Text

Proposed Tech Specs Moving Tech Specs Re Core Max Fraction of Limiting Power Density to Core Operating Limits Rept
	ML20011D921
Person / Time
Site:	Perry
Issue date:	12/19/1989
From:	; CLEVELAND ELECTRIC ILLUMINATING CO.
To:	;
Shared Package
ML20011D920	List: ML20011D919; ML20011D921;
References
	NUDOCS 9001030030
	Download: ML20011D921 (41)
	v • d • e

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I-I 9001030030 891219 PDR ADOCK 05000440 P PDC

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-.....m-4 Attachesnt _3 PY-CEI/NRR-1104 L Page i of 20

, h.7yu E DEFINITIONS ,_,

l SECTION 4

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

1. 2 AVERAGE PLANAR EXPOSURE...................................... 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................... 1-1 1.4 CHANNEL CALIBRATION.......................................... 1-1 1.5 CHANNEL CHECK................................................ 1-1

1. 6 CHANNEL FUNCTIONAL TEST...................................... 1-1 1.7 CORE ALTERATION.............................. . ............. 1-2 -

orsnar#Ns umTs RaroRT4 Merlue d%

1. 8 CORE (MAXIMUM FRACTION OF LIMITING POWER DENSIT JYo ~....... ... 1-2

1. 9 CRITICAL POWER RATI0......................................... 1-2' 1.10 DOSE EQUIVALENT I-131........................................ 1-2

(

1.11 DRYWELL INTEGRITY.....,....................................... 1-2 1 1.12 E-AVERAGE DISINTEGRATION ENERGY.............................. 1-3 l l

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 THERMAL P0WER. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.17 FREQUENCY N0TATION........................................... 1-4 1.18 FUEL HANDLING BUILDING INTEGRITY............................. 1-4 1.19 GASEOUS RADWASTE TREATMENT (OFFGAS) SYSTEM. . . . . . . . . . . . . . . . . . . 1-4 1.20 IDENTIFIED LEAKAGE........................................... 1-4 1.21 ISOLATION SYSTEM RESPONSE TIME............................... 1-4 1.22 LIMITING CONTROL ROD PATTERN................................. 1-5

,- 1.23 LINEAR HEAT GENERATION RATE.................................. 1-5 k

PERRY - UNIT 1 i

, .. . _ , - . ~. -

e Attachment 3 _

PY-CEI/NRR-1104 L Page 1 of to LIMIT ~ LNG 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 IV ITY AN0 MAL I ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1 - 2

, 3/4.1.3 CONTROL RODS Control Rod Operabili ty. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-3 Control Rod Maximum Scram Insertion Times.............. 3/4 1-6 Control Rod Scram Accumulators......................... 3/4 1-8 i Control Rod Drive Coupling. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-10 Control Rod Position Indication. . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-12 ,

1 Control Rod Drive Housing Support. . . . . . . . . . . . . . . . . . . . . . 3/4 1-14 l

3/4.1.4 CONTROL R0D PROGRAM CONTROLS Control Rod Wi thdrawa1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-15 Rod Pattern Control System............................. 3/4 1-16 3/4.1.5 STANDBY L IQUID CONTROL SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-18 Figure 3.1.5-1 Sodium Pentaborate Solution-Concentration / Volume Require-me n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 1 - 2 0 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE. . . . . . . . . . . . . 3/4 2-1 Figure 3.2.1-1 MAPFAC f ........................... 3/4 2-2 Figure 3.2.1-2 MAPFAC ........................... 3/4 2-3 p

Figure 3.2.1-3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types BP8 SRB 219....... 3/4 2-4 PERRY - UNIT 1 v Amendment No.20 )

l At tschunt 1 )

PY-CEI/NRR-1104 L d Page 3 of A8 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i 1

SECTION PAGE POWER DISTRIBUTION LIMITS (Continued) p, M,

[ Figure 3.2.1-4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial I

Core Fuel Types BP8 SRB 176....... 3/4 2-5 l 1

< 1 l~

Figure 3.2.1-5 Maximum Average Planar Linear Heat '

! Generation Rate (MAPLHGR) Versus j Average Planar Exposure Reload Core f uel Types BS301E. . . . . . . . . . . . 3/4 2-6 Figure 3.2.1-6 Maximum Average Planar Linear Heat ,

Generation Rate (MAPLHGR) Versus Average Planar Exposure Reload Core Fuel Types BS301F............ 3/4 2-6a !

3/4.2.2 MINIMUM CRITICAL POWER RATI0........................... 3/4 2-% A l

bFig u r e 3. 2,2- 1 MCP R . . / . . . . . . . . . . . . . . . . . . . . . . . . . _ .

f 3/ 4 2- 8 p, j,1, 1

mus====

l Figure 3.2.2-2 MCPR ............................. 3/4 2-9 p

- 1 3/4.2.3 LINEAR HEAT GENERATION RATE. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-)( 3 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION. . . . . . . . . . . . . . 3/4 3-1 Table 3.3.1-1 Reactor Protection System Instrumentation................... 3/4 3-2 Table 3.3.1-2 Reactor Protection System l Response Times.................... 3/4 3-6 Table 4.3.1.1-1 Reactor Protection System l Instrumentation Surveillance l Requirements...................... 3/4 3 1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.................... 3/4 3-9 i l

Table 3.3.2-1 Isolation Actuation Ins trumenta tion. . . . . . . . . . . . . . . . . . . 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 Time..................... 3/4 3-21 Table 4.3.2.1-1 Isolation Actuation Instrumenta- ,

tion Surveillance Requirements. . . . 3/4 3-23 1 PERRY - UNIT 1 vi Amendment No. 20 l l

4 Attachment 3 PY-CEI/NRR-1104 L l Page 1 of _Jo l

ADN!N!$TRATIVE CONTROLS E .

