ML20072C750
| ML20072C750 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 08/10/1994 |
| From: | Thadani M Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072C753 | List: |
| References | |
| NUDOCS 9408180081 | |
| Download: ML20072C750 (32) | |
Text
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UNITED STATES T
- ,f NUCLEAR REGULATORY COMMISSION if,g-
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yi WASHINGTON, D.C. 20555-0001 PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND CAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET N0. 50-277 PEACH BOTTOM ATOMIC POWER STATION. UNIT NO. 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 192 License No. DPR-44 1.
The Nuclear Regulatory Commission (the Conunission) has found that:
A.
The application for. amendment by Philadelphia Electric Company, et al. (the licensee) dated April 1,1993, and supplemented by letters dated April 7, July 16, and August 20, 1993, and June 8, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I.
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) 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 Commission'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 Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, 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:
i 9408100081 940810 PDR ADOCK 05000277 P
PDR i
l
- (2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 192, are hereby incorporated in the license.
PEC0 shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective following startup from Refueling Outage 2R10.
FOR THE NUCLEAR REGULATORY COMMISSION CMLW tatt Mohan C. Thadani, Acting Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 10, 1994' d.
i
f ATTACHMENT TO LICENSE AMENDMENT NO. 192 FACILITY OPERATING LICENSE NO. OPR-44 DOCKET NO. 50-277 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Remove Insert 1
1 2
2 3
3 4
4 9
9 9a 9a 10 10 11 11 lla lla 15 15 16 16 19 19 20 20 35 35 37 37 40 40 54 54 73 73
)
74 74 74a 74a 133a 133a
Unit 2, DRP-44 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised areas are indicated by marginal lines.
Remove Insert 133b 133b 140 140 140b 140b 140c 140c 141a 141a 141b 141b 256 256
Unit 2 PBAPS 1.0 DEFINITIONS The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieved.
Alteration of the Reactor Core - The act of moving any component in the region above the core support plate, below the upper grid and within the shroud with the vessel head removed and fuel in the vessel.
Normal control rod movement with the control drive hydraulic system is not defined as a core alteration.
Normal movement of in-core instrumentation and the traversing in-core probe is not defined as a core alteration.
Averace Planar Linear Heat Generation Rate (APLHG_R_). - The APLHGR shall be applicable to a specific planar height and is equal to the sum of the heat gen-eration rate per unit length of fuel rod, for all the fuel rods in the specific bundle at the specific height, divided by the number of fuel rods in the fuel bundle at that height.
Channel - A channel is an arrangeznt of a sensor and associated components used to evaluate plant variables and produce discrete outputs used in logic. A channel terminates and loses its identity where individual channel outputs are combined in logic.
Cold Condition - Reactor coolant temperature equal to or less than 212 F.
Cold Shutdown - The reactor is in the shutdown mode, the reactor coolant temperature equal to or less than 212 F, and the reactor vessel is vented to atmosphere.
Core Operatino Limits Reoort (COLR) - The COLR is the unit-specific document that provides the core operating limits for the current Operating Cycle. These cycle-specific core operating limits shall be determined for each Operating Cycle in accordance with specification 6.9.1.e.
Plant operation within these limits is addressed in individual Specifications.
Critical Power Ratio (CPR) - The critical power ratio is the ratio of that assembly power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor condition of interest as calculated by application of the GEXL correlation.
(Reference NED0-10958).
Dose Eouivalent I-131 - That concentration of I-131 (Ci/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134, and 1-135 actually present.
Downscale Trio Set Point (DTSP) - The downscale trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting. Amendment No. 191, 171, 17b I H, 192
Unit 2 PBAPS 1.0 DEFINITIONS (Cont'd)
Encineered Safecuard - An engineered safeguard is a safety system the actions of which are essential to a safety action required in response to accidents.
