ML20234B713

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Amend 123 to License DPR-44,revising Tech Specs to Incorporate Operating Limits for All Fuel Types for Cycle 8 Operation & Changes in Slope of Flow Biased APRM Scram & Rod Block Setpoints
ML20234B713
Person / Time
Site: Peach Bottom Constellation icon.png
Issue date: 09/11/1987
From: Butler W
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20234B716 List:
References
NUDOCS 8709210052
Download: ML20234B713 (38)
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{{#Wiki_filter:- _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ ,[ga arcy,?4 UNITED STATES i NUCLEAR REGULATORY COMMISSION j f WASH 6NGTON, D. C. 20555 Septaser 11, 1987 l 'g...../ PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY I ~ ATLANTlC CITY ELECTRIC COMPANY l l DOCKET NO. 50-277 PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE I Amendment No. 123 License No. DPR-44 1. The Nuclear Regulatory Commission (the Comission) has found that: A. The application for amendment by Philadelphia Electric Company, et al. (the licensee) dated January 9,1987, as supplemented by letters i dated February 6, March 24, and May 13, 1987, 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 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 comen 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 i 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 Nn. DPR-44 is hereby i amended to read as follows: %[2hO P l

8 P (2) Technical S,ecifications J The Technical Specifications contained in Appendices A and B, as revised through Amendment No.123, are hereby incorporated in the license. PEC0 shall operate the facility in accordance with the Technical Specifications. l 3. This license amendment is effective prior to startup of Unit 2 in Cycle 8. FOR THE NUCLEAR REGULATORY COMMISSION /s/ 1 Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II I

Attachment:

Changes to the Technical Specifications Date of Issuance: September 11, 1987 i C L0cect.h eac/ ccn cLCM o cA # Y 1/2.7/S'7 0" fy e PDng/ PCI-?/P OGC PDI-2/D M0 Uf f; RClark: WButler }/Q/87 04 /29/87 / /87 / /87

j -?- (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.123, are hereby incorporated in the license. DECO shall operate the facility in accordance with the i Technical Specifications. 3. Tnis license amendment is effective prior to startup of Unit 2 in Cycle 8. FOR THE NUCLEAR REGULATORY COMMISSION j i 1 Walter R. Butler, Director C Project f)irectorate I-2 Division of Reactor Projects I/II i

Attachment:

l l Changes to the Technical l Specifications l Date of issuance: Seotember 11, 1987 a l i i l l l L__._________.______

i ATTACHMENT TO LICENSE AMENOMENT NO.123 FACILITY OPERATING LICENSE NO. DPR-44 DOCKET NO. 50-277 Replace the following pages of the Appendix A Technical Specifications l with the enclosed pages. The revised areas are indicated by marginal lines. I k Remove Insert ) iv iv iva iva { 1 1 9 9 ) i 9a 9a 10 10 l 11 11 t 'lla 11e l 13 13 I I 15 15 16 16 l 17 17 i 18 18 l 33 33 l 37 37 ) 40 40 73 73 74 74 133a 133a 133c 133c 133d 133d 133e 133e 140 140 140b 140b 140c 140c 140d 142 142a-1 142a-2 142a-3 142a-4 142a-5 142L 142m 142n I O

