PY-CEI-NRR-0496, Suppls & Provides List of Addl Changes to Be Included in Full Power OL Tech Specs.Changes Do Not Affect Ability to Safely Operate Plant Under Current License. Justifications & marked-up Tech Specs Encl
| ML20203E113 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 07/18/1986 |
| From: | Edelman M CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| PY-CEI-NRR-0496, PY-CEI-NRR-496, NUDOCS 8607240053 | |
| Download: ML20203E113 (44) | |
Text
{{#Wiki_filter:1 f THE CLEVELAND ELECTR P.o. Box 5000 - CLEVELAND CHIO 44101 - TELEPHONE (216) 622-9800 - lLLUMINATING BLOG - 55 PUBLICSGUARE Serving The Best Location in the Nation MURRAY R. EDELMAN July 18, 1986 [uc A PY-CEI/NRR-0496 L Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Perry Nuclear Power Plant Docket No. 50-440 Additional Changes to Technical Specifications for Full Power Licensing
Dear Mr. Denton:
This letter supplements our letter dated June 18, 1986 and provides a list of additional five changes that the Cleveland Electric Illuminating Company requests be included in tha Technical Specifications which will accompany the full power operating licenue for the Perry Nuclear Power Plant - Unit 1. These items represent clarification and enhancements to the Technical Specifications. The justifications and the proposed markup pages are attached. None of these changes affect CEI's ability to safely operate the Perry Nuclear Power Plant-Unit I under its current license. Thus, no amendment of the present low power license is being requested. If you have any questions, please call me. Very truly, ur , d6 Murray R. Ed man Senior Vice President Nuclear Group Attachments MRE:njc cc: Jay Silberg, Esq R. Vollmer J. Keppler John Stefano (2) R. Bernero C. Norelius J. Grobe W. Butler C. Paperiello 8607240053 860718 PDR ADOCK 05000440 g P PDR y
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Attachment PY-CEI/NRR-0496 L Justification Technical Specification Definition 1.7 Core Alteration The Technical Specification definition of Core Alteration presently does not consider normal movement of the SRM's, IRM's, TIP's, or special moveable detectors a core alteration. This change request would provide the same consideration for LPRM's. The Technical Specification definition of core alteration provides specific exception for the movement of incore instrumentation. Since the LPRM strings are only removed from the core for replacement, and have no normal drive mechanism, a similar exception from the definition of core alteration is requested. Presently, Technical Specification 3.3.1, Reactor Protection System Instrumentation, contains a footnote which does not consider replacement of an LPRM string a core alteration. This change request would clarify the present definition and modify it to include an exception presently contained in the Technical Specifications. l l
- DEFINITIONS CORE ALTERATION L 9 R F\ s;
- 1. 7 CORE ALTE hall be/the addition, removal, relocation or movement of fuel, sources, re instrupentijor reactivity controls within the reactor i pressure vessel with the ve movement of the SRMs, IRMs,fsel Normal TIPS, nead or special removed movableand fuel in the is detectors vessel.
not considered 4 a CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude comple-tion of the movement of a component to a safe conservative position. CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY
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1.8 Thevalue highest COR'E of MAXIMUM the FLP0 which FRACTION exists inOF theLIMITINGcore. POWER DENSITY (CMFLPD) s CRITICAL POWER RATIO 1.9 The CRITICAL POWER RATIO (CPR) shall be the ratie of that power in the assembly which is calculated by application of the GEXL 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 1 i 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 j conditions are either: i 1. Capable of being closed by an OPERABLE automatic isolation i system, or
- 2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, I
except as provided in Table 3.6.4-1 of Specification 3.6.4.
- 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. l
- e. The drywell leakage rates are within the limits of l ,
Specification 3.6.2.2. l PERRY - UNIT 1 1-2
Attachment PY-CEI/NRR-0496 L Justification Technical Specification 4.5.1. Emergency Core Cooling Systems - Operating Surveillance Requirement 4.5.1.e.2.c presently lists the low pressure alarm system setpoint for the ADS as 2475 + 25 psig. This change request would change this setpoint to greater than or equal to 155 psig. Perry presently has a high pressure air system (Safety Related Instrument Air, P57), which supplies the ADS accumulators. This high pressure air is fed through a pressure regulator to reduce it to the nominal operating pressure of the ADS (150 psig). The plant is currently installing a design change which would utilize a low pressure air supply system for the ADS system. The details of this design modification are described in our letter dated July 10, 1986 (PY-CEI/NRR-0497L). This Technical Specification change request would reflect the new design setpoint for the P57 system. This design change will be implemented concurrently with receipt of the full power Technical Specifications. I i
EMERGENCY CORE COOLING SYSTEMS (<- SURVEILLANCE REQUIREMENTS (Continued)
- e. For the ADS by:
- 1. At least once per 31 days, performing a CHANNEL FUNCTIONAL. TEST of the safety related instrument air system low pressure alarm system.
- 2. At least once per 18 months:
a) Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation. b) Manually opening each ADS valve when the reactor steam dome pressure is greater than or equal to 100 psig* and observing that either: -
- 1) The control valve or bypass valve position responds accordingly, or
- 2) There is a corresponding change in the measured steah flow, or
- 3) The safety relief valve discharge pressure switch
( indicates the valve is open. c) Performing a CHANNEL CALIBRATION of the safety related instrument air system low pressure alarm system and verifying an alarm setpoint of 2?70 1 20 psig on decreasing pressure. > jj 3-
*The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.
i l l i l PERRY - UNIT 1 3/4 5-5 w- <y a - - - e--,- - , , , - . , - - , - - - -
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. 1rY-CEI/NRR-0496 L
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' l 4 Justification Technical Specification 3.8.2.1
. r. Technical Specification 3.8.2.2 f D.C. Sources A-The Technical Specifications presently place minimum requirements on the D.C. electrical power sources." This change request would allow credit'to be taken for the Unit 2 batteries to' meet the Limiting Condition for Oper[ tion (LCO). Perry FSAR Section 8.3.2.1.2.1 presently describes the . ~ Class 1E di, visional 125 , Volt D.C. system. The discuasion describes the use"of the maintenance tie bus circuit breakers. ThesebreakerEarenormallyopenbutmaybec1bsedunder
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administrative control to allow maintenance or equalizing a battery. IEEE Standard 308-1980, Section 8.5 states that it is permissible t'o provide t inter-unit ties between the Class 1E buses of the units in a multi-usit a station, provided any single component failure does not degrade the'$.fiss 1E power systems of any unit below an accsptable level add provides that the independence of the redundant systems are maintained. The'cifcuit, design meets these criteria and is such that the Unit 2 batteries may be powered from the respective divisional Unit 1 battery chargers. - This change request would allow credit to be toden for the Unit 2 batteries explicitly to meet the LCO for D.C. sources. s. O c P
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ELECTRICAL POWER SYSTEMS 3/4.8.2 D.C. SOURCES D.C. SOURCES - OPERATING - LIMITING CONDITION FOR OPERATION - 3.8.2.1 As a minimum, the following D.C. electrical power sources shall be OPERABLE:
- a. Division 1, consisting of:
1.- 125 volt battery 1R42-5002x oc 2. R 4 2. - f 002. .