Fraction of Limitino 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 l'
providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
Hich (oower) Trio 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 coolant temperature greater than 212 F, system pressure less than 1055 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 of the unit and the importance of the required action. Amendment No. 102,192 v
Unit 2 i
PBAPS ~
l 1.0 DEFINITIONS (Cont'd)
Instrument or Channel Calibration - An instrument or channel calibration means the adjustment of an instrument or channel signal output so that it corresponds, within acceptable range, and accuracy, to a known value(s) of the parameter which the instrument or channel monitors. The known value of the parameter shall be injected into the channel or instrument as close to the primary sensor as practicable.
Instrument or Channel Check - An instrument or channel check is a qualitative determination of acceptable operability by observation of instrument or channel behavior during operation.
This determination shall include, where possible, comparison of the instrument or channel with other independent instruments measuring the same variable.
Instrument or Channel Functional Test - An instrument or channel functional test means the injection of a simulated signal into the channel or instrument as close to the primary sensor as practicable to verify the proper instrument channel response, alarm and/or initiating action.
Intermediate (Dower) Trio Set Point (ITSP) - The intermediate power trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting applicable between 65% and 85% reactor thermal power.
Limitina Conditions for Ooerations (LCO) - The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility. When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled.
Limitina safety System Settino (LSSS) - The limiting safety system settings are settings on instrumentation which initiate the automatic protective action at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represents margin with normal operation lying below these settings. The margin has been established so that with proper operation of the instrumentation, the safety limits will never be exceeded.
Loaic - A logic is an arrangement of relays, contacts and other components that produces a decision output.
. Amendment No. N 2, 192 n
v
+
Unit 2 PBAPS 1.0 DEFINITIONS (Cont'd)
(a)
Initiatina - A logic that receives signals from channels and produces decision outputs to the actuation logic.
(b) Actuation - A logic that receives signals (either from initiation logic or channels) and produces decision outputs to accomplish a protective action.
Loaic System Functional Test - A Logic System Functional Test shall be a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements etc., of a logic circuit, from sensor through and including the actuated device, to verify Operability. The Logic System Functional Test may be performed by any series of sequential, overlapping or tota' system steps such that the entire logic system is tested.
Low (oower) Trio Set Point (LTSP) - The low power trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting applicable between 30% and 65% reactor thermal power.
L4APFAC(F) (MAPLHGR Flow Factor) - A core flow dependent multiplication factor used to flow bias the standard Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit.
MAPFAC(P) (Power Decendent MAPLHGR Multiolier) - A core power
~
dependent multiplication factor used to power bias the standard Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit.
g MEMBERS OF THE PUBLIC - Members of the public shall include all persons who are not occupationally associated with the plant.
This category does not include employees of the utility, its contractors, or. vendors.
Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.
Minimum Critical Power Ratio (MCPR) - The minimum in-core critical power ratio corresponding to the most limiting fuel assembly in the core.
Associated with the minimum critical power ratio is a core flow dependent (MCPR(F)) and core power dependent (MCPR(P)) minimum critical power ratio.
Mode of Operatior. - A reactor mode switch selects the proper interlocks for tie operational status of the unit. The following i
are the modes anc interlocks provided: Refuel Mode, Run Mode, l
Shutdown Mode, Startup/ Hot Standby Mode. Amendment No. 102, 178, 192
Unit 2 l
PBAPS SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY 1
Aeolicability:
Acolicability:
')
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
- l fuel thermal behavior, cladding integrity Safety Limits from being exceeded.
Ob.iectives:
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 Soecification:
integrity Safety Limits from being exceeded.
A. Reactor Pressure > 800 osia and Core Flow > 10% of Rated Soecification:
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 Settino 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 + 71% - 0.66 AW specification 2.1. A for longer (Clamp 9 120%)
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 be assumed to be exceeded if thermal power (3293 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. Amendment No. IB, 24, 42, 48, 78, 86,123, Z57,152,192
- f Unit 2 PBAPS
-SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY AW =
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 AW) is. accomplished l
by correcting the flow input of the flow biased scram to preserve the-original (two loop) relationship between APRM scram setpoint:and recirculation drive flow or by adjusting the APRM flux trip setting.
AW = 0 for two loop. operation.
The APRM flux scram trip setting shall not exceed 120% of rated thermal power.