6.5.2 NUCLEAR SAFETY REVIEW Cop 0GTTEE (H5RC) .

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

6-10 Composition ........................................... 5-12 I r

A l te rna te s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F11 Consultants............................................ 6-11 i

)

Meeting Frequency...................................... 6-11' i

I Q uo run . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-11 '

t Review................................................. 4-12 Audits................................................. .

5-12 Records................................................ F13 6.5.3 TECHNICAL REVI EW AND CONTROL ACTIVITIES. . . . . . . . . . . . . . . . 5-14 6,6 REPORTABLE EVENT ACT10N.................................,,... 6-15 6.7 SAFETY LIMIT V10LATION...................................... 6-15 6.8 PROCEDURES. INSTRUCTIONS Ale panap_ms,,,,,,,,,,,,,,,,,,,,,,, pg$ -

6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS ....................................... F17 -

Startup Report......................................... F17 '

Annual Reports ........................................ FIS Annual Radiological Environmental Operating Report..... F18 Semiannual Radioactive Effluent Release Report......... 5-13 Monthly Operating Reports.............................. 6-21 c oR E dra A ArtN C LIN l rs REroRT . . . . . . . . . . . . . . . . . . . . . . . . 6-Al AAA tugues 6.9.2 SPECIAL REP 0Rfs........................................ 6-21 6.10 RECORD RETENTION........................................... -

6-21 A

/

i 6.11 RADIATION PROTECTION Panne_"............................... F23 PERRY - UNIT 1 xxv1 Amendient No.22

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i Attachment 3 '

PY-CEI/NRR-Il04 L DEFINITIONS Page _ .5- of Jt o CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor >

pressure vessel with the vessel head removed and fuel in the vessel. Normal movement of the SRMs, IRMs. LPRMs, TIPS, or special movable detectors is not considered a CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to.a safe conservative position, fCOREMAXIMUMFRACTION0FLIMITINGPOWERDENSITY g 1.8 The CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY (CMFLPD) shall be the oA thighest value of the FLPD which exists in the core. rerlut

, with l .

CRITICAL POWER RATIO I .,t 1

[

! 1.9 The CRITICAL POWER RATIO (CPR) chall be the ratio of that power in the , .,

assemblylfiich is calculated by application of a General Electric critical

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

- transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 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 mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

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

DRYWELL INTEGRITY 1.11 DRYWELL INTEGRITY shall exist when:

a. All drywell penetrations required to be closed during accident l conditions are either

! 1. Capable of being closed 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.

I

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

Attccht nt-3 ;'

PY-CEI/NRR-1104 L Page 6 of Ro

-Insert 1 CORE OPERATING LIMITS REPORT 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. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Plant operation within these-operating limits is addressed in individual' Specifications, l

, . . ~ .

Attachment ,J_

POWER DISTRIBUTION LIMITS Pas o gj y 3/4.2 POWER DISTRIBilTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATLON RATES (APLHGRd shall not exceed t result obtained from multiplying the appl' cable 1%PLHGR va' ves* by the ler of either the flow dependent MAPLHGR factor (MAPFAC ) of Figure 3.2.1-L 7A -

f or the power dependent MAPLHGR factor (MAPFAC p

) of Figure 3.2.1-2[ limits specWeA in the CoAE OPEAAT/M LIMITS APPLICABILITY _: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or REPS RT, equal to 25% of RATED THERMAL POWER.

limitt specified in the coAE ofERAT/W LIMITS MEPoKT)

If t any time durina operation it is deMined that an APLHGR is exceeding th result of the above multiplicationDinitiate corrective 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 r 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l SURVEILLANCE REQUIREMENTS l(

l 4. 2.1 All APLHGRs shall be verified to be equal to or less than the above j limits:

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,:andi

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.

These applicable MAPLHGR values are:

l

1) Those that have been approved for the respective fuel and lattice type as a function of the average planar exposure (as determined by the NRC approvedmethodologydescribedinGESTAR-II) or

2) When hand calculations are required, the MAPLHGR as a function of the average planar exposure for the most limiting lattice (excluding natural l

uranium) shown in the Figures 3.2.1-3, 3.2.1-4, 3.2.1-5, and 3.2.14 for the applicable type of fuel. j-PERRY - UNIT 1 3/4 2-1 Amendment No. 20 r - ,

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20 40 60 100 120 CORE THERMAL POWER (% RATED),

l POWER DEPENDENT MAPLHGR FACTOR.

