ML20012B577

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Proposed Tech Specs Re Core Operating Limits Rept, Instrumentation That Initiates Control Rod Blocks & Engineered Safeguards Compartments Cooling & Ventilation
ML20012B577
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 03/08/1990
From:
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20012B574 List:
References
NUDOCS 9003150348
Download: ML20012B577 (50)
v • d • e


Text

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, ATTACHMENT 2 1

L.- .--

I PEACH BOTTOM ATOMIC POWER STATION l UNITS 2 and 3 j l

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V Docket Nos. 50-277 l 50-278 r

. License Nos. DPR-44 '

DPR-56 (

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l REVISED TECHNICAL SPECIFICATION PAGES .

i List of Attached Pages Unit 2 Unit 3 i iv iv iva- vi vi 1 1 10 10 -11a lla 17 17 24 24 40

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40- 73 73 74 74 74a 74a 133a 133a 133b 133b-- 133c 133c. 133d 133d 140

'140 140a 140a 140b~

-140b 140c 140c 141a 141a 142

^142 256'

.142d 256a 1429 256 256a

$ 7 m.

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PBAPS i LIST DTTTGURES  !

Figure Title Page r

.1.1-1 APRM Flow Bias Scram Relationship To Normal 16 Operating Conditions  !

4.1.1 Instrument Test Interval Determination Curves 55  !

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-4.2.2 Probability of System Unavailability vs. Test Interval 98 f

3.3.1 SRM Count Rate vs. Signal-to-Noise Ratio 103a l 3.4.1- DELETED 122 l 3.4.2 DELETED 123 .

i i 3.5.K.1-1 DELETED 142 l

3.5.K.1-2 DELETED 142

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3.5.K.1-3 DELETED 142 {

3.5.K.2 - DELETED 142 1

3.5.K.2-1 DELETED 142 i 3.5 K.2-2 DELETED 142 .

3.5.K.2-3 DELETED 142  !

3.5.K.3 DELETED 142b 3.5.1.E DELETED 142d l 3.5.1.F DELETED 142e  ;

3.5.1.G DELETED: 142f '

3 5.1'.H DELETED 1429 l r

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PBAPS LIST OF FIGUkES  !

i Figure Title Page

-3.5.1.1 DELETED 142g

! 3.5.1.J DELETED 1429 3.5.i.K DELETED 142g

[ 3.5.1.L DELETED 142g 3.5.1.M DELETED 142g 3.5.1.N DELETED 142g 3.5.1.0 DELETED 1429 3.6.1 Minimum Temperature for Pressure Tests 164 such as required by Section XI 3.6.2 Minimum Temperature for Mechanical Heatup 164a or Cooldown following Nuclear Shutdown 3.6.3 Minimum Temperature for Core Operation 164b [

(Criticality)  !

3.6.4 Deleted 164c  !

3.6.5 -Thermal Power and Core Flow Limits 164d 3.8.1 Site Boundary and Effluent Release Points 216e 6 2-1

. Management Organization Chart 244 6.2-2 Organization for Conduct of Plant Operations 245 -

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Unit 2 PBAPS i

LIST OF TABLES Table _ Title P,ag l 4.2.8 Minimum Test and Calibration frequency 81 a for CSCS j

4.2.C Minimum Test and Calibration Frequency 83 for Control Rod Blocks Actuation l l

4.2.0 Minimum Test and Calibration Frequency 84 for Radiation Monitoring Systems 4.2.E Minimum Test and Calibration Frequency 85 for Drywell Leak Detection 4.2.F Minimum Test and Calibration Frequency 86 fcr Surveillance Instrumentation 4.2.G Minimum Test and Calibration frequency 88 for Recirculation Pump Trip '

3.5.K.2 DELETED 133d 3.5.K.3 DELETED 133d 4.6.1 In-service Inspection Program for Peach 150 Bottom Units 2 and 3 3.7.1 Primary Containment Isolation Valves 179 3.7.2 Testable Penetrations with Double 184 0-Ring Seals '

3.7.3 Testable Penetrations with Testable 184  :

Bellows 3.7.4 Primary Containment Testable Isolation 185 Valves 4.8.1 Radioactive Liquid Waste Sampling and 216b-1 Analysis 4.8.2 Radioactive Gaseous Waste Sampling and 216c-1 Analysis 4.8.3.a Radiological Environmental Monitoring 216d-1 Program 4.8.3.b Reporting Levels for Radioactivity by 216d-5 Concentrations in Environmental Sample

-vi-L.

l Unit 2 PBAPS 1.0 DEFINITIONS The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieved.

Alteration of the Reactor Core - The act of moving any component in the region above '

the core support plate, below the upper grid and within the shroud with the vessel head removed and fuel in the vessel.

Normal control rod movement with the control drive hydraulic system is not defined as a core alteration. Normal movement of in-core instrumentation and the traversing in-ctre probe is not defined as a core alteration.

Average Planar Linear Heat Generation Rate (APLHGR) - The APLHGR shall be applicable to a specific planar height and is equal to the sum of the heat generation rate per unit length of fuel rod, for all the fuel rods in the specific bundle at the specific ,

height, divided by the number of fuel rods in the fuel bundle at that height.

Channel - A channel is an arrangement of a sensor and associated components used to ,

evaluate plant variables and produce discrete outputs used in logic. A channel t terminates and loses its identity where individual channel outputs are combined in i logic.

L Cold Condition - Reactor coolant temperature equal to or less than 212 F.

Cold Shutdown - The reactor is in the shutdown mode, the reactor coolant temperature equal to or less than 212 F, and the reactor vessel is vented to atmosphere.

Core Operatina limits Report (COLR) - The COLR is the unit-specific document that provides the core operating limits for the current Operating Cycle. These cycle-specific core operating limits shall be determined for each Operating Cycle in accordance with specification 6.9.1.e. Plant operation within these limits is

  • addressed in individual Specifications.

Critical-Power Ratio (CPR) - The critical power ratio is the ratio of that assembly >

power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor condition of interest as calculated by application of the GEXL correlation. (ReferenceNEDO-10958).

Dose Equivalent I-131 - That concentration of I-131 (C1/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, 1-133 I-

-134, and 1-135 actually present.

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PBAPS UNIT 2 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING I

2.1.A (Cont'd) j In'the event of operation with a maximum fraction of )

limiting power density (MFLPD) j greater than the fraction of l ratedpower(FRP),thesetting j shall be modified as follows.  ;

S<(0.58W+62%-0.58AW) where, t FRP = fraction of rated thermal power (3293MWt) l MFLPD = maximum fraction of .

limiting power density where the limiting  ;

power density is the value specified in the CORE i OPERATING LIMITS REPORT.  ;

The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value-of 1.0, in which case the actual i..

operating value will be used. '

2. APRM--When the reactor mode '

' switch is in the STARTUP position, the APRM scram shall l be set at less than or L equal to 15 percent of rated power. '

3. IRM--The IRM scram shall be set at less than or equal to

! 120/125 of full scale.

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  • init 2  ;

PBAPS l l SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING B. Core Thermal Power Limit B. APRM Rod Block Trip Setting i (Reactor Pressure 5 800 psia) j SRB $ (0.58 W + 50% - 0.586W) where:

L FRP = fraction of rated  ;

thermalpower(3293MWt).  !

MFLPD = maximum fraction of -

limiting power density >

where the limiting power density is the value specified in the CORE ,

OPERATING LIMITS REPORT.

The ratio of FRP to MFLPD  :

shall be set equal to 1.0 i, unless the actual operating i value is less than the design value of 1.0, in which case +

the actual operating value will be used.

C. Whenever the reactor is in the C. Scram and isolation- > 538 in. above

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shutdown condition with reactor low water vessel zero  !

irradiated fuel in the reactor level (0" on level vessel, the water level shall instruments) I not be less than minus 160 -

inches indicated level (378 inches above vessel zero).