I
-9a-Amendment No. 192
Unit 2 PBAPS SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A (Cont'd) i l
2.
APRM--When the reactor mode switch is in the STARTUP position, the APRM scram shall be~ set at less than or equal to 15 percent of rated power.
3.
IRM--The IRM scram shall be set at less than or equal-to 120/125 of full scale.
-; Amendment No. 23, 34, 36, #2,
- S, 70, 78, 123, t
72), 154,192
+
4 0
't r.---w.
Unit 2 PBAPS
' SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING B.
Core Thermal Power limit B.
APRM Rod Block Trio Settino (Reactor Pressure s 800 psia)
When the reactor pressure is Su 5 (0.66 W + 59% - 0.66 AW)
$ 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.
Su = Rod block setting in percent of rated thermal power (3293 MWt)
W = Loop recirculation flow rate in percent of design.
l 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 setting (-0.66 AW) is l
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.
AW = 0 for two loop operation.
The APRM rod block trip setting shall not exceed 108% of rated thermal power.
, Amendment No. 23, 34, #2,
- S, 70, 78, 123, 192
~
m
Unit 2 PBAPS SAFETY' LIMIT LIMITING SAFETY SYSTEM SETTING B.
Core Thermal Power Limit B..APRM Rod Block Trip Setting (Reactor Pressure s 800 psia)
C.
Scram and isolation-2 538. in.
reactor low water above vessel C.
Whenever the reactor is in level' zero (0 on the shutdown condition with level irradiated fuel in the reactor instruments) vessel, the water level shall
.not be less than minus 160 inches indicated level (378 inches above vessel zero).
-11a-
'AmendmentNo.78,III,123,16, 192
Unit 2
~
PBAPS 1.1.C BASES (Cont'd.)
However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a Safety Limit, provided scram signals are operable, is supported by the extensive plant safety analysis.
The computer provided with Peach Bottom Units 2 and 3 has a sequence i
annunciation program which will indicate the sequence in which events such as a scram, APRM trip initiation, pressure scram initiation, etc. occur. This program also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient.
Thus, computer information
.l normally will be available for analyzing scrams; however, if the computer information should not be available for any scram analysis, Specification 1.1.C will be relied upon to determine if a Safety Limit has been violated.
D.
Reactor Water Level (Shutdown Condition)
During periods when the reactor is shutdown, consideration must also be given to water level requirements due to the effect of decay heat.
If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core i
can be cooled sufficiently should the water level be reduced to two-thirds the core height.
Establishment of the safety. limit at minus 160 inches indicated level (378 inches above vessel zero) provides adequate margin to assure sufficient cooling during shutdown conditions. This level will be continuously monitored.
E.
References 1.
General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, January 1977 (NED0-10958-A).
2.
Process Computer Performance Evaluation Accuracy, General Electric Company BWR Systems Department, June 1974 (NEDD-20340).
3.
" General Electric Standard Application for Reactor Fuel",
NEDE-24011-P-A (as amended).
4.
" Peach Bottom Atomic Power Station Units 2 and 3 Single-Loop Operation", NEDO-24229-1, May 1980.
S.
" 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.
' Amendment No. 23, 36, #8, 70, III,123,192 4
130 Power / Flow points APRM Flux Scram 120~
- 100Pn5F
.e f
- 100P/100F 110 -
APRM Rod Block
- 100P/105F
/
- 70P/110F
,/' @
100 -
@: 65P/104F 18
@
- 50P/106F
,/
m
/
p' 90 -
s'
/
~
Q:
y
/
MELLL 80 -
/,/
Rc8 "
i j/
~
O 0.66W + 71%
j
/
\\
/
100% Rod Line 70 -
g y
? 4
/
Increase /
d en E h 60 -__ _ _ og6_W + 52%_ _ _ _
siana, r,<iooo
=,,ie.
Core D
i ui,,a no,io Fiow 1 L --.
