Text

Proposed Tech Specs,Supporting Reload 1,Cycle 2 for Plant Operations
	ML20248J493
Person / Time
Site:	Fermi
Issue date:	04/03/1989
From:	; DETROIT EDISON CO.
To:	;
Shared Package
ML19297H484	List: ML20248B419; ML20248J493; NRC-89-0052, Supplemental Reload Licensing Submittal for Fermi Plant Unit 2,Reload 1,Cycle 2;
References
	CON-NRC-89-0052, CON-NRC-89-52 NUDOCS 8904140448
	Download: ML20248J493 (41)
	v • d • e

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9 8 ATTACHMENT 2

SUMMARY

OF PROPOSED TECHNICAL SPECIFICATION / BASES CHANGES W

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l ATTACHMENT 3 i TECHNICAL SPECIFICATION AND BASES CHANGES i

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- - _ - . ____ --__ _ _-__ _ _- - - - _ . .. .w

DEFINITIONS CORE ALTERATION 4

1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed an.1 fuel in the vessel.

Normal movement of SRMs. IRMs, TIPS, or special movable detectors is not considered a CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.

C o N F" ,

CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of at power in the assembly which is calculated by application of tk j^T correlations.o cause some point. in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alonej would produce the same thyroid dose as'the quantity and isotopic mixtufe of I-131, I-132, I-133. I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

E-AVERAGE DISINTEGRATION ENERGY 1.10 E shall t!e the average, weighted in proportion to the concentration of each radionuclides in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in HeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until .the ECCS equipment is capable of performing its safety function, i.e. , the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

FRACTION OF LIMITING POWER DENSITY 1.12 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing at a given location divided by the specified LHGR limit for that bundle type.

FRACTION OF RATED THERMAL POWER l 1.13 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL POWER divided by the RATED THERMAL POWER.

i FERMI - UNIT 2 1-2

DEFINITIONS b .

2. Closed by at least one manual' valve, blank flange, or

- deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6.3.

I

b. ' All primary containment equipment hatches are closed and seale'd.

c. Each primary containment air lock is in comp 1,iance with the requirements of Specification 3.6.1.3.

d. The primary containment leakage rates are within the limits of Specification 3.6.1.2. "

e. The suppression chamber is in compliance with the requirement of Specification 3.6.2.1.

f.

The sealing mechanism associated with each primary containment penetration, e.g. , welds, bellows, or 0-rings, is OPERABLE.

g. The suppression chamber to reactor building vacuum breakers are in compliance with Specification 3.6.4.2.

THE PROCESS CONTROL PROGRAM 1.30 The PROCESS CONTROL PROGRAM (PCP) shall.contain the provisions to as the SOLIDIFICATION of wet radioactive wastes results in a waste fo properties that meet' the The requirements of 10 FCP shall identify CFRparameters process Part 61 and of low-leve active waste disposal sites. influencing SOLIDIFICATION, such as pH, oil c ratio of solidification agent to waste and/or necessary additives for each type s of anticipated waste, and the acceptable boundary conditions for the process 9 parameters shall be identified forThe each waste type, based on laboratory s PCP shall also include an identi-and full scale testing or experience.

fication of conditions that must be satisfied, based on full scale testing, to assure that dewatering of bead resins, powdered resins, and filter sludges will result in volumes of free water, at the time of disposal, within the limits of 10 CFR Part 61 and of low-level radioactive waste disposal sites.

PURGE - PURGING 1.31 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.32 RATED THERMAL POWER shall be a total reactor core heat tra thereactorcoolantof)29fNWT.

'5M O FERMI - UNIT 2 1-5

. l

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_2_ . 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER. Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow ,

less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. .

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10%

of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1. !

O hq. ( k U e4 g,n1 ' ;

THERMAL POWER, High Pressure and High Flow -Fre"9 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than .

with the reactor vessel steam dome pressure greater than 785 psig and core flow gaeater than 10% of rated flow.

APPLICABILITY _: OPERATIONAL CONDITIONS 1 and 2.

i ACTION: g k h Q g Q ,44 twt' Nil, d \.07 With MCPR less than,1<6f and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, l l

I REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4. )

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

2-1 FERMI - UNIT 2

%'l i

4 SAFETY LIMITS !

BASES 2.1.2 THERMAL POWER. Hioh Pressure and Hich Flow The fuel cladding integrity Safety Limit is set such that no mechanistic Since the fuel damage is calculated to occur if the Ifnit is not violated. ~

parameters which result in fuel damage are not directly observable during reac-tor operation, the thermal and hydraulic conditions resulting in a fuel damage could occur. Although it is recognized that a departure from !

nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has beenHo adopted as a convenient limit.

core operating state and in the procedures used to calculate the critical Therefore, thepower '

result in an uncertainty in the value of the critical power. fuel cladding assembly for whici) em4 than 99.9% of th'e fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties. ' ~ - - = = _~ p

' .., .n=rma V '

e ' Safety', Limit hej lysis Basis, GETAB"MCPR is determined DsEthe e uencrui c.iec tainties in operating parameters and the procedures used to calc on is I ont crit determin power. The probability of the occurrence of boiling trans ength (L))

using the General Electric Critical Quality (X) Boili ](

GEXL, corre on.

tion is valid over the range of co ons used in the

" The GEXL cor f t<:sts of the data use o develop the correlation. O the uncertainties listed The required input to e statistical model the core parameters listed i i Bases Table B2.1.2-1 and t ominal values j i n Bases Table B2.1.2-2.

i e core parameters are given in ,

The bases for the uncertainties n b ertain in the GEXL correlation is give 1 IE00-20340 and the basis for the n NE00-10958-A*. The power ribution is ba onatypical764assemblyd tr as arbitrarily chos to produce a skewed po J: ore in which the rod patte ' est number of assemblies t the highest power a

distribution having the ution during any fuel cycle 1d not be as severn (3evels. The worst dis ed in the analysis.

