ML20045E154
| ML20045E154 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 06/23/1993 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML19303F669 | List: |
| References | |
| NUDOCS 9307010187 | |
| Download: ML20045E154 (68) | |
Text
{{#Wiki_filter:. . - - _ _ _ _ _ - _ - . __ 4 ATTACHMENT 2 i PEACH. BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 i Docket Nos. 50-277 50-278 License Nos. DPR-44 i DPR-56 l PROPOSED OPERATING LICENSE CHANGES No. 93-12-0 l l List of Attached Pages Unit 2 Unit 3 Operating License (0L), Page 5 Operating License (OL), Page 5 OL, Appendix A OL, Appendix A
" Technical Specifications" " Technical Specifications" Pages: 2*, 6, 9*, 11*, 16*, 17, Pages: 2*, 6, 9*, 11*, 16*, 17, 18, 24, 29, 30, 37*, 39, 18, 24, 29, 30, 37*, 39, l
40*, 49, 50, 73*, 74*, 117, 40*, 49, 50, 73*, 74*, 117, 129, 130, 137, 140a, 140c*, 129, 130, 137, 140a, 140c*, 157, 164d, 189, 193, 195 157, 164d, 189, 193, 195 OL, Appendix B " Environmental Technical OL, Appendix B " Environmental Technical Specifications" Specifications"
- Cover Sheet - Cover Sheet
- Page 2 - Page 2
- These pages reflect proposed changes due to APRM Rod Block Monitor Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS /MELLLA) implementation submitted by the NRC by letter dated April 1,1993. prior to Power Rerate. Page 140c reflects also changes due to the use of SAFER /GESTR Loss-of-Coolant Acsident (LOCA) methodology submitted to the NRC by letter dated March 18, 1993.
9307010187 930623 Y DR ADOCK 05000277 jf3 PDR gg _ _ _ _ _ ____ _
' ' Unit 2 ;
~' ' PBAPS 1.0 DEFINITIONS (Cont'd)
Engineered Safeguard - An engineered safeguard is a safety system the actions of which are essential to a safety action required in response to accidents. , Fraction of Limiting Power Density (FLPD) - The ratio of the' / linear heat generation rate (LHGR) existing at a given location to the design LHGR for that bundle type. Functional Tests - A functional test is the manual operation or' ' initiation of a system, subsystem, or component to verify' that it functions within design tolerances (e.g., the manual start of a , core spray pump to verify that it runs and that it pumps the required volume of water). Gaseous Radwaste Treatment System - Any system designed and installed to reduce radioactive gaseous effluents by collecting i primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the t total radioactivity prior to release to the environment. i> High (power) Trip Set Point (HPTS) - The high power trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting ' applicable above 85% reactor thermal power. Hot Shutdown - The reactor is in the shutdown mode and the reactor coolant temperature greater than 212 F. Hot Standby Condition - Hot Standby Condition means operation with cooiant temperature greater than 212 F, system pressure less than 1085 psig, and the mode switch in the Startup/ Hot Standby l position. The main steam isolation valves may be opened to provide steam to the reactor feed pumps. f Immed, te - Immediate means that the required action.will be initiated as soon as practicable considering the safe operation l of the unit and the importance of the required action. 4 r 4 i a __ __.m _ _ _ . _ _ = _ _ _ . _ _ __- _ __ _ -
Unit 2 8 .o (1) Maximum Power Level' f PECo is authorized to operate the Peach Bottom Atomic Power Station, Unit 2, at steady state reactor core power levels not to exceed 3458- l l [ megawatts thermal. , 5 (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 172 are hereby_ incorporated in the license. PECo shall operate the facility in accordance with the Technical Specifications. (3) The licensees may perform modifications to the Low Pressure Coolant Injection System as described in the licensees' application for license [ t amendment dated July 9, 1975. The licensees shall not operate the. I t facility prior to receipt of the Commission's authorization. , (4) Physical Protection The licensee shall fully implement and maintain in effect all . 1 provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and ; i 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR L 73.21, are entitled: " Peach Bottom Atomic Power Station, Units 2 and 3, i l Physical Security Plan," with the revisions submitted through December 16, I i . ' 1987; " Peach Bottom Atomic Power Station, Units 2 and 3 Plant Security Personnel Training and Qualification Plan," with revis'ons submitted Page 5
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i PBAPS ' 1.0 DEFINITIONS (Cont'd) Protective Action - An action initiated by the protection system when a limit is recched, A protective action can-be at a channel : or system level. , . Protective Function - A system protective action-which results from the protective action of the channels monitoring a - particular plant condition. Purge - Purging - Purge or Purging is the controlled process of 1 l discharging air or gas from a confinement to maintain ' temperature, pressure, humidity, concentration or other. operating condition, in such a manner that replacement air or gas is required to purify the confinement. 1 Rated Power - Rated power refers to operation at a reactor power e of 3458 MWt; this is also termed 100 percent power and is the l maximum power level authorized by the operating license. Rated steam flow, rated coolant flow, rated neutron flux, and rated. nuclear system pressure refer to the values of these parameters ; when the reactor is-at rated power. ,, Reactor Power Operation - Reactor power operation is any operation with the mode switch in the "Startup" or "Run" position i with the reactor critical and above 1% rated power. l e Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel. steam space detectors. ; Refuel Mode - With the mode switch in the refuel position, the : reactor is shutdown and interlocks are established so that only one control rod may be withdrawn. Refueling Outage - Refueling outage is the period of time between I the shutdown of the unit prior to a refueling and the startup'of , the unit after that refueling. For the purpose of designating , frequency of testing and surveillance, a refueling outage shall-mean a regularly scheduled outage; however, where-such outages occur within 8 months of the completion of the previous refueling l l
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" ' Unit 2 PBAPS LIMITING SAFETY SYSTEM SETTING -SAFETY LIMIT 2.1 FUEL CLADDING INTEGRITY 1.1 FUEL CLADDING INTEGRITY Applicability:
Applicability: _The Limiting Safety System Settings The Safety Limits established to preserve the fuel cladding apply to trip settings of the integrity apply to those instruments and devices which are provided to prevent the fuel variables which monitor the cladding integrity Safety Limits
. fuel thermal behavior. from being exceeded.
Objectives: Objectives: The objective of the Safety The objective of the Limiting Safety Limits is to establish limits System Settings is to define the which assure the integrity of the fuel cladding. level of the process variables at which automatic protective action is initiated to prevent the fuel cladding Specification: integrity Safety Limits from being exceeded. A. Reactor Pressure 2 800 psia Specification: , i and Core flow 2 10% of Rated The limiting safety system settings The existence of a minimum shall be as specified below: critical power ratio (MCPR) less than 1.06 for two recirculation loop operation, A. Neutron Flux Scram or 1.07 for single loop APRM Flux Scram Trip Setting operation, shall constitute , 1. violation of the fuel cladding (Run Mode) ' integrity safety limit. When the Mode' Switch is in the To ensure that this safety RUN position, the APRM flux limit is not exceeded, neutron scram trip secting shall be: flux shall not be above the l i scram setting established in S s 0.66W + 66% - 0.66 AW (Clamp 0120%) specification 2.1.A for longer than 1.15 seconds as indicated by the process computer. When where: the process computer is out of service this safety limit shall S= -Setting in percent of rated thermalpower(3458MWt) l be assumed to be exceeded if the neutron flux exceeds its Loop recirculating flow rate ! scram setting and a control W= in percent of design. rod scram does not occur. l _g_ s
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. Unit 2 1 1
~- PBAPS !
LIMITING SAFETY SYSTEM SETTING SAFETY LIMIT j Core Thermal Power limit B. APRM Rod Block Trio Settina B. (Reactor Pressure s 800 psia) When the reactor pressure is Su s (0.66 W + 54% - 0.66 AW) l
$ 800 psia or core flow is (Clamp 9108%) .
less than 10% of rated, the core thermal power shall not where: exceed 25% of rated thermal ' power. Sg - Rod block setting in ' percent of rated thermal power (3458 MWt) l W = Loop recirculation flow
-rate in percent of design. ,
aW = Difference between two loop and single loop effective recirculation drive flow at the same core flow. During single loop operation, the reduction in trip i> setting (-0.66 AW) is accomplished by correcting the flow input of the , flow biased rod block to preserve the original L(two loop) relationship between APRM Rod block setpoint and recirculation drive flow or by adjusting the APRM Rod block trip setting. AW = 0 for two loop operation. The APRM rod block trip setting shall not exceed 108% of rated thermal power.
i Figure 1.1-1 APRM Flow Biased Scram Relationship to Normal Operating Conditions . 130
-~~~~
120 - Power / P.ow points 33 110 - :iogp3ig p@M
- 100P/100F / @ b b 100- : 100P/105F p
- 67P/110F /
; 62P/104F '
, g, /
f tft t 48P/106F S0 . / MELLhegion M / / w 70 - 0.66wd + 66% _ _ _ f ..g. 3004* [y b p ], stismiv rututina nnte
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g_ . plow @
< 1 e 1: setected and inseaim Domain
> I'I' or flowIncrease
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2: Starmp Operations Only 40 - Two Pump 5 # Speed (20%) p uolLD* 30 - MWW ""' Namral Canuladon 20 - 10 - 6b 7b 8b 9b Id0 110. 120 0 lb 2b 3b 4b Sb Notes : P = Uprated Thermal Power (100% upoted Power = 3458 MWt) CORE FLOW (%) F = M!bhrCore Dow (100% Core flow = 101.5 MtMw) r, _ m we 4 e -s'*p'T b-+*e a w-
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Unit 2 PBAPS
2.1 BASES
FUEL CLADDING INTEGRITY P The abnormal operational transients applicable to operation of 1 the Peach Bottom Atomic Power' Station Units have been analyzed
- throughout the spectrum of planned operating conditions up to or '
above the thermal power condition required by Regulatory Guide 1.49 The analyses were based upon plant operation in accordance . with the operating map given in Figure 3.7.1 of the FSAR. In j' addition, 3458 MWt is the licensed maximum power level of each ' Peach Bottom Atomic Power Station Unit, and this represents the maximum steady state power which _shall not knowingly be exceeded. (See Reference 4). Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity
- coefficient, control rod scram worth, scram delay time, peaking- ,
factors, and axial power shapes. These factors are selected conservatively with respect to their effect on the applicable . _ , transient results as determined by the current analysis .model. , l Conservatism incorporated into the transient analysis is documented in NEDE-24011-P-A (as amended). E h t
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Unit 3 . n -- , 12.1 BASES (Cont'd) , For analyses'of the thermal consequences of the transients, a 1
-MCPR equal to'or greater than the operating limit MCPR given in. ;
Specification 3.5.K is conservatively assumed to exist This choice prior to of using~ -!' initiation of the limiting transients. conservative values of controlling parameters and initiating transients at the design power level produces more pessimistic " answers than would result by using expected-values of control . parameters and analyzing at higher power levels. . Steady state operation without forced recirculation will not be permitted. The analysis to support operation at various power ' and flow relationships has considered operation with either one . or two recirculating pumps. In summary:
- 1. The abnormal operational transients were analyzed at or above the maximum power level required by Regulatory Guide 1.49 to determine operating limit MCPR's.