(MAPFAC, )

FIGURE 3.2.1-2 PERRY - UNIT 1 3/42-3 Amendment No.20

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i Attachmtat J ,

PY-CE1/NRR-1104 L Page _I+ of Jo POWER DISTRIBUTION LIMITS specifkA in the. C88E -'

3/4.2.2 MINIMUM CRITICAL POWER RATIO opgg4yms t,girs LIMITING CONDITION FOR OPERATION 8""7 # U" 3.2.2 "3 IMUM CRITIC POWER than%)pP%aneMcPRQ11 sits TIO (MCPR) indicated core flow, shall HERMAL be eP0ual R, to AT*ordrester and core averana exnostce connared to End of Cycle Exposure (E0CE)*gs shown ink (Figures 3.2.2-1 anc 3.2.2-2.Q APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

limits specifieA in the CORE ol'EAATMG UNITS REPs RT, With MCP ssthanthefaDD11 cab' e McPR limit shown in Figures 3.2.2-1 andW (3.2. N-2.linitiate corrective act'on within 15 minutes and restore swrn to -

witn' n the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS l

4.2.2 MCPR shall be determined to be eaua' to or greater than the(MCP,g) P imitf (determinedfromFigures3.2.2-1and3.2.2-2 specifjeA (, the cagc

a. . At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . *"A*' # * * ~

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL P0WER 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 R0D PATTERN for MCPR.

l-

d. The provisions of Specification 4.0.4 are not applicable.

4 "This AT refers to the planned reduction of rated feedwater temperature from nominal rated feedwater temperature (420*F), such es prolonged removal of feedwater heater (s) from service.

- End of Cycle Exposure (E0CE) is defined as 1) the core average exposures at which there is no longer sufficient reactivity to achieve RATED THERMAL POWER with rated core flow, all control rods withdrawn, all feedwater heaters in service and equilibrium Xenon, or 2) as specified by the fuel vendor, t

PERRY - UNIT 1 Amendment No. 20 3/42-X z !

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FIGURE 3.2.2-1 PERRY - UNIT 1 3/42-8 dment No. 20

-, . . _ . _ , _ _ _ . , _ - _ ..._._ _ _ . ---m__ -

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POWER DEPENDENT MCPR CTOR !

(MCPRp)

FIGURE 3.2.2-2 PERRY - UNIT 1- 3/4 2-9 Amend. t No. 20

, ...:,, . ,c

-i-Attachment 2 i

PY-CEI/NRR-1104 L Page y of y

-t POWER DISTRIBUTION LIMIT b

3/4.2.3 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.3' The LINEAR HEAT GENERATION RATE (LHGR) shall not excee he limit speifieJt, in he CoM

a. 13.4kw/ftforBP8x8Rfuel]JL ofERATING,

b. 14.4 kw/ft for GE8x8EB fuel.J LIMITS AE PoltT.

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

ACTION:

With the LHGR of any fuel- rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

-t SURVEILLANCE REQUIREMENTS 4.2.3 LHGR's shall be determined to be equal to or less than the limit: ,

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

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

1 i I

PERRY - UNIT 1 3/4 2- Amendment No. 20 l 3 :

l l 1 . .

l

l Attachment 't M RR-Il04 L DESIGN FEATURES Page It of 26

' i DESIGN TEMPERATURE AND PRESSURE (Continued) l

b. Maximum internal temperature:

1. Drywell 330'F.

2. Suppression pool 185'F.

j~ c. Maximum external to internal differential pressure:

1. Drywell 21 psid.

2. Containment 0.8 psid.

SECONDARY CONTAINMENT 1 5.2.3 The secondary containment consists of the annulus httween the shield building and the primary containment and has a minimum free volume of 392,548 cubic feet. .

fanA shall be limited to those fuel assemblies which have been 3

5. 3 REACTOR CORE analyzed with NRC approved codes and methodsyhave been '

shown to comply with all Safety Design Bases in the%SAR, am A are identifisA la the COR8 OP8AATime U FUEL ASSEMBLIES LIMITS ASPoRT'. j 5.3.1 The reactor core shn11 contain 748 fuel assemb'ieEwith each fuel firssembly containing 62 fue' rods and two wa;er rods c' ad with Zircaloy-2.

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

initial core loading shall have a maximum average enrichment of 1.9 weight percent U-235. Reload fuel shall be similar in physirsi design to the initial 4

Qoreirsding. Jg CONTROL ROD ASSEMBLIES ,

5.3.2 The reactor core shall contain 177 control rod assemblies each consisting of a crucifom array of stainless steel tubes containIng 143.7 inches of boron carbide, 8 C, 4 powder surrounded by a crucifo m shaped stainless steel sheath.

