning, etc. The time requirement for inserting the boron solution Mas selected to override rhe rate of reactivity insertion caused by cooldown of the reactor fol- ~oMLng the xenon 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 reli.ef valve at too loM a pressure. The upper limit on thc relief valve settings provides system protection Erom overprcssure. B. Only one of the tMo standby liquid control pumping loops is needed for operating thc system. One inoperablc pumping circuit does not immed-iately threaten shutdovn capability, and reactor operation can continue Mhile the circuit is being repaired. Assurance that the remaining system Mill perform its intended function and that the long-term average svstinbility of thc system is not reduced Ls obtained Ero a one-out-of-ten>>y>>tern bt nd allowable equipmcnt out-of-ser>Lcc time of one-third of thr normal surveillance frequency. This method detc rminea an equip-ment out-of-icrvlcc time of tcn days. Additional conservatism ie introduced by reducing the,alloMsblc out-of-service time to seven

days, snd by increased testing of th>> operable redundant component.

C. l.evc1 indication snd alarm indicate Mhether the solutLon volume has

changed, Mhich might Lndicate a possLblc solution concentration change.

The test interval has been established in const leration of these factors. Temperature and liquid level alarms for the system arc annuncia'ted in the control room. The solution ie kept at least l0'F above the saturation temperature to guard against boron preqipitat ton. The margin is included in Figure 3,4.2. The volume concentration requirement of the solution are such that should evaporation occur Etom any point Mithin the curve, a loM level alarm Mill annunciate before the temperature-concentration requirements are exceeded. 140 .iTINC COMIT IAHS YOII OPF RAT IOlV .H Maintenance o{'illed P{"char e Pipe ~e suction of the i{CIC and {iT'CI pumps shall be aligned to.the condensate storage

tank, and the pres"ure suppres-sion chamber head tank shall normally be aligned to serve the d{scharg.

pipin,", of %he R{N and CS pum s. Tt c "cond'ensate head tank may be used to serve .,he R((B and CS discharge pi'ping if the PSC head tank is unavailable. T"ne pressure 'indicators on the discharge c!'he R{P. and CS pumps shall indicate not less than listed below. Pl-75-20 48 psig Pl-75-46 4{) psig Pl-74-51 48 psig Pl-74-6S 4{I psig Rate During steady state power operation, the Haximum A'verage Planar Heat Generation 'e (MAPLHGR) for each type o! fuel as unction of average planar exposure snail not exceed the Limiting value shown in Tables 3,5.1-1~ 2~-3p "4pand -5, f't any time during operat{on it is 4rtermined bv normal surveillance that the limiting value for APL((GR is being

exceeded, action shall be initiated with-in 15 minutes tn restore eperation co within the prescribed" limits. lf the APLHGR is not recurned to vichtn the prescribed 1{m{ts within two (2) hours, the reactor shall be brought to the Cold Shutdown con<licion within 36 hours.

Surveillance and corresponding action shall continue until reactor operation is within the prescribed 'comics. J'. Linear Heac Generation Rate (LHGR) During steady state pover operation, the linear heat generation race (LHGR) of any rod in any fuel assembly at any axial location shall not e.~'ceed the maximum al'ovable LHGR,as calculated by the following equation: 54elVF aLLAj{CI, >'I. <I!,i!<<,>INTS 4.5.H Haincenunre of Fil'ed Dischar . Pie~ l. Every month prior to the c..scLng of the RI{RS (LPCI and ConcainmencI Spray) and core spray systems, the discharze piping of these systems shaLL be vented from the high point and water flov determined. 2. Following any period where the LPCI or core spray systems have not been required to bc operable, the'dis-c{~arge p'ping of the inoperabls sys-tem shall be ~ented from the high poinc prior to the return of the system to service. 3. Whenever the HPCI or RCIC system lined up to take suction from the condensate storage

tank, che dis-charge piping of the HPCI and RCIC shall be vented from the high point of the system and vater flow observed on a monthly basis.