- d**"'""
c:
Rcgion 50 --
5""'"'*""' "*
~
h Min Pump y
40 -
/ speed (20%)
con" #
w ginim"* go OU 30 -
Natural Circulation /
00*C c #,4c
~
e 20 -
go ?
c,,;,,gon li letP**Psucson 10 -
mo 0
(
i 0
10 20 30 40 50 60 70 80 90 100 110 120 a"
o CORE FLOW (%)
~
El Relationship to Normal Operating Conditions
- 8 Figure 1.1-1
Unit 2
'I PBAPS 2.1.A BASES (Cont'd)
An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurious scrams which have an adverse,
I effect on reactor safety because of the resulting thermal stresses.
Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams.
Analyses of the limiting transients show that no scram adjustment is required to assure MCPR greater than the fuel cladding integrity safety limit when the transient is initiated from MCPR greater than the operating limit given in Specification 3.5.K, adjusted for power and flow as specified in the COLR.
For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provides adequate thermal mar-gin between the setpoint and the Safety Limit, 25 percent of rated. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup.
Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod pat-terns are constrained to be uniform by operating procedures backed up by the Rod Worth Minimizer. Worth of individual rods is very low in a uniform rod pattern.
Thus, of all possible sources of reactivity input, uniform control rod with-drawal is the most probable cause of significant power rise.
Because the flux distribution ' associated with uniform rod withdrawals dos not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of change of power is very slow. Generally, the heat flux is in near equilibrium with the fission rate.
In an assumed uni-form rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the Safety Limit.
The 15 percent APRM scram remains active until the mode switch is placed in the RUN position.
This switch occurs when the reactor pressure is greater than 850 psig. Amendment No. M, 48, 70, IH,192
Unit 2 PBAPS
'l 2.1.A BASES (Cont'd.)
The IRM system consists of 8 chambers, 4 in each of the reactor protection system logic channels. The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5-decades are covered by the IRM by means of a range switch and the 5-decades are broken down into 10 ranges, each being one-half of a decade in size. The IRM scram trip setting of 120 divisions is active in each range of the IRM.
For example, if the instrument were on range 1, the scram setting would be at 120 divisions for that range; likewise, if the instrument were on range 5, the scram would be 120 divisions on that range.
Thus, as the IRM is ranged up to accommodate the increase in' power level, the scram trip setting is also ranged up. The most significant sources of reactivity change during the power increase are due to control rod withdrawal.
For in-sequence control rod withdrawal the rate of change of power is slow enough due to the physical limitation of withdrawing control rods, that heat flux is in equilibrium with the neutron flux and an IRM scram would result in a reactor shutdown well before any Safety Limit is exceeded.
In order to assure that the IRM provided adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just subcritical and the IRM system is not yet on scale. This condition exists at quarter rod density. Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is bypassed.
The results of this analysis show that the reactor is scrammed and peak power limited to one percent of rated power, thus maintaining MCPR above the fuel cladding integrity safety limit. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and. continuous withdrawal of control rods in-sequence and provides backup protection for the APRM.
B.
APRM Rgd Block Trio Settina The APRM system provides a control rod block to avoid conditions which would result in an APRM scram trip if allowed to proceed.
The APRM rod block trip setting, like the APRM scram trip setting, is automatically varied with recirculation loop flow rate.
The flow variable APRM rod block trip setting provides margin to the APRM scram trip setting over the entire recirculation flow range. Amendment No. 23, 36, 48, 70, 192
1 Unit 2 PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicability:
Applicability:
Applies to the instrumentation and Applies to the surveillance associated devices which initiate of the instrumentation and associated devices which a reactor scram.
initiate reactor scram.
Objective Objective To assure the operability To specify the type and of the reactor protection frequency of surveillance to be applied to the system.
protection instrumentation.
Specification:
Specification:
A.
When there is fuel in the vessel A.
Instrumentation systems shall be functionally the setpoint, minimum number tested and calibrated of trip systems, and minimum as indicated'in Tables number of instrument channels 4.1.1 and 4.1.2 that must be operable for each position of the reactor respectively.
mode switch shall be as given in Table 3.1.1.
B.
DELETED B.
The designed system response times from the opening of the sensor contact up to and including the opening of the trip actuator contacts shall not exceed 50 milliseconds.
