ML20101K963

Proposed Tech Specs Revising Table 3.3.1-2, Reactor Protection Sys Instrumentation

ML20101K963
Person / Time
Site:	Perry
Issue date:	06/30/1992
From:	CENTERIOR ENERGY
To:
Shared Package
ML20101K954	List: ML20101K953 PY-CEI-NRR-1509, Proposed Tech Specs Revising Table 3.3.1-2, Reactor Protection Sys Instrumentation
References
PY-CEI-NRR-1509, NUDOCS 9207060176
Download: ML20101K963 (11)

Text

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i TABLE 4.3.1.1-1

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REACTOR PROTECTION SYSTEM INSTRUMENTATION SORVEILLANCE REQld.*EMENTS

! "O e CHANNEL OPERATIONAL c CHANNEL FUNCTIONAL CilANNEL CONDITIONS IN WHICH  !

go CALIBRATION I *)

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09

% FUNCTIONAL UNIT CHECK TEST SURVEILLANCE REQUIRED i ** " 1. Intermediate Range Monitors: ,

i . ooOf a. Neutron Flux - High S/U.S,(b) W R 2 l j i S W R 3,4,5 t i oo-  ;

@g b. Inoperative NA W MA 2,3,4,5 ias t o 2. Average Power Range Monitor

(f)

a. Neutron Flux - High, S/U,5,(b) W SA 2 l  !

Seldown S W SA 3, 5 f D. Flow Blased Simulated -

Thermal Power - High 5,0(h) y y(d)(e) g(n). ,(1)

, 1

c. Heutron Flux - High S W ;W I) , SA 1

{ d. Inoperative NA W NA 1, 2. I, 5 l l'

! 3. Reactor Vessel Steam Dome Pressure - High 5 M R III 1, 2 III-

, 4. Reactor Vessel Water Level - ,

Low, Letel 3 S M Rg g) 1, 2 >

5. Reactor Vessel Water Level - " %y

, High, Level 8 5 M g

A g) MH2 3 o 1 3

$ym

,h"

6. Main Steam Line Isolation Valve - Closure NA M R I.

@3[y p "

1 *

~H fg

M # 5 3

7. ' Mal., Steam Line Radiation -

3 i k' h+j High S M &

J -+- M 1, 2(N.

jC $j

8. Drywell Fressure - High 5 M n'SI [' 1, 2 III h K

}'  ? 9. Scram Discharge Volume Water p 3

] $ Level - High s i

a. Level Transmitter 5 M R III 1, 2, S Ik)

'b. Float Switches MA M R 1,2,5(k) i i

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TABLE 4.3.I.1-1 (Centinued)

A REACTOR PROTECTIOM SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS m" OPERATIONAL

" CHANNEL

• CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH CHELK TEST CALIBRATION SURVEILLANCE REQUIRED E FUNCTIONAL UNIT

-* 10. Turbine Stop Valve - Closure NA M R 1

11. Turbine Control Valve fast Closure Valve-Trip System 011 M R 1 Pressure - Low NA

12. Reactor Mode Switch 1,2,3,4,5 Shutdown Position NA R NA

13. Manual Scram NA M NA 1,2,3,4,5 (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during cach startup af ter entering OPERATIONAL CONDITION 2 and the IR*1 and APRM channels shall be determined to overlap for 1:*

at least 1/2 decades during each controlled shutdown, if not performed within ti.e previous 7 days.

• l (c) Deleted ,

';" (d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values 5" calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL power. Adjust the APRM channel if the absolute difference is greater than 2% oT RATED THERMAL POWER.

The provistens of Spect f tcation 4.0.4 are not appitcable provided the survet11ance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter reachino 25% of RATED THERMAL POWEP..

(e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a cal!*xated flow signal.

(f) The eRMs shall be calibrated at least once per 1000 MWD /T using the TIP system. 7 -- '

(g) Calibrate trip unit setpoint at lecst once per 31 days. d >+ <o g (h) Verify measured core flow (total core flowl to be greater than or equal to established core ficw at the ts Gh.tthe]

7$

<=

E existing loop flow (APRM % flow). y (1) This calibration shall consist of verifyingg h: i10.C;;;;.Lsimulatedthermalpowertimeconstantgf*W{N P m s

H P. (j) This function is not required to be OPERA 8LE when the reactor pressure vessel head is resoved 1

% per Specification 3.10.1. '

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2 (k) With any control rod withdrawn. Not applicable to control rods removed per P Specification 3.9.10.1 or 3.9.10.2.

g (1) This function is not required to be OPERA 8tE when Drywell Integrity is not required. Mi h (m)

TheCHANNELCALIBRATIONshallexcludetheflowreferencetransmitters,thesetransmittersshallbe) calibrated at least once per 18 months. $

fs wMkin % Iimi41 yewa4 c., m cou.

