Text

Amend 33 to License NPF-58,revising Tech Specs to Remove cycle-specific Parameters
	ML20064A825
Person / Time
Site:	Perry
Issue date:	09/13/1990
From:	Hannon J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20064A826	List: ML20064A825; ML20064A827;
References
	NUDOCS 9009280140
	Download: ML20064A825 (14)
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o UNITED STATES r

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q NUCLEAR REGULATORY COMMISSION el WASHINGTON, D C. 20655

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-1 THE CLEVELAND ELECTRIC ILLUMINATING COMPANY ET AL.

DOCKET NO. 50-440-PERRY NUCLEAR POWER PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE

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Amendment No. 33 License ~No. NPF-58

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1.

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

A.

The application for amendment by The-Cleveland Electric Illuminating -

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. Company,' Duquesne Light Company, Ohio Edison Company, Pennsylvania Power Company, and Toledo Edison Company (the licensees) dated December 19, 1989 as supplemented March 30, 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 j

'forth in 10 CFR Chapter I; B.

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

~C.

There is-reasonable assurance (i) that'the activities authorized by

-this amendment can be conducted without endangering the health and l

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

U-D.

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

'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..

L 2.:lAccordingly, the license is amended by changes to the' Technical Specifications L

as, indicated in the attachment to this license amendment, and-paragraph 2.C.(2) of Facility Operating License No. NPF-58 is hereby amended to read "as follows:

9009280140 900913 PDR ADOCK 05000440

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-(2): Technical Specifications The Technical Specifications: contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 33 are hereby incorporated into this license.t The Cleveland Electric Illuminating Company shall operate the facil-ity in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of^ issuance and is to be

-implemented within 10 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION w-John N. Hannon, Director Project Directorate III-3 DivisionofReactorProjects-III; IV, Y and Special Projects Office of Nuclear Reactor Regulation

Attachment:

, Changes to the Technical Specifications iDate of issuance: September 13, 1990 m.;

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l ATTACHMENT TO LICENSE AMENDMENT NO. 33 r

FACILITY OPERATING LICENSE NO. NPF-58 i

DOCKET NO. 50-440 Replace the following pages of the Appendis "A": Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages are provided to maintain document completeness.

Remove Insert i

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v vi vi xxv1 xxvi.

l 1-2 1-2 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4:2-4 3/4 2-5 3/4.2-6 3/4 2-6a 3/4 2-7 L

3/4 2-8 l

3/4 2-9 L

3/4 2-10 l 4 5-4 6-21.

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DEFINITIONS SECTIDW

- 1. 0 DEFINITIONS PAGE 1.1 ACTI0N.......................................................

1-1 1.2c AVERAGE PLANAR EXP0$0RE......................................

1-1 1.3-AVERAGE PLANAR LINEAR HEAT GENERATION RATE...................

1-1

-- 1. 4 C HANN E L CALIB RATION..........................................

1-1

1. 5 CHANNEL CHECK................................................

1-1 1.6 CHANNEL FUNCTIONAL TEST......................................

1-1

1. 7 - CO R E A LT E RATI ON..............................................

1-2

1. 8 CORE OPE RATING L IMITS REP 0RT............ ~.....................

1-2 1.9 CRITICAL. POWER RATI0.........................................

1-2

~l'.10.00SE EQUIVALENT I-131........................................

1-2 1.11 DRYWELL INTEGRITY.....,......................................

1 1.12 E-AVERAGE DISINTEGRATION ENERGY..............................

1-3 1.13 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME............

1-3

1.14 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME....

1-3 1.15 FRACTION OF LIMITING POWER DENSITY...........................

1-3 1.16 FRACTION OF RATED THELHal P0WE.R,.............................

1-3 1.17 FREQUENCY N0TATION...........................................

1-4 1.18 FUEL HANDLING BUILDING INTEGRITY.............................