- 2. 125 volt full capacity charger 1R42-5006 or 1R42-5007.
- b. Division 2, consisting of:
- 1. 125 volt battery 1R42-S0034 ee 2 A. 41 - S o o 3 .
- 2. 125 volt full capacity charger 1R42 ,5008 or 1R42-5009.
- c. Division 3, consisting of: '
- 1. 125 volt battery 1E22-5005x ,c 2. E 2. 2. - S o o s~,
- 2. 125 volt full capacity charger IE22-5006 or 1R42-5011.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. ACTION: ( ,w
- a. With eithe ivision 1 ba
-- Wchargers
^
ry and/or bo r~~or Di ision 2 battery d/or both ch ers of the ab e required D.C. ectrical power urces inoper e, restore th inoperable divi on battery t OPE LE status wi n 2 hours or e in at least H0 SHUTDOWN wit nl next 12 hour and in COLD S OWN within th ollowing 24 urs.)
- b. )
l With Divisio battery a r both chargers f the above r utred D.C. elect cal power sou es inoperable, clare the HP systeminj) operable take the AC ION required by Specification . 5.1. / d SURVEILLANCE REQUIREMENTS LN.CEAT C 4.8.2.1 Each of the above required 125 volt batteries and chargers shall be demonstrated OPERABLE:
- a. At least once per 7 days by verifying that:
- 1. The parameters in Table 4.8.2.1-1 meet the Category A limits, i
and
- 2. Total battery teminal voltage is greater than or e' qual to 129 volts on float charge. ~
l lC PERRY - UNIT 1 3/4 8-12
INSERT C
- a. With the Unit 1 and the Unit 2 Division 1 batteries and/or both chargers of the above required Division 1 D.C. electrical power sources inoperable, restore an inoperable Division 1 battery or charger to OPERABLE status within 2 hours or be in at least HOT SHUTDOWN within the.next 12 hours and in COLD SHUTDOWN within the following 24 hours,
- b. With the Unit I and Unit 2 Division 2 batteries and/or both chargers of the above required Division 2 D.C. electrical power sources inoperable, restore an inoperable Division 2 battery or charger to OPERABLE status within 2 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
. t
- c. With the Unit 1 and Unit 2 Division 3 batteries and/or both chargers of the above required Division 3 D.C. electrical power sources inoperable, declare the HPCS system inoperable and take the ACTION required by Specification 3.5.1.
1
ELECTRICAL POWER SYSTEMS D.C. SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, Division 1 or Division 2, and, when the HPCS system
.is required to be OPERABLE, Division 3, of the D.C. electrical power sources shall be OPERA 8LE with:
- a. Division I consisting of:
- 1. 125 volt battery 1R42-5002, or- 2 M 2.- f o o 2 -
- 2. 125 volt full capacity charger 1R42-5006 or 1R42-5007.
- b. Division 2 consisting of:
- 1. 125 volt battery 1R42-5003, ., 2 R%- Coo 7,
- 2. 125 volt full capacity charger 1R42-S008 or 1R42-5009.
- c. Division 3 consisting of:
- 1. 125 volt battery 1E22-5005, e 2 E 2.2. - roon
- 2. 125 volt full capacity charger 1E22-5006 or 1R42-5011.
APPLICABILITY: OPERATIONAL CONDITIONS 4, 5 and 8 - ACTION: Insecf @
, g , vv m c.~<- m ^ r r DW f' a. -With-both-94visica i bettery-and/or-both-chari;;r: : d-Division-2 '
-battery and/or-both-chargers-of-the-above-required-0.CWeetricalh
-power- sources-inoperable, :::p;nd CORE-ALTERATIONSrbandling-of j
-irradiated-fuel-in-the-primary-containment-and-operatica: with-a
-potential -fee-draining th; reecter ve:sel.
%%v v" A m g :
dn b. Withl Division 3 batte[y and/or both chargers of the above required 6 l[1 j f/;6 D.C. electrical power sources inoperable, declare the HPCS system inoperable and take the ACTION required by Specification 3.5.2 and gg4A 3.5.3. ' kbf A c. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENTS 4.8.2.2 Each of the above required battery and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.1.
*WhenhandlingirradiatedfuelintheFuelHandlingBuildingorpfimary i containment.
l l t PERRY - UNIT 1 3/4 8-16 l
INSERI D
- a. With the Unit I and the Unit 2 Division 1 batteries and/or both chargers of the above required Division 1 D.C. electrical power sources and the Unit 1 and Unit 2 Division 2 batteries and/or both chargers of the above required Division 2 D.C. electrical power sources inoperable, suspend CORE ALTERATIONS, handling of irradiated fuel in the fuel handling building or primary containment and operations with a potential for draining the reactor vessel.
l
N Attachnant PY-CEI/NRR-0496 L Justification
. Technical Specification Surveillance 4.11.1.3 Liquid Effluents The Technical Specifications do not presently identify the specific surveillance requirements associated with the continuous releases of radioactive liquid effluents as described in Table 4.11.1.1.1-1. This change request would add Surveillance Requirement 4.11.1.1.3, which would specifically address the sampling requirements of continuous radioactive liquid effluents.
