ML20087A789
| ML20087A789 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 07/18/1995 |
| From: | Russell W NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20087A791 | List: |
| References | |
| NUDOCS 9508070172 | |
| Download: ML20087A789 (34) | |
Text
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UNITED STATES p*
4
. j j
NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20666-0001 o
't,.....,d PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-278 PEACH BOTTOM ATOMIC POWER STATION. UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 211 License No. DPR-56 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Philadelphia Electric Company, et al.
)
(the licensee) dated June 23, 1993, as supplemented by letters dated April 5, May 2, June 6, June 8, July 6 (two letters), July 7, July 20, July 28 (two letters), September 16, September 30, and October 14, i
1994 and June 22, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I.
B.
The facility will reperate in conformity with the application, the j
provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health or safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been t
satisfied, 9508070172 950718 DR ADOCK 050002 8 i
^
. 2.
Accordingly, Facility Operating License No. OPR-56 paragraph 2.C.(1) is hereby amended to read as follows:
(1) Maximum Power level PECO Energy Company is authorized to operate the Peach Bottom Atomic Power Station, Unit 3, at steady state reactor core power levels not in excess of 3458 megawatts thermal.
3.
Further, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-56 is hereby amended to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 211, are hereby incorporated in the license. PECO Energy Company shall operate the facility in accordance with the Technical Specifications.
4.
This license amendment is effective as of its date of issuance and is to be implemented prior to startup in Cycle 11 currently scheduled for October 1995.
FOR THE NUCLEAR REGULATORY COMMISSION h
-A William T. Russell, Director Office of Nuclear Reactor Regulation Attachments:
1.
Page 3 of License
- 2.
Changes to the Technical Specifications Date of Issuance: July 18, 1995
- Page 3 is attached, for convenience, for the composite license to reflect this change.
ATTACHMENT TO LICENSE AMENDMENT NO. 211 FACILITY OPERATING LICENSE NO. DPR-44 DOCKET NO. 50-277 Replace the following pages of the Facility Operating License (FOL), the Appendix A Technical Specifications, and the Appendix B Environmental Technical Specifications, with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Insert FOL 3
3 Appendix A 2
2 6
6 9
9 11 11 16 16 17 17 18 18 24 24 29 29 30 30 37 37 39 39 40 40 49 49 50 50 73 73 74 74 117 117 129 129 130 130 137 137 140a 140a 140c 140c 157 157 164d 164d 189 189 193 193 195 195 Appendix B Cover sheet Cover sheet 2
2 i
i t
(5) PECO Energy Company, pursuant to the Act and 10 CFR Parts 30 and 70, i
to possess, but not to separate, such byproduct and special nuclear i
material as may be produced by operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:
Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections i
50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and 2'
orders of the Commission now or hereafter in effect; and is subject to i
i the additional conditions specified below:
i j
(1) Maximum Power Level PECO Energy Company is authorized to operate the Peach Botton Atomic i
Power Station, Unit 3, at steady state reactor core power levels not in excess of 3458 megawatts thermal.
(2) Technical Soecifications j
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
are hereby incorporated in the license.
PECO Energy Company shall operate the facility in accordance with the Technical Specifications.
l (3) Physical Protection i
l The licensee shall fully implement and maintain in effect all provisions of the Comission-approved physical security, guard i
l 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 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards j
Information protected under 10 CFR 73.21, are entitled: " Peach Bottom Atomic Power Station, Units 2 and 3, Physical Security Plan,"
with revisions submitted through December 16, 1987; " Peach Bottom i
Atomic Power Station, Units 2 and 3 Plant Security Personnel Training and Qualification Plan," with revisions submitted through i
July 9,1986; and " Peach Bottom Atomic Power Station, Units 2 and 3 Safeguards Contingency Plan," with revisions submitted through i
March 10, 1981. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
(4) The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility, and as approved in the NRC SER dated May 23, 1979 and Supplements dated August 14, 1
September 15, October 10 and November 24, 1980, and in the NRC SERs dated September 16, 1993 and August 24, 1994, subject to the following provision:
Page 3 Amendment No.
i 17, 53, 138, 198, 201, 211 f
y
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Unit 3 PBAPS
]
1.0 DEFINITIONS (Cont'd) d Engineered Safeauard - An engineered safeguard is a safety system the actions of which are essential to a safety action required in response to accidents.
e Fraction of limitino Power Density (FLPD) - The ratio of the linear heat generat-ion rate (LHGR) existing at a given location to the design LNGR for that bundle type.