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Unit 2 PBAPS 2.1 BASESr FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the Peach Bottom Atomic Power Station Units have been analyzed throughout the spectrum of planned operating conditions up to or above the thermal power condition required by Regulatory Guide 1.49. The analyses were based upon plant operation in accordance with the operating map given in Figure 3.7.1 of the FSAR. In addition, 3293 MWt is the licensed maximum power level of each Peach Bottom Atomic Power Station Unit, and this represents the maximum steady state power which shall not knowingly be exceeded.

Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and axial power shapes. These factors are selected conservatively with respect to their effect on the applicable transient results as determined by the current analysis model.

Conservatism incorporated into the transient analyses is documented in References 2 and 3. j l

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Unit 2 2.1 BASEST (Cont'd).

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n L. References n-

1. Linford, R. B., " Analytical Methods of Plant E

Transient Evaluations for the General Electric Boiling Water Reactor", NEDO 10802, February 1973.-

2. " General Electric Standard Application for Reactor.

Fuel", NEDE-240ll-P-A.(as amended).

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3. PECo-FMS-0006, " Methods for_ Performing BWR Reload Safety Evaluations"'(latest approved revision)

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Unit 2 PBAPS NOTES FOR TABLE 3.1 d (Cont'd)

10. The APRM downscale trip is automatically bypassed when the IRM '

instrumentation is operable and not high.

11. An APRM will be considered operable if there are at least 2 LPRM inputs per level and at least 14 LPRM inputs of the normal complement.
12. This equation will be used in the event of operation with a maximum fractionoflimitingpowerdensity(MFLPD)greaterthanthefractionof '

rated power (FRP), where:

FRP = fraction of rated thermal power (3293 MWt).

MFLPD = maximum fraction of limiting power density where the  ;

limiting power density is the value specified in the CORE OPERATING LIMITS REPORT. .

The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the  ;

actual operating value will be used.

W= Loop Recirculation flow in percent of design. W is 100 for core flow of 102.5 million ib/hr or greater.

AW = the difference between two loop and single loop effective recirculation drive flow rate at the same core flow. During single loop operation, the reduction in trip setting (-0.58 '

AW) is accomplished by correcting the flow input of the flow biased High Flux trip setting to preserve the original (two loop) relationship between APRM High Flux setpoint and ,

recirculation drive flow or by adjusting the APRM Flux trip setting. AW = 0 for two loop operation.

Trip level setting is in percent of rated power (3293 MWt).

13. See Section 2.1.A.I.

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TABLE.3.2.C o Unit 2 INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS Minimum No. Instrument of Operable Trip Level Setting Number of Instrument Action Instrument Channels Provided Channels Per by Design Trip System 4 (2) APRM Upscale (Flow

$(0.58W+50-0.58aW) x 6 Inst. Channels (10)

Biased) FRP MFLPD I 4 APRM Upscele (Startup 2 12% 6 Inst. Channels (10)

Mode) 4 APRM Downscale 12.5 indicated on 6 Inst. Channels (10) scale 1 (2)(7)(11) Rod Block Monitor 1(0.66W+(N-66)-0.66aW)x 2 Inst. Channels 4 (flow Blased) FRP (1) y MFLPD with a maximum of 2N%

1 (7) Rod Block Monitor 12.5 indicated on Downscale 2 Inst. Channels (1) scale 6

IRM Downscale (3) 12.5 indicated on 8 Inst. Channels scale (10) 6 IRM Detector not in (8) 8 Inst. Channels (10)

Startup Position 6 1RM Upscale

$108 indicated on 8 Inst. Channels- (10) '

scale 2 (5) SRM Detector not in (4)

Startup Position 4 Inst. Channels (1) 5 2 (5)(6) SRM Upscale

$10 counts /sec. 4 Inst. Channels (1) 1 Scram Discharge $25 gallons ~I Inst. Channel Instrument Volume (9)

High Level

e . Unit 2 PBAPS 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 and IRM blocks need not be operable in "Run" mode, and the APRM 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 lasts longer than seven days, the system shall be tripped. If the first column cannot be met for both trip systems, the systems shall be tripped.

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2. This equation will~be used in the event of operation with a maximum fraction of -

limiting power density (MFLPD) greater than the fraction of rated power (FRP)  !

where:

FRP = fraction of rated thermal power (3293 MWt) '

MFLPD = maximum fraction of limiting power density where the limiting power density is the value specified in the CORE OPERATING LIMITS REPORT.  !

The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.

W = Loop Recirculation flow in percent of design. W is 100 for core flow of 102.5 million lb/hr or greater.

Trip level setting is in percent of rated power (3293 MWt).

AW is the difference between two loop and single loop effective recirculation drive flow rate at the same core flow. During single loop operation, the reduction in trip setting is accomplished by correcting the flow input of the flow biased rod block to preserve the original (two loop) relationship between the rod block setpoint and recirculation drive flow, or by l adjusting the rod block setting. AW = 0 for two loop operation, i L -3. IRM downscale is bypassed when it is on its lowest range.

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4. This function is bypassed when the count rate is 1 100 cps.
5. One of the four SRM inputs may be bypassed.
6. This SRM function is bypassed when the IRM range switches are on range 8 or above.
7. The trip is bypassed when the reactor power is 1 30%.
8. This function is bypassed when the mode switch is placed in Run.

1 Unit 2 PBAPS NOTES FOR TABLE 3.P.C (Cont.)

9. If'the number of operable channels is less than required by the minimum operable channels per trip function requirement, place the inoperable '

channel in the tripped condition within one hour. This note is applicable in the "Run" mode, the "Startup" mode and the " Refuel" mode if more than one control rod is withdrawn.

10. For the Startup (for IRM rod block) and the Run (for APRM rod block) positions of the Reactor Mode Selector Switch and with the number of OPERABLE channels:
a. One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour,
b. Two or more less than required by the Minimum OPERABLE Channels per  :

Trip function requirement, place at least one inoperable channel in '

the tripped condition within one hour.

11. The value of N is specified in the CORE OPERATING LIMITS REPORT.

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1 -Q PBAPS Unit 2 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.I Average Planar LHGR 4.5.I Average Planar LHGR

.During power operation, the APLHGR The APLGHR for each type of fuel for each type of fuel as a function as a function of average planar of axial location and average planar exposure shall be checked daily exposure shall be within limits during reactor operation at based on applicable APLHGR limit > 25% rated thermal power.

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values which have been determined '

by approved methodology for the ,

-respective fuel and lattice types.  !

When hand calculations are required, i the APLHGR for each type of fuel as  ;

a function of average planar  ;

exposure-shall not exceed the limit i for the most limiting lattice (excluding natural uranium) specified ,

in the CORE OPERATING LIMITS REPORT >

during two recirculation loop '

operations. During single loop operation, the APLHGR for each fuel i type shall not exceed the '

values multiplied by the r,aboveeduction ,

factors specified in the CORE OPERATING  !

LIMITS REPORT. If at any time during {

operation it is determined by normal  :

surveillance that the . limiting value  :

of APLHGR is being' exceeded, action '

shall be initiated within one (1) hcur

.to rectorc-ALPEGR to within prescribed limits. If the APLHGR is not returned i to within prescribed limits within.five (5)-hours, reactor  ;

power shall be decreased at a rate which would bring the reactor 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 /> unless APLHGR is returned to within limits during this period. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

3.5.J Local LHGR' 4.5.J Local LHG_R_

During power operation, the linear The LHCR as a function of core *

-heat generation rate (LHGR) of any height shall be checked daily rod in any fuel assemb3y at any during reactor operation at axial location shall not exceed ~> 25% rated thermal power.

design LHGR. >

LHGR _< LHGRd LHGRd = Design LHGR The values for Design LHGR for each fuel type are specified in the CORE OPERATING LIMITS REPORT.

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. LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS i 3.5.J Local LHGR (Cont'd) ~

If at any time during operation it 10' determined by normal surveillance ,

th0t limiting value for LHGR is be-

  • ing exceeded, action shall be initi- ,

Ctcd within one (1) hour to restore

  • LHGR to within prescribed limits.