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l l i f f 3/4.11 RADI0 ACTIVE EFFLUENTS ( - 3/4.11.1 LIQUID EFFLUENTS CONCENTRA'T ION LIMITING CONDITION FOR OPERATION I l 3.11.1.1 The concentration of radioactive material released in liquid i effluents to UNRESTRICTED AREAS (see Figure 5.1.1-1) shall be limited to ' the concentra.tiors specified in 10 CFR Part 20, Appendix 8, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 4 microcuries/m1 total activity. APPLICA8ILITY: At all times. ACTION: With the concentration of radioactive material released in liquid effluents , to UNRESTRICTED AREAS exceeding the above limits, immediately restore the I concentration to within the above limits. ( SURVEILLANCE REQUIREMENTS 4.11.1.1.1 The radioactivity content of each batch of radioactive liquid
, waste shall be determined prior to release by sampling and analysis in accord-1 ance with Table 4.11.1.1.1-1. The results of pre-release analyses shall be used with the calculational methods in the 00CM to assure that the concentration at the point of release is maintained within the limits of Specification 3.11.1.1.
l 4.11.1.1.2 Post-release analyses of samples composited from batch releases shall be performed in accordance with Table 4.11.1.1.1-1. The results'of the radioactivity analysis shall be used in accordance with the methodology
- and parameters in the 00CM to assure that the concentrations at the point of l release are maintained within the limits of Specification 3.11.1.1.
3 4.11.1.1./ Continuous releases of radioactive liquid effluents shall be sampled and analyzed in accordance with Table 4.11.1.1.1-1. The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1. l PERRY - UNIT 1 , 3/4 11-1
Attachment PY-CEI/NRR-0496 L Justification Technical Specifications pages vi xviii 2-4 B 2-7 B 2-9 B 2-10 3/4 2-1 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-9 3/4 2-10 3/4 3-5 3/4 3-8 3/4 3-46 3/4 3-58 B 3/4 2-1 B 3/4 2 B 3/4 2-3 B 3/4 2-4 B 3/4 2-5 B 3/4 2-6 The changes to the above are necessary to support the operation of the plant in the Maximum Extended Operating Domain (MEOD). The MEOD analysis for Perry is contained in Appendix 15E to FSAR Chapter 15, with additional information supplied to the staff in Amendment 25 transmitted by letter (PY-CEI/NRR-0484 L, dated June 27, 1986). D16/48
( LINITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE POWER DISTRIBUTION LIMITS (Continued) Figure 3.2.1-3 Maximum Average Planar Linear Heat' Generation Rate (MAPLNGR) Versus Average Planar Exposure Initial Core Fuel Types P85R8071. . . . . . . . . . . 3/4 2-4 ar ,... nrnn a6 rvania.......................................... ai ,6 a - 3/4 2.g MINIMUM CRITICAL POWER
- 2. RATI0............................ 3/4 2-p #7 z
Figure 3.2.g-1 MCPR f
.............................. 3/4 2-/ 6 2 -
3 Figure 3.2.f-2 MCPR P
.............................. 3/4 2-g 'I 3/4.2.A LINEAR HEAT GENERATION RATE............................. 3/42-f10 3/4.3 INSTRW4ENTATION 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 I Table 3.3.1-2 Reactor Protection S Response Times......ystem
................ 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 i
Table 3.3.2-1 Isolation Actuation Instrumentation..................... 3/4 3-11 Table 3.3.2-2 Isolation Actuation Instrumentation Setpoints........... 3/4 3-17 Table 3.3.2-3 Isolation System Instrumen-tation Response Time................ 3/4 3-21 Table 4.3.2.1-1 Isolation Actuation Instrumen- ' tation Surveillance Requirements...................... 3/4 3-23 F19 ec 3. 3.1 - 1 MRPR%; . 2 .C ( hyan 3. 2. / - s' M/3PFAGp 3/4 2 ~6 4 PERRY - UNIT 1 vi 4
l 8ASES SECTION PAGE 3/4.0 APPLICA8ILITY........................................... 8 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN...................................... 8 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES................................. 8 3/4 1-1 3/4.1.3
, CONTROL R005.......................................... B 3/4 1-2 3/4.1.4 CONTROL R00 PROGRAM CONTR0LS......................... 8 3/4 1-3 l
3/4.1.5 STAND 8Y LIQUID CONTROL SYSTEM........................ 8 3/4 1-4 3/4.2 POWER DISTRI8dTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION t RATE................................................. 8 3/4 2-1 l Bases Table 8 3/4 2.1-1 Significant Input Para-I . meters to the Loss-Of-(. Cooling Accident Analysis............... 8 3/4 2-3 1/w.s.s nr nn ac u ruIttis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 2 3/4.2.J' MINIMUM CRITICAL POWER RATI0. . . . . . . . . . . . . . . . . . . . . . . . . 8 3/4 2-4
- 2. '2 Bases Figure 8 3/4 2.y-1 Power to Flow o Map............perating ....... 8 3/4 2-6 3/4.2.y LINEAR HEAT GENERATION RATE.......................... 8 3/4 2-5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENT.' ,.,N............. B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.................. 8 3/4 3-2 3/4.3.3 EMERGENCYCORECOOLINGSiYSTEMACTUATION i
INSTRUMENTATION...................................... 8 3/4 3-2 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.... 8 3/4 3-3 , 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION...................................... 8 3/4 3-4 ( 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION.................... 8 3/4 3-4 . PERRY - UNIT 1 , xviij i I
' - - - --..,_.ew, ,,,_ n ,n. _ _n,- - - . . _ __,.n.. . , , . . . , _ , . _ _ _ . - _ - . _ _ _ . _ _ . , . . - - - .- -.
.n a m TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS E
Q
, FUNCTIONAL UNIT ALLOWABLE TRIP SETPOINT VALUES c
- 1. Intermediate Range' Monitor
- a. Neutron Flux-High H < 120/125 divisions < 122/125 divisions
- b. Inoperative of full scale of full scale NA NA
- 2. Average Power Range Monitor:
- a. Neutron Flux-High Setdown < 15% of RATED < 20% of RATED
- b. Flow Blased Simulated Thermal Power-High THERMAL POWER Np o THERMAL POWER 08
- 1) Flow Biased 1 0.66 W+ M with 1 0.66 M , with a maximum of a maximum of
- 2) High Flow Clamped < 111.0% of RATED < 113.0% of RATED THERMAL POWER THERMAL POWER
- c. Neutron Flux-High < 118.0% of RATED < 120.0% of RATED m THERMAL POWER 1 d. Inoperative NA THERMAL POWER NA
- 3. Reactor Vessel Steam Dome Pressure - High 1 1064.7 psig i 1079.7 psig
- 4. Reactor Vessel Water Level - Low, level 3 > 177.7 inches above > 177.1 inches above Top of active fuel
- Iop of active fuel *
- 5. Reactor Vessel Water Level-High, Level 8 < 219.5 inches above < 220.1 inches above Top of active fuel
- Top of active fue1*
- 6. Main Steam Line Isolation Valve - Closure 1 8% closed i 12% closed
- 7. Main Steam Line Radiation - High < 3.0 x full power < 3.6 x full power Eackgr.ound Eackground
- 8. Drywell Pressure - High 1 1.68 psig i 1.88 psig "See Bases Figure B 3/4 3-1.
ur ng a o am, t e APRM rip setpoin and allowa le value may e pers tied a be increased to , the MEOD values (trip setpoint of 10.66 W + 64% and allowable value of 10.66 W + 67%). '
i i LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) Average Power Rance Monitor (Continued) ~ 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit. The 1S% neutron flux trip remains active until the mode switch is placed in the Run position. The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the sys-4
- 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
1.e.. for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neut'ron flux due to the time constants of the heat trans-for associated with the fuel. For the Flow Biased Simulated Thermal Power-High setpoint, a time constant of 6
- 0.6 seconds is introduced into the flow biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1.
