ML20043E289

Amends 154 & 155 to Licenses DPR-44 & DPR-56,respectively, Removing cycle-specific Parameter Limits from Tech Specs,Per Generic Ltr 88-16

ML20043E289
Person / Time
Site:	Peach Bottom
Issue date:	05/21/1990
From:	Butler W Office of Nuclear Reactor Regulation
To:
Shared Package
ML20043E290	List: ML20043E289 ML20043E291
References
GL-88-16, NUDOCS 9006120251
Download: ML20043E289 (56)

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NUCLEAR REGULATORY COMMISSION n

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PHit ADELPHIA ELECTRIC COMPANY PUBl1C 5ERVICE ELECTRIC AND GA5 COMPANY DELMARVA PQwER AND LIGHT CumwANY ATLANTIC CITY [L4CTRIC COMPANY DOCKET N0. 50-277 PEACH BOTTOM ATOMTC POWER STATION UNIT N0. 2

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AMENDMENT TO FACILITY OPERATING LICENSE l

Amendment No. 154 i

l License No. DPR-44 l

l-1.

The Nuclear Regulatory Comission (the Comission) has found that:

1 A.

The application for amendment by Philadelphia Electric Company, et al. (the licensee) dated March 8, 1990 as supplemented on April 26, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter 1.

B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety-of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health or safety of the public; and l

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. DPR-44 is hereby amended to read as follows:

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Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.154, are hereby incorporated in the license. PECO shall operate the facility in accordance with the Technical Specifications.

3.

This iteense amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMIS$10N

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Walter R. Butler Director Project Directorate 1-2 Division of Reactor Projects - 1/11

Attachment:

Changes to the Technical Specifications Date of issuance: May 21, 1990 f

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Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.154, are hereby incorporated in the i

license. PECO shall operate the facility in accordance with the l-Technical $pecifications.

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3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGUtATOR1 COMM15510W t

1 Walter R. Butler Director Project Directorate 1 2 Division of Reactor Projects. 1/11 l

Attachment:

Changes to the Technical Specifications Date of Issuance: May 21, 1990 i

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ATTACWENT TO LICENSE AMENDMENT N0.154 FACILITY OPERATING LICENSE NO. DPR-44 DOCKET NO. 50 277 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.

Remove Insert iv iv iva iva vi vi 1

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11a lla 40 40 73 73 74 74 74a 74a 133a 133a 133b 133b 133c 133c 133d 133d j

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, ATTACWENT TO LICFt!SE AMENDMENT N0.154 t

FACILITY OPERATING LICENSE N0. DPR-44 i

1 DOCKET NO. 50-277 1-

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142a-1 142a 2 142a-3 142a-4 142a-5 142d 142d 1429 1429 142h

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Unit 2 PBAPS LIST OF FIGURES Figure Title

,Page 1.1-1 APRM Flow Bias $ cram Relationship To Normal 16 Operating Conditions 4.1.1 Instrument Test Interval Determination curves 55 4.2.2 Probability of System Unavailability vs. Test 98 Interval 3.3.1

$RM Count Rate vs. Signal-to-Noise Ratio 103a 3.4.1 DELETED 122 3.4.2 DELETED 123 3.5.K.1 1 DELETED 142 3.5.K.1-2 DELETED 142

'3.5.K.1-3 DELETED 142 3.5.K.2 DELETED 142 3.5.K.2-1 DELETED 142 3.5.K.2-2 DELETED 142 3.5.K.2-3 DELETED 142 3.5.K.3 DELETED 14;b

3. 5.1. E DELETED 142d 3.5.1.F DELETED 142e 3.5.1.G DELETED 142f

3. 5.1. K, DELETED 142g l

l Amendment No. 36, 40, 45, 48, 70, 86, 108.jv.

122, 123, 140, 154 l-

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PBAPS LIST OF FIGURES Fioure Title h-3.5.1.1 CELETED 142g 3.5.1.J DELETED 142g

3. 5.1. 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 142g 3.6.1 MInimumTemperatureforPressureTests 164 such as required by Section XI 3.6.2 Minimum Temperature for Mechanical Heatup 164a of 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

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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 f45 o

' Amendment No. 86, 102, 123, 158. 154

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Unit 2 P8APS LIST OF TABLES Table Title h

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

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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 o

-Amendment No. 33, 48, 86. 192. 154

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PBAPS V

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

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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

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defined as a core alteration. Normal movement of in-core instrumentation and the traversing in core probe is not defined as a core alteration.

E Average Planar Linear M3 Generation Rate (APLHGR$ - The APLHGR shall be applicable to a speciff;hnar height and is equa' to the sum of the heat gen-i eration rate per unit 1.ngth of fuel rod, for all the fuel rods in the specific L-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 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 with,i_n 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 NEDO-10958).

Dose Eo'61 valent I-131 - That concentration of I-131 (Ci/ge) writch alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132,1-133,1-134, and 1-135 actually present.

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

In the event of operation with a maximum fraction of limiting powce density (MFLPD) greater than ths fraction of-rated power (FRP), the setting shall be modified as hilows.

S 1 (0.58W + 62% - 0.586W) g 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 l

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 switch i

is in the STARTUP position, the APRM scram shall be set,at less f

than or equal-to.15 percent of rated power.

3.

IkM--The IRM scram shall be set at less than or equal to 120/125 of full scale.

Amendment No. 23, 34, 36, 42, 48, 76, 78,.10 123, 125, 154

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SAFETY LIMIT-LIMITING SAFETY SYSTEM SETTING

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' Core Thermal Power Limit B.