1-4

-1.19.GASE0VS RADWASTE TREATMENT (0FFGAS) SYSTEM...................

1-4 1.20 IDENTIFIED LEAKAGE...........................................- 4

1. 21 ISO LATION ' SYSTEM RESPONSE TIME...............................

1-4 1.22 LIMITING. CONTROL ROD PATTERN.................................

1-5 l

1.23 LINEAR HEAT GENERATION RATE..................................

1-5 l

PERRY - UNIT l' i

Amendment No. 33 c.

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DEFINITIONS SECTION DEFINITIONS-(Continued)

PAG [

1.24 LIQUID RADWASTE TREATMENT SYSTEM....................

..c.....

1-5 1.25 LOGIC SYSTEM FUNCTIONAL TEST..........................c....

1-5 1.26 MEMBER (S) 0F THE PUBLIC.........................

1-5 a

3 1.27 MINIMUM CRITICAL POWER RATI0......................-r...r,....-

1-5

- 1.28 0FFSITE DOSE CALCULATION MANUAL..............................

1-5 1.29 OPERABLE - OPERABILITY.......................................

.1-6

1. 30 OPERATIONAL 1 CONDITION - C0NDITION............................

1-6 1.31 PHYSICS ~TESTS................................................

1-6 1.32. PRESSURE BOUNDARY LEAKAGE....................................

1 1.13. PRIMARY _ CONTAINMENT INTEGRITY................................

1-6 i

3 1.34 PROCESS CONTROL PR0 GRAM......................................

1-7

1. 3 5 ' PURG E - PURG I NG..............................................

1-7 1.36-RATED THERMAL P0WER..........................................

1-7

' 37 REACTOR PROTECTION SYSTEM RESPONSETIME......................

1-7 i

- 1.38 REPORTABLE EVENT.............................................

1 '

1.39 ROD DENSITY..................................................

1-7 1.40 SECONDARY CONTAINMENT.INTEGRIT.Y.........

1-7

' 1.41 SHUTDOWN MARGIN..............................................

1-8 1.42 SITE B00NDARY................................................

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-1.43 SOLIDIFICATION...............................................

1-8 1,44 SOURCE CHECK.................................................

1-8 l'

1.45 STAGGERED TEST BASIS.........................................

1-8 1.46 THERMAL P0WER................................................

1-9

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PERRY - UNIT 1 11 L

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- LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

- SECTION PAGE

- 3/4.0- APPLICABILITY............................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN........................................

3/4 1-1 3/4.1.2 RE ACT I V IT Y AN0 MAL I ES...................................

3/4 1 - 2 3/4.1.3 CONTROL RODS Control; Rod Operability................................

3/4 1-3 Control Rod Maximum Scram Insertion Times..............

3/4 1 - Control Rod Scram Accumulators.........................

3/4 1-8 Control Rod Drive Coupling.............................

3/4 1-10

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= Control Rod Position Indication........................

3/4 1-12 Control' Rod Drive Housing Support.....'.................

3/4 1-14 l 3/4.1.4 CONTROL ROD PROGRAM CONTROLS-Control Rod Withdrawa1.................................

3/4'l-15 Rod Pattern Control System.............................

3/4 1-16 3/4.1.5-STANDBY LIQUID CONTROL SYSTEM..........................

3/4 1-18

' Figure 3.1.5-1 : Sodium Pentaborate Solution

. Concentration / Volume Require-ments.............................

3/4 1-20 i

'3/4.2: POWER DISTRIBUTION ~ LIMITS e

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATICN RATE..............

3/4 2-1 T

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PERRY

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AmendmentNo.ff,33

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LIMITING CONDIT10NS FOR OPERATION AND SURVEILLANCE REQUIREMENTS j

1 SECTION P_Ag A

i POWER DISTRIBUTION LIMITS (Continued).

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3/4.2.2 MINIMUM CRITICAL POWER RAT 10...........................