If the LHGR is not returned to '

within prescribed limits within five (5) hours, reactor power shall be decreased &t a rate which '

would bring the reactor 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 />  !

unless LHGR is returned to within  ;

limits during this period. Surveil-  ;

1cnce and corresponding action shall continus until reactor operation is within the prescribed limits.

l 3.5.K Minimum Critical Power 4.5.K Minimum Critical Power Ratio (MCFR) Ratio (MCPR)

1. During power operation the MCPR

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1. MCPR shall be checked daily for the' applicable incremental during reactor power operation cycle core average exposure and at >25% rated thermal power.

l for each type of fuel shall be 2. Except as provided in Specif1- '

Cqual to or greater than the value cation 3.5.K.3, the verifica-  ;

lgiven-in Specification 3.5.K.2 or tion of the applicability of i 3.5.K.3 times Kf,.where Kf is as 3.5.K.2.a Operating Limit MCPR cp;cified in the CORE OPERATING Values shall.be performed every LIMITS REPORT. If;at any 120 operating days by scram time time during operation it is testing 19 or more control rods-  ;

datermined by normal surveillance on a rotation basia and per- i th0t the limiting value for MCPR forming the following:

is being exceeded, action shall be

. initiated within one (1) hour to a. The average scram time to rCstore MCPR to within prescribed the 20% insertion position linits. If the MCPR is not returned shall be '

.to within prescribed limits within five (5) hours, reactor Tave<IB _

power shall be decreased at a b. The average scram time to rote which would bring the the 20% insertion position.

rocetor to the cold shutdown is determined as follows:

condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />

.unless MCPR is returned to 7Iave = Ni i within limits during this period. i=1 -

Surveillance and corresponding n Cetion shall continue until re- >~~ N i cetor operation is within the i=1 proscribed limits, where: n = number of surveillance tests performed to date in the cycle.

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o Unit 2 PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.K Minimum Critical Power 4.5.K Minimum Critical Power

' ' Ratio (MCPR) (Cont'd) Ratio (MCPR) (Cont'd)

2. Except as specified in 3.5.K.3, Ni = number of active control i the Operating Limit MCPR Values rods measured in the ith shall be as specified in the surveillance test.

CORE OPERATiHG LIMITS REPORT y for when Li= average scram time to the 20% insertion position a) requirement 4.5.K.2.a of all rods measured in is met, and for when the ith surveillance test, b) requirement 4.5.K.2.a is c. The adjusted analysis mean not met, where: r scram time (T3 ) is calculated i as follows:

[= Iave - IB 0.90 - T3 {B=p+1.65[ N1 \1/2 l, ,

3. If the Surveillance Requirement k, of Section 4.5.K.2 to scram Ni ]

i=1 time-test control rods is .

not performed, the Operating Where:

Limit MCPR values shall be as specified in the p =

mean of the distribution '

CORE OPERATING LIMITS REPORT for average scram insertion for this condition.

time to the 20% position =

0.694 sec. ,

N1 = total number of active 'l control rods measured in i specification 4.3.C.1 7 = standard deviation of the distribution for average scram insertion ~ time to the 20% position = 0.016.

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'5 Tables 3.5.K 2 and 3.5 K.3 have been

- removed from former Technical Specification pages 133d and 13?,h respectively, and the-associated information has been-relocated to the Core Operating Limits Report.

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PBAPS  !

3.5 BASES (Cont'd.) I H.- -Engineered Safeguards Compartments Cooling and Ventilation l r

One unit cooler in each pump ~ compartment is capable of providing

.cdequate ventilation flow and cooling. Engineering analyses indicated that the temperature rise in safeguards compartments without adequate vontilation flow or cooling is such that continued operation of the '

cafeguards equipment-or associated auxiliary equipment cannot be assured. Ventilation associated with the High Pressure Service Water Pumps is also associated with the Emergency Service Water pumps, and is specified in Specification 3.9.

F I. Average Planar LHGR This specification assures that the peak cladding temperature I following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR Part 50,_ Appendix K. ,

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat

- g3neration 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 aseembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which_is equal to or less than the design LHGR. This LHGR times 1.02 is used in the haat-up code along.with the exposure dependent steady state gap conductance and rod-to-rod local peaking factors. The Technical Specification APLHGR is the LHGR of the highest powered rod divided by its local peaking factor. The limiting value for APLHGR is shown in

- the applicable-figure for each fuel type in the CORE OPERATING LIMITS REPORT.

Only the most. limiting APLHGR operating limits are shown in the i .

figures for the multiple lattice fuel types. Compliance with the lettice-specific APLHGR limits is ensured by using'the process l 1

computer. When an alternate method to the process computer is required (i.e. hand calculations and/or alternate computer simulation), the most limiting lattice APLHGR limit for each fuel type shall be applied to every lattice of that fuel type.

'The calculational procedure used to establish the APLHGR is based on a loss-of-coolant accident analysis. The analysis was performed using G2neral Electric (G.E.) calculational models which are consistent with

.the requirements of Appendix K to 10 CFR Part 50. A complete discussion of each code employed in the analysis is presented in R3ference 4. Input and model changes in the Peach Bottom loss-of-coolant analysis which are different from the previous analyses ptrformed with Reference 4 are described in detail in Reference 8.

These changes to the analysis include: (1) consideration of the

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counter current flow limiting (CCFL) effect, (2) corrected code

-inputs, and (3) the effect of drilling alternate flow paths in the bundle lower tie plate.

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Unit 2 PBAPS 3.5. BASES (Cont'd) l

-J.- Local LHGR This specification assures that the linear heat generation rate in any 8X8 fuel rod is less than the design linear heat generation. The m ximum LHGR shall.be checked daily during reactor operation at > 25%

power to determine if fuel burnup, or control rod movement'has caused changes in power distribution. For LHGR to be at the design LHGR balow 25% rated thermal power, the peak local LHGR must be a factor of cpproximately ten (10) greater than the average LHGR which is precluded by a considerable margin when employing any permissible control rod pattern.

K. Minimum Critical Power Ratio (MCPR)

Operating Limit MCPR

-The required operating limit MCPR's at steady state operating

-conditions are derived from the established fuel cladding integrity S2fety Limit MCPR and analyses of the abnormal operational transients presented'in Supplemental Reload Licensing Analysis and References 7 and-10. For any abnormal operating transient 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 d3 crease below the Safety. Limit MCPR at any time during the transient cssuming instrument trip setting given in Specification 2.1. ,

To assure that-the fuel cladding integrity Safety Limit is not violated 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-transients evaluated are as described in References 7 and 10. l 1

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s . . -Unit 2 PBAPS 3.5.A. , BASES (Cont'd)

.The largest reduction in critical power ratio is then added to the fuel: cladding integrity safety limit MCPR to establish the MCPR '

Loperating. Limit for each fuel type, l

Analysis of the abnormal operational transients is presented in

-Raferences 7 and 10. Input data and operating conditions used in this  !

cnalysis are shown in References 7 and 10 and in the Supplemental  !

Raload Licensing Analysis..

'l 3.5.L. Average Planar LHGR (APLHGR), Local LHGR and Minimum Critical Power Ratio (MCPR)

In the event that the calculated value of APLHGR, LHGR or MCPR exceeds its limiting value, a determination is made to ascertain the cause and initiate corrective actions to restore the value to within prescribed i limits. The status of all indicated limiting fuel bundles is reviewed j cs well as input data associated with the limiting values such as l power distribution, instrumentation data (Traversing In-Core Probe - l TIP, Local Power Range Monitor - LPRM, and reactor heat balance l Linstrumentation), control rod configuration, etc., in order to i L

1 datermine whether the calculated values are valid. 1 In the event that the review indicates that the calculated value  !

cxceeding liraits is valid, corrective action is immediately undertaken '

to restore the value to within prescribed limits. Following corrective action, which may involve alterations to the control rod configuration and consequently changes to the core power distribution, rovised instrumentation data, including changes to the relative neutron flux distribution, for up to 43 in-core locations is obtained sna the power distribution, APLHGR, LHGR and MCPR calculated.-

Corrective action is initiated within one hour of an indicated value cxceeding limits and verification that the indicated'value is within prescribed limits is obtained within five hours of the initial

-indication.