5.4 REACTOR COOLANT SYSTEM ,

i DESIGN PRES $URE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

)

i 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, l

b. For a pressure of:

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

PERRY - UNIT 1 5-4

l Attachment 1 R 0 L ADMINISTRATIVE CONTROLS h,.,M T SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) i The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the

. OFFSITE DOSE CALCULATION MANUAL (ODCM), pursuant to Specifications 6.13 and 6.14, J

respectively, as well as any major change to Liquid, Gaseous, or Solid Radwaste Treatment Systems pursuant to Specification 6.15. It shall also include a list- l ing of new locations for dose calculations and/or environmental monitoring iden-t tified by the Land Use Census pursuant to Specification 3.12.2. ,

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

' 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 1 4 leading to liquid holdup tanks exceeding the limits of Specification 3.11.1.4.

i i MONTHLY OPERATING REPORTS 6.9.1.8 Routine reports of operating statistics and shutdown experience shall i be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission. Washington, D.C. 20555, with a copy to the ,

Regional Administrator of the Regional Office no later than the 15th of each j moJ th following the calendar month covered by the report. I SPECIAL REPORTS "d A }

6.9.2 Special reports shall be submitted to the Regional Administrator 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 License Event Report system within 30 days. ;

l 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 l reported to the Regional Administrator of the Regional Office of the NRC via the Licensee Event Report system within 30 days.

6.10 RECORD RETENTION Move t e new 4-h I'ese 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:

l a. Records and logs of unit operation covering time 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. y PERRY - UNIT 1 6-21

Attachnent 3 PY-CEI/NRR-1104 L Page 18 of 16 Insert 2 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. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in the Gentral Electric reload licensing topical report General Electric Standard Application for Reactor Fuel - GESTAR II: NEDE-24011-P-A and US (latest approved revision). The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOVN MARGIN, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements therets, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

l I

ATTACHMENT 4 t

i TECHNICAL SPECIFICATION i

BASES CHANGES ,

f t

1 r

l A - - _ _ ._ _ _ _ _ _ _

Attachment _ Y f' i

PY-CEI/NRR-1104 L

3/4.2 POWER DISTRIBUTION LIMITS Page l of ,3,, i l

BASES j

-~

The specifications of this section assure that the peak cladding temper- i ature following the postulated design basis loss-of-coolant accident will not

exceed the 2200'F limit specified in 10 CFR $0.46.

{

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE l, This specification assures that the peak cladding temperature (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, t 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 densification. .

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 Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATIO TE(APLHGR) is this LHGR of the biobest onwered rod divided by its loca aking factor.

The MAPLHGR limits (of Figures 3. 2.1-1. 3. 2.1-2. and 3. 2. ' -37tre multiDlied by the smaller of either the flow dependent MAPLH5R factor (MAP'ACf ) ur the powe dependent MAPLHGR factor (MAPFAC ) corresponding to existing core flow and sped % A p

power state to assure the adherence to fuel mechanicalW esign bases during

N' the most limiting transient. MAPFACf 's are determined using the three-8A[

dimensional BWR simulator code to analyze slow flow runout transients. t:Mits MAPFAC p

's are generated using the same data base as the MCPR p to protect the MEPeRT core from plant transients other than core flow increases. ,

The Technical Specification MAPLHGR value is the most limiting composite '

i of the fuel mechanical design analysis MAPLHGR and the ECCS MAPLHGR.

Fuel Mechanical Design Analysis: NRCapprovedmethods(specifiedin ,

Reference 1) are used to demonstrate that all fuel rods in a lattice. I operating at the bounding power history, meet the fuel design limits I specified in Reference 1. This bounding power history is used as the ,

basis for the fuel design analysis MAPLHGR value. !

1 LOCA Analysis: A LOCA analysis is performed in accordance with 10 CFR i Part 50 Appendix K 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.

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

-_-___--____A

Attachment Y PY-CEI/NRR-1104 L Page_2 of 1 I l d POWER DISTRIBUTION LIMITS l l

l BASES

l. I 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. Wheq hand calculations are required, these Technical Specification MAPLHGR figure talues for that fuel type are used for all lattices in that bundle.

For some GE fuel bundle designs MAPU,GR depends only on bundle type and burnup. Other GE fuel bundles have MAPU,GRs that vary axially depending upon

the specific combination of enricheo urtnium and gadolinia that comprises a >

fuel bundle cross section at a particu'ar axial node. Each particular combination of enriched uranium and grdolinia, for these fuel bundle types, is called a lattice type by GE. The'e particular fuel bundle types have MAPLHGRs that very by lattice type (axially) as well as with fuel burnup.

Approved MAPLHGR values (limiting values of APLHGR) as a function of fuel and lat': ice typos and as a function of the average planar exposure are provided in["echnica 5pecification Figures 3.2.1-3 through 3.2.1-6) ,

(( the CORE o rER ATING LIMITS REPORT.

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

___a --

I Attachment f PY-CEI/NRR-1104 L ,

Page 1 M 1 POWER DISTRIBUTION LIMITS BASES l 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 l operational transients. For any abnomal 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 t below the Safety Limit MCPR at any time during the transient assuming instrument i 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 transiente have been analyzed to determine which result in the largest reduc- i 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 systum initial Darameters shown in USAR Chaoter 15 and/or Reference % and c' eveland3 J2 fE' ectric's November 28 and December 29. 1988 submitta a Inst are input to V a GE-core dynamic behavtor transient computer program. The codes used to evaluate these events are described in Reference 1.

The purpose of the i1CPRf and :1CPa p

is to define operating limits at other 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 MCPRf s are established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit .

requirement can be assured.

