ML18025B736

Proposed Tech Specs Supporting Amend to Licenses DPR-33, DPR-53 & DPR-68 Re Operation at Reduced Power W/Single Recirculation Loop in Svc.Bases & Justification Encl

ML18025B736
Person / Time
Site:	Browns Ferry
Issue date:	03/04/1982
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18025B735	List: ML18025B734 ML18025B736 ML20041E472
References
NUDOCS 8203100579
Download: ML18025B736 (119)

Text

ENCLOSURE, 1 PROPOSED 'CHANGES TO TECHNICAL SPECIFICATIONS BROWNS FERRY NUCLEAR PLANT (50-259, -260, -296)

" 820304 "8203100579 '05000259

<a"'<<PDR /DOCK

- 't l

UNIT 1 PROPOSED CHANGES

SAFETY LL'IIT LWITING SAFETY SYSTE'I S~aZNG iiEL CI MDII!G INTEGRITY 2.1,'FUEL CLADDING INTEGRITY

~Alie~ab 'lit ~il~iaal I.1ii Applies to the interrelated vari- Applies to trip uettinRs of the ables saciated Mith fueL Instruments and devices Mhich are thermal behavior. provided tc prevent th reactor system safety limits from be&S exceeded.

~Cb eat i e ~Oh act v Tp establish Limits which ensure To define the level of the process the integrity of the fuel clad- variables at uhich autanatic pro-d Leg ~ tective action Ia Initiated to are-vcnt tl.e fuel claddinS inteI;rity safety limit fran acing exceeded.

S ecificctions S ecificatian A. Reactor prcssure > 800 psia The lhaiting safety systen settings shall and Core Flaw > l0% of Rated. be as specified bcloat li'nen the reac or pressure is A. Neutron Flux Scr~~.

greater hen 800 psia, the existence of a minimum criti- 1. APLI Flux Scram Trip Setting, cal po-er r"tio (KCPR) less (Run Ilode) than 1.07 for two 'recircula-tion loop operation, or Ellen th'lode SMitch I l';08 for single-loop RL".I position, the A"-?.".

'he operation, shall constitute flux scran trip setting shall be:

violation of the fuel clad-ding integrity safety limit.

S < 0.66 (W-gW) + 54%

a herc S Setting in percent of rated thcrnal po~er (3293 NJt) 0 " Loop tecirculaticn flow ra e ia pere nt a; ated (rated loop recirculation flav rate equals 34.2a10~ lb/hr)

Difference between two loop and single-loop indicated re-circulation drive flow rate at the same core flow.

hW ~ 0 for two loop operation.

7 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING

1. 1 FUEL CLADDING INTEGRITY 2 ' FUEL CLADDING INTEGRITY or core coolant flow is less SRB < 0.66 (W- A'W) + 42K than 10$ of rated;. the core thermal*power shall not ex-, wher e:

ceed 823 MWt (about 25$ 'of rated thermal'power). SRB = Rod block setting is percent of-rated thermal power (3293 MWt)

W = Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate equals 34.2 x 106 lb/hr) h W Difference between two loop and single-loop indicated recirculation drive flow at the same core flow.

AW 0 for two loop operation.

C. Whenever the reactor is C. Scram 5 isolation- >538 in. above in the shutdown condition reactor low water vessel zero level reactor vessel, the water level shall not be less D. Scram turbine stop ~ 10 percent than 17.7 inches above the valve closure valve closure.

top of the normal active fuel zone. E. Scram turbine control valve Upon trip

1. Fast Closure of the fast acting solenoid valves.

2. Loss of Control ~ 550 psig oil pressure F. Scram low con- + 23 inches denser vacuum Hg vacuum G.

Scram main steam ~ 10 percent

'ine isolation valve closure H. Main steam isolation > 825 psig valve closure nuclear system low pressure 10

1.1 BA ES PURL CLADDING INTEGRITY SAPETY LI IT The fuel oladding represents one of the physical barriers which separate radioactive materials fxom environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion ox use-related cracking may occur during the lifo of the cladding, fission product migxati'on from this source is incrementally cumulative and continuously measurable. Puel cladding perforations, however,. can result from thermal stresses which occur from reactor operation significantly above design oonditions and the protection system se tpoints. %bile fission product migration from cladding perfoxation is just as measurable as that fxom use-related cracking, the thermally-caused cladding perfox'ations signal a threshold, beyond which

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still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding safety limit is defined in terms of the reactor operating "oonditions 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

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fuel cladding safety limit is defined with maxgin to the conditions which would produce onset transition boil ing (MCPR of 1.0). This establishes a Safety Limit such that the minimum critical power ratio greater than the safety limit represents a conservative margin re1ative to the cnnddtlons renuired to ma4nta4n fuel cladding inteuritv.

Onset of transition boiling results in a decrease in heat tx ansfer from the clad and, therefore> elevated clad temperatuxe and the possibility of clad failure. Since boiling transition is not a directly observable paxameter, the margin to boiling transition is calculated'rom plant operating parameters such as oore 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 aotual bundle power. The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (hICPR). It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrumented variables, i.e., normal plant operation presented on Pigure 2.1.1 by the nominal expected flow control line. The Safety Limit has suf f icient conservatism to assure that in the event of an abnormal operational transient intitiated from a normal operating condition (htCPR ) limits specif ied in spec ifi os t ion 3. S.K) more than 99. 9% of the fuel rods in the core are expected to avoid transition. The margin between

'CPR of 1 0 (onset of boiling

~ transition boiling) and the safety limit is derived generically in Reference 1. The MCFR fuel cladding safety limit is increased by 0.01 for single-loop operation as discussed in Reference 2.

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l, 1 .1 -

Because Ph S~S't the boiling transition correlation is based on a large quantity of full scale date, there is a very high confidence that opoxation of a fuel assembly at the condition of MCPR ~safety limit would- not produce boiling transition. Thus, although it is not, required to establ ish the safety limit additional margin the safety limit and the actual occurence of loss ofexists'etween oladding integrity.

However, if boiling transition were to occur, clad perforation would not be expected. Cladding temperatures would increase to approximately 1100oF which is below the perforation temperature o'f the cladding material. This has been verified by tes'ts in the General Eleotrio Test Reactor (GETR) where fuel similar in design to<<BFNP operated above the critical heat flux for a significant period of time (30 minutes) without clad perforation.

If reactor +1 ~ <<

pressure should ever exceed 1400 psia during normal power operating (the limit of applicability of the boiling transition correlation), it would be assumed that the fuel cladding integrity safety limit has been violated.

power, 0 flow) is greater than 4.56 ps't ht pressures below 800 psia, the core elevation pressure drop (0 low powers and flows, this pressure differential is maintained in the bypass "rogion of the core. Since the pressure drop in the bypass region is essontially all elevation head, the core pressure drop at low powers and flow will always be greater than 4.56 ps'nalyses show that with a flow of 28 x 10~ lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and. has a value of 3 ' psi. Thus, the bundle flow with a 4.56 psi driving head will be greater than 28 x 10~ Ibs/hr. Full scale ATLAS test data t'aken at pressures from 14.7 psia to 800 psia indicate. that the fuel assembly critical power at this flow is approximately" 3 .35 MWt. With the design peaking factors this corresponds to a ooro thermal power of more than 50%. Thus' core thermal power limit of 25% for reactor pressures below 800 psia is conservative.

For the fuel in the core during periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay hest. If water level should drop below the top of the fuel during this time, the ability to remove decay heat is reduced. This reduction in oooling capability could load to elevated cladding temperatures and clad perforation. As long as the fuel remains covered with water, sufficiont cooling is available to prevent fuel clad perforation.,

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1. 1 BASES The safety 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 provide adequate margin. This point corresponds approrimately to the top of the actual fuel assemblies and also to the lower reactor low water level trip (378 inches above vessel sero).

REFERENCE

1. General Electric BMR Thermal Analysis Bases (GETAB) Data, Correlation and Design Application, NEDO 10958 and NEDE 10958.

2. "Browne Ferry Nuclear Plants, Units 1, 2, and 3, Single-Loop Operation,"

NED0-24236, May 198I.

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2. 1 'ASES Analyses of the limiting transients show that no scram adjustment is required to assure the MCPR safety limit is not violated when the transient is initiated from MCPR> limits specified in specification 3.5.K.

2i APRM Flux Scram Tri Settin (Refuel or Start 8 Hot Standb Mode)

For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate ther mal margin between the setpoint and the safety limit, 25 percent of rated. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not muoh colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer and the Rod Sequence Control System. Thus, all of possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribdtion associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram leve, the rate of power rise is no more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit.

The 15 percent APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressurer is greater than 850 psig.

3, IRM Flux Scram Tri Setting The IRM System oonsists of 8 chambers, 4 in each of the reactor protection system logic channels. The IRM is a 5-decade instrument which covers -the range of power level between that covered by the SRM and the APRM. The 5 decades are covered 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 size.

The IRM scram setting of 120 divisions is active in each range of the IRM.

For 21

g .1 BASFS IRR Flux 5'era~ Trf~ Setx.in" (Conc inued) exaloplc i ~.f thec e instrument ns me Mere on range 1, the scram setcing would be at 120 divisions for that range; likeMioe, if the instrument Mas on rang e 5 the sero+

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setting Mou 1d bee 120 divisions on chat range. Thus, as the IRK is ranged up to accoltcTlo d ate th e inczease in paver level, the scram setcing is aalso s ron g ed up. h scracL at 120 divisions on the ILf instruments remains in cffcct as ion ong as the thc staztu mode. In addition, the APRM 15Z scram prevents highez pouez opex'acion without being in che RUN mode, Thee IRH sc sczam p zovides c on for c'hanges which occur boch locally and over the entire core. Thc most significant sources of reactivity change during thee pooMer increase aze

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duc to control rod Michdraval. Foz insequence concrol rod withdrawal; the f chan e of ower is sloM enough due to the physical limitation o vichdzaMing control rods, that heac flux is in equilibr um w e flux and an XRf scram would result in aWebctoz shucdovn well before any safety d d For thc case of a single concxol rod vithdravol error, a an e of zod withdrawal accidents was analy-cd. This analy~'ls include start ng the accident at various poMer levels. The uost severe ca~e in 1 a e involves an initial in which the reactor is )ust subczitical and che IRi system is noc 'ondition yec on sca 1 e.. This condition o ex'scs ac quarter rod density. Quarter rod density is illusczated in paragraph 7.5,5 of the FSAR. Addicional conse

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was taken in this analysis by assucting chat the IR1 channel close" t to the d i b ed Th zcsults of this analysis shou chat the reactor is scrammed and peak pouer lied.ted to one percent, of rated poMer, chus 'xoaincain ing the MCPR above safety limit. B'ase3 on the above analysis, the L ARM ojovpdes prote tio' against local control roc MxcnazaMal errors and concinuous vitharaua~ o control rods in sequence.

, 8. APAM Contxol Rod Block Reactor power level may be varied by moving control rods or the recirculation flow rate. The APRM system provides a control by varying block to pzevent rod wichdrawal beyond a given point at constant recir- rod culation flow rate, and thus to protect against the condition oE a MCPR less than the MCPR safety limit.

f This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excess values due to control rod withdrawal. The flow variable trip setting provides substantial margin

2,1 BASES from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 108K of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system.

C. Reactor water Low Level Scram and Isolation (Exce t Main Steamlines)

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

This level has been used in transient analyses dealing with coolant inventory decrease. The results. reported in FSAR subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel

'and the'pressure barrier, bdcause MCPR is greater than the safety limit in all cas'es,.and sy'tem press'ure does not reach the sifety v'alve settings. The scram setting is approximately 31 inches below normal operating range and is thus

. adequate to avoid spurious scrams.

D. Turbine Sto Valve Closure Scram The turbine . top valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from closure of the stop valves.

lilith a trip setting of 10K of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained even during the worst case transient that assumes the turbine bypass valves remain closed. (Reference 2)

E. Turbine Control Valve Scram

1. Fast Closure Scram This tuhbine 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 re)ection coincident wt.th failures of the turbine bypass valves. The Reactor Protection System initiateq a scram when fast closure of the control valves is initiated by the fa'st 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 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-out-of-two-twice logic input to the reactor protection system.

This trip setting, a nominally 50% greater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce 'transients very similar to that for the stop valve. No signifi<<

cant change in MCPR occurs. Relevant transient analyses are discussed in References 2 and 3 of the Final Safety Analysis Report. This scram is bypassed when turb'ine steam flow is below 30X of rated, as measured by turbine first state pressure.

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iA8t.t: 3 1 C IKSTRQHQTATtOX THAi INtttATKS LO BLOCK stto L~ %a.

Operabto Per

~trl S ~ $ Fuhctton Trt tevet Sctttor .

) t(li ltLt Upscale (ttou lies) 0.66 (M-bM) + 42X (note 2) 2(L) APRH Vpacale (Stsrtup Node) (8) 2(I) APRX Douascale (9) 2CL) APOC lnoperattve. (LO >

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EP) 1St Vpscale (Plou Btaa) 0.66 (M-hg)+ 40X (note 2)

~ -- .-'-I(I) RAN Dounocale (9) s $Z ASK looperattve (LO )

2(L) ~ It't Upscale (8) < IOB/12$ of full scale IOf Downscale (3) (8) 5/125 of ful1 ocalo IRH Detector not fn Startup Fosftfoo {8)

(IL) ttR tnoperattvc (8) (10 )

X(t) (6) Q!l VpacaLc (8) Ix 5

-- - --'2(L) (6) QQt Downscale (C)(8)

< IO counts/oec.

countoI ~ ec ~

2(L) {6) SRl Detector not fn Stortup Poeftfon (4)(d)

- 2(2)(C) SQt tnoperittvc (8) (104) 2(l) Plou bl s Cceparotor < IOX dtffercncc in rccirculatfon floes 2(l) ~ Plov 8Ias Vpacole < ILDI recfrculattou flou Rnd block Loctc ~

H/A WAS HestratnL

{pS-Ss-6>A a 147 pslg turbtne PS-S5-618) ftrst stage prcssure (approx:,-.stall )Ot po er)

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~ 1 NOTES FOR TABLE 3.2.C For Che startup and run positions of the Reactor Mode Selector Switch, there shall. be two operable or tripped trip systems for each function.

The SRM, IRM, and APRM (Startup mode), blocks need not be operable in "Run" mode, and the APRM (Flow biased) and RBM rod blocks need not be operable in "Startup" mode. If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition last longer Chan seven days, the system with the inoperable channel shall be tripped.