4. @hen the R{!RS and the CSS are re'- quired to be operable, the pressure indicators which monitor the dis-charge lines shall be monitored daily and the pressure recotded. X. Max&mum Ayers-"e P'ansr Linear Heat Gene;a-tion Rate ("IAPLHGR) The HAPL((GR for ach ype o! f.el a ' fu. c-tion of average planar exposure shall bc determined daily during reactor operation at + 25K rated thermal power. J. Linear Heat Generation Rate (LHGR) The LHGR as a function rf cere ~e!ghc sha be checked daily during reactor eperacion at > 25X rated thermal pover. '5.lVITI HG CONa)ITIOHS FOR OPERA'I 'OM LHGR LHGRd [I - 5 P/P) (L/LT)j LHGR - Oesign LHGR L9.5 kv/ft. for 7x7fuel ~13. 4 kt"/ft for Rx8fuel 'd P/P)max Maxim+a pqyrf spiking penalty ~ 0.022 for 8x8 and 8x8R fuel LT Total core length~ 12.0 feet for 7x7 and 8x8 fuel = 12.5 feet for 8x8R fuel L ~ Axial position above bottom of core. If at any time during operation ic is deter-mined by rormal surve!1 lance c'hat the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to ~ithin the prescribed limits. If the LHGR is not returned to vithin the prescribed limits within cwo (2) hours, the reactor shall be brought co the Cold Shutdown condition vithin 36 hours. Surveillance and corresponding action shall continue until reactor operation is wichin,che prescribed limits. K. Minimum Critical Power Ratio (MCPR) The MCPR operating limit for BFHP 2 cycle 3 is 1.33 for 7x7,1.30 for SxS/8xSR fue1. Thes~ limits apply to steady state pover operation t at rated pover ana flov. For core flovs other than rated>the MCPR shall be greater than the above limits times Kf. Kf is the value shovn in Figure 3.5.2. SURVF.ILI&hCF. RF. UIRgUNTS Minimum Critical Power Ratio (HCPR) MCPR shall be determined daily during reactor power operation at i 25X rated thermal pnwer and !ol-lowlng nny change in power level ot distribution th t vnuld causa opera-tion with a lioiting control rod pattern as described in the bases fc Specification 3.3. If'at any thee during operation it ia determined by normal surveillance chat the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes co restore operation to within che prescribed limits. If,the steady state ".!CPR is noc returned to ~ithin the prescribed limits within tvo (2) hours, che reactor shall be broughc co the Cold Shutdown condition within 36 hours. Surveillance< and cnrrespond ing ace inn shall tnntinur unt tl ra'.i; tnr operac(nn is wich tn che prescribed 15m5ts. L. ~Ns ort tn ~ue uk."cravat. If any nf the i,im'.cinp v >la~~ !.!encl 'ed Specif(cacions 3.5. I, 3, or R are exceeded 'nd the spec ified remedial acc inn Ls taken, rhe event shal) b'e logged ard reported in a 30-da> tvritten report. Amendment Ho. 35 f fn of Che cote spray, LPCf, HPCIS, and RC1CS are not

tills4, a vater ha! ear can develop fn thLs pipfnr.

an c a peep an or s art sca!cad. o s! n a rc T i f L c da!!!acc co the dischar!ta piping and co ansu,c ac%ad I!ar%fn fn t>,e operat on o e ti n of chess syace!!!s, this Technical Specfffcacfon sequfrca the dfsc nrre nes d h lines co be filled vhenave'r the syste!!! Ls fn an cable cendf cion. If a discharge pipe fs net fflled, the pumps that supp y su 1 d b f opernble for Technfcal Specffiiation pur-thac line ii!vsc be assueed to e nopecn possess The core spray an sysceii! d RHR c dfscharre pfpfna hfah point venc Ls visually che checked for,voter ov once n ~ flov once n moth prfoc co cescfnr, to ensure that c e fffid. Th vfs!!al checking Mt,ll avefd startina the cote spray or NN sywce~ vfch n discharge ) fnc not fflied. Ia addition to the visual a prosoure suppression chamber e orvation aad to ensure a e orva I'ill d discharge line other than prfo. to testfag, h ad tank Ls located app!roximately 20 feet abovi s terna. The the disc ge ac har li highpoiat to supp~ makeup vater for these systems. e schar hL h coadenscbte ea h d tank located approximately 100 f'eet abevo the di ge g c ressfoa chamber point serves as a ac p c b ku charging system vhea thc pressure suppressio ure indicators are used to head tank Ls not La service. System discharge pressure d t iae the vater level above the discharge 1fne gh po e hi iat. The indicators villreflect approximately 30 psfg for a vater level a th g po e ere ae t e hi h iat aad 5 psig for a vater level ia the preasuresuppression chamber head tank aad are mc itored daily to easure that the discharge lines are filled. Mhe I th Lr nor!Inl standby cnndit ion, the s!!ccfen for the IlPCl nnd RCIC p!~a arc zffrncf cn thr condinsacc scornpe