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- 11. The licensed maximum power level is 3458 MWt.
111. Analyses of transients employ adequately conservative values of the controlling reactor parameters.- ,, , iv. The analytical procedures now used result in a more logical' i answer than the alternative method of assuming a higher . starting power in conjunction with the expected values for the parameters. The. bases for individual trip settings- are discussed in.the following paragraphs. A. Neutron Flux Scram > The Average Power Range Monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (3458'MWt). l Because fission chambers provide the basic input signals, the
- APRM system responds directly to average neutron flux. . During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous T'. c ~~ fore,neutron during flux due to the time constant of the fuel. of the fuel abnormal operational transients, the thermal b will be less than that indicated by the neutron flux at the scram -
setting. Analyses demonstrate that with a 120 percent scram-trip setting, none of the abnormal operational transients analyzed violate the fuel Safety Limit and there is a substantial margin from fuel' damage. Therefore, the use of flow referenced scram trip proxides even additional margin. ' r
Unit 2
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2.1 BASES
(Cont'd) L. References .
- 1. Linford, R. B., " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," NED010802, February 1973.
- 2. " General Electric Standard Application for. Reactor fuel",
NEDE-24011-P-A (as amended). ' 3. " Methods for Performing BWR Reload Safety Evaluations," PECo-FMS-0006-A (as amended).
- 4. " Power Rerate Safety Analysis Report for Peach Bottom 2 & >
3," NEDC-32183P, May 1993.
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Unit 2 I LIMITING SAFETY SYSTEM SETTING l SAFETY' LIMIT 1.2- REACTOR COOLANT SYSTEM INTEGRITY 2.2: REACTOR COOLANT SYSTEM i INTEGRITY r Aeolicability: Apolicability: i Applies to limits on reactor coolant system pressure. Applies to trip settings / of the instruments-and devices ' which are provided to prevent the reactor system safety limits from being exceeded. Ob.iectives: Ob.iectives: To establish a limit below which the integrity of the To define the level of the process variables at which j reactor coolant system is not ' threatened due to an automatic protective action-overpressure condition. is initiated to prevent the pressure safety limit from being exceeded. Soecification: Specification:
- 1. The reactor vessel dome pressure shall not exceed 1. The limiting safety system 1325 psig at any time when settings shall be as irradiated fuel is pree 'nt specified below:
in the reactor vessel. Protective Action /limitina Safety System Settinq , A. Scram on Reactor Vessel high pressure
$1085 psig B. Relief valve settings 1135 psig ( 11 psi) l (4 valves) 1145 psig (ill psi) l !
(4 valves) 1155 psig ( 11 psi) l - (3 valves)
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Unit 2 ..~ i 7 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1 C. Safety valve settings ;
- 2. The reactor vessel dome pressure shall not ex- 1260 psig i 13 psi-
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ceed 75 psig at any (2 valves) time when operating the . Residual Heat Removal 2. The shutdown cooling iso-pump in the shutdown lation valves shall be ; cooling mode, closed whenever the reac- l tor vessel dome pressure ; is >75 psig. t k 7 l 1. t e 9 i s
PBAPS Unit
- Table 3.1.1 , ,
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Modes In which Number of Minimum No. Instrument $
, of Operable Function Must Be .
Trip Level Operable Channel s Action
- Instrument Provided (1)
Channels Trip Function Setting Refuel Startup Run by Design per Trip (7) Item System (1) , x x 1 Mode Switch A 1 Mode Switch In x 1 Shutdown (4 Sections) x x x 2 Instrument A 2 1 Manual Scram Channels g E x x (5) 8 Instrument A 3 3 IRM High Flux 5120/125 of Full Channels Scale IRM Inoperative x x (5) 8 Instrument A 4 3 Channels
$' x 6 Instrument A or B l 5 2 APRM High Flux (0.66W+66%-0.66AW)
(Clamp 0 120%) Channels (12) (13) x x x 6 Instrument A or B 6 2 APRM (11) Channels Inoperative (10) 6 Instrument A or B 7 2 APRM Downscale 22.5 Indicated Channels on Scale APRM High Flux $15% Power x x 6 Instrument A 8 2 Channels in Startup High Reactor $1085 psig x(9) x x 4 Instrument A l 9 2 Channels Pressure ' High Drywell $2 psig x(8) x(8) x 4 Instrdment A 10 2 Channels Pressure Reactor Low 20 in. Indicated x x x- 4 Instrument A 11 2 Channels Water Level Level M
-w -v-rvs- - - W w w,w w w w we , e +--
Unit 2
' PBAPS NOTES FOR TABLE 3.1.1-
- 1. There shall be two operable or tripped trip systems for each function. If the minimum number of operable sensor channels -
for a trip system cannot be met, the affected trip system - shall be placed in the safe (tripped) condition, or the. '; appropriate actions listed below shall be taken. i A. Initiate insertion of operable rods and complete insertion of all operable rods within four hours.
. Reduce power level to IRM range and place mode switch in B.
the start up position within 8 hours. , C. Reduce turbine load and close main steam line isolation valves within 8 hours. . D. Reduce power to less than 30% rated. Permissible to bypass, in refuel and shutdown positions of 2. the reactor mode switch. :
- 3. Deleted.
- 4. Bypassed when turbine first stage pressure is less than that which is equivalent to.30% of rated thermal power.
- 5. IRM's are bypassed when APRM's are onscale and the reactor mode switch is in the run position.
- 6. The design permits closure of any two lines without a scram being initiated.
- 7. When the reactor is subtritical and the reactor water temperature is less than 212 degrees F, only the following trip functions need to be operable: ,
A. Mode switch in shutdown B. Manual scram C. High flux 1RM D. Scram discharge instrument volume high level
- 8. Not nequired to be operable when primary containment inteQrity is not required. ,
- 9. Not required to be operable when the reactor pressure vessel head'is not bolted to the vessel.
Unit 2 PBAPS-4 NOTES FOR TABLE 3.1.1 (Cont'd)
- 10. The APRM downscale trip is automatically bypassed when the IRM instrumentation is operable and not high.
- 11. An APRM will be considered operable if there are at least 2 LPRM inputs per level and at least 14 LPRM inputs of the normal complement.
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- 12. W = Loop Recirculation flow in percent of design.
Delta'W = The difference between two loop and single loop effective recirculation drive flow rate at the same core flow. During single loop operation, the reduction in trip setting (-0.66 delta W) is accomplished by correcting the flow input of the flow biased High Flux trip setting to preserve the original (two loop) rela-tionship between APRM High Flux setpoint and recirculation drive flow or by adjusting the APRM Flux trip setting. Delta W equals zero for two loop operation. ,
]
Trip level setting is in percent of rated power (3458 MWt).
- 13. See Section 2.1.A.I.
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Unit 2 PBAPS 3.1 BASES (Cont'd)' the amount of water which must be accommodated during a scram. , During normal operation the discharge.. volume is empty- ! however, should it fill with water, the water discharged.to' the piping from the reactor could not be accommodated which would result in slow scram times or partial control rod .c insertion. To preclude this occurrence, level. switches have . been provided in the instrument volume which alarm and scram ; the reactor when _the volume of water reaches 50 gallons. As - indicated above, there is sufficient volume in the piping to - accommodate the scram without impairment of the scram times ' or amount of insertion of the control rods. This function shuts the reactor down while sufficient volume remains to accommodate the discharged water and precludes the situation , in which a scram would be required but not be able to perform its function adequately. A source range monitor (SRM) system is also provided to j 7' , supply additional neutron level information during start-up ' but has no scram functions (reference paragraph 7.5.4 FSAR). . 3 Thus, the IRM and APRM are required in the " Refuel" and .
" Start / Hot Standby". modes. In the power range the APRM system provides required protection (reference paragraph 7.5.7 FSAR) . Thus the .IRM System is not required in the f "Run" mode. The APRM's cover only the power range. The ;
IRM's and APRM's provide adequate coverage in the start-up and intermediate range. ! The high reactor pressure, high drywell pressure, reactor low water level and scram discharge volume.high level scrams are required for Startup and Run modes of plant operation. They j l are, therefore, required to be operational for these modes of reactor operation. .l-The requirement to have the scram functions indicated in , Table 3.1.1 operable in the Refuel mode assures that shifting to the Refuel mode during reactor power operation does not diminish the protection provided by the reactor protection system. i The turbine condenser low vacuum scram is only required 1 during power operation and must be bypassed to start up the 'i unit. The main condenser low vacuum trip is bypassed except in the run position of the mode switch. Turbine stop valve closure occurs at 10% of valve closure. When turbine l first stage pressure is below that which corresponds to 30% of rated thermal power, the scram signal due to turbine stop valve closure is ,
. bypaised because the flux and pressure scrams are adequate to protect the reactor.
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- Unit 2-i 3.1 BASES (Cont'd.)
Turbine control valves fast closure initiates-a scram base'd' on pressure switches sensing Electro-Hydraulic Control. (EHC) system oil pressure. The switches are located be-tween f ast closure solenoids and the disc dump valves, and , are set relative (500<P<B50 psig) to the normal EHC oil i pressure.of 1600 psig gauge that, based on the small system ' volume, they can rapidly detect valve closure or loss of hy-draulic pressure. This scram signal is also bypassed when the turbine first stage pressure indicates.that reactor power is less than 30% of rated. The requirement that the IRM's be inserted in the core when the APRM's read 2.5 indicated on the scale in the Startup ~j and Refuel modes assures that there is proper overlap in the neutron monitoring system functions and thus, that adequate coverage is provided for all ranges of reactor operation. > i t
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PBAPS Unit 2 TABLE 3.2.C INSTRUMENTATION THAT INITIATES CONTROL R0D BLOCKS Trip Level Setting Number of Instrument Action Minimum No. Instrument of Operable Channels Provided by Design Instrument
<. Channels Per Trip System APRM Upscale (Flow Biased) (0.66W+54%-0.66AW) 6 Inst. Channels (10)- l 4 (2) (Clamp at 108% max) 4 APRM Upscale (Startup 512% 6 Inst. Channels (10)
Mode) 4 APRM Downscale 22.5 indicated on scale 6 Inst. Channels (10) Rod Block Monitor (RTP 285%), S., sHTSP 2 Inst. Channels (1) 1 (7)(11) (65% sRTP <85%), S,, sITSP (Power Biased) (30% $RTP <65%), S,, sLTSP Rod Block Monitor 2DTSP 2 Inst. Channels (1) 4 1 (7)(11) Y Downscale 6 IRM Downscale (3) 22.5 indicated on scale 8 Inst. Channels (10) 8 Inst. Channels IRM Detector not in (8) (10) 6 Startup Position-6 IRM Upscale s108 indicated on scale 8 Inst. Channels (10) SRM Detector not in (4) 4 Inst. Channels . (1) - . 2 (5) Startup Position SRM Upscale $10' counts /sec. 4 Inst. Channels (1) l 2 (5)(6) 1 Scram Discharge s25 gallons 1 Inst. Channel (9) f Instrument Volume I High Level 4 w
Unit 2
- PBAPS NOTES FOR TABLE 3.2.C
- 1. For the startup and run positions of the Reactor Mode Selector. Switch, there shall be two operable or tripped trip systems for each function.