I 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 affluents 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.
i High (power) Trio Set Point (HPTS) - The high power trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting i
appitcable 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 coolant 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.
Immediate - Inmediate means that the required action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.
- Amendment No. 104, 184, 211
O Unit 3 o
~
PBAPS 1.0 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.
Purae - Purcina - 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 ODeration - 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 j
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.
Refuelina Outace - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the unit after that refueling.
For the purpose of designating i
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 See the term " Refuel" under the Definition of " Surveillance Frequency" for specific time limits on surveillances with a frequency that includes the term " Refueling Outage."
i Amendment No. IO#, 182, 211
Unit 3 PBAPS SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Acolicability:
Aeolicability:
l 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 which are variables which monitor the provided to prevent the fuel fuel thermal behavior, cladding integrity Safety Limits from being exceeded.
Obiectives:
Obiectives:
The objective of the Safety Linits 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 Specification:
integrity Safety Limits from being exceeded.
A. Reactor Pressure > 800 osia and Core Flow 1 10% of Rated Joecification:
The existence of a minimum The limiting safety system settings critical power ratio (MCPR) shall be as specified below:
less than 1.07 for two recirculation loop operation, A.
Neutron Flux Scram or 1.08 for single loop operation, shall constitute
- 1. APRM Flux Scram Trio Settina 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 0.66W + 66% - 0.66 4W l
specification 2.1. A for longer (Clamp 9120%)
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 l be assumed to be exceeded if thermal power (3458 MWt) the neutron flux exceeds its scram setting and a control W=
Loop recirculating flow rate rod scram does not occur.
in percent of design.
i
~9-Amendment No. 14, 41, 77, 79, !%0, 159, 183, 184, 211
Unit 3 PBAPS SAFETY LIMIT LIMITINC SAFETY SYSTEM SETTING 3.
Core Tharumi Power Limit B.
APRM Rod Block Trin Settinn (Reactor Pressure s 800 psia)
~
When the reactor pressure is S
s (0.66 W + 544 - 0.66 av) l u
s 800 psia or core flow is -
(Clamp 6 1086) 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 MWe) l V-Isop recirculation flow race 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 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 1
setpoint and recirculation drive flow or by adjusting the APRM Rod block trip setting.
1 av - O for two loop operation.
The APRM rod block trip setting shall not exceed 108% of rated thermal power.
J
-11 Amendment No. 14, 33, 41, 62, 77, 150, IB4, 211 l
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i 130 f _ _APR_M F_ lux serem _ _ _
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Relationship to Normal Operating Conditions Fignre 1.1-1 5
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r Unit 3 PBAPS r
2.1 8ASES
FUEL CLADDING INTEGRITY t
i The abnormal operational. transients applicable to operation of 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 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 Botton Atomic Power l
Station Unit, and this represents the maximum steady state power which shall not knowingly be exceeded.
(See Reference 6).
l Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scras worth, scram delay time, peaking factors, and axial power shapes. These fac-tors are selected conservatively with respect to their effect on the applicable transient results as determined by the current analysis model. Conservatism incorporated into the transient analysis is documented in References 4 and 5.
l
- Amendment No. 33, 79, 150, 159 211 v
Unit 3 2.1 BASES (Cont'd)
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 conf #UVativa 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 i
permitted.
The analysis to support operation at various power and flow relationships has considered operation with either one I
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.
1 11.
The licensed maximum power level is 3458 MWt.
l iii. Analyses of transients employ adequately conservative values of the contrclling reactor parameters.