Ins the distribution

[i

rrelation, and

a. "Gener lectric BWR Thermal Analysis Bases (GETAB) Data, l

Desi Application," NE00-10958-A.

eneral Electric " Process Computer Performance Evaluation Accuracy

g. and b .; d os p rnn- 9 _A M dg.ted Laa 1973 c.d neembe h NE00-20340

..WespectWely. 4 y

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~TN5 ERT 8 2-2 FERMI - UNIT 2

j Insert 1 The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in the operating parameters and critical power. The in the procedures used to calculate probability of the occurrence of boiling transition is determined using the approved critical power correlation. Details of the fuel ,

cladding integrity safety limit calculation are given in Reference !

1.

the Reference 1 includes determination a tabulation of the Safety of the uncertainties Limit MCPR and of the nominal used in values of parameters used in the Safety Limit MCPR statistical analysis.

I l

Reference 1.

" General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P-A (latest approved revision).

i I

b* Bases Table 82.1.2-1 UNCERTAINTIES USED IN THE' DETERMINATION OF THE F0EL CLADDING SAFETY LINIT*

Standard '

Deviation Quantity (% of Point) 1.76 Feedwater Flow

a Feedwater Temperature 0.76 0 4 Reactor Pressure Core Inlet Temperature 0.2 (MTlk Core Total Flow- ( 2.5 ,

3.0 A Gt Channel Flow Area Friction Factor Hultiplier 10.0 Channel Friction Factor 5.0 Hultiplier

( 6.3 3

( TIP Readings R Factor 1.5 Critical Power 3.6

The uncertainty analysis used to establish the core wide Safety Limit MCPR is based on'the assumption of quadrant power symmetry for the reactor core.

FERMI - UNIT 2 B 2-3

l Bases Table 82.1.2-2 g NOMINAL VALUES OF PARAMETERS USED IN THE STATISTICAL ANALYSIS OF FUEL CLADDING INTEGRITY S LIMIT THERMAL POWER 3323 HW Core Flow 108.5 Mlb/hr Dome Pressure 1010.4 ps ,

2 Channel Flow Area 0.108 ft R-Factor H h enrichment - 1.043 edige enrichment - 1.039 Low enrichment - 1.030 ,

i YT\$

I h(e leLETED .

FERMI - UNIT 2 B 2-4

i .

REACTIVITY CONTROL SYSTEMS

)

3/4.1.4 CONTROL R00 PROGRAM CONTROLS R00 WORTH MINIMIZER f

! LIMITING CONDITION FOR OPERATION c

f 3.1.4.1 The rod worth minimizer (RWM) shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2*, when THERMAL POWER is less than or equal to 20% of RATED THERMAL POWER, the minimum allowable preset power 1svel.

ACTION:

) a. With the RWM inoperable, verify control rod movement and compliance with the prescribed control rod pattern by a second licensed operator or other technically qualified member of the unit technical staff who is present at the reactor control console. Otherwise, control

) rod movement may be only by actuating the manual scram or placing the reactor mode switch in the Shutdown position.

1 I b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS ,

0 4.1.4.1 The RWM shall be demonstrated OPERABLE:

h r -

a. In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to withdrawal of b '

control rods for the purpose of making the reactor critical, and in OPERATIONAL CONDITION 1 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months af ter RWM automatic initia-tion when reducing THERMAL POWER, by verifying proper indication of the selection error of at least one out-of-sequence control rod.

b. In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to withdrawal of control rods for the purpose of making the. reactor critical, by veri-fying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod (after selection of first control 0 rod).

g c. In OPERATIONAL CONDITION 1 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after RWM automatic initiation when reducing THERMAL POWER, by demonstrating the withdraw block and insert block functipns.

% W,4 9o 0 % WINM

d. By demonstrating that the ; ntM' 7;d p;tt;r e-end sequence input to the RWM computer e,ctly loaded following any loading of the program into the computer. is

" Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.

g FERMI - UNIT 2 3/4 1-16

r-3/4.2 POWER DISTRIBUTION LIMITS I

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1 l

LIMITING CONDITION FOR OPERATION I skaAA nok we,a.ad *. I 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) ' r ::ch type f r- r ,.