- The SRM and IRM blocks need not be operable in "Run" mode, and the APRM and RBM rod blocks need not be operable in "Startup" mode. If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than seven days, the system shall be tripped. If the first column cannot be met for both trip systems, the systems shall be tripped.
- 2. W = Loop Recirculation flow in percent of design.
Trip level setting is in percent of rated power (3458 MWt). l AW is the difference between two loop and single loop effective recirculation drive flow rate at the same core flow. During single loop operation, the reduction in trip setting is accomplished by correcting the >* flow input of the flow biased rod block to preserve the original (two loop) relationship between the rod block setpoint and recirculation drive fl ow. 4W = 0 for two loop operation. ,
- 3. IRM downscale is bypassed when it is on its lowest range.
- 4. This function is bypassed when the count rate is 2100 cps.
- 5. One of the four SRM inputs may be bypassed.
- 6. This SRM function is bypassed when the IRM range switches are on range 8 or above.
- 7. The trip is bypassed when the reactor power is s 30%.
- 8. This function is bypassed when the mode switch is placed in Run.
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.,a Unit 2 ;
I PBAPS SURVEILLANCE REQUIREMENTS LIMITING CONDITIONS FOR OPERATION 4.4 STANDBY LIQUID CONTROL' SYSTEM 3.4 STANDBY LIQUID CONTROL SYSTEM (Contd.) .; (Cont'd.)
- 3. The Standby Liquid Control System '
conditions must satisfy the following equation: C 1( Q f E p 19.8%ato,-{. 13% wt. j (86 gpm ; where, C= Sodium Pentaborate Solution . Concentration (% weight) Pump Flow Rate (gpm) 3. Pump Flow Rate: At least i i Q= once per month each pump against a system head of 1255 psig. loop shall be functionally l ,
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tested by pumping boron solution to the test tank. At least once per quarter ' check and record pump flow rate against a system head of 1255 psig. l
- 4. Enrichment: Following each E= Boron-10 Enrichment (% atom Boron-10) addition of boron to the solution tank, calculate enrichment within 8 hours.
Verify results by analysis within 30 days.
- 5. Solution Volume: At least once per day check and record.
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. Unit-2 .
PBAPS SURVEILLANCE REQUIREMENTS LIMITING CONDITIONS FOR OPERATION 1 3.5.C HPCI subsystem-(cont'd.) . 4.5.C HPCI Subsystem (cont'd.) Item ' Frequenc"y (b) Pump .Once/ month Operability (c) Motor Operated. Once/ month Valve
, Operability (d) Flow Rate at once/3 months approximately 1030 psig l Reactor Steam.
Pressure. (e) Flow Rate at once/ operating ! 150 psig cycle
-l Reactor Steam Pressure s>
The HPCI pump shall deliver ' at'least 5000 gpm for a system head corresponding to a reactor.. pressure of approximately 1030 to 150 psig.
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- 2. From and after the date that 2. DELETED the HPCI Subsystem is made or
'found to be inoperable for any reason, continued reactor operation.is permissible only during the succeeding seven days unless such subsystem is r sooner made operable, provi-ding that during such seven days all active components of the ADS subsystem, the RCIC system, the LPCI subsystem ,
and both core spray subsys-tems are operable.
- 3. If the requirements of 3.5.C cannot be met, an orderly shut-down shall be initiated and !
the reactor shall be in a cold shutdown Condition , within 24 hours. e
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__s- > c' Unit 2- l l PBAPS SURVEILLANCE REQUIREMENTS LIMITING CONDITIONS FOR OPERATION ~ 3.5.D Reactor Core Isolation 4.5.D Reactor Core Isolation Cooling (RCIC) Subsystem Cooling (RCIC) Subsystem
- 1. RCICSubsystemtestingshaIl
- 1. The RCIC Subsystem shall be be performed as follows:
operable whenever there is irradiated fuel in the reactor Item Frequency. vessel, the reactor steam pressure is greater than 105 psig, and once/ Operating prior to. reactor startup from (a) Simulated a Cold Condition, except as Automatic Cycle specified in 3.5.D.2 below. Actuation
-Test *
(b) Pump Once/ Month , operability (c) Motor Operated once/ Month Valve operability (d) Flow Rate at once/3 Months approximately i 1030 psig i4i Reactor Steam Pressure ** (e) Flow RateLat once/ Operating ' approximately Cycle ' 150 psig Reactor Steam l Pressure ** (f) Verify auto- Once/ Operating *** matic transfer Cycle , from CST to suppression pool , on low CST water + level
- 2. DELETED
- 2. From and after the date that the RCIC Subsystem is made or found to be inoperable for any reason, continued reactor power opera-tion is permissible only during
- Shall include automatic restart the succeeding seven days on low water level signal, provided that during such seven days the HPCI Subsystem ** The RCIC pump shall deliver is operable. at least 600 gpm for a system
- 3. If the requirements of 3.5.D head corresponding to a reactor i'
cannot be met, an orderly shut- pressure of approximately 1030 to l down shall be initiated and 150 psig. the reactor pressure shall be reduced *to 105 psig within *** Effective at lat refueling outage . after Cycle 7 reload. 24 hours.
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PBAPS Unit 2
,,s i
3.5 BASES (cont'd.) C. HPCI The limiting conditions for operating the HPCI System are derived from the Station Nuclear Safety Operational Analy- , l sis (Appendix G) and a detailed functional analysis of the c ; HPCI System (Section 6.0). The HPCIS is provided to assure that the rieactor core is ' adequately cooled to limit fuel clad temperature in the : event of a small break in the nuclear system and loss-of-coolant which does not result in rapid depressurization of the reactor vessel. The HPCIS permits the reactcr to be shut down while maintaining sufficient reactor vessel i water level inventory until the vessel is depressurized. The HPCIS continues to operate until reactor vessel pres-sure is below the pressure at which LPCI operation or Core _ Spray System operation maintains core cooling. [ L The capacity of the system is selected to provide this re-quired core cooling. The HPCI pump is designed to pump 5000 gpm at reactor pressures between 1150 and 150 psig. 4 Two sources of water are available. Initially, deminera-lized water from the condensate storage tank is used in-stead of injecting water from the suppression pool into the reactor. When the HPCI System begins operation, the reactor depres- > surizes more rapidly than would occur if HPCI was not ini-tiated due to the condensation of steam by the cold fluid [ pumped into the reactor vessel by the HPCI System. As the reactor vessel pressure continues to decrease, the HPCI flow momentarily reaches equilibrium with the flow through
- the break. Continued depressurization causes the break flow to decrease below the HPCI flow and the liquid inven-tory begins to rise. This type of response is typical of ,
the small breaks. The core never uncovers and is continu-ously cooled throughout the transient so that no core damage of any kind occurs for breaks that lie within the capacity range of the HPCI. The analysis in the FSAR, Appendix G, shows that the ADS provides a single failure proof path for depressurization for postulated transients and accidents. The RCIC serves as an alternate to the HPCI only for decay heat removal . I when feed water is lost. Considering the HPCI and the ADS plus RCIC as redundant paths, reference (1) methods would give an estimated allowable repair time of 15 days based on the one month testing frequency. However, a maximum , allowable repair time of 7 days is selected for conservatism. t
-137-
t Unit 2 PBAPS , 3.5 BASES (Cont'd)
.i J. Local LHGR This specification assures that the linear heat generation rate in any 8X8 fuel +
rod is less than the design linear heat generation. The maximum LHGR shall be ! 2 checked daily during reactor operation at 25% power to determine if fuel i burnup, or control rod movement has caused changes in power distribution. For LHGR to be at the design LHGR below 25% rated thermal power, the peak local : LHGR must be a factor of approximately ten (10) greater than the average LHGR l which is precluded by a considerable margin when employing any permissible control rod pattern. K. Minimum Critical Power Ratio (MCPR) ; O Operating Limit MCPR ,, The required operating limit MCPR's at steady state operating conditions are , derived from the established fuel cladding integrity c,afety Limit MCPR and analyses of the abnormal operational transients presented in Supplemental Reload l Licensing Analysis and References 7 and 10. For any abnormal operating tran- ! sient analysis evaluation with the initisi condition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.1. : To assure that the fuel cladding integrity Safety Limit is not violated during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in critical power ratio (CPR). See Reference 12. The transients eyaluated are as l. l described in References 7 and 10. P i I ' -140a-(
Unit 2 : PBAPS '[ i 3.5.L. BASES (Cont'd) Operating experience has demonstrated that a calculated value of APLHGR, LHGR i or MCPR exceeding its limiting value predominately occurs due to this latter
-i cause. This experience coupled with the extremely unlikely occurrence of con-
- current operation exceeding APLHGR, LHGR or MCPR and a loss-of-Coolan't? Accident or applicable Abnormal Operational Transients demonstrates that the times required to initiate corrective action (1 hour) and restore the calculated ;
value of APLHGR, LHGR or MCPR to within prescribed limits (5 hours) are adequate including MELLL operation with implementation of ARTS restrictions (Ref. 11). 3.5.M. References
- 1. " Fuel Densification Effects on General Electric Boiling Water Reactor j Fuel", Supplements 6, 7 and 8, NEDM-10735, August 1973. ,
- 2. Supplement I to Technical Report on Densifications of General Electric Reactor Fuels, December 14,1974 (Regulatory Staff).
i i Communication: V. A. Moore to I. S. Mitchell, " Modified GE Model for Fuel i
- 3. !
Densification", Docket 50-321, March 27, 1974. : Letter, C. O. Thomas (NRC) to J. F. Quirk (GE), " Acceptance for l 1
- 4. 1 Referencing of Licensing Topical Report NEDE-23785, Revision 1, Volume III (P), 'The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident'," June 1, 1984. ,
- 5. DELETED.
- 6. DELETED.
t
- 7. " General Electric Standard Application for Reactor Fuel", NEDE-24011-P-A (as amended). .
i
- 8. " Peach Bottom Atomic Power Station Units 2 and 3 SAFER /GESTR - LOCA '
Loss-of-Coolant Accident Analyses," NEDC-32163P, January,1993. t
- 9. DELETED. ,.
- 10. " Methods for Performing BWR Relocd Safety Evaluations," PECo-FMS-0006-A -
t (as amended). !
- 11. " Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for 4 Peach Bottom Atomic Power Station Units 2 and 3," NEDC-32162P, I
Revision 1, February, 1993. !
- 12. " Power Rerate Safety Analysis Report for Peach Bottom 2 & 3," NEDC-32183P, May 1993.
