Text

Amend 44 to License NPF-43,revising Tech Spec Figure 3.2.3-2, Flow Correction (Kf) Factor
	ML19332E084
Person / Time
Site:	Fermi
Issue date:	11/21/1989
From:	Zwolinski J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML19332E085	List: ML19332E084; ML19332E089;
References
	NUDOCS 8912060215
	Download: ML19332E084 (17)
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DETROIT EDISON C0etPANY WOLVERINE POWER SUPPLY COOPERATIVE, INCORPORATED DOCKET NO. 50-341

' FERMI-2 AMENDMENT TO FACILITY OPERATING LICENSE

s.

Amendment No. 44 License No. NPF-43 1.-

The Nuclear Regulatory Comission (the Commission) has1found that:

'A.

The application for amendment by the Detroit Edison Company (the licensee)datedNovember116,1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

L and the Comission's rules and regulations set forth in 10 CFR l

Chapter I; B.

The-facility will operate in conformity with the application, the provisions of the Act, and the. rules and regulations of the Comission; LC.

Thereis.reasonableassurance(1)-thattheactivitiesauthorizedby this amendment can be conducted without endangering the health and 1

safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be ininical to the common defense and security or to the health and safety of the public; and L

L E.

The issuance of this amendment is in accordance-with'10 CFR Part 51 of L

the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly,-the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph

'2.C.(2) of Facility Operating License No. NPF-43 is hereby amended to read as follows:

1 Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 44, and tne Environmental Protection Plan contained in L

Appendix B, are hereby incorporated in the license. Deco shall operate l-the facility in accordance with the Technical Specifications and the L

Environmental Protection Plan.

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

e FOR THE NUCLEAR REGULATORY' COMMISSION l

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4 John wolinski, Assistant Director -

for Region.III Division of Reactor Projects - III, IV, V & Special Projects Office of Nuclear Reactor Regulation-

'ttachment:-

A Cher.ges to the Technical-Specifications Date.of Issuance: November 21, 1989 o

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d ATTACHMENT TO LICENSE _ AMENDMENT NO. 44 FACILITY OPERATING LICENSE NO.;NPF-43 DOCKET NO. 50-341

) Replace tne following pages of the Appendix "A" Technical Specifications with

-f the attached pages. -The revised pages are identified by Amendment number and contain a vertical line' indicating the area of change. The corresponding overleaf pages are also provided'to maintain document completeness.

REMOVE INSERT 4

n B 2-1 B 2-1 3/4 2-4 3/4 2-4 3/4 2-4a 3/4 2-4a 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8' 3/4 2-8a.

3/4 2-8a 3/4 2-8b 3/4 2-8b 3/4 2-9 3/4 2-9 3/4 2-10 3/4 2-10 u

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0verleaf page provided to maintain document completeness.

No changes contained on this page.

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'2.1 SAFETY LIMITS-BASES t:

1

. 2:0 INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping are.the principal barriers to the release of radioactive materials to the l

environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated: transients. The fuel cladding integrity Safety Limit is set such-that no fuel damage is calculated

.to occur if the limit is not violated. Because fuel: damage is not directly L

-observable, a step-back approach is used to establish a Safety Limit. MCPR l

greater than-the Safety Limit represents a conservative margin relative to 1

the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is j

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related to its relative freedom from perforations or cracking. Although scme

'1 corrosion or use related cracking may occur during the life of the cladding, h

-fission product migration from this source is incrementally cumulative and l

continuously measurable.

Fuel cladding perforations, however, can result L

from thermal stresses which occur from reactor operation significantly above I

design conditions and the Limiting Safety System Settings. While fission l

product migration from cladding perforation is just as measurable as that p

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, the fuel cladding q

l.

Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a signi-ficant departure from the condition intended by design for planned operation.

2.1.1 : THERMAL POWER, Low Pressure or Low Flow The use of the approval critical power correlation is not valid for all l

critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore,.the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting i

l' condition on core THERMAL POWER with the following basis. Since the pressure

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drop in the bypass region is essentially.all elevation head, the core pressure L

drop-at. low power and flows will always be greater than 4.5 psi. Analyses 3

i show that with a bundle flow of 28 x 10 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. 3Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 10 lbs/hr.

Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel. assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER cf more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

FERMI - UNIT 2 B 2-1 Amendment No. 44

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MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PL'ANAR EXPOSURE RELOAD FUEL TYPE BC318D N

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AVERAGE PLANAR EXPOSURE RELOAD FUEL TYPE BC318E FIGURE 3.2.1-4 l

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- POWER DISTRIBUTION LIMITS-

' SURVEILLANCE' REQUIREMENTS q

h 4.2.3.1 -MCPR, with:

a.

-t =,1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4.1'.3.2, or b.

.t as defined in Specification 3.2.3 used to determine the limit

,within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time' surveillance

._ test required by Specificatior. 4.1.3.2, shall be determined _to.be equal to or greater than the applicable MCPR limit determined from Figures 3.2.3-1 through 3.2.3-1B and 3.2.3-2:

a.

-At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b '.

~Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL R0D PATTERN for MCPR.

1 d.

The provisions of Specification 4.0.4 are not applicable.

4~.2.3.2 Prior to:the use of Curve A and whenever Surveillance Requirement i

4.2.3.1 is performed'while using Curve A of Figures 3.2.3-1 through 3.2.3-1B, verify that all non-CCC control rods are fully withdrawn from the core..

Non-CCC control rods are all control rods excluding A2 rods, Al shallow rods inserted'less than or equal to notch position 36, all peripheral rods, and rods inserted to position 46.

Normal control rod operability checks, coupling checks, scram time testing, and friction testing of non-CCC control rods does l

not require the utilization of the more restrictive non-CCC operational mode L

MCPR limits.

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FERMI - UNIT 2 3/42-7 Amendment No. 97,44 l

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CURVE A - WCPR limit for CCC operational mode with both turbine bypast and moleture deperator reheater in service.

CURVE B

WCPR limit f ar non CCC operational mode with both f urbine bypass and moleture separator reheater in service.

CURVE C - MCPR limit for both CCC or non-CCC operational

- modes with either turbine bypees of moisture L

teparator reheeler out of service.

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CURVE D - MCPR limit for both CCC and non CCC operational l

modes with both turbine bypese and moleture separator l

rehester out of service.

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BOC TO 12,700 MWD /ST l

MINIMUM CR!TICAL POWER RATIO (MCPR) VERSUS TAU AT RATED FLOW FIGURE 3.2.3-1 FD MI - UNIT 2 3/4 2-8 Amendment No. 37,4p/,

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CURVE B - MCPR limit for non-CCC operational mode with both l

turbine-bypese and moleture separator reheater in service.

CURVE C MCPR limit for both CCC or non CCC operational medes with either turbine bypass or moleture separator reheat 2r out of service.

CURVE D - MCPR limit for both CCC or non-CCC operational modes with both turbine bypese ano moleture separator rehester out of service.

12,700 MWD /ST TO 13,700 MWD /ST MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS TAU AT R ATED FLOW FIGURE 3.2.3-1A FERMI - UNIT 2 3/4 2-8a AmendmentNo./1/,44 L

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MCPR limit for CCC operational mode with both turbine bypese and moleture separator reheater in service.

CURVE B - MCPR limit for non CCC operational mode with both turbine bypass and moleture separator reheater in service.

CURVE C - MCPR limit for both CCC and non-CCC operational modes with either turbine bypese of molature separator reheater out of servloe.

CURVE D + MCPR limit for both CCC of non CCC operational modea -with both turbine bypese and moleture separator r.h...t oui ei.or io..

13.700 MWD /ST TO EOC MINIMUM CRITICAL POWER RATIO VERSUS TAU AT RATED FLOW FIGURE 3.2.0-1B FERMI - UNIT 2 3/4 2-8b Amendment No. 47/,44

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FLOW CORRECTION (K,) FACTOR FIGURE 3.2.3-2 FERMI - Unit 2 3/4 2-9 Amendment No. ff,44 I

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3/4.2.4 LINEAR MEAT GENERATION RATE g

LIMITING mm! TION FOR OPERATION 3.2.4.The LINEAR HEAT GENERATION RATE (LNGR) shall not exceed 13.4 kw/ft for bundle types 8CR183 and 8CR233 or 14.4 aw/ft for bundle types BC3186 and SC318E.

APPL CABILI'TY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equa' to zam of RATED THERMAL. POWER.

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With the LNGR of.a4 fuel rod exceeding the limit, initiate corrective action L

within 15 minutes and restora the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or L

reduce THERMAL POWER to less than 25% of RATED THERMAL POWER trithin the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE ~ REQUIREMENTS 4.2.4 LHGR's shall be determined to be equal to or less than the limit:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , e

b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 155-of RATED THERMAL POWER, and b'

c.