PBAPS Unit 2 LIST OF FIGURES j Figure Title Page 1.1-1 APRM Plow Bias Scram Relationship To 16 Normal Operating Conditions 4.1.1 Instrument Test Interval Determination 55 Curves 4.2.2 Probability of System Unavailability 98 vs. Test Interval 3.4.1 Deleted 122 3.4.2 Deleted 123 3.5.K.1-1 MCPR Operating Limit vs. T, BP/P8X8R Fuel, 142 l Standard Operating Conditions 3.5.K.1-2 MCPR Operating Limit vs,lI, GE8X8EB Fuel, 142a-1 Standard Operating Conditions 3.5.K.1-3 MCPR Operating Limit vs.Y, LTA310 Fuel, 142a-2 Standard Operating Conditions 3.5.K.2-1 MCPR Operating Limit vs.(, BP/P8X8R Fuel, 142a-3 Increased Core Flow /* 3.5.K.2-2 MCPR Operating Limit vs.6, GE8X8EB Fuel, 142a-4 l Increased Core Flow 3.5.K.2-3 MCPR Operating Limit vs. LTA310 Fuel, 142a-5 Increased Core Flow 3.5.K.3 Deleted 142b 3.5.1.E Kf Factor vs. Core Flow 142d 3.5.1.P Deleted 142e' 3.5.1.G Deleted 142f 3.5.1.H MAPLHGR vs. Planar Average Exposure, 142g l Unit 2, P 8X8R Fuel, Type P8DRB285, l 100 mil channels Amendment No. 26, AQ, 45, 45, -iv-l 70, E6, 108, I22, 123

4 7 PBAPS Unit 2 LIST OF FIGURES Figure Title Page 3.5.1.I MAPLHGR vs. Planar Average Exposure 142h Unit 2, P8X8R Fuel, Type P8DRB284H, 80 mil & 100 mil channel & 120 mil channels 3.5.1.J MAPLHGR vs. Planar Average Exposure 142i Unit 2, P8X8R and BP8X8R Fuel, Type P8DRB299 and BP8DRB299, 100 mil channels l 3.5.1.K MAPLHGR vs. Planar Average Exposure 142j Unit 2, P8X8R Fuel (Generic) 3.5.1.L MAPLHGR vs. Planar Average Exposure 142k Unit 2, BP8X8R Fuel, Type BP8DRB299H 3.5.1.M MAPLHGR vs. Planar Average Exposure 1421 Unit 2, GE8X8EB Puel, Type BD319A 3.5.1.N MAPLHGR vs. Planar Average Exposure 142m Unit 2, GE8X8EB, Type BD321A 3.5.1.0 MAPLHGR vs. Planar Average Exposure 142n Unit 2, GE8X8EB, Type LTA310 l l 3.6.1 Minimum Temperature for Pressure Tests 164 1 such as required by Section XI 3.6.2 Minimum Temperature for Mechanical Heatup 164a j or Cooldown following Nuclear Shutdown j l 3.6.3 Minimum Temperature for Core Operation 164b (Criticality) i i 3.6.4 Transition Temperature Shift vs. Fluence 164c l 3.8.1 Site Boundary and Effluent Release Points 216e l 1 6.2-1 Management Organization Chart 244 j 6.2-2 Organization for Conduct of Plant Operations 245 1 1 i Amendment No. 56, 102, 123 -1va-

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~ 1 i PBAPS UNIT 2 1.0 DEP1NITIONS i The succeeding frequently used terms are explicitly defined so l 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 remove 6 and fuel in the vessel. l Normal control rod movement with the control drive hydraulic cystem is not defined as e core alteration. Normal taovement of in-core instrumentation and the traversing in-core probe is not I defined as a core alteration. Average Planar Linear Heat Generation Rate (APLHGR) - The APLHGR 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. l Cold Condition - Reactor coolant temperature equal to or less I than 212 F. 1 i l 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 Ecu1 valent 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, I-132, I-133, I-134, and I-135 actually present. a Amendment No. 102, 127, 123 ] l Unit 2 l PBAPS a d SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING l.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING JNIEGRITY Applicability: Applicability: The Safety Limits established The Limiting Safety System Settings to preserve the fuel cladding apply to trip setting of the j 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: Objectivest i The objective of the Safety The objective of the Limiting Safety j Limits is to establish limits System Settings is to define the l which assure the integrity of level of the process variables at j the fuel cladding. which automatic protective action is initiated to prevent the fuel cladding integrity Safety Limits from being exceeded. Specification: Specification. ) i A. Reactor _ Pressure 2 800 psia The limiting safety system settings and Core Flow 2 10% of Rated shall be as specified below: i A. Neutron Flux Scram i The existence of a minimum