. .[k , 2

' LMi APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

1 ACTION: ?

skus. ^

"';r : 3.".' :, ;. .

a limitsj^'inutes , v. . , . . . , ,

With an APLHGR exceeding the and restore APLHGR to within initiate corrective action within 15 m the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

r SURVEILLANCE REQUIREMENTS f

4.2.1 u -- r a All APLHGRs

--es-,,.. shall,be,

,.,verified

, , ,.,to be

.a,equal o ato or less than the limits f4 b O Ie \1

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 R00 PATTERN for APLHGR.

d. The provisions of Specification 4.0.4 are not applicable.

FERMI - UNIT 2 3/4 2-1

..j

1 Insert 1 {

i A. The MAPLHGR limit which has been approved for the respective I I

fuel and lattice type as a function of the average planar exposure (as determined by. the NRC approved methodology described in GESTAR-II), or B. When hand calculations are required, the most limiting lattice type MAPLHGR limit as a function of the average planar exposure shown in the Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, and 3.2.1-4 for the applicable bundle type.

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POWER OfSTRIBUTION LIMITS b

3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit shown in Figureg3 .2.3-1 times the Kg shown in Figure 3.2.3-2, with:

, bu. 'E ."1."1 - % %

,, (Tave 38 ) ]

tg - 28

~

where:

. Tg = 1.096 seconds, control rod average scram insertion time limit to notch 36 per Specification 3.1.3.3, N

I Yg 0.852 1.65[ ) .06, n

N -

,96 I i 01 i=1 n

Z t,y, , y,3 Ny tg ,

1 Ng [

i=1 n = number of surve111ance tests performed to date in cycle, '

th N g = number of active control rods measured in the i surveillance test, ,

Ty = average scram time to , notch 36 of all rods measured in the i th surveillance test, and N total number of active rods measured in Specification

g = 4.1.3.2.a. ,

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION I

a. W MCP ess th he a icable Ilmit own in gures 3.2 -1 and

.2.3 , initi corr tve act within minut and restor CPR to glg wit the r utred mit wit 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reduc HERMAL P0 to less t n 25% RATED ERMAL P R withi he next hours.

9,

b. ith t main rbine b ss syste noperab per Spec ication 3. e no oper ion sa continue nd the p visions Speci 'ic fon 3.0.4 one ho ,MCPR 1 termined be al ap icable rovided at, with ain or gr ter than ' e MCP

~

mit as s wn in Fi e 3.2. ' I y th turbine ypass in erabl rve times the applic le Kg s n in gure 3.2 J FERMI - 11 NIT 2 3/4 2-6 i I

Insert 2

a. Operating in the Control Cell Core (CCC) operating mode and MCPR less than the applicable MCPR limit shown in Figures 3.2.3-1.thru 3.2.3-1B (Curve A) times the applicable K, curve shown in Figure 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

b. Operating in the non-CCC operating mode and MCPR less than the applicable MCPR limit shown in Figures 3.2.3-1 thru 3.2.3-1B (Curve B) times the applicable K, curve shown in Figure 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or t

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

c. Operating in either the CCC or non-CCC operating mode with either the main turbine bypass system inoperable per' Specification 3.7.9 or the moisture separator reheater inoperable, operation may continue and the provisions of specification 3.0.4 are not applicable provided that, within one hour, MCPR is determined to be equal to or greater than the MCPR limit as shown in Figure 3.2.3-1 thru 3.2. 3-1B (Curve C) by the main turbine bypass or moisture separator reheater inoperable curve times the applicable K, shown in Figure 3.2.3-2.

d. Operating in either the CCC or non-CCC operating mode with both the main turbine bypass system inoperable per Specification 3.7.9 and the moisture separator reheater inoperable, operation may continue and the provisions of ,

Specification 3.0.4 are not applicable provided that, within one hour, MCPR is determined to be equal to or greater _than the MCPR limit as shown in Figure 3.2.3-1 thru 3.2.3-1B by the main turbine bypass and moisture separator reheater inoperable curve times the applicable K, shown in Figure 3.2.3-2.

The CCC operating mode includes operation with only A2 rods, Al shallow rods less than or equal to notch position 36, all peripheral rods inserted in the core, 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 not require the utilization of the more restrictive non-CCC operational mode MCPR limits. Any other operation is a non-CCC operating mode.

_ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ .._____m __

e 4

POWER DISTRIBUTION LIMITS .

_SURVE!LLANCE REQUIREMENTS 4.2.3 MCPR, with:

a. t = 1.0 prior to performance of the initial scras 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 deterstne 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 Specification 4.1.3.2 I l

shall be determined to be equal to or greater than the applicable MCPR limit deterstned from Figures 3.2.3-1 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 THERNAL 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 bid a LIMITING CONTROL ROD PATTERN for MCPR.

d. The provisions of Specification 4.0.4 are not appifcable.

?

V -

4.2.3.2 Prior to the use of curve A and whenever Surveillance Requirement 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, i Al shallow rods inserted less thar v- equal to notch

! position 36, all peripheral rods, :tnd rods inserted to position 46. Normal control rod or erability checks, coupling checks, scram time testing, and friction testing of non-CCC control rods does not require the utilization

of the more restrictive non-CCC operational mode MCPR l limits.

l FERMI - UNIT 2 3/4 2-7

1.46 1.46 1.4 - -

1.4 CURVE D x1(1.39 J 1.36, ,_ _ x

-x - X ,, ,x _ -x - x . -x - X _1.36 _

1.36 ,4,,,,,,4...-+g j p C["~ -

1.36 1.3a_. + -+~~

M C p + u- CMRVE B .

p 1.3:

1.3 R

CURVE A 1.26 -

1.26 1.2 - -

1.2 1.16 1.16 0 0.10.20.30.40.60.60.70.80.91.0 TAU l CURVE A - MCPR limit for CCC operational mode with both turbine bypaes and moisture separator reheater in service.