APRM Rod Block Trip Settinc 1

(Reactor Pressure < 800 psia) g SRB 1 (0.58 W:+ 50% - 0.58AW) g where:'

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_ thermal power (3293 MWt).

MFLPD = maximum fraction of limiting power density where the limiting power density is the value-specified in:the CORE 4'

OPERATING LIMITS REPORT.

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The tatio of FRP to MFLPD shall m,

be set equal'to 1.0 unless the e

actual operating value is~1ess than the design value of 1.0, in which case the actual operating value will be used.

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Whenever the reactor.is in

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Scram and isolation--> 538 in.

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the shutdown condition with reactor low water '

-above vessel p'"

irradiated fuel in the reactor level zero (0" on.

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vessel, the water level shall level l .

not.be less than minus 160

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i inches indicated level (378 inches above vessel zero).

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NOTES FOR TABLE 3.1.1 (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.

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12. 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 1

power density is.the value specified in the CORE OPERATING i

LIMITS REPORT.

The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual l

operating value is less than the design value of 1.0, in which case che 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.

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 recir-culation drive flow or by adjusting the APRM Flux trio setting.

AW = 0 for two loop operation.

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

' 13. See Section 2.1.A.1.

.l Amendment No. 33, 41, 62, 78, 123, 154 40 P ~y, EEL *Z'Tf;y75% '"fCI(Qi9 %EE * "' '

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INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS k.

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Minimum No.

Instrument Trip Level Setting Number of Instrument Action-p@

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of Operable Channels Provided o

Instrument by Design W

.. Channels Per 3

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Trip System

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APRM Upscale (Flow Blased) 5(0.58W+50-0.586W) x-6 Inst. Chanr.els (10).

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4 APRM Upscale (Startup

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6 Inst. Channels (10).

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4 APRM Downscale 12.5 indicated on scale 6 Inst. Channels (10)-

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Rod Block Monitor

<(0.66W+(N-66)-0.66aW)x 2 Inst. Channels (1) y (Flow Biased)

FRP l

MFLPD with a maximum of 1ME n

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Rod Block Monitor

>2.5 indicated on scale 2 Inst. Channels (1)

Downscale 6

IRM Downscale {3) 22.5 indicated on scale 8 Inst. Channels (10).

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IRM Detector not in (8) 8 Inst. Channels

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Startup Position i

6 IRM Upscale 1108 indicated on scale 8 Inst. Channels (10)-

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SRM Detector not in (4) 4 Inst. Channels (1)

Startup Position 5

P-2 (5)(6)

SRM Upscale 110 counts /sec.

4 Inst. Channels (1)

Scram Discharge'

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PBAPS NOTES FOR TABLE 3.2.C-1..

For the startup and run positinns 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 l

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 j

system is functionally tested immediately and daily thereafter; if this condition ~ 1asts longer than seven days, the system shall be tripped.

If

-l the first column cannot be met for both trip systems, the systems shall be

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tripped.

2.-

The equation for Trip Level Setting will be used in the event of operation l

with a maximum fraction of limiting power density (MFLPD) greater than

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the fraction of rated power (FRP) where:

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

MFLPD = maximum fraction of limiting power density where the i

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.

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

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.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 corr. acting the 1

flow-input of the. flow biased-rod block to preserve the original- (two loop)

L relationship between the rod block setpoint and recirculation drive flow,

'y 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.

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'4.

Thjs~ function is bypassed when the count rate is > 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 i

or above.

7.

.The trip is bypassed when the reactor power is 1 30%.

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l 8.

This function is bypassed when the mode switch is placed in Run.

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Amendment No. 23, 48, 70, 78, 123.154

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,.i Unit 2-PBAPS NOTES FOR TABLE 3.2.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:

One less than required by the Minimum OPERABLE Channels per Trip a.

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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j PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.I Averaoe Planar LHGR 4.5.I Averaoe Planar LHGR During power operation, the APLHGR for The APLGHR for each type of each type of fuel as a function of axial fuel as a function of average location and average planar exposure planar exposure shall be shall be within limits based on applic-checked daily during reactor

-able APLHGR limit values which have been operation at 1 25% rated determined by approved methodology for thermal power.

the. respective fuel and lattice types.

When hand calculations are required, the APLHGR for each type of fuel as a func-tion of average planar 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 opera-tion, the APLHGR for each fuel type shall not exceed the above values multiplied by the reduction factors specified in the CORE OPERATING LIMITS REPORT.

If at any

-time during operation it is determined by i

. normal surveillance that the limiting

'value of APLHGR is being exceeded, action i

shall be initiated within one (1) hour to restore ALPHGR to within prescribed limits.

If the APLHGR is not returned 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 LHGR During power operation, the linear heat The LHGR as a function of core generatJon rate (LHGR) of any rod in any height shall be checked daily fuel assembly at any axial location shall durin 1 25%g reactor operation-at not exceed design LHGR.

rated thermal power.

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

Amendment No. 40, 48, 70, 78, 86, 108, -133a-123,154 si

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

If at any time during operation it is determined by normal surveillance that limiting value for LHGR is being exceeded, action shall be initiated within one (1) hour to restore LHGR to within prescribed limits.

If the LHGR is not returned to l

within prescribed limits within five (5) l hours, reactor power shall be decreased E

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 LHGR is returned to r

within limits during this period.