conk, Mhfch Ls phvsfcal yI nc h

ltPCIS nd ACICS plpfnr.. Thfs assures!Chat Che llPC1 revved b and aCIC dfachnrac pipfn< rcr afns filled. Further assurance Ls prov c y nbaervfnq vncer flov frow these aysccas high points ~nthly, Wafaua Ivazago Planar Linear Heat Generation Rate'HhPQK ) that che peak ciaddfng tc!sperature tollovfng tho This specification assures t at e e @coed the pcNtQ14 t0 c s gn d d L b sis loss<<of-coolant accident vill not emcee e lie!Lt specified Ln the 10CFR50, Appendix K, 4 oak cladding tewperacure follovfng a postulaced loss-of-coolnnc acci-a function of the average hcoc generation race of all the dent ia prficarfl'y a unct on o e onl dc endcnc second-rodo of a fuel asseI!bly ac any a!cfal locacfon and fs on y p h rod co rod peMer dl crLbucfon vfchfn an assembly. Sjnce cx-arily on th. ro co r po e f 1 assi.'rib! affect ec ccd loca ~ voc a 0 ia .the d 1 cfons fn povcr df"crfbvtfen vithfn a fue i.ri.y the calcula:c pro c a d k 1 d teiapcrncvrc by less ch~n 20 P relacfvc to. e oak tci!ipervcvrc or a cypica uc f I fuel design thc lf!!!Ic on the svcrape linear boat gcnaracLen race Ls suf ffcfcnc co assure chat calculaccd tenpecacurcs Ln che LOCFR50 Aopcndfx K If~Le. The limiting value for MAPLHGR is ses su tin these limitiag rslues is preserted ~C-24088 and NED0-241.69. 3.5.J. ~Lfnenr }}cat Generation Rate l.llCR T}:fs specification assures that the linear heat generation rote in any fs less Chan the desfpn linear henc generation if fuel pellet dcnnfffcat fan is postulated, Thc pover spike penalty specified is based on the anal-ysis presented fn Section 3.2 ~ 1 of Rcfcrcncc 1 as modified in References 2 and 3, and assumes n lfnrarly lncrca ing variatfo>> ill Ov}al ga})s bc-Cvecn cora bottom and