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ADMINISTRATIVE CONTROLS f'yc ?dC

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SEMIANNUAL RA0!0 ACTIVE EFFLUENT RELEASE REPORT (Continued)

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 MNUAL (00CM), pursuant to Specifications 6.13 and 6.14, i respectively, as well as any major change to Liquid, Gaseous, or Solid Radwaste Treat:nent 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 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 j 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 Commission. 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 4

6.9 1.) Core operating limits shall be established and documented in the CORE OPOOING 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 Pcwer Ratio (MCPR) for Technical Specification 3.2.2.

(3) The Linear Heat Generation Rate (LHGR) for Technical Woecificgtion t ? t L(t) The si mi de A Th..l I'.we r T

• ihe analytical metFods used to cetermdbc ust.at b Ttha'ul Sreaf e'atdn L

3. 3.;

l r T6Fe operatin mTts shall be those previously reviewed and approved by NRC in NEDE-240ll-P-A, General Electric Standard Application for Reactor Fuel. (The approved revision at the time reload analyses are performed shall be. identified in the COLA.)

The core operating limits shall be detemined 50 that all applicable limits (e.g. , fuel thernal-mechanical limits, core thermal-hydraulic limits. ECCS limits, nuclear limits such as SHUT 00WN MARGIN, and transient and accident analysis limits) of the safety analysis are m?t.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or

( supplements thereto, shall be provided upon issuance, for each reload cycle.

to the U.S. Nuclear Regulatory Comission Document Control Desk with copies to the Regional Administrator and Resident inspector.

PERRY - UNIT 1 5-21 Amendment No. 33

j P)'- c eI/AJ 8R -f ro 7 l.

1 LIMITING SAFETY SYSTEM SETTINGS Abdmut 3 k e. 5 o f f j BASES ,

4 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) i

/sverage Power Rance Monitor (Continued)

! 5% of RATED THERMAL POWER per minute and the APRM system would be more than

adequate to assure shutdown before tne power could exceed the $afety Limit.

l The 15% neutron flux trip remains active until the mode switch is placed in ,

the Run position. r l The APRM trip system is calibrated using heat balance data taken during-1 steady state conditions. Fission chambers provide the basic input to the sys- '

[ tem and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Neutron Flux-High setpoint; i.e for a power increase, the THERMAL POWER of the fuel will be less than j that indicated by the neutron flux due to the time constants of the heat trans-i fer associated with the fuel. For the Flow issed Simulated Thermal Power-High i setpoint, a time constant c' 5 o ^.! :::: d : s introduced into the flow

biased APRM in order to s!mulate the fuel thermal transient characteristics. A more conservative maximum valus is used for the flow biased setpoint as shown in Table 2.2.1-1.

Wita in fhe cotR)

The APRM setpoints were selected to provide adequate margin for the Safety

! Limits and yet allow operating margin that reduce? the possibility of unneces-j sary shutdown.

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3. Reactor Vessel Steam Dome Pressure-High i

High pressure in the nuclear system could cause a rupture to the nuclear

, system process barrier resulting in the release of fission products. A pres-

' sure increase while operating will also tend to increase the power of the

reacter by compressing voids thus adding reactivity.- The trip will quickly i reduce the neutron flux, counteracting the pressure increase. The trip set-ting is slightly higher than the operating pressure to permit no-al operation 4

without spurious trips. The setting provides for a wide margin to the maximum-allowable design pressure and takes into account the location of the pressure

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measurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power / flow conditions when the turbine control valve fast closure and turbine stop valve closure trips are bypassed. For a load rejection or turbine trip under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

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PERRY - UNIT 1 8 2-7 l

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PDB-F0001 j

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PLANT DATA BOOK ENTRY SUBMITTAL SHEET 4

5 i TITLE: CORE OPERATING LIMITS REPORT FOR THE PERRY NUCLEAR POVER PLANT, i

UNIT 1 CYCLE 4 (RELOAD 3)

PDB - F0001 /Rev . 2 EFFECTIVE DATE: _ . _ ,

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[ )f g;kI PDB-F0001 Pages 1

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J Rev.: 2 UNIT 1 CORE OPERATING L1 HITS REPORT INDEX Specification Page INTRODUCTION AND REFERENCES 3

! AVERAGE PLANAR LINEAR llEAT uCNF5 TAT 10N RATE  !.

