ML20003E147
| ML20003E147 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 03/27/1981 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20003E138 | List: |
| References | |
| NUDOCS 8104020330 | |
| Download: ML20003E147 (41) | |
Text
{{#Wiki_filter:r PBAPS Unit 3 LIST OF TABLES Table Title Page 4.2.B Minimum Test and Calibration Frequency 81 for CSCS 4.2.C Minimum Test and Calibration Frequency 83 for Control Rod Blocks Actuation 4.2.D Minimum Test and Calibration Frequency 84 for Radiation Monitoring Systems 4.2.E Minimum Test and Calibration Frequency 85 for Drywell Leak Detection
- 4.2.F Minimum Test and Calibration Frequency 86 for Surveillance Instrumentation 4.2.G Minimum Test and Calibration Frequency 88 for Recirculation Pump Trip 3.5.K.2 Operating Limit MCPR Values for 133d Various Core Exposures 3.5.K.3 Operating Limit MCPR Values for 133e Various Core Exposures 3.5-1 DELETED 4.6.1 In-Service Inspection Program for Peach 150 Bottom Units 2 and 3 3.7.1 Primary Containment Isolation Valves 179 3.7.2 Testable Penetrations With Double 184 0-Ring Seals 3.7.3 Testable Penetrations With Testable 184 Bellows 3.7.4 Primary Containment Testable Isolation 185 Valves 4.8.1 Radioactive Liquid Waste Sampling 210 and Analysis 4.8.2 Radioactive Gaseous Waste Sampling 211 and Analysis' 3.11.D.1 . Safety Related Shock Suppressors 234d 3.14.C.1 -Fire Detectors 240k
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PBAPS Unit 3 LIST OF FIGURES Fiqure Title Pace 1.1-1 APRM Flow 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 Required Volume and Concentration of 122 Standby Liquid Control System Solution 3.4.2 Required Temperature vs. Concentration 123 for Standby Liquid Control System Solution 3.5.K.1 MCPR Operating Limit Vs. Tau 142 8x8 and 8x8R Fuel 3.5.K.2 MCPR Operating Limit Vs. Tau, P 8X8R Fuel 142a
- Figures 3.5.1.A and 3.5.1.B (7x7 Fuel) deleted (PB3. Cycle 5 - all 8X8 core) 3.5.1.C MAPLHGR Vs. Planar Average Exposure, 142b Unit 3, 8x8 Fuel, Type H 3.5.1.D MAPLHGR Vs. Planar Average Exposure, 142c Unit 3, 8x8 Fuel, Type L 3.5.1.E Kf Factor Vs. Core Flow 142d 3.5.1.F MAPLHGR Vs. Planar Average Exposure, 142e Unit 3, 8x8 PTA Fuel 3.5.1.G MAPLHGR Vs. Planar Average Exposure, 142f Unit 3, 8x8R Fuel 3.5.1.H MAPLHGR Vs. Planar Average Exposure, 142g Unit 3, P 8X8R Fuel (P8DRB284H)
_ 3.5.1.I MAPLHGR vs Planar Average 142h Exposure, Unit 3, P8X8R Fuel (P8DRB299) 3.5.1.J MAPLHGR vs Planar-Average Exposure, Unit 3, P8X8R Fuel (Generic) 1421 3.6.1 Minimum Temperature for Pressure Tests 164 such as required by Section X1 3.6.2 Mini:aum Temperature -for Mechanical Heatup 164a or Cooldown following Nuclear Shutdown 3,6.3 Minimum Temperature for Core Operation 164b (Criticality) 3.6.4 -Transition Temperature Shift vs. Fluence 164c 6.2-1 Management Organization Chart 244 6.2-2 Organization for Conduct of Plant Operation 245 r
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r PBAPS 1.0 DEFINITIONS (Cont'd) component, or device to perform its function are also capable of performing their related support function. Operatino - Operating means that a system or component is performing its intended functions in its required manner. Operatino Cycle - Interval between the end of one refueling outage for a particular unit and the end of the next subsequent refueling outage for the same unit. Primary Containment Intecrity . Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:
- 1. All non-automatic containment isolation valves on lines connected to the reactor coolant system or containment which are not required to be open during accident conditions are closed. These valves may be opened to perform necessary operational activities.
- 2. At least one door in each airlock 10 closed and sealed.
- 3. All automatic containment isolation valves are operable or deactivated in the isolated position.
- 4. All blind flanges and manways are closed.
Protective Action - An action initiated by the protection system when a limit is reached. A protective action can be at a channel or system level. Protective Function - A system protective action which results from the protective action of the channels monitoring a particular. plant condition. Rated Power - Rated power refers to operation at a reactor power of 3,293 MWt; this is also termed 100 percent power and is the maximum power level authorized by the operating license. Rated steam flow, rated coolant flow, rated neutron flux, and rated nuclear system pressure refer to the values of these parameters when the reactor is at rated power. F 4_
4 PBAPS Unit 3 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING ' 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY . Applicability: Applicability: The Safety Limits established The Limiting Safety System Settings to preserve the fuel cladding apply to trip settings of the instru-integrity apply to those ments and devices which are provided variables which monitor the to prevent the fuel cladding integrity l ' fuel thermal behavior. Safety Limits from being exceeded.
