ML20247J049
| ML20247J049 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  (DPR-56-A-150) |
| Issue date: | 09/01/1989 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20247J054 | List: |
| References | |
| NUDOCS 8909200183 | |
| Download: ML20247J049 (33) | |
Text
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NUCLEAR REGULATORY COMMISSION n
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WASHINGTON, D. C. 20555
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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. 150 License No. DPR-56 1.
The Nuclear Regulatory Comission (the Comission) has found that:
The app (lication for amendment by Philadelphia Electric Company, A.
et al.
the licensee) dated July 7, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (theAct),ChapterI; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and 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 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 amendment to read as follows:
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.(2) Technical Specifications The Technical Specifications. contained in Appendices A and B,-as revised through Amendment No.150, are hereby incorporated-in the
' Technical Specifications..
3.
2This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
~
I
.Mohan C.'Thadani for-Walter R. Butler, Director Project Directorate I-2 Division of Reactor' Projects 1/11
Attachment:
Changes to the Technical Specifications Date of Issuance:
. September 1, 1989 i
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(2) Technical Specifications The Technical Specifications _ contained in Appendices.A and B, as revised through Amendment No.150, are hereby incorporated in the Technical Specifications.
3; This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 1
k' alter R. Butler, Director Project Directorate I-?
Division of Reactor Projects I/II
Attachment:
Changes to the Technical-Specific 6tions Date of' Issuance:
-September 1, 1989
ATTACHMENT TO LICENSE AMENDMENT NO. 150 FACILITY OPERATING LICENSE NO. DPR-56 1
DOCKET NO. 50-278 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Pages iv iv 1
1 9
9 9a 9a 10 10 11 11 Ila 11a 13 13 15 15 16 16 17 17 18 18 33 33 37 37 40 40 73 73 74 74 133a'-
133a 133b 133b 133c 133c 133d 133d 133e 133e 140 140 140b 140b 140c Id0c 140d 142 142 142a 142a 142e 142f
b,3
- f UNIT 3
.c PBAPS LIST OF FIGURES lY L
Figure Title Page 1.1-1 APRM Flow Bias Scram Relationship To Normal Operating 16 Conditions 4.1.1 Instrument Test Interval Determination Curves 55
- 4.2.2 Probability of System Unavailability vs. Test Interval 98 3.3.1 SRM Count Rate vs. Signal-to-Noise Ratio 103a 3.4.1 DELETED 122 3.4.2 DELETED 123 3.5.K.1 MCPR Operating Limit vs. Tau, BP/P8X8R, LTA, GE8X8EB Fuel, Standard Operatitg Conditions 142 3.5.K.2 MCPR Operating Limit vs. Tau, BP/P8X8R, LTA, GE8X8EB Fuel, Increased Core Flow 142a
- 3. 5.1. A : DELETED 3.5.1.B DELETED
- 3. 5.1. C DELETED
'3.5.1.D DELETED
-3.5.1.E-Kf Factor vs. Core Flow 142d
- 3. 5.1. F MAPLHGR vs. Planar Average Exposure, Unit 3, GE8X8EB Fuel (Type BD 319A) 142e 3.5.1.G MAPLHGR-vs. Planar Average Exposure, Unit 3, GE8X8EB Fuel (Type BD321A) 142f
- 3. 5.' 1. H MAPLHGR vs. Planar Average Exposure, Unit 3 142g P8X8R Fuel (P8DRB284H)
- 3. 5.1. I MAPLHGR vs. Planar Average Exposure, Unit 3 142h P8X8R and BP8X8R Fuel (P8DRB299 and BP80RB299) 3.5.1.J MAPLHGR vs. Planar Average Exposure, Unit 3 142i BP8XBR Fuel (BP8DRB299H)
L 3.5.1.K MAPLHGR vs. Planar Average Exposure, Unit 3 142j P8X8Q LTA (P80QB326) 3.6.1 Minimum Temperature for Pressure Tests such as 164 required by Section XI
~
3.6.2 Minimum Temperature for Mechanical Heatup or 164a Cooldown following Nuclear Shutdown 3.6.3 Minimum Temperature for Core Operation (Criticality) 164b 3.6.4 Transition Temperature Shift vs. Fluence 164c 3.6.5 Thermal Power Limits of Specifications 164d 3.6.F.3, 3.6.F.4, 3.6.F.5, 3.6.F.6 and 3.6.F.7 3.8.1 Site Boundary and Effluent Release Points 216e 6.2-1 Management Organization Chart 244 l
6.2-2 Organization for Conduct of, Plant Operation 245 "lV~
Amendment No. 14 41, 45, 46, 62, 79, 92, 194, 197,,114, 126, 142, 150
m.