(CORRESPONDS TO TS 3.2.1) j Figure 3.2.1-1 Flow Dependent MAPLHGR Factor (HAPFACg ) 5 Figure 3.2.1-2 Power Dependent HAPLHGR factor (MAPFACp ) 6 1 Figure 3.2.1-3 Deleted 7 Figure 3.2.1-4 HAPLHGR Verr,us Average Plant. Exposurs, Fuel Type 'oP8 SRB 176 8 Figure 3.2.1-5 HAPLllP. Versus Average Planar Exposure, Fuel Type BS301E 9 Figure 3.2.1-6 MAPLHGR Versus Average Planar Exposure, Fuel Type BS301F 10

Figure 3.2.1-7 HAPLllGR Versus Average Planar Exposure, Fuel Type GE8B-P8505320-9GZ-120H-150-T 11 Figure 3.2.1-8 HAPLilGR Versus Average Planar Exposure, Fuel Type GE8B-P8 SOB 322-7CZ-120H-150-T 12 I

Figure 3.2.1-9 MAPLilGR Versus Average Plenar Exposure, Fuel Type GE10-P8SXB306-10GZ2-120M-150-T 13 Figure 3.2.1-10 HAPLilGR Versus Average Planar Exposure, Fuel Type GE10-P8SXB306-11GZ3-120H-150-T 14 HINIMUH CRITICAL POVEk RATIO (CORRESPONDS TO TS 3.2.2) 15 Figure 3.2.2-1 Flow Dependent MCPR Limit (MCPRg )

.uel Types GE8X8EB, BP8X8R 16 Figure 3.2.2-2 PesorDependentMCPRLimit(HCPik) p

-Fuel Types GE8X8EB, BP8X8R 17 Figure 3.2.2-3 Flow Dependent MCPR Limit (MCPRf )

Fuel Type GE8X8NB-3 18 CYCLE 4 4 CORE OPERATING PERRY UNIT 1 LIMITS REPORT

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Rev.: 2 UNIT 1 CORE OPERATING LIMITS REPORT INDEX (Continu'ed}

Speciffcation gg Figure 3.2.2 4 Power Dependent MCPR Limit (HCPRp )

fuel Type GE8X8ND-3 19 LINEAR llEAT GENERATION RATE (CORRESPONDS TO TS 3.2.3) 20  ;

i Linear Heat Generation Rate of each Fuel Type j REACTOR PROTECTION SYSTEM INSTRUMENTATION 21 (CORRESPONDS TO TS 3.3.1)

Simulated Thermal Power Time Constant CYCLE 4 CORE OPERATING PERRY UNIT 1 LIMITS REPORT

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Of T Rev.s 2 INTRODUCTION AND REFERENCES INTRODUCTION This Core Operating Limits Report for PNPP Unit 1 Cycle 4 is prepared in accordance with the requirements of PNPP Technical Specification 6.9.1.9. J The core operating limits presented vere developed using NRC-approved <

methods (Reference 2). Results from the reload analyses for the General i

Electric fuel in PNPP Unit 1 for Cycle 4 are documented in References 3, 4, 5 and 6.

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

1. Average Planar Linear Heat Generation Rate (APLHGR) Limits for each fuel / lattice type, including the power and flow dependent HAPFAC curves. (Technical Specification 3/4.2.1)

2. Minimum Critical Power Ratio Operating Limit including the power and flov dependent MCPR curves. (Technical Specification 3/4.2.2)

3. Linear Heat Generation Rate (LilGR) Limit for each fuel type.

(Technical Specification 3/4.2.3)

4. The simulated thermal power time constant. (Technical Specification 3/4.3.1) e REFERENCES

1. J.R. Hall (USNRC) to H.D. Lyster (CEI). Amendment No. 33 to Facility Operating License No. NPF-58, September 13, 1990,

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

NEDE-24011-P-A-10 and NEDE-24011-P-A-10-US (US Supplement),

April 1991.

3. " Supplemental Reload Licensing Report for the Perry Nuclear Power Plant Unit 1, Reload 3, Cycle 4," GE Document 23A7147 Rev. 0 (March, 1992).

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. 0 (October 1988).

5. " Supplement 1 to the Supplemental Reload Licensing Submittal for the Perry Nuclear Power Plant Unit 1, Reload 2, Cycle 3," GE Document 23A6492AA Rev. 0 (September 1990).

CYCLE 4 CORE OPERATING PERRY UNIT 1 LIMITS REPORT

) PDB-F0001 l O il

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6. " Supplement 1 to the Supplemental Reload Licensing Submittal for the Perry Nuclear Pover Plant Unit 1, Reload 3, Cycle 4," GE Document 23A7147AA, Rev. 0 (January 1992).

7. Petty Nuclear Power Plant Updated Safety Analysis Report, Unit 1, Appendix 158-Reload Safety Analysis.

8. R. E. Parr (GE) to H. S. Rupp (CEI), PY1004R03 - Rotated Bundle Analysis, PY-GEF/CEI-439, May 13, 1992.

9. Fax transmittal from J. Vorthington (GE) to h. S. Rupp (CEI),

Reanalysis of GE10 Rotated Dundle, May 19, 1992.

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CYCLE 4 CORE OPr. RATING PERRY UNIT 1 TC/VAX/Page 1 of 2 LIMITS REPORT

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Page: 21-Last g  !

f km Rev.: 2 REACTOR PROTECTION SYSTEM INSTRUMENTATION (TS 3.3.1)

The simulated thermal power fline constant shall be 610.6 seconds.

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l CTCLE 4 CORE OPERATING PERRY UNIT 1 LIMITS REPORT

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