. Obiectives: Obiectives: ,
The objective of the Safety The objective of the Limiting Safety Limits is to establish limits System Settings is to define the level which assure the integrity of of the process variables at which auto - the fuel cladding. matic protective action is initiated
- to prevent the fuel cladding integrity Safety Limits from being exceeded.
Specification: Specification: The limiting safety system settings shall be as specified below: A. Reactor pressure 2E00 psia A. Neutron Flux Scram and Core Flow 210% of Rated The existence of a minimum 1. APRM Flux Scram Trip Settino critical power ratio MCPR less (Run Mode) than 1.07 shall constitute violation of the fuel cladding When the Mode Switch is in the integrity safety limit. RUN position, the APRM flux scram trip setting shall be: To ensure that this safety limit is not exceeded, neutron S 5 0.66W +54% flux shall not be above.the scram setting established in where: specification 2.1.A for longer than 1.15 seconds as. S = Setting in percent of - indicated by the process com- ' rated thermal power puter.'When the process com- (3293 MWt) puter is out of service this
. safety limit shall-be assumed W = Loop recirculating flow to be exceeded if the neutron rate in percent of design flux exceeds its scram setting W is 100 for core flow and a control rod scram does of 102.5 million Ib/hr not occur. / or greater.
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PBAPS Unit 3 SAFETY LIMIT LIMITING 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 modified as follows: S < (0.66 W + 54%) (FRP) 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 all 8x8 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.
- 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.
- 4. When the reactor mode switch is in the STARTUP or RUN position, the reactor shall not be operated in the natural circulation flow mode.
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( PBAPS Unit 3 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING B. Core Thermal Power Limit B. APRM Rod Block Trip Setting (Reactor Pressure 5 800 psia) SRB $ 0.66W + 42% When the reactor pressure is 5 300 psia or core flow is where: less than 10% of rated, the core thermal power shall not SRB = Rod block setting in exceed 25% of rated thermal percent of rated thermal power. power (3293 MWt) W = Loop recirculation flow rate in percent of design W is 100 for core flow of 102.5 million Ib/hr
. or greater.
In the event of operation with a maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows: SRB 1 (0.66 W + 42%) ( FRP) 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 all 8X8 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--1538 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 17.1 in. above the top of the normal active fuel zone. r f PBAPS Unit 3
1.1 BASES
FUEL CLADDING INTEGRITY A. Fuel Claddino Inteority Limit at Reactor Pressure > 800 psia and Core Flow >10% of Rated The tuel cladding integrity safety limt 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 not directly observable during reactor operation the thermal and 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 transition is calculated to occur has been adopted as a convenient limit. 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 li/it 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. l I l r L
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PBAPS Unit 3 1.1.A BASES (Cont'd) B. Core Thermal Power Limit (Reactor Pressure < 800 psia on Core Flow < 10% of Rated) The use of the GEXL correlation is not valid for the critical power calculations at pressures below 800 psia or core flows less than 10% of rated. Therefore, the fuel cladding integrity safety limit is established by other means. This is done by establishing a limiting condition of core thermal power operation with the following basis. Since the pressure drop in the bypass region is essentially all elevation head which is 4.56 psi the core pressure drop at low power and all flows will always be greater than 4.56 psi. Analyses show that with a flow of 28 x 103 lbs/hr bundle flow, bundle pressure drop in nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.56 psi driving head will be greater than 28 x 102 lbs/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors this corresponds to a core thermal power of more than 50%. Therefore a core thermal power limit of 25% for reactor pressures below 800 psia or core flow less than 10% is conservative. C. Power Transient Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 1.1.A or 1.1.B will not be exceeded. Scram times are checked periodically to assure the insertion times are adequate. The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage. t l r
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2.1 BASES
FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the Peach Bottom Atomic Power Station units have been analyzed throughout the spectrum of planned operating conditions up to the thermal power condition of 3440 MWt. 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 analysis of fast pressurization transients as described in reference 3. Conservatism is incorporated in all other transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and axial power shapes. These factors are selected conservatively with respect to their effect on the applicable transient results as determined by the current analysis models. These transient models evolved over many years, have been substantiated in operation as a conservative tool for evaluating reactor dynamic performance. Results obtained from a General Electric boiling water reactor have been compared and results are summarized in Reference i for cold water events, and in Reference 2 for pressurization events. The absolute value of the void reactivity coefficient used in the analysis is conservatively estimated to be about 25% greater than the nominal maximum value expected to occur during the core lifetime. The scram worth used has been derated to be equivalent to approximately 80% of the total scram worth of the control rod. The scram delay time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by Technical Specifications. Active coolant flow is equal to 88% of total core flow. The effect of scram worth, scram delay time and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion. The rapid insertion of negative reactivity is assured by the time requiremente for 5% and 25% insertion. By the time the rods are 60% inserted, approximately fcur dollars of negative reactivity have been inserted which strongly turns the transient, and accomplishes the desired effect. The times for 50% and 90% insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady state condition, r
PBAPS Unit 3 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 effect on reactor safety because of the reculting 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. The scram trip setting must be adjusted to assure that the LHGR transient peak is not increased for any combination of maximum fraction of limiting power density (MFLPD) and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.1.A.1, when the MFLPD is greater than the fraction of rated power (FRP). 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 HCPR greater than the operating limit given in Specification 3.5.K. 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 margin between the setpoint and the Safety Limit, 25 percent of rated. The margin ic 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 patterns are constrained to be uniform by operating procedures backed up by the Rod Worth Minimizer and Rod Sequence Control System. Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probabic cause of significant power rise. Because the flux distribution sa? ciated with uniform rod withdrawals does not involve hi9n local peaks,'and because several. rods must be, moved to change power by a significant percentage of rated power, the rate of power is very slow. Generally, the heat flux is in near' equilibrium with the fission rate. In an assumed uniform 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.