If the first column cannot be met, for both trip systems, both trip systems shall be tripped.

2. W is the recirculation loop flov in percent of design. Trip level setting is in percent of rated power (3293 HWt) i See Specification 2.1 for APRM control rod block setpoint.

AW is the difference betveen tvo loop and single>>loop indicated recircu-lation drive flov rite at the same core flow rate. During single-loop operation, the reduction in trip settings (0.66hW) is accomplished by correcting the flow signals to the RBM and to the flow biased APRM scram system, to preserve the original (two loop) relationship between the RBH and APRM setpoints and the total recirculation flow, or Qy properly biasing the RBM trip and the APRM scram settings.

hW << 0 for tvo loop operation.

3. IRM dovnscale is bypassed vhen it iu on its lowest range.

4. .This function is bypassed when the count rate is > 100 cps and IRM above range 2.

5. One instrument channel; i.e., one APRM or IRM or RBM, per trip system may be bypassed except only one of four SRM may be bypassed.

6. IRH channels A, E, C> G all in range 8 bypasses SRM channels A 6 C functions.

IRM channels B, F, D, H all in range 8 "bypasses SRM channels B 6 D functions.

7. The folloving operational restraints apply to the RBM only:

a. Both RBM channels are bypassed when reactor power is < 30X.

b. The RBM need not.be operable in the "startup" position of the reactor mode selector svitch.

c. Two RBM channels are provided and only one of these any be bypassed from the console. An RBM channel be out of service for testing and/or maintenance prov'ided thismaycondition does not last longer than '24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. in any tljirty day period.

d. If minimum conditions for Table 3.2.C are not mqt, administrative controls shall be immediately imposed to prevent contrbl rod withdrawal 74

3.2 QASBS The HPCI high flow and tempera'ture instrumentation are provided to deteot a break in the HPCI steam piping. Tripping of this instrumentation results in aotuation of HPC) isolation valves.

Tripping logic for the high flow is a 1 out of 2 logic, and all sensors aro required to be operable.

High temperature in the vicinity of the HPCI equipment is sensed by 4 se'ts of 4 bimetallic temperature switches. The 16 temperature switches are arranged in 2 trip systems with 8 temperature switohes in each trip system.

Tho HPCI trip'settings of 90 psi for high flow and 2004F for high temperature aro such that core unooyery is prevented and fission product rolease is within limits.

The RCIC high flow and temperature instrumez,tation are arranged

'he same as that for the HPCI. The trip setting of 450 inoh H20 for high flow and 2004F for temperature are based on the same oriteria as the HPCI.

High temperature at the Reactor Cleanup System floor drain could indicato a break in the cleanup system. When high temperature ooours. the cleanup system is isolated.

The instrumenta tion which initiates CSCS action is arranged in a dual bus system. As for other vital instrumenta tion arranged in this fashion the specification preserves the effectivoness of the system even during periods when maintenance or testing is being performed. An exception to this is when logic functional testing is being performed.

Tho oontrol rod block functions are provided to prevent, excessive oontrol rod withdrawal so that MCPR does not decrease below the safety limit. Tho trip logic for this function is 1 out of any trip one of six APRM's, eight IRM's, or four SRM's will n.'.g

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result in a rod block.

Tho minimum instrument channe 1 requirements assure sufficient, instruments tion to assure tho s ingle failure criteria is met ~

Th'e minimum instrument channel requirements for the RBM may be reduood by ono for maintenance testing, or calibration. This timo period is only 9% of tho o perating time in a month and does not significantly inorease the risk of preventing an inadvertent control rod withdrawal.

Tho APRM rod blook function's flow biased and prevents a I

significant reduotion in MCPR, especially during operation at

,reduced flow. Tho APRM provides gross core protection: i.e.,

~ 1'imits tho gross core power increase from withdrawal of control

-rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than the safety limit.

Tho RBM rod block function provides local protection of the core:

ice.+ tho prevention of critical power in a local region of tho oore for a singlo rod withdrawal error from a limiting oontrol rod pattern.

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does. provide the operator with a visual indication of neutron level.

The consequences of reactivity accidents are functions of the initial neutron flux. The requirement of at least 3 counts per second assures

'that any tragsient, should it occur, begins at or above the initial value of 10 of rated power used in the analyses of transients from cold conditions. One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal. A minimum of tw'o operable SRM's are provided as an added conservatism.

5. The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high per level operation. Two RBM channels are provided, and one of these may be bypassed from the console for maintenance and/or testing.

Automatic rod withdrawal blocks from one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due 'to rod withdrawal errors when this condition exists.

A limiting control rod pattern is a pattern which results in the core being on a thermal hydraulic limit, (i.e., MCPR given by Specification 3,5,k or LHGR of 13.4 kw/ft.

During use of such patterns, it is

)udged that testing of the RBM system prior to withdrawal-of such rods to assure its operability will assure that improper withdrawal does not occur. Xt is normally the responsibility .

of the Nuclear Engineer to identify these limiting patterns and the designated rods ei.ther when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns. Other personnel qualified to perform these functions may be designated by thc plant superintendent to perform these functions.

Scram Insertion Times The control rod system is designated to bring the reactor subcritical at 'the rate. fast enough to prevent fu<<l damage: i.e., to prevent the MCPR from becoming less than the safety limit. The limiting power transient is t

given in Reference 1. Analysis of this transient ~hews that the negative reactivity rates resulting from the scram witn ". "verage response of. all the drives as given in the above specification provide the required protection, and MCPR remains greater than the safety limit.

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

LIMITING CONDITIONS POR OPERATION SURVEILLANCE RE UIREMlBITS 3.5aK Minimum Critical Power Ratio 4.5.K. Minimum Critical Power Ratio CPR ~{MCPR The minimum critical pover ratio (MCPR) as

1. MCPR shall be determined daily during a function of scram time and core flow, reactor power operation at  ?'25X rated shall be equal to or greater than shown in thermal power and folloving any change one vv.vv recirculation loop in operation, in power level or distribution that would cause operation with a limiting these values should be increased by control rod pattern as described in the 0.01) multiplied by the Kf shown in bases for Specification 3.3.

Figure 3.S.2, vhere:

2. The MCPR limit shall be determined for g~ 0 or ave -7 B , vhichever is each fuel type 8X8, 8XBR, P8X8R, from

~A 7 B greater figure 3.5.K-1 respectively using:

7 A OR90 sec (Specification 3.3.C.l scram a. ~~ 0.0 prior to initial scram time limit to ZOX insertion from fully time measurements for the cycle, vR RMRvava) performed in accordance v'ith specification 4.3.C.1..

73r0.710+1.65 t M {0.053) LRaf 5$

b. 7 as defined in specification 3.5.K folloving the conclusion of 7ava j/ n each scram time surveillance test required by specifications 4.3.C.1 n ~ number of surveillance rod tests and 4.3.C.2.

performed to date in cycle (including BOG .test). The determination of the limit must be completed vithin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of Scram time to ZOX insertion from I each scram time surveillance required fully vithdrawn of the ith rod. by specification 4.3.C.

N ~ total number of active rods measured in specification 4.3.C.1 at BOC If at any time during steady state operation it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shall be initiated vithin 15 minutes to restore operation to vithin the prescribed limits. If the steady state MCPR is not returned to vithin the prescribed limits

'vithin tvo (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 limits.

I 160

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I

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Limitinc Conditions for Ooeration Surveillance Reo i)2emants 3.5 Core and Containment Coolin(); S stems 4.5 Core and Co ta ent L. APRM Setooints L. APRH Setnoints

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1. Whenever the core thermal FRP/CM." LPD shall power is > 25~~ of'ated, the be determined ration of FRP/CMFLPD shall be daily when the or the APRM scram and '>'1.06 reactor is~

rod block setpoint equations thermal 25'f'ated listed in sections 2. 1.A and power.

2.1.9 shall be multiplied by FRP/CMFLPD as f'ollows:

' 'FRP Wc (0.66(W-4W)+542)~~

c (0.66{W-4W)+422)

CNPLPW

2. Rien it is determined, that 3.5.L.1 is not being met, 6 hou'rs is allowed to correct .

the cond) tion.

3. Zf 3.5.L.1 and 8.5.L.2 cannot be met, the reactor power shall be reduced to c 25~+ of rated thermal power within 4 hours.

M. Report'n Reouinements If any of the limiting values

'dentified in*Specifications

.3.5.'I, J, K, or L.3 are ex<<-

ceeded and the'pecified remedial action is taken, the event shall be logged and reported in a 30-day

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writ en report.

160A c c ~ $ .6 ~ ~

5.5 BASES 3.5.H Ma i t of Fi led Dischar e Pi e If the disoharge piping of the core spray, LPCI, HPCIS, and RCICS are not filled, a water hammer can develop in this piping when the pump and/or pumps are started. To minimize damage to the discharge piping and to ensure added margin in the operation of these systems, this Technical Specification requires the discharge lines to 'be filled whenever the system is- in an operable condition. If a discharge pipe is not filled, the pumps that supply that line must be assumed to be inoperable for Technical Specification purposes.

The core spray and RHR system discharge piping high point vent is vi'su'ally checked 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 filled. In addition to the visual observation and to ensure a filled discharge line other thani's prior to testing, a pressure suppression ohambor head tank located approximately 20 feet above the discharge line high point to supply maI5eup water for those systems. The condensate head tank located approximately 100 feet above the discharge high point serves as a backup charging system when the pressure suppression chamber head tank is not in service. System discharge pressur'e indicators axe used to determine the water level above the discharge line high point The indicators will reflect approximately 30 psig for a

~

water level at the high point and 45 psig for a water level in the pressure suppression chamber head tank and are monitored daily to e'nsure that the discharge lines are filled.

When in their normal standby condition, the suction for the HPCI and RCI C pumps are a 1 i gned to the condensa te storage tank, which is physically at a higher elevation than the HPCIS and RCICS piping This assures that the HPCI and RCIC discharge piping

~

remains f illed. Further assurance is provided by observing water flow from these systems high points monthly ~

3.5 ~ I Maximum Average Planar Linear Hea t Genera t ion Rate (MAPLHGR)

This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specif ied in 10 CFR 50, Appendix K.

The peak oladding temperature following a postulated loss-of-

,coolant accident is primarily a function of the average heat

.generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within an assembly. Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than + 200F relative to the peak temperature for a typical fuel design. the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10 CFR 50, Appendix,K limit. The limiting value for MAPLHGR is shown in Tables 3 .S.I-15 '-2; -3 5 -4, and -5. The analyses supporting those limiting values is presented in References 4 and 6 ~

168

~ i 0 i

3.5iN References

1. "Fuel Densification Effects on General Electric Boiling Water Reactor Fuelt Supplements 6i 7i, and 8, HEIM-10735, August 1973.

2. Suplement 1 to Technical Report on Densification of General Electric Reactor Fuels, December 10, 1970 (USA Regulatory Staff).

3. Communication: V. A. Moore to I.S. Mitchell, "Modified QE Model for Fuel Densification," Docket 50 321, March 27, 1974.

Generic Reload Fuel Application, Licensing Topical Report)

NEDE-24011-P-A and Addenda.

5. Letter from R. H. Buchholv. (GE) to P. S. Check (NRC), "Response to NRC request for information on ODYN computer model," September 5>

1980'.

6. "Browne Perry Nuclear Plant, Units l, 2, and 3, Single-Loop Operation",

NED0-24236, May 1981.

169A

0

~ ~

Table 3.5.I-l VAPuiCR VERSUS AVERACE rueAR EXPnSURE Fuel Type: IIDDTFDL Average Planar Expoeure -

RlPLHGR Bed t ~kul I e

'00 11. 2

'18000 11.3 58000 11. 9 108000 12.1 15,000 12. 2 208000 12.

1'1.6 258000 308000 10. 9 358000 9.9 40,000 9.3 Table 3.5.I-2 HAPLHOFR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: DDDTII8 hverage Planar Exposure mZLEaa (1).

arcs/c ~buy I e 200 1,000 ll.2 3,000 11.8 10,000 12.1 15,000 12. 2 208000 12. 0 25,000 11.5 30,000 10. 9 35,000 10,0 9.3 408000 only one recirculation lloop in service multiply given

~ ~For operation Mith

.HAFLBGR .uelu)e by 0.83.

171

g ~

~ ~

Table '3.5.Z-3 MAPLHCR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: RDRB265H Average Planar ~ ~

. Exposure MAPLHGR t ~kMltt) 200 11.5 lo000 11. 6 5o000 11. 9

'0,000 12. 1 15,000 .12. 1 20o000 lie 9

, 25 F 000 11.3 30,000 10.7 35o 000 10.2 40,000 9.6 Table 3.5.I-4 MAPLHliit VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: RDRB265L and PRDRB265L hverage Planar Exposure d/t 200 1e000 ll.6 5,000 12. 1 10,000 12+ 1 15 I 000 12. 1

$ 0,000 ll.9

• 25,000 11.3 30>000 10.7 35,000 10+ 2 40,000 (I)Eor. operation vith only one recirculation loop in service multiply given MAPLHGR values by 0.82.

172

Table 3.5.1-5 WLPLHCR VERSUS AVERAGE PLANAR EXPOSURE Puel Type: P80RM84L, CLTA-1, CLTA>>2 Exposure MAPLHGR 0)

~>>d/t ~kW Et 200 11.2 10QO ll.3 50QQ. ~ 11.8 10,000 12.0 15>000 12.0 20>000 11.8

'5,000 11;2 30,000 10. 8 35>000 10.2 40,000 9.5 (I)For operation with only one recirculation loop in service multiply given MPLHGR values by 0.82.

172a

LIMITING CONDITIONS I OR OPERATION S11RVEILI hNCE REOUIRENENT 4.6. E Jc t Porn~a

+,6,7 Recirculation 'ration b. The indicated value of core flow xate varies

1. The reactor sh'all not be from thc value derived operated with only one from loop. flow recirculation loop in mcasuremcnts by more service for more than,24 than 10%.

hours, unless designated adjustments for single>>loop c. The diffuser to lower operation are made in APRM rod block and scram set-plenum differential pxcssure reading on an points (Technical Specifi-cation 2.1.A, 2.1.B, and individual jet pump Table 3.2.C),

varies from the mean of (Table 3.2.C),

RBM setpoints fuel all jet,pump MCPR cladding integrity safety diffexential pressures and operating limits (Tech- by more than 10%.

nical Specification 1.I.A) 2. Whenever there is and MAPLHGR limits (Tech-nical Specification 3.5.I). rocixcula tion flow with the If this specification cannot reactor in the Startup or be met within the stated Run Node and one 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then the plant recirculation pump is shall be placed'n hot shut- operating with the down within an additional equalizer valvo closed, the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the adjust- diffuser to lower plenum ments are made sooner, differential pressure shall bc checked daily and the I'ollowing one pump differential pressure of an operation the discharge individual jet pump in a valve of the low speed pump loop shall not vary from may not b c opened unless the mean of all jot pump the speed of the faster differential pressuxes in pump is 1 ess than 50% of that loop by more than its rated speed. 10%.