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 (reactor thermal power) is less than 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 scram trip setting, none of the abnormal operational transients analyzed viola.te 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. Amendment No. 33, 4I, 42, 62, 150, 211
~ _ -
.. ~ _
Unit 3
2.1 BASES
(Cont'd)
~
L. References
- 1. Linford, R. 8., " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," NEDO 10802, February 1973.
- 2. " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors", NEDO 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 NEDD-24154 and NEDE 24154-P, Volumes I, II, and III.
- 4. " General Electric Standard Appifcation for Reactor Fuel",
NEDE-24011-P-A(asamended).
i
- 5. " Methods for Performing BWR Reload Safety Evaluations,"
PECo-FMS-0006-A (as amended).
- 6. " Power Rerate Safety Analysis Report for Peach Botton 2 & 3,"
NEDC-32183P, May 1993.
l Amendment No. IA, 79, !$9, 211 l
(
unit 4 LIMITING SAFETY SYSTEM SETTING SAFETY LIMIT 1.2 REACTOR COOLANT SYSTEM INT.EGRITY 2.2 EEACTOR C00LA E }H U,g IPlfd8111 Anolicabilitv:
Aeolicabilitv:
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.
Obiectives:
Qbinctives:
To establish a limit below I
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.
hification:
Soecification:
- 1. The reactor vessel dome pressure shall not exceed
- 1. The limiting safety system 1325 psig at any time when settings shall be as 4
irradiated fuel is present specified below:
in the reactor vessel.
Protective Action /Limitina Safety System Settina i
A. Scram on Reactor Vessel high pressure 51085 psig l
B. Relief valve settings 1135 psig ( 11 psi) l (4 valves) 1145 psig (211 psi) l (4 valves) 1155 psig ( 12 psi) l (3 valves) Amendment No. U, 211 i
Unit 3 SAFETY LIMIT LIMITING SAFETY SYSTEM j
SETTING I
2.
The reactor vessel done pressure shall not ex-1260 psig
- 13 psi l
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-tor vessel done pressure is >75 psig.
. Amendment No. 41, 211
PBAPS Un.it 3 o Table 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Minimum No.
Modes In which Number of of Operable Function Must Be Instrument Instrument Trip Level Setting Operable Channels Action Channels Trip Function Provided by (1)
Refuel Startup Run Design per Trip Items System (7) 1 1
Mode Switch In x
x x
1 Mode Switch A
Shutdown (4 Sections) 2 1
x x
2 Instrument A
Channels d.
3 3
IRM High Flux s120/125 of Full x
x (5) 8 Instrument A
l' Scale Channels 4
3 IRM Inoperative x
x (5) 8 Instrument A
Channels 5
2 APRM High Flux (0.66W+66%-0.66aW) x 6 Instrument A or B (Clamp 9 120%)
Channels (12) (13) 6 2
APRM (11) x x
x 6 Instrument A or B g
Inoperative Channels 0
7 2
APRM Downscale 22.5 Indicated (10) 6 Instrument A or B k
on Scale Channels a
8 2
APRM High Flux s15% Power x
x 6 Instrument A
in Startup Channels gg 9
2 High Reactor
$1085 psig x(9) x x
4 Instrument A
y-Pressure Channels l
.k.E N
10 2
High Drywell
$2 psig x(8) x(8) x 4 Instrument A
Pressure Channels
!3' 11 2
Reactor low 20 in. Indicated x
x x
4 Instrument A
~O Water Level Level Channels M
= -
Unit 3 PBAPS i
NOTES FOR TABLE 3.1.1 1.
If the required actions and associated completion time of specification 3.1.A, Actions 1 or 2 or 3 are not met, take the action listed below for the affected trip function as I
required by Table 3.1.1.
A.
Initiate insertion of operable rods and complete insertion of all operable rods within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
B.
Reduce power level to IRM range and place mode switch i
in the start up position within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
C.
Reduce turbine load and close main steam line isolation valves within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D.