-140c-
- _ _ _ _ _ _ _ _ . . _ - ____m < _ _ _ _ __--_ _.__
i Unit 2 PBAPS 3.6.D & 4.6.D BASES Safety and Relief Valves The safety / relief and safety valves are required to be operable , above the pressure (122 psig) at which the core spray system is not designed to deliver full flow. The pressure relief sy' stem for each unit at the Peach Bottom APS has been sized to meet two , design bases. First, the total capacity of the safety / relief and the safety valves has been established to meet the overpressure protection criteria of the ASME code. -Second, the distribution ; of this required capacity between safety / relief valves and safety l valves has been set to meet design basis 4.4.4.1 of subsection 4.4 of the FSAR which states that the nuclear system safety / relief valves shall prevent opening of the safety valves during normal plant isolations and load rejections. The details of the analysis which shows compliance with the ASME i code requirements is presented in subsection 4.4 of the FSAR and the Reactor Vessel Overpressure Protection Summary Technical Report presented in Appendix K of the FSAR. Eleven safety / relief valves and two safety valves have been ,l installed on Peach Bottom Unit 2 with a total capacity of 75.30% , of rated-steam flow. The analysis of the worst overpressure transient demonstrates margin to the code allowable overpressure - limit of 1375 psig. To meet the power generation design basis, the total pressure relief system capacity of 75.30% has been divided into 62.21% safety / relief (11 valves) and 13.09% safety (2 valves). The analysis of the plant isolation transient shows that the 11 safety /relie. valves limit pressure at the safety valves below the setting of the safety valves. Therefore, the safety valves will not open. Experience in safety / relief and safety valve operation shows that a testing of 50 per cent of the valves per year is adequate to detect fall .e or deteriorations. The safety / relief and safety i valves are Denchtested every second i 4 3 9 9
-157-
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-164d-
Unit'2 PBAPS 3.7.A/4.7.A BASES Primary Containment The integrity of the primary containment and operation of the core standby cooling system in combination, limit the-off-site doses to. values less than those suggested in 10CFR100 in the event of a break in the primary system pip-ing. Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists. Concern about such a violation exists whenever the reactor is critical and above atmospheric pressure. An exception is made to this requirement during j initial core loading and while the low power test program- j is being conducted and ready access to the reactor vessel j is required. There will be no pressure on the system at ' this time, thus greatly reducing the chances of a pipe l break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring. Procedures and the Rod Worth Minimizer would limit control worth such that a rod drop would not result in any fuel ,, damage. In addition, in the unlikely event that an excur-sion did occur, the reactor building and standby gas treat-ment system, which shall be operational during this time, offer a sufficient barrier to keep off-site doses well below . 10CFR100 limits. The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blow-down from 1038 psig. Since all of the gases in the drywell l are purged into the pressure suppression chamber air space during a loss-of-coolant accident, the pressure resulting from isothermal compression plus.the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber ' maximum pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is dis-charged to the suppression chamber and that the drywell volume is purged to the suppression chamber. Using the minimum or maximum water volumes given in the speci-fication, containment pressure during the design basis acci-dent is approximately 49.1 psig which is below the maximum of 62 psig. Maximum water volume of 127,300 f t' results in a downcomer submergence of 4.4 feet and the minimum volume of 122,900 f t' results in a submergence approximately 0.4 feet less.
-189-
Unit 2 . PBAPS 3.7.A & 4.7.A BASES (Cont'd) ! The design ba- 'osy-of-coolant accident was evaluated in the SER at the , primary conti maximum allowable accident leak rate of 0.5%/ day at 56
-i psig, a stanu .
as +.reatment system filter efficiency of 90% for halogens and ' assuming the fission product release fractions stated in TID-14844. The SER i shows that the maximum two hour dose is about 1.0 REM whole body and 14 REM thyroid at 4500 meters from the stack. The resultant doses in the SER that would occur for the duration of the accident at the low population zo~ne~ distance of 7300 meters are about 2.5 REM total whole body and 105 REM total thyroid. As a result of uprating the power to 3,458 MWt, the corresponding doses calculated in UFSAR Subsection 14.9 are more conservative since they are-based on a containment leak rate of 0.635% per day and larger dispersion (X/Q) values. These UF5xx analyses result in two hour doses at the Exclusion Area l Boundary of about 1.0 REM whole body and 15 REM thyroid. The UFSAR analyses also result in doses at the low population zone distance (7300 meters) for the duration of the accident of about 3.9 REM whole body and 239 Rem thyroid. Thus, the doses reported are the maximum that would be expected in the unlikely event of a design basis loss-of-coolant accident, These doses are also based on the assumption of no holdup in the secondary containment resulting in a direct release of fission products from the primary containment f through the filters and stack to the environs. Therefore, the specified primary containment leak rate and filter efficiency are conservative and ' provide margin between expected off-site doses and 10 CFR 100 guidelines. . The water in the suppression chamber is used only for cooling in the event of , an accident; i.e., it is not used for normal operation; therefore, a daily check of the temperature and volume is adequate to assure that ade- l quate heat removal capability is present. i Drywell Interior , The interiors of the drywell and suppression chamber are l painted to prevent rusting. The inspection of the paint ' during each major refueling outage, approximately once j per year, assures the paint is intact. Experience with this type of paint at fossil fueled generating stations , indicates that the inspection interval is adequate. , Post LOCA Atmosphere Dilution In order to ensure that the containment atmosphere remains , inerted, i.e. the oxygen-hydrogen mixture below the flam- ' mable limit, the capability to inject nitrogen into the containment after a LOCA is provided. During the first j year of operation the normal inerting nitrogen makeup sys- - tem will be available for this purpose. After that time the specifically designed CAD system will serve as the post-LOCA Containment Atmosphere Dilution System. By maintaining a minimum of 2000 gallons of liquid N, in the storage tank it is assured that a seven-day supply of N, for post-LOCA containment inerting is available. Since ' the inerting makeup system is continually functioning, no
-193- ,
.._ : . i Unit 2
/ .
PBAPS , 3.7.A & 4.7.A BASES (Cont'd) Due to the nitrogen addition, the pressure in the containment . after a LOCA will' increase with time. Under the worst expected conditions, repressurization of the containment ; will reach 30 psig. If and when that pressure is reached, venting from the containment shall be manually , initiated. The venting path will be through the Standby .l Gas Treatment system in order to minimize the off site dose. Following a'LCCA, periodic operation of the drywell and torus sprays.will be used to assist the natural convection and diffusion mixing of hydrogen and oxygen. i
;i i
+
t t
-195- :
APPENDIX B i TO FACILITY OPERATING LICENSE DPR-44 AND FACILITY c OPERATING LICENSE DPR-56 ENVIRONMENTAL , TECHNICAL SPECIFICATIONS AND BASES FOR THE FULL POWER FULL TERM OPERATION OF PEACH BOTTOM ATOMIC POWER STATION i'- UNIT 2 MAY 31, 1989 . l 4 YORK COUNTY, PENNSYLVANIA PHILADELPHIA ELECTRIC COMPANY ! DOCKET NO. 50-277 f w f
Unit 2 PBAPS-
- 1. Protection Limit - A numerical limit on a plant effluent or operating parameter which, when not exceeded, should not result in an unacceptable environmental impact. ..
- m. Rated Thermal Power - Rated thermal power refers to operation at a reactor power of 3458 MWt.
- n. Report Level - The numerical level of an environmental park'-
meter below which the environmental impact is considered e reasonable on'the basis of available information.
- o. Special Study Program - An environmental study program -
designed to evaluate the impact of plant operation on an environmental parameter.
- p. Total Residual Chlorine - The sum of the free chlorine and the combined chlorine.
1.2 ABBREVIATIONS 1
- a. AEC - Atomic Energy Commission ii
- b. BWR - Boiling Water Reactor
- c. 10CFR20 - Code of Federal Regulations; Title 10 - Atomic Energy -
Part 20 - Standard for Protection Against Radiation
- d. 10CFR50 - Code of Federal Regulations; >
Title 10 - Atomic Energy Part 50 - Licensing of Production and Utilization Facilities
- e. FSAR - Final Safety Analysis Report i
- f. NEPA - National Environmental Policy Act
- g. MPC - Maximum Permissible Concentration
- h. MSL - Mean Sea Level
- i. PBAPS - Peach Bottom Atomic Power Station Units No. 2 and 3 f
- j. POR - Plant Operations Review
- k. 6&SR - Operation and Safety Review >
- 1. PMF - Probable Maximum Flood
- m. PSAR - Preliminary Safety Analysis Report
':^
I d Unit 3 , (1) Maximum Power Level , PECo is authorized to operate the Peach Bottom Atomic Power Station, Unit 3, at steady state reactor core power levels not in excess of 3458' l ; 1 , megawatts thermal. ,
. (2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.175 are hereby incorporated in the license. PECo shall operate the facility in accordance with' the Technical Specifications. (3) Physical Protection The licensee shall fully implement and maintain in effect all i provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including , amendments made pursuant to provisions of the Miscellaneous Amendments, , and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and , The 27822) and to the _ authority of 10 CFR 50.90 and 10 CFR 50.54(p). plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: " Peach Bottom Atomic Power Station, Units 2 and 3, Physical Security Plan," with the revisions submitted through December 16, ; 1987; " Peach Bottom Atomic Power Station, Units 2 and 3 Plant Security -l i Personnel Training and Qualification Plan," with revisions submitted through July' 9,1986; and " Peach Bottom Atomic Power Station, Units 2 l and 5 Safeguards Contingency Plan," with revisions submitted through Page 5 I
- ~^ '
~- -
Unit 3- - PBAPS 1.0 DEFINITIONS (Cont'd)
~
Engineered Safeguard - An engineered safeguard is a safety system the actions of which are essential.to a safety action required in l response to accidents. Fraction of Limiting Power Density (FLPD) - The ratio of the ; linear heat generation rate (LHGR) existing at a given location to the design LHGR for that bundle type. Functional Tests - A functional test is the manual operation or . initiation of a system, subsystem, or component to verify that' it ' functions within design tolerances (e.g., the manual ' start of a' core spray pump to verify that it runs and that it pumps the required volume of water). Gaseous Radwaste Treatment System - Any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. High (power) Trip Set Point (HPTS) - The high power trip setpoint : associated with the Rod . Block Monitor (RBM) rod block. trip setting (, applicable above 85% reactor thermal power. . Hot Shutdown - The reactor is in the shutdown mode and the p reactor coolant temperature greater .than 212 F. Hot Standby Condition - Hot Standby Condition means operation ' with coolant temperature greater than 212 F, system pressure less than 1085 psig, and the mode switch in the Startup/ Hot Standby .: position. The main steam isolation valves may be opened to provide steam to the reactor feed pumps. . Immediate - Immediate means that the required action will be initiated as soon as practicable considering the safe operation 4 of the unit-and the importance of the required action. .