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!( The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow _ margin that reduces the )ossibility of unneces-sa shutdown. ow erfced tVs@ofnt' must ,5e'adjutttid 'tiy't'he o la in Specification 3.2.2 in order to maintain these margins # when MFLPD is greater than or equal to FRTP. I
- 3. Reactor Vessel Steam Dome Pressure-Hiah l
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 reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip set-ting is slightly higher than the operating pressure to pemit normal operation 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 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 B 2-7
( LIMITING SAFETY SYSTEM SETTINGS - 8ASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
- 8. Drywell Pressure-Hich
- High pressure in the drywell could indicate a break in the primary pressure boundary systems. The reactor is tripped in order to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant and to the primary containment. The trip setting was selected as low as possible without causing spurious trips.
- 9. Scram Discharge Volume Water Level-Hich The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram. Should this volume fill up to a point where there is insufficient volume to accept the displaced water at pressures below 65 psig, control rod insertion would be hindered. The reac-tor is therefore tripped when the water level has reached a point high enough
' to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods at pressures below 3 I 65 psig when they are tripped. The trip setpoint for each scram discharge volume is equivalent to a contained volume of approximately 24 gallons of water. {
- 10. Turbine Stop Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutron
' flux, and heat flux increases that would result from closure of the stop ' valves. With a trip setting of 5% of valve closure from full open, the resul-
, tant increase in heat flux is such that adequate thermal margins are maintained t
during the worst case transient. Jfgerya /)
- 11. Turbine Control Valve Fast Closure. Trip 011 Pressure-Low The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the I
turbine control valves due to load rejection with or without coincident failure of the l turbine bypass valves. The Reactor Protection System initiates a trip when - fast closure of the control valves is initiated by the fast acting solenoid l valves and in less than 20 milliseconds after the start of control valve fast closure. This is achieved by the action of the fast acting solenoid
- valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves. This loss of pressure is sensed by j pressure switches whose contacts fem the one-out-of-two twice logic input to i
the Reactor Protection System. This trip setting, a slower closure time, and a different valve characteristic from that of the turbine stop valve, combine to produce transients which are very similar to that for the stop valve. Relevant i transient analyses are discussed in Section 15.2.2 of the Final Safety Analysis , Report. g g g I PERRY - UNIT 1 , 8 2-9 f
- _ - - - - - - _ . . - . _ ~ . . _ . _ _ _ _ _ - - _ ~ ~ _ _ , _ _ _ .~m.-__ _ . -
LIMITING SAFETY SYSTEM SETTING 8ASES REACTOR PROTECTION SYSTEM INSTRUNENTATION SETPOINTS (Continued)
- 12. Reactor Mode Switch Shutdown Position The reactor mode switch Shutdown position provides additional manual reactor trip capability.
- 13. Manual Scram The Manual Scram provides manual reactor trip capability. The manual scram function is composed of four push button switches in a one-out-of-two taken twice logic.
1 I i I 9 ( . FERRY - UNIT 1 . 8 2-10
4 ( POWER DISTRIBUTION LIMITS 3/4.2 POWER OISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE l LIMITING CON 0! TION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type ' of fuel as.a. function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3) + l APPLICA8ILITY: OPERATIONAL CONDITION 1, when THERMAL POWER Is greater than or equal to 25% of RATED THERMAL POWER. l ACTION: l 1 With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, or 3.2.1-3, initiate corrective action within 15 minutes and restore APLNGR to within { ! the required limits within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. i as multiplied by the smaller of either the flow dependent MAPLHGR factor ( (MAPFAC f
) of Figure 3.2.1-4 or the power dependent MAPLHGR factor (MAPFAC )
of Figure 3.2.1-5. p
' SURVEILLANCE REQUIREMENTS i
4.2.1 All APLNGRs shall be verified to.be equal to or less than the limits
<detershied-from-Figures-3.2.1=1, 3.2.1-2, ent. 3.2.1-3 4 l
- a. At least once per 24 hours,
- b. Within 12 hours 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 hours when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR.
- d. The provisions of Specification 4.0.4 are not applicable.
e
- (
l { PERRY - UNIT 1 . 3/4 2-1
--..v-..--.-_----,.-. . _ - - ,- - --- - m m _ -,-.,--w-
INSERT A As indicated in Table 3.3.1-1, this function is automatically bypassed below the turbine first stage pressure value equivalent to thermal power less than 40% of RATED THERMAL POWER. The automatic bvpass setpoint is temperature dependent due to the subcooling
- changes that affect the turbine first stage pressure - reactor power relationship.
For RATED THERMAL PCWER operation with feedwater temperature greater than or equal to 420 F, an allowable setpoint of L26.9% of control valve wide open turbine first stage pressure is provided for the bypass function. This se. point is also applicable to operation at less than RATED THERMAL POWER w'th the correspondingly lower feedwater temperature. The allowable setpoint is reduced to [22.5%, [19.5%, and 116.5% of control valve wide open turbine first stage pressure for RATED THERMAL POWER operation with a feedwater temperature between 370 F and 420 F; 370 F and 320 F, and 320 F and 250 F, respectively. Similarly, the reduced setpoint is applicable to operation at less than RATED THERMAL POWER with the correspondingly lower feedwater temperature. i INSERT B f As with the Turbine Stop Valve-Closure, this function is also bypassed below l 40% of RATED THERMAL POWER. The basis for the setpoint is identical to that described for the Turbine Stop Valve-Closure. i I _ , - . - - , _ , ~ , - - , _ , _ . - - . _ , _ _ - . - _ . _ . . _ _ _ , . . . . . . _ - _ _ _ . - - _ _ _ . _ . . . _ . _ _ _ _ _ . - - . - - - - . - . . - . _ _ - , .