In the event that the calculated value of APLHGR, LHGR or MCPR exceeding its limiting value is not valid, i.e., due to an erroneous-instrumentation indication, etc., corrective action is initiated within one-hour of an indicated value exceeding limits. Verification that the indicated value is within prescribed limits'is obtained within five hours of the initial indication. -Such'an-invalid indication would not be a violation of the-limiting condition for operation and therefore would not constitute a reportable occurrence, i

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Unit 2 PBAPS I 3.5.L. BASES (Cont'd)  ;

Operating experience has demonstrated that a calculated value of

'APLHGR, LHGR or MCPR exceeding its limits value predominantely  !

occurs due to this.latter cause. This experience coupled with the l extremely unlikely occurrence of concurrent operation exceeding '

APLHGR, LHGR or MCPR and a Loss-of-Coolant Accident or applicable Abnormal Operational Transients demonstrates that the timee required to initiate corrective action (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) and restore the calculated value of APLHGR, LHGR or MCPR to within prescribed limits (5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) are adequate.

3.5.M. References

1. " Fuel Densifica31on Effects on General Electric Boiling Water Reactor Fuel", Supplements 6, 7 and 8, NEDM-10735, August 1973,
2. Supplement 1 to Technical Report on Densifications of General

-Electric Reactor Fuels, Decembec 14, 1974 (Regulatory Staff). '

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

4.- General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE 20566 (Draft), August 1974.

5. General Electric Refill Reflood Calculation (Supplemerit to SAFE i Code Description) transmitted to the USAEC by letter, G.-L.

Gyorey to Victor Stello, Jr., dated December 20, 1974.

6. DELETED.

'7. " General-Electric Standard Application for Reactor Fuel", NEDO- :t 240ll-P-A (as amended).

i

8. Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2, NEDO-24081, December 1977, and for Unit 3, NEDO-24082, December 1977.
9. Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2, Supplement 1, NEDE-24081-P, November 1986.
10. PECo-FMS-0006, " Methods for Performing BWR Reload Safety Evaluations" (latest approved revision)

-140c-

, - -. - - - . - ~

4 1

'. .. . Unit 2 PBAPS i.5.K Minimum Critical Power Ratio (MCPR) - Surveillance Requirement At core thermal power levels less than or equal to 25%, .the reactor will be operating at, minimum recirculation pump speed and the moderator void content will be very usmall. .For-all designated control rod patterns which may be employed at this point,

, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a consider &ble margin. With this low void content, any inadvertent core flow-increase would only place operation in a more conservative mode relative to ,

MCPR. During initial start-up testing of the plant, a MCPR evaluation will be made at-25f thermal power level with minimum recirculation pump speed. The MCPR margin iwill thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR above

~ 25% rated thermal power is sufficient since power distribution shifts are very slow when there heve not been significant power or control rod changes. The requirement

,for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a themal limit.

4.5.L MCPR Limits for Core Flows Other than Rated The purpose of the Kr factor is to define operating limits at other than rated flow conditions. At less than 100% flow the required MCPR is the product of the cperating limit MCPR and the Kr fcctor. Specifically, the Kr factor provides

the required thermal margin to protect against a flow increase transient. The most '

limiting transient initiated from less than rated flow conditions is the recirculation pump speed up caused by a motor-generator speed control failure.

For' operation in the automatic flow control mode. the Kr factors assure that the cperating limit MCPR will not be violated should the most limiting transient occur at

.lsss than rated flow. In the manual flow control mode, the Kr factors assure that the Safety Limit MCPR will not mhe violated for the same postulated transient Gvent.

The.Kr factor curves in the CORE OPERATING LIMITS REPORT were developed j

,g2nerically and-are applicable to all BWR/2, BWR/3, and BWR/4 reactors. :The Kr

factors were derived using the flow control line corresponding to rated thermal power at rated core' flow.

For the manual flow control mode, the Kr factors were calculated such that at the maximum flow rate (as limited by the pump scoop tube set point) and the corresponding core power (along the rated flow control line), the liraiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPR's were calculated at different points along the rated flow control line corresponding to different. core flows. The ratio of the MCPR calculated et a given point of the core flow, divided by the operating limit MCPR determines the Kr.

Fcr 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 power and flow.

The Kr factors specified in the CORE OPERATING LIMITS REPORT are acceptable for Peach Bottom operation because the operating limit MCPR is greater than the original 1.20 operating limit MCPR used for the generic derivation of Kr.

-141a-

to- 4 1 ~

r l40 et t-i.- , c,

.; q[ ,.. Unit 2 .

PBAPS L

~ 1 t

t i

t l

The following Figures have been removed from the Technical Specifications and the .

associated information has been relocated to the Core Operating Limits Report.

o j Figure 3.5.K.1-1, former page 142 i Figure 3.5.K.2, former page 142a  !

Figure 3.5.K.1-2, former page 142a-1

- Figure 3.5.K.1-3. former page 142a-2 Figure 3.5.K.2-1, former page 142a-3

. Figure 3.5.K.2-2, former page 142a Figure 3.5.K.2-3, former page 142a-5 .

L  !

4 L

t

-142-31  :

4

s: ,

Unit 2-s b. , ,-  ;

-PBAPS Is t

. i l

Figure 3.5.1.E has been removed from this page of the Technical Specifications and the associated information has been relocated to the Core Operating Limits Report.

.i

-s V

A .\

f L

E

-142d-d

i' ._

ro'

(! ...; , Unit 2~

i a:

('i s

~

Lk

e i

I The-following Figures have been removed from the Technical Specifications and the associated'information has been relocated to the Core Operating Limits Report:

1 Figure 3.5.1.H. former page 1429-Figure 3.5.1.I, former page 142h Figure 3.5.1.J,-former page 1421 L Figure 3.5.1.K, former page 142j 1 i

Figure 3.5.1.L. former page 142k 1 Figure-3.5.1.M. former page 1421 1 Figure'3.5.1 N. former pageLl42m .!

Figure 3.5.1.0, former.page'142n i l

~

1 l I

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-142g-  ;

l .o

( ,

Unit 2 PBAPS-6.9.1 RoutineReports(Cont'd) ,

l

-c. Annual Safety / Relief Valve Report Describe all challenges to the primary coolant system safety and relief valves. Challenges are defined as the automatic opening of the primary coolant safety or relief valves in response to high reactor pressure.

d. Monthly Operating Report Routine reports of operating statistics and shutdown experience and a narrative summary of the operating experience shall be submitted on a monthly basis to the Office of Management and Program Analysis (or its successor), U.S. Nuclear Regulatory Commission, Washington, DC 20555, with a copy to the appropriate. Regional Office, to be submitted no later tnan the 15th of the month following the calendar month covered by the report.
e. Core Operating Limits Report (1) Core operating limits shall be established and shall be documented in the CORE OPERATING LIMITS PEPORT prior to each Operating Cycle, or prior to any remaining portion of an Operating Cycle, for the following:
a. The APLHGR for Specification 3.5.I,
b. The MCPR for Specification 3.5.K.
c. The Kr core flow adjustment factor for Specification 3.5.K.
d. The LHGR for Specification 3.5.J.
e. The upscale flow biased Rod Block Monitor setpoint and the upscale high flow clamped Rod Block monitor setpoint of Specification 3.2.C.

(2) The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents as amended and approved:

a. HEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel"
b. Philadelphia Electric Company Methodologies as described in:

(1) PECo-FMS-0001-A, " Steady-State Thermal Hydraulic Analysis of Peach Bottom Units 2 and 3 using the FIBWR Computer Code" l'

(2) PECo-FMS-0002-A, " Method for Calculating Transient Critical Power Ratios for Boiling Water Reactors (RETRAN-TCPPECo)"

-256-

h .g L$ , ,

Unit 2 PBAPS  !

6,9.1-- Routine Reports'(Cont'd) a:'-.

(3) PECo-FMS-0003-A, " Steady-State fuel Performance Methods Report" (4) PEco-FMS-0004-A, " Methods for_ Performing BWR Systems Transient Analysis" (5) PECo-FMS-0005, " Methods for Performing BWR Steady-State Reactor Physics Analysis" (6) PECo-FMS-0006, " Methods for Performing BWR >

Reload Safety Evaluations" (3) The core operating limits shall be determined such that allapplicablelimits(e.g.,fuelthermal-mechanical limits, core thermal-hydraulic limits. ECCS limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.