PBAPS Unit 3 of.;
L, 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, i
Normal control rod movement with the control drive hydraulic system is not defined as a core alteration.
Normal movement of l'
in-core instrumentation and.the traversing in-core probe is not defined as a core alteration.
l Average Planar Linear Heat Generation Rate (APLHGR) - The APLEGR shall be applicable to a specific planar height and.is equal to the sum of the heat generation 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 arrangement 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 P.
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.
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 NEDO-10958).
Dose LVuivalent I-131 - That concentration of I-131 (Ci/gm) wnich alone would produce the same thyroid dose as the quantity and-isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.
Amendment No. 105, 125, 150.
l Unit 3 4
IBAPS I
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1-FUEL CLADDING INTEGRITY Applicability:
Applicability:
j 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.
Objectives:
Obiectives:
The objective of the Safety The objective of the Limiting Safety Limits is to establish limits System Settings is to define the
{
which assure the integrity of level of the' process variables at the fuel cladding.
which automatic protective action is initiated to prevent the fuel cladding integrity Safety Limits from being
)
exceeded.
Specification:
Specification:
A. Reactor Pressure t 800 psia The limiting safety system settings and Core Flow t 10% of Rated shall be as specified below:
A. Neutron Flux Scram The existence of a minimum
(Run Mode) less than 1.04 for two recirculation loop operation, When the Mode Switch is in the or 1.0b for single loop i
RUN position, the APRM fluF operation, shall constitute scrad trip setting shall be:
violation of the fuel cladding integrity safety limit.
S < 0.58W + 62% - 0.58 AW l
To ensure that this safety where:
limit is not exceeded, neutron flux shall not be above the S = Setting in percent of rated scram setting established in thermal power (3293 MWt) specification 2.1.A for longer than 1.15 seconds as indicated W = Loop recirculating by the process computer. When flow rate in percent the process computer is out of of design. W is 100 for L
service this safety limit shall core flow of 102.5 be assumed to be exceeded if million Ib/hr or greater.
the neutron flux exceeds its scram setting and a control rod scram does not occur.
{
Amendment No. If, (I, 77. 79 150
V Unit 3
.,.g FBAPS SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING
- 1. l' FUEL CLADDING _ INTEGRITY 2.1 FUEL CLADDING INTEGRITY a W = Difference between two
' loop and single loop effective recirculation drive flow rate at the same core flow.
Durin single loop operation,gthe reduction in trip setting
(-0.584 W) is accomplished l
by correcting the flow 1
input of the flow biased scram to preserve the original (two loop) relationship between APRM i
scram setpoint and recirculation drive flow or by. adjusting the APRM flux trip setting.
A W = 0 for two loop operation.
i Amendment No. 37,150
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Ja-PBAPS
- .L UNIT 3
_; g 4
SAFETY' LIMIT LIMITING SAFETY SYSTEM SETTING f.
2.1.A (Cont'd)
In the event of operation with a maximum fraction of limiting power density (MFLPD) greater tnan the fraction of rated power (FRP), the setting shall be modified as follows.
S < ( 0. 5 8W + 6 21 - 0. 5 8 A W) (FRP)
- where, FRP = fraction of rated thermal power (3293 MWt)
MFLPD d maximum fraction of limiting power density where the limiting power density is 13.4 KW/ft for BP/P8X8R and LTA 0
fuel and 14.4 KW/ft for GEBX8EB fuel.
The rntio of FRP to MFLPD shall be set equal to 1.0 unless the actual opersting value is less than the design value of 1.0, in which case the actual operating value will be used.
- 2. ACRM--When the reactor mode switch is in the STARTUP position, the APRM ceram shall be set at less than or equal to 15 percent of rated power.
Amendment No. If, 33, $1, 62, 77, 79, 107. 150
PBAPS Unit 3 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING B.