1. APRM Flux Scram Trip Setting j

critical power ratio MCPR (Run Mode) less than 1.07 for two recirculation loop operation, When the Mode Switch is in tne or 1.08 for single loop RUN position, the APRM flux j operation, shall constitute scram trip setting shall be: j l violation of the fuel cladding l l 2ntegrity safety limit. S < 0. 58W + 62% - 0. 58 A W l 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 ceconds as indicated W = Loop recirculating by the process computer. When flow rate in percent i the process computer is out of of design. W is 100 for service this safety limit shall core flow of 102.5 be assuned to be exceeded if million lb/hr or greater, the neutron flux exceeds its scram setting and a control rod ceram does not occur. huendment No. 15, 2/o A2r abo 78:~9~ B E,123

Unit 2 ) PBAPS SAFETY LIMIT LIMITING CAPETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY I 204 = Dif ference between two loop and single Icop effective recirculation drive flow rate at the same core flow. During single loop operation, the reduction in trip setting (-0.58 AW) is accomplished l by correcting the flow input of the flow biased scram to preserve the original (two loop) 1 relationship between LPRM j scram setpoint and i recirculation drive flow or by adjusting the APRM i flux trip setting. d W = 0 for two loop operation. l l i l 1 l l i Amendment No. 75, 323 -9a-e l

PBAPS UNIT 2 SAFETY LIMIT LIM 2 TING SAFETY SYSTEM SETTING 2.1.A (Cont'd) In the event of operation with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modifitd as follows. l l S 5 (0.58 W + 62% -0.58 AW)(FRP) l MFLPD

where, FRP = fraction of rated thermal power (3293 MWt)

MFLPD = maximum fraction of j limiting power density where the limiting power density is 13.4 KW/ft for BP/P8X8R fuel and 14.4 KW/ft for GE8X8EB and LTA310 fuel. The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the d6 sign value of 1. 0, in which case the actual operating value will be used. l

2. APRM--When the reactor mode switch is in the STARTUP pcsition, 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.
4. flhen the reactor mode switch iB in the STARTUP or RUN position, the reactor shall not be operated in the natural circulation flow mode.

1 ~ Amendment No. If, 84, 26, 42, 48, 70. 78, 123

Unit 2 l PBAPS SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING B. Core Thermal Power Limit B. APRM Rod Block Trip Settino (Reactor Pressure 5 800 psia) i When the reactor pressure is SRB $ (0.58 W + 50% - 0.584 W) l < 800 psia or core flow is Tess than 10% of rated, the where: core thermal power shall not exceed 25% of rated thermal SRB e Rod block setting in power. percent of rated thermal , power (3293 MWt) l l W = Loop recirculation flow rate in percent of design. l W is 100 for core flow of l 102.5 million lb/hr or I greater. AW = Difference between two 1 loop and single loop effective recirculation drive flow at the same 3 core flow. During j single loop operation, ( the reduction in trip setting (-0.58 A N) is l l accomplished by correcting j the flow input of the flow biased red block to preserve the original l (two loop) relationship between APRM Rod block I setpoint and recirculation j drive flow or by adjusting l the APRM Rod block trip setting. l 4 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 l follows. l l l l r Amendment No. pp, Jg, 92, pp,.' 79, 7s, 123

Unit 2 PBAPS SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING B. Core Thermal Power Limit B. APRM Rod Block Trip Settina (Reactor Pressure 5 800 psia) SRB 5 (0.58 W + 50% - 0.58 A W) (FRP)l MFLPD 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 fuel and 14.4 KW/ft for GE8X8EB and LTA310 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 reacter 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 (318 inches above vessel zero). l l l Amendment No. 75, Ill, 123 -lla- ~

t i Unit 2

1.1 BASES

FUEL CLADDING INTEGRITY A. Fuel Cladding Integrity Limit at Reactor Pressure t 800 psia and Core Flow 2 10% of Rated l l l 1 The fuel cladding integrity safety limit is set such that no l l fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are l not directly observable during reactor operation the thermal l hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage esuld occur. Although it is recognized that l a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient lir: However, the uncertainties in monitoring the core operating 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 1 and 3 for two l l recirculation. loop operation. The Safety Limit MCPR is increased by 0.01 for single-loop operation as discussed in reference 4. l l l l \\ l ' Amendmer.t No. J6, 56, 123