Otherwise, the affected trip system shall be placed in
'the tripped condition, or
.the action listed in Table 3.1.1.for the specific trip function shall be taken.
35 Amendment No. 65, 70, 71,-76, 192 4
m.
3-
PBAPS Unit 2 Table 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REOUIREMENT 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) per Trip Refuel Startup Run Design Items System (1)
(7) 1 1
Mode Switch In x
x x
1 Mode Switch A
Shutdown (4 Sections) 2 1
x x
2 Instrument A
Channels 3
3 IRM High Flux 5120/125 of Full x
x (5) 8 Instrument A
Scale Channels i
5 4
3 IRM Inoperative x
x (5) 8 Instrument A
Channels 5
2 APRM High Flux (0.66W+71%-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 Inoperative Channels k
7 2
APRM Downscale 22.5 Indicated (10) 6 Instrument A or B k"
on Scale Channels a
8 2
APRM High Flux 515% Power x
x 6 Instrument A
in Startup Channels 9
2 High Reactor
$1055 psig x(9) x x
4 Instrument A
u Pressure Channels 10 2
liigh Drywell
$2 psig x(8) x(8) x 4 Instrument A
ww Pressure Channels 11 2
Reactor Low 20 in. Indicated x
x x
4 Instrument _
A j y."
Water Level Level Channels l
Unit 2 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 cdjusting the APRM Flux trip setting. Delta W equals zero for two loop operation.
l Trip level setting is in percent of rated power (3293 MWt).
13.
See Section 2.1.A.1.
l l
l Amendment Ne
?,, 41, 62, 78,,
23, 134,.192 i
4 Unit 2 PBAPS 4.1 BASES (Cont'd)
Experience with passive type instruments in generating stations and substations indicates that the specified calibrations are adequate. For those devices which employ amplifiers, etc., drift specifications call for drift to be less than 0.4% month; i.e., in the period of a month a maximum drif t of 0.4% could occur, thus providing for adequate margin.
For the APRM systems, drift of electronic apparatus is not the only consideration in determining a calibration frequency.
Change in power distribution and loss of chamber sensitivity dictate a calibration every seven days. Calibration on this
- ncy assures plant operation at or below thermal limits.
A comparison of Tables 4.1.1 and 4.1.2 indicates that two instrument channels have not been included in the latter i
table.
These are: modo switch in shutdown and manual scram. All of the devices or sensors associated with these scram functions are simple on-off switches, and, hence, calibration during operation is not applicable.
B.
' sensitivity of LPRM detectors decreases with exposure to
_utron flux at a slow and approximately constant rate. This is compensated for in the APRM system by calibrating twice a week using heat balance data and by calibrating individual LPRM's every 6 weeks, using TIP traverse data. Amendment No. 23, 70, 192 j
l
PBAPS Unit 2 TABLE 3.2.C INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS Minimum No.
Instrument Trip Level Setting Number of Instrument Action of -Operable Channels Provided Instrument by Design Channels Per Trip System 4 (2)
APRM Upscale (Flow Biased)
(0.66W+59%-0.66AW) 6 Inst. Channels (10)
(Clamp at 108% max) 4 APRM Upscale (Startup s12%
6 Inst. Channels (10)
Mode) 4 APRM Downscale 22.5 indicated on scale 6 Inst. Channels (10) 1 (7)(11)
Rod Block Monitor (RTP 285%), S, sHTSP 2 Inst. Channels (1)
(Power Biased)
(65% sRTP <85%), S, sITSP (30% sRTP <65%), S, sLTSP i
1 (7)(11)
Rod Block Monitor 2DTSP 2 Inst. Channels (1)
'l Downscale g
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 EI 6
IRM Upscale s108 indicated on scale 8 Inst. Channels (10) k 2 (5)
SRM Detector not in (4) 4 Inst. Channels (1)
R Startup Position 2 (5)(6)
SRM Upscale s10' counts /sec.
4 Inst. Channels (1)
.~%
1 Scram Discharge s25 gallons 1 Inst. Channel (9)
- UPP Instrument Volume liigh Level g
- g..
L.E.M p.