3 S toady state operation with Recirculation Pum 0 eration bet h rc c i r cu1 ation pumps out of service for up to 12 1. Recirculation pump speeds hours is permitted. During shall bc checked and logged such interval restart of at least once per day.

the recirculation pumps) i s pcrmi t ted, provided the loop discharge temperature is within 75oP of the saturation tempcraturc of the reactor vessel watex as determined by domo pressuro. Thc total elapsed time in natural circulation and one pump operation must be no greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

G. Structura Intc rit G. Structural Intc rit 4 .6.h together with The structural integri ty of supplementary notes, thc,pr imary system shul 1 be specifics thc 182

.8.6/4.6 BASES If they do di f fez'y 10 percent or more, the coro flow rate

~

measur'ed by the jet pump diffuser differential pressure system must. be cheoked against the core flow rate derived from the moasux'ed values of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percont or moro (with the derived value highex) diffusor measuremonts will be taken to define the location within the

'vessol of failed jet pump nozzle (or riser) and the unit shut down for ropairs. If the potential blowdown flow area is inoreaseds tho system resistanoe to the xecirculation pump is also reduoeds hence, the affected drive pump will 'run out'o a substantially higher flow rate (approximately 115 percent to 120 poroent for a single nozzle failuxe). If the two loops are balanoed in flow at the same pump speed, tho xesistance charaoteristics cannot have ohanged. Any imbalance between drive loop flow rates would be indica ted by the plant process instrumentation. In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicatod by a positive differential pressure but the not affoot would be a slight decxease (3 percent to 6 perco'nt) in the total coro flow measured. This decrease, tog,ethox <<ith the loop flow increase, would result in a lack of correlation between measured and dex'ivod cox'e flow rate. Pinally, the affected,jet pump diffuser differential pressure signal would be reduced beoause the baokflow would bo less than the normal forward flow.

A nozzle-risor system failure could also generate 'the coinoident failure of a jet pump diffuser body; howevex', the converse is not true. Tho lack of any substantial stress in the jet pump diffusor body makes failure impossible without an initial nozzle-riser system failure.

3. 6. F/4. 6. F R on P 0 e a o Steady-state operation without forced recirculation will not bo 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 reoirculation pump from tho natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 754F. This reduoes the positive reactivity insertion to an acoeptably low value Single-loop oporation is permitted for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without making the adjustmonts in core limits and rod lines doscribed,in speoification 3.6.F. This is justified by the fact, that in most

.oases, recixculation pump maintenance can be performed in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Por periods of extended operation; the adjus tmonts in specifica tion 3 .6.F. should be made. It is noted that those restrictions are conservative compared to two pump operation in all cases.

221

UNIT 2 PROPOSED CHANGES

LQIITIVC SAFETY SYSTEll SETTDG SAFETY LLIIKT 2,1 FUEL CLAUDIUC IhaEGRITY

l. 'siEL CI MDItlC IIITEGRITY A licabilit A 1 icah ilit Applies to the interrelated vpri- Applies to trip settin8s of thc ables <<ssociatcd vith fuel instruaents and devic s Mhich ar>>

thermal behavior. provided tc prevent th reactor systcc: safety 1&it f rota bein8 exceeded.

~Gb tl e To establi,sh limni.ts which ensure To define the lcvcl of the process the intcRrity of the fuel clad- variables at which autonatic pro-Lcs.

tective action I 'nitiated to pr<<-

d vent th fuel claCdinS inteFr ity safety liai frets cairg exceeded.

S ccific tions S ecification Reactor Pressure > 800 psia The 1&citing safety systeta settings shall A.

of Rated. be as specified bcloM:

and Core Plov > 10%

I;nen the reactor pressure is A. Ne hara Plr "c".<<.

8rcater hen 800 psia, the Trip Setting existcnr.e of a tainir.ua criti- 1. APLI Flux Scrata (Run llod<<)

=al po;er r" tio (".CPR) less than 1.07 for two recircula-tion loop operation, or Wlan the t¹ position, RU)I Bode SMitth ir the A-"p-".

1':08 for single-loop flux scrarL trip setting shall operation, shall constitute be:

violation of the fuel clad-ding integrity safety limit. S < 0.66 (W-f,W) +, 54%

where:

S " Settin8 in percent of rated thartaal poMar (3293 Mt) 0 " I.oop rccir-uiaticn flent rate in percent o; rated (rated loop rcc'rculat'oa flora rate equals

34. 2x10 lb/hr )

gg = Difference between two loop and single-loop indicated re-circulation drive flow rare at the same core flow.

8 ~ hW = 0 for two loop operation.

0

)lltT/t5G SAFPQ Sv.'TPt SETT j lln 8 A~'P6"I 't.l>IJT HIP.I. t:t.inllte: imrlttm P l Pljgle CtdhtIDlttG THTFCRTTT

- In the event of operation with the core maximum fraction of limiting

'over density (CIPD) greater than fraction of rated thermal'ower (FYJ')

the setting shall be modified os follovs:

6 c (0.66 (W-4W) + 545) PDP l CHP(PD For no combination of loop recircu-lation flow rate and core thermal

'oser shall the APRM flux scram trip setting be alloved to exceed 120K

. of rated thermal pover.

(Note: These settings assume operation within the basic thermal hydraulic design criteria. These criteria are LHCR+18.Skv/ft for.gx8, kv/ft for 7x7 fuel and 613.4 SxSR. end PSx8R. and MCPR vithin limits of Specification 3.S.k. If it is deter(((ined that either of these design criteria is being viola. cd .

during operation, acrioo sho11 bc initiated within 15 mirutcs to'rc core ~

operation within piescribhd limpet: ~

Surveillance requirements for APF:::

scram setpnint are given in

~ pacification 4 1'-B

2.

APLt Mhen the reactor mode avicch is in the STARTUP POSITIOND the APRH scram shall be oet at loss than or equal to ISZ of .rated pover.

3, IN-The IO scram shall be set at less than or equal to 120/1ZS of full scale.

B. APRH Rod Slo"r. Trio CI4re .'hewial Pove DtPWit SctttnP'he

~t'enp Lor Preriure < OQ t4st ~ ) Af'lV>> Ro'lock trip sectjntt rhai)

E be:

l'Ion Cho reactor pressure Ls )eel than or equal,to 8 pate, I 9

'4 4 ~ ~

~

'.sAFF.i Y LItlIT 1 1 FUnL CLADDING IlliECRITY . 2 .1 FUEL CLADDIRG ItiTEGRITI or core coolant f1ov is less RB < 0.66 (W-hW) + 42K than Lt4 of rated, the core thermal pover shall not ex- 'vherc:

ceed 823 Wt, (about 25'f block setting is percent rated thecal pover). RR

~ Rod of rated thermal pover (3293 Hwt)

M . = Loop recirculation flov rate in percent of rated (ral.cd loop recircula)ion flcv rate equals 34.2 X l0 lb/hr) hW ~ Difference between two loop and and single loop indicated recir-culation drive flow at the sane core flow.

hW 0 for two loop operation In the event of operation with the core maximum fraction of limiting power den-sity (CMFLPD) greater than fraction of rated thermal power (FRP) the setting shall be modified as follows:

S RB < 0.66 (W-dW) + 42/ FRP C. Whenever the reac or is in CMFLPD the ahutdovn condition vith irradiated fuel in the reac- C. Scram and isoluation >538 in. above tor vessel, the vatcr level, reactor low water vessel zero level.

shall not be 1 ss than 17.7, in. 'above the top of the DE Scram turb'.nc stop < 10 percent normal active fuel xone.. valve clcsurc yelve closure E. Sera~ turbine control valve Upon trip of

l. Fast closure fast actir,-

'he solcnoid valves Loss nf co.itrol > 550 psi~,

oil pressu"e F. Scram-lov con- 23 inches denser vacuum Hg vacuum G. Serac nain stezsn < 10 percent line isolation valve closure H. Pain stew isolation > 825 psig valve closure--nuclear system lov pr=ssure.

10 y ~ ~ e ~ ~

1.1 BASES FUEL CLADDING INTEGRITY SAFETY LIMIT The fuel cladding represents one of the physical barriers which separate radioactive materials from environs. The integrity of

,this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incxcmentally cumulative and continuously measurable. Fuel cladding perforations, howevex,'an result from thermal stresses which occur from reactor opera tion significantly above design conditions and the protection system se tp'oints. Whil'e fission product migration from cladding perforation is just as measurable as that from use-related cracking, the thermally-caused cladding perforations signal a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, thc fuel cladding safety limit is defined in terms of the reactor operating conditions which can result in cladding perforation.

The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational transient. Because fuel damage is not directly observable, the fuel cladding safety limit is defined with margin to the conditions which would produce onset transition boiling (HCPR of 1.0). This establishes a Safety Limit such that the minimum Powe<<<<io greater than the safety limit represents a conservative margin relative to the. conditions reouired to ma9ntain fuel claddina intearitv.

Onset of transition boiling results in a decrease in heat transfer from thc clad and, therefore, elevated clad temperature and the possibility of clad failure. Since boiling transition is not a directly observablc parameter, the margin to boiling transition is calculated from plant opera ting parame ters such as core power, core flow, feedwater temperature, and core power, distribution. The margin fox 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 bundle powers The minimum value of this ratio for any bundle in the core is the minimum critical powex ratio (htCPR) It is assumed that the plant operation is

~

controlled to thc nominal protective setpoints via the instrumented variables, i.e., normal plant operation presented on Figure 2.1.1 by the nominal expected flow control line. The Safety Limit has sufficient conservatism to assure j that in the event of an abnormal operational transient intitiated from a normal operating condition (hfCPR ) limits specified in specification 3.5.K) morc than 99.9% of the fuel rods in the coro are expected to avoid'boiling transition. The margin between h!CPR of 1.0 (onset of transition boiling) and the safety limit is derived generically in Reference 1. The MCPR fuel cladding safety limit is increased by 0.01 for single-loop operation as discussed in Reference 2.

15

i. A ~BASE Beoauso tho boiling transition correlation is based on a large quantity of full soalo data, there is a very high confidence that operation of a fuel assembly at the condi tion of MCPR safety limit would not produoe boiling transi tion. Thus, although it is not required to establish the safety limit additional margin exists botwoon tho safety limit and the actual occurence of loss of cladding integrity.

Howevers if boiling would not be ozpeoted.

transition were to occur, clad perforation Cladding temperatures would increase to approx'imately 11004F whioh is below the perforation temperature of tho oladding materials This has been verified by tests in the General Elootrio Tost Reactor (GETR) where fuel similar in design to BFNP operated above the critical heat flux for a significant period of time (30 minutes) without clad perforation.

If reactor pressure should .ever ezceed 1400 psia during normal power operating (the limit of applicability of the boiling transi.ti.on oorrolation), it would be assumed that the fuel cladding integrity safety limit has been violated.

In addtion to the boiling transition limit oporation is oonstrainod to a maximum LHGR of 18.5 kW/ft for 7x7 fuel and 13.4 kW/ft for gzg, 8z8R, and Pgz8R. This limit is rea'ohod whon the Core Maximum Fraction of Limiting Power Density esluals 1 0 (CMFLPD ~ 1

~ '). For the case where Core Maximum Fraotion of Limiting Power Density exoeeds the Fraction of Rated Thermal Power, operation is permitted only at less than 100% of ratod power and only with reduced APRM scram settings as required by spocification 2.1.A.1, At pressures bolow 800 psia> the core elevation pressure drop (0 power> 0 flow), is greater than 4.56 psi. At low powers and flows this pressure differential is maintained in the bypass rogion of tho oore. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low powers and flow will always bo greater than 4.56 psi. Analyses show that with a flow of 28 x 10'bs/hr bundle flow, bundle pressure drop is noarly independont of bundle power and has a value oi 3.5 paid Thus, the bundle flow with a 4.56 psi driving head will bo greater than, 28 x'104 lbs/hr ~ Full soalo ATLAS test

'data. taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly oritioal power at this flow is approximatoly 3.35 MWt ~ With the design peaking factors this corresponds to a core thermal power of more than 50%. Thus, a oore thermal power limit of 25% for reaotor prossures below 800 psia is conservative.

For tho fuel in the oore during periods when the reactor is shut down, oonsidoration must also beiven to water level requirements due to the effeot of decay heat. If water level should drop below the top of the fuel during this time> the ability to remove'decay heat is reduced. This reduotion in

'oooling oapab ility oould lead to elevated cladding temperaturos and clad perforation. As long as the fuel remains coverod with waters suffioiont oooling is available to prevent fuol clad perforation.

1.1 BASES The safety limit has been established at 17.7 in'bove the top of the irradiated fuel to provide a point which can be monitored and also provide adequate margin. This point corresponds approximately to the top of the actual fuel assemblies and also to the lower reactor low water level trip (378 inches above vessel aero).

REFERENCE 1 ~ General Electric BWR Thermal Analysis Bases (GETAB) Data, Correlation and Design Application, NEDO 10958 and NEDE 10958.

2. "Browns Ferry Nuclear Plants, Units 1, 2, and 3, Sin8le-Loop Operation,"

NFD0-24236, May 198l.

17

LIMITINC SAFETY SYSTF.'t SETTINGS R..LATED TO FUEL CLADD:NC INTFCRITY abnorL'al operatfonal cransients applicable to operation of che Broun Ferry N<<I<sr p]anc have been analyzed throughout the spectrum of planned operating can-dicions up to the design thermal porer condition of 3440 Nc. The analyses uere based upon plant operation in accordance uith the operating map gfven in Figure 3. 7-]

of cne FSAE, In addition, 3293 tQc is the licensed maximum pover level Brouns Ferry Nuclear Plant, and this represents the maximum steady-state pouer uhich shall noc'knouingly be exceeded.