Core Thermal Power Limit (Reactor Pressure 1 800 psia)
B. APRM Rod Block Trip Setting j
When the reactor pressure is
< 800 psia or core flow is SRB 1 (0.58 W + 50% - 0.584 W) l Tess than 10"$ of rated, the where:
core thermal power shall not exceed 25% of rated thermal power.
SRB = Rod block setting in percent of rated thermal power (3293 MWt)
W
= Loop recirculation flow rate in percent of design.
W is 100 for core flow of 102.5 million lb/hr or greater.
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 (-0.58 4 W) 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.
A W = 0 for two loop operation.
In the event of operation with maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP),
the setting shall be modified as follows.
Amendrnent If, 33, #I, 62, 77 150
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i Unit 3 d
PBAPS SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING B.
Core Thermal Power Limit B. APRM Rod Block Trip Setting (Reactor Pressure < 800 psia)
SRB < (0.58 W + 50% - 0.586 W)
FRP = fraction of rated thermal power (3293 MWt).
MFLPD n maximum fraction of limiting power density where the limiting i
power density is 13.4 KW/ft for BP/P8X8R' and LTA fuel and 14.4 KW/ft for GE8X3CB fuel.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
C.
Whenever the reactor is in the C. Scram and isolation--> 538 in. above shutdown condition with reactor low water vessel zero irradiated fuel in the reactor level (0" on level vessel, the water level shall instruments) not be less than minus 160 inches indicated level (378 inches above vessel zero).
-lla-Amendment No. 77, 79, 115, 150
l l
o PBAPS Unit 3
1.1 BASES
FUEL CLADDING INTEGRITY A.
Fuel Cladding Integrity Limit at Reactor Pressure > 800 psia and Core Flow > 10% of Rated The fuel cladding integaity safety limit is set such that no fuel-damage is calculated to occur if the limit is not
)
violated.
Since the parameters which result in fuel damage are i
not directly observable during reactor operation'the thermal I
i hydraulic. conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur.
Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling I
transition is calculated to occur has been adopted as a convenient limit.
However, the uncertainties in monitoring the core operatino state and in the procedure used to calculate the critical power result in an uncertainty in the value of the critical power.
Therefore, the fuel cladding integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.
The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis described in references 2 and 3 for two recirculation loop operation.
The Safety Limit MCPR is increased by 0.01 for single-loop operation as discussed in reference 4.
Amendment No. 14, 79, 150.
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PBAPS Unit 3
'4.
1.1.C BASES (Cont'd.)
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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 Unit 3 has a sequence i
annunciation program which will indicate the sequence in which 1
events such as scram, APRM trip initiation, pressure scram l
initiation, etc. occur.
This program also indicates when the scram setpoint is cleared.
This will provide information on how l
long a scram condition exists and thus provide some measure of the energy added during a transient.
Thus, computer information 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 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 (NEDO-10958-A).
2.
Process Computer Performance Evaluation Accuracy, General Electric Company BWR Systems Department, June 1974 (NEDO-20340).
3.
" General Electric Standard Application for Reactor Fuel",
I NEDE-240ll-P-A (as amended).
4.
" Peach Bottom Atomic Power Station Units 2 and 3 Single-Loop L
Operation", NEDO-24229-1, May 1980.
Amendment No. 33, fl, 62, IZE, 150 1
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PBAPS Unit 3 l
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APRM FLOW BIAS SCRAM RELATIONSHIP TO NORMAL OPERATING CONDITIONS FIGURE I l-1 Amendment No. H, 150,.
7 PBAPS Unit 3
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2.1 BASES
FUEL CLADDING INTEGRITY L.-
l 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 accordance with the operating map.given in Figure 3.7.1 of the FSAR.
In addition, 3293 MWt is the licensed maximum power level of each Peach Bottom Atomic Power-Station Unit, and this represents the maximum steady state power which shall not knowingly be exceeded.
Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity j
coefficient, control rod scram worth, scram delay time, peaking factors, and axial power shapes.
These factors are selected conservatively with respect to their effect on the applicable transient results as determined by the current analysis model.
Conservatism incorporated into the transient analyses is documented in NEDE-240ll-P-A (as amended).
Amendment No. 33, 79, 150 }
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- 2. l*.fASES-(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 conservative values of controlling parameters and initiating transients at the design power level produces more pessimistic answers than would~ result.by using expected values of control parameters and analyzing at higher power levels.
e Steady state operation'without forced recirculation will not be permitted.