Surveillance and corresponding action

.shall continue.until reactor operation is within the prescribed limits.

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

Ratio (MCPR)

1. During power operation the MCPR for

1. MCPR shall be checked daily I

the applicable incremental cycle core during reactor power operation average exposure and for each type of at >25% rated thermal power.

fuel shall be equal to or greater than

2. Except as provided in Spec-the value given in Specification 3.5.K.2 ification 3.5.K.3, the verifi-or 3.5.K.3 times Kf, where Kf is as cation of the applicability of specified in the CORE OPERATING LIMITS 3.5.K.2.a Operating Limit MCPR REPORT.~ If at any time during operation Values shall be performed every it is determined by normal surveillance 120 operating days by scram that the limiting value for.MCPR is being time testing 19 or more control exceeded, action shall be initiated rods on a rotation basis and within one (1) hour to restore MCPR to performing the following:

within prescribed limits.

If the MCPR is not returned to within prescribed

a. The average scram time to limits within five (5) hours, reactor the 20% insertion position power shall be decreased at a rate which shall be:

would bring the reactor to the cold shut-t ave < t B down 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 within limits during

b. The average scram time to this peelod.

Surveillance and corres-the 20% insertion position ponding action shall continue until is determined as follows:

reactor operation is within the prescribed limits, tave=ENiti i=1 ENi i=1 where: n = number of surveillance tests performed' to date in the cycle.

. Amendment No. 38, 48, 86, 154

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

OPERATING LIMITS REPORT for when 11 = average scram time to a) requirement 4.5.K.2.a the 20% insertion posi-is met, and for when ion of all rods measured-in the ith surveillance b) requirement 4.5.K 2.a is test.

not met, where:

c.

The adjusted analysis t = 1 ave - t B mean scram time (t ) is B

calculatedasfol1$ws:

0.90

-t 3.

If the Surveillance Requirement of IB = p + 1.65 [ N1 }1/2 i

Section 4.5.K.2 to scram time test control rods is not performed, the ENi 1 4 i

Operating Limit MCPR values shall

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be as specified in the CORE OPERATING LIMITS REPORT for this condition.

Where:

mean of the distribution p =

for average scram inser-tion time to the 20%

position = 0.694 sec.

N1 = total number of active l

control rods measured in specification.4.3.C.1.

standard deviation of o =

the distribution for average scram insertion time to the 20% position 0.016.

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removed,from former Technical Speci-fication pages.133d and 133e,'recpec-tively, and the associated information l

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has been relocated to the Core Operating Limits Report.

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' Amendment No. 86. 108, 123.154-133d-c.,

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PBAPS 3.5 BASES (Cont'd.)

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Enaineered Safeauards Compartments coolino and Ventilation 1'

one unit cooler in each pump compartment is capable of providing adequate j

ventilation flow and cooling.

Engineering analyses indicated that the tempera-ture rise in safeguards compartments without adequate ventilation flow or 3

cooling is such that continued operation of the safeguards equipment or asso-t ciated auxiliary equipment cannot be assured.. Ventilation associated with the L

High Pressure Service Water Pumps is also associated with the Emergency Service i

Water pumps, and is specified in Specification 3.9.

I.

Averaoe Planar LHGR 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-or oolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial-location and is only dependent, secondarily, on the rod-to-rod power distribution within an assembly.

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.

• l Only the most limiting APLHGR operating limits are shown in the figures for the multiple lattice fuel types.

Compliance with-the lattice-specific APLHGR limits is ensured by using the process computer.

When an alternate method to the pro-cess computer is required (i.e. hand calculations and/or alternate computer simulation), the most limiting lattice APLHQR limit for each fuel type shall be l

i 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 General 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 enalysis 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 Reference 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.

I Amendment No. 23,'36, 40, 48, 70, 86, -140 123,154

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

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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 changes in power distribution.

For LHGR to be at the design LHGR below 25% rated thermal power, the peak local L)mR must be a factor of approximately 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)

Operatina 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 anal-yses of the abnormal operational transients presented in Supplemental Reload Licensing Analysis and Reference 7.

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 decrease below the Safety Limit MCPR at any time during the transient assuming

. instrument trip setting given in Specification 2.1.

l 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 criti-cal power ratio (CPR).

The transients evaluated are as described in Reference 7.

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The largest reduction in critical power ratio-is then added to the fuel cladding integrity safety limit MCPR to establish the MCPR Operating Limit for each fuel type.

Analysis of the abnormal operational transients is presented in Reference 7.

Input data and operating conditions used in this analysis are shown in Reference 7 and in the Supplemental Reload Licensing Analysis.

L 3.5.L Average Planar LHGR (APLHGR). Local LHGR and Minimum Critical Power I

_ Ratio (MCPR)

In the event that'the calculated value of APLHGR, LHGR or MCPR exceeds its limiting va~lue, a determination is made to ascertain the cause and initiate corrective actions to restore the value to within prescribed limits.

The status of all indicated limiting fuel bundles is reviewed as well as input data associated with the limiting values such as power distribution, instrumen-tation data (Traversing In-Core Probe - TIP, Local Power Range Monitor - LPRM, l

and reactor heat balance instrumentation), control rod configuration, etc., in order to determine whether the calculated values are valid.

In the event ~that-the review indicates that the calculated value exceeding

' 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 l

to the core power distribution, revised instrumentation data, including changes to the relative neutron flux distribution, for up to 43 in-core locations is obtained and the power distribution, APLHGR, LHGR and MCPR calculated. Correc-tive action is-initiated within one hour of an indicated value exceeding limits and verification that the indicated value is within prescribed limits is I

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 indica-tion,-etc., corrective action is initiated within one hour of an indicated value exceeding limits.