Conservatism fs incorporated fn the transient analyses in estimating thc controlling f ac cors ~ such as void reacclvf ty coef f i ci"nt, control rod scram scram delay cine, peaking factors, and axial pover shapes. These factors are selected conservatively- uich.respect to cheiz effect on che applicable transient rcs'ulcs as determined by the current analysis model.

transient model, evolved over many years, has been substantiated n opera-tion as a conservative cool or evaluating reactor oynamic performance.

obtained from a Cencral Electric boiling uacer reaccor,have been compazed uich predictions made by the model. The comparisions and res'Its az'e summarfaed in Reference X.

The absolute value of the void rcactiv.',ty coefficient used in thc analysis fs conscrvatfvcly estimated to be about 25% greater than the nom'nal maximum value expec:ed tu occur during thc core lifetime. The scram vorch used has been dcraccd co bc equivalent to approx=.,scely Sty- of the total scram vorth of chc control rods. The scram delay cfme and race o( rod insertion allovc(

hi + I > i ~+ v%i'0:lrc consciva tive) y sc c equal co thc longest delay ar d "lou>>

est insertion rate acceptable by Technical Specfffcatfsns.

The effect of scrsm vorth, screts delay tfme and rod insertion rate, all conservatively applied, are of greatest ofgnfffcance.

fn the early portion of the negative reactivity insertion. The rapfd fnsertfon of negative reactivity fs assured by the time requirements for 5Z and 202 fnaercion.

By the time the rods are 60i inserted, approximately fouz dollars of negative zeac-tlvicy has been inserted vhich strongly turns the transient, and accomplishes the desired effect. The times for 50% and 90% insertion are given to assure proper

. completion of the expected perfonnance in the earlier portion of the transient, co eacablfsh the ultfmate fully shutdovn steady-state condition.

For hoalyses of the thermal consequences of che cransf cnts a HCPR) 19mf ts

,".pacified in specification 3.5,k is conservstfvelv Assumed to exist prior to initiation of the transients.

chal co of using conservative valu.s of controlling parameters and inl c 'acing transients at the design pover level, produces morc pessimis i ansuers chan uoul J regni by using expected values uf control parameters and analyzing ut higher pouer "vela.

1 19

2. 1 BASES

\

The scram trip setting must be ad)usted to ensure that the LSCR transient peak is not increased for any combination of QfFLPD and FRP. The scram setting is ad)usted in accordance with. the formula in specification'2.l.h.l when the CMFLFD exceeds FRP.

Analyses of the limiting transients show that no scram adjustment is required to assure the MCPR safety limit is not violated when the transient is initiated from MCPR) limits specified in specification 3.5oK.

20 APRM Flux Scram Tri Settin (Refuel. or Start & Hot Stnndb Mode)

For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate thermal margin between the setpoint and the safety limit, 25 percent of rated. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns ar e constrained to be uniform by operating procedures backed up by the rod worth minimizer and the Rod Sequence Control System. Thus, all of possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distrib8tion associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change .power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram leve, the rate of power rise is no morc than 5 percent of rated power pcr minute, and the APRM system would bc more than adequate to assure a scram before the power could exceed the safety limit.

The 15 percent APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when reactor pressurer is greater than 850 psig.

34 IRM Flux Scram Tri Setting The IRM System consists of 8 chambers, 0 in each of the reactor protection system logic channels. The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are covered 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 size.

The IRM scram setting of 120 divisions is active in each range of the IRM.

For 21

0' Rol bhsKs example, "if the instrument vere on rnnge 1, the scram netting vould be ct 12D divisions for that range; likevioe, if the instrument vas on range S, the scram setting vould be 120 divisions on that range. Thus, as the IlR is ranged up to accawnadatc the inczease in paver level, the sctam setting is also ranged up. A scram nt 120 divisions on the IRti in truments temains in effect as long as the or is in the stnttup mode. In addition, the AFRM ISZ scram prevents r actor reac hi h ver operation vithout be'ing in the RUN mode, The IRH scram provide s protcccion for changes vhich occur both locally nnd over the entire+ core. Thee

~

most significsnt sources of reactivity change during the paver increase are due to cont'zol rad vithdrnvnl. Foz insequence control rod vithdtnval, the rate of change of paver is slav enough due ta the physical IM.tation af vithdraving control rods, that heat flux is in equilibrium vith the neutton flux and an IRH scram vould result in a tcactor shutdovn veil before any safety limit is exceeded.'or the case ot a single control rod vithdravcl errar, a

~"

range of rod vithdraval nccidents vas nnaly'red. This analyils includod starting the accident at variaus pover levels. The most severe ci'e involves an initial condition in vhich the reactor is )ust subcritical and the IR'! system is nat yet an scale. This conditian exists nt quarter rod density. Quarter rad density is illustrated in paragraph 7.S.S of the FSAR. Additional conservntism vss taken in this anplysis by assuming that the IRi channel closest to the vithdravn zod is bypassed. The results af this analysis shou that the tcnctat is scrammed an) peak paver limited to one percent of raced pover, thus maintaining MCPR above safety limiK,Based an the above analysis, the IRM provides protection against loc'al 'control rod withdr'awal errors..and continuous withdrawal of control rods ia sequence;

b. AF RH Control. Rod Black Reactor paver level may be varied by moving contro1 rods or by varying the tecltculation flov rate. The APRH system provides a control rad bio& to prevent rod vithdre:al beyond a given point at constant zecir- .

cuclotian f lou tate. and thus ta protect against the condition of a HCFR less chan the safety limit. This rad'black trip setting,'hich.is automatically v'arried vlth recirculation loop flow rate. prevents nn increase tn thd reactor paver Level. to excess values due to'ontrol rod vith-drevnl. The flov variable" trip setting provides substantial mnr,<in W

22

'a

~ g ~ i

2. 1 BASES from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 108% of rated thermal power becau'se of the APRM rod block trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system.

C. Reactor water Low Level Scram and Isolation (Exce t Main Steamlines)

The set point, for t'e low level scram is above the bottom of the separator skirt.

This level has been used in transient analyses dealing with coolant inventory decrease. The results reported in FSAR subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than the safety limit in all cases,. and sy'tem pressure does not reach the safety v'alve settings. The scram setting is approximately 31 inches below normal operating range and is thus adequate to avoid spurious scrams.

D. Turbine Sto Valve'Closure Scram The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from 'closure of the stop valves.

With a trip setting of 10% of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained

. even during the worst case transient'hat assumes the turbine bypass valves remain closed. (Reference 2)

E. Turbine Control Valve Scram Fast Closure Scram This tulbine 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 refection 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 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-out-of-two-twice logic input to the reactor protection system.

This trip setting, a nominally 50% greater closure time and a different valve characteristic from that of the turbine stop valve, combine to produce transients very similar to that for the stop valve. No signifi-cant change in MCPR occurs. Relevant transient analyses are discussed

~ in References 2 and 3 of the Final Safety Analysis Report. This scram is bypassed when tukb'ine steam flow is below 30% of rated, as measured by turbine first state pressure.

23

TABM 3,2.C IUSGtlOfENTATIOM THAT IHITIATES ROD BMCKS

~

NIoIcaaa boi Operable Pcr Fubct Ion Trt Level Set tfa 2(1) APRH Upscale (Plou Btas) 0.66 (H-511) + 42X (note 2)

.. ~(I) APRH Upscale (Startup Hade) (8) < 12X 2(1) APRH Dovascalc (9) > 3I 2(1) APRH InoperatIve (lob)

'L ~

1(2) R& Upscale (Plou BLaa) 0.66 (P-hM) + 40X (note 2) 1(l) iQH Doveacalc (9) > 3Z 1 (1) RDH.Ieoperattve (10 )

3(l) IRH Upscale (8) <108/12S of full scale 3(L) ZRH Dovoacale (3) (8) > y/125 of full ecale 3(1) IRH Detector aot Ia Stertup Poeitioa (8) (Ii) .

3(1) IRH Inoperative (8) (10 )

2(1)(6) 5

, SRH Upacalc (8) < 1 x 10 couate/oec.

2(1) (6) SRH Dornacale (4) (8) 3 counts/eeet 2(1) (6) SRH Detector aot ia Startup PoettCoa (4)(8) (ll)

. -."2(1) (6) SRH Iaoperatfvc (8) (105) 2(1) .

Flou Mes Ccaparotor <lOI dkffercacc Cn rcctrculatlo<<leuc PIou BIao UpocaIo < lLOX recQ'culatloa Clou Rod'lock LoeIc H/A f4CS Hestreia>

(%AS-6XA a 147 paid turbine SS4S-61n) first ste8e pressure (approxIaetely 30I pover)

0

~ ~

N S FOR TABLE 3.2.C

1. For the startup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function. The SRM, IRM, and APRM (Startup mode), blocks need not be operable in "Run" mode, and the APRM (Flow biased) and RBH rod blocks need not be operable in "Startup" mode.

. If the first column cannot be met for one of the two trip systems, this condition may exist for up to'seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition last longer than seven days, the system with the inoperable channel shall be tripped. If the first column cannot be met for both trip systems,

~

both trip systems shall be tripped.

2. W is the recirculation loop flow in 'peicent of design. Trip level setting is in percent of rated power (3293 MWt). A ratio of FRP/CMFLPD <1.0 is permitted at .reduced power. See Spe'cification 2.1 for APRM control rod block setpoint.

AW is the difference between two loop and single-loop indicated recirculation drive flow rate at the same core flow rate. 'uring single-loop operation, the reduction in trip settings (0.66AW) is accomplished by correcting the flow

'ignals to the RBM and to the flow biased APRM scram system, to preserve the

'riginal (two loop) relationship between the RBH and APRM setpoints and the total recirculation flow, or by properly biasing the RBH trip and the APRM scram settings.

AW ~ 0 for two loop, operation.

3. IRM downscale is bypassed when it is on its lowest range.

4. This function is bypassed when the count rate is > 100 cps and IRM above range 2.

5. One instrument channel; i.e., one APRM or IRM or RSM, per trip system may be bypassed except only one, of four SRM may be bypassed.

6. IRM channels A, E, C, 6 all in range 8 bypasses SRM channels .A and C functions.

IRM channels B, F, D, 8 all in range 8 bypassed SRH channels B and D functions, 7.. The following operational restraints apply tp the RBM only.

a. Both RBM channels are bypassed when reactor power is < 30X.

b. The RBH need not be operable in the "startup" position of the reactor mode selector switch.

ci Two RBM channels are provided and only one of these may be bypassed from the console. An RBH channel may be out of service for testing and/or maintenance provided this condition 'does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any thirty day period.

d. If minimum conditions for Table 3.2.C are not met, administgative controls, shall be immediately imposed to prevent control rod withdrawal.

74

~ ~

'I

0

't 0

3.2 BASES The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves.

Tripping logic for the high flow is a 1 out of 2 logic, and arc requixed to be operable. a11'ensors High temperature in the vicinity of thc HPCI equipmcnt is sensed by 4 sots of 4 bimetallic temperature switches. The* 16 temperature switches are arranged in 2 trip systems with g temperature switches in each txip system.

Thc HPCI trip settings of 90 psi for high flow and 200oF for high temperature are such that core uncovery is prevented and fission product release is within limits.

The RCIC high flow ~ and temperature'i nstxumentation ax'e arranged the same as that for the HPCI. The trip setting of 450 inch H20 for high flow and 200oP for temperature are based on the same criteria as the HPCI.

High temperature at thc Reactor Cleanup System floor dxain could indicate a break in the cleanup system. When high temperatuse occurs> the cleanup system is isolated.

The instrumentation which initiates CSCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this fashion the specification preserves the effectiveness'f the system even during periods when maintenance or testing is being performed. An exception to this is when logic functional testing

'is being performed.

The control rod block functions are provided to prevent excessive control rod withdrawa) so that MCPR does not decrease below the

,safety limit. The trip logic for this function is 1 out of n'.

e.g., any trip onc of six APRM's, eight IRM's, or four SRM's will result in a rod block.

The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met.

The minimum instrument channel requirements for the RBM may be reduced by one for maintenance testing, or calibration. This time,period is only 3% of the* operating time in a month and docs not significantly increase the risk of preventing an inadvertent control rod withdx'awal.

The APRN rod block function's flow biased and prevents a signi f icant reduction in NCPR, especially during opera t ion at rcduccd flow. The APRM provides gross coro protection: 1 e ~ ~ y limits the gross coro power increase from withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than the safety limit.

The.RP>M rod block function provides local protection of the core:

ice.; the prevention of critical power in a local region of the core for a single xod withdrawal error from a limiting control rod pattern.

0

BASES:

does provide the operator with a visual indication of neutron level.

The cons'equences of reactivity accidents are functions of the initial neutron flux. The requirement of at least 3 counts per second assures that any transient, should it occur, begins at or above the initial value of 10 of rated power used in the analyses of transients from cold conditions. One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered

.control rod withdrawal. A minimum of two operable SRM's are provided as an added conservatism.

5. The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high per level operation. Two RBM channels are provided, and one of these may be bypassed from the console for maintenance and/or" testing.

Automatic rod withdrawal blocks from one of the channels will block erroneous zod 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.

I A limiting control rod pattern is a pattern which results in the core being on a thermal hydraulic limit, (i.e., MCPR given by Specification 3.5.k oz LHGR of 18.5 kw/ft for 7 x 7 or 13.4 for 8 x 8, 8 'x 8R 6 P8 x 8R fuel]. During use of such patterns, judged that testing of the RBM system prior to withdrawal-it is 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 th designated rods either when the p'atterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns. Other personnel qualified to perform these functions may be designated by the plant superintendent to perform these functions.

Scram Insertion Times The control rod system is designated to bring the reactor subcritical at the rate fast enough to prevent fuel damage: i.e., to prevent the MCPR from becoming less than the safety limit. The limiting power transient is given in Reference 1. Analysis of this transient shows that the negative reactivity rates resulting from the scram with the average response of all

~

the drives as given in the above specification provide the required protection/

and MCPR remains greater than the safety limit.

On an early BWR, some degradation of control rod scram performance occurred during plant st'artup and was determined to be caused by 131

SURVFII.LANCE RF. UIRI'.CLIENTS I.IHITING CONDITIONS FOR OPERATION LIII:II < I.IIGI:.t I - (AP/I') (L/I.T))

LIIGIL ~ l)csijn LIIGR ~ 18.5 kW/ft d

(Ap/p) max o IIaximum power spiking penalty p p2(

LT Total core length 12. 0 ft L ~ Axii) position abov" bottom of core If at .iny tine during operation it is deter" mined by normal surveillance that the limiting valuo for L!IGR is being exceeded, action shall bc 'initiated Mithin 15 minutes to restore operation to Mithin the prescribed limits.