3/4 2-2 i

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i 3/4.2.3 LINEAR HEAT GENERATION RATE............................

3/4 2-3 i

J 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTAT10!i..............3/4 3-1 l

Table 3.3.1-1 Reactor' Protection SyMem Ins trumenta tion...................

3/4 3-2 Table 3.3.1-2 Reactor F.'otection. System Response Times....................

3/4 3-6 Table 4.3.1.1-1 Reactor Protection System Instrumentation Surveillance Requirements......................

3/4 3-7 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION....................

3/4 3-9 Table 3.3.2-1 Isolation Actuation Instrumentation...................

3/4 3,

Table 3.3.2-2 Isolation Actuation.

Instrumentation Setpoints.........

3/4 3-17 Table 3.3.2-3 Isolation System Instrumentation Response T1me.....................

3/4 3-21 Table 4.3.2.1-1 Isolation Actuation Instrumenta-i' tion Surveillance Requirements....

3/4 3-23 PERRY - UNIT 1 vi Amend:nent No. 29,33

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ADMIN!$TRATIVE CONTROLS SECTION NG[

l 6.1 RESpoNstI L I Tv...............................................

5-1 6.2 URGANIZAT!0N.................................................

5-1 6.2.1

. Corporate...............................................

6-1 Figure 6.2.1-1 Corporate Organization.............

6-3

- 6.2.2 Unit Staff..............................................

6-1 Figure 6.2.2-1 Unit Organization..................

6-4 Table 6.2.2-1 Minimun shift Crew Composition.........................

6-6 r

6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP Function................:.............................

6-7 i

i Composition............................................

6-7 t F Responsibilities.......................................

6-7 L

t Records................................................

6-7 6.2.4 SHIFT TECHNICAL ADVIS0R................................

6-7 t

6.3 UNIT STAFF QUALIFICATIONS...................................

6-7 6.4-TRAINING....................................................

6-4 1

6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC)

L Function.................'............................

6-8 Composition...........................................

6-8 Alternates.............................................

6-8 Meeting Frequency.....................................

6-8 Quorum.................................................

6-9 f-L Responsibilities......................................

6-9 I

Records................................................

6-10 l

PERRY,- UNIT 1 xxv f

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$ECTION f,g[

6. 5. 2 wucuAR sArrTi arvrew comrTTtt (nsec)

Function..............................................

6-10 Composition...........................................

5-11 A l te rn8 tts.............................................6*11 c o n s u l ta n t s............................................

6 11 Meeting Frequency......................................

6-11 Q uo r um.........'........................................

5 11 Review.................................................

6-12 Audits.................................................

5 12

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Records................................................

6-13

6. 5. 3 TECHNICAL REVIEW AND CONTROL ACTIVITIES................

5-14 6.6 REPORTABLE EVENT ACT!0N......................................

4-15 j

6.7 $AFETY LIMIT V!0LAT!0N......................................

5-15 6.8 PROC EDURES. INSTRUCTIONS AND PR0 GRAMS.......................

5-15 6._9 REPORTINGRE0V!REMEG 6.9.1 ROUTINE REP 0RTS........................................

6-17 Startup Report.........................................

6-17 Annual Reports.........................................

6-18 Annual Radiological Environmental Operating Report.....

6-18 Semiannual Radioactive Effluent Release Report.........

6-19 j

Monthly Operating Reports..............................

6-21 l

i Co re Opera ting Limi ts Report...........................

6-21 l

6.9.2-SPECIAL REP 0RTS........................................

6-21 6.10 RECORD RETENTION...........................................

6-21a l

6.11 RADI AT ION PROTECTION PR0 GRAM...............................

6-23 PERRY - UNIT 1 xxv1 Amendment No. 22. 33

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)0 DEFINITIONS The following tems are defined so that unifem interpretation of these i

i specifications may be achieved.

The defined teres appear in capitalized type and shall be applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.