O /The ScfAp figure containtA in the C*RE d6TRATlA)4 LM4875 REPaRT* 1 (Fioure 3.2.2-27also reflects the required MCPR values resulting from the analysis perfomed 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 f s were calculated such that for the maximum core flow rate and the corresponding THERMAL POWER along a conservative steep generic power flow control line, the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the 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 . f l

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

.m---wt,- - - , - - ._w ., - - , _ _ . m +-- - - _ _ - _ - _ . _ _

, . 4. . .m.

,, . e., =

.. n.-' '

8

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'k4 I,

+ '- F- -s '\I' 0

o - '..e'3'.

~

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, ATTACBnNT 5' .

. f' "

- a

% 1

. : CORE'0PERATIM LIMITS REPORT: .,

.% y h y'

f. ';

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m

I Attnehment 5 OM18: PDB-F-001 PY-CEI/NRR-1104 L Page: 1 of 4 :

Page 1 of lj[ Rev.: 0 ,

F THE CLEVELAND ELECTRIC ILLUMINATING COMPANY i

i PERRY NUCLEAR POWER PLANT OPERATIONS MANUAL Plant Data Book - TAB F TITLE: UNIT 1 CORE OPERATING LIMITS REPORT CYCLE 2 (RELOAD 1) j REVISION : 0 EFFECTIVE DATE:

P PREPARER:

/ Date REVIEVER:

/ Date PORC NEETING NO.:

/ Date APPROVED:

/ Date CORE' OPERATING

(

PERRY - UNIT 1 LIMITS REPORT l

I Attachment 5 OM18: PDB-F-001 PY-CEI/NRR-1104 L Page 2 of 4 Page 2 of if Rev.: O CORE OPERATING LIMITS REPORT PERRY NUCLEAR POVER PIM T - UNIT 1 CYCLE 2 (RELOAD 1) REVISION O TABLE OF CONTENTS PDB Entry Number _ Title MPL Page(s)' Revision PDB-F-001 Title Page J11 1 0 Table of Contents Scope of Revision Introduction and References PDB-F-002 (Reserved)

PDB-F-100 Average Planar. Linear' Heat Generation Rate (corresponds to TS 3.2.1) J11 1 0 PDB-F-101 Flow Dependent MAPLHGR Factor (MAPFACf )

J11 1 0 PDB-F-102 Power Dependent MAPLHGR Factor (MAPFACp ) J11 1 0 PDB-F-103 MAPLHGR Versus Average Planar Exposure, Fuel Type BP85RB219 J11 1 0 PDB-F-104 MAPLHGR Versus Average Planar Exposure, Fuel Type BP8 SRB 176 J11 1 0 PDB-F-105 MAPLHGR Versus Average Planar Exposure, Fuel Type BS301E J11 1 0 PDB-F-106 MAPLHGR Versus Average Planar Exposure, Fuel Type BS301F J11 1 0 PDB-F-200 Minimum Critical Power Ratio (corresponds to TS 3.2.2) J11 1 0 PDB-F-201 Flow Dependent MCPR Factor (MCPRg ) J11 1 0 CORE OPERATING PERRY - UNIT 1 LIMITS REPORT !

Att ehment 5 OM18: PDB-F-001 PY-C'2/NRR-1104 L Page: 3 of 4 Page 3 of _[f Reva 0 TABLE OF C0hTMfS PDB Entry Number Title MP Page(s) Revision PDB-F-202 Power Dependent MCPR Factor (MCPRp ) Jll 1 0 PDB-F-300 Linear Heat Generation Rate (corresponds to TS 3.2.3) J11 1 0 NOTE: The preparer and reviewer and responsible for verJfying that all Plant Data Book entries shown in the above Table of contents are correct and reflect the fuel / core design information. Individual entries for this section of the Plant Data Book, Tab F - Core Operating Limits Report, may not be updated without a corresponding

~

revision to this entry, PDB-F l5D1 and subsequent PORC review of this and any other revised entries. Perry Plant Technical Dapartaent Director's approval is required for issuance of any reviaions to the Core Operating Limite Report.

SCOPE OF REVISION Affected Affected Revision PDB Entry (s) Summary of Changes Da g 0 All Original Issue 3/15/$0 CORE' OPERATING PERRY - UNIT 1 LIMITS REPORT l

J

1 i

Attachment 5 OM18: PDB-F-001 l PY-CEI/NRR-1104 L Page: 4 of 4 l Page f of ff Rev.: 0 INTRODUCTION AND REFERENCES i

INTRODUCTION This Core Operating Limits Report for PNPP Unit 1 Cycle 2 is prepared in accordance with the requirements of PNPP Technical Specification 6.9.1.9. The core operating limits presented here vere developed using NRC-approved methods !

(Reference 2). Results from the reload analyses for the General Electric fuel i in PNPP Unit 1 Cycle 2 are documented in References 3 and 4. ]

The cycle-specific core operating limits for the following PNPP Unit 1 Technical Specifications are included in this report:

1. Average Planar Lineer Heat Generation Rate (APLHGR) Limit (Technical Specification 3/4.2.1)

2. Minimum Critical Power Ratio Operating Limit (Technical Specification 3/4.2.3) ]

3. Linear Heat Generation Rate (LHGR) Limit (Technical Specification 3/4.2.4)

REFERENCES

1. USNRC Generic Letter 88-16, " Removal of Cycle-Specific Parameter Limits from Technical Specifications," October 4, 1988.