Reduce power to less than 30% rated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
2.
Permissible to bypass, in refuel and shutdown positions of
)
the reactor mode switch.
3.
Deloted.
4.
Bypassed when turbine first stage pressure is less than that i
which is equivalent to 30% of rated thermal power.
5.
IRMs are bypassed when APRMs 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.
i 7.
When the reactor is subcritical and the reactor water j
temperature is less than 212 degrees F, only the following i
trip functions need to be operable A.
Mode switch in shutdown j
B.
Manual scram C.
High flux IRM D.
Scram discharge instrument volume high level 8.
Not required to be operable when primary containment integrity is not required.
9.
Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.
l Amendment No. 33 208,108211 121, 193,
Unit 3 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 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.1.
i i Amendment No. 33, 41, 62, 77, 79, 106, 132, 150, 155, 184, 211 l
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 insertion.
To preclude this occurrence, level switches have been provided in the instrument volume which alans 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 perfone 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 functior.s (reference paragraph 7.5.4 FSAR).
l Thus, the IRM and APRM are required in the " Refuel" and t
" Start / Hot Standby" modes.
In the power range the APRM system provides required protection (reference paragraph 7.5.7 FSAR).
Thus the IRM Systes 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.
i The high reactor pressure, high drywell pressure, reactor low water level and scram discharge volume high level scrans 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 scrans are adequate to protect the reactor.
l
! Amendment No. 121, 193, 211
Unit 3 3.1 BASES (Cont'd.)
Turbine control valves fast closure initiates a scram bared 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, a6d are set relative (500<P<850 psig) to the normal EHC oil 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 IRN's be inserted in the core when the APRN'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 l
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1 Amendment No. 193, 211 j
Unit 3 TABLE 3.2.C INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS Minimum No. of Operable Number of Instrument Channels instrument Channels Per Trip System instrument Trip Level Setting Provided by Design Action 4
(2)
APRM Upscale (Flow Biased)
(0.66W+ 54%0.66aW) 6 Inst. Channels (10) (14)
(Clamp at 108% max) 4 APRM Upscale (Startup Mode) s 12%
6 Inst Channels (10) (14) 4 APRM Downscale a 2.5 indicated on scale 6 Inst. Channels (10) (14) 1 (7) (11) (13)
Rod Block Monitor (RTP 285%), S. sHTSP 2 Inst. Channels (12) (14)
(Power Biased)
(65% sRTP <85%), S. sITSP (30% sRTP <65%), S sLTSP 1
(7) (11) (13)
Rod Block Monitor Downscale aDTSP 2 Inst. Channels (12) (14) 6 IRM Downscale (3)
= 2.5 indicated on scale 8 Inst. Channels (10) 6 IRM Detector not in Startup Position (8) 8 Inst. Channels (10) 6 1RM Upscale s 108 indicated on scale 8 Inst. Channels (10) g B
(
2 (5)
SRM Detector not in Startup (4) 4 Inst. Channels (1)
Position j
g 2 (5)(6)
SRM Upscale s 10' counts /sec.
4 Inst. Channels (1) l t
N~w 1
(15)
Scram Discharge Instrument Volume s 25 gallons 1 inst. Channel (9)
High Level U "."
l ~.%
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l PBAPS Unit 3 i
i NOTES FOR TABLE 3.2.C 1.
If the first column cannot be met for one of the two trip systems, this condition may l
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).
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 flow.
t 4W = 0 for two loop operatien.
3.
IRM downscale is bypassed when it is on its lowest range.
4.
This function is bypassed when the count rate is a 100 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.
1 l Amendment No. 33, (I, 62, 77, 79, 150, Iss, 184, 206, 211
Unit 3 PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.4 STANDBY LIOUID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL SYSTEM (Cont'd)
(Cont'd) 3.
The Standby Liquid Control System conditions must satisfy the following equation:
/
C 0
h E
)
(13%wt.