cnp, and assures vlth a

95% confidence, ch>>t no morc than one fuel rod exceeds the dcign 1}near heat Gencrnt.lon rnte duc co povcr spfl:fng. Tnc LIICR a" a functfon of core hcfgl!t sh" 11 bc checked daily dur-jng reactor oper>>t ion at 25% pover to dcccrn>inc if fuel burnup, or con-troll 'f9d movcmes t ha s cau" cd changes In povc r d }s t rebut Ion I For L}ICR co bc a 1faf tinp value bclov 25% rated thermal povcr, thc !}Tpp vould have co be greater than 10 uter}ch f precluded by n considerable margin vhen employing ~ny~crmf~sfble contro1 rod pattern. 3.5.K. Mfafmum Critical Pover P~cfo MCPR At core thens~l pcver levels less than or equal to 25, the reactor vill be operating at minimus recirculation pump speed and Che moderntor vofd contra vill bs very small. Por all designated control rod patterns vhfch may be em-ployed at this point, operating plane experience and thermal hydraulic anal-ysis &dfcaced thrust tbe resulting HCPR value is in excess of requirements by a considerable mnrgfa. Vfth this lou void content, any inadveztent core flou increase uoufd only place operation fn a more conservative mode rela-tive to HCPR. The daily requireaent for calculating MCPR nbove 25% rated thermal pover is sufficient ei'nce pover distribution shifts are very siov vhen cher en sere have aot been sfgnif cant pcver oz control rod changes. The requirement for calculating HCPR vhen a lfafting control rod pattern is approached ensures that HCPR vill be kaoun follovfag a change fn pover or pover shape (regardless of magaftude) that could place operation at a thermal limit. ~Re orcfa Re ufreaeats The LCO's associated vfth monitoring the fuel rod operating conditions are required to be mec at all times, i.e., there fs na ellovable time fn vhich Che p'nat caa kaovfngly exceed the lfnfcfng values for MAPLHCR, LHGR, and HCPR. It $ o s requfremeat, as stated in SPecifications 3.5.I,.J, r~d chat if at any time duriog steady state paver operation, ft fn determined Chat the 1fmfting values for MPJ'LHGR, L1IGR, or HCPR are exceeds} act'on fs chen initiated to restore operation to vichin the prescribed 1fmfcs. This action ks iaftiated as soon ne noraal surveillance in icates tha an operating has been reached. Each event involving steady stace operation beyond n specff fed lfmft shall be logged sad reported quarcerly. It aust be reccgniaed that there is alvays en action uhich vould return any of the p rareters (MAP~MCR, L}fGR, or HCPR) co vfthia prescribed limits, aameiy pover reduction. l?ader most circumstances, this vill aot be the only alternative. I H. References 1, "Fu}!I Deaeificacfoa Effects oa General Electric Boil ng i(a' g(aa 'cor purl," Supplements 6, 7, and 8, NET';10735, August 19?3, Supplement 1 co Techn fcs-Report on Dens'eat fons of Cence ai g1eccrfc P~accor

Fuels, Deceaber 14, 1974 (USA Regulatory Staff).

4. Commuaf cncf on: V. A, Moore to I. S. Mitchell, "Moo if 'c!d CE t:odei for Puel Densif icatfoa," Docket 50-321, March 27, 1974. General Electric BVR Reload 2 LicensingAmendment for BFNP Unit 2, NED0-24169,January 1979. TABLE 3.5.I-5 HAPLHGR VERSUS AVERAGE PLANAR EXPOSURE AVERAGE PLANAR EXPOSURE NM/t 200 1,000 5,000 10,000 15,000 20,000 25,000 30,000 MAPLHGR kW/ft: 11.2 11.3 11.8 12.0 12.0 11.8 11.2 10.8 Fuel Type: 8DRB284 PCT l."F) 1685 1667 1671 1647 1669 1672 1633 1596 1728 ~ 4 ~ lf - detected reasonably in a matter of few hours util'zing the-available leakage detection sc'nemes, and if the origin cannot be determined in n reasonably snort, time tne unit should be shut down to allo'<< further investigation and corrective action. The total leakage rate consists of all leakage, identified and unidenti-fied, which flows to the drywell floor drain ante d eouipment drain sumps. The capacity of thc dry~rell floor sump pump is 50 gpm and the capacity of the drywell equip:ient su,"..p pump is also 50 =.p.-.. Removal of 25 gpn from eith r of these sumps can be accomplished with considerable m rgin. 4 RDr >% i'CES 1, Nuclear System Leakage Hate Limits (BRA'SAH Suosection 4.10) 3,6.D/4,6,D Safet and Relief Valv s The safety and relief valves are reouired to b o,ersble above the pres-sure (105 psig) at which the core spray sys e.s is not designed to deliver full flow, Tne pressure relief system for each unit at the Drowns Ferry Nuclear Plant has been sized to m et two design bases. Firs", the total safety/relief valve capacity has been estaclish d to meet the overoress "e protection criteria of the ASiZ Cca,

Sccona, the distrioution of this required capacity between safety valves and relief valves has

'oeen se to meet design basis 4,4.4-1 of subsection 4,4 which states that the nuclear system relief valves shall prevent opening of the safety valves during normal plant isola ions and load regections, The details of the analysis which shows compliance, as mod.ified by Reference ", with the ASfK Code requirem nts is presented in subsection 4.4 of the FSAR and the Reactor V ssel Overpressure Protection Summary Technical Report submitted in Amendment 22 in response to question 4.1 dated Decem'oer 6, 1971 To meet the safety design basis, thirteen safety/relief valves have been install on unit 2 with a,otal capacity of'4.2~p of n clear boiler rated steam flow, The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure of 1285 psig if a neutron flux scram is assumed, 'his results in a 90 psig to the co"e allowable overpressure limit of 1375 psig, To meet the operaticn 1 design bes's the o al safety/relief capaci g c 84,2 r of nuclear boiler rated has been diviaed. in o 7(V~ relief (ll val;es) and 14.2'afety (2 valves) The analysis of the plant isolation transient (turbine trip 'with bypass Amendment No. 35 219