(4) The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be submitted upon issuance for each Operating Cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

.yl 4

-256a-

b>

' PBAPS' Unit 3L

f. - LIST OF FIGURES pc .c ..

h KFigurei Title g 1 k

O fl.1 1 APRM Flow Bias Scram Relationship To Normal' Operating Conditions 16 b ?.1;1 1= < Instrument Test Interval Determination Curves 55

4. 2. 2 - 'Probabil:ity of System Unavailability.vs'. Test Interval 98 13;3.1- SRM Count Rate vs. Signal-to-Noise Ratio- 103a-4 3.4.1. DELETED 122  ;

i 3.4.2 DELETED' 123 3.5.K.1 DELETED 142 3.5.K.2' DELETED- 142 3.5.1.A DELETED

~

3.5.1 B DELETED 3.5.1.C. DELETED 3.5.1.D DELETED l 3.5;1.E DELETED 142 3.5.1.F DELETED 142 13.5.1.G DELETED 142 7

3.5.1.H DELETED 142-3.5.'1. I : DELETED 142 0

13.'5.1.J -DELETED '142 3.5.1.K DELETED 142

3.6.1

~ Minimum Temperature-for Pressure Tests such as required by Section XI 164

.3.6.2

. Minimum-Temperature for Mechanical Heatup or Cooldown following 164a Nuclear Shutdown F 3.6.3

Minimum Temperature-for Core Operation.(Criticality) 164b 13;6.4' . Transition Temperature Shift vs. Fluence 164c L 3 ; 6. Thermal Power Limits of Specifications 3.6.F.3. 3.6.F.4 164d 3.6.F.5.-3.6.F.6 and,3.6.F.7

-3.8.1 Site Boundary and Effluent Release Points 216e 6;2-1 Management Organization Chart 244 6.2-2 Organization for Conduct of Plant Operation 245 i

-iv-

  • = '
  • Unit 3 PBAPS i LIST OF TABLES
Table Title Page '

4.2.B Minimum Test and Calibration Frequency 81 ,

for CSCS 4.2.C Minimum Test and Calibration Frequency 83 for Control Rod Blocks Actuation 4.2.0 Minimum Test and. Calibration Frequency 84' for Radiation Monitoring Systems 4.2.E -Minimum Test and Calibration Frequency 85 for Drywell Leak Detection ,

4.2.F Minimum Test and Calibration Frequency 86 for Surveillance Instrumentation 4.2.G Minimum Test and Calibration Frequency 88 for Recirculation Pump Trip 3.5.K.2 DELETED 133d 3.5.K.3 DELETED 133d'

'4.6.1 In-service Inspection Program for Peach 150 ,

Bottom Units 2 and 3 ,

3.7.1 Primary Containment Isolation Valves 179 3.7.2 Testable Penetrations with Double 184-0-Ring Seals 3.7.3 -Testable Penetrations with Testable 184 Bellows 1

-3.7.4 Primary Containment Testable Isolation 185 Valves 4.8.1: Radioactive Liquid Waste Sampling and 216b-1 Analysis 1

4.8.2 Radioactive Gaseous Waste Sampling and 216c-1 Analysis 4.8.3.a_ Radiological Environmental Monitoring 216d-1 Program

4.8.3.b Reporting Levels for Radioactivity 216d-5 Concentrations in Environmental Samples

-vi- ,

1

Unit-3 i

PBAPS i

.- 1.0 DEFINITIONS

--The succeeding frequently used terms are explicitly defined so that a uniform

. interpretation of the specifications may be achieved.

Alteration of the Reactor Core - The act of moving any component in the region above the core support plate, below the upper grid and within the shroud with the vessel

-head removed and fuel in the vessel.

Normal control rod movement with the control drive hydraulic system is not defined as a core alteration. Normal movement of in-core instrumentation and the traversing in-core probe is not defined as a core alteration.

Average Planar Linear Heat Generation Rate (APLHGR) - The APLHGR shall be applicable to a specific planar height and is equal to the sum of the heat generation rate per unit length of fuel rod, for all the fuel rods in the specific bundle at the specific height, divided by the number of fuel rods in the fuel bundle at that height.

Channel - A channel is an arrangement of a sensor and associated components used to evaluate plant variables and produce discrete outputs used in logic. A channel terminates and loses its identity where individual channel outputs are combined in logic..

Cold Condition -. Reactor coolant temperature equal to or less than 212 F.

Cold Shutdown - The reactor is in the shutdown mode, the reactor coolant temperature equal to or less than 212 F, and tne reactor vessel is vented to atmosphere.

Core Operating Limits Report (COLR) - The COLR is the unit-specific document that provides the core operating limits for the current Operating Cycle. These cycle-specific core operating limits shall be determined for each Operating Cycle in

accordance with specification 6.9.1.e. Plant operation within these limits is addressed in individual Specifications.

? Critical-Power Ratio (CPR) - The critical power ratio is the ratio of that assembly power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor condition of interest as-calculated by application of the GEXL correlation. (Reference HE00-10958).

Dose Equivalent I-131 - That concentration of I-131 (Ci/gm) which alone would-produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

i

[ .o l l .'t PBAPS UNIT 3

- g .; . . .. . '

SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING i 2.1.A (Cont'd) ,

ir the event of operation witi. a maximum fraction of

,' limitir.; oower density (MFLPD) greater thari L;.c fraction of - -

,j

rated power (FRP), the setting _

shall be modified as follows.

I S < (0.58W + 62% - 0.58AW) where,

^

FRP = fraction of rated thermal power (3293 MWt)

MFLPD = maximum fraction of-limiting power density where the limiting power density is the value

-specified in the CORE 1 OPERATING LIMITS REPORT.

The ratio of FRP to MFLPD-shall be set. equal to 1.0 unless the actual operating value is less than the design ~value of 1.0, in which case the actual .

operating value will be used.

2. APRM--When the reactor mode
r switch is.in-the STARTUP i position, the APRM scram shall be set at less than or. o equal to'15 percent of rated power.
3. IRM--The IRM scram shall be set at less than or equal to 120/125 of full scale.

- 4:

.kf 1 t.

i  !

- '. u .' Unit 3 PBAPS-SAFETYLLIMIT' LIMITING SAFETY SYSTEM SETTING

8. Core Thermal Power Limit B. APc' Rod Block Trip Setting

~ - ~

'{ Reactor Pressure < 800 psia)

SRB < (0.58 W + 50% - 0.586W) (FRP)

MFLPD where:

FRP = fraction of rated thermalpower(3293MWt)..

MFLPD = maximum fraction of limiting power density where the limiting power density is the value specified in the CORE OPERATING LIMITS REPORT.

.The ratio of FRP to MFLP0 shall be set. equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.

3 C. Whenever the reactor 'is in the C. Scram and isolation--> 538 in. above shutdown condition with reactor low water - vessel zero irradiated fuel in the reactor level (0" on level

. vessel, the water level shall not be..less than minus 160 instruments)

= inches indicated level (378 inchesabovevesselzero).

-11a-u

._ _ -_-____--__--__-D

I 4 . ..

Unit 3 PBAPS E

2.1 BASES

FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of p the Peach Bottom Atomic Power Station Units have been analyzed L throughout the spectrum of planned operating conditions up to or above the thermal power condition required by Regulatory Guide

'l.49. The analyses were based upon plant operation in accordance with the operating map given in Figure 3.7.1 of the FSAR. In addition, 3293 MWt is the licensed maximum power level of each

. Peach Bottom Atomic Power Station Unit, and this represents the maximum steady state power which shall not knowingly be exceeded.

Conservatism is incorporated in the transient analyses in

=

estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and axial power shapes. These factors are selected conservatively with respect to their-effect on the applicable transient results as determined by the current analysis model.