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 (3293 MWt).
6W 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. 4W = 0 for two loop operation.
l 3.
IRM downscale is bypassed when it is on its lowest range.
4.
This function is bypassed when the count rate is t 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.
' Amendment No. fA, 48, 10 18, 123,1%,1b2
_ ~ _ _ _ ~ _
Unit 2 PBAPS
~
NOTES FOR TABLE 3.2.C (Cont.)
9.
If the number of operable channels is less than required by the minimum operable channels per trip function requirement, place the inoperable channel in the tripped condition within one hour. This note is applicable in the "Run" mode, the "Startup" mode and the " Refuel" mode if more than one control rod is withdrawn.
10.
For the Startup (for IRM rod block) and the Run (for APRM rod block) positions of the Reactor Mode Selector Switch and with the number of OPERABLE channels:
a.
One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour.
b.
Two or more less than required by the Minimum OPERABLE Channels per i
Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.
11.
The values for HTSP, ITSP, LTSP, and DTSP are specified in the CORE
,0PERATING LIMITS REPORT.
-74a-Amendment No. SS, 91, 15#,192
Unit 2 l
PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 3.5.I Averace Planar LHGR 4.5.I Averaae Planar LHGR During power operation, the APLHGR for The APLHGR for each type of each type of fuel as a function of axial fuel as a function of location and average planar exposure and average planar exposure reactor power / flow multipliers (provided and reactor power / flow in the CORE OPERATING LIMITS REPORT) multipliers (provided in the shall be within limits based on appli-CORE OPERATING LIMITS REPORT) cable APLHGR limit values which have been shall be checked daily determined by approved methodology for during reactor operation at the respective fuel and lattice types.
a25% rated thermal power.
When hand calculations are required, the APLHGR for each type of fuel as a func-tion of average planar exposure shall not exceed the limit for the most limiting lattice (excluding natural uranium) specified in the CORE OPERATING LIMITS REPORT during two recirculation loop operations.
If at any time during operation, it is determined by normal surveillance that the limiting value of APLHGR is being exceeded, action shall be initiated within one (1) hour to restore APLHGR to within prescribed limits.
If the APLHGR is not returned to within prescribed limits within five (5) hours, reactor power shall be decreased at a rate which would bring the reactor to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> unless APLHGR is returned to within limits during this period.
Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
3.5.J Local LHGR 4.5.J Local LHGR During power operation, the linear heat The LHGR as a function of core generation rate (LHGR) of any rod in any height shall be checked daily fuel assembly at any axial location shall during reactor operation at not exceed design LHGR.
a25% rated thermal power.
LHGR s LHGRd LHGRd - Design LHGR The values for Design LHGR for each fuel type are specified in the CORE OPERATING LIMITS REPORT.
-133a-Amendment No. Id, 48, 1d, /8, 86 Y05, Y23, Y M,
19E
Unit 2 PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 3.5.J Local lHGR (Cont'd) 4.5.K Minimum Critical Power Ratio (MCPR)
If at any time during operation it is determined by normal surveillance that limiting value for LHGR is being exceeded, action shall be initiated within one (1)
- 1. MCPR shall be checked daily hour to restore LHGR to within prescribed during reactor power operation limits.
If the LHGR is not returned to at 225% rated thermal power.
within prescribed limits within five (5)
- hours, reacter power shall be decreased
- 2. Except as provided in Spec-at a rate which would bring the reactor ification 3.5.K.3, the verifi-to the cold shutdown condition within cation of the applicability of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> unless LHGR is returned to 3.5.K.2.a Operating Limit MCPR within limits during this period.
Values shall be performed every Surveillance and corresponding action 120 operating days by scram shall continue until reactor operation time testing 19 or more control is within the prescribed limits, rods on a rotation basis and 3.5.K Minimum Critical Power Ratio (MCPR)
- a. The average scram time to the 20% insertion position
- 1. During power operation the MCPR for the shall be:
applicable incremental cycle core average x
ave s t B exposure and for each type of fuel shall be equal to or greater than the value given in Specification 3.5.K.2
- b. The average scram time to or 3.5.K.3, or MCPR(F), or the MCPR the 20% insertion position operating limit as determined by is determined as follows:
application of MCPR(P), whichever is greater. MCPR(F) and MCPR(P) are provided in the CORE OPERATING LIMITS REPORT. If at n
t ave -
E Ni t i any time during operation it is determined by normal surveillance that the limiting j.1 value for MCPR is being exceeded, action shall be initiated within one (1) hour to n
restore MCPR to within prescribed limits.