2. " General-Electric Standard Application for Reactor Fuel-GESTAR II,"

NEDE-24011-P-A (latest approved revision) and NEDE-24011-P-A-US (US Supplement - latest approved revision).

3. " Supplemental Reload Licensing Submittal for the Perry Nuclear Power Plant Unit 1, Reload 1, Cycle 2," GE Document 23A5948 Rev. 1. (November.

1988).

4. " Supplement 1 to the Supplemental Reload Licensing Submittal for the Perry Nuclear Power Plant Unit 1, Reload 1, Cycle 2," GE Document 23A5948AA Rev. O. (October 1988).

CORE OPERATING PERRY - UNIT 1 LIMITS REPORT ,

l Attachment 5 OM18: PDB-F-100 PY-CEI/NRR-1104 L Page 1 of 1 Page f of pr Rev.: 0 AVERAGE PLANAR LINEAR REAT GENERATION RATE (TS 3.2.1) l All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) shall not exceed the f result obtained from multiplying the applicable MAPLHGR values

by the smaller of either the flow dependent MAPLHGR factor (MAPFACg ) entry PDB-F-101, l or the power dependent MAPLHGR factor (KAPFAC p

) entry PDB-F 102. {

l t

These applicable MAPLHGR values are:

1. Those that have been approved for the respective fuel and lattice type as a function of the average planar exposure (as described by the NRC approved methodology described in GESTAR-II) i or,

2. When hand calculations are required, the MAPLHGR as a function of the average planar exposure for the most limiting lattice (excluding natural uranium) shown in entries PDB-F-103 through PDB-F-199 for the applicable type of fuel.

I l

CORE OPERATING I PERRY UNIT - 1 LIMITS REPORT l

i i

t OM18 3 . PDB-F-101 Page 1 1 - Last-

Attachment 1 Rev.'t 0 PY-CEI/NRR-1104 L Page i of X i 4

PIPP 100. $767 Afv.12/99

, 1.1 ,,,

III

l III l j' Ii1 !

III

! III III l a III

j. III l

i III

. 1.0 - -

iii r ,

j III JF '

IIl 2 r l 1II E ,

III J' III JP ;;

III .F .

f' '! I1I E I a

Ill J' i

4 0.9 = -

'll III E f ;

j i III J' i

!. III JF .I

!. III ,F i III E .

4 44 II] J' i l 'O III JF i

( III >F !

& 0.8 = -

111/ .i

' (' 1IF .l 2 I WI Im I .'

EII j

miI .

r WII J' IRI I -

jr gg ] l 0.7 = :

A'

/ l'l I IRI l

JF l IE

,F IIR 5 IIE <

> r

III' i I

1 0,6 = -

j - MACf = MIN'(1.0, 0.4574 4 0.006758F) !

- i 0.5 ; .'. i i i 0 20 40~ 60 80. :100 120 l CORE FLOW (%' RATED), F i FLOW DEPENDENT MAPLHGR FACTOR l (MAPFACf ) j

.. g CORE OPERATING PERRY - UNIT 1 LIMITS REPORT '

t b %v+ -w y *t f w *g * *ew*-" -- *- +*-=-'Nw+'*L- = v- --=------+-,w** e--4-+w

ON188 PDB-F-102-Attachment 5 .Page 1 - Last PY-CEI/NRR-1104 L Rev.t' O Page 7 of IC PW P No. 848 hev. 12/89 1.1 1.0 = -

a ,

~

d -

A d

JF dF JP JF JP 0.9 : =

~

AV 2r i r I

Q .,

I< 0.8 - -

/

y ,..

a-6'

^

a -

0'7 ~ ~

j ', 1~0 + 0.0052;(P-100)

MAPFACp ..

Jr t .for f "

, 405 ('P 131005 ;LAll core flows f '

255 1 P 1'405 ;-Core flow F-1 505 ar 0.6 - -

. s sdP' I b

+

'b MAPFACp = O.6'+~0;002 (P-40) -

fOr q --

255 1 P 1'405.;' Core flow Fl> 505 ".-

~

0.5 , , 7 ,. .,

0 20 '40 60 80- :100 l120 CORE THERMAL POWER-(% RATED), P POWER DEPENDENT MAPLHGR-FACTOR -

(MAPFACp )

l CORE OPERATING ?

PERRY - UNIT 1 LIMITS REPORT

m "'"

g W -13.5 EXPOSURE MAPLMGR g jA I I I "# )

< z 13.0 3 i(k 0.0 11.9 i

g 12.5 1.0 12.0 C 12.1 22.2 a 12.3 12.1-

~

- 5.0

.12.07 "'

g y ,,

N .. 10.0 12.2 -

12.3 m Z 15.0 O {2 20.0 12.1

>W REGION OF 30.0 11.2 gy; OPERATION

< $ 10.0-- ' 35.0 10.6 *OE 2O

'* 40.0 9.9 hP P., "

Q - 9.0 h 0 5000 10000 15000 .20000 25000 30000 35000 40000 45000 50000 Qc. 5 n

FO ggm AVERAGE PLANAR EXPOSURE (mwd /t) dO 5 MAXIMUM AVERAGE PLANAR LINEAR BEAT Bote: Intermediate MAPLEGR F7g yg mp GENERATION RATE (MAPLBGR) YERSUS values are obtained ,gg by linear interpolation T> AVERAGE PLANAR EXPOSURE,BP8x8R between adjacent points.