Verification that the indicated value is within pre-scribed 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 teportable occurrence.

Amendment No. 23, 36, 48, 70, 86, 123,

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Unit 2 PBAPS 4.5.K Minimum Critical Power Ratio (MCPR) - Surveillance Reouirement 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 this point MCPR value is in excess, operating plant experience indicated that the resulting of requirements by a considerable 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 2

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 suf-ficient since power distribution shifts are very slow when there have 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 thermal limit.

4.5.L MCPR Limits for Core Flows Other Than Rated The purpose of the K factor is to define operating limits at other than rated f

flow conditions. At less than ?.00% flow the required MCPR is the product of the operating limit MCPR and the K factor.

Specifically, the K factor provides f

[

the required thermal margin to protect against a flow increaIe 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 K, factors assure that the operating limit MCPR will not be violated should the most limiting transient.

occur at less than rated flow.

In the manual flow control mode, the K, factors assure that the Safety Limit MCPR will not be violated for the same po&tulated transient event.

factor curves in the CORE OPERATING LIMITS REPORT were developed I

The K genericallyandareapplicabletoallBWR/2,BWR/3,andBWR/4 reactors.

The K factorswerederivedusingtheflowcontrollinecorrespondingtoratedtherma$

power at rated core flow.

For the manual flow control mode, the K factors were calculated such that at the maximum flow rate (as limited by the, pump scoop tube set point) and the cor-respondng core power (along the rated flow control line), the limiting bundle's relative power was adjusted until the MCPR was sligntly 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 j_

operating limit MCPR determines the K.

f

-For operation in the automatic flow control mode, the same procedure was employed i

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

1 i

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4.5.L~ MCPR Limits' for Core Flows' 0ther Than Rated (Cont'd.)

The K factors specified in the. CORE OPERATING LIMITS REPORT are acceptable l

forP$achBottomoperationbecausetheoperatinglimitMCPRisgreaterthanthe l:

original 1.20 operating slimit MCPR used for the generic derivation of K.

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.from the Technical Specifications and the associated information has been relocated to the Core Operating Limits Report:

'l Figure 3.5.K.1-1, former page 142 1

1

-Figure 3.5.K.2, former page 142a Figure-3.5.K.1-2, former page 142a-1 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-4 Figure 3.5.K.2-3, former page 142a-5 o

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Ls Figure 3.5.1.E has been removed from this page of the Technical Specifications and the associated information has been L

relocated to the Core Operating Limits Report.

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The following Figures have been removed from the Technical Specificar. ions and the associated information has been relocated to the Core Operating Limits Report:

Figure 3.5.1.H. former page 142g 1."

1 Figure 3.5.1.I. former page 142h~

Figure 3.5.1.J former page 1421 Figure 3.5.1.K, former page 142j Figure 3.5.1.L. former page 142k Figure 3.5.1.H former page 1421 Figure 3.5.1.N, former page 142m --

Figure 3.5.1'.0,-former page 142n

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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 primar coolant safety or relief valves in response to high reactor pressure. y d.

Monthly Operatina Reoort 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 spbmitted no later than the 15th of the month following the calendar month i

covered by theireport, e.

Core Operatino Limits Report I

(1) Core operating limits shall be established and shall be documented L

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 3.5.1, b.

The MCPR for Specification' 3.5 K.

c.

The K core flow adjustment factor for Specification 3.5 K, g

d.

The LHGR for Specification 3.5.J.

The upscale flow biased Rod Block Monitor setpoint;and the e.

upscale high flow clamped Rod Block monitor setpoiiht 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, spe-cifically those described in the following documents as amended and approved:-

o NEDE-24011-P-A, " General Electric Standard Application for a.

Reactor Fuel" (latest approved version) 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"

/

? Amendment No.10it,110.154

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6.9.1' Routine Reports'(Cont'd) 3 (2) PEco-FMS-0002-A, " Method for Calculating Transient l'

Critical Power Ratios for Boiling Water Reactors (RETRAN-TCPPEco)"

?

(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-A, " Methods for Performing BWR Steady-State Reactor Physics Analysis" (3) The core operating limits shall be determined such that all L

applicable limits (e.g., fuel thermal-mechanical limits, core L

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 1

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.

3 L

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.i Amendment No. 154'

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e-UNITED STATE 8 NUCLEAR REGULATORY COMMISSION

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L PHitADELPHIA' ELECTRIC COMPANY PUBl1C 5ERVICE ELECTRIC AND SAS COMPANY l

DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-278-I PEACH BOTTOM ATOMIC POWER STATION. UNIT N0. 3' AMENDMENT TO FACILITY OPERATING LICENSE i

Amendment No.155 License No. DPR-56 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Philadelphia Electric Company, et al. (the licensee) dated March 8,1990 as supplemented on April 26, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter 1.

B.

The facility will operate in conformity with the application, the L

provisions of the Act, and the rules.and regulations of the l-Comission;-

L C.

There.is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted

.in compliance with the Comission's regulations; D.

The' issuance of this amendment will not be inimical to the comon defense and security or to the health' or safety of the putrlic;-and E.

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

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

satisfied.

2.