; a s
~ ~
- Unit 3 PBAPS ,
l.0 DEFINITIONS-(Cont'd) Protective Action - An action initiated by the protection system when a limit is reached. A protective action can be at a channel ; or system level. , Protective Function - A system protective action which results from the protective action of the channels monitoring a particular plant condition. _l Purge - Purging - Purge or Purging is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. Rated Power - Rated power refers to operation at a reactor power of 3458 MWt; this is also termed 100 percent power and is,the l maximum power level authorized by the operating license. Rated
- steam flow, rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the values of these parameters ,
when the reactor is at rated power. .. Reactor Power Operation - Reactor power operation is any operation with the mode _ switch in the "Startup" or "Run" position with the reactor critical and above 1% rated power. . Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those ; measured by the reactor vessel steam space detectors. Refuel Mode - With the mode switch in the refuel position, the j reactor is shutdown and interlocks are established so_that only ~ one control rod may be withdrawn. Refueling Outage - Refueling outage is the period of time between a the shutdown of the unit prior to a refueling and the startup of the unit after that refueling. For the purpose of designating ' frequency of testing and surveillance, a refueling outage shall-mean a regularly scheduled outage; however, where_such outages. occur within 8 months of the completion of the previous refueling f S
.- _ m _
Unit 3 PBAPS
. SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING .1.1- FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Aonlicability:
Aan11cability: i
'The Safety Limits established The Limiting Safety' System Settings .to preserve the fuel cladding apply to trip settings of the -integrity apply to those instruments and devices'wh,1ch are T variables which monitor the provided to prevent: the fuel: -i fuel thermal behavior. cladding integrity Safety Limits from being exceeded.
Obiectives: Obiectives: The objective of the Safety Limits is to establish limits The objective of the Limiting Safety which assure the integrity of System Settings is to define the , the fuel cladding. level of the process variables at which automatic protective action is initiated to prevent the fuel cladding Suecification: integrity Safety Limits from being ' exceeded. A. Reactor Pressure 2 800 usia Specification: and Core Flow k 10% of Rated The existence of a minimum The limiting safety system settings i>- critical power ratio (MCPR) shall be as specified below: less than 1.06 for two recirculation loop operation, A. Neutron Flux Scram or 1.07 for single loop . operation, shall constitute 1. APRM Flux Scram Trin Setting violation of the fuel cladding (Run Mode) integrity safety limit. When the Mode Switch is in the To ensure that this safety RUN position; the APRM flux limit is not exceeded, neutron scram trip setting shall be: flux shall not be above the scram setting established in S s 0.66U + 66% - 0.66 av l specification 2.1.A for longer (Clamp @ 120%) than 1.15 seconds as indicated by the process computer. When where: 1 the process computer is out of _ service this safety limit shall S- Setting.in percent of rated . . be assumed to be exceeded if thennal power (3458 MWt) l the neutron flux exceeds its scram setting and a control V- Loop recirculating flow rate. rod scram does not occur. in percent of design. b 9 t
- a Unit 3- '
' PBAPS LIMITING SAFETY SYSTEM SETTING SAFETY LIMIT Core Thermal Power Limit B. APRM Rod Block Trin Settinr.
B. (Reactor Pressure s 800 psia) When the reactor pressure is S gg 5 (0.66 W + 54% - 0.66 AW) f s 800 psia or core flow is (Clamp @ 108%) ' less than 10% of rated, the core thermal power shall not where: exceed 25% of rated thermal power. SRs - Rod block setting in percent of rated thermal power (3458 MVt) l W- Loop recirculation flow rate'in percent of design. AV = Difference between two loop and single loop effective recirculation drive flow at the same , core flow. During single loop operation, the reduction in trip s 's-setting (-0.66 AW) is accomplished by correcting the flow input of the flow biased rod block to - preserve the original (two loop) relationship between APRM Rod block ' setpoint and recirculation drive flow or by adjusting the APRM Rod block trip setting. - t AW - O for two loop operation. ; The APRM rod block trip setting shall not exceed 108% of rated . thermal power. t 5 9 4 i
Figure 1.1-1 APRM Flow Biased Scram Relationship to Normal Operating Comlitions , l- 130 1 120 - Power / Flow points e------- Scth #
~
- 100P/81F
- 100P/100F y @ @Q
/
100 - : 100P/105F j
- 67Pil10F /
- 62P/10tF
/
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. tR : 48P/106F /
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/
1 is 0 70 - 0.66Wd + 66% __ f 3 g .9,# g . Increased Core / g 4 cn H m A # ' stahtitiv r<etu=w neet= - 8 a 60 -
- pio, @
2 1: Selected Rod insenam Domain
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2: Stamp Opuntions Only h. i W p: 40 - Two rump
# Speed (20%) yog Ltne 30 - ggmom Natural Circulation 20 -
h 10 - 0 , , 120 30 40 50 - 60 70 80 90' 100- 110.- - 0 10 20 < t Notes : P = Uprated %crmal Power _ (100% uprated Power - 345s Mwt) ' CORE FLOW (%)
- F = MitWrCore Flow (100% Core Flow - 10295 Mtter)
C
- l. .
I LJ N e 4 @' _.-4 P-aw 'T w m u m-3ww - M C Yvt'-" '9m 'ti++wwWeewe e m m mye w p-wn-eTPe,e --wp"<.-WafyMe'h ver % @w w w w* Cc ww--'T- - y>fy-we W wv' 4 y* - w ev gw- 7 y pp g it gy #*-' sw+%m- F T+- 1
.c Unit 3 PBAPS .
t
2.1 BASES
FUEL CLADDING INTEGRITY ; The abnormal operational transients applicable to operation of the Peach Bottom Atomic Power Station Units have been analyzed throughout the spectrum of planned 1 operating conditions up to or above the thermal power condition required by Regulatory Guide 1.49. The analyses were based upon plant operation in accor- , dance with the operating map given in Figure 3.7.1 of the FSAR. In' addition, 3458 MWt is the licensed maximum power level of each Peach Bottom Atomic Power l l i Station Unit, and this represents the maximum steady state power which shall l not knowingly be exceeded. (See Reference 6). t i Conservatism is incorporated in the transient analyses in estimating the _ controlling f actors, such as void reactivity coefficient, control rod scram worth,' scram delay _ time, peaking factors, and axial power shapes. These fac- ', tors are. selected conservatively with respect to their effect on the applicable j transient results as detenmined by the current analysis model. Conservatism ' incorporated into the transient analysis is documented in References 4 and 5. t f t i n e e I 9
Unit 3
- '2 .1 BASES (Cont'd) i For analyses of the thermal consequences of the transients, a '
MCPR equal to or greater than the operating limit MCPR given in Specification 3.5.K is conservatively assumed to exist prior.to initiation of the limiting transients. This choice of using conservative values of controlling parameters and initiating transients at the design power level produces more pessimistic ' answers than would result by using expected values of control parameters and analyzing at higher power levels. Steady state operation without forced recirculation will not be permitted. The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculating pumps. . In summary:
- 1. The abnormal operational transients were analyzed at or above the maximum power level required by Regulatory Guide 1.49 to determine operating limit MCPR's.
The licensed maximum power level is 3458 MWt. l ii. 111. Analyses of transients employ adequately conservative values of the controlling reactor parameters. ci iv. The analytical procedures now used result in a more logical answer than the alternative method of assuming a higher starting power in conjunction with the expected values for - the parameters. The bases for individual trip settings are discussed in the , following paragraphs. A. Neutron Flux Scram , The Average Power Range Monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated' thermal power (3458 MWt). l Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel i (reactor thermal power) is less thatt the instantaneous neutron flux due to the time constant of the fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that Indicated by the neutron flux at the scram setting. Analyses demonstrate that with a 120 percent Lcram trip , setting, none of the abnormal operational transients analyzed 3 violate the fuel Safety Limit and there is a substantial margin from fuel damage. Therefore, the use of flow referenced scram trip provides even additional margin. a 1
. . . . ~ . .
l I Unit 3 i
2.1 BASES
(Cont'd) L. References
- 1. Linford, R. B., " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," NED010802, February 1973. ,
i
- 2. " Qualification of the One-Dimensional Core Transient'Model for Boiling Water Reactors", NED0 24154 and NEDE 24154-P, Volumes I, '
II, and III.
- 3. " Safety Evaluation for the General Electric Topical Report Qualification of the One-Dimensional Core Transient-Model for ;
Boiling Water Reactors NED0-24154 and NEDE 24154-P, Volumes I, II, and III.
- 4. " General Electric Standard Application for Reactor Fuel", , ,
NEDE-24011-P-A (as amended).
- 5. " Methods for Performing BWR Reload Safety Evaluations," -
PECo-FMS-0006-A (as amended).
- 6. " Power Rerate Safety Analysis Report for Peach Bottom 2 & 3," <
NEDC-32183P, May 1993.
'f i
t, o
Unit 3-- I SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.2 REACTOR COOLANT SYSTEM INTEGRITY 2.2 REACTOR COOLANT SYSTEM INTEGRITY Acolicability: Applicability: Applies to limits on reactor coolant system pressure. Applies to trip settings.; of the instruments and devices which are provided.to prevent the reactor system safety limits from being exceeded. Objectives: Objectives: To establish a limit below which the integrity of the To define the level of the-reactor coolant system is not process variables at which threatened due to an automatic protective action overpressure condition. is initiated to prevent the pressure safety limit from being exceeded. Soecification: Specification:
- 1. The reactor vessel dome
pressure shall not exceed 1. The limitir., safety system 1325 psig at any time when settings shall be as irradiated fuel is present specified below: in the reactor vessel. ,, Protective Action /Limitina Safety System Settina A. Scram on Reactor Vessel high pressure
$1085 psig l B. Relief valve settings 1135 psig (ill psi) l-(4 valves) l 1145 psig (ill psi) l-(4 valves) 1155 psig ( 11 psi) l (3 valves)
-~ - --
Unit-3 , i SAFETY LIMIT LIMITING SAFETY SYSTEM- - SETTING .; . C. Safety valve settings
- 2. The reactor vessel dome pressure'shall not ex- 1260 psig i 13 psi. "
l l
~
ceed 75 psig at any . (2 valves) time when operating the
- 2. The shutdown cooling' iso-Residual Heat Removal
'lation valves shall be pump in the shutdown cooling mode, closed whenever the reac- ~ t tor vessel dome pressure. ,
is >75 psig. 5 I i I L t I t 6 I k k J
PBAPS Unit 3 Table 3.1.1 , REACTOR PROTECTION SYSTEM (SCRAMI INSTRUMENTATION REQUIREMENT Modes In which Number of Minimum No.
, of Operable Function Must Be Instrument Trip Level Operable Channels Action
- Instrument Provided Channels Trip Function Setting (1)
Refuel Startup Run by Design per Trip (7) Item System (1) 1 Mode Switch In x x x 1 Mode Switch A 1 Shutdown (4 Sections) Manual Scram x x x 2 Instrument A 2 1 Channel s 3 3 IRM High Flux s120/125 of Full x x (5) 8 Instrument A 3 Scale Channels 3 IRM Inoperative x x (5) 8 Instrument A
, 4
"' Channels 7 x 6 Instrument A or B l 5 2 APRM High Flux (0.66W+66%-0.66AW) (Clamp @ 120%) Channels (12) (13) x x x 6 Instrument A or B 6 2 APRM (11) Channels Inoperative APRM Downscale 22.5 Indicated (10) 6 Instrument A or 8 7 2 on Scale Channels 2 APRM High Flux $15% Power x x 6 Instrument A 8 Channels in Startap 2 High Rea: tor $1085 psig x(9) x x 4 Instrument A l 9 Pressure Channels 1 2 High Drywell $2 psig x(8) x(8) x 4 Instrdment A 10 Channels - Pressure 2 Reactor Low 10 in. Indicated x x x 4 Instrument A 11 Channels Water Level Level M
+- .__%-~ ..wm
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_ _.. m,___.m._.. . _ _ . _ . - _ _ _ a_ mm___._m_ - m___.____ . _ . _ __._.___ _ _ _ _ . . . _
Unit'3 <
' PBAPS i
NOTES FOR TABLE 3.1.1 7
^
- 1. There shall be two operable or tripped trip systems for each ;
function. If the minimum number of operable sensor channels l for a trip system cannot be met, the affected trip system ' shall be placed in the safe (tripped) condition, or the appropriate actions listed below shall be taken. . A. Initiate insertion of operable rods and complete , insertion of all operable rods within four hours. , B. Reduce power level to IRM range and place mode switch in the start up position within 8 hours. C. Reduce turbine load and close main steam line isolation valves within 8 hours. D. Reduce power to less than 30% rated. .