E Ni If the MCPR is not returned to within I"I prescribed limits within five (5) hours, fc where: n - number of surveillance o
r th r r
he c 1 shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> unless tests performed to date in the MCPR is returned to within limits during cycle.
this period.
Surveillance and corres-ponding action shall continue until reactor operation is within the prescribed limits.
-133b-Amendment No. 38, 48, 86, U#,
192
PBAPS -
3.5 BASES (Continued)
H. Enoineered Safeauards Comoartments Coolina and Ventilation One unit cooler in each pump compartment is capable of providing adequate ventilation flow and cooling.
Engineering analyses indicated that the tempera-ture rise in safeguards compartments without adequate ventilation flow or cooling is such that continued operation of the safeguards equipment or asso-ciated auxiliary equipment cannot be assured. Ventilation associated with the High Pressure Service Water Pumps is also associated with the Emergency Service Water pumps, and is specified in Specification 3.9.
I. Averace Planar LHGR This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR Part 50, Appendix K.
The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods' of a fuel assembly at any axial location and is only dependent, second-arily, on the rod-to-rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR. This LHGR times 1.02 is used in the heat-up code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factors. The Technical Specification APLHGR is the LHGR of the highest powered rod divided by its local peaking factor. The limiting value for APLHGR is shown in the applicable figure for each fuel type in the CORE OPERATING LIMITS REPORT, and must be adjusted for power and flow by application of MAPFAC(P) and MAPFAC(F).
MAPFAC(P) and MAPFAC(F) are provided in the CORE OPERATING LIMITS REPORT.
Only the most limiting APLHGR operating limits are shown in the figures for the multiple lattice fuel types.
Compliance with the lattice-specific APLHGR limits is ensured by using the process computer. When an alternate method to the pro-cess computer is required (i.e. hand calculations and/or alternate computer simulation), the most limiting lattice APLHGR limit for each fuel type shall be applied to every lattice of that fuel type.
The calculational procedure used to establish the APLHGR is based on a loss-of-coolant accident analysis.
The analysis was performed using General Electric (G.E.) calculational models which are consistent with the requirements of Appendix K to 10 CFR Part 50. A complete discussion of each code employed in the analysis is presented in Reference 4.
The plant specific results using the Reference 4 methodology are presented in Reference 8.
-140-Amendment No. 23, 36, 40, #S, /0, 86, I23, I H,192
4 Unit 2 PBAPS 3.5.K.
BASES (Cont'd)
The largest reduction in critical power ratio is then added to the fuel cladding integrity safety limit MCPR to establish the MCPR Operating Limit for each fuel type.
Analysis of the abnormal operational transients is presented in References 7,10 and 11.
Input data and operating conditions used in this analysis are shown in References 7,10 and 11 in the Supplemental Reload Licensing Analysis.
3.5.L.
Averaae Planar LHGR (APLHGR). Local LHGR and Minimum Critical Power Ratio (MCPR)
In the event that the calculated value of APLHGR, LHGR or MCPR exceeds its limiting value, a determination is made to ascertain the cause and initiate corrective actions to restore the value to within prescribed limits. The status of all indicated limiting fuel bundles is reviewed as well as input data associated with the limiting values such as power distribution, instru-mentation data (Traversing In-Core Probe - TIP, local Power Range Monitor -
LPRM, and reactor heat balance instrumentation), control rod configuration, etc., in order to determine whether the calculated values are valid.
In the event that the review indicates that the calculated value exceeding limits is valid, corrective action is immediately undertaken to restore the value to within prescribed limits.