Accordingly, the license is amended by changes to.the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. DPR-56 is hereby amended to read as follows:

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Technical Specifications The Technical Specifications contained in Appendices A and B, as:

revised through Amendment No.155, are hereby incorporated in the license.

pEC0 shall operate the facility in accordance'with the i

Technical Specifications.

. 3.

This license amendment is effective as of its date of issuance.

I FOR THE NUCLEAR REGULATORY COMMISS10N' i

/s/

Walter R. Butler, Director-Project Directorate 1-2 Division of Reactor Projects - 1/11

Attachment:

Changes to the Technical Specifications

. Date of issuance: May 21, 1990 i

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Technical Specifications The Technical Specifications contained in Appendices A and B as

.i revised through Amendment No.155, are hereby incorporated in the license.- PEC0 shall operate the facility in accordance with the Technical _.~ Specifications.

3.

This license amendment is effective as of its date of issuance.

l l-FOR THE NUCLEAR' REGULATORY COMMISSION

~!

Walter R. Butler, Director Project Directorate 1-2

]

Division of Reactor Projects - 1/11

Attachment:

. Changes to the Technical Specifications Date of Issuance: May 21, 1990

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. 1 ATTACHMENT TO LICENSE AMENDMENT NO.155 FAcit1TY OPERATING LICENSE NO. DPR-56 1

DOCKET NO. 50-278

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Replace the fellowing pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.

Remove Insert iv iv t.-

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74 74 l

1 74a 74a 133a 133a 133b.

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- ATTACHMENT TO LICENSE AMENDMENT NO.155

. FACILTTY-OPERATING LICENSE N0. DPR.56

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LIST DF FIGURES Fiaure Tit 1e Page a

L cl.1-1 APRM Flow Bias Scram Relationship to Normal Operating 16 1

Conditions ~

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- 4.1.1 Instrument Test Interval Determination Curves 55 4.2.2

.PrGability of System Unavailability vs. Test Interval 98 f

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3.3.1 SRM Count Rate vs. Signal-to-Noise Ratio 103a

-3.4.1 DELETED 122 3.4.2 DELETED 123 3.5.K 1 DELETED 142 3.5.K.2 DELETED 142 3.5.2.A DELETED 3.5.1.B DELETED 3.5.1.C DELETED-23.5.1.D DELETED-

-3.5.1.E DELETED-142-3.5.1.F DELETED 142 s

-3.5.1.G DELETED 142 3.5.1.H DELETED

<142 a

3.5.1.1 DELETED L

142 3.5.1.J DELETED-g 142 L

3.5.1 K DELETED 142 3.6.1-

• Minimum Temperature for Pressure Tests'such as required 164 by Section XI 3.6.2 Minimum Temperature for Mechanical Heatup or Cooldown 164a

-following Nuclear Shutdown L

3.6.3 Minimum Temperature for Core Operation (Criticality) 164b 3.6.4 Transition Temperature Shift vs. Fluence 164c Amendment No. 14, 41, 45, 46, 62, 79,-

92, 104, 607,'114, 126,~iV' l

142, 150, 155

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?e LIST OF FIGURES w

Fjgre Title

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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 i

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3.8.1 Site Boundary and Effluent Release Points 216e 1

s 6.2-1 Management Organization Chart

- 244 y

6.2-2 Organization for Conduct of Plant Operation 245 l

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PBAPS LIST OF TABLES Table Title flat 4.2.B Minimum Test and Calibration Frequency for CSCS 81

4. 2. C Minimum Test and Calibration Frequency for Control Rod 83 Blocks Actuation 4.2.0 Minimum Test and Calibration Frequency for Radiation 84 Monitoring Systems 4.2.E Minimum Test and Calibration Frequency for Drywell 85 Leak Detection 4.2.F Minimum Test and Calibration Frequency for Surveillance 86 Instrumentation 4.2.G Minimum Test and Calibration Frequency for Recirculation 88 Pump Trip

3. 5. K. 2 DELETED 133d 3.5.K.3 DELETED 133d 4.6.1 In-service Inspection Program for Peach Botton Units 2 150 and 3 3.7.1 Primary Containment Isolation Valves 179 3.7.2 Testable Penetrations with Double 0-Ring Seals 184 3.7.3 Testable Penetrations with Testable Bellows 184 3.7.4 Primary Containment Testable Isolation Valves 185 4.8.1 R.dioactive Liquid Waste Sampling and Analysis

216b 4. 8. 2 Radioactive Gaseous Waste Sampling and Analysis 216c-1 4.8.3.a Radiological Environmental Monitoring Program 216d-1 4.8.3.b Reporting Levels for Radioactivity Concentrations in 216d-5 Environmental Samples e

i Amendmen No. 41, 42. 79.184.155

-vi-4

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Unit 3 l.

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

Alk. oration 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.

Avernoe Planar Linear Heat Generation Rate (APLHGR) - The APLHGR shall be impplicable to a specific planar height and is equal to the sum of the heat gen-eration rate per unit length of fuel rod, for all the fuel rods in the specific bundle at the specific height, divided by the nLabor 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 the reactor vessel is vented to atmosphere.

Core Operatino Limits Report (COLR) - The COLR is the unit-specific document that provices 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 NED0-10958).

Dose Eoeivalent 1-131 - That concentration of I-131 (Ci/ge) which alone would produce'the same thyroid dose as the quantity and isotopic mixture of 1-131 I-132,1-133,1-134, and 1-135 actually present.

Amendment No. 164, 125, 150, 155

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$AFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1. A (Cont'd)

In the event of operation with a maximum fraction of-limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows.