*Ie=_km> m.mm__e 1.0 s
}
~
R .__. :: _ :: g09 ___ _ ___ . --- _j_ R 1:EE __/_lZ y __
.___ __ p;_ -- ___ .____
'~~~~~~ ~~-~'- ~---- -- ~~~--
0.8 t u S _____ _
- _ _l:::_ .
V..-..__
~ ~ '
y jl. TsQ g,, k f' % ( W
~~---
p^d::: __ ___ N;--- . ___.__.._ __ . . 12
--\ MAPFACr = MIN (1.0, 0.4574
' ---- ~ - - --
~~ '- - -
+ 0.006758F) _
0.6 _.___._ . ____
- --{ __ __
___ ::::: :g____ ___ 0.5 ___ __ _____ _____ iisis._ 0.4 0 20 40 60 80 100 120 CORE FLOW (% RATED), F fija< e 3. 2. I - 4 fcr<y - hl/ / 3l4 2 ~ 6"
l.1 1.0 ; r p/ r f O 0.9 / u\ .____ j
/ _
[ b__y_ j_._ _. D
~~~~
0.8 jf t / 100%; All Core Flows 4o ----- -
--_/-
j
-. For 40% <-PP ' 40%; Core Flow F ~ 50 %
- MAPFACp For 25% = 1.0 + 0.0052 (P-100)
- ~
Q j..::- : .
---~-
g 0.7 f _ i g j
- g[_:: .:::_:_-
g -~~~
..__/.-
--g .-__.__
0.6 __ ___
~~ -
For 25% P 40%; Core Flow F >50% 2:_('T MAPFACp = 0.6 + 0.002 (P-40)
~ -~
0.5 ._____
----- ~----
___ _ :- ::: ~~---
- ~--- ----
.__ T:: :::: .. ____ ::::: ::::: _
0.4
'~~~~- '
0 20 40 60 80 100 120 CORE THERMAL POWER (% RATED), P f}} urc 7 ./ / - S~ l0,.,.y c . Mas } } T/4 2 - G
- VI) >
f\"
#OA m
POWER DISTRIBUTION LIMITS'
/
3/4.2.2 APRM SETPOINTS j LIMITINGCONDITIONF0dOPERATION /
/
3.2.2 The APRM f biased simulated the 1 power-high scram trip setpoint (S) and flow bia d neutron flux-upscale shall be estab1 hed according to the fo owing ntrol relationships: rod block trip setp int (SRB) TRIP SETPOINT ALLOWABLE VALUE S 1 (0.66W + 485*)T 51 (0.66W + S R8 1 (0.66W + 42%*)T S R8 1 (0.66W45%*)T
)51%*)T where: and 5 are in percent pf RATED THERMAL POWER,
= Loch 8 recirculation ow as a percentage of the oop recirculation flow which produce a rated core flow of 104.. million 1bs/hr.
' T = Lowest value of t ratio of FRACTION OF RA (O THERMAL POWER (FRTP) divided by the C E MAXIMUM FRACTION OF LI TING POWER DENSITY (CMFLPD). T is applied o if less than or equal 1. 0.
^
APPLICABILITY: OPERATIO 'L CONDITION 1, when THE L POWER is greater than or
; equal to 25% of RATED RMAL POWER.
j ACTION: ' With the APRM flow _ ased simulated thermal r-high scram trip setpoi and/or the flow biased ne ron flux-upscale control rod block trip setpoint I s conser- ) , vative than the v ue shown in the Allowab ValuecolumnforSorSg,asabove ( determined, int ate corrective action w in 15 minutes and adjust S and/or S R8 to be consist t with the Trip Setpoin value** within 6 hours or reduce THERMAL [ POWER to les than 25% of RATED THE L POWER within the next 4 ours.
/
- SURVEILLANCE REQUIREMENTS l 4.2.2 The FRTP and CMFLPD sha e determined, the val of T calcu-lated, and the most recent ac al APRM flow biased sin ted thermal power-high scram and flow biased neutro flux-upscale control ro lock trip setpoints verified to be within the ove limits or adjusted, s required
- a. At least once r 24 hours,
- b. Within 12 hours after completion of af ERMAL POWER increase of t least 15% o RATED THERMAL POWER in ne hour, and
- c. Initially and at least once per ours when the reactor is perat-ing wit CMFLPD greater than or . ual to FRTP.
- d. The p visions of Specificatio 4.0.4 are not appifcable.
During the ftartup test program, th, APRM trip setpoint and al,lowable values "may be pe tted to be increased t,d the ME00 values (S trip 'petpoint of ( $0.66W 4% and allowable valu of < 0.66W + 67%, and S R8/t rip setpoint of
<0.66W 58% and allowable val of 7 0.66W + 61%).
**9tth CHFLPD greater than the FRTP, rather than adjustin he APRM setpoints, I the APRM gain may be adjusted such that the APRM scale / readings are greater i
( than or equal' to 100% times CMFLPD provided that the,4djusted APRM scale read-ing does not exceed 100% of RATED THERMAL POWER, and a notice of the adjustmen is posted on the reactor control panel. N PERRY - UNIT 1 3/4 2-5 1
{ POWER DISTRIBUTION LIMITS 2
- 3/4.2.X MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 2
' 3.2 7 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than both MCPR f and MCPR, limits at indicated core flow, THERMAL POWER AT* and l core average exposure. compared to End of Cycle Exposure (E0CE)** as shown in Figures 3.2.J-1 and 3.2 7 2.