If the LIIGR is not Mithin returned to Mithin the prescribed limits t<o (2) hours, the shall be brought to the Cold Shutdown 'eactor condl.tion Mithin 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shill continue until rr,actor operation is Mithin the prescribed l.issue,ts ~

trItical Pover Rat fo (HCPR R. Htnloun limit for BFNP 2 cycle K. Minimum Critical Power Ratio The operating MCPR 1.27 for 8X8, 8X8R, and (HCPII) is 1;32 for 7X7, 4

P8X8R fuels. These limits apply to steady shall be determined daily state power operation with at rated power and MCPR during reactor power operation at flow. For operation only one recir- 25X rated thermal poMer and these values culation loop in operation, For core following any change in power. level should be increased by 0.01. stated above, dr distribution that vould cause flow conditiorsother than operation with a limiting contrcl greater than the above rod pattern as described in the the MCPR shall be shown in limits times Kf. Kf is the value bases d'or Specification 3.3.

Figure 3.5.2. i If at any normal time during operation it is the deter-limit-mined by surveillance that action is being exceeded, ing value for MCPR within shall be initiated to within the prescribed 15 minutes to .-

restore operation MCPR is not return-limits. If the steady prescribed limits within ed to within the shall be two (2) hours, the reactor condition Shutdown brought to the Cold surveillance and corres-within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, continue until reactor ponding action shall prescribed limits.

operation is within the L. P cnor inc Reouiremcnts valves identified

. I.'any of the limiting 3.5.1, J, or K are i.n Specifications action is exceeded and the specified and taken, the event shall be logged report.

reporfed in a 30-day Mrittcn 160

0 AS NB 5 .5,. BASES 3.5.H Maintenance of Filled Dischar e Pi e If the discharge piping of tho core spray, LPCI, HPCIS, and RCICS are not filled, a water hammer can develop in this piping when

.the pump and/or pumps are started. To minimize damage to the discharge piping and to ensure added margin in the operation of these systems, this Technical Specification requires the discharge lines to be filled whenever the system -is in an oper'able condition. If a discharge pipe is not filled, the pumps that supply that line must be assumed to be inoperable for Technical Speoifioation purposes.

The core spray and RHR system discharge piping high point vent is visually checked 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 filled. 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 high point to supply makeup water for these systems. The condensate head tank located approximately 100 feet above thc discharge 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 will reflect approximately 30 psig for a

~

water level at tho high point and 45 psig for a water level in the pressure suppression chamber head tank and are monitored daily to ensure that the discharge lines are filled.

When in their normal standby condition, the suction for the HPCI and RCIC pumps are aligned to,the condensate storage tank, which is physically at a higher elevation than the HPCIS and RCICS.

piping. 'his assures that the HPCI and RCIC discharge piping remains filled. Further'ssurance is provided by observing water flow from these systems high points monthly.

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)

This specification assures that the peak cladding the postulated design basis loss-of-coolant accident temperature'ollowing will not e'xceed the limit specified in 10 CFR 50, Appendix K.

The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial

,location and is only dependent secondarily on the rod to rod power distribution within an assembly. Since expected local variations in power distribution within a fuel assembly affect the 'calculated peak clad temperature by less than + 20OF'elative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within tho 10 CFR 50, Appendix K limits The limiting value for MAPLHGR is shown in Tables 3 5 I- ~ ~

2 ~ 35 4 and -5 ~ The analyses supporting these limiting value's is presented in References 4 and 5.

168

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3.5.J. Linear Heat Generation Rate (LHGR)

This specification assures that the Linear heat generation rate in any rod is Less than the design linear heat generation il'ueL pellet densif'ioation is postulated. The power spike penalty specif'ied is based on the analysis presented in Section 3.2.'1 of Reference 1 as modified in References 2 and 3, and assumes a linearly increasing variation in axial gapa between core bottom and top, and assures with a 95$ conf'idence, that no more than one fuel rad exceeds design Linear heat generation rate due to power spiking. The LHGR (as a f'unction ol'ore height f'r 7x7 fuel and as a constant for Bx8, BxBR, and PBxBR f'uel) shall be checked daiLy during reactar operation at > 25$

power to determine if fuel burnup, or control rod movement has caused changes in power distribution. For LHGR to be a limiting value below 25$ rated thermaL power, the R f'actor would have to be Less than 0.241 which is pr'ecluded by'a considerable margin when employing any permissible oontrol rod

'attern.

3.5.K. Minimimum Critical Power Ratio (MCPR)

At core thermal pawer Levels Less than or equaL to 25$ , the reactor wilL be operating rrt minimum recirculation pump speed and the modeiator void content, will.be very smaLL. For all designated control rod patterns, which may be employed at this point, operating p'lant experience and thermal hydraulic analysis indicated that the resulting MPCR value is in excess af'equirements by a considerabLe margin. Mith this low void content, any inadvertent core

. f'Low increase would only place operation in a more conserative made relative

~ to HGPR. The daily requirement f'ar calcuratin(r HGPR above 25$ rated thermal

~

power is suf'ficiant since power distribution shifts are very slow when there have not been signif'icant power or control rad changes. The requirement for

. oalculating MCPR when a Limiting control rod pattern is approaohed ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.

r3.5.L. Re ortin Requirements The LCO's associated with monitoring the fuel rod operating *conditions are required to be met at aLL times, i.e., there is no allowable time in which the plant can knowingly exceed the limiting values f'r HAPLHGR, LHGR, and It is a'equireient, as stated in Specifications 3.5.Iis , determined .J., and .K.,

~

MCPR.

that if't any time that the Limiting values during steady f'r state power oper atian, it.

HAPLHGR, LHGR, ar 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 than an operating Limit has been reached. Each event involving steady state operation beyond a specified Limit shall be logged and reported quarterly.

It must be recognized that there isoralways an actian which would return any MC?R) to withi'n prescribed Limits, of the parameters (MAPLHGR, LHGR, namely power reduction. Under most circumstances, this will not be the only alternative.

3.5aM'ef'erences

1. "Fuel Densification'ffects on General Electric Boiling Mater Reactor Fuel," Supplements 6 ~ 7~ and 8, NEDM-10735, August 1973.

2. Supplement 1 to Technical Report on Densif'icatians of'eneral Electric Reactor Fuels, December 14, 1974 (USA Regulatory Staff'.

3. Communication: V. A. Moore to I. S. Hitcheli, "Hadif'ied GE Model for Fuel Densification," Docket 50-321', Harch 27, 1974.

4 ~ Generio'eload Fuel Application, Licensing Topical Report, NEDE-24011-P-A, and Addenda.

5. "Browns Ferry Nuclear Plant, Units 1, 2, and 3, Single-Loop Operation",

NED0-%4236, May 1981.

r 169

~

'able

~ ~

~ ~

Average Planar Exposure Hwd/t

3. 5. I-1 Fuel Type:

MAPLHGR

~kM/8t MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Initial Core - T e 1 6 3 200 15.0 1,000 15. 1 5, 000 16.0 10,000 16.3 15, 000 16.1 20,000 15.4 25,000 14. 2 30,000 13. 1 35,000* 11. 8 40,000* 10. 5 Table 3.5.I-2 MAPLHGR VERSUS 'AVERAGE PLANAR EXPOSURE Fuel Type: Initial Core - T e 2 Average Planar Exposure MAPLHGR Had/t ~kW/fe 200 15. 6 1,000 15.5 5,000 16,2 10,000 16.5 15,000 16.5

'3.

200000 15. 8 25I000 . ~ 14.5 30,000 3 35s000 ll.9 40,000* 10.6 The average level of irradiation of the irradiated fuel from the reactor shall not exceed 33,000 MWD/MIU in accordance Mith Tables S-3 Ad S-4 of 10 CFR Part 51. l i

~ ~For operation +th only one recirculation loop in service multiply MAPLHGR values by 0.70. 'iven

Table 3.5.I-3 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Puel Type: 8D274l Average Planar Exposure MAPLHGR d/t ~kW/f t 200 ll.2 1,000 Il.3 5,000 11.9 10,000 12. 1 15,000 12. 2 20,000 12. 1 258000 ll.6 30,000 10. 9,,

Table 3 '. 1-4 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Puel Type: 8D2748 Average Planar Exposure MAPLHGR Had/t) ~W/kft) 200 11.1 1,000 ll. 2 5,000 11. 8 10,000 12. 1 15,000 12. 2 208000 12. 0 25,000 11. 5 30, 000 10.9 (2) For operation Mith only one recirculation loop in service multiply given NAPLHGR values by 0. 83.

172

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Table 3.5.I-5 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: 8DRB284 and P8DRB284 Average Planar Exposure -

MAPLHGR

/t ~kW/~ft 200 11. 2 1,000 11.3 5,000 11. 8 10,000 12. 0 15,000 12. 0 20,000 ll. 8 25,000 11.2 30,000 10.8 (3) For operation with only one recirculation loop in service multiply given MAPLHGR values by 0.82.

172a

LIMITING CONDITIONS FOR OPERATION SURVFTLLANCF. RFOUIREthFNT 3.6.F Recirculation Pum 0 aration b. The indicated value of

1. The reactor shall not be oper- core flow rate varies ated with only one recircula- from the value derived tion loop, in service for more from loop flow than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, unless designa- measurements by mo'e ted adjustments for single-loop than 10%.

operation are made in APRM rod block and scram setpoints c. The dif fuscr to lower (Technical Specification 2.1.A,. plenum dif f exential 2.1.B, and Table 3.2.C), RBM pressure reading on an setpoints (Table 3.2.C), MCPR individual jot pump fuel cladding integrity safety varies from thc mean of and operating limits (Technical all jet pump Specification 1.1.A) and MAPLHGR limits (Technical Specification differential pressures by more than 10%.

3.5.I). If this specification cannot be met within the stated 2. whenever there is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then. the plant shall ba placed in hot shutdown with-recirculation flow with the

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

'reactor in the Startup Node and one ox'un the adjustments are made sooner. rocixculation pump is opoxating with the equalizer valve closed, thc diffuscr to lower plenum

-2. Following one pump differential prcssure shall operation, tho di schargc bc chcckcd daily and thc valve of the low speed pump differential prcssure of an may not be opened unless individual jet pump in a the speed of the faster loop shall not vary from pump is less than 50% of the mean of all jet pump its rated speed. differential pressures in that loop by more than Steady s ta to opera t ion with 10%.

both recirculation pumps out of service for up to 12 F. Recirculation Pum 0 eration hours is poxmitted. During such interval res tart of 1. Recirculation pump speeds the recirculation pumps is shall bc checked and logged permitted', provided the at least onco per day ~

loop dischargo tempcraturc is within V5 F of thc saturation tompcraturc of the reactor vessel water as determined by dome pressure. The total elapsed 'time in natural circulation and one pump operation must be no greater than 24 'hours.

Structural Inte rit G. Structural Into rit Thc structural integrity of the primary system shall bc Tablo 4.6.A togc ther wi th supplementary note s, specifies the 182

6.6/4.6 EASES.

If they do differ by 10 percent or more, the core flow rate measured by the jet pump d if f user d if f erent ial pressure system must be checked against the core flow rate derived from the measured vilues of loop flow to core flow correlation. If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuscr measurements will be taken to define the lo'cation within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs. If the potential blowdown flow area is "increased, the system resistance to the recirculation pump is also reduced; hence, the affected drive pump will 'run'ut'o a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure). If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation. In addition, the affected jet pump would provide a leakage path past the core thus reducing the core flow rate . The reverse flow through thc inactive jet pump would .still be indicated by a positive differential pressure but the net affect would be a slight decrease (3 percent to 6 percent) in the total core flow measured. This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate ~ Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.

A nozzle-riser sy'tem failure could also generate the coincident failuie of a jet pump diffuser body; however, the converse is not true. The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.

3.6,F / 4.6.p Reel cu ion Pum 0 e ation Steady-state operation without forced recirculation wil l not bc 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 pump from the natural circulation cond it i on will not be permitted unless the temperature difference betw cen the loop to be started and thc core coolant temperature is less tEan 75 F. This reduces the positive reactivity insertion t o an acceptably low value.

Single-loop operation is permitted for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without making the adjustments in core limits and rod lines described in specification 3.6.F. This is justified by the fact, th at ln most cases, recirculation pump maintenance can be performed in less tEan 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />., For periods of e'xtended operation; the adjustments in specification 3 .6.F. should be made. It is noted that these restrictions are conservative compared to tw 0 puslp operation in all cases.

221

UNIT 3 PROPOSED CHANGES

SAFETY LIHIT LIHITING SAFETY SYSTEH SETTING L' NG INT R Y 2' OEL ADDING GRITY

'cab' A icab

Applies to the interrela ted Applies to trip settings of the var'iables associated with fuel 'instruments and devices which thermal behavior. are provided to prevent the reactor system safety limits irom being exceeded.

Ob ec ve

~beche "To define the level of the To establish limits which process variables at which ensure the integrity of the automatic protective action is fuel cladding initiated to prevent the fuel cladding integrity safety 1'rom being exceeded.

limit ec'cat o at ons The limiting safety system As Reactor Pressure > 800 settings shall be as specified psia and Core Flow > 10% below of Rated.

A eutron lux Scram Qhen the reactor pressure is greater than 800 psia, a. APRH Flux Scram Trip the existence of a minimum Setting (Run Hode) critical power ratio (HCPR) less than 1.07 When the Node Switch for two recirculation loop is in the RUN operation, or 1.08 for position, the APRH single>>loon operation, flux scram trip shall constitute violation setting shall be:

of the fuel cladding integrity safety limit. S < 0.66 (N-hW) + 54~

'here:

S a Setting in per-cent of rated thermal power (3293 HNt)

SAFETY l IYIT I IHITING ShFKTY SYSTKN SKTTIHG

2. l FUEL CLADDING INTEGRITY l I o FUEL CLADDING INTEGRITY W ~ Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate equals 34.2 X 10 lb/hr) dW Difference between two loop and single-loop indicated recirculation drive flow rate at the same core flow.

hW ~ 0 for two loop operation.

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:

FRP 6 + (0.66 (((-4M) + 54K).CNFLPD For no co()(bination of loop recirculation-flow rate and core ther()(al poDFer shall the hp$ N flux scram trip settinq be allo46ed to exceed 120% of rated thermal powers fHOTt: These settings aea(4(F)e operation within the basic ther(FDai hydraulic deeign criteria'hese criteria are L?lGA i 13.akMJft and HC>A within 1inite of specification 3.5.X.

10.