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

CURVE C - MCPR limit for both CCC or non*CCC operational modes without either turbine bypass or moleture separator reheater.

CURVE D = MCPR limit *or both CCC and non CCC operational mCfsv without both turbine bypese and moisture separator l re h e ate r. l BOC TO 12,700 MWD /ST MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS TAU AT RATED FLOW FIGU RE 3.2.3-1 d /4 2-6

l I

1.45 1,46 q X ifgg 1.4 curve o l

-X -

X , ,x _ -X - c 1.36 ~X ~ x 1.36

(

-X' , x -X CURVE C --

1.36 - +~"'4.. -

1.36 1.33. _...-+-"+.".4.-4,,,,,4..+-+"~

(

1.34 M CURVEB C y,3: ' -' <> -' _

3,3 P

R 1.27 CURVE A 1.26 - -

1.25 1.2 - -

1.2 i

1.16 1.16 0 0.10.20.30.40.60.60.70.80.91.0 TAU CURVE A = 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. j CURVE C - MCPR limit for both CCC or non-CCC operational enodes without alther turbine bypass or moleture separator reheater.

CURVE D - MCPR limit for both CCC or non-CCC operational modes without both turbine bypese and moleture separator reheater.

12,700 MWD /ST TO 13,700 MWD /8T MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS TAU AT RATED FLOW FIGURE 3.2.3-1 A 3/4 2-8A

s s 1.45 1.45 l*4 ~

CURVE D $,' 4

-X - M~'" 1.39

_ _ .y -X - M ' ' X.

1.36 3. -X - X _ CURVE C . 1.36 1.35 &-+""'4..-..+...,... 1.35 -

6 1.34 1.33- - ~~ + """ 4 ...... & + ~"" 4 ... . ,

M C pu Rv,'E s 1.3' 13 R

1.28 CURVE A 1.25 - -

1.25 1.2 - -

1.2 1.15 1.15 0 0.10.20.30.40.50.60.70.80.91.0 TAU CURVE A = MCPR limit for CCC operational mode with both turbine bypass and moleture separator reheater in service.

CURVE s 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 without either turbine hypese or moisture separator reheater.

CURVE D - MCPR limit for both CCC or non-CCC operational modes without both turbine bypass and moleture separtor rehester.

13,700 MWD /ST TO EOC MINIMUM CRITICAL POWER RATIO VERSUS TAU AT RATED FLOW FIG U RE 3.2.3-1B L

c _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -

1.4- -

r c

1.3 -

AUTOMATIC FIDW CONTROL 1.2 -

K f ,

1.1 -

MANUAL FLOW CONTROL 3C00P TUBE SETPOINT CAUBRATION POSITIONED SUCH THAT

' FLOYMAX = 102.5%

1.0 -

107.07.

= 112.0%

= 117.07.

I I I I I I I O

30 40 50 60 70 80 90 100 l CORE FLOW (%)

l l FLOW CORRECTION (Kf ) FACTOR FIGURE 3.2.3-2 3/42-9 l

t POWER DISTRIBUTION LIMITS

( '

3/4.2.4 LINEAR HEAT GENERATION RATE ,

LIMITING CONDITION FOR OPERATION i

f 3.2.4 The LINEAR HEAT GENERAT t ew wak % scmt3 %4 *WukON or W W RATE w i( (LHGRp.khall

<ML sesseD w a sc.ss@ ,

not ekcoed APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or 13.4 k equal.to 25% of RATED THERMAL POWER.

c ACTION:

With the'LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore 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 reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within 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 , ,

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 on a LDiiTING CONTROL ROD PATTERN FOR LHGR.

d. The provisions of Specification 4.0.4 are not applicable.

i i

FERMI - UNIT 2 3/4 2-10

(

t

1

[ DESIGN FEATURES i

5.3 REACTOR CORE FUEL ASSEMBLIES p 5.3.1 Thereactorcoreshallcontain764fuelassemblieMitheachfuel l embly containing 62 fuel rods and two water rods clad with Zircaloy-2.

Each fuel rod shall have a nominal active fuel length of 150 inches. The '

q initial core loading shall have a maximum average enrichment of 1.89 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum assembly average enrichment of 3.20 weight (percent U-235. au 4 s kai awa=%na.s %k % w and

.oA mer., ac,<.vsA =&as awa as%4s and %va. su.w %e \y.4. w.

I

^ " ** " '

CONTROL ROD ASSEMBLIES "Ma\

5.3.2 The reactor core shall contain 185 control rod assemblies, each consisting of a cruciform array of stainless steel tubes containing 143 inches of boron carbide, B C, powder surrounded by a cruciform shaped stainless steel sheath.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE l

5.4.1 The reactor coolant system is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressare of:

1. 1250 psig on the suction side of the recirculation pump.