The analysis to support operation at various power and flow relationships has considered operation with either.one or two. recirculating pumps.
In summary:
.i.
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.
11.
The licensed maximum power level is 3293 MWt.
i lii.
Analyses of transients employ adequately conservative values.of the controlling 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 (3293 MWt).
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 3
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 tiansients analyzed violate the fuel Safety Limit and there is a substantial margin from fuel damage.
Therefore, the use of flow referenced scram trip provides even additional margin.
Amendment No. 33, H, n, 62,150.
.___.m._.__.___._.m__
PBAPS Unit 3 1
2.2 BASES REACTOR COOLANT SYSTEM INTEGRITY The pressure relief system for each unit at the Peach Bottom Atomic Power Station has been sized to meet two design bases.
First, the total capacity of the safety / relief valves and safety valves has been established to meet the overpressure protection criteria of the ASME Code.
Second, the distribution of this required capacity between safety valves and relief valves has been set to meet design basis 4.4.4.1 of subsection 4.4 of the FSSR which states that the nuclear system safety / relief valves FD l'1 prevent opening of the safety valves during normal plant it elations and load rejections.
(
The details of the analysis which show compliance with the ASME Code requirements are presented in subsection 4.4 of the PSAR and the Reactor Vessel Overpressure Protection Summary Technical Report submitted in Appendix K.
Eleven safety / relief' valves and two safety valves have been installed on Peach Bottom Units 2 and 3.
The analysis of the I
worst overpressure transient is provided in the Supplemental Reload Licensing Submittal and demonstrates margin to the code l
allowable overpressure limit of 1375 psig.
The safety / relief valve settings satisfy the Code requirements that the lowest valve setpoint be at or below the vessel design pressure of 1250 psig.
These settings are also sufficiently above the normal operating pressure range to prevent unnecessary cycling caused by minor transients.
~
The design pressure of the shutdown cooling piping of the Residual Heat Removal System is not exceeded with the reactor vessel steam dome less than 75 psig.
Amendment No. 33, 41, 42, 63, 79, 150
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Unit 3 l ?.j.
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 APPM 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.
This equation will be used in the event of operation with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP), where:
FRP = fraction of rated thermal power (3293 MWt).
MFLPD = maximum fraction of limiting power density where the limiting power density is 13.4 KW/ft for BP/P8X8R and LTA fuel and 14.4 KW/ft for GE8X8EB fuel.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
W=
Loop Recirculation flow in percent of design.
W is 100 for core flow of 102.5 million lb/hr or greater.
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 (-0.58 delta W) is I
accomplished by correcting the flow input of the flow biased High Flux trip setting to preserve the original (two loop) relationship betws2n 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 (3293 MWt).
13.
See Section 2.1.A.l.
I i
Amendment No. 33, 41, 62, 77, 79, 106, 132, 150 ff
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34 Unit 3
)
PBAPS NOTES FOR TABLE 3.2.C I
- 1. -
For the.startup and;run positions-of the Reactor Mode Selector
$ witch, 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 dail condition lasts. longer.than seven days, y thereafter; if this the system shall be tripped.
If the first column cannot be met for both trip systems,.the systems shall be tripped.
2.-
This equation will.be used in the event of operation with a i
maximum fraction.of limiting power density-(MFLPD) greater than the fraction of rated power (FRP) where:
FRP =ffraction of rated thermal power (3293 MWt)
MFLPD = maximum fraction of limiting power density where the limiting power density is'13.4 KW/ft for BP/P8X8R and LTA fuel and 14.4 KW/ft for GE8X8EB fuel.
The ratio of-FRP to MFLPD shall be set equal to 1.0 unless the actual. operating value is less than the design value of 1.0, in which case the actual operating value will be used.
W = Loop Recirculation flow in percent of design.
W is 100 for core-flow.of 102.5 million Ib/hr or greater.
Trip level setting is in percent of rated power (3293 MWt).
4 W 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, or by adjusting the rod block; setting.
A W =-0 for two loop operation.
3.
IRM downscale'is' bypassed when it is on its-lowest range.
.4.
This function is bypassed when the count rate is > 100 cys.
L p
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 < 30%.
I 8.
This function is bypassed when the mode switch is placed in Run.