I p SAFETY LIHIT LIHITIHG SAFFTY SYSTFA GETTLHC 1 ~ 1 F L C C H CR Y 2 1 FUEL CLADDING IHTECMITY 8~ Core Therraa Po44 S, APRH Rod Block Tri Seetin R o rsssu 5800 s a Tho APRH Rod block trip Mhsn tho reactor pressure setting shall bot is less than or equal to 800 psia, or core coolant S 'e 0.66 (W-dW) + 42Z flcF~ ia less than 10% of rated, the core therF)4al RFhere:

power shall not exceed 823 Hut (about 25% of rated S+ a Rod block setting thera4al pe)RFer] ~ in percent of rated ther)))a 1 poRFer (3293 N4t) l4 Loop recirculation flow rate in percent of rated (rated loop recirculation floRF rate equals 38 2 x 10 'b/hr) dW ~ Difference between two loop and single-loop indicated recirculation drive'flow rat<<

at the same core flow.

hW ~ 0 for two loop flow C. Fover Tranalent In the event of operation with the cori To anaure that the Saf ety Ll)alt maximum fraction of limiting power den-i eat ah 1 ahed ln Spec 1 I icat ion sity (CMFLPD) greater than fraction of 1 ~ 1.A and 1.1,5 ia oot cxcocdodp each required acta(a shall be rated thermal power (FRP) the setting initiated by ito expected act(2(a shall be modified as follows:

atsoal. Tha Safety Ll)alt shall be oaau(aed to bo ceca()dad RFhon R ( (P 44 (W RW) + 422)FRP aerraa ia aceaap1iohcd by a)eaoo  ! (2(F(PD other than the oxpoctod serac

~ lanai. 4

'2

0 1.1 IIASHS FUEI. CLADDTNG TNTHGRTTY SAHHTY I,TMTT The fuel cladding represents one of thc physi cal barriers which separate radioactive materials from environs. The integrity of this cladding basxier is selated to its relat ive freedom from perforations ox'racking. Although some corx osion os usc-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perfox'ations, however, can result from therma 1 stresses which occus from reactor operation significantly ab ovc design

-conditions and thc psotcction system setpoint s. While fission product migsation from cladding perforation i s just as measurable as that from use-related cracking, the thermal ly-caused cladding perforations signal a thresho ld, beyond which still greatcs- thermal stresses may cause gx'os s sathcr than incremental cladding deterioration. Therefor e, thc fuel cladding safety limit is defined in terms of the react or operating conditions which can result in cladding perfo ration.

onset The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal opcsational transient. Because fuel damage is not directly observable, the

'fuel cladding safety limit is defined with margin to the conditions which would px oduce 'transition boiling (MCPR of 1.0). This establishes a Safety Limit such that the minimum critical power ratio greater than the safety limit represents a conservative margin relative to the condo'alons ran>>%red to ma4ora8n f>>el cladding inteeritv.

Onset of transition boiling results in a decrease in heat transfer from thc clad and, therefore, elevated clad transition temperature and the possibility of clod failure. Since boiling is not a directly observable parameter, the maxgin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater tempesature, and core power distribution. The margin for each fuel assembly is characterized by'he critical power ratio (CPR) which is the ratio of the

.bundle power which would produce onset of transition boiling divided by the actual bundle power. The minimum value of- this ratio for any bundle in the core is thc minimum critical powcx ratio (MCPR) It is assumed that thc plant operationthe is

~

controlled to thc nominal protective setpoints via instrumented variables, i.c., normal plant operation presented on Figure 2.1.1 by the nominal expected flow control linc. Theassux'e Safety Limit has sufficient consexvatism to that in the event of an abnormal operational txansicnt intitiated

'from a normal operating condition (MCPR > limits specified in specification 3.5.K) morc than 99.9% of the fuel rods in the core arc expected to avoid boiling transition. The margin between hICPR of 1.0 (onset of transition boiling) and the safety limit is derived generically in Reference 1. The MCPR fuel cladding safety limit is increased by 0.01 for single-loop operation as discussed in Reference 2.

15

1.1 I BASES Because the boiling transition correlation is based on a large quantity of full scale data, there iscondia very high confidence t ion f ~safety that limit operation of a fuel assembly at the o NCPR would not produce boiling trans i ti on. Thus, a though i t is not 1

required to establish the safety limit additional margin exists between the safety limit and the actual occurence of loss of cladding integri ty.

However, if boiling transition were to occur, clad perforation would not be expected. Cladding temperatures would increase to approximately 11004F which is below the perforation temperature of the cladding material. This has been verified by tests in the General Electric Tost Reactor (GETR) where fuel similar in design to BFNP- operated above the critical heat flux for a significant period of time (30 minutes) without clad perforation.

If reactor pressure should ever exceed 1400 psia during normal power operating (the limit of applicability of thc boiling transition correlation), it would be assumed that thc fuel cladding integrity safety limit has been violated.

In addtiom to the boiling transition limit operaAti on i s constrained to a maximum LHGR of l3 4 kW/ft.

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 Rated Thermal Power, operation is permi tted only at less than 100% of rated power and only with reduced APRM scram settings as required by sp'coif ication 2.1.A.1.

At pressures bolow 800 psia, the core elevation pressure drop (0 power, 0 flow) is greater than 4.56 psi. At low powers and flows this pre'ssure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass xegion is essentially all elevation head, the core pressure drop at .low powers and flow will always be greater than 4 .56 psi. Analyses show, that with a flow of 28 x 10~ lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.56 psi driving I

head will be greater than 28 x 10'bs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that tho fuel assembly critical power at this flow is approximately 3.35 MWt ~ With the design peaking factors this corresponds to a core thermal power of more than 50%.'hus, a core thermal power 1 imi t of 25% for reactor pressures below 800 ps i a .i s conservative.

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

16

The safety limit has been established at 17.7 in. above the top of the irradiated fuel to provide a point which can bc monitored and also provide adequate margin. This point corresponds approzimately to the top of the actual fuel asscmblics and also to thc lower reactor low water level trip (37 8 inches above vessel zero).

REFFRHNCE

1. General Electric BWR Thermal Analysis Bases (GETAB) and Design Application, NEDO 10958 and NEDE Data'orrelation 10958.

2. "Browns Ferry Nuclear Plants, Units 1, 2, and 3, Single-Loop Operation,"

NED0-24236, Hay 198).

17

~ ~

hi i..iuus 1t pruvLdi.s ad<<qu~f

~ a marqin for the tuel cl~ddlnq anteqcity safety limit yet allows operatinq marqin that reduces the possibility of unnecessary scrams ~

The scram trip setting must Ue adjusted to ensure that the LHGR transient peak ia ~ot increased for any combinat ion of CMFl.W nndFRI'. The scram settinq is adjusted in accordance with the formula 1n Specaf ication 2 1.A.1 when the CHF PI exceeds FOP.

~ ~

Analyses of the limiting transients show that no scram adjustment is required to assure the HCPR safety limit is not violated when the transient is initiated from HCPR

> the safety limit.

PRM Flux Sc am Tr Sett n el o Sta t 8 ot S and Mode For operation in the startup mode while the reac"or is, at lov pressure, the APRN scram setting of 15 percent of rat<<d Power provides adequate thermal margin between the setpoint and the safety limit, 25 percent of rated. The marqin is adequate to accomodate anticipated maneuvers associated with ~er plant startup. Effects of increasinq pressure at xero or low void content are cold ~ater from sources available during startup 'inor.

is not much colder than that already in the system/

temperature coefficients are small, and control rod pat terns are constrained to be uniform by operatinq procedures backed up by the rod ~orth minimirer and the Rod Sequence Control System. Worth of individual rods is very low in a uniform rod pattern. Thus, all of possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of. significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local p.aks, and because several rods must be moved to change power by a siqnif icant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent

,of rated power p r minute. and the APRN system would be more tnan adequate to assure a scram before the power could exceed the safety limit. The 15 percent APRN scram remains active until the mode switch is placed in the RUM position. This switch. occurs when reactor pressure is greater than 850 psig.

'P

3. IRM Fli>x Spc'an V'r t ti The IRM System consists of 8 chambers, a in earth of the reactor protectaon system logic channeLS. The 11M is a

h S-el> cad~ instrument which covers the range of paver 1'evel bctveon that covered by the SRH and the APRN. The S decades are covered by the IRM by means of a range switch and thv 5 decades are broken down into 10 ranges, each being anc-half of a decade in oixe. The IRH scram Oettin11 of 120 divisions IRN. FOr example, if i active in each range of thc the '.nstrument vere on range the scram setting would be at 120 divisions for that range; likewise, if the instrument vas on range 5 ~ the scram setting should be 120 divisions on that, range.

Thus, as the IRH is ranged up to accommodate the increase in power level the scram setting is also ranqed up. A scram at 120 divisiano on the IRN instruments remains in effect, ao long as the reactor is in the startup mode. The hPRH 15 percent scram vill prevent higher power operation vithout being in the run mode. V The IRN scram provides protection for changes which occur both locally and aver the entire core.

The most significant sources of reactivity change during the power increase are duc to control rod withdrawal.

For insequence control rod withdrawal, the rate of change of power is olov enaugh'ue ta the physical limitatipn of vithdravinq control rods, that heat flux is in equilibrium with thc neutron flux and an IRH scram would result in a reactor shutdown well before any saf ety limit is exceeded. For the case of a single control rad withdrawal error this transient hao been analyxed in paragraph 7.5.5.S of the FSAR. In order to ensure that the IRN provides adequate pratection against the single rod withdrawal error, a range of rod vithdraval accidents vas analyxed. This analysis included startinq the accident at various paver levels.

The most severe case involves an initial condition in which the reactor is 5uot oubcritical and the IRM system is not yet on scale. This condition exists at quarter rod dens ity. Quartet rod density io illustrated in paraqraph 7.5.5 af the FsAR. Additional conservatism .

vas taken in this analysis by assuming that. the IRH channel closest to the vithdravn rod is bypassed. The results af thxs analysis ohov that the reactor is scrammed and peak power limited to one percent of rated power, thus maintaxninq Ac MCPR safety limit. Based on the ebovo analysis, the IRH provides protection against local control rod vithdraval errors and continuous vithdroMal of control rods in sequence.

APRN Control Rod Black Reactor paver level may be varied by movinq control rods or by varying the recirculation flow rate. The APRH syste~

provides a control rod block to prevent rod withdrawal beyond 2l

~ ~

~ ~

qlvr n point. ar constant accarculation f low cato, and thus I ~ to pr~itict aqolnat the condition of a HcPR less than che safety limit. This'rod block trip setting, which is automatically varied wath recarculatxon loop flow rate, prevents an increaoe in trre reactor power level to excess values due to control rod

~ rathdrawal. The flow variable <rip setting provides substantial margin.from fuel da o~ge, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the Safety Limit increases ao the flow decreases for the specified trip setting versus flow relationship; therefore, the worst case HCPR which could occur durinq the steady-state operation io at l08% of rated therraai power because of the APRON rod block trip setting.

The actual power distribution in the core is established by specified control rod sequences and io monitored continuously C',

by the in core LP$ N syotea.

Reactor wa er w Levc YR!'hus. As with the'PN scram trip setting, the APRH rod block trip setting io ad5uoted dcnrnward if the CWLl'!J exceeds preserving the APRN rod block safety margin.

sc aa and k

t ce t Hai Steam s The set point for the low level ac".aa is above the bottom of t.he separator skirt. Thio level hao been used in transient analyses dealing with coolant inventory decrease. The results reported in FSAR subsection N10.5 show that scram and isolation of all process lines fexcept main steam) at this l.evel adequately protects the fuel and the pressure barrier/

I because

'fifth NCPR ia greater thar. the safety limit in all cases, and system pressure does not reach the safety valve settings. The scram setting is approximately 31 inches below the normal operating range and is thus 'adequate to avoid spurious scrams.

p Turbi e Sto Valv C osure S aa The turbfne stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would r<<suit from closure of thc stop valves.

a trip setting of lOX of valve closure from full open, the resultant fricreasc fn heat flux fs such that adequate thermal margins are maintained even during the worst case transient that asstxnes the turbine bypass valves remain closed. (Reference 2).

f, %Urbane rontrol Valve Serac

1. Fast Closure Scraa This turbine control va lvc fast closure scram anticipates the pressure, neutron flux, and heat flux fncrease that could result from fast closure of the'urbine control valves due to load refection coincident with failures of the turbine bypass valves. The Reactor Protection System fnftfates a scram when fast closure of the control valves is initiated by the fast acting solenoid valves and fn less than 30 milliseconds after the, start of control valve fast clbqure. This is achieved by the action of the fast acting solcnofd valves in rapidly reducing hydraulic con'rol 22

TASLS 3 loC ZNSTRQKXKTATZOK TSAT INITIATES ROO SLOCRS xinisus No.

Operable Psr

~c~sS ct ton T Lava st APRK Opscale (Plav bias) 0.66 (14-hH) + 42X (note 2)

APRK Opscals (Startup Node) (4) a 11S l(1) . APRK Ootmscale (9) 3S 3 (1) APRN Zooperative (10b)

Opscal ~ (Flov bLas) ~ 0.66 (u-~M) + 40X (note 2)

I 1(7) RSK RSK Dottuscale (9)

RRK Inoperative (10c)

C ZRK Opscale (4) . 5 104/1lS oC Cull scale 3 l1) ZRK Ootmscais (3) lb) KS/1lS of Cull scale

-3 (1) 'IRK Oetector Sot LK'Startup Position (4) 3 (1) IRK Inoperative (4) (10a) l (1) (4) SRK Opscale (4) ~ 1 c 14 counts/ssc.

~ l (1) (4) SRN Ootmscale (4) (4} >3 counts/sec I l1} (4) SRN Oetector not in Startup position (4) (4) (11) l(1) (4) SRN Inoperative (4) (10a)

Plov bias Cotttpacator ~ 10S 4ifference ia rscicculatlon floss Plov bias Opscale +110S cecicculation Ciov Ro4 block Logic S/A RSCS Restraint 149 psig turbine (PS SS 4)A'nd first-stage pressure PS 4$ 41B)

l NOTES FOR TABLE 3.2.C I

1. For the startup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function. The SRM, IBM, and APRM (Startup mode), blocks need not be operable in "Run" mode, and the APRM (Flow biased) and RBM rod blocks need not be operable in "Startup" mode.

If the first column cannot be met for one of the two trip systems, this condi-tion may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; condition last longer than seven days, the system with the inoperable channel if this shall be tripped. If the first column cannot be met for both trip systems, both, trip systems shall be tripped.