2. 1500 psig from the recirculation pump discharge to the outlet f.ide of the discharge shutoff valve.

3. 1500 psig from the discharge shutoff valve to the jet pumps:

c. For a temperature of 575*F.

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,034 cubic feet at a nominal steam dome saturation temperature of 540 F.

i FERMI - UNIT 2 5-5

} . . .

REACTIVITY CONTROL SYSTEMS ,_

~

8ASES CONTROL R005 (Continued)

The surveillance requirement to measure and record the time that the accumu-lators maintain pressure above the alarm setpoint is intended to provide infor- ,

mation rather than establish OPERABILITY of the accumulators. No action is !

required if tha accumulator pressure does not remain above the alarm setpoint for at least 10 minutes.

Control rod coupling integrity is required to ensure compliance with the J analysis.of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have affected the control rod coupling integrity. The subsequent check is performed as a backup to the initial demonstration.

In order to ensure that the control rod patterns can be followed and therefore that other parameters are within their limits, the control rod position indication system must be OPERABLE.

The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant sy!! tem. The support is not required when there is no ,

l pressure to act as a driving, force to rapidly eject.a drive housing.

The required l surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system j

components.

3/4.1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to ;

result in a peak fuel enthalpy greater t an 280 cal /gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous,

- scattered patterns of control rod withdrawal. When THERMAL POWER is greater than l- 20% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus requiring the RSCS and RWM to be OPERABLE when THERMAL POWER is less than or equal to 20% of RATED THERMAL POWER provides adequate control.

The RSCS and RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

The analysis of the rod drop accident is presented in Section 15 4.9 of theUFSAR and the techniques of the analysis are presented in a topical report, Referencer i Rd - --rrk R U. E';72;;; 2d?

The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs

[ up the written sequence used by the operator for withdrawal of control rods.

FERMI - UNIT 2 B 3/4 1-3 l

' ' ]I t i REACTIVITY CONTROL SYSTEMS g

)

' BASES f 3/4.1.5 STAND 8Y LIQUID CONTROL SYSTEM The standby liquid control systen provides backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity of boron which produces a concen-tration of 660 ppe in the reactor core and other piping systems connected to !

the reactor vessel. To allow for potential leakage and improper mixing this concentration is increased by 25L The required concentration is achieved by having a minimum available quantity of 4640 gallons of sodium pentaborate solution containing a minimum of 5500 lbs of sodium pentaborate. This quantity of solution is a net amount which is above the pump suction, thus allowing for the portion which cannot be. injected. The pumping rate of 41.2 gpm provides a negative reactivity insertioa rate over the permissible pentaborate solution volume range, which adequately compensates for the positive reactivity effects due to temperature and xenon during shutdown. The temperature requirement is necessary to ensure that the sodium pentaborate remains in solution.

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to ,

continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable. ,

Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron con- /)" . '

centration will not vary unless more boron or water is added, thus a check on j l the temperature and volume once each 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months assures that the solution is available for use.

Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.

C. J. A;,agf " Sti- r .,J J. i - S ""- ' "- Adt" ^-td s

- -1.

- 9 m N '7" W -i. ~,eL e Lpui6 nEv640~a, a e .,;E _

~

-2 g v.J.raone,k.6."US,-,d DW *g e M *"JG""~'^^ O ,

~. q ,, , , . . . . . .

3. , J , ". ";.,a. 0. J.-.";;c.; 47.." f. E ^^. ' " ^^4 = 4 W ' " *d " %

k LeeL ;0t ".1;5^9_.l?lU=

w.Z??

l, "fuktseA TAac\rck MM (pp\\ca) don (, bb M[ I was -ww.-e-A %d os ue_wh wood aM.

e 8 3/4 1-4 f FERMI - UNIT 2

q l

1 3/4.2 POWER DISTRIBUTION LIMITS BASES 1 The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident I will not exceed the 2200*F limit specified in 10 CFR 50.46.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE I j

The peak cladding temperature (PCT) following a postulated loss-of-coolant .1 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 dependent only secondarily on the rod to rod power distribution within an assembly. The peak clad {

j temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR !

times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod diviced by its local peaking factor. The limiting value for APLHGR is shown in Figures 3.2.1-1, 3.2.1-2 and 3.2.1-3 f a.4 .

__ _h;d.

m _. - !

,- __...........i e s ~3;W%7.~2}p,,y>wavrw 2fand 3.'2.T-v3vu w ..... . .g _ ._ y n

< analy The analysis was perform ed using %1s'b'a' sed EnT16ss-6f General Electric (GE ~ coolant ~a cM culation maodels are consistent with the *equirementsofAppendt to 10 CFR 50.

' fA complete' ion of each code em iloyew .., .a_ 'nalysiths presented in

. [

4 eference 1. es in this ana ysis compared. previous analyses can

$e broken down'as Md .t . .

e. Input Channes j I.

f -

Coefficients in the vaporizatic )

f 1. Corrected Vaporiza alcula t:

r correlation u . , - the REFLOOD cooe corrected.