Amendment No. 33, 41, 63, 77, 79,
,74, 150 L
m
PBAPS Unit 3
. LIMITING COMDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.I Average Planar LHGR 4.5.I Average Planar LHGR During power operation, the APLHGR The APLGHR for each type of fuel for each type of fuel as a function as a function of average planar of axial location and average planar exposure shall be checked daily exposure shall be within limits during reactor operation at based on applicable APLHGR limit values which have been approved for
> 25% rated thermal power.
the respective fuel and lattice types.
When hand calculations are required, the APLHGR for each type of fuel as a function of average planar exposure shall not exceed the limit for the most limiting lattice (excluding natural uranium) shown in the applicable figures for BP/P8X8R, LTA and GE8X8EB fuel types during two recirculation loop operations. During single loop operation, the APLHGR for each fuel type shall not exceed the above values multiplied by the following reduction factors: 0.81 for BP/ PBX 8R and LTA fuel and 0.73 for GE8X8EB fuel.
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 ALPHGR 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 LEGR 4.5.J Local LEGR During power operation, the linear The LEGR as a function of core heat generation rate (LEGR) of any height shall be checked daily rod in any fuel assembly at any axial location shall not exceed during reactor operation at design LHGR.
> 25% rated thermal power.
l LHGR < LHGRd I
l LHGRd = Design LHGR 13.4 KW/ft for BP/P8X8R and LTA fuel 14.4 KW/ft for GE8X8EB fuel Amendment 33, 41, 62, 77. 79, 92, - 133a -
150
PBAPS Unit 3 LfMITING CONDITIONS FOR OPERATION SURVEILLANCE RLOUIREMENTS 3.5.J Local LHGR (Cont'd)
If at any time during operation it is determined by normal surveillance that limiting value for LEGR is be-ing exceeded, action shall be initi-ated within one (1) hour to restore LHGR to within prescribed. limits.
If the LHGR 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 LHGR is returned to within limits during this period.
Surveil-lance and corresponding action shall continue until reactor operatic ( is within the prescribed limits.
l 3.5.K Minimum Critical Power 4.5.K Minimum Critical Power Ratio (MCFR)
Ratio (FEPR)
- 1. During power operation the MCPR
- 1. MCPR shall be checked daily for the applicable incremental during reactor power operation cycle core average exposure and at >25% rated thermal power.
for each type of fuel shall be
- 2. Except as provided in Specifi-equal to or_ greater than the value cation 3.5.K.3, the verifica-given in Specification 3.5.K.2 or tion of the applicability of 1.5.K.3 times Kf, where Kf is as 3.5.K.2.a Operating Limit MCPR shown in Figure 3.5.1.E. If at Values shall be performed every any time during operation it 120 operating days by scram time is determined by normal survell-testing 19 or more control rods lance that the limiting on a rotation basis and per-(
value for MCPR is being exceeded, forming the following:
action shall be initiated within one (1) hour to restore MCPR to
- a. The average scram time to within prescribed limits.
If the 20% insertion position the MCPR is not returned shall be:
to within prescribed limits 17 ave < 1EB within five (5) hours, reactor power shall be decreased at a
- b. The average scram time to rate which would bring the the 20% insertion position reactor to the cold shutdown is determined as follows:
condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> n
unless MCPR is returned to Y ave = 2~N11 7
within limits during this period.
i=1 ___
Surveillance and corresponding n
action shall continue until re-T-~ Ni actor operation is within the i=1 prescribed limits.
where: n = number of surveillance tests performed to date in the cycle.
l Amendment No. 150
- 133b -
l t
PBAPS Unit 3 LIMITING COND1TXONS FOR CPERATION SURVEILLANCE REQUIREMENTS o
3.5.K Minimum Critical Power 4.5.K Minimum Critical Power Ratio (MCPR) (Cont'd)
Ratio (MCPR) (Cont'd)
- 2. Except as specified in 3.5.K.3, Ni = number of active control the Operating Limit MCPR Values rods measured in the ith are as follows:
surveillance test.
I a.
If requirement 4.5.K.2.a is met:
The Operating Limit MCPR values y
are as given in Table 3.5.K.2 Li = average scram time to the 20% insertion position of all rods measured in b.
If requirement 4.5.K.2.a is not the ith surveillance test.
met:
The Operating Limit MCPR c.
The adjusted analysis mean values as a function of T scram time (TB) is calculated are as given in Figures as follows:
3.5.K.1 and 3.5.K.2
[
\\ 1/2 N1 L B = p + 1. 6 5 l n
(p-
=!