Initially and at least once p r 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating on a LIMITING CONTROL ROD PATTERN FOR LHGR.

d.

The provisions of. Specification 4.0.4 are not applicable.

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FERMI - UNIT 2 3/4 2-10 Amendment No g2,44 l

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'3/4.1 REACTIVITY CONTROL SYSTEMS BASES s

3/4.1.1 SHUTDOWN MARGIN A sufficieht SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits,.and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be subcritical by at-least R + 0.38% delta k/k or R + 0.28% delta k/k, as appropriate.

L The value of R in units of % delta k/k is the difference between the calculated

_value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of R must be positive or zero and must be determined for each fuel loading cycle.

Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demor.stration of the SHUTDOWN MARGIN.

j The highest worth rod may be de'termined analytically'or by test.

The SHUTDOWN l

MARGIN is demonstrated by an insequence control rod withdrawal at the beginning

of life fuel cycle conditions,. and, if necessary, at any future time in the cycle if the first demonstration indicates that the required margin could be l

reduced as a function of exposure.

Observation of subcriticality in this condition assures subcriticality with the most reactive control-rod fully withdrawn.

l'

. This reactivity characteristic has been a basic assumption in the analysis of plant ~ performance and can be best demonstrated at the time of fuel loading, L

but the margin must also be determined anytime a control rod is incapable of insertion, j

'3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is small, a careful check on actual conditions to the predicted conditions is necessary, and the changes in reactivity can be inferred from these comparisons of rod patterns.

Since the comparisons are. easily done, frequent checks are not an imposition on normal operations.

A 1% change is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluated.

A change as large as 1% would not exceed the design conditions of the reactor and is on the safe side of the postulated transients.

FERMI - UNIT 2 B 3/4 1-1

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REACTIVITY 2CONTROLSYLT, EMS E

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3/4.1.3' CONTROL RODS.

The specjfications of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with i

i cthose used in thel safety analyses, and (3) limit the potential effects of the t

rod drop acciient. The ACTION statements permit variations from the basic requirements but~at the same-time impose more restrictive criteria for continued operation. 'A limitation on~ inoperable rods is set such that the resultant effect

-on total rod worth and scram shape will be kept to a minimum. The requirements for the-various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

. _ Damage-within the' control rod drive mechanism could be a generic problem, 1

~therefore with a control rod immovable because of excessive friction or mechan-l~

ical interference, operation of.the reactor is limited to a time period which

is reasonable to determine the cause of the inoperability and at the same-time prevent operation with a.large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be l

taken out.of service provided that those in the nonfully inserted position are consistent with~the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable.could be more than

.(

~the.eight allowed by the specification, but the occurrence of eight inoperable L

rods could be1 indicative of a generic problem and the reactor must be shut down p

for investigation and resolution of the problem, The control rod system is designed to bring the reactor subtritical at a rate fast'enough to prevent the MCPR from becoming less than the Safety Limit 1

MCPR during the~1imiting power transient analyred in Chapter 15 of the UFSAR.

This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given 'in the specifications,. provide the required protection and MCPR remains greater than the Safety Limit MCPR. The occurrence of scram. times longer than those

specified should be viewed as an indication of a systematic problem with the i

rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially L

-serious problem.

The-scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a B

reactor screm and will isolate the reactor coolant system from the containment when required.

I Control rods with inoperable accumulators are declared inoperable and l.

.Specificatior 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still~be inserted with normal drive water pressure. Operability of the accu-mulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.

FERMI - UNIT.2 B 3/4 1-2 Amendment No. 44

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BASES TABLE B 3.2.1 _1 SIGNIFICA_NT_ INPUT PARAMETERS TO THE LOSS-OF-COOLANTACQDENTANALYSIS Plant Parameters:.

Core-THERMAL P0WER...............

3430 HWt* which corresponds to 105% of rated steam flow i

6

' Vessel - Steam Out put............... ' 14.86' x 10 lbm/hr which corresponds to 105% of rated steam flow l

Vessel Steam Dome Pressure.......

1055 psia Design Basis Recirculation Line Break Area for:

m 2

a '.

Large Breaks ' 4.1 f t 2

b..

Small Breaks 0.1 ft Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL

' FUEL BUNDLE _

GENERATION RATE PEAKING POWEF.

FUEL TYPE-GE0 METRY

( kW/f t)

FACTOR RATIO i

Initial Core 8-x 8 13.4 1.4

'1.18 First Reload 8x8 14.4 1.4 1.18-l A.more detailed listing of input of each model and its source is presented in Section II of Reference I and subsection 6.3 of the FSAR.