( ss areas in the

! 2. Inco ( more accurate bypass areas - The ide were recalculated using a more accurate nique.

t E ,d*'

[ ,

Corrected guide tube thermal resistance.

5 Anzv; . h E k F -- " %wS FERMI - UNIT 2 8 3/4 2-1

l 1

i

{ ,

Insert 3 The Technical Specification MAPMGR value is the most limiting composite of the fuel mechanical design analysis MAPMGR and the ECCS MAPMGR.

Fuel Mechanical Design Analysis: NRC approved methods (specified in Reference 1) are used to demonstrate that all fuel rods in a lattice, operating at the bounding power history, meet the fuel design limits specified in Reference j(

~

1. This bounding power history is used as the basis for the fuel design analysis MAPMGR value.

I4CA Analysis: A LOCA analysis is performed in accordance with 10CFR50 Appendix K to demonstrate that the MAPMGR values comply with the ECCS limits specified in 10CFR50.46. The analysis is performed for the most limiting break size, break location, and single failure combination for the plant.

Only the most limiting MAPMGR values are shown in the Technical Specification figures for multiple lattice fuel. When hand calculations are required, these Technical Specification MAPMGR '

figure values for that fuel type are used for all lattices in that bundle.

For some fuel bundle designs MAPMGR depends only on bundle type and burnup. Other fuel bundles have MAPMGRs that vary axially depending upon the specific combination of enriched uranium and gadolinia that comprises a fuel bundle cross section at a particular axial node. Each particular combination of enriched j uranium and gadolinia, for these fuel bundle types, is called a lattice type. These particular fuel bundle types have MAPMGRs that vary by lattice (axially) as well as with fuel burnup.

Peference

1. " General Electric Standard Application for Reactor Fuel," i l NEDE-24011-P-A (latest approved revision).

1

POWER DISTRIBUTION LIMITS BASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)

" i _-=nce

^

L _. _ . .. . , , . . - , m J__ ._ _- --

3

. 1. Core CCFL pressure differential - 1 psi - Incorporate t /sumption' I that flow from the bypass to lower plenum must overc 1 psi li ssure drop in core.

2. In orate NRC pressure transfer assumpti The assumption used/j in t ' FE-REFLOOD pressure transfer he pressure is increas ing ,

was cha 1

(

N-A few of the change 4 f ct the de~nt calculation irrespective of i l

CCFL. These changes are I Q d 1 7 ; .

Input Change ?

a. - j

1. Break Areas ' 5BAbreakare s calculated more accurately. -

,b. Model Chan f 1. mproved Radiation and Conduction Calculation

~

corporation of 4

,/ CHASTE 05 for heatup calculation. {

62Et K tbaajentf4 cant +1 ant 4nputgeremet,9g';.40si-p oolan .

. - ' _ _u.t . .aiysis is Tresented in Bases Table' B-3M-1-1. ,

\' 3/4.2.2 ApRM SETPOINTS '

The fuel cladding integrity Safety Limits of Specification 2.1 were based j on a power distribution which would yield the design LHGR at RATED THERMAL l POWER. The flow biased simulated thermal power-upscale scram setting and flow I biased simulated thermal power-upscale control rod block functions of the APRM gg instruments must be adjusted to ensure that the MCPR does not become less than

.kS6 or that > 1% plastic strain does not occur in the degraded situation. The l,' ,

" cram settings and rod block settings are adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and MFLPD l indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.

i l FERMI - UNIT 2 8 3/4 2-2

1

. i

r, BASES TABLE B 3.2.1-1

)

SIGNIFICANT INPUT PARAMETERS TO THE )

LO$5-0F-C00LANT ACCIDENT ANALYSIS Plant Parameters:

Core THERMAL P0WER.................... 3430 MWt* which corresponds to j 105% of rated steam. flow Vessel Steam Output................... 14.86 x 108 lbm/hr which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure............ 1055 psia Design Basis Recirculation Line Break Area for: .{

u

a. Large Breaks 4.1 fts [,

b. Small Breaks 0.1 ft

./ Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kW/ft) FACTOR RATIO Initial Core 8x8 13.4 1.4 1.18

4d RaAcad %58 ~ ~ W.T (.4 Qg l A more detailed listing of input of each model and its source is presented in Section II of Reference 1 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 j powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION l RATE limit.

]

l l

l FERMI - UNIT 2 B 3/4 2-3 1

. . . . t POWER DISTRIBUTION LIMITS g )

. BASES v i

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady-state operating conditions as specified in Specification 3.2.3 a g derived from the established fuel cladding integrity Safety Limit MCPR.a' 1.v , and an analysis of abnormal operational $ ;

transients. For any abnormal operating transients analysis evaluation with the 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 y Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients 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 temperature decrease. The limiting transient yields the largest delta MCPR. When added to the Safety Lirit MCPR.a m c06, the required minimum l operating limit MCPR of Specification 3.2.3 is obtained and presented in l Figure 3.2.3-la g 1.L W 3 A s,%M c. .

bW The evaluation of a given transient begins with the system initial parameters shown inUFSAR Table 150.0-1 that are input to a GE-core dynamic behavior transient computer program. T'e ::d: .. J 6. evelum, v . . . . . . m. i o n

. 4. ;e J,..--; W h u nn-pal r,d tt; ;=;== 2:e : n: ;7;.,3.,;2a;;c,c, ,

, a, J.a ; ;2t ie d= - it'd I- 590-19902 '" 02t~'t! Of thir ? sMr ;I; s 5 tir i;.. ; m .ial rn,rA 'v. . the i..;,;t fr fr-th r :::!y :: Of the 15: m:113 1:esiti .g

[ {9

Lu.. iw w i Li es . : ..;1: 05:nn:1 t =n $ nt th: m;l hyd =;li; T C ;;d: d::rit;d in G E-i M ^I4I.- The principal result of this evaluation is the reduction in I "M l

MCPR caused by the transient.