0d FUEL TYPE BP85RB219 o **

HZ s E.

O E

5?

w.

g.m_

t-l -

l-l.

l

{ av .

"~" '

13.5

, exF0sune w ei. Man-mg:

X Z s 13.0 12.9 12.9 W R) @WM) 1 .w ; 12. [ x ,_g 0.0 12.0 C <g 12.5 1.0 12.2 Z Z H:

f ' BP85RB176 -

i p12.2 5.0 12.7-q <<: 12.04 *" - .;

- :g M 11 7 10.0 12.9-W z 11.5 O

15.0 12.9 ,

11.0 l jP4; ~

l o.s 20.0 12.6 W M:" 10.5- 'FERMISSIBLE 25.0 11.7 REGION OF'

>4 h 10.0_.-. OPERATION.

d2 -30.0 10.8-m,>

I Oi

D P' '

35.0, 10.2 E**"

b*

.2 <; 9*5 40.0 9.6-

.Jss 0 9 ~'

m:

X

-4 xi 9.0 53 o se e

2 0 5000 10000 15000 20000 25000 30000 35000 40000 45000 50000 ..

n

-- ~-

9

%2 t~ o jw en

/ AVERAGE PLANAR EXPOSURE (mwd /t)- "

bO Intermediate MAFLEGE iW m MAXIMUM AVERAGE PLANAR LINEAR BEAT Enlm:

ltn N GENERATION RATE (MAFLEGR) VERSUS values are obtained lok jg -

AVERAGE PLANAR EIFOSURE,.BF8x8R by linear interpolation between adjacent points. ? % li;;

F 7 @.

iH Z O FUEL TYPE BP85RB176 .

o av .,

I t,e 5,.

.,,c -n , . , .- - ,..n-~.-,-..,,-.,,-.=.-n---,- w~-.,--.-...-.,,,_-------ww

o 13.5 u.2 / EXPOSURE MAPI,MGR l

~

f ( (mwd /t) (kW/ft) ;

(, ,,

0.0 12.4-12.5 (e '

1.0 12.4 -i

- 12.0 - 5.0. 12.8 ;

3 e

M

< Zb 3 y'11.5 ( 7.0 8.0 13.2 g a g i L

11.0 c < u2 10.0 13.5-l- Z ZH

-3

< < 10.5 3 12.5 13.5 3g 10.4 -

- n. 20.0 12.5 Z 10.0

$O 25.0 - j g 9.5 35.0 10.4 !

u2 M -

> u2 9.0 ,,,a 45.0 8.9

<Z l

X y 8.5 PERMISSIBLE \_

50.0 6.9 <

p REGION OF 1 **"

l 2

~ u2 Q 8.0 OPERATION 3

\

Sh{! ~

y% 7.5 E y:r

~

I 7.0 E "8. !

6.0 , , .

__ g %

[o 6.5 D* -;

gxm t*

g 6.0 mO l gy 0 5000 10000 15000 20000 25000 30000 35000 40000 45000 50000 mx I > AVERAGE PLANAR EXPOSURE (mwd /t)

Od

~

-3 Z F7E O MAXIMUM AVERAGE PLANAR LINEAR BEAT Botes: 1. Intermediate MAFL5GR values are obtained linear

? 5.5 GENERATION RATE (MAFLBGR) VERSUS ,, ,, ,,

AVERAGE PLANAR EXPOSURE, GE8x8EB *

yats. ,,

- FUEL TTFE BS301E 2. This curve is a composite of , g_

the most limiting enriched fuel e-lattices. For lattice specific E7 4 ~ values consult Supplement 1- " g~

, to the Supplemental Reload w-Licensing Submittal.

13.5 ""'""~' ""

En.g f L EXPOSURE (mwd /t)

MAPLHCR (kW/ft) 0.0 12.6' x 12.5 < 12.6 12.9 ,12.5 1.0 -

5.0 12.9 12.0

(*

m I

zD 7.0 15.2 h 3

8.0 -

q $}11.0 10.0 -15.5

~

kk10.5 y 12.5 13.5 20.0 12.5 d "z 10.0 4

Wo O- L 25.0 11.8 i

9.5 35.0 10.A

$31-9.0 .,,, A5.0 - 8.9

'8.5 PERMISSIBLE .