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 Unit 2 has a sequence annunciation program which will indicate the sequence in which I events such as 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 i i 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 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 l. 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", NEDE-240ll-P-A (as amended). 4. " Peach Bottom Atomic Power Station Units 2 and 3 Single-Loop Operation", NEDO-24229-1, May 1980. o l hmendment No. 23, 36, fB, 70, ZZZ, ' 123 l

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f I j PBAPS UNIT 2

2.1 BASES

FUEL CLADDING INTEGRITY ) l l The abnormal operational transients applicable to operation of l 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 j maximum steady state power which shall not knowingly be exceeded. Conservatism is incorporated in the transient analyses in l estimating the cont' rolling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking i factors, and axial power shapes. These factors are selected j conservatively witn respect to their effect on the applicable transient results as determined by the current analysis model. l Conservatism incorporated into the transient analyses is j documented in NEDE-24011-P-A (as amended). l l l l l l l l l l l l l Amendment No. 22, 26, 123 _y7_ 1 l

PBAPS Unit 2 2.1 BASES (Cont'd) For analyses of the thermal consequences of the transients, a MCPR equal to or greater than the operating limit MCPR given in Specification 3.5.K is conservatively assumed to exist prior to initiation of the limiting transients. This choice of using conservative values of controlling parameters and initiating transients at_the design power level produces more pessimistic l I answers than would result by using expected values of control parameters and analyzing at higher power levels. I Steady state operation without forced recirculation will not be permitted. The analysis to support operation at various power I and flow relationships has considered operation with either one l or two recirculating pumps. l 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. ii. The licensed maximum power level is 3293 MWt. 1 iii. 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 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 l setting, none of the abnormal operational transients analyzed i violate the fuel Safety Limit and there is a substantial margin I from fuel damage. Therefore, the use of flow referenced scram l trip provides even additional margin. Amendment No. 77, 76. 48, 70, 123 -1B-

n l PBAPS Unit 2 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 FSAR which states that the nuclear system safety / relief valves shall prevent opening of the safety valves during normal plant isolations and load rejections. The details of the analysis which show compliance with the ASME Code requirements are presented in subsection 4.4 of the FSAR and i 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 Unit 2. The analysis of the worst overpressure transient is provided in the Supplemental Reload Licensing Submittal and demonstrates margin to the code 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. I i l l s Amendment No. 23, 36, 48, 7 0, 123 - _ _ _ _ _ _

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1 t i PBAPS Unit 2 1 II NOTES FOR TABLE 3.1.1 (Cont'd) 10. The APRM downscale trip is automt,tically bypassed when the IRM instrumentation is operable and net high. 11. An APRM will be considered operable if there are at least l LPRM inputs per level and at least 2 normal complement. 14 LPRM inputs of the l { 12. This equation will be usec in the event of operation with a maximum fraction of limiting power density.(MFLPD) ( than the f raction of ratt.d power (FRP), greater where: l FRP = fraction of ratec thermal power (-3293 MWt). 4 MFLPD = maximum f ractio:. of limiting l power density where the f limiting power density is 4 13.4 KW/ft for BP/ fax 8R fuel ) and 14.4 KW/ft for GE8XBEB and LTA310 fuel. I I i The ratio of FRP to MFLPD shall be sut equal to 1.0 unl'ess the actual operating value is Jess tnan the design value of ( l.0, in which case the actual operating value will be used t i. i W Loop Recli culation flow in 96t cent of design. 1 = is 100 for core flow of 102.,5 million Ib/hr or W i greater. l / AW = the difference between two loop and singl'ellcop l effective ' recirculation drive flow rate at the i same ccre flow. During single loop,cperation, 4 the reduction in trip setting (-0.5E' A.W) is. l accomplished by correcting tt.e flow input of flow biased High Flux trip setting to, preserve ' the-the original (two loop) relationship between, APRM High. Flux setpoint and recirculation drive flow or ty adjusting the APRM Flux trip set'til:g. I o W = 0 for two loop operation. 'd Trip level setting is in percent of rated power &3293 MWt). 13. See Section 2.1.A.l. 1 l I ' Amendment No. 23, AI, H2, 75, ~40-1 23 i r-a