86gpmj 19.8% atom ~j,
- where, C-Sodium Pentaborate Solution Concentration (% weight)
Q-Pump Flow Rate (gpm) 3.
Pump Flow Rate: At least against a system head of once per 92 days each pump l
1255 psig.
loop shall be functionally 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.
E-Boron-10 Enrichment (% atom 4.
Enrichment:
Following each Boron-10) addition of boron to the solution tank, calculate enrichment within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Verify results by analysis within 30 days.
5.
Solution Volume: At least once per day check and record.
-117-Amendment No. I26, 201, 211 l
i Unit 3 PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.5.C HPCI Subsystem (cont'd.)
4.5.C HPCI Subsystem (cont'd.)
11.3m Frecuency (b) Pump once/ month Operability (c) Motor Operated Once/ month Valve Operability (d) Flow Rate at once/3 months approximately 1030 peig l
Reactor Steam Pressure (e) Verify, with Once/ operating reactor pressure cycle sl75 peig, the HPCI pump can develop a flow rate 25000 gpm against a system head correspod-ing to reactor pressure.*
- 2. From and after the date that The HPCI pump shall deliver the HPCI Subsyrtem is made or at least 5000 gpm for a system found to be inoperable for head corresponding to a reactor any reason, continued reactor pressure of approximately 1030 to l
operation is permissible only 150 psig.
during the succeeding seven days unless such subsystem is
- 2. DELETED 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 cannet be met, an orderly shut-down shall be initiated and the reactor shall be in a cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test.
-129-Amendment No. 162, 202, 211
~
Unit 3 PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.D Reactor Core Isolation.
4.5.D Reactor Core Isolation Cooling (RCIC) Subsystem Cooling (RCIC) Subsystem 1.
The RCIC Subsystem shall be 1.
RCIC Subsystem testing shall operable whenever there is be performed as follows:
irradiated fuel in the reactor vessel, the reactor steam pressure Item Frequency is greater than 105 peig, and prior to reactor startup from (a) simulated once/ operating a Cold condition, except as Automatic Cycle specified in 3.5.D.2 below.
Actuation Test-(D) Pump Once/ Month operability (c) Motor operated once/ Month valve operability (d) Flow Rate at once/3 Months approximately 1030 psig Reactor steam Pressure **
(e) Flow Rate at once/ operating approximately Cycle 150 peig i
Reactor Steam I
Pressure **
(f) verify auto-once/ operating"*
matic transfer Cycle from CST to suppression pool 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, cortinued reactor power opera-tion is permissible only during the succeeding seven days
- Shall include automatic restart provided that during such on low water level signal.
seven days the NPCI Subsystem is operable.
- The RCIC pump shall deliver at least 600 gym 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 l
j down shall be initiated and 150 peig.
4 the reactor pressure shall be reduced to 105 peig within
- Iffective at let refueling outage 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
after Cycle 7 reload.
-130-Anendment No. 49,102, II7, I62, 211
PBAPS Unit 3 3.5 BASES (cont'd.)
C.
HPCI The limiting conflitions 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 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 reactor to be shut down while maintaining sufficient reactor vessel 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.
The capacity of the system is selected to provide this re-quired core cooling. The HPCI pump is designed to pump l
5000 gpa at reactor pressures between 1150 and 150 psig.
l Two sources of water are available. Initially, dominera-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 Sys tem. 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-i 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 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.
-137-Amendment No. 211
i Unit 3 PSAPS 3.5 BASES (Cont'd)
J.
Local LHGR l
This specification assures that the linear heat generation rate in any 8X8 fuel I
rod is less than the design linear heat generation. The maximum LHGR shall be I
2 checked daily during reactor operation at 254 power to determine if fuel burnup, or control rod movement has caused changes in power distribution. For l
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 LNGR
.which is precluded by a considerable margin when employing any permissible control rod pattern.
K.
Minimus Critical.fower Ratio (MCPR)
Operatino 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 l-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 evaluated are as l
described in References 7 and 10.