3. 6/4. 6 BABBB:

valve failure to open) assuming a turbine trip scram is presented in Reference 5. This analysis shows that the ll relief valves limit pres-sure at the safety valves to 1211 psig, well below the setting of the safety valves. Therefore, the safety valves will not open. Thi.s analysis shows that peak system pressure is limited to .1236 psig which is 139 psig below the allowed vessel overpressure of 1375 psig. 4 Experience in relief and safety valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations. The relief and safety valves are benchtested every second operating cycle to ensure that their set points are within the + 1 percent tolerance. The relief valves are tested in place once per operating cycle to establish that they will open and pass steam. The requirements established above apply when the nuclear system can be pressurized above ambient conditions. These requirements are appli.cable at nuclear system pressures below normal operating pressures because abnormal operational trans'ients could possibly start at these conditions such that eventual overpressure relief would be needed.

However, these transients are much less

severe, in terms of pressure, than those starting at rated conditions.

The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized. t REFERENCES 2 ~ Nuclear System Pressure Relief Syst: em (BFNP FSAR Subsection 4.4) e A'mendment 22 in response to AEC Question 4.2 of December 6, 1971. 3. "Protection Against Overpressure'! (ASME Boiler and Pressure Vessel

Code, Section III, Article,9) 4.

Browns Ferry Nuclear Plant Design Deficiency Report--Target Rock Safety-Relief Valves, transmitted by J. E. Gilleland to F. E. Kruesi, August 29, 1973 ~ 5. General Electric BWR Reload 2 Licensing Amendment for BFNP Unit 2, NED0-24169,January 1979. 3.6.3/4.6.6 ~Jet Pum e 6 Fai'lure of a ]et pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown followfng the design basis double-ended line break. Also, fai.lure of the diffuser would eliminate the capabi.lity to refload the core to two-thirds height level following a reci.rculation line break. Therefore, if a failure occurre'd, repairs must be made. The detection technique is as follows. With the two recirculation pumps balanced in speed to within + 5 percent, the flow rates in both recircula-tion loops will be verified by control room monitoring instruments. If the two flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity has been verified. 220 Amendment t(o. 35 S. O'AJOR ACS tOV FEATURES 5 1 SiTK FEATURL'S Browns Ferry unit 2 is located at Browne 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 Wheeler Lake at Tennessee River Nile 294 in Limestone

County, Alabama.

The minimum distance from the outside of the secondary containment b'uilding to the boundary of the exclusion area as defined in 10 CFR 100.3 shall be 4,000 feet. 5.2 REhCTOll A. The core shall consist of 364 fuel assemblies of 49 fuel rods

each, l68 fuel assemblies of 63 fuel rods each, and 232 fuel assemblies of 62 fuel rods each.

B. The reactor core shall contain 185 cruciform-shaped control rods. The control material shall be boron carbide powder (8 C) compacted to approximately 70 percent oi theoretical . density. 5.3 REACTOR VESSEL The reactor vessel shall be as described fn Table 4.2-2 of the FSAR. The applicable design codes shall be as described in Table 4.2-1 of the FSAR. 5.4 COHTA1HHENT A. The principal design parameters for thc 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 aa described in Section 5.3 of the FSAR. C. Pcnetrations to the primary containment and piping passing through such penetrat iona shall be designed in accordance with the standards set forth in Section 5.2.3.4 of the FSAR. 5.5 FUEL STORAGE E A. The arrangement of fuel in the new-fuel storage facility shall be such that k ff, for dry conditions, is less than 0.90 and flooded is fess than 0.95 (Section 10.2 of FSAR), Amendment NO 35 Q~ C2 ENCLOSURE 2 'r 0}}