'n M M eA"E..

The purpose of the K7 factor of Figure 3.2.3-2 is to define operating limits at other than rated core flow conditions. At less than 100% of rated flow the required MCPR is the product of the MCPR and the Kg factor. The K f factors assure that the Safety Limit MCPR will not be violated during a flow increase transient resulting from a motor generator speed control failure. The Kg factors may be applied to both manual and automatic flow control modes.

The K, factor values shown in Figure 3.2.3-2 were developed generically and are applicable to all BWR/2, BWR/3, and BWR/4 reactors. The Kg factors were derived using the flow control line corresponding to RATED THERMAL POWER at rated core flow.

For the manual flow control mode, the Kg factors were calculated such that 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. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the Kf . h FERMI - UNIT 2 B 3/4 2-4

Insert 4 The MCPR curves illustrated in Figures-3.2.3-1 thru 3.2.3-1B were derived as described - above for the following assumed operating conditions:

Curve A - MCPR limit with turbine bypass system, moisture l separator reheater systems in service and CCC (Control Cell Core) operational mode (A2 rods, Al shallows inserted less than or equal to notch position 36, all peripheral rods, and all rods .

inserted to position 46) inserted in the core. 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 operational mode (any control rod inserted in the ,

core). The operating domain includes 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 moisture separator reheater system available or the main turbine bypass system available and the moisture separator reheater system 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 100% power / flow region and the extended load line region with 100% power and reduced flow.

Curve A provides the MCPR limit assuming operation above 25 percent RATED THERMAL POWER with the turbine bypass system and moisture separator reheater in service. The curve was developed based upon i the operating MCPR limits for a rod withdrawal error transient '

(UFSAR, Section 15.4.2) for operating within the CCC control rod .

patterns and a Main Turbine Trip with Turbine Bypass Failure !

transient (UFSAR, Section 15.2.3). CCC control rods are A2 rods, Al shallow rods (inserted lens than or equal to notch position 36),

all peripheral rods, and all rods inserted to position 46. The analysis of the Main Turbine Trip with Turbine Bypass Failure takes credit for the steam flow to the moisture separator reheater.

Curve B provides the MCPR limit assuming operation above the 25 percent RATED THERMAL POWER with the turbine bypass system and

moisture separator reheater system in service and non-CCC control rods inserted in the core. Non-CCC control rods are all rods excluding A2 rods, Al shallow rods (inserted less than or equal to notch position 36), all peripheral rods, and all rods inserted to position 46. The curve was developed based upon the operating MCPR limits for a rod withdrawal error transient (UFSAR, Section 15.4.2) for any operating withdrawal sequence.

Curve C provides the MCPR limit assuming operation above the 25 percent RATED THERMAL POWER with the moisture separator reheater operable and turbine bypass system inoperable or the moisture separator reheater inoperable and the turbine bypass system operable. The curve was developed based upon the operating MCPR <

11tnits for several combinations of Feedwater Controller Failure.

Curve D provides the MCPR limit assuming operation above the. 25 percent RATED THERMAL POWER with both the moisture separator reheater inoperable and the turbine bypass system inoperable. The curve was developed based upon the operating MCPR limits from the Feedwater Controller Failure.

9

f POWER DISTRIBUTION LIMITS 1 BASES MINIMUM CRITICAL POWER RATIO (Continued)

For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at RATED THERMAL POWER and rated thermal flow.

The K factors shown in Figure 3.2.3-2 are conservative for the General ,

f Electric plant operation because the operating limit MCPRs of Specification 3.2.3 are greater than the original 1.20 operating limit MCPR used for the generic derivation of K7.

At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial .startup testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum ,

j recirculation pump speed. The MCPR margin will thus be demonstrated such that '

future MCPR evaluation below this power level will be shown to be unnecessary.

The daily requirement for calculating MCPR when THERMAL POWER is greater than

/ or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod l

changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will b'e known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

i 3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in I any rod is less than the des.ign linear heat generation even if fuel pellet densification is postulated.

References:

1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.

l

~2. R. C. Linf., d, ^nalytical 'tthede of Pioni T. ansi:nt O aluationa Lc tk CE St!", NE00-10002, Tanwary 1373.