Where:
Where:
lI = lIave i B p
mean of the distribution
=
0.90 - 17B for average scram insert time to the 20% position-0.694 sec l
- 3. The Operating Limit MCPR values N1 = total number of active shall be as given in Table 3.5.K.3 control rods measured in if the Surveillance Requirement specification 4.3.C.1 of Section 4.5.K.2 to scram time test control rods is not (T'= standard deviation of the performed.
distribution for average scram insertion time to the 20% position = 0.016 l
Amendment No. 79, 150
-133c-
d 4
PBAPS Unit 3
~
Table 3.5.K.2 OPERATING LIMIT MCPR VALUES FOR VARIOUS CORE EXPOSURES
- j MCPR Operating Limit Fuel Type For Incremental Cycle Core Average Exposure **
{
BOC to 2000 MWD /t 2000 MWD /t before EOC Before EOC To BOC Standard Operating Conditions BP/P8X8R 1.21 1.26 LTA 1.21 1.26 l
GE8XSEB 1.21 1.26 Increased Core Flow BP/P8X8R 1.21 1.27 LTA 1.21 1.27 GE8X8EB 1.21 1.27 If requirement 4.5.K.2.a is met.
These values shall be increased by 0.01 for single loop operation.
Amendment No. 42, 62, 77, 79, 85,
-133d-92, 107, 114,150
- - - - - - - - - - - - - - - - - - - - - - - - - - ~ ~ - - - ~ - ~ - ~
~ - -
e
~
PBAPS Unit 3 Table 3.5.K.3 OPERATING LIMIT MCPR VALUES FOR VARIOUS CORE EXPOSURES
- MCPR Operating Limit Fuel Type For Incremental Cycle Core Average Exposure **
BOC to 2000 MWD /t 2000 MWD /t before EOC Before.EOC To EOC t
Standard Operating Conditions BP/P8X8R 1.26 1.30 LTA 1.26 1.30 GE8X8EB 1.26 1.30 Increased Core Flow BP/P8X8R 1.26 1.31 LTA 1.26 1.31 GE8X8EB 1.26 1.31 If surveillance requirement of section 4.5.K.2 is not performed.
These values shall be increased by 0.01 for single loop operation.
Amendment No. 79, 85, 93, 197, 114'
-133e-150
a
,1 J,
sh 7
PBAPS-Unit 3 3.5' BASES (Continued)
H.
Engineering Safeguards Compartments Cooling and Ventilation One unit cooler in each pump compartment is capable of providing adequate ventilation flow and cooling.
Engineering analyses indicated that the l
temperature rise in safeguards compartments without adequate ventilation flow or cooling is such that continued operation of_the safeguards equipment or associated auxiliary equipment cannot be assured.
Ventilation associated witn the High Pressure Service Water Pumps is also associated with the Emergency-Service Water pumps, and is specified in Specification 3.9.
1
~I.
Average 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 tenerature (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, secondarily, on the rod-to-rod power distribution within an assembly.
The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR.
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 factcr The limiting value for APLHGR is shown in the applicable figures for l each fue? type.
(
1 Only the most ?imiting and least limiting APLHGR operating limits are shown in i
the figures for the multiple lattice fuel types.
Compliance with the lattice-specific, approved APLHGR limits is ensured by using the process computer; When an alternate method to the process 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.
l The calculational procedure used to establish the APLHGR for each fuel type is
-based on a loss-of-coolant accident analysis.
The analysis was performed l
using General Electric (G.E.) calculational models which are consistent with l-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.
Input and model changes in the Peach Bottom loss-of-coolant analysis which are different from the previous analyses performed with Reference 4 are described in detail in Reference 8.
These changes to the analysis include:
(1) consideration of L
the counter current flow limiting (CCFL) effect, (2) corrected code inputs, and (3) the effect of drilling alternate flow paths in the bundle lower tie plate.
I Amendment No. 33, 41, 42, 62, 39,
-140-150 j
i-J
p
^l T.it C-p *f I.
PBAPS Unit 3 sc 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 l
{
' Reference 7.
Input data and operating conditions used in this
-]
-analysis are shown in Reference 7 and in the Supplemental Reload Licensing Analysis.
3.5.L.