- 2. Permissible to bypass, in refuel and shutdown positions of '
the reactor mode switch.
-ie
,}
- 3. Deleted. ,
Bypassed when turbine first stage pressure is less than
- 4. -
that which is equivalent to 30% of rated thermal power. ; I
- 5. IRM's are bypassed when APRM's are onscale and the reactor mode switch is in the run position. ,
- 6. The design permits closure of any two lines without a scram being initiated.
- 7. When the reactor is subtritical and the reactor water I temperature is less than 212 degrees F, only the following trip functions need to be operable:
l A. Mode switch in shutdown B. Manual scram r C. High flux IRM D. Scram discharge instrument volume high level
- 8. Not required to be operable when primary containment inteOrity is not required.
- 9. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel, i
i t
l Unit 3
.. - . j PBAPS NOTES FOR TABLE 3.1.1 (Cont'd)
- 10. The APRM downscale trip is automatically bypassed when the IRM instrumentation is operable and not high.
- 11. An APRM will be considered operable if there are at least 2 LPRM inputs per level and at least 14 LPRM inputs of the normal complement. .
- 12. W = Loop Recirculation flow in percent of design.
Delta'W = The difference between two loop and single loop effective recirculation drive flow rate at the same core flow. During l single loop operation, the reduction in trip setting (-0.66 delta W) is accomplished by correcting the flow input of the flow biased t High Flux trip setting to preserve the original (two loop) rela-
~
tionship between APRM High Flux setpoint and recirculation drive ' flow or by adjusting the APPJi Flux trip setting. Delta W equals zero for two loop operation. Trip level setting is in percent of rated power (3458 MWt). .
- 13. See Section 2.1.A.1.
?
O b
Unit 3 PBAPS "3.1 BASES (Cont'd) the amount'of water which must be accommodated during a l scram. During normal ope;ation the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not be accommodated which would result in slow scram times or partial control rod ; insertion. To preclude this occurrence,_ level switches have been provided in the instrument volume which alarm and scram the reactor when the volume of water reaches 50 gallons. As indicated above, there is sufficient volume in the piping to accom.aodate the scram without. impairment of the scram times or imount of insertion of the control rods. This function shuts the reactor down while sufficient volume remains to accommodate the discharged water and precludes the situation in which a scram would be required but not be able to perform its function adequately. A source range monitor (SRM) system is also provided to supply additional neutron level information during start-up but has no scram functions (reference paragraph 7.5.4 FSAR). i Thus, the IRM and APRM are required in the " Refuel" and ,
" Start / Hot Standby" modes. In the power range the APRM system provides required protection (reference paragraph 7.5.7 FSAR). Thus the IRM System is not required in the -
"Run" mode. The APRM's cover only the power range. The IRM's and APRM's provide adequate coverage in the start-up and intermediate range.
The high reactor pressure, high drywell pressure, reactor low water level and scram discharge volume high level scrams are required for Startup and Run modes of plant operation. They are, therefore, required to be operational for these modes of reactor operation. The requirement to have the scram functions indicated in Table 3.1.1 operable in the Refuel mode assures that shifting to the Refuel mode during reactor power operation does not diminish the protection provided by the reactor protection system. The turbine condenser low vacuum scram is only required during power operation and must be bypassed to start up the unit. The main condenser low vacuum trip is bypassed except in the run position of the mode switch. Turbine stop valve closure occurs at 10% of valve closure. When turbine first stage pressure is below that which corresponds to 30% of rated thermal power, the scram signal due to turbine stop valve closure is bypassed because the flux and pressure scrams are adequate to protect the reactor.
Unit 3 3.1 BASES (Cont'd.) ; , Turbine control valves fast closure initiates a scram. base'd - on pressure switches sensing Electro-Hydraulic Control (EHC) system oil pressure. The switches are located be- : tween fast closure solenoids and the disc dump. valves, and are set relative (500<P<850 psig) to the normal EHC oil pressure.of 1600 psig gauge that, based on the small system g volume, they can rapidly detect valve closure or loss of hy-draulic pressure. This scram signal is also bypassed when the turbine first stage pressure indicates that reactor power is less than 30% of rated. I i The requirement that the IRM's be inserted in the core when the APRM's read 2.5 indicated on the scale in the Startup and Refuel modes assures that there is proper overlap in the neutron monitoring system functions and thus, that adequate , coverage is provided for all ranges of reactor operation. i I
~ )
J t e O e l
PBAPS Unit 3'
- t. =
TABLE 3.2.C ' ' INSTRUMENTATION THAT INITIATES CONTROL R0D BLOCKS Trip Level Setting Number of Instrument Action Minimum No. Instrument of Operable Channels Provided by Design Instrument
.Channelg Per Trip System APRM Upscale (Flow Biased) (0.66W+54%-0.66AW) 6 Inst. Channels .(10) l
- - 4 (2) (Clamp at 108% max) 4 APRM Upscale (Startup $12% 6 Inst. Channels (10)
. Mode) ,. 4 APRM Downscale 22.5 indicated on scale 6 Inst. Channels (10) Rod Block Monitor (RTP 285%), S , sHTSP 2 Inst. Channels (1) 1 (7)(11) (65% sRTP <85%), S , sITSP (Power Biased) (30% _ sRTP <65%), S , sLTSP 1 (7)(11) Rod Block Monitor 2DTSP 2 Inst. Channels (I) i-Downscale gi 6 IRM Downscale (3) 22.5 indicated on scale 8 Inst. Channels (10) 6 IRM Detector not in (8) 8 Inst. Channels (10) Startup Position 6 IRM Upscale $108 indicated on scale 8 Inst. Channels (10) SRM Detector not in (4) 4 Inst. Channels- (1) 2 (5) i Startup Position SRM Upscale $10' counts /sec. 4 Inst. Channels (1) , 2 (5)(6) ' 1 -Scram Discharge $25 gallons- 1 Inst. Channel (9)' Instrument Volume High Level , 4 e e t
% + - -
rr n.. - , , e .s.
~4' Unit 3 PBAPS NOTES FOR TABLE 3.2.C
- 1. For the startup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each fun'ction.
The SRM and IRM blocks need not be operable in "Run" mode, and the APRM and RBM rod blocks need not be operable in "Startup" mode. .If the first-column cannot be met for one of the two trip systems, this condition may. exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than seven days, the system shall be tripped. .If the first column cannot be met for both trip systems, the systems shall be tripped. I
- 2. W = Loop Recirculation flow in percent of design.
Trip level setting is in percent of rated power (3458 MWt). l AW is the difference between two loop and single loop effective recirculation drive flow rate at the same core flow. During single loop operation, the reduction in trip setting is accomplished by correcting the i+ flow input of the flow biased rod block to preserve the original (two loop) relationship between the rod block setpoint and recirculation drive fl ow. AW = 0 for two loop operation.
- 3. IRM downscale is bypassed when it is on its lowest range.
- 4. This function is bypassed when the count rate is 2100 cps.
- 5. One of the four SRM inputs may be bypassed.
- 6. This SPJi function is bypassed when the IRM range switches are on range 8 or above.
- 7. The trip is bypassed when the reactor power is s 30%.
- 8. This function is bypassed when the mode switch is placed in Run.
Unit 3 , PBAPS SURVEILLANCE REQUIREMENTS LIMITING CONDITIONS FOR OPERATION 4.4 STANDBY LIQUID CONTROL 35YSTEM 3.4 STANDBY LIQUID CONTROL SYSTEM * (Cont'd.) (Cont'd.)
- 3. The Standby Liquid Control System conditions must satisfy '
the following equation: 13% wt. C fQ \ E (86gpmf(19.8% atomy} j where, ' C= Sodium Pentaborate Solution Concentration (% weight) Pump Flow Rate (gpm) 3. Pump Flow Rate: At least ,1 Q= once per month each pump against a system head of 1255 psig. loop shall be functionally . l. 1 tested _by pumping boron solution to the test tank. At least once per quarter check and record pump flow rate against a system head of 1255 psig. l j '
- 4. Enrichment: Following each E= Boron-10 Enrichment (% atom Baron-10) addition of boron to the solution tank, calculate enrichment within 8 hours.
Verify results by analysis within 30 days. ,
- 5. Solution Volume: At least once per day check and record. l r:
t
. -117-i
_ _ _ _ _ - - _______i___ __m -r - =
_ _ .m - i a Unit 3 PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.C HPCI Subsystem (cont'd.) 4.5.C HPCI Subsystem (cont'd.) Item Frequenc'y (b) Pump once/ month Operability (c) Motor Operated Once/ month Valve Operability (d) Flow Rate at once/3 months approximately 1030 psig l Reactor Steam Pressure (e) Flow Rate at once/ operating 150 psig cycle Reactor Steam l Pressure
! I The HPCI pump shall deliver at least 5000 gpm for a system head corresponding to a. reactor pressure of approximately 1030 to l*
150 psig.
- 2. From and after the date that 2. DELETED the HPCI Subsystem is made or found to be inoperable for any reason, continued reactor operation is permissible only during the succeeding seven days unless such subsystem is sooner made operable, provi-ding that during such seven days all active components of the ADS subsystem, the RCIC system, the LPCI subsystem and both core spray subsys-tems are operable.
- 3. If the requirements of.3.5.C cannot be met, an orderly shut-down shall be initiated and the reactor shall be in a
' Cold Shutdown Condition within 24 hours.
. -129-
'+ L.. j j i
.- b Unit 3 ,
PBAPS l SURVEILLANCE REQUIREMENTS LIMITING CONDITIONS FOR OFFaTTON 3.5.D Reactor Core Isolation 4.5.D Reactor Core Isolation Cooling (RCIC)' Subsystem Cooling (RCIC) Subsystem The RCIC Subsystem shall be 1. Pf IC Subsystem testing shkIl 1. operable whenever there is .a performed as follows- i irradiated fuel in the reactor Item Frequency i vessel, the reactor steam pressure is greater than 105 psig, and once/ Operating prior to reactor startup from (a) Simulated a Cold Condition, except as Automatic Cycle specified in 3.5.D.2 below. Actuation Test * , (b) Pump Once/ Month' operability (c) Motor Operated Once/ Month Valve ' Operability (d) Flow Rate at once/3 Months approximately i 1030 psig il. Reactor Steam , Pressure ** (e) Flow Rate at Once/ Operating
- approximately Cycle 150 psig i Reactor Steam I Pressure **
(f) Verify auto- Once/ Operating *** , matic transfer Cycle ? from CST to suppression pool. -t on low CST water level .