3. ?_M 4fic=+4nn nf +ke One O!-:n:ienal cere Trentient u;gc; ;cc Be4! h 0 un te oge,-te 3, ugnn_3q5$ , cctcy 137g d--

'^$" ^' ^ enmniitor Program fne the T, eng ygt an !p{ :( g;z];

( Ctrn:1, T: chic:1.02: ription, WEDE-25110, J nuo,, 1000._

?.. " Cma.cd %c.bic. Mcmau A AeebMon Qoc bche Ad '"

WEM -Mon- ?- % 1 cb.t. o+ecovs.d endh .

l FERMI - UNIT 2 / B 3/4 2-5 I

I!gE,.E i b g . LIST OF FIGURES -

FIGURE PAGE 3.1.5-1 SODIUM PENTABORATE VOLUME / CONCENTRATION REQUIREMENTS ................................... 3/4 1-21 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (NAPLHGR) VS. AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE SCR183 .................. 3/4 2-2 j 3.2.1-2 MAXIMUMAVERAGEPLANARLINEARHF1TGENERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, 1NITIAL CORE FUEL TYPE SCR233 .................. 3/4 2-3 y_. . t. . . ) ,,, . . . . . . - - . . . .. . . y on e , ,,,

MI

.................. 3/4 2-4 3.2.3-1 9!"!"LH C"!?r ~~ ?2 2rr ("?""} u,nens

^.T ^ ^JG F ;.e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-8 a i

3.2.3-2 FLOW CORRECTIONf(K ) FACTOR . . . . . . . . . . . . . . . . . . . . 3/4 2-9 l 3.4.1.1-1 THERMAL POWER VS CORE FLOW ..................... 3/4 4-3 3.4.6.1-1 MINIMUM REACTOR PRESSURE VESSEL METAL TEMPERATURE -

VS. REACTOR VESSEL PRESSURE .................... 3/4 4-21 t

4.7.5-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST ..... 3/4 7-21 B 3/4 3-1 REACTOR VESSEL WATER LEVEL ...................... B 3/4 37 8 3/4.4.6-1 FAST NEUTRON FLUENCE (E>1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE ........................ B 3/4-4-7 8 3/4.6.2-1 LOCAL POOL TEMPERATURE LIMIT .................... B 3/4 6-5 B 3/4.7.3-1 ARRANGEMENT OF SNORE BARRIER SURVEY POINTS ...... B 3/4 7-6 5.1.1-1 EXCLUSION AREA ................................. 5-2 5.1.2-1 LOW POPULATION ZONE . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5.1.3-1 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID -

EFFLUENTS ...................................... 5-4 6.2.1-1 0FFSITE ORGANIZATION . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 6.2.2-1 UNIT ORGANIZATION .............................. 6-4 FERMI - UNIT 2 xxi

l Insert 5 3.2.1-3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR Versus Average Planar Exposure, Reload Fuel Type BC318D). . . . . . . . . . . . 3/4 2-4 3.2.1-4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR Versus Average Planar Exposure, Reload Fuel Type BC318E). . . . . . . . . . . . 3/4 2-4 3.2.3-1 BOC to-12,700 MWD /ST, Minimum Critical Power Ratio (MCPR) Versus Tau at Rated Flow . . . . . 3/4 2-8 .

l 3.2.3-1A 12,700 MWD /ST to 13,700 MWD /ST, Minimum i' l

Critical Power Ratio (MCPR) Versus Tau at Rated Flow . . . . . . . . . . . . . . . . . . 3/4 2-8A 3.2.3-1B 13,700 MWD /ST to EOC, Minimum Critical Power Ratio (MCPR) Versus Tau at Rated Flow . . . . . . . . . . . . . . . . . . . . . 3/4 2-8B l

I, INDEX l

!i

r. '

LIST OF TABLES I

TABLE ,

PAGE 1.1 1-9 SURVEILLANCE FREQUENCY NOTATION,................

1.2 OPERATIONAL. CONDITIONS ......................... 1-10 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ...................................... 2-4

. .. ! _ . . _ _ . . ^: 1 _: _ Z : ^

T"*""'" ^" ^ - -": -

TZ'.. CL.".-^..Z .,a.i. 6inii ,.................. .. ^R ,

I

^^~~T.' W ,ii .................................... e : '. l 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION ...... 3/4 3-2 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES ....... 3/4 3-6 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ...................... 3/4 3-7 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION ............ 3/4 3-11 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS .. 3/4 3-15 3.3.2-3 ISOLATION ACTUATION SYSTEM INSTRUMENTATION RESPONSE TIME .................................. 3/4 3-18 I

4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS ................................... 3/4 3-20 3.3.3-1 EMERGENCY CORE COOLING SYSTEM ACTUATION

- INSTRUMENTATION ................................ 3/4 3-24 <

3.3.3-2 EMERGENCY CORE COOLING SYSTEM ACTUATION i INSTRUMENTATION SETPOINTS ...................... 3/4 3-27 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES ... 3/4 3-29 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS ...... 3/4 3-30 3.3.4-1 ATWS RECIRCULATION PUMP TRIP SYSTEM q INSTRUMENTATION ................................ 3/4 3-33 3.3.4-2 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS . . . . . . . . . . . . . . . . . . . . . . 3/4 3-34

( i FERMI - UNIT 2 xxii 1

I

ML20248J493

Text

SUMMARY

References:

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