Conservatism incorporated into the transient analyses is

} documented in References 4 and 5. l I

__ I

_ __ _,_________.____.m______-____-__---"a-- _ - - - --- ~ - '

g .,

o.
p. .

o ., ,

Unit 3 2.1 BASES (Cont'd)

L. References

1. Linford, R. B., " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor", NEDO 10802, February 1973.

I

2. " Qualification of the One-Dimensional Core i Transient Model for Boiling Water Reactors", NEDO l 24154 and NEDE 24154-P, Volumes I, II, and III. '
3. " Safety Evaluation for the General Electric Topical Report-Qualification of the One-Dimensional Core l

Transient Model for Boiling Water Reactors NEDO-24154 and NEDE 24154-P, Volumes I, II, and III,

4. - " General Electric Standard Application for Reactor Fuel", NEDE-24011-P-A (as amended).
5. PECo-FMS-0006, " Methods for Performing BWR Reload Safety Evaluations" (latest approved revision) i I

)

\

i l

I

'~

. . . l 3 ,. ,

Unit 3 PBAPS L

' NOTES FOR TABLE 3.1.1 (Cont'd) lc

10. The APRM downscale trip is automatically bypassed when the IRM instrumentation is operable and not high,
11. An APRM will be considered operable if there are at least 2 LPRM inputs per level and at least 14 LPRM inputs of the normal complement.

J t

12. This equation will be used in the event of operation with a maximum  ;

fractionoflimitingpowerdensity(MFLPD)greaterthanthefractionof ratedpower'(FRP),where:

FRP = fraction of rated thermal power (3293 MWt).

MFLPD = maximum fraction of limiting power density where the

' limiting power density is the value specified in the CORE OPERATING LIMITS REPORT.

The ratio of FRP to'MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.

W =' Loop Recirculation flow in percent of design. W is 100 for core flow of 102.5 million 1b/hr or greater.

Delta W = The difference between two loop and single loop effective p recirculation drive flow rate at-the same core flow. During single loop operation, the reduction in trip setting (-0.58 delta W) is accomplished by correcting the flow . input of the flow biased High Flux trip setting to preserve the original (two loop) relationship between APRM High Flux s7tpoint and recirculation drive flow or by adjusting the APRM Flux trip setting. Delta W equals zero for two loop operation.

~

Trip level setting is in percent of rated power (3293 MWt).

13. .See Section 2.1.A.1.

~ _ . _

-s; TABLE 3.2.C~ Unit T3. -

0-INSTRUMENTATION THAT.IN1?IATES CONTROL ROD BLOCKS

~

Minimum No. Instrument Trip Level Setting Number of Instrument Action of Operable Channels Provided Instrument by Design Channels Per Trip System 4 (2) APRM Upscale (Flow- 5(0.58W+5G-0.58AW) x 6 Inst. Channels (10)

Biased) FRP MFLPD i 4 APRM Upscale (Startup 5 12% 6 Inst. Channels (10)

Mode) 4 APRM Downscale 6 Inst. Channels 12.5 indicated on (10) scale 1 (2)(7)(11) Rod Block Monitor 5(0.66W+(N-66)-0.66AW)x 2 Inst. Channels (1) 4 (Flow Biased) FRP y MFLPD with a maximum of $N%

1 (7) Rod Block Monitor- 12.5 indicated on 2 Inst. Channels (1)

Downscale scale 6 8 Inst. Channels IRM Downscale (3) 12.5 indicated on (10) scale 6 IRM Detector not in (8) 8 Inst. Channels (10)

Startup Position 6 IRM Upscale $108 indicated on 8 Inst. Channels (10) scale 2 (5) SRM Detector not in .(4) 4 Inst. Channels .(1)

Startup Position 5

2 (5)(6) SRM Upscale 50 counts /sec.

1 4 Inst. Channels (1) 1 Scram Discharge $25 gallons 1 Inst. Channel (9)

Instrument Volume High Level

. o. -

, , , Unit 3 PBAPS JLOTES 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 and IRM blocks need not be operable in "Run" mode, and the APRM 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 lasts longer than seven days, the system shall be tripped. If the first column cannot be met for both trip systems, the systems shall be tripped.
2. This equation will be used in the event of operation with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP) where:

FRP = fraction of rated thermal power (3293 MWt)

MFLPD = maximum fraction of limiting power density where the limiting power density is the value specified in the CORE OPERATING LIMITS REPORT.

The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.

W = Loop Recirculation flow in percent of design. W is 100 for core flow of 102.5 million lb/hr or greater.

Trip level setting is in percent of rated power (3293 MWt).

AW is the difference between two loop and single loop effective-recirculation drive flow rate at the same core flow. During single loop g operation, the reduction in trip setting is accomplished by correcting the flow input of-'the flow biased rod block to preserve the original (two loop) relationship between the rod block setpoint and recirculation drive. flow, or by adjusting the rod block setting. 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.

l 5. One of the four SRM inputs may be bypassed.

1

6. This SRM function is bypassed when the IRM range switches are on range 8 or above.
7. The trip is bypassed when the reactor power is < 30%.
8. This function is bypassed when the mode switch is placed in Run.

l' I

  • Unit 3 J
s '

.- c PBAPS l

NOTES FOR TABLE 3.2.C (Cont.)

1

9. -If the number of operable channels is less than required by the minimum operable channels per trip function requirement, place the inoperable 1

channel in the tripped condition within one hour. -This note is applicable in the "Run" mode, the "Startup" mode and the " Refuel" mode if more than one control rod is withdrawn, i

10. For the Startup (for IRM rod block) and the Run (for APRM rod block) positions of the Reactor Mode Selector Switch and with the number of OPERABLE channels:

a.

One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 7 days or place the inoperable c~)annel in the tripped condition within the next hour,

b. Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.
11. The value of N is specified in the CORE OPERATING LIMITS REPORT.

l l

1 l

I

-74a-

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?

PBAPS- Unit 3 4-s. . >

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.'5.I Average Planar'LHGR -4,5.I Average Planar LHGR During power operation, the APLHGR- The APLGHR for each type of. fuel-for each type of fuel as a function as a function of average planar of-axial 11ocation.and average planar exposure shall beJehecked daily

. exposure shall be within-limits during reactor operation at based on applicable APLHGR limit > 25% rated thermal power.

values which have been determined -i by approved methodology for the respective fuel and. lattice types.

When hand calculations are required, the APLHGR for each type of fuel as a function of average planar i exposure shall not exceed the limit for the=most limiting lattice (excluding natural uranium)'specified in the: CORE' OPERATING LIMITS REPORT during two' recirculation loop operations. During single loop i operation, the APLHGR for each fuel type shall not exceed the above  ;

values' multiplied by the reduction-

-factors specified in the CORE OPERATING l LIMITS REPORT. If at any time during I operation it is determined by normal surveillance that the limiting value  !

of APLHGR is being exceeded, action shall be initiated within one (1) hour 3

Lto. restore ALPHGR to<within prescribed limits. If the APLHGR is not returned to within prescribed. limits within five-(5) hours, reactor- 1 power shall be decreased at a rate which'would bring the reactor to the i cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> unless'APLHGR is returned 'l

-to within limits during this 1 period. Surveillance and corresponding action shall continue until <

reactor operation is within the prescribed limits.

3.5.J. Local LHGR 4.5.J Local LHGR I

.During power operation, the-linear The LHGR as a function of core i heat generation rate (LHGR) of any height shall be checked daily  ;

rod in any fuel assembly at any during reactor operation at  ;

axial location shall not exceed > 25% rated thermal power.

design LHGR. -

LHGR _< LHGRd

.LHGRd'= Design LHGR The values for Design LHGR for each fuel type 1 are specified in the CORE OPERATING LIMITS REPORT.

- 133a -

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@c[ _

g -PBAPS Unit 3  ;

>e..,

iLIMITING CONDITIONS FOR' OPERATION' SURVEILLANCE REQUIREMENTS

13. 5.J ' Local ' LHGRL ( Cont ' d )

If-at any time during operation it=

.ic; determined by1 normal surveillance ,

that-limiting-value for'LHGR is be-ing exceeded, action shall be initi-

~ cted within.one (1) hour to restore b ,LHGR to wi th i n"prescr i bed limits.