This power level meets the Appendix K requirement of 102%. The core heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.

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FERMI - UNIT 2 B 3/4 2-3 Amendment No. 42,44

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POWERDISTRIBUTIONLIMi]s 3

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3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limiting MCPRs at steady-state operating conditions as_specified in Specification 3.2;3 are derived from the established fuel clad-ding integrity Safety Limit MCPR, and an analysis of abnormal operational tran--

sients. For any abnormal operating' transients analysis evaluation-with the h

initial _ condition of the reactor being at the steady. state operating limit, it

'is required that the resulting.MCPR does not-decrease below the Safety Limit MCPR at any time during:the transient assuming instrument trip setting given in

Specification 2.2.

To assure =that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-

.sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase.in pressure and power, positive reactivity insertion, and coolant tem-p, V

,perature decrease. The limiting transient yields-the largest delta MCPR. When og &

indded to the Safety Limit MCPR, the required minimum operating limiting MCPR of Specification 3.2.3 is obtained and presented in Figures 3.2.3-1, 3.2.3-1A, g"9, j!

and 3.2.3-18.

s The MCPR curves illustrated in Figures 3.2.3-1 thru 3.2.3-1B were derived

'.7

asidescribed above-for the following assumed operating conditions

-Curve A - MCPR limit with turcine bypass system, moisture separator reheater systems.in service and CCC (Control Cell Core) operational mode (A2 rods, Al shallows inserted ~1ess than or equal to notch posi-tion 36, all peripheral rods, and all rods inserted to position p

46). The operating domain includes the 100% power / flow region

~

and extended load line region with 100% power and reduced flow.

Curve B - MCPR limit with the turbine bypass system, moisture separator reheater systems in service and non-CCC o (any

-non-CCC control rod inserted in the core)perational mode The operating L

domain includhs the 100% power / flow region and the extended load line region with 100% power and reduced flow.

Curve C - MCPR limit for either CCC or non-CCC operational modes with either the main turbine bypass system inoperative and the mois.

l ture separator reheator system available or the main turbine i

bypass system available and the moisture separator reheater sys-tem inoperable. The operating domain includes the 100%

power / flow region and the extended load line region with 100%

power with reduced flow.

Curve D - MCPR limit for either CCC or non-CCC operational modes with the main turbine bypass system inoperative and the moisture separator reheater system inoperable. The operating domain includes the FERMI - UNIT 2 B 3/4 2-4 Amendment No. 39,g,44

I-q l b.QC VN

. c :.'

?.

POWER DISTRIBUTION LIMITS

BASES f

j 3/4.2.3 MNIMUM CRITI(,AL POWER RATIO (Continued) bypass' system' or the moisture separator reheater be inoperable as 25-percent

~ RATED. THERMAL' POWER-is exceeded, the MCPR check must be completed within one i

hour.

The' evaluation of a given. transient begins with the system initial

- parameters shown in UFSAR Table 15.0.1 that are input to a GE-core dynamic n

~ behavior ~ transient computer program. The codes used to evaluate transients are described in GESTAR:II.. The principal result of this evaluation is the 1

reduction in MCPR caused by.-the. transient.

.The purpose of'the K, factor of Figure 3.2.3-2 is to define operating.

.. limits at other than rated core flow conditions. At.less then 100% of rated flow the required MCPR is the product of the MCPR and the K factor. The f

K fa'ctors assure that the Safety Limit MCPR will not be violated during'a flow iberease.transientresultingfromamotor-generatorspeedcontrolfailure. The K factors'may be applied to both manual and automatic flow control modes.

f

The K factor values shown in Figure 3.2.3-2 were developed generically f

and are~ applicable to all BWR/2, BWR/3, and BWR/4 reactors.

The K, factors were' derived.using the flow control line corresponding to RATED THERMAL POWER at rated ' core flow, although they are applicable for the extended operating.

region.

/

factors were calculated-such that

For the manual flow control mode, the K f

> for the maximum flow rate, as limited by.the pump scoop tube setpoint and the

- corresponding-THERMAL POWER along the rated flow control line, the limiting bundle's relative power was adjusted until the MCPR changes with different core flows.3 The ratio of.the MCPR calculated at a given point of core flow, divided by the' operating limit MCPR, determines the K.

f p

1 l

t FERMI - UNIT 2 B 3/4 2-4b Amendment No. M 42.44

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ML19332E084

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