AVERAGE PLANAR EXPOSURE 1.2 The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLNGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter witch the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alam and/or trip functions, and shall include the CHANNEL FUNCTIONAL TCST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.

This determination shall include, where pos-sible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:

a.

Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips, b.

Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.

PERRY - UNIT 1 1-1 r <

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DEFINITIONS i

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

Nonnal movement of the SRMs. IRMs. LPRMs. TIPS, or special movable detectors is not considered a CORE ALTERATION. Suspension of CORE ALTFRATIONS shall not preclude j

completion of the movement of a component to a safe conservative ponition.

CORE OPERATING LIMITS REPORT i

1.8 The CORE OPERATING LIMITS REPORT is the Perry Unit 1-specific document that provides_ the core operating limits-for the current operating reload cycle.

i These cycle-specific core operating limits shall be detennined for each reload cycle in accordance with Specification -6.9.1.9.

Plant operation within these operating limits is addressed in individual Specifications.

CRITICAL POWER RATIO 1.9 The CRITICAL POWER RATIO (CPR) shall be the ratio of that. power in the assembly which is calculated by application of a General Electric critical

~

power correlation to cause some point in the assembly to experience boiling i

transition, divided by the actual assembly operating power.

DOSE EQUIVALEfG I-131 s 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic

- ixture of I-131. I-132,1-133.1-134, and I-135 actually.present. The thyroid m

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

DRYWELL INTEGRITY 1.11 DRYWELL INTEGRITY shall exist when:

a.

All drywell penetrations required to be closed during accident conditions are either:

1.

Capable of being clo' sed by an OPERABLE automatic isolation system, or

=2.

Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position.

b.

The drywell equipment hatch is closed and sealed.

c.

The drywell head is installed and sealed.

d.

The drywell air lock is in compliance with the requirements of Specification 3.6.2.3.

e.

The drywell leakage rates are within the limits of Specification 3.6.2.2.

PERRY - UNIT 1 1-2 Amendment No. 20,33 u

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- POWER DISTRIBUTION LIMIT _!i 3/4.2 POWER DISTRIBUTION LIMITS f

f 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE tIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR UINEAR HEAT GENERATION RATES (APLHGRs) thall not exceed the limits specified in the CORE OPERATING LIMITS REPORT.

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

i ACTION:

If at any time during operation it is determined that an APLHGR is exceeding the limits specified in the CORE OPERATING LIMITS REPORT, initiate corrective l

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

l SURVEILLANCE REQUIREMENTS 1

4.2.1 All APLHGRs shall be verified to be equal to or less than the above limits:

l a.

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

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

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

d.

The provisions of Specification 4.0.4 are not applicable.

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PERRY - UNIT 1 3/4 2-1 Amendment No. 20,33 I,9

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I POWER DISTRIBUTION LIMITS

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3/4.2.2 MINIMUM CRITICAL POWER RATIO i

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LIMITING CONDITION FOR OPERATION I

i 3.2.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater

)

than the MCPR, limits specified in the CORE OPERATING LIMITS REPORT at the indicated core flow. THERMAL POWER, AT* and core average exposure compared to End of Cycle Exposure (EOCE)**.

APPLICABILITY:

OPERATIONAL' CONDITION 1, when THERMAL POWER is greater than or i

equal to 25% of RATED THERMAL POWER.

I ACTION:

With MCPR less than the limits specified in the CORE OPERATING LIMITS REPORT, i

L initiate corrective action within 15 minutes and restore MCPR to within the 1

l required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of l

RATED THERML POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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SURVEILLANCE REQUIREMENTS I

4.2.2 MCPR shall be determined to be equal to or greater than the limits specified in the CORE OPERATING LIMITS REPORT:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 1

b.

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

c.

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

d.

The provisions of Specification 4.0.4 are not applicable.