If the LHGR:is.not returned to

- within prescribed limits within  ;

- five (5) hours, reactor power ,

'chall be decreased at a rate which  !

would bring:the reactor to the cold

chutdown-condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> >

~ unless LHGR-is returned to within n ' limits during this period. Survell- '

'lcnce-andz corresponding action shall  :

, continue until. reactor operation.is ,

' iwithin"the prescribedilimits. l 4

L3.5 K Minimum Critical Power 4.5.K Minimum Critical Power Ratio (MCPR) Ratio (MCPR) _ .

l'. During. power operation the MCPR 1. MCPR shall be. checked daily; for the-applicable incremental during reactor' power operation cycle-core average-exposure and at >25% rated thermal power._ f

.for eachLtypetof; fuel shall be 2. Except as provided in Specifi- t equal to or' greater than the value cation 3.5.K.3, the verifica . f

.givenLin Specification 3.5.K.2 or tion of the-applicability of o 3.5.~K.3 times Kf, where Kf is as 3.5.K.2.a-Operating Limit MCPR spacified in the CORE OPERATING Values shall_be performed every 1 LIMITS REPORT. If at any 120 operating days by scram time time during operation it is testing 19 or more-control rods t

'datermined by normal surveillance on a rotation basis and per-

, th2t the limiting value for MCPR forming the following:

L is being-exceeded,; action shall be

. initiated within one (1) hourEto a. The average scram time to-the 20%' insertion position ~

rsstore MCPR'to within prescribed '

-limits. If the MCPR is not returned sha)1 be:

L .to within prescribed limits Lave < LIB within five-(5) hours, reactor

_ power shall~be decreased at a b. The average scram time to -

' rate which would bring the the 20% insertion position rcactor to the cold shutdown is determined as follows: '

condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

L unless MCPR is returned to l ave ='}n[. nit ~i fwithin limits during this period. i=1 1 Surveillance and corresponding n action shall continue until re- 1[_ Ni

'cctor operation is'within the i=1 prescribed limits.

where: n = number of surveillance tests performed to date in the '

cycle.

- 133b -

+ +

+

Unit 3_ .;

-PBAPS i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS-  !

3.5 K: Minimum Critical Power 4.5.K Minimum Critical Power I

. Ratio (MCPR) (Cont'd) Ratio (MCPR) (Cont'd) 2 'Except.as specified in 3.5.K.3, Ni-= number of active control- ,

.the Operating Limit MCPR Values rods measured =in the ith shall be as specified in the surveillance test.

CORE OPERATING LIMITS REPORT for when T =i average scram time to -

the 20% insertion position

' a)" requirement 4.5.K.2.a of all rods measured in-is met, and for when the ith surveillance test. ,

, b) requirement 4.5.K.2.a is c. The adjusted anal i not met, where: scram time (T3 ) ysis mean.

is calculated i

I as follows:

I = Iave B 1/2 i 0.90 - T3 {0=p+1.65 N1 3.> If the Surveillance-Requirement I of Section 4.5.K.2 to scram ,

i=[Ni-1 i time test control rods is not performed, the Operating Where: t Limit MCPR values shall '

be as specified in the p'

=

mean of the distribution LCORE OPERATING LIMITS REPORT for average scram-. insert

for this condition. time to the 20% position = "

0.694 sec s N1 = total number of active control rods measured in

. specification 4.3.C.1 7 = standard deviation of the distribution for average scram insertion time to.

the 20% position.= 0.016 "

1

-133c-f

y ;n - - -

7<

. .. .' Unit 3 t.

PBAPS Tables 3.5.K.2 and 3.5.K.3 have been removed from former Technical Specification pages-133d and 133e, 4

respectively, and the associated information has.been relocated to the Core Operating Limits Report.

.]

I' l

-133d-

L

.,. .,- Unit 3 PBAPS '

3.5' BASES (Continued)

,H. 1 Engineered Safeguards Compartments Cooling and Ventilation 1 l

One unit cooler in-each pump compartment is capable of-providing cdequate ventilation flow and cooling. Engineering analyses indicated ,

I that the temperature rise in safeguards compartments without adequate  :

Lvantilation flow or cooling is such that continued operation of the l ocfeguards equipment or associated auxiliary equipment cannot be casured. Ventilation associated with the High Pressure Service Water 1

-Pumps is also associated with the. Emergency Service Water pumps, and is specified in Specification 3.9.

I. Average Planar LHGR I This specification assures that the peak cladding temperature

'following the postulated design basis loss-of-coolant accident will '

not exceed the limit specified in the 10 CFR Part 50, Appendix K.

'The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat

gsneration 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. The peak clad temperature is +

calculated assuming a LHGR for the highest powered rod which is equal

,to or-less than the design LHGR. This LHGR times 1.02 is used in:the- -

heat-up code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factors. The Technical Specification APLHGR is the LHGR of the highest powered rod divided by its local peaking factor. The limiting value for APLHGR is shown in the applicable figure for each fuel type in the CORE OPERATING LIMITS-REPORT.

Only-the most limiting APLHGR operating limits are shown'in the l figures for the multiple lattice fuel types.- Compliance with the lcttice-specific APLHGR limits is ensured by using the process l computer. When an alternate method to the process computer is

' required-(i.e. hand calculations and/or alternate' computer simulation), the most limiting lattice APLHGR limit for each fuel type shall be applied to every lattice of that fuel type.

The calculational procedure used to establish the APLHGR for each fuel ftype is based on a loss-of-coolant accident analysis. The' analysis iwas performed using General Electric (G.E.) calculational models which Gre consistent with the requirements of Appendix K-to 10 CFR'Part 50.

A. complete discussion of each code employed-in the analysis is presented in Reference 4. Input and model changes in the Peach Bottom loss-of-coolant analysis which are different from the previous analyses performed with Reference 4 are described in detail in RQference 8. These changes to the analysis include: (1) consideration of the counter current flow limiting (CCFL) effect, (2) corrected code inputs, and (3) the effect of drilling alternate flow paths in the bundle lower tie plate.

-140-

.s

e

_. .- . Unit 3 PBAPS 3.5 BASES (Cont'd) l .r

. J . .. . Local LHGR This-specification assures that the linear heat generation rate in any 8X8 fuel rod is less than the design linear heat generation. The maximum LHGR shall be checked-daily during reactor operation at > 25%

power to determine if fuel burnup, or control rod movement has caused i changes in power distribution. For LHGR to be'at the design LUGR below-25% rated thermal power, the peak local LHGR must be a factor of cpproximately ten (10) greater than the average LHGR which is precluded by a considerable margin when employing any permissible i control rod pattern.

~

'K. Minimum Critical' Power Ratio (MCPR)

' Operating Limit MCPR '

The. required operating limit MCPR's at steady state operating conditions are derived from the established fuel cladding. integrity Safety Limit MCPR and analyses of the ibnormal operational transients presented in Supplemental Reload Licensing Analysis and References 7 '

cnd 10. For any abnormal operating transient 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 '

Ldacrease below the Safety Limit MCPR at any time during-the transient it casuming instrument trip, setting given in Specification 2.1.

To assure that the fuel cladding integrity Safety Limit-is not violated 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 transients evaluated are as described in References.7 and 10. l 5

-140a-l l

Unit 3 PBAPS 1

L 3.5.K. BASES (Cont'd) '

The largest reduction in critical power ratio is then added to the ,

i fuel cladding integrity safety limit MCPR to establish the MCPR l Operating Limit for each fuel type.

Analysis of the abnormal operational transients is presented in  !

1 References ? and 10. Input data and operating conditions used in this I o

Cnalysis are shown in References 7 and 10 and in the Supplemental  !