-140a-Amendment No. 33, 41, 42, 62, 79, 155, 159, 211 i
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-Coolant Accident or applicable Abnomal Operational Transients demonstrates that the times required to initiate corrective action (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) and restore the calculated value of APLHGR, LHGR or MCPR to within prescribed Itaits (5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) 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 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).
3.
Communication:
V. A. Moore to I. S. Mitchell, " Modified GE Model for Fuel Densification", Docket 50-321, March 27, 1974.
t 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 Botton Atomic Power Station Units 2 and 3 SAFER /GESTR - LOCA Loss-of-Coolant Accident Analyses," NEDC-32163P, January,1993.
9.
DELETED.
10.
" Methods for Perfoming 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 Staticn Units 2 and 3," NEDC-32162P, Revision 1, February, 1993.
'12.
" Power Rerate Safety Analysis Report for Peach Botton 2 & 3,"
NEDC-32183P, May 1993.
-140c-Amendment No. 33, 34, #1, #2, 62, 79, 150, I59, IB#,
211
Unit 3 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 system 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 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 show 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.
Eleven safety / relief valves and two safety valves have been installed on Peach Bottom Unit 3 with a total capacity of 75.30%
l 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 4
analysis of the plant isolation transient shows that the 11 safety / relief 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 cycle is adequate to detect failure or deteriorations.
The safety / relief and safety valves are benchtested every second i
-157-Amendment No. 33, 35, 41, 42, 62, 79, 182, 211
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Unit 3 PBAPS 3.7.A/4.7.A BASES Primary Containment The integriti 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 l
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 t
is required.
There will be no pressure on the system at this time, 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 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.
i Using the minimum or maximum water volumes given in the speci-fication, containment pressure during the design basis acci-dont is approximately 49.1 psig which is below the maximum 3
of 62 psig.
Maximum water volume of 127,300 ft results in a downcomer submergence of 4.4 feet and the minimum volume 2
of 122,900 ft results in a submergence approximately 0.4 4
feet less.
-189-Amendment No. 40, 211
l Unit 3 1
PBAPS 3.7.A & 4.7.A BASES (Cont'd.)
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 l
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 zone distance of 7300 meters are about 2.5 REM total whole body and 105 REM total i
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 i
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 bases loss-of-coolant accident. These doses are also based on the assumption of no holdup in the secondary containment i
resulting in a direct release of fission products from the primary containment through the filters and stack to the environs. Therefore, the specified i
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 adequate heat removal capability is present.
Drywell Interior The interiors of the drywell and suppression chamber are painted to prevent rusting. The inspection of the paint during each major refueling outage, 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 flammable limit, the capability to inject nitrogen into the containment after a LOCA is provided.
During the first year of operation the normal inerting nitrogen makeup system 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 Amendment No. 182, 211
Unit 3
~
PBAPS 3.7.A & 4.7.A BASES (Cont'd)
Due to the nitrogen addition, the pressure in the containment after a LOCA wi.11 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.
-195-Amendment No. 23, 211
APPENDIX 8 TO FAC-ILITY OPERATING LICENSE DPR-44 AND FACILITY OPERATING LICENSE DPR-56 ENVIRONMENTAL TECHNICAL SPECIFICATIONS AND BASES FOR THE FULL POWER FULL TERM OPERATION OF PEACH BOTTOM ATOMIC POWER STATION UNIT 3 MAY 31, 1989 YORK COUNTY, PENNSYLVANIA PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-278 l
Amendment No. 25, 211
Unit 3 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.
l n.
Report Level - The numerical level of an environmental para-meter below which the environmental impact is considered 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 a.
AEC - Atomic Energy Comission b.
8WR - 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; 1
Title 10 - Atomic Energy Part 50 - Licensing of Production and i
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.
PBAPS - Peach Botton Atomic Power Station Units No. 2 and 3 j.
POR - Plant Operations Review k.
0&SR - Operation and Safety Review 1.
PMF - Probable Maximus Flood m.
PSAR - Preliminary Safety Analysis Report
! Amendment No. 211 j
..