This AT refers to the planned reduction of rated feedwater temperature from nominal rated feedwater temperature (420'F), such as prolonged removal of.

i feedwater heater (s) from service.

I

- End of Cycle Exposure (EOCE) is defined as 1) the core' average exposures at which there is no longer sufficient reactivity to achieve RATED THERMAL i

POWER with rated core flow, all control rods withdrawn, all feedwater heaters in service and equilibrium Xenon, or 2) as specified by the fuel i:

vendor.

l PERRY - UNIT 1 3/4 2-2 Amendment No. 29,33 l

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t POWER DISTRIBUTION LIMIT 3/4.2.3 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.)fied in the CORE OPERATING LIMITS REPORT.The LINEAR HEAT GENER speci APPLICABILITY:

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

A,CTION:

With the LHGR of a;,y fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit withir i hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER witl the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SUR7EILLANCE REQUIREMENTS 4.2.3 LHGR's shall be determined to be equal to or less than the linit:-

a.

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

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

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

d.

The provisions of Specification 4.0.4 are not applicable.

PERRY - UNIT 1 3/423 Amendment No. eik 33 l

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DES)GN FEATURES

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DESIGN TEMPERATURE AND PRES 5URE (Continued) b.

Maximum internal temperature:

1.

Drywell 330*F.

2.

Suppression pool 185'F.

Maximum external to internal differential pressure:

c.

1.

Drywell 21 psid.

2.

Containment 0.8 psid.

$ECONDARY CONTAINMENT 5.2.3 The secondary containment consists of the annulus between the shield i

building and the primary containment and has a minimum free volume of 392,548 cubic feet.

5. 3 REACTOR CORE FUEL ASSEMBLIES I

5.3.1 The reactor shall contain 748 fuel assemblies.

Each assembly shall consist of a matrix of Zircaloy clad fuel rods with an initial composition of slightly enriched uranium dioxide (UD,) as fuel material.

Fuel assemblies shall be limited to those fuel designs approved by the NRC Staff for use in BWR's.

QNTROLRODASSEMBLIES 5.3.2 The reactor core shall contain 177 control rod assemblies each t

consistingofacruciformarrayofstainlesssteeltubescontainIng i

143.7 inches of boron carbide, 8 C, powder surrounded by a cruciform shaped stainless steel sheath.

4 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a.

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

For a pressure of:

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

6 PERRY - UNIT 1 5-4 Amendment No. 33

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l ADMINISTRATIVE CONTROLS l

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l 1rMIANNUAL RADIDACTIVE EFFLUENT RELEASE REPORT (Continued) l

, The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the GIFSITE DOSE CALCULATION MANUAL (ODCM), pursuant to Specifications 6.13 and 6.14, i

respectively, as well as any major change to Liquid, Gaseous, or Solid Radweste Treatment Systems pursuant to Specification 6.15.

It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Specification 3.12.2.

The Semiannual Radioactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gaseous s

effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.7.9 or 3.3.7.10, respectively; and description of the events leading to liquid holdup tanks exceeding the limits of Specification 3.11.1.4.

MONTHLY OPERATING REPORTS

6. 9.1. 8 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Resource Management.

U. S. Nuclear Regulatory Comission, Washington, D. C. 20555, with a copy to the Regional Administrator of the Regional Office no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT 6.9.1.9 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

(1) The Average Planar Linear Heat Generation Rate (APLHGR) for Technical Specification 3.2.1.

(2) The Minimum Critical Power Ratio (MCPR) for Technical Specification 3.2.2.

(

(3) The Linear Heat Generation Rate (LHGR) fo Technical Specification 3.2.3.

The analytical methods used t'o determine the core operating limits shall be those previously reviewed and approved by NRC in NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel.

(The approved revision at the time reload analyses are performed shall be identified in the COLR.)

4

)

The core operating limits shall be determined 50 that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as FiUTDOWN MARGIN, and transient and accident analysislimits)ofthesafetyanalysisaremet.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, i

to %e U.S. Nuclear Regulatory Commission Document Control Desk with copies to

.c Regional Administrator and Resident Inspector.