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s m2 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. 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 = mhximum fraction of limiting power density where the limiting power density is 13.4 KW/ft for BP/P8X8R fuel and 14.4 KW/ft for GE8X8EB and LTA310 fuel. J 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). A 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. o 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 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 < 30%. 8. This function is bypassed when the mode switch is placed in Run. Amendment No. 23, /$, 70, 75,1Z3 _

PBAPS Unit 2 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.I Average Planar LHGR 4.5.I Average Planar LHGR During power operation, the APLHGR The APLHGR 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 3 25% rated thermal power. values which have been approved for the respective fuel and lattice types. j When hand calculations are required, I the APLEGR for each type et fuel as l a function of average planar l exposure shall not exceed the limit 3 for the most limiting lattice (excluding natural uranium) shown in the applicable figures for BP/P8X8R, GE8X8EB and LTA310 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 i reduction factors: 0.79 for SP/P8X8R i fuel and 0.73 for GE8X8EB and LTA310 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 1 initiated within one (1) hour to l restore APLHGR to within prescribed limits. If the APLEGR is not returned to within prescribed limits j within five (5) hours reactor power i shall be decreased at a rate which would bring the reactor to the cold shutdown condition within 36 hours unless APLHGR is returned to within l limits during this period. l Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits. i l l 3.5.J Local LHGR, 4.5.J Local LHGR 1 l During power operation, the linear The LHGR as a function of core 1 j heat generation rate (LHGR) of any height shall be checked daily l rod in any fuel assembly at any during reactor operation et axial location shall not exceed 1 25% rated thermal power. design LEGR. LHGR < LHGRd LHGRd = Design LHGR 13.4 KW/ft for BP/P8X8R fuel 14.4 KW/ft for GE8X8EB and LTA310 f uel r 1 Amendment No. 40, AS, 7g, 7g, gg,-133a-l J 5,123 i 1

l PBAPS Unit 2 l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.K Minimum Critical Power 4.5.K Minimum Critical Power Ratio (MCPR) (Cont'd) Ratio (MCPR) (Cont'd) number of active control 2. Except as specified in 3.5.K.3, Ni = l the Operating Limit MCPR Values rods measured in the ith are as follows: surveillance test. a. If requirement 4.5.K.2.a is met: The Operating Limit MCPR values are as given in Table 3.5.K.2. 17i = 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 i values as a function of T' scram time (TTB) is calculated are given in Figures as follows: s /2 3.5.K.1-1, 3.5.K.1-2, l 3.5.K.1-3, 3.5.K.2-1, Ni T n J C' 3.5.K.2-2, and 3.5.K.2-3. TIB = p +1.65 l' Ni/ i=1 Where: Where: l l 7 = f ave ' TB p= mean of the distribution for 1

0. 90 - TB average scram insertion time to the 20% position = 0.694 sec.