$ 1 (0.58W + 625 - 0.5 taw) g

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 n

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 switch is in the STARTUP position, the APRM scram shall be set at less than or 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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faendment No. 14. 33, 41, 62, 77. 79. 307, 169,155 K~

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SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 8.

Core Thermal Power Limit 8.

APRM Rod Block Trio Settino (Reactor Pressure 1500 psia) t SRB (0.58 W + 5 3 - 0.5 hW) FRP where:

I FRP = fraction of rated 1

thermal power (3293 MJt).

MFLPD = maximum fraction of limiting power density where the limiting power i

density is the value i

specified in the CORE OPERATING LIMITS REPORT.

The ratio of FRP to MFLPD shall f

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.

C.

Whenever the reactor is in C.

Scram and isolation--> 538 in.

the shutdown condition with reactor low water above vessel

~

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

inches indicated level (378 inches above vessel zero).

I 1

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1 Amendment No. 77, 79, 115, 150, 155

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P8APS NOTES FOR TABLE 3.1.1 (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.

j

12. This equation will be used in the event of operation with a maximum fraction 1

l of limiting power density (MFLPD) greater than the fraction of rated power (FRP),where:

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

1 MFLPD = maximum fraction of limiting l

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

)

i LIMITS REPORT.

l l

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 1

actual operating value will be used.

W=

Loop Recirculation flow in percent of design. W is 100 for core i

flow of 102.h million 1b/hr or greater.

1 Delta W = The difference between two loop and single loop effective recirculation drive flow rate at the same core flow.

During sin W) gle loop operation, the reduction in trip setting (-0.58 delta.

is accomplished by correcting the flow input of the flow biased' High Flux trip setting to preserve the original (two loop) rela-tionship between APRM High Flux setpoint and recirculation drive flow or by adjusting the APRM Flux trip setting.

Delta W equals zero for two 1 cop operation.

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

13. See Section 2.1.A.1.

i Amendment No. 33, 41, 62, 77, 79, 106 132, 150, 155

,.73,, ;;q m;3x, e

e PSAPS Unit 3 TABLE 3.2.C g

INSTRUMENTATION THAT IIITIATES CONTROL N00 BLOCKS E

Minimum No.

Instrument Trip Level Setting Number of Instrument Action-

}'

of Operable Channels Provided Instrument by Design g - Channe1s Per Trip System

'3 EU 4 (2)

APM Upscale (Flow Blased)

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

Eg FRP MFLPD 3:

t w

l0 4

APM Upscale (Startup 3

M)

-<12%

6 Inst. Channels (10)

j g

4 APf51 Downscale 12.5 indicated on scale 6 Inst. Channels (10)

'I.

1 (2)(7)(11)

Rod Block Monitor

<(0.66W+(N-66)-0.66aW)x 2 Inst. Channels (1) i;I;

,E (Flow Biased)

FRP MFLPD with a maximum of <NE

,d D

U 1 (7)

Rod Block Monitor lE Downscale

->2.5 indicated on scale 2 Inst. Channels (1) 6 IRM Downscale (3) 12.5 indicated on scale 8 Inst. Channels (10)

((

6 IRM Detector not in (8) 8 Inst. Channels (10) n Startup Position 6

IlWI Upscale

$108 indicated on scale 8 Inst. Channels (10)

5 2 (5)

SAM Detector not in (4) 4 Inst. Channels (1) l if Startup Position t

'^

5 2 (5)(6)

SRM Upscale 110 counts /sec.

4 Inst. Channels (1) o l

1 Scram Discharge

<25 gallons 1 Inst. Channel (9)

Instrument Vo1ume

~

High Level I

l

.i

Unit 3 P8APS l

NOTES 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

't 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 i

exist for up to seven days provided that during that time the operable sys-l tem is functionally tested immediately and daily thereafter; if this condi-tion lasts longer than seven days, the system shall be tripped.

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

2.

The equation for Trip Level Setting will be used in the event of operation l

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 riensity 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 i

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.

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 setp or by adjusting the rod block setting. gint and recirculation drive flow, 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 3,100 cps.

e 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 5,30%.

8.

This function is bypassed when the mode switch is placed in Run.

AmendmentIo. 33, 41, 62, 77, 79 158,155 L

• (,.

Jf' y,7*, Z { ['T U fI E

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~ ' '

Unit 3 PBAPS i

NOTES FOR TABLE 3.2.C (Cont.)

i 9.

If the number of operable channels is less than required by the minimum i

operable channels per trip function requirement, place the inoperable channel in the tripped condition within ene hour.

This note is applicable in the "Run" mode, the "startup" mode and the " Refuel" mode if more than one control rod is withdrawn.

l L

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 t

l-OPERABLE channels:

One less than required by the Minimum OPERABLE Channels per Trip a.

Function requirement, restore the inoperable channel to OPERABLE l

status within 7 days or place the inoperable channel in the tripped l

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.

e i

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Amendment No. 88, 93, 155

-74a-l 2 i~,;c,:6 % T..' V i G C ~~;'lE f.Z W I W E.

~

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p Unit 3 o

PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.1 Average Planar LHGR 4.5.1 Averaoe Planar LHGR During power operation, the APLHGR for The APLGHR for each type of each type of fuel as a function of axial fuel as a function of average location and average planar exposure planar exposure shall be shall be within limits based on applic-checked daily durin able APLHGR limit values which have been operation at > 25% g reactor rated determined by approved methodology for thermal power 7 the respectiva fuel and lattice types.