PERK.

llNIT 1 6-21 Amendment No. 33 J

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ADMINISTRATIVE CONTROLS I

i SPECI AL REPORTS l

6.9.2 Special reports shall be submitted to the Regional Administraton of the Regional Office within the time period specified for each report.

6.9.3 Safety-relief valve failures will be reported to the Regional Administrator of the Regional Office of the NRC via the Licensee Event Report system within 30 days.

6.9.4 Violations of the requirements of the fire protection program described in the Final Safety Analysis Report which would have adversely affected the ability to achieve and maintain safe shutdown in the event of a fire shall be reported to the Regional Administrator of the Regional Office of the NRC via.

the Licensee Event Report system within 30 days.

n 6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.2 The following records shall be retained for at least 5 years:

a.

Records and logs of unit operation covering the interval at each power level.

b.

Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear safety.

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PERRY - UNIT 1 6-21a Amendment No. 33

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1 ADMINISTRAT!VE CONTROLS-RECORD RETENTION (Continued) 1 c.

All REPORTABLE EVENTS.

J d.

Records of surveillance activities inspections, and calibrations requiredbytheseTechnicalSpecifIcations.

Records of changes made to the procedures required by Specifiestion e.

6.8.1.

f.

Records of radioactive shipments.

g.

Records of sealed source and fission detector leak tests and results.

I h.

Records of annual physical inventory of all sealed source material of record.

i l

6.10.3.The following records shall be retained for the duration of the unit Operating License:

' i, Records and drawing changes reflecting unit design modifications made a.

to systems and equipment described in the Final safety Analysis Report, b.

Records of new and irradiated fuel inventory, fuel transfers, and l

assembly burnup histories.

\\

Records of radiation exposure for all individuals entering radiation c.

centrol areas.

d.

Records of gaseous and liquid radioactive material released to the

environs, Records of transient or operational cycles for those unit components e.

identified in Table 5.7.1-1.

i f.

Records of reactor tests and experiments.

g.

Records of training and qualification for current members of the unit staff.

h.

Records of inservice inspections performed pursuant to these Technical Specifications.

1.

Records of quality assurance activities required by the Operational Quality Assurance Manual.

j.

Records of reviews performed for changes made to procedures or equip-mont or reviews of tests and experiments pursuant to 10 CrA 50.59.

k.

Records of meetings of the PORC and the NSRC.

PERRY - UNIT.1 6-22

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j A7.

j 3/12 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temper-ature following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in 10 CFR 50.46.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1

i Thisspecificationassuresthatthepeakcladdingtemperature(PCT) following the postulated design basis Loss-of-Coolant Accident (LOCA) will not exceed the limits specified in 10 CFR 50.46 and that the fuel design analysis limitsspecifiedinGESTAR-II(Reference 1)willnotbeexceeded, l

1 The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only 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 corrected for deDSification.

This LHGR times 1.02 is used in the heatup code along with the exposure depen-dent steady state gap conductance and rod-to-rod local peaking factor. The i

Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor.

The MAPLHGR limits specified in the CORE OPERATING LIMITS REPORT are l

l multipliedbythesmallerofeithertheflowdependentMAPLHGRfactor(MAPFAC) 1 g

or the power dependent MAPLHGR factor (MAPFAC ) corresponding to existing core p

I flow and power state to assure the adherence to fuel mechanical detign bases during the most limiting transient.,,MAPFAC 's are detemined using the three-f dimensional BWR simulator code to analyze slow flow runout transients.

[

MAPFAC 's.are generated using the same data base as the MCPR to protect the l

p p

core from plant transients other than core flow increases.

,- The Technical Specification MAPLHGR value is the most limiting composita of the fuel rechanical design analysis MAPLHGR and the ECCS MAPLHGR.