- 2. 2 APPLICA8ILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 253 of RATED THERMAL POWER. '
ACTION: g
! With 2 R 1ess than the applicable MCPR limit 3.2. 2, initiate corrective action within 15 minutes'and restore MCPR shown in Figures 3.2.f i and I to ' within the required limit within 2 hours or reduce THERMAL POWER to less than i
25% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS i 2 ' 4.2 7 MCPR shall be determined to be equal to or greater than the MCPR limit determined from Figures 3.2.X-1 and 3.2 7 2: ( 2. 2 { a. At lea,st once per 24 hours,
- b. Within 12 hours 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 hours when the reactor is operating l with a LIMITING CONTROL R00 PATTERN for MCPR. I
- 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 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 i
heaters in service and equilibrium Xenon, or 2) as specified by the fuel vendor. PERRY - UNIT 1' 3/42-/ r?' '
l w" -
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20' 40 /' 60 80/ - 100 s f CORE - FLOW (% OF RATED) ., - MCPR f i Figure 3.2./-1 2 i J PERRY - UNIT 1 3/42.f G
m 1.7
-__pum _
1.6 3%--- _ :MCPRf = MAX (1.18, 1.8134-0.006948F) :
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-~~-- '----- - -
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- __.___ :::: . :::: :::Z- ::::- _ ._____ :::Z-
---- '---- '~-'-- - ~ ~---
1.0 0 20 40 60 80 100 120 CORE FLOW (% RATED), F MCP&q Rge v.z.z-t
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v , s' , ,' , . 8-p,- m ,. ia _ ge i i + ,
.p , A5 core a rage / /
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..i. s , , i .m :w. .x , s e / J i ex .
es / u-C, - - w m w w. e
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/
0 60 80 100 120 CORE POWER (% RATED)
"These MCPR limits assume that the TCV/TSV scram bypass setpoint (see note (h) of Table 3 P3.1-1) and EOC-RPT bypass setpoint (see note (b) of Table 3.3.4.2-1) are consistent with corresponding steam flow. For At>0, the trip setpoint shall be conservatively reduced from <25.4% of calibrated span on increasing turbine first stage pressure to <15%.~ The allowable value shall'be consistently reduced from _<26.9 f of calibrated span to _<16.5%.
MCPR P ( Figure 3.2.g-2 2 PERRY - UNIT 1 3/42-f 7 1
l1i
'L .
THERMAL POWER 25% sP s40% ~~-
. CORE FLOW > 50%
/
] _____
2.2 f_' ____ _
- - ~
THERMAL POWER 25% sP s40% __ _\ , . CORE FLOW s50% 2.0 h l V, 1 i _____ l.8 ----~ ~~~ -- ( _____ _____ g _____ ____._
A_A' l.6 THERMAL POWER 40%
70% A-A' Core average exposure > EOCE and :_- ;b 100' F < AT s 170' F and M M-H- - - -- - - - - - Core flow s 105%. :: hs A-A' -- -- C-C'N ~ k ~-'-~ - B-B' All core average exposures and 1.2 (50' F <aT s 100' F and core flow s 100%) or -- - -- -N B-B' (AT 5100 F and core flow > 100% and 5105%). :- ____._ C-C' All core average exposures and __ ._____ _,_,___ _____ (AT 5 50 F and core flow s 100%) or -- ~~ ~-~~~ ~~~~- ------ ZZZ-(AT = 0' F and core flow > 100% and 5105%) _ _ .__ ___ 0 20 40 60 80 100 120 CORE THERMAL POWER (% RATED), P M C f) /d p f9ec 7<.2-L & y - 46n '/ .2 $ E'9 POWER DISTRIBUTION LIMIT i ( 3 ' 3/4.2.# LINEAR HEAT GENERATION RATE l
- LIMITING CONDITION FOR OPERATION
\ 3 i 3.2 # The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed 13'.4 kw/ft.
- APPLICA8ILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or i 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 lieft within 2 hours or ' - reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next I 4 hours. . I l I SURVEILLANCE RCOUIREMENTS i (' J l 4.2.g LNGR's shall be determined to be equal to or less than the limit: i .. .it least once per 24 hours,
- b. Within 12 hours efter completion of a THERMAL POWER increase of at least ISE of RATED THERMAL POWER in one hour, and
- c. Initially and at least once per 12 hours when the reactor is operating on a LIMITING CONTROL R00 PATTERN for LHGR.
- d. The provisions of Specification 4.0.4 are not appitcable.
( - PERRY - UNIT 1 , 3/42-//0 0 TA8LE 3.3.1-1 (Continued) ( ' REACTOR PROTECTION SYSTEM INSTRUMENTATION TA8LE NOTATIONS (a) A channel may be placed in an inoperable status for up to 2 hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERA 8LE channel in the same trip system l is monitoring that parameter. (b) Unless adequate shutdown margin has been demonstrated per Specifica-tion 3.1.1 and the "one-rod-out" Refuel position interlock has been 2 demonstrated OPERA 8LE per Specification 3.9.1, the shorting links shall be removed from the RPS circuitry prior to and during the time any control red is withdrawn." (c) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel. (d) This function is not required to be OPERA 8LE when the reactor pressure
- vessel head is removed per Specification 3.10.1.
-(e) This function shall be automatically bypassed when the reactor mode switch is not in the Run position. ! ( (f) This function is not required'to be OPERABLE when DRYWELL INTEGRITY is not required. (g) With any control rod withdrawn. Not appifcable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. (h) This function is automatically bypassed when turbine first stage pressure is less than the value of turbine first stage pressure corresponding to ! 40% f RATED THERMAL POWER. MM ** The initial setpoint shall be f 25.4% of the calibrated span on increasing turbine first stage pressure for AT (see 3/4.2.2 for definition) = 0 F .{21%for0 F( o T j 50 F; f 18% for 50 F < AT i 100 F and $ 15% for 100 F< AT .i 170 F. The allowable value shall be 26.9%, 22.5%, 19.5%, and f16.5%respectively. < < { *Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2. ( PERRY - UNIT 1 3/4 3-5 i ' m j TABLE 4.3.1.1-1 (Continued) j 3 1 = REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS j ' 7 CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL ! _E FUNCTIONAL UNIT . CHECK CONDITIONS FOR WHICH TEST CALIBRATION SURVEILLANCE REQUIRED ! [ 10. Turbine Stop Valve - Closure NA M R .' 1 l 11. Turbine Control Valve Fast i Closure Valve Trip Systen 011 Pressure - Low NA M R 1 ! l'2. Reactor Mode Switch , { Shutdown Position MA R NA 1,2,3,4,5 l 13. Manual Scram NA M NA l 1,2,3,4,5 (a) Neutron detectors may be excluded from CHANNEL CALIBRATION. l (b) i The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup ! R after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for
- at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.