2. W is the recirculation loop flow 'in percent of design. Trip level setting is in percent of rated power (3293 MWt). A ratio of FRP/CMFLPD < 1.0 is permitted at reduced power. See Specification 2. 1 for APRM control rod block setpoint.

bW is the difference between two loop and single-loop indicated recirculation drive flow rate at the same core flow rate. During single-loop operation, the reduction in trip settings (0.66AW) is accomplished by correcting the flow signals to the RBM and to the flow biased APRM scram system, to preserve the original (two loop) relationship bet~can the RBM and APRM setpoints and the total recirculation flow, or by properly biasing the RBM trip and the APRM scram settings.

AW ~ 0 for two loop operation.

3. IRM downscale is bypassed when it is on its lowest range.

4. This function is bypassed when the count rate is > 100 cps and IBM above range 2.

5. One instrument channeL; i.e., one APRM or IRM or RBM, per trip system may be bypassed except only one of four SRM may be bypassed.

II

6. IRM channels A, E, C, G all in range 8 bypasses SRM channels A & C functions.

IRM channels B, F, D, H all in range 8 bypasses SRM channels B & D functions.

7. The following operational restraints apply to the RBM only:

a. Both RBM channels are bypassed when reactor power is <30X.

b. The RBM need not be operable in the "startup" position of the reactor mode selector switch.

c. Two RBM channels are provided and only one of these may be bypassed from the console. An RBM channel may be out of service for testing and/or maintenance provided this condition does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any thirty period. 'ay L

d. If minimum conditions for Table 3.2.C are not met, administrative controls shall'e immediately imposed to prevent control rod withdrawal.

77

Pressure instrumentation is provided to close the main steam isolation valves in Run,Mode when the main steam line pressure drops below 825 psig.

The HPCI hiqh flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results 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.

High- temperature in the vicinity of the HPCI equipment is sensed by 4 sets of 4 bimetallic temperature switches. The 16 temperature 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 and 200oF for high temperature are such that core uncovery is prevented and fissi'on product release is within limits.

The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI. The trip setting of 450n water for high flow and 200oF for temperature are based on the same criteria as the HPCI.

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

The instrumentation which initiates CSCS action is arranged in a dual bus system. As for other vital instrumentation arranged in this fashion, the Specification preserves the effectiveness of the system even during periods when maintenance'r testing is being performed. An exception to this is when logic functional testing is being performed.

The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to the safety limit.

The trip logic for this function is 1 out of n: e.g., any trip on one of six APRM's, eight IRM', or four SRM's will result in a rod block.

The minimum instrumeht channel requirements assure sufficient inst rumenta ion to assure the single failure criteria is met, Tvo RBN channe s are provided ard on'y'ne of these may be bypassed from the console, for

'l maintenance and/or testing provided tha" this condition does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any thirty day period. This time period is only 3$ of the operating time in a month and doe not significantly increase the risk of preventing an.

inadvertent control rod vithdraval.

The APRM rod block function is flow biased and prevents a siqnificant reduction in MCPR, .especially during operation at reduced flow. The APRM provides gross core protection; i.e. ~

limits the gross core power increase from withdrawal of control 110

rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than the safety limit.

1he RBM 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 rod withdrawal error from a limiting control rod pattern.

If the IRM channels are in the worst condition of allowed bypass, the scaling arrangement is such that for unbypassed IRM channels, a rod block signal is generated before the detected neutrons flux has increased by more than a factor of 10.

A downscale indication is an indic'ation the instrument has failed or the instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented.

The refueling interlocks also operate one logic channel, and are required for safety only when the mode switch is in the refueling position.

Foi effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapid enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate. The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip settings given in the specification are adequate to assure the above criteria are met. The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.

Two post treatment off-gas radiation monitors are provided and, when their trip point is reached, cause an isolation of the off-gas line. Isolation is initiated when both instruments reach their high trip point or one has an upscale trip and the other a downscale trip or both have a downscale trip.

Both instruments are required for trip but the instruments are set so that any instruments are set so, that the instantaneous stack release rate limit given in Specification 3 8 is not exceeded.

Four radiation monitors are provided for each unit which initiate Primary Containment Isolation (Group 6 isolation valves) Reactor Building Isolation and operation of the Standby Gas Treatment System. These instrument channels monitor the radiation in the Reactor zone ventilation exhaust ducts and in the Refueling Zone.

~ ~

sit tub> otii'\ .1liL sllmi, 5

~~ a L t'CVVa ~ Ii.'ll du dli dddei) contlervat iuwo

~

Tli . H4il Blue k tutti tot <t "v+tit l iti I d.tm~ q

~ ~ ~~ 't t lie ev tit of errt>>tuttut rod

~ ~

M>thrJrawal frotn locations of ltaqh [~er densest du provided, and one oC these od ithd al bl k om one of the ch inn ~> ls will lilock erroneous rod '~ithdrawal to ptevent fut l dtmaqe. The s .ecxfi ied restrictions Mith out o service consexvatively assure that f'uei damaqe will not, occur due to rod withdrawa wltnn this condition ettints.

A Limitinq cont.rol rod pattern is a pattern which results in tht core'heinq on a thermal hydraulic limit

(,a.:BCPL'~ the safety limit or LHCR 13.4). During use of such patterns, it is fudged that testittg of the RBH system

. prior to withdraMal of such rods to assure its operability wtll a"sure that improper withdrawal does tiot occuc. It xs normally the responsibilit f Hucl ea r fttqineer to identify these limting patterns and the desianatel rods either when the patterns are inxtially established or as they develop due to the. 'th occurr ence of inoperable contxol rods i th hami irig pattertis. Ocher personnel qualified to perform these functions may be designated by the plant, superintendent to perform these functions.

' nset'tion C. am T mes I The control rod system is designed to b rinq t e reactor subcritical aat a rate fast enouqh to prevent fuel damage;

.e., tc prevent the MCPR from becoming less th an th e safety

- limit.

Analysis of thxs tra n ient shoits that the neqative rea c tiv resultxnqq from rom the th scram with the average response of s'ates it s>~

given i'n the above 'specification, provide the required the drives as protection, nnd HCPR remains greater than the safety limit.

On ati t arly BWR, somit I

deqr<<dation of control rod scram performs>>cc. occurrtnl i)ut.'inq p laant startup and was determitied ause y p)trticulate material (probably construction i'ris) 'pluqqinq an xntct'nal control rod drive p ent control rod drive (Model 78081ttttB) is filt h qrossly improved by tli. relocation of t h e filter to a loca tiotion out of tl<<scram drive ath. i.

i tef 'th cram performance, f even if completely bldcked 0 ,~ *a v ~

LIHITINC CONDITIONS, FOR OPERATION SURVEILLANCE REQUIREMENTS 3~5 CORE AND COHTAI'i%KNT 5 QO~R~hH S~S+S QQ~~Q~~SK COOL INC SYSTEMS and corresponding action shall continue until reactor operation is uithin the prescribed limits.

K. Minimum Critical Power Racio (MCPR)

Tha MCPR operating limit is K. nimum t ca 1.24 for'x8 fuel, and 1.25 Power for 8x8R fuel, and 1.25 for P8x8R fuel. These limits N"PR shall be determined apply to steady scate power daily during reactor power operation at raced pouer and I operation at 251 rated thermal power and flow. For operation with only one recirculation loop in opera- following any change in tion these values should be in- power level or creased by 0.01. For core flows distribution that uould cause operation with a other than rated, the MCPR shall limiting control rod be greater than the above limits pattern as described in times Kf, Kfis the value shown the bases for in Figure 3.5.2. If at any time Specification 3.3.

during operation, it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shell be initia-ted within 15 minutes to restore operation to within the prescrib-ed limits. If the steady state MCPR is not returned to within

.the prescribed limits within the .

(2) hours, the reactor Shall be brbught 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 />. Sur>>

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

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

167

tescinq to ensure that the lines 'are filled. The visual checkanc wi11 avoid startinq the core spray or RHR system with a discharqe line not filled. In addition to the visual observation and to ensure a filled discharqe line other than prior to testinq, a pressure suppression chamber head tank is located approximately 20 feet above the discharqe line hiqhpoint to supply makeup water for these systems. The condensate head tank located approximately 100 feet above the dfi,scnarqe high point serves as a backup charqinq system when the pressure suppression chamber head. tank is not in service.

System discharee pressure indicators are used to determine the water level above the discharqe line hiqh point.. The indicators will reflect approximately 30 psiq for a water level at the nich point and 45 psiq for a water level, in the pressure suppression chamber head tank and are monitored daily to ensure that the discharge lines are filled.

when in their normal standby condition, the suction for the HPCI and RCIC pumps are aliqned to the condensate storage tank, which is physically at, a hiqher elevation than the HPCIS and RCICS PiPinq. This assures that the HPCI and RCIC discharqe pipinq remains filled. Further assurance is provided by observing water flow from these systems high points monthly.

This specification assures that, the peak cladding temperature follcMing the postulated design basis loss-of-coolant accident will not exceed the limit, specified in the 10 CFR 50 Appendix K The peak cladding temper'ature following a postulated loss-of coolant accident is primarily a function of the average heat qeneration rate of all tne rods of a fuel assembly at any axial loc'ation and is only dependent secondarily on the rod co rod gower distribution ~ithin an assembly. since expected locaL variations in power dis'tribution within a fuel assembly affecc. the calculated peak clad temperature by less than i 20~F relative to the peak .temperature for a typical fuel desiqn, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10 CFR 50 Appendix X limit. The limitinq value for tApLHGR iq shown in Tables 3.5.I-I, 3. Tl.e analyses support nS these limitinS values is presenteh in reference .4 and 5.

near Kea Generation ate LHGR

'this specification assures that the linear heat generation rate in any 'rod is lese than the design linear heat, 176

~~Bll E8 loacred and. reported quarterly. 'It must be recoqnixed that there is alMays an action vhich ~ould return any ot" the parameters (HApLHGR, LHQR, or BcpR) to Mithin prescribed 1imats, namely pmer reduction. Under most circumtances, thos ~ill not be the only alternative.

ll, ~Rfe hC85 Cencric Reload Fuel Application, Licensing Topical Rcport NKDE 24011-P-A and Addenda,

5. uBrovas Ferry Nuclear Plant, Units 1, 2, and 3, Single-Loop Operation", NED0>>24236, May 1981.

178

TABLE 3.5. I-1 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: Initial Core - Type 2 Plant: BF-3 Average Planar Exposure MAPLHGR(1)

(Mwa/t) (kW/ft) 200 11. 4 17000. 11.6 5,000 12.0 10,000 12.2 15,000 12.3

. 20,000 12.1 25,000 11.3 30,000 10.2 TABLE 3.5. I-2 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Type: Initial Core - Type 1 Plant: BF-3 Average Planar Exposure MAPLHGR (Mwd/ t) (kW/ft) 200 11.2 1,000 11.3 5,000 11.8 10,000 12.1 15,000 12.3 20,000 12.1 25,000 11.3 30,000 10.2 (1) For operation with only '.one recirculation loop in service multiply given MAPLHGR values by 0.70.

) V 1%44

TABLE 3.5. X-3 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Fuel Types: 8DRB265L and P8DRB265L Plant: BF-3 Average Planar Exposure MAPLHGR (Mwd/t) (kw/ft) 200 11.6 11000 11.6 5,000 12.1 101000 12.1 15,000 12.1 20,000 11.9 25,000 11.3 30>000 10.7 (2) For operation with only one recirculation loop in service multiply given MAPLHGR values by. 0.82.

The values in this table are conservative for both prepressurized and nonpressurized fuel.

182

I LINITING CONDITIONS FOR OPERATION SURVEII LANCE REQUIREMENTS 6 PR MARY SYSTEM BOUNDARY e 6 R MARY SYSTEM BOUNDARY Recirculation. um 0 eration p Recirculation Pum 0 aration

1. The reactor shall not be 1. Recirculation pump operated Mith only one re- speeds shall be circulation loop in servic checked and logged at for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, un least once per day.

less designated adjustments for single-loop operation are made in,APRM rod block and scram setpoints (Tech-nical Specification 2.1.A, 2.1.B, and Table 3.2.C),

RBM setpoints (Table 3.2.C)

MCPR fuel cladding integri safety and operating limits (Technical Specification

1. 1.A) and MAPLHGR limits (Technical. Specification 3.5. I). If this specifica-tion cannot be met Mithin the stated 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then the plant shall be placed in hot shutdown Mithin an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless.

the adjustments are made sooner.

2t Following one-pump opera-tion, the discharge valve of the'oM speed pump may not be opened unless the speed of the faster pump is less than 50K of its rated speed.

195

LIMITING CONDITIONS POR OPERATION SURVEILLANCE REQUIREMENTS

$ 6 ~P RY S M D Y 4~6 PRIMARY SYSTEM B UNDARY

3. Stead/ state cPC! itic>> wf th lluth G Structural Inte rit rccirculati ">> p<<:!ps l<<t of scr- Table 4.6.A together vicc for up to )2 hr; is per- 1 ~

with supplementary mi t tcd. Ouri>>9 s>>~:h i>>tcrva restart of thc recirculation 1

notes, specifies the is permitted, provided th inservice inspection surveillance

'ump(s)

. loop dischal9c tern')cia'turc ls requirements of the l(ithi'n 75oF of tnc saturation reactor coolant terrperatui o Of the r aCtOr system as follows:

~

'vessel water as d tcrnincd by do!""c pres sul c ~ a. areas to be inspected

b. percent of areas to be inspected during the G>> Stru tural Inte rit inspection interval 1 ~ The structural integrity of the c. inspection primary, system shall frequency be maintained at the level required by the d. methods used for original acceptance inspection standards throughout the life of the 2~ Evaluation of plant. The reactor inservice inspections shall be maintained will be made to the in a cold shutdown acceptance standards condition until each specified for the indication of a original equipment.

defect has been investigated and 3 ~ The inspection evaluated interval shall be 10 years.

Additional inspections shall be performed on certain circumferential pipe welds as listed to provide additional protection against pipe whip, which could damage auxiliary and control systems.

Feedwater- GPW-9, KFW-13, GFW 12r GFW 26 GFW-29, 'FW-31r KPW-39 r GFW- 15 r 196 .KPW-38, and GPW-32

I 'I Jt~. '.,v '%W It+44 0

3.6/0.6 BARF,.

A norxle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true. The lack of any substantial stress in the jet pump diffuser body makes failure impossible without. an initial noxzle-riser system failure.