.0 6.9 3

-p ! REGION OF' *5o y l 8.0 OPERATION 1 Of

,W.

mt \ --

yz;7.5 yg C 2 7.O Eg ,,,i,. q %

en Q 6.5 e r.n C m

4 W tn 6.0

In M i

] O. 5000 10000.15000 20000 25000 30000-35000 A0000 45000 50000 x

H gZ AVERAGE PLANAR EXPOSURE (mwd /t)

MAXIMUM AVERAGE PLANAk LINEAR BEAT Estas: 1.' Intermediate NAFLEGR values x:s '

7 I E.

GENERATION RATE-(NAFLBGR)-TERSUS are obtained by linear , , , , ,

AVERAGE PLANAR EIFOSURE, GE8x8ER'

Th* St

, o .,

FUEL TTFE BS301F 2. This curve is's composite of , S .'

the most limiting enriched fuel

1attices. For lattice specific E7 values consult Supplement 1 ,~

to the Supplemental Reload- " E --

Licensing Submittal.

l

.- . - _. . _ _ _ . - . _,-..u__._-........._.___-......_ . . _ _ _ _ _ _ _ _ _ . . . _ . . -

l Attachment 5 OM18: PDB-F-200 PY-CEI/NRR-1104 L Page: 1 of 1 Page y of if Rev.: 0 MINIMUN CRITICAL POWER RATIO (TS 3.2.2)

The MINIMUM CRITICAL POVER RATIO (MCPR) shall be equal to or greater than both the MCPRf and MCPR p limits at the indicated core flow, THERMAL POWER, delta T

and core average exposure compared to the End of Cycle Exposure (EOCE)** as specified in entries PDB-F-201 and PDB-F-202.

ThisdeltaTreferstotheplannedreductionofratedfeegvater temperature from nominal rated feedvater temperature (420 F), such as prolonged removal of feedvater heater (s) from service.

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

CORE' OPERATING PERRY - UNIT 1 LIMITS REPORT

Attach ent 1 ON18 : PDbF-201 PY-CEI/NRR-1104 L ..Page : 1 - last -

p g, g og g we n ). ms w. me, Rev. : 0 1.8 iii1 ii1 iiiii1 1 iiiiiiiii>>iiii,iiiiiii>>>iiiiii1.

MCPR g = (1.6134 - 0.006948F) * [1.0 + 0.0032 (40 - F))

^

'i for F-f 405-

'lt -

'l 2 P

. dP 1.7 _- -

i , 2,-

r ar I t .S'

'l 2P

' r l .d K AF 1 J P 1.6 = :

L i

k u MCPRg= MAX (1.18, 1.8134L 0.006948F)-

t for.

1.5 _

, F F 40s u '

r g t, < r Fm 1 J Q 'k N g 4 L

1m B c

e c 1.4 " '

1i ,

/

't . F 1 .I --

, l PERMIsslatE REGION OFc ::

i OPERATION ~

1.3 _- -

1.

E

't 1

RATED OPERATING.. k I'2 ; '

. LIMIT MCPR ==1.18.:

1.15 - l, .; ; l-0 20 40- 60 80- .100~ 120?

CORE FLOW (% RATED), F FLOW DEPENDENT MCPR FACTOR' (MCPRg ) -

CORE OPERATING PERRY - UNIT 1 LIMITS REPORT l

\\

Attechnent L ONIB : PDB-F-202 PY-CEI/NRR-1104 L Page : l'- Last Page d of If Rev. -0

- e m. sm k mp l lllllllllllllllllllll!!!!!!!!!

THERMAL POWER 255 $ P f 405

'2.2= - ,, : CORE FLOW > SOE.

av am a s-2.1= -

~~

e THERMAL POWER 255 $ P f 405-2.0 - -

i

CME FLW '$ M :

y tt 1.9 -' '

,*et 1 - PERMISSIBLE ~

REGION OF OPERATION-1.82 -

$1 .7 = -

E ~

.y . , , THERMAL POWER 405 ( P f 705 ; -

[

1,6 - /

m. y

m. i 1,5 = '

fi

. m

-.m

'% THERMAL POWER P > 705=

1,4 = -

{

22 BEFORE.AND AT-END OF_ CYCLE:

,, j'

.~.. All core average exposures ,

m:

1.3 " ~

__ and 0 $ AT f 100*F-and m Core flow 11'1055 :, 6

~~ AFTER END OF CYCLE:

t' 1.2 = :: Core average. exposure > EOCE 'm

and 0 $ AT f 170'F and 1.1 -

Core flow f 1055 1,0 - ,

e i- e 1 0 20 40 60 80 100 120 CORE THERMAL POWER (% RATED), P POWER DEPENDENT MCPR FACTOR (MCPRp)

CORE OPERATING PERRY - UNIT 1 LIMITS REPORT l

i Attachment 5 OM16: PDB-F-300 !

PY-CEI/NRR-1104 L Page 1 of 1 pagelfoiff Rev 0 l

l LINEAR HEAT GENERATION RATE (3.2.3) l The LINEAR HEAT GENERATION RATE (LHGR) shall not exceeds

a. 13.4 kv/ft for the following fuel types:

1. BP8 SRB 219 j

2. BP85RB176 i i

b. 14.4 kv/ft for the following fuel types:

1. BS301E

2. BS301F ,

l l

i i

1 i

l 1

1 l (

i 1

1

(

I i

i i l

l t

l l CORE OPERATING l

PERRY - UNIT 1 LIMITS REPORT l

ML20011D921

Text

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