When hand calculations are required, the APLHGR for each type of fuel as a func-tion of average planar exposure shall not exceed the limit for the most limiting lattice (excitding natural uranium) specified in the CORE OPERATING LIMITS REPORT durin operations. g two recirculation loop During single loop opera-tion, the APLHGR for each fuel type shall not exceed the above values multiplied by the reduction factors specified in the CORE OPERATING LIMITS REPORT.

If at any time during operation it is determined by i

normal surveillance that the limiting value of APLHGR is being exceeded, action shall be initiated within one (1) hour to restore ALPHGR to within prescribed limits.

If the APLHGR is not returned to within prescribed limits within five (5) d 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 actipn shall continue until reactor operation is within the prescribed liikits.

3.5.J Local LHGR 4.5.J Local LHGR

~

During power operation, the linear heat The LHGR as a function of core generation rate (LHGR) of any rod in any height shall be checked daily fuel assembly at any axial location shall during reactor operation at not exceed design LHGR.

?,25% rated thermal power.

LHGR3,LHGRd LHGRd = Design LHGR The values for Design LHGR for each fuel type are specified in the CORE OPERATING LIMITS REPORT.

l i

Amendment No. 33, 41, 62, 77, '79, 92 -133a-150, 155 i-dij5Ef,y o.7,OMYf72i Y'.

~

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p Unit 3 PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.J Local LHGR (Cont'd)

If at any time during operation it is determined by normal surveillance that limiting value for LHGR is t*eing exceeded, action shall be initiated within one (1) hour to restore LHGR to within prescribed limits.

If the LHGR is not returned to 1

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 LHGR is returned to within limits during this period.

Surveillance and corresponding action l

i shall continue until reactor operation is within the prescribed limits.

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

Ratio (MCPR) l

1. During power operation the MCPR for

1. MCPR shall be checked daily the applicable incremental cycle core during reactor power operation i

I average exposure and for each type of at >25% rated thermal power, fuel shall be equal to or greater than

2. Except as provided in Spec-the value given in Specification 3.5.K.2 ification 3.5.K.3, the verif1-or 3.5.K.3 times Kf, where Kf is as cation of the applicability of specified in the CORE OPERATING LIMITS 3.5.K.2.a Operating Limit MCPR -

REPORT.

If at any time during operation Values shall be performed every i

it is determined ey normal surveillance 120 operating days by scram I

that the limiting value for MCPR is being time testing 19 or more control exceeded, action shall be initiated rods on a rotation basis and l

within one (1) hour to restore MCPR to performing the following:

within prescribed limits.

If the MCPR l

is not returned to within prescribed

a. The average scram time to i

limits within five (5) hours, reactor the 20% insertion position l

power shall be decreased at a rate which shall be:

would bring the reactor to the cold shut-t ave < t B down 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 within limits during

b. The average scram time to this petiod.

Surveillance and corres-the 20% insertion position l

ponding action shall continue until is determined as follows:

reactor operation is within the prescribed limits.

Iave=INiti i=1 "I Ni i=1 where: n = r umber of surveillance tests performed to date in the cycle.

Amendment No. 150, 155

-133b-i l

7,7?T.W pum m "p*.

,_ j

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Unit 3

. p 1

.i 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) l, 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 CORE surveillance test.

OPERATING LIMITS REPORT for when ti = average scram time to a) requirement 4.5.K.2.a the 20% insertion posi-is met, and for when ion of all rods measured h

in the ith surveillance L

b) requirement 4.5.K.2.a is test.

not met, where:

c.

Theadjustedanalysis tB mean scram time (t ) is t = t ave calculated as follbs:

0.90 - T B tB=p+1.65[nN1\\1/2 3.

If the Surveillance Requirement of Section 4.5.K.2 to scram time test control rods is not performed, the INi operating Limit MCPR values shall k i=1

/

be as specified in the CORE OPERATING LIMITS REPORT for this condition.

Where:

mean of the distribution p =

for average scram insert time to the 20% position

= 0.694 sec.

N1 = total number of active control rods measured in specification 4.3.C.1.

standard deviattion of o =

i the distribution for average scram insertion time to the 20% position l

= 0.016 o

L l

l Amendment No. 79, 150,155

-133c-

• .i...

J a j i :'

, : :: :i:~ Lh _.

L-

_y 3 f

O Unit 3 i

PBAPS l

i l

i Tables 3.5.K.2 and 3.5.K.3 have been removed from former Technical Speci-fication pages 133d and 133e, respec-tively, and the associated information j

has been relocated to the Core Operating Limits Report.

l l

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5 i.

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- Amendment No. 42, 62, 77, 79, 85, 92. -133d-107, 114, 150, 155

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Unit 3 PBAPS 3.5 BASES (Continued)

H.

Encineered Safeauards Compartments Coolino and Ventilation l

One unit cooler in each pump compartment is capable of providing adequate ventilation flow and cooling.

Engineering analyses indicated that the tempera-ture rise in safeguards compartments without adequate ventilation flow or cooling is such that continued operation of the safeguards equipment or asso-ciated auxiliary equipment cannot be assured. Ventilation associated with the

=

High Pressure Service Water Pumps is also associated with the Emergency Service L

. Water pumps, and is specified in Specification 3.9.

I.

Averece Planar LHGR This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit g

specified in the 10 CFR Part 50, Appendix K.