Fuel Mechanical Design Analysfs:

NRC approved methods (specified in Reference 1) are used to demonstrate that all fuel rods in a lattice, operating at the bounding power history, meet the fuel design limits-specified in Reference 1.

This bounding power history is used as the i

j basis for the fuel design analysis MAPLHGR value.

LOCA Analysis: A LOCA analysis is perfomed in accordance with 10 CFR Part 50 Appendix X to demonstrate that the MAPLHGR values comply with the ECCS limits specified in 10 CFR 50.46. The analysis is performed for the most limiting break size, break location, and single failure combination for the plant.

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PERRY - UNIT 1 B 3/4 2-1 AmendmentNo.h33 l

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

Only the most limiting MAPLHGR values are shown in the Technical Specifi-cation figures for multiple lattice fuel. When hand calculations are required, these Technical Specification MAPLHGR figure values for that fuel type are used for all lattices in that bundle.

For some GE fuel bundle designs MAPLHGR depends only on bundle type and burnup.

Other GE fuel bundles have MAPLHGRs that vary axially depending upon the specific combination of enriched uranium and gadolinia that comprises a fuel bundle. cross section at a particular axial node.

Each particular combination of enriched uranium and gadolinia, for these fuel bundle types, is called a lattice type by GE.

These particular fuel bundle types have MAPLHGRs that vary by lattice type (4xia11y) as well as with fuel burnup.

ApprovedMAPLHGRvalues(limitingvaluesofAPLHGR)asafunctionof fuel and lattice types, and as a function of the averaoe olanar exposure are provided in the. CORE OPERATING LIMITS REPORT.

s.

PERRY - UNIT 1 B 3/4 2-2 Amendment No. 4X133 l

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PERRY - UNIT 1 B 3/4 2-3 Amendment No. 20 t

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_ POWER DISTRIBUTION LIMITS BASES 3/4.2.2 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.2 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.07 and an analysis of the limiting operational transients.

For any abnonnal operating transient analysis evalua.

tion 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 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the inrgest reduc-l tion in CRITICAL POWER RATIO (CPR). The type of transients evaluated are documented in the USAR and Reference 1.

The limiting transient _ yields the.

largest delta CPR.

When added to the Safety Limit MCPR, the required operating limit MCPR of Specification 3.2.2 is obtained. The power-flow map of Figure B 3/4 2.2-1 defines the analytical basis for generation of the MCPR operating limits.

The evaluation of a given transient begins with the system initial parameters shown in USAR Chapter 15 and/or Reference 1 that are input l

to a GE-core dynamic behavior transient computer program. The codes used to evaluate these events are described in Reference 1.

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The purpose of the MCPRf and 1CPa is to define operating limits at other p

than rated core flow and power conditions.

At less than 100% of rated flow and

[

power the required MCPR is the larger value of the MCPR and MCPR at the f

p existing core flow and power state.

The MCPR s are established to protect the f

core from inadvertent core flow increases such that the 99.9% MCPR. limit L

j requirement can be assured.

L The MCPR figure contained in -the CORE OPERATING LIMITS REPORT also l

L reflects the Eequired MCPR values resulting from the analysis performed to justify operation with the feedwater temperature ranging from 420'F to 320'F at 100% RATED THERMAL POWER steady state conditions, and also beyond the end of cycle with the feedwater temperature ranging from 420'F and 250'F.

The MCPR s were calculated such that for the maximum core flow rate and f

the corresponding-THERMAL POWER along a conservative steep generic power flow control line, the limiting bundle's relative )ower was adjusted until the MCPR was slightly above the Safety Limit.

Using t11s relative bundle power. the MCPRs were calculated at different points along this conservative steep power flow control line corresponding to different core flows.

The calculated MCPR at a given point of core flow is defined as MCPR.

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PERRY - UNIT 1 8 3/4 2-4 Amendment No. /f/,33 J

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