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PBAPS Unit 3 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 120 divisions for cnat 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 d'ue 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 end the IRM system is not yet on scale. This condition exists at quarter rod density. Additional conservatism was taken in this analyses by assuming that the IRM channel closest to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scramed 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 Rod Block Trip 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. As with the APRM scram trip setting, the APRM rod block trip setting is adjusted if the maximum fraction of limiting power density exceeds the fraction of rated power, thus preserving the APRM rod block safety margin. As with the scram setting, this may be accomplished by adjusting the APRM gain.
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/ PBAPS Unit 3 2.1 BASES (Cont'd.) C. Reactor Water Low Level Scram and Isolation (Exceot Main Steamlinesl The set point for the low level scram is above the botto= of the separator skirt. This level has been used in transient analyses dealing with coolant inventory decrease. The results reported in FSAR subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than the fuel cladding integrity safety limit in all cases and system pressure does not reach the safety valve settings. The scram setting is approximately 31 in, below the normal operating range and is thus adequate to avoid spurious scrams. D. Turbine Stop Valve Closure Scram The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid _ closure of the turbine stop valves. With a scram trip setting of less than or equal to 10 percent of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the fuel cladding integrity safety limit even during the worst case transient that assumes the turbine bypass is closed. This scram is bypassed when turbine steam flow is below 305 of rated, as measured by turbine first stage pressure. E. Turbine Control valve Scram The turbine control valve fast closure scram anticipates the pressure,. neutron flux and heat flux-increase that could result from fast closure of the turbine control valves due to a load rejection exceeding the capacity of the bypass valves or a failure in the hydraulic control system which results in a loss of oil pressure. This scram is initiated from pressure switches in the hydraulic control system which sense loss of oil pressure i due to 'he opening of the fast acting solenoid valves or a failure in the hydraulic control system piping. Two turbine first stage pressure switches for each trip system initiate automatic bypass of the turbine control valve fast closure scram when the first stage pressure is below that required to produce 30% of rated power. Control valve closure time is approximately twice as long as that for'stop valve closure. F t 2.1 BASES (Cont'd) L. References
- 1. Linford, R. B., " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor", NEDO 10802, February 1973.
- 2. " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors", NEDO 24154 and NEDE 24154-P, Volumes I, II, and III.
- 3. " Safety Evaluation for the General Electric Topical Report Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors NEDO-24154 and NEDE 24154-P, Volumes I, II, and III.
9 F PBAPS Unit 3 1.2 BASES The teactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products. It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel. The pressure safety limit of 1325 psig as measured by the vessel steam space pressure indicator assures not exceeding 1375 psig at the lowest elevation of the reactor coolant system. The 1375 psig value is derived from the design pressures of the reactor pressure vessel (1250 psig at 5750F) and coolant system piping (suction piping: 1148 psig at 5620F; discharge piping: 1326 psig at 5620F). The pressure safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes: ASME Boiler and Pressure Vessel Code, Section III for the pressure vessel and ANSI B31.1.0. for the reactor coolant system piping. The ASME Boiler and Pressure Vessel Code permits pressure transients up to 10% over design pressure (110% X 1250 = 1375 psig), and the ANSI code permits pressure transients up to 20% ozer the design pressure (120% X 1148 = 1378 psig; 120% X 1326 = 1591 psig). A safety limit is applied to the Residual Heat Removal System (RHRSi when it is operating in the shutdown ., ccaling mode. At this time it is included in the reactor coolent system. 6 e
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l i . PBAPS Unit 3 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's compliance with the ASME Code requirements are presented in subsection 4.4 of the FSAR 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 Unit 3. The analysis of