(c) Within 24 hours prior to startup. If not performed within the previous 7 days. i T * (d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values i j calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL { e g 2% of RATED THERMAL POWER. i ,CL '.E L'sG ~ Ei..E*'d ~ ~ ' ~ ~'~~ ~~~ "'"'-"'"~~-' ~'~~" " "~ -' '" ! (e) Th s ca cons s of t adjustment of the APRM flow biased channel to conform to a ' calibrated flow signal. * (f) The LPRMs shall be calibrated at least once per 1000 WD/T using the TIP system. (g) Calibrate trip unit setpoint at least once per 31 days. (h) Verify measured core flow (total core flow) to be greater than or equal to established core flow at the existing loop flow (APRM % flow). j (1) This calibration shall consist of verifying the 6 i 0.6 second simulated thermal power time constant. , (j) This function is not required to be OPERABLE when the reactor pressure vessel head is removed j per Specification 3.10.1. . I (k) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. - l (1) This function is not required to be OPERABLE when Drywell Integrity is not required. 4 (m) The CHANNEL CALIBRATION shall exclude the flow reference transmitters, these transmitters shall be j calibrated at least once per 18 months. . i i i 1 . 4 TABLE 3.3.4.2-1 ' END-0F-CYCLE RECIRCULATION PUW TRIP SYSTEM INSTRUMENTATION E
- MINIM 41 p . TRIP FUNCTION OPERABLECHANNElgI .
PER TRIP SYSTEM $
- 1. Turbine Stop Valve - Closure 2(b)
- 2. Turbine Control Valve - Fast closure 2(b) w (a)A trip system may be placed in an inoperable status for up to 2 hours for required surveillance provided that the other trip system is OPERABLE.
1 (b)This function is automatically bypassed when turbine first stage pressure is less than the w value of turbine first stage pressure corresponding to 405 f RATED THERMAL POWER. 8 \ -k
- The initial setpoint shall be < 25.4% of the calibrated span on increasing turbine first stage pressure for AT (see 3/4.2.2 for definition) = 0 F; f 21% for 0 F < AT f 50 F; 3 18% for 50 F < AT i 100 F and 15% for 100 F < AT 1 170 F. The allowable value shall be 26.9%, 22.5%,(19.5%, andg 16.5% respectively.
, t 4 i TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS
- , TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE
@ 1. ROD PATTERN CONTROL SYSTEM 7 a. Low Power Setpoint 20 + 15, - 0% of RATED THERMAL POWER ** 20 + 15. - 0% of RATED THERMAL POWER ** c b. RWL - High Power Setpoint 70 + 0, - 15% of RATED THERMAL POWER ** 70 + 0, - 15% of RATED THERMAL POWER ** 5 m 2. APRM ,o s ,,o
- a. m .: n m m e n -
~ n ac u . -u~ d -" fM" Q - -- u - d
- b. nop
- c. Downscale
~> 4% of RATED THEEMAL POWER > 3% of P.ATED THERMAL POWER
- d. Neutron Flux - Upscale -
Startup i 12% or RATED THERMAL POWER $ 14% of RATED THERMAL POWER
- 3. SOURCE RANCE MONTTORS
- a. Detector r.ot full fn NA NA
- b. Ups ale' < 1 x 105 gs < 1.6 x 105 cps
- c. Icoperative -
NA RA ! d. Downscale 1 0.7 cps p 1 0.5 cps { 4. INTERMEDIATE RANGE MONITORS w 't. Detectar not full in NA a 't. 'pscale J < 108/125 dhision of full scale . NA
- c. Inoperative .
NA ~ < 110/125 division of full scale ' i NA >
- d. Downscale 1 5/125 d M sit,n of full scale 13/125 division of ft>11 scale S. SCRAM DISCHARGE VOLUME -
- a. Water Letci . High < 16.6 inches *** < 17.48 inches ***
- (624' 3.3" elevatica) ~ (624' 4.17" elevation) ,
- 6. REACTOR COOLAYr SYSTEM RECIRCULATION FLCti .
- a. Upscale 5 4085 of rated flow 1-HEM-o.f rated flow ,
- 7. REACTOR MODE SWITCH SHUTDOWN POSITION
# # 70 *Y7o - NA NA 4 % *The Avera e Power Range Monitor rod block function is varied as a function of recirculation loop flow 4 (W). ..e trl ' "in eT d.is f.I..2!ie **The actuil setp he correspo W~ ~ = s . . vii a.E. h of the rb ressure or t ese power levels. *** Level zero is 622' 10.69" elevation; level transultter readout. . g - ..eet ,. .... rs ,. _ , .. -. - e.._::_.e ve.se sey .,e pereitted .e be ircree;;d *h- F^^ :h::_ 'tr'; :^--i.-t af 0.55 " ^ 5,"."' er.d ellentle velse of 0.00 " : '1%). rovided s gna ~ - o noise rat o 1 2. 4 INSERT C 1
- a. Flow Biased Neutron Flux- Upscale , ,
! 1) Flow Biased 1 0.66 W+58% , with i 0.66 W+61% , with a maximum of a maximum of
- 2) High Flow Clamped i 108.0% of RATED f 110% of RATED
, THERMAL POWER THERMAL POWER 4 1 i l 4 t t b. i ) l i I l 3/4.2 POWER DISTRIBUTION LINITS ! (' .' BASES The specifications of this section assure that the peak cladhing temper-ature following the postulated design basis loss-of-coolant accident will not exceed the 2200*F limit specified in 10 CFR 50.46. - i 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE L 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 u:2d in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor. The li;;;itir.g velse fe. ,"#LHGR -4s-shown-in Fi;;r;; 3.2.1-1. 3.2.1-2 er.,: 3.2.1-3. Jose,./ // . /ihn lC The calculational procedure used to establish the APLHGR ch == n Figer;; -3.2.1-1, 0.2.12 ec4 3.2.1-3 is based on a loss of-coolant accident analysis. The analysis was performed using General Electric (GE) calculational models ( which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1. Differences in this analysis compared to previous analyses can i be broken down as follows,
- a. Input Chances
- 1. Corrected Vaporization Calculation - Coefficients in the vaporization correlation used in the REFLOOD code were corrected.
- 2. Incorporated more accurate bypass areas - The bypass areas in the top guide were recalculated using a more accurate technique.
- 3. Corrected guide tube thermal resistance.
- 4. Correct heat capacity of reactor internals heat nodes.
l PERRY - UNIT 1 B 3/4 2-1 __ . ._ . .. . _ - - . - . . , _ _ _ _ _ . _ _ _ = - - - - -. . INSERT H The MAPLHGR limits of Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3 are multiplied by the smaller of either the flow dependent MAPLHGR factor (MAPI'AC F ) r the power dependent MAPLHGR factor (MAPFAC p ) corresponding to existing core flow and power state to assure the adherence to fuel mechanical design bases during the most limiting transient. MAPFAC f 's are determined using the three-dimensional BWR simulator code to analyze slow flow runout transients. MAPFACp's are generated using the same data base as the MCPR to protect the core from plant transients other than p core flow increases. i l 1 POWER DISTRIBUTION LINITS ( BASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)
- b. Model Chance -
1. Core CCFL pressure differential - 1 psi - Incorporate the assumption that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.