3.6.F/4.6.F Recirculation Pum 0 ezation Steady-state operation without forced recirculation will not be permitted for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. And the start of a recirculation pump from the na'tural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75 F. This reduces the positive reactivity insertion to an low value. 'cceptably Single-loop operation is permitted for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without making the adjustments in core limits and rod lines described in specification 3.6.F. This is Justified by the fact, that in most cases, recirculation pump maintenance can be performed in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For periods of extended operation, the adJustments in specification 3.6.F. should be made. It is noted that these restrictions are conservative compared to two pump operation in all cases.

Requiring the discharge valve of the lour speed loop to remain closed until the speed of the faster pump is belo~ SO% of its rated speed provides assurance when going from one to two pump operation that excessive vibration of the )et pump risers will not occur.

3.6.C/4.6.C Structural Inte zit

. The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the AS51E Boiler and Pzessure Vessel Code.

The program reflects the built-in limitations of access. to the reactor coolant systems.

P 227.

0 ENCLOSURE 2 JUSTIFICATION AND DESCRIPTION OF CHANGES

UNIT 1 Page 8, TS 1 ~ A, U-1 Add 0.01 to the MCPR safety limit for single recirculation loop operation to compensate for increased uncertainty in core flow parameters and nucleonics parameters during single-loop operation.

Page 8, TS 2. 1.A. 1, -U-1 Add -0.66 AW term to existing APRH flux scram trip setpoint equation.

This modifies the APRM flux scram trip setpoint equation to compensate for backflow through the get pumps on the inactive recirculation loop during single-loop operation. The equation reduces to the current equation for two recirculation loop operation since hW = 0 for two recirculation loop operation.

Page 10, TS 2.1.B, U-1 Add=-0.66 AW term to existing APRM rod block trip setpoint equation.

This modifies the APRM rod block trip setpoint equation to compensate for backflow through the get pumps on the inactive recirculation loop during single-loop operation. The equation reduces to the current equation for normal two recirculation loop operation since AW = 0 for two recirculation loop operation.

Page 15, Bases l. 1 (both changes), U-1 Page 16, Bases 1.1, U-1 Page 21, Bases 2.1.A.1, U-1 Page 22, Bases 2.1.A.3 5 2.1.B, U-1 Page 23, Bases 2. 1. 6, U-1 Page 113, Bases 3.2 (both changes), U-1 Page 131, Bases 3.3/4.3 (both changes), U-1 Remove reference to HCPR safety limit >1.07 since the MCPR safety limit depends on whether the unit is operating with one or two recirculation loops in service. For two-loop operation, the MCPR safety limit is still >1.07, but for single recirculation loop operation, the MCPR safety limit >1.08. So references to the MCPR safety limit being, greater than a specific limit are changed to a generic statement that MCPR be greater than the MCPR safety limit.

Page 17, References, U-1 Add references to support single recirculation loop operation.

Page 73, Table 3 2 'y U 1 Add -0.66 AW term to APRM upscale (flow bias) and RUM upscale (flow bias) equations. This modifies the equations to compensate for backflow through the get pumps on the inactive recirculation loop during single-loop operation. The equations reduce to the current equations for normal two recirculation loop operation since AW = 0 for two recirculation loop operation.

Page 74, Note for Table 3.2.C, note 2, V-1 Insert definition of AW term for APRM upscale (flow bias) and RBM upscale (flow bias) equations.

Page 160, TS 3.5.K, U-1 Insert note above increasing HCPR by 0.01 for single-loop operation.

(See explanation/justification for change on page 8, TS 1.A.)

Page 160A, TS 3.5.L.1, U-1 Add -0.66 AW (FRP/CHFLPD) term to existing APRH flux scram trip set point equation for operation in single recirculation loop operation mode with the core maximum fraction of limiting power density (CMFLPD) greater than the fraction of rated thermal power (FRP). This modifies the equation to compensate for backflow through the jet pumps on the inactive recirculation loop during single-loop operation. The equation reduces to the current equation for two recirculation loop oper ation since AW = 0 for two recirculation loop operation.

Add -0.66 AW (FRP/CMFLPD) to the existing APRM rod block trip setpoint equation for oper ation in single recirculation loop operation mode with the core maximum fraction of limiting power density greater than the fraction of rated thermal power (FRP). This modifies the equation to compensate for backflow through the jet pumps on the inactive recirculation loop during single-loop operation'. The equation reduces to the current equation for normal two recirculation loop operation since bW = 0 for two recirculation loop operation.

Page 168, Bases 3.5.I, U-1 Correct text to refer to proper tables and references.

Page 169A, References, U-1 Add reference to support single recirculation loop operation.

Pages 171, 172, 5 172A, Tables 3;5.I-l through 3.5.I-5, Footnotes, U-1 Add footnote to show required MAPLHGR reduction factors for single recirculation loop operation.

Page 182, TS 3.6.F.1, U-1 Insert requirements for remaining in single recirculation loop operation for more than 48 hours.

I y ~ ~ ~ ~

1

Page 182, TS 3.6.F.3, U-1 Change the word pumps to pump(s).

Page 221, Bases 3.6.F/4.6.F, U-1 Include material in bases to Justify single-loop operation.

1

~ ~

~~

Page 8, TS 1.A, U-2 Add 0.01 to the MCPR safety limit for single reoirculation loop operation to oompensate ior increased uncertainty in core flow parameters and nuoleonics parameters during single-loop operation.

'Page 8, TS 2.1.A.1, U-2 Add -0.66 dM term to existing APRM flux scram trip setpoint equation.

~

This modifies the APRH flux scram trip setpoint equation to compensate for backflow through the Jet pumps on the inactive recirculation loop during single-loop operation. The equation reduces to the current equation for two reoirculation loop operation since AM = 0 for two reciroulation loop operation.

Page 9, TS 2.1.A.1, U-2 Add <<0.66 AM (FRP/CMFLPD) term to existing APRM flux scram trip set point equation for operation in single recirculation loop operation mode with the core maximum fraction of limiting power density (CHFLPD).

greater than the fraction of rated thermal power (FRP). This modifies the equation to compensate for backflow through the Jet pumps on the inaotive recirculation loop during single-loop operation.'he equation reduces to the current equation for two reoirculation loop operation since Sl = 0 for two recirculation loop operation..

Page 10, TS 2.1.B, U-2 Add -0.66 4M term to existing APRM rod blook trip setpoint equation.

This modifies the APRM rod block tr ip setpoint equation to compensate for backflow through the get pumps on the inactive recirculation loop during single-loop operation. The equation reduces to the current equation for normal two reoiroulation loop operation since AW = 0 for two reoirculation loop operation.

Page 10, TS 2.1.B, U-2 Add >>0.66 N (FRP/CMFLPD) to the existing APRM rod block trip setpoint equation for operation in single recirculation loop operation mode with the core maximum fraction of limiting power density greater than the fraction of rated thermal power (FRp). This modifies the equation to compensate for baokflow through the Jet pumps on the inactive recirculation loop during single-loop operation. The equation reduces to the current equation for normal two recir culation loop operation since AM = 0 for two reoiroulation loop operation.

4

Page 15, Bases 1.1. (all changes), U-2 Page 16, Bases 1.1 (both changes), U-2 Page 21, Bases 2.1.A.1, U-2 Page 22, Bases 2.1.A.3 8 2.1.B, U-2 Page 23, Bases 2.1.6, U>>2 Page 113, Bases 3.2 (both ohanges), .U-2 Page 131, Bases 3.3/4.3 (both changes), U-2 Remove reference to MCPR safety limit >1.07 since the MCPR safety limit depends on whether the unit is operating with one or two reciroulation loops in service. For two-loop operation, the MCPR safety limit is still >1.07, but for single reoirculation loop operation, the MCPR safety limit >1.08. So referenoes to the MCPR safety limit being greater than a specific limit are changed to a generio statement that MCPR be greater than the MCPR safety limit.

Page 17, References, U-2 Add references to support single reoirculation loop operation.

Page 73, Table 3.2.C, U-2 Add -0.66 AW term to APRM upscale (flow bias) and RBM upscale (flow bias) equations. This modifies the equations to compensate for backflow through the jet pumps on the inactive recirculation loop during single-loop operation. The equations reduce to the current equations for normal two recirculation loop operation since AW = 0 for two recirculation loop operation.

Page 74, Note for Table 3.2.C, note 2, U>>2 Inser t definition of AW term for APRM upscale (flow bias) and RBM upscale (flow bias) equations.

Page 160, TS 3.5.K, U-2 Insert note above increasing MCPR by 0.01 for single-loop operation.

(See explanation/Justification for change on page 8, TS 1.A.)

Page'68, Bases 3.5.I, U-2 Correot text to refer to proper tables and references.

Page 169, Referenoes, U-2

~

Add reference to support single reoirculation loop operation.

Pages 171, 172, h 172A, Tables 3.5.?-1 through 3.5.I-5, Footnotes, U-2 Add footnote to show required MAPLHGR reduction factors for single recirculation loop operation.

~ ~

Page 182, TS 3.6.F.1, U-2 Xnser t requirements for remaining in single recirculation loop operation for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Page 182, TS 3.6.F 4 4.6.F, Speoification Heading, U-2 Change specifioation heading from Jet Pump Flow Mismatch to Recirculation Pump Operation. This change provides a more correct heading for the speoification.

Page 182, TS 3.6.F.3, V-2.

Change the word pumps to pump(s).

Page 221, Bases 3.6.F/4.6.F, U-2 Xnclude material in bases to )ustify single-loop operation.

UNIT 3 Page 9, TS 1.1.A, U-3

~ Add 0.01 to the MCPR safety limit for single recirculation loop operation to compensate fov increased uncertainty in core flow parameters and nucleonios parameters during single-loop operation.

Page 9, TS 2.1.A, U>>3 Add -0.66 AW term to existing APRM flux scram trip setpoint equation.

This modifies the APRM flux scram trip setpoint equation to oompensate

'or .backflow through the )et pumps on the inactive recirculation loop during single-loop operation. The equation reduces to the current equation for two recirculation loop operation since AW = 0 'ior two recirculation loop operation.

Page 10, TS 2.1.A, U-3 Add -0.66 AW (FRP/CMFLPD) term to existing APRM flux scram trip set point equation for operation in single reoirculation loop operation mode with the cove maximum fraction of limiting power density {CMFLPD) greater than the fraction of rated thermal power {FRP). This modifies the equation to compensate for backflow through the get pumps on the inactive recirculation loop during single-loop operation. The equation reduces to the ourrent equation for two recirculation loop operation since AW "- 0 for two recirculation loop operation.

Page 12, TS 2. 1.B (both changes), U-3 Add -0.66 AW term to existing APRM rod block trip setpoint equation.

This modifies the APRM rod block trip setpoint equation to compensate ior backflow through the )et pumps on the inactive recirculation loop during single-loop operation. The equation reduces to the ourvent equation for normal two recirculation loop oper ation, since AW = 0 fov two reoirculation loop operation.

'dd -0.66 bW (FRP/CMFLPD) to the existing APRM rod blook trip setpoint equation for operation in single recirculation loop operation mode with the core maximum fraotion of limiting power density greater 'than the fraction of rated thermal power (FRP). This modifies the equation to compensate for backflow through the get pumps on the inactive veoirculation loop dur ing single-loop operation. The equation reduces to the curd ent equation for normal two recirculation loop operation since AW = 0 for two recirculation loop operation.

Page 15, Bases 1.1. (both changes), U-3 Page 16, Bases 1. 1. (both changes), U-3 Page 20, Bases 2.1.A.1, U<<3 Page 21, Bases 2.1.A.3, U-31 Page 22, Bases 2.1.B 8 2.1.C, U>>3 Page 110, Bases 3.2, U-3 Page 111, Bases 3.2, U-3 Page 134, Bases 3.3/4.3.B.5 8 3.3/4.3.C, U-3 Remove reference to MCPR safety limit >1.07 since the MCPR safety limit depends on whether the unit is operating with one or two recirculation loops in service. For two-loop operation, the MCPR safety limit is still >1.07, but for si,ngle recitculation loop operation, the MCPR safety limit >1.08. So references to the MCPR safety limit being greater than a specific limit are changed to a generio statement that HCPR be greater than the MCPR safety limit.

Page 17, References, U-3 Add references to support single recirculation loop operation.

Page 76, Table 3.2.C, U-3 Add -0.66 AW term to APRM upscale (flow bias) and RBM upscale (flow bias) equations. Th's modifies the equations to compensate for backflow through the )et pumps on the inactive recirculation loop during single-loop operation. The equations reduce to the current equations for normal two recir culation loop operation since AW = 0 for two recirculation loop operation.

Page 77, Note for Table 3.2.C, note 2, U-3 Insert definition of AW term for APRM upscale (flow bias) and RBM upscale (flow bias) equations.

Page 167, TS 3.5.K, U-3 Insert note above increasing MCPR by 0.01 for single-loop operation.

(See explanation/Sustiflcat5.on for change on page 9, TS 1. 1.A.) .

Page 176, Bases 3.5.I, U-3 Cor reot text to refer to proper tables and references.

Page 178, References, U-3 Add reference to support single. recirculat5on loop operation.

Pages 181 4 182, Tables 3.5.I-l through 3.5.I-3, Footnotes, U-3 Add footnote to show required HAPLHGR r eduction factors for'ingle reoirculation loop operation.

Page 195, TS 3.6.F.1, U-3 Insert requirements for remaining in single recirculation loop operation for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Page 195, TS 3.6.F 5 4.6.F, Specification Heading, U-3 Change specification heading from Jet Pump Flow Mismatch to Recirculation Pump Operation. This change provides a more cor rect heading for the speoification.

Page 196, TS 3.6.F.3, U-3 Change the word pumps to pump(s).

Page 227, Bases 3.6.F/4.6.F, U-3 Include material in bases to /ustify single-loop operation.

ENCLOSURE 3 DESCRIPTION OF NEED l

Information provided in NED0-24236, "Browns Ferry Nuclear Plant Units 1, 2, and 3 Single-Loop Operation," demonstrates that it is unnecessary to completely shut down the unit in the event of a failure in one recirculation loop which cannot be repaired in,24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Analysis indicates that with proper adjustment for single recirculation loop operation the unit oan be operated at reduced power levels with no impact on public health and safety. In cases where a long outage (possibly inoluding removing all oi the fuel from the reactor) is needed to repair the inoperative recirculation loop, the ability to operate with a single recirculation loop in service means that the unit can continue to operate at reduced power until a scheduled refueling outage when the inoperative

-recirculation loop can be repaired.