- 2. From and after the date that 2. DELETED ,
the RCIC Subsystem is made or found to be. inoperable for any reason, continued reactor power opera-tion is permissible only during
- Shall include automatic restart j the succeeding seven days on low water level signal.
provided that during such ' eeven days the HPCI Subsystem is operable.
** The RCIC pump shall deliver <i at least 600 gpm for a system
- 3. If the requirements of 3.5.D head corresponding to a reactor cannot be met, an orderly shut- pressure of approximately 1030 to f down shall be initiated and 150 poig.
the reactor pressure shall be reduced. to 105 psig within *** Effective at let refueling outage ) 24 hours.
- after Cycle 7 reload. <
l
-130- 7 r
4 PBAPS Unit 3 . 4 . 3.5 BASES (cont'd.) C. HPCI , The limiting conditions for operating the HPCI System are derived from the Station Nuclear Safety Operational Analy- . sis (Appendix G) and a detailed functional analysis of the HPCI System (Section 6.0). The HPCIS is provided to assure that the rleactor core is 1 adequately cooled to limit fuel clad temperature in the ! event of a small break in the nuclear system and loss-of-coolant which does not result in rapid depressurization of the reactor vessel. The HPCIS permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurizei ' The HPCIS continues to operate until reactor vessel pres-sure is below the pressure at which L.PCI operation or Core Spray System operation maintains core cooling. The capacity of the system is selected to provide this re-quired core cooling. The HPCI pump is designed to pump ,J ! 5000 gpm at reactor pressures between 1150 and 150 psig. Initially, deminera-Two sources of water are available. lized water from the condensate storage tank is used in-stead of injecting water from the suppression pool into - the reactor. t When the HPCI System begins operation, the reactor depres-
- surizes more rapidly than would occur if HPCI was not ini-tiated due to the condensation of steam by the cold fluid As the pumped into the reactor vessel by the HPCI System.
reactor vessel pressure continues to decrease, the HPCI flow momentarily reaches equilibrium with the flow through i the break. Continued depressurization causes the break ' flow to decrease below the HPCI flow and the liquid inven- l tory begins to rise. This type of response is. typical of the small breaks. The core never uncovers and is continu- l ously cooled throughout the transient so that no core i damage of any kind occurs for breaks that lie within the capacity range of the HPCI. The analysis in the F3AR, Appendix G, shows that the ADS provides a single failure proof path for depressurization for postulated transients and accidents. The RCIC serves as an alternate to the HPCI only for decay heat removal - when feed water is lost. Considering the HPCI and the ADS. : plus RCIC as redundant paths, reference (I) methods would ! give an estimated allowable repair time of 15 days based
- on the one month' testing frequency. However, a maximum
, allowable repair time of 7 days is selected for conservatism.
i
-137-
b Unit 3 PBAPS 3.5 BASES (Cont'd) J. Local LHGR This specification assures that the linear heat generation rate in any 8X8 fuel The maximum LHGR shall be rod is less than the design linear heat generation. checked daily during reactor operation at 225% power.to determine-if fuel For burnup, or control rod movement has caused changes in power distribution. LHGR to be at the design LHGR below 25% rated thermal power, the peak local ' LHGR must be a factor of approximately ten (10) greater than the average LHGR which is precluded by a considerable margin when employing any permissible 3 control rod pattern. K. Minimum Critical Power Ratio (MCPR) Operating Limit MCPR The required operating limit MCPR's at steady state operating conditions are derived from the established fuel cladding integrity Safety Limit MCPR and - analyses of the abnormal operational transients presented in Supplemental Reload Licensing Analysis and References 7 and 10. For any abnormal operating tran-sient analysis evaluation with the initial condition of the reactor being at - the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient
- assuming instrument trip setting given in Specification 2.1.
To assure that the fuel cladding integrity Safety Limit is not violated during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in l critical power ratio (CPR). See Reference 12. The transients evaluated are as , described in References 7 and 10. i 1 k
-140a-a t
Unit 3 PBAPS 3.5.L. BASES (Cont'd) Operating experience has demonstrated that a calculated value of APLHGR, LHGR or MCPR exceeding its limiting value predominately occurs due.to this latter ! cause. This experience coupled with the extremely unlikely occurrence of con-current operation exceeding APLHGR, LHGR or MCPR and'a loss-of-CoolaritJ Accident or applicable Abnormal Operational Transients demonstrates that the times required to initiate corrective action (1 hour) and restore the calculated l value of APLHGR, LHGR or MCPR to within prescribed limits (5 hours) are ! adequate including MELLL operation with implementation of ARTS restrictions l (Ref. 11). i 3.5.M. References
- 1. " Fuel Densification Effects on General Electric Boiling Water Reactor Fuel", Supplements 6, 7 and 8, NEDM-10735, August 1973.
- 2. Supplement 1 to Technical Report on.Densifications of General Electric-Reactor Fuels, December 14,1974 (Regulatory Staff).
V. A. Moore to I. S. Mitchell, " Modified GE Model for Fuel ..
- 3. Communication:
Densification", Docket 50-321, March 27, 1974.
- 4. Letter, C. O. Thomas (NRC) to J. F. Quirk (GE), " Acceptance for. -
Referencing of Licensing Topical Report NEDE-23785, Revision 1, Volume III (P), 'The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident'," June 1, 1984.
- 5. DELETED.
- 6. DELETED.
- 7. " General Electric Standard Application for Reactor Fuel", NEDE-24011-P-A (as amended).
- 8. " Peach Bottom Atomic Power Station Units 2 and 3 SAFER /GESTR - LOCA Loss-of-Coolant Accident Analyses," NEDC-32163P, January,1993,
- 9. DELETED.
- 10. " Methods for Performing BWR Reload Safety Evaluations," PECo-FMS-0006-A-(as amended).
- 11. " Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Peach, Bottom Atomic Power Station Units 2 and 3," NEDC-32162P, Revision 1, February, 1993.
NEDC-'
- 12. " Power Rerate Safety Analysis Report for Peach Bottom 2 & 3,"
32183P, May 1993.
-140c- ,
f i Unit 3 . l PBAPS i e 3.6.D & 4.6.D BASES , Safety and Relief Valves ! The safety / relief and safety valves are required to be operable - above the pressure (122 psig) at which thepressure The core spray reliefsyst,em systemis ; not designed to deliver full flow. i for each unit at the PeachtheBottom APS hasofbeen the sized safetyto/ meet relief two design bases. First, total capacity and the safety valves has been established to meet the Second, the distribution overpressure ! l protection criteria of the ASME code.of this required capacity between safe valves has been set to meet design basis 4.4.4.1 of subsection 4.4 of the FSAP which states that the nuclear system safety / relief valves shall prevent opening of the safety valves during normal plant isolations and load rejections. The details of the analysis which shows compliance with the ASME code requirements is presented in subsection 4.4 of the FSAR and
- the Reactor Vessel Overpressure Protection Summary Technical Report presented in Appendix K of the FSAR.
ii Eleven safety / relief valves and two safety valves have been ] installed on Peach Bottom Unit 3 with a total capacity of 75.30% of rated-steam flow. The analysis of the worst overpressure transient demonstrates margin to the code allowable overpressure - l limit of 1375 psig. t To meet the power generation design basis, the total pressure ' relief syster. capacity of 75.30% and has been 13.09% divided safety (2 into 62.21% valves). The safety / relief (11 valves) analysis of the plant isolation transient shows that the 11 safety / relief valves limit pressure at the safety Therefore, thevalves below safety valves the setting of the safety valves. will not open. Experience in safety / relief and safety valve operation shows that ' a testing of 50 per cent of the valves per The year/ is safety adequate relief and safetyto detect failure or deteriorations. valves are benchtested every second t i
-157-
Unit 3 PBAPS o
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Unit 3
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PBAPS 3.7.A/4.7.A BASES Primary Containment The' integrity of the primary containment and operation of the core standby cooling system in combination, limit the ~~ i off-site doses to values less than those suggested in 10CFR100 in the event of a break in.the primary system pip-ing. Thus, containment integrity is specified whenever the potential for violation of the primary reactor system " integrity exists. Concern about such a violation exists : whenever the reactor is critical and above atmospheric pressure. An exception is made to this requirement during : initial core loading and while the low power test program . is being conducted and ready. access to the reactor' vessel is required. There will be no pressure on the' system at this timo, thus greatly reducing the chances of a . pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be'in effect i again to minimize the probability of an accident occurring. ' Procedures and the Rod Worth Minimizer would limit control worth such that a rod drop would not result in any fuel - i In addition, in the unlikely event that an excur- ,i damage. ~ sion did occur, the_ reactor building and standby gas treat-ment system,.which shall be operational during this time, offer a sufficient barrier to keep off-site doses well below _ 10CFR100 limits. The pressure suppression pool water provides the heat sink , for the reactor primary system energy release following'a postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blow-down from 1038 psig. Since all of the gases in the drywell l are purged into the pressure suppression chamber air space during a loss-of-coolant accident, the pressure resultin.j from isothermal compression plus the vapor pressure.of the : liquid must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is dis-charged to the suppression chamber and that the drywell volume is purged to the suppression chamber. Using the minimum or maximum water volumes ' given in the speci-fication, containment pressure during the design basis acci- , dent is approximately 49.1 psig which is below the maximum , of 62 psig. Maximum water volume of 127,300 ft results in 2 a downcom,er submergence of 4.4 feet and the minimum volume of 122 900 f t' results in a submergence approximately 0.4 feet less.
-189-l
# - +~t a A Unit 3 ,
PBAPS 3.7. A & 4.7. A BASES (Cont'd) i The design basis loss-of-coolant accident was evaluated in the SER at the primary containment maximum allowable accident leak rate of 0.5%/ day at 56 psig, a standby gas treatment system filter efficiency of 90% for halogens and assuming the fission product release fractions stated in TID-14844. The SER shows that the maximum two hour dose is about 1.0 REM whole body and 14 REM thyroid at 4500 meters from the stack. The resultant doses in the SER that would occur for the duration of the accident at the low population z6ne distance of 7300 meters are about 2.5 REM total whole body and 105 REN total thyroid. As a result of uprating the power to 3,458 MWt, the corresponding doses calculated in UFSAR Subsection 14.9 are more conservative since they are based on a containment leak rate of 0.635% per day and larger dispersion (X/Q) values. These UFSAR analyses result in two hour doses at the Exclusion Area - Boundary of about 1.0 REM whole body and 15 REM thyroid. The UFSAR analyses also result in doses at the low population zone distance (7300 meters) for the duration of the accident of about 3.9 REM whole body and 239 Rem thyroid. Thus, the doses reported are the maximum that would be expected in the unlikely event of a design basis loss-of-coolant accident. These doses are also based on the assumption of no holdup in the secondary containment resulting in a direct release of fission products from the primary containment through the filters and stack to the environs. Therefore, the specified primary containment leak rate and filter efficiency are conservative and
- provide margin between expected off-site doses and 10 CFR 100 guidelines.