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is prime.rily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent, second-arily, on the rod-to-rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the hightst 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 E

for APLH3R 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 figures for the l

multiple lattice fuel types.

Compliance with the lattice-specific APLHGR limits

]

is ensured by using the process computer. When an alternate method to the pro-cess 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 type is based on a loss-of-coolant accident analysis.

The analysis was performed using General 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 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 Reference 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.

Amendment No. 33, 41, 42, 62, 79, 150, -140-155 M

gg

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O

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' ^

Unit 3 PBAPS 3.5 BASES (Cont'd)

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 changes in power distribution.

For LHGR to be at the design LHGR below 25% rated thermal power, the peak local LHGR must be a factor of approximately ten (10) greater than the average LHGR which is precluded by a considerable margin when employing any permissible control rod pattern, i

K.

Minimum Critical Power Ratio (MCPR)

Operatino Limit MCPR 1

(

The required operating limit MCPR's at steady state operating conditions are derived from the established fuel cladding integrity Safety Limit MCPR and anal-i l

yses of the abnormal operational transients presented in Supplemental Reload Licensing Analysis and Reference 7.

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 decrease below the Safety Limit MCPR at any time during the transient assuming 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 i

l have been analyzed to determine which result in the largest reduction in criti- '.

cal power ratio (CPR).

The transients evaluated are as described in Reference 7.

l l

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Atendment No. 33, 41, 42, 62, 79

-140a-155 i

s. n o '.l

' ~

~

'3

• 3 s

^l'

~.

Unit 3 g.

i j

PBAPS 4.5.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 small.

For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting l

MCPR value is in excess of requirements by a considerable 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 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 suf-

{

ficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

The requirement for calculating MCPR when a limiting control rod pH torn is approached ensures that MCPR will be L

known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.

4.5.L MCPR Limits for Core Flows Other Than Rated The purpose of the K factor is to define operating limits at other than rated g

flow conditions.

At less than 100% flow the required MCPR is the product of the operating limit MCPR and the K factor.

Specifically, the K, factor provides g

the required thermal margin to protect against a flow increate 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 K factors assure that the operating limit MCPR will not be violated should the, most limiting transient,

occur at less than rated flow.

In the manual flow control mode, the K factors assure that the Safety Limit MCPR will not be violated for the same po,tulated s

transient event.

The K factor curves in the CORE OPERATING LIMITS REPORT were developed genericallyandareapplicabletoallBWR/2,BWR/3,andBWR/4reactof.s..TheK l

factorswerederivedusingtheflowcontrollinecorrespondingtoratedtherma$

power at rated core flow.

For the manual flow control mode, the K factors were calculated such that at the maximum flow rate (as limited by the, pump scoop tube set point) and the cor-responding core 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 t!!is 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 l

operating limit MCPR determines the K.

g 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 power and flow.

6 Amendment No. 18. 33. 155

-141a-6 i

,A s

1 6 'of a g 4 %a A

.j E

8

e Unit 3 PBAPS 4.5.L MCPR Limits for Core Flows Other Than Rated (ContM i

The K factors specified in the CORE OPERATING LIMITS REPORT are acceptable for g

Peach Botton Unit 3 operation because the operating limit MCPR is greater than the original 1.20 operating limit MCPR used for the generic derivation of K.

y r

v P

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e Amendment No. 18. 41.155

-141b-a,3:.4 j.': [.. - fv. 9.~., s ;u.9. 4; ur..y.-,. _

_ ;af *. f

. p L r,,

Unit 3 l

t 7

PBAPS l

N l

1 1

i The following Figures have been removed from the Technical Specifications and the associated information has been relocated 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 j

Figure 3.$ 1.F, former page 142e Figure 3.5.1.G, former page 142f i

Figure 3.5.1.H former page 1429 l

Figure,3.5.1.1, former page 142h l

Figure 3.5.1.J. former page 1421 Figure 3.5.1.K, former page 142j 1

I o

Amendment No. 41, 79, 85, 92, 114, 156.-142-155 4

.if*' e#W p a W I.-

e D' N h le '. $ 94 8

hAf * + #

p-

=A

+e p

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Unit 3 l

(

PBAPS 6.9.1 Routine Reoorts (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 Operatina Reoort 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.

I e.

Core Operatino 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 3.5.1, b.

The MCPR for Specification 3.5.K.

c.

The X core flow adjustment factor for Specification 3.5.K f

d.

The LHGR for Specification 3.5.J.

The upscale flow biased Rod Block Monitor setpoint and the e.

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, spe-cifically those described in the fellowing documents as amended and approved:

NEDE-24011-P-A, " General Electric Standard Application for e

a.

Reactor Fuel" (latest approved version) 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" Amendment No. 104. !!S. 155

-256-

.. Q.. y L-

+

g...

Unit 3 PSAPS

}

6.9.1 Routine Reports (Cont'd)

(2) PEco-FMS-0002-A, " Method for Calculating Transient Critical Power Ratios for Boiling Water Reactors i

(RETRAN-TCPPEco)"

(3) PEco-FMS-0003-A, "Steaty-State Fuel Performance Methods i

Report" (4) PEco-FMS-0004-A, " Methods for Performing BWR Systems Transient. Analysis" t

(5) PEco-FMS 0005-A, " Methods for Performing BWR Steady-State Reactor Physics Analysis" i

(3) The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core i

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.

l 1

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e i

M Amendment No.155

-256a-i

.a w.

4 f

s

.,,. =

A i

.+<

- 1/ -

_