I 3. The Operating Limit MCPR values Ni total number of active control = shall be as given in Table 3.5.K.3 rods measured in specification if the Surveillance Requirement 4.3.C.1 of Section 4.5.K.2 to scram time test control rods is not performed. c' = standard deviation of the distribution for average scram insertion time to the 20% position = 0.016. I 1 Amendment No. 86, 123 -133c-I i

PBAPS Unit 2 Table 3.5.K.2 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 Standard Operating Conditions BP/P8X8R 1.24 1.27 l GE8X8EB 1.24 1.28 LTA310 1.24 1.28 Increased Core Flow BP/P8X8R 1.24 1,29 GE8X8EB 1.24 1.30 LTA310 1.24 1.31 If requirement 4.5.K.2.a is met. These values shall be increased by 0.01 for single loop operation. Amendment No. 56, J 0 E, 123 -133d-1 l l l L_--_________-____-_-____-___

l PBAPS Unit 2 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 Standard Operating Conditions l BP/P8X8R 7 1.28 GE8X8EB 1.31 1.29 LTA310 1.32 1.29 1.32 _ Increased Core Flow BP/P8X8R 1.28 GE8X8EB 1.33 1.29 LTA310 1.34 1.29 1.35 If Surveillance Requirement 4.5.K.2 is not performed. These values shall be increased by 0.01 for single loop operhtion. I I I 1 l l l l Amendment No. gg, 708,123 -133e-I

^ PBAPS Unit 2 3.5 BASCS (Cont'd.) 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 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 with the High Pressure Service Water Pumps is also associated with the Emergency Service Water pumps, and is specified in Specification 3.9. I. Average Planar LHCR 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 3 l generation rate of all the rods of a fuel assembly at any axial location and is only dependent, distribution within an assembly. secondarily, on the rod-to-rod power The peak clad temperature is calculated assuming a LHGR for the highgst powered rod which is equal 4 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 its local peaking factor.the LHGR of the highest powered rod divided by The limiting value for APLHGR is shown in the applicable figure for each fuel type. Only the most limiting and least limiting APLHGR operating limits are shown in the figures for the multiple lattice fuel types. Compliance with the lattice-specific, approved APLHGR limits is ensured by using I i the process computer. When an alternate method to the process is required (i.e. hand calculations and/or alternate computer computer i 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 j Reference 4. Input and model changes in the Peach Bottom loss-of-I i l coolant analysis which are different from the previous analyses performed with Reference 4 are described in detail in Reference 8. These changes to the analyGis include: (1) consideration of the counter current flow limiting (CCFL) effect, (2) corrected code i inputs, and (3) the effect of drilling alternate flow paths in the bundle lower tie plate. i 1 1 Amendmbnt No. 2b, 26, AO $2, l -140-l 70, BE, 123

PBAS Unit 2 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. l Analysis of the abnormal operational transients is presented in Reference 7. Input data and operating conditions used in this analysis are shown in Reference 7 and in the Supplemental Reload Licensing Analysis. i l 3.5.L. Averace Planar LHGR (APLHGR), Local LEGR and Minimum l Critical Power Ratio (MCPR) l In the event that the calculated value of APLHGR, LHGR or MCPR l exceeds its limiting value, a determination is made to ascertain the cause and initiate corrective action to restore the value to within prescribed limits. The status of all indicated limiting fuel hundles is reviewed as well as input data associated with the limiting values such as power distribution, instrumentation 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. i 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 I 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. 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. i 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 indication, etc., corrective action is initiated within one hour of an indicated v.alue exceeding limits. Verification that the indicated value is within prescribed limits is obtained within five hours of the initial indication. Such an i invalid indication would not be a violation of the limiting condition fcc operation and therefore would not constitute a reportable occurrence. Wmendment No. 22, 3g, f8, 70, gg,-140b-1 23

i i PBAPS Unit 2 j I 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 concurrent 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 hour) and restore j the calculated value of APLHGR, LHGR or MCPR to within prescribed limits (5 hours) are adequate. 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 UE Model fcr 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, i NEDE 20566 (Draft), August 1974. I 5. General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by letter, l G. L. Gyorey to Victor Stello, Jr., dated December 20, 1974. l 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 l Power Station Unit 2, NEDO-34081, December 1977, and for Unit 3, NEDO-24082, December 1977. 9. Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2, Supplement 1, NEDE-24081-P, November 1986. Amendment No. 27, 36, 38, 45,-140c-7%, 86, 123

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