The water in the suppression chamber is used only for cooling in the event of - an accident; i.e., it is not used for normal operation; therefore, a daily check of the temperature and volume is adequate to assure that ade-quate heat removal capability is present. Dry'well Interior , The interiors of the drywell and suppression chamber are painted to prevent rusting. The inspection of the paint during each major refueling outage, approximately once per year, assures the paint is intact. Experience with this type of paint at fossil fueled generating stations ; indicates that the inspection interval is adequate. i Post LOCA Atmosphere Dilution i In order to ensure that the containment atmosphere remains inerted, i.e. the oxygen-hydrogen mixture below the flam-mable limit, the capability to inject nitrogen into the 1 containment after a LOCA is provided. During the first year of operation the normal inerting nitrogen makeup sys-tem will be,available for this purpose. After that time the specifically designed CAD system will serve as the post-LOCA Containment Atmosphere Dilution System. By maintain #ng a minimum of 2000 gallons of liquid N, in the storage tank it i; esured that a seven-day supply of N, for post-LOCA containment inerting is available. Since the inerting makeup synem is continually functioning, no
-193-
-a: .
dnit3
-PBAPS
~3.7.A & 4.7.A BASES (Cont'd)
Due to the nitrogen addition, the pressure in the containment after a LOCA will increase with time. Under the worst expected conditions, repressurization of the containment . will reach 30 psig. If and when that pressure is reached, venting from the containment shall be manually ' initiated. The venting path will be through the Standby Gas Treatment system in order to minimize the off site dose. Following a 'LOCA, periodic operation of the drywell and torus sprays will be used to assist the natural convection and diffusion mixing of hydrogen and oxygen. 1 4 s. _,6 q
< .f
)
-l i
.i!
h5 n
-195-
-I
-e ,-
APPENDIX B TO N FACILITY OPERATING LICENSE DPR-44 AND FACILITY OPERATING LICENSE DPR ENVIRONMENTAL TECHNICAL SPECIFICATIONS AND BASES FOR THE FULL POWER FUL!. TERM OPERATION OF PEACH BOTTOM ATOMIC POWER STATION i i UNIT 3
~
MAY 31, 1989 YORK COUNTY, PENNSYLVANIA . PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-278 e
- e.
- Unit 3 :;
' PBAPS
- 1. Protection Limit - A numerical limit on a plant effluent or '
operating parameter which, when not exceeded, should not- : resultfin an unacceptable environmental impact. :
- m. Rated Thermal Power - Rated thermal power refers to operation !
at a reactor power of 3458 MWt.
- n. Report Level - The numerical level of.an environmental pari- ,
meter below which the environmental impact is considered reasonable on the basis of available information. 7
- o. Special Study Program - An environmental study program designed to evaluate the impact of plant operation on an t environmental parameter.
- p. Total Residual Chlorine - The sum of the free chlorine and the combined chlorine.
1.2 ABBREVIATIONS .
- a. AEC - Atomic Energy Commission a i
- b. BWR - Boiling Water Reactor
- c. 10CFR20 - Code of Federal Regulations; -
Title 10 - Atomic Energy Part 20 - Standard for Protection Against Radiation n
- d. 10CFR50 - Code of Federal Regulations; Title 10 - Atomic Energy
- Part 50 - Licensing of Production and .
Utilization Facilities
- e. FSAR - Final Safety Analysis Report
- f. NEPA - National Environmental Policy Act
- g. MPC - Maximum Permissible Concentration ,
- h. MSL - Mean Sea Level i
- i. PBAPS - Peach Bottom Atomic Power Station Units No. 2 and 3 l j. POR - Plant Operations Review
- k. 9'&SR - Operation and Safety Review
- 1. PMF - Probable Maximum Flood
- m. PSAR - Preliminary Safety Analysis Report
.. ~.
r e t i ATTACHMENT 3 PEACH BOTTOM ATOMIC POWER STATION Units 2 and 3 , Docket Nos. 50-277 , 50-278 . License Nos. DPR-44 DPR-56 l NEDC-32183P, " Power Rerate Safety Analysis Report : for Peach Pottom 2 & 3," Class III, r,ted May 1993 1 Affidavit supporting General Electric Company's request to withhold NEDC-32183P'from public' disclosure. y e i
..J
= -
t GENERAL ELECTRIC COMPANY AFFIDAVIT I, ROBERT C. MITCHELL, being duly sworn, depose and state as follows: (1) I am Project Manager, Safety and Communications, General Electric Company ("GE") and have been delegated the ' function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding. (2) The information sought to be withheld is the entirety of GE proprietary report NEDC-32183P, " Power Rerate Safety Analysis Report for Peach Bottom 2 & 3", dated May, 1993. This document, taken as a whole, constitutes a proprietary compilation of information, some of it also independently proprietary, prepared by the General Electric Company. The independently proprietary elements are delineated by bars marked in the margin adjacent to the specific material. (3) In making this application for withholding of proprietary-information of which it is the owner, GE relies upon'the exemption from disclosure set forth in the Freedom'of ; Information Act ("FOIA"), 5 USC Sec. 552 (b) (4) , and the ; Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17 (a) (4) , 2.790 (a) (4) , and 2.790(d) (1) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is ; here sought is all " confidential commercial information", ; and some portions also qualify under the narrower definition of " trade secret", within the meanings assigned to those terms for purposes of FOIA' Exemption'4 in, respectively, Critical Mass Enercy Proiect v. Nuclear Reculatory Commission, 975F2d871 (DC Cir. 1992), and Public l Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983). (4) Some examples of categories of information which fit into the definition of proprietary information are: i
- a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over a
/ .
3 h other. companies; ,
- b. Information which, if used by a competitor, would l reduce his expenditure of resources or improve his :
competitive position in the design, manufacture, 1 shipment, installation, assurance of quality, or l licensing of.a similar product; !
- c. Information which reveals cost or price information, 'I production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers;
- d. Information which reveals aspects of past, present,' '
or future General Electric customer-funded development plans and programs, of potential l commercial value to General Electric;
- e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
Both the compilation as a whole and the marked independently proprietary elements incorporated in that compilation are considered proprietary for the reason - described in items-(4)a. and - (4 ) b. , above. , (5) The information sought to be withheld is being submitted to i NRC in confidence. That information (both the entire body , of information in the form compiled in this' document, and the marked individual proprietary elements) is of a sort customarily held in confidence by GE, and has, to the best of my knowledge, consistently been held in confidence by GE, has not been publicly disclosed, and is not available in public sources. All disclosures to third parties, including any required transmittals to NRC,'have been.made pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in + confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in (6) and (7) following. (6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and , sensitivity of the informationlin relation to industry ; knowledge. Access to such documents within GE is limited on a "need to know" basis. , (7) The procedure for approval of external release of such a , document typically requires review by the staff manager, { l i
l project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing. ; function (or his delegate), and by the Legal. operation, for technical content, competitive effect, and determination of : the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and ; potential customers, and their agents, suppliers, and > licensees, and others with a legitimate need for the i information, and then only in accordance with appropriate regulatory provisions or proprietary agreements. ; (8) The information identified by bars in the margin is classified as proprietary because it contains detailed ' results and conclusions from these evaluations, utilizing' . analytical models and methods, including computer codes, which GE has developed, obtained NRC. approval of,.and applied to perform evaluations of transient and accident events in the GE Boiling Water Reactor ("BWR"). The development and approval of these system, component, and thermal hydraulic models and computer codes was achieved at a significant cost to GE, on the order of several million - dollars. The remainder of the information identified in paragraph (2) is classified as proprietary because it constitutes a confidential compilation of information,. including detailed ' results of analytical models, methods, and processes, 3 including computer codes, and conclusions from these applications, which represent, as a whole, an integrated ; process or approach which GE has developed, obtained NRC approval of, and applied to perform evaluations of the safety-significant changes necessary to demonstrate the regulatory acceptability of a given increase in licensed power output for a GE BWR. The development and approval of this overall approach was achieved at a significant additional cost to GE, in excess of a million dollars, over and above the very large cost of developing the underlying individual proprietary analyses. To effect a change to the licensing basis of a plant requires a thorough evaluation of the impact of the change , on all postulated accident and transient events, and all other regulatory requirements and commitments included in the plant's FSAR. The analytical process to perform and document these evaluations for a proposed power uprate was developed at a substantial investment in GE resources and expertise. The results from these evaluations identify those BWR systems and components, and those postulated events, which are impacted by the changes required to accommodate operation at increased power levels, and, just as importantly, those which are not so impacted, and the technical justification for not considering the latter in
_4_ changing the licensing basis. The scope thus determined forms _the basis for GE's offerings to support utilities in both performing analyses and providing licensing consulting services. Clearly, the scope and magnitude of effort of any attempt by a competitor to effect a similar licensing change can be narrowed considerably based upon these results. Having invested in the initial evaluations and developed the solution strategy and process described in the subject document GE derives an important competitive advantage in selling and performing these services. However, the mere knowledge of the impact on each system and component reveals the process, and provides a guide to the solution strategy. (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GE's comprehensive BWR technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond_the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition,_the technology base includes the value derived from providing analyses done with NRC-approved methods, including justifications for not including certain analyses in applications to change the licensing basis. GE's competitive advantage will be lost if its competitors are able to use the results of the GE experience to avoid fruitless avenues, or to normalize or verify their own process, or to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions. In particular, the specific areas addressed by any document and submittal to support a change in the safety or licensing bases of the plant will clearly reveal those areas where detailed evaluations must be performed and specific analyses revised, and also, by omission, reveal those areas not so affected. While some of the underlying analyses, and some of the gross structure of the process, may at various times have been publicly revealed, enough of both the analyses and the detailed structural framework of the process have been held in confidence that this information, in this compiled form, continues to have great competitive value to GE. This value would be lost if the information as a whole, in the context and level of detail provided in the subject GE document, were to be disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of
4 5
+
resources, including that required to determine the areas that are not affected by a power uprate and are therefore blind alleys,'would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to' seek an adequate return on its , _large investment in developing its analytical process. , STATE OF CALIFORNIA ) '!
) SS:
COUNTY OF SANTA CLARA ) , _ Robert C..Mitchell, being duly sworn, deposes and says: That he has read the foregoing affidavit and'the matters stated therein are true and correct to the best of his knowledge, information, and belief. Executed at San Jose, California, this /> day of M A/ 6 , 1911 bLJ1 C %t G As _R Q ~ Robert C.'Mitchell General Electric Company Subscribed and sworn before me this day-of LAL- , 19 0 0f4 ft_ $]" ~ 'f ggqq] Notary PubIic, StatepfCalifornia h PAULA F. HUSSEY tjoy&v (* ic - CAUTC;W A j SANTA CW A COUNTY : My comm. expires APR 5,1994 } _ _ - ~ , - ~ ~ *
- b i
6/14/93RTH f
-- -- ________________________________________________________.___.____.____.___}}