' Average Planar LHGR (APLHGR), Local LHGR'and Minimum l
f Critical Power Ratio (MCPR)
In the event that the calculated value of APLHGR, LHGR or MCPR s
exceeds its limiting value, a determination is made to ascertain i
the cause and initiate corrective action 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, instrumentation data (TraversingLIn-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 11s obtained and the power distribution, APLEGR, LHGR and MCPR calculated.
Corrective 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.
4 In the event that the calculated value of APLEGR, LHGR.or MCPR exceeding its limiting value.is not valid, i.e.,
due to an erroneous instrumentation indication, etc., corrective action is initiated within one hour of an indicated value exceeding limits.
Verification that the indicated value is within prescribed 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.
1 Amendment No. 33, 41, 42, 62, ;:9
-140b-
-150 l
A
,4
'.-[.
PBAPS Unit 3 1
i
)
l 3.5.L.
BASES (Cont'd)
Operating experience has demonstrated that a calculated value of i
APLHGR, LEGR or MCPR exceeding its limiting value predominately occurs due to this latter cause.
This experience coupled with the extremely unlikely occurrence of concurrent operation exceeding APLEGR, 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 APLEGR, 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.
I 3.5.M.
References
" Fuel Densification Ef$ects on General Electric Boiling 1.
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.
General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE 20566 (Draft), August 1974.
5.
General Electric Refill Reflood Cal'culation (Supplement to SAFE Code Description) transmitted to the USAEC by letter, G. L. Gyorey to Victor Stello, Jr., dated December 20, 1974.
6.
DELETED.
7.
" General Electric Standard Application for Reactor Fuel",
NEDO-240ll-P-A (as amended).
8.
Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2, NEDO-24081, December 1977, and for
(
Unit 3, MEDO-24082, December 1977.
9.
Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2, Supplement 1, NEDE-24081-P, November 1986, and for Unit 3, NEDE-24082-P, December 1987.
l Amendment No. ?3, 34, 41, 43, 63,
- 79. 150
_-___-___A
w
. 4
,. +
c.
r PEACH BOTTOM UNIT 3 l
FIGURE 3.5.K.1 MCPR OPERATING UMIT vs 7
- FUEL TYPES: BP/P8X8R,LTA,GE8X8EB (STANDARD OPERATING CONDITIONS) 1.40 1.40 1.38-
""'~~
" : " " " ~ ~ 4 " " " " ": ' " ; " ":" "- -1.38 1
1.36- ~ - - " 4 " ":" " ": " " " " " + ~ '" " "4" " + " - -1.36 4
" "-[ " ","
-l
" [" " " "
"[~ ~ -
1.34-
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-1.32
.1. 3 0 -
": ""-" ;""";"""'to goc "";"-
-1.30 gpoRE Cr h
1.28 - """""4000 N 1""""""""""- -1.28
.C
' - ' ".I
" + " " - -
E.O
-1.26 1.26-4 e
o
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-1.24 10 o
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.--l i/
l
- -1.18 1.18 -
1.16,
1.16 i
i i
i i
i i
0 0.1 0.20.30.4 0.5 0.6 0.7 0.8 0.9 1
OPTB T
OPT A Amendment No. 41, 29, SE, 92
-- M2 --
114, 150
h ?.
PEACH BOTTOM UNfT 3
?
FIGURE 3.5.K2 1
MCPR OPERATING UMIT.vs T FUEL TYPES: 8P/P8X8R.LTA.GE8X8EB (INCREASED. CORE FLOW) u l
1.40 1.40 1.38-
". " " i " " i " " -
- - i "t " - 2 " " : " - -1.38 1.36- ~ - ' ;' ~~'
"- -1.36 1
v 1.34-
- j '
~,';~-
-1.34 1.32-
- -1.32
"" " "[ ~ " " gogE EOC W GOC 1.30- " " ":-
-1.30
/TB
-.2000 p o e
ct i
-. -1.28 o
1.28-2 toc 1.26 1
1.26-oB.t
- j gj 1.24 -
-1.24 2
c to 1.22-mo
-1.22 1.20- " " 9. " " i " ~ ; " "
-1 20
...........{......;.........{.....!......~..........:....
1.13 _
_ i.is 1.16 i
1.16 i
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1
OPTB T
OPT A
- Amendment NO. 41, 79, 86' 92'
-142a-
. 197, 114, 150 r -
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