- 2. Incorporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed.
A few of the changes affect the accident calculation irrespective of CCFL. These changes are listed below,
- a. Input Chance
- 1. Break Areas - The D8A break area was calculated more accurately.
- b. Model Chance
- 1. Improved Radiation and Conduction Calculation - Incorporation of CHASTE C 05 for heatup calculation.
A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2.1-1. cecccm 3/4.2.2 APRM SETPOINTS erMm / The f I cladding integrity Safety Limits of / y Specification /2.1 were based on a powe distribution which would yield tp design LHGR at ED THERMAL POWER. e flow biasedgmulated thermal, power-htgh ser ip setpoint and / the biased neutro( flux-upscale control rod block tions of the ins nts must be djusted to ensure'that the MCPR snotbecomeles% s than 1(D6 or that > 1astic strain does not occur in degraded situat'fon. The scram settings,and rod block setJfngs are adjustegin accordance wifh the for-mula in this dpecification 7 when the combinatforpef THERMAL POWE and CMFLPD indicatesjt peak power distr,fbution to ensure than an LHGR tra sient would not be increased in degraded conditions. I i 9# PERRY - UNIT 1 8 3/4 2-2 ( .' Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parameters; Core THERMAL POWER . . . . . . . . . . . . . . . . . . . . 3729 Mwt* which corresponds to 105% of rated steam flow Ves sel Steam Output . . . . . . . . . . . . . . . . . . . 16. 2 x 108 lba/hr which ' corresponds to 105% of rated steam, flow Vessel Steam Dome Pressure............. 1060 psia Design Basis Recirculation Line Break Area for:
- a. Large Breaks 2.7 ft2
- b. Small Breaks 0.09 ft2
( Fuel Parameters: PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNOLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kW/ft) FACTOR RATIO Initial Core P8 x 8R 13.4 1.4 4 MCPR f l A more detailed listing of input of each model and its source is presented in Section II of Reference 1 and subsection 6.3 of the FSAR. "This power level meets the Appendix K requirement of 102%. The core t heatup calculation assumes a bundle power consistent with operation of i the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit. PERRY - UNIT 1 B 3/4 2-3 i POWER DISTRIBUTION LIMITS ( BASES 2 3/4.2.1 MINIMUM CRITICAL POWER RATIO / .- The required operating ifait MCPRs at steady state operating conditions as specified in Specification 3.2.J are derived from the established fuel cladding integrity Safety Limit MCPR of 1.06, and an analysis of abnormal l operational transients. For any abnormal 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 setting Limit given in MCPR at any time Specification 2.2. during the transient assuming instrument trip To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR . The type of transients evaluated were loss of flow, increase in ' pressure an)d power, positive reactivity insertion, a d temperature decrease. The limiting transient yields the largest del PR. When added to the Safety Limit MCPR ef 1.00, the required ;;..!; operatingA limit MCPR of Specification 3.2./ is obtained e..4 p. m ..ii.4 tii Ti p. The power-flow map of Figure 8 4/4 2. -1 defines the analytical basis 3.2.f. for 1. generation of the MCPR operating 1 ts. '] J J,s. E. 3-/ W /f 6 S-2 The evaluation of a given transi begins with the system initial parameters shown in FSAR Table 15.0-1 that are input to a GE-core dynamic ( behavior transient computer progrg The code used to evaluate pressurization l events is described in NEDO-24154 events is described in NEDO-10802(2) and the program used in non pressurization The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundlewithtgsinglechanneltransientthermalhydraulicTASCcodedescribed in NEDE-25149 The principal result of this evaluation is the reduction in MCPR caused by the transient. I i The purpose of the MCPR and MCPR efFigree0.2.gler.d0.2. define operating limits at o[her than Eated core flow and power con /2Atis to ditions. less than 100% of rated flow and power the required MCPR is the larger value of the MCPR and MCPR at the existing core flow and power state. The MCPR s are establis$ed to pro @.ect the core from inadvertent core flow increases suck that the 99.9% MCPR limit requirement can be assured. 2 also Figure 3.2./-24 reflects the required MCPR values resulting from the analysis . l 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.
- a. conumide s fecp ger,er Ic fcuer Oca The MCPR f s were calculate at for the maximum core flgw rate and the corresponcing THERMAL POWER along J,e 100% of reted stee. fiew 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 ti. 100% of .eted 5:ee; flew control, line corresponding to different core flows. The calculated MCPR at a given point of core flow is defined as MCPR'. ,
Na, coacNa//fe de90 PERRY - UNIT 1 B 3/4 2-4 fcuct f/CcJ L _ .-_ _ - . _ _ _ . _ . _ _ _ . _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ - _ - _ - _ . . _ _ _ i POWER DISTRIBUTION LIMITS ( . BASES
- MINIMUM CRITICAL POWER RATIO (Continued)
The MCPR s are established to protect the core from plant transients other than core floS increases, including the localized event such as rod withdrawal error. The MCPR s were calculated based upon the most limiting transient at the P given core power level. pgy At THERMAL POWER levels less than or equal to 25% of RATED THERNAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indi- - cates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with mi'nimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of. RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The require-ment for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit. 3 3/4.2.g LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.
References:
- 1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.
- 2. R. 8. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, NEDO-10802, February 1973.
- 3. Qualification of the One Dimensional Core Transient Model For Boiling Water Reactors, NEDO-24154, October 1978.
- 4. TASC 01-A Computer Program For The Transient Analysis of a Single Channel, Technical Description, NEDE-25149, January 1980.
( PERRY - UNIT 1 8 3/4 2-5
INSERT J For core power less than or equal to 40% of RATED THERMAL POWER, where the EOC-RPT and the reactor scrams on turbine stop valve closure and turbine control valve' fast closure are bypassed, separate sets of MRPR p limits are provided for high and low core flows to account for the significant sensitivity to initial core flows. For core power above 40% of RATED THERMAL POWER, bounding power dependent MCPR limits were developed. l D16/48
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i i i i i i i i i i i ! O 0 10 20 30 40 50 60 70 80 90 10 0 110 120 PERCENT CORE FLOW I i , POWER-FLOW OPERATING MAP BASES FIGURE B 3/4 2J-l 4
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