Following corrective action, which may involve alterations to the control rod configuration and consequently changes to the core power distribution, revised instrumentation data, including changes to the relative neutron flux distribution, for up to 43 in-core locations is obtained and the power distribution, APLHGR, LHGR and MCPR calculated. Correc-tive action is initiated within one hour of an indicated value exceeding limits and verification that the indicated value is within prescribed limits is obtained within five hours of the initial indication, j
In the event that the calculated value of APLHGR, LHGR or MCPR exceeding its limiting value is not valid, i.e., due to an erroneous instrumentation indica-tion, etc., corrective action is initiated within one hour of an indicated value exceeding limits.
Verification that the indicated value is within pre-scribed limits is obtained within five hours of the initial indication. Such an invalid indication would not be a violation of the limiting condition for operation and therefore would not constitute a reportable occurrence.
140b Amendment No. 23, 36, #S, 70, 86,
)
773, Is#, I37,192 l
i
Unit 2 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 Abnormal 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 limits (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.
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.
-140c-Amendment No. 27, 36, 38, 48, 70, 86, I23, Is7, 192
Unit 2 PBAPS 4.5.K Minimum Critical Power Ratio (MCPR) - Surveillance Reouirement At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed-and the moderator void content will be very small.
For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% thermal power level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary.
The daily requirement for calculating MCPR above 25% rated thermal power is suf-ficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal limit.
4.5.L MCPR limits for Core Flows Other Than Rated A flow dependent MCPR limit, MCPR(F), is necessary to assure that the safety limit MCPR is not violated during recirculation flow increase events. The design basis flow increase event is a slow-power increase event which is not terminated by scram, but which stabilizes at a new core power corresponding to the maximum possible core flow.
Flow runout events are analyzed along a constant xenon flow control line assuming a quasi steady state heat balance.
The flow dependent MCPR limit, MCPR(F), is provided in the CORE OPERATING LIMITS REPORT.
The MCPR(F) is independent of the rated flow limit provided in Specification 3.5.K 2 and 3.5.K.3.
To verify applicability of this curve to PBAPS, recirculation flow runout events were analyzed with a PBAPS specific model at a typical mid cycle exposure condition. These flow runout events were simulated along the Maximum Extended Load Line Limit rod line to the maximum core flow runout value of 105%. The results of the analyses indicated that application of the MCPR(F) curve will preclude a violation of the MCPR safety limit in the event of a recirculation flow runout. The MCPR(F) curve is cycle independent.
-141a-Amendment No. 23, I M, 192
- + +,
Unh 2 4
pggp3 "LEFT BLANK INTENTIONALLY =
-141b-kendment No, yg, gg y,,, 192
roard unit 4 6.9.1 Routine Reoorts (cont'd)
- c. Annual Safety / Relief Valve Reoort Describe all challenges to the primary coolant system safety and relief valves.
Challenges are defined as the automatic opening of the primary coolant safety or relief valves in response to high reactor pressure.
- d. Monthly Operatina Reoort Routine reports of operating statistics and shutdown experience and a narrative summary of the operating experience shall be submitted on a monthly basis.
Each report shall be submitted no later than the 15th of the month following the calendar month covered by the report.
e.
Core Ooeratino limits Reoort (1)
Core operating limits shall be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each Operating Cycle, or prior to any remaining portion of an Operating Cycle, for the following:
a.
The APLHGR for Specification 3.5.1, b.
The MCPR for Specification 3.5.K, c.
The core flow and power adjustment factors for Specification 3.5.K and 3.5.I, d.
The LHGR for Specification 3.5.J, e.
The upscale power biased Rod Block Monitor setpoints and corresponding power levels.
(2) The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those -
described in the following documents as amended and approved:
a.
NEDE-24011-P-A, " General Electric Standard Application for Reactor fuel" (latest approved version) b.
" 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 c.
Philadelphia Electric Company Methodologies as l
described in:
(1)
PECo-FMS-0001-A, " Steady-State Thermal Hydraulic Analysis of Peach Bottom Units 2 and 3 using the FIBWR Computer Code"
-256-Amendment No. Z22, 110, I H, 160, 192
-