Reload Licensing Analysis. '

i 3.5.L. Average Planar LHGR (APLHGR),,, Local LHGR and Minimum  ;

L Critical Power Ratio (MCPR1 l In the event that the calculated value of APLHGR, LHGR or MCPR exceeds its limiting value, a determination is made to ascertain the cause and initiate corrective action to restore the value to within prescribed  :

H limits. The status of all indicated limiting fuc1 bundles is reviewed

cs well as input data associated with the limiting values such as b

power distribution, instrumentation data (Traversing In-Core Probe -

TIP, Local Power Range Monitor - LPRM, and reactor heat balance

  • instrumentation), control rod configuration, etc., in order to 4 determine whether the calculated values are valid.  :

i In the event that the review indicates that the calculated value caceeding limits is valid, corrective action is immediately undertaken

-to restore the value to within prescribed limits. Following corrective action, which may involve alterations to the control rod ,

configuration and consequently changes to the core power distribution, rGvised instrumentation data, including changes to the relative noutron flux distribution, for up to 43 in-core locations is obtained cnd the power distribution, APLHGR, LHGR and MCPR calculated.

Corrective action is initiated within one hour of an indicated value cxceeding limits and verification that the indicated value is within prescribed limits is obtained within five hours of the initial indication.

In the event that the calculated value of APLHGR, LHGR or MCPR cxceeding its limiting value is not valid, i.e., due to an erroneous instrumentation indication, etc., corrective action is initiated within one hour of an indication value exceeding limits. Verification tnat the indicated value is within prescribed limits is obtained <

.within five hours of the initial indication. Such an invalid indication would not be a violation of the limiting condition for operation and therefore would not constitute a reportable occurrence.

140b-

Unit 3 PBAPS 3.5.L. BASES (Cont'd)

Operating experience has demonstrated that a calculated value of APLHGR, LHGR or MCPR exceeding its limiting value predominantely occurs due to this latter cause. This experience coupled with the extremely unlikely occurrence of concurrent operation exceeding APLEGR, LHGR or MCPR and a Loss-of-Coolant Accident or applicable Abnormal Operational Transients demonstrates that the times required to initiate corrective action (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) and restore the calculated value of APLHGR, LHGR or MCPR to within prescribed limits (5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) are adequate.

3.5.M. References

1. " Fuel Densification Effects on General Electric Boiling Water Reactor Fuel", Supplements 6, 7 and 8, NEDM-10735, August 1973,
2. Supplement 1 to Technical Report on Densifications of General Electric Reactor Fuels, December 14, 1974 (Regulatory Staff).
3. Communication: V. A. Moore to I. S. Mitchell, " Modified GE Model for Fuel Densification", Docket 50-321, March 27, 1974.
4. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE 20566 (Draft), August 1974.
5. General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by letter, G. L.

Gyorey tt Victor Stello, Jr., dated December 20, 1974.

6. DELETED.
7. " General Electric Standard Application for Reactor Fuel", NEDO-240ll-P-A (as amended).
8. Loss-of-Coolant Accident Analysis for Pea'ch Bottom Atomic Power Station Unit 2, NEDO-24081, December 1977, and for Unit 3,

!: NEDO-24082, December 1977.

9. Loss-of-Coolant Ace'??nt Analysis for Peach Bottom Atomic Power Station Unit 2, Suppimaent 1, NEDE-24081-P, November 1986, and for Unit 3, NEDE-24082-P, December 1987,
10. PECo-FMS-0006, " Methods for Performing BWR Reload Safety Evaluations" (latest approved revision)

-140c-

, . o Unit 3 PBAPS 4.5.K Minimum Critical Power Ratio (MCPR) - Surveillance Requirement e

At core thermal power levels less than or equal to 25%, 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 tnis point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void con ent, any inadvertent

- core flow increase would only place operation in a more conservative mode relative to MCPR. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% thermal power level with minimum 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 above 25% rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement fcr calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude)thatcouldplaceoperationatathermallimit.

4.5.L MCPR Limits for Core Flows Other than Rated The purpose of the Kr factor is to define operating limits at other than rated flow conditions. At less than 100% flow the required MCPR is the product of the operating limit MCPR and the Kr factor. Specifically, the Kr factor provides the required thermal margin to protect against a flow increase transient. The most limiting transient initiated from less than rated flow conditions is the recirculation pump speed up caused by a motor-generator speed control failure.

fer operation in the automatic flow control mode, the Kr factors assure that the cperating l'.mit MCPR will not be violated should the most limiting transient occur at less than rated flow. In the manual flow control mode, the Kr factors assure that the Safety Limit MCPR will not be violated for the same postulated transient l event.

The Kr factor curves in the CORE OPERATING LIMITS REPORT were developed l

g;nerically and are applicable to all BWR/2, BWR/3, and BWR/4 reactors. The Kr

( f actors were derived using the flow control line corresponding to rated thermal power i

at rated core flow.

fcr the manual flow control mode, the Kr factors were calculated such that at the maximum flow rate (as limited by the pump scoop tube set point) and the corresponding care power (along the rated flow control line), the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPR's were calculated at different points along the rated flow control line corresponding to different core flows. The ratio of the MCPR calculated ,

at a given point of the core flow, divided by the operating limit MCPR determines the Kr.

Fcr operation in the automatic flow control mode, the same procedure was employed .

cxcept the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at rated power and flow.

The Kr factors specified in the CORE OPERATING LIMITS REPORT are acceptable for I Peach Bottom Unit 3 operation because the operating limit MCPR is greater than the original 1.20 operating limit MCPR used for the generic derivation of Kr.

l -141a-

r - - - - - - - - - _ - - _ - - - - _ - _ _

So .e' 'o Unit 3 ,

EP-  !

L i F

PBAPS i

f I

)

i i

I i

The following Figures have been removed  !

from the Technical Specifications and the i associated information has been relocated 1 to the Core Operating Limits Report' '

Figure 3.5.K.1. former page 142  !

Figure 3.5.K.2, former page 142a  ;

Figure 3.5.1.E former page 142d .!

, figure 3.5.1.F. former page.142e i Figure 3.5.1.G. former page 142f i Figure 3.5.1.H. former page 142g- 4 Figure 3.5.1.I former page 142h !

Figure 3.5.1.J. former page 1421 Figure 3.5.1.K former page = 142j -  ;

-f l

6 I

i h

i i

r

-142- l r

e r _, m _m - , _ - - , , , , , . - , , - , , - ,

e

, . o Unit 3 PBAPS 6.9.1 _ Routine Reports (Cont'd)

c. Annual Safety / Relief Valve Report Describe all challenges to the primary coolant system safety and relief valves. Challenges are defined as the automatic opening of the primary coolant safety or relief valves in response to high reactor pressure,
d. Monthly Operatino Report Routine reports of operating statistics and shutdown experience and a narrative summary of the operating experience shall be submitted on a monthly basis to the Office of Management and Program Analysis (or its successor), U.S. Nuclear Regulatory Commission, Washington, DC 20555, with a copy to the appropriate Regional Office, to be submitted no later than the 15th of the month following the calendar month covered by the report.
e. Core Operating limits Report (1) Core operating limits shall be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each Operating Cycle, or prior to any remaining portion of an Operating Cycle, for the following:
a. The APLHGR for Specification 3.5.I,
b. The MCPR for Specification 3.5.K.
c. The Kr core flow adjustment factor for Specification 3.5.K.
d. The LHGR for Specification 3.5.J.
e. The upscale flow biased Rod Block Monitor setpoint and the ups: ale high flow clamped Rod Block monitor setpoint of Specification 3.2.C.

(2) The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents as amended and I approved:

1

a. NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel" l b. Philadelphia Electric Company Methodologies as described in:

1 (1) PECo-FMS-0001-A, ' Steady-State Thermal Hydraulic Analysis of Peach Bottom Units 2 and 3 using the FIBWR Computer Code" (2) PECo-FMS-0002-A, " Method for Calculating Transient Critical Power Ratios for Boiling Water Reactors (RETRAN-TCPPECo)"

-256-

C 4 o Unit 3'

_PBAPS <

5.9.1 Routine Reports (Cont'd)

~

(3) PECo-FMS-0003-A, " Steady-State fuel Performance Methods Report" (4) PECo-FMS-0004-A, " Methods for Performing BWR Systems Transient Analysis" (5) PECo-FMS-0005, " Methods for Performing BWR Steady-State Reactor Physics Analysis" (6) PECo-FMS-0006. " Methods for Performing BWR Reload Safety Evaluations" (3) The core operating limits shall be deterreined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.

(4) The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be submitted upon issuante for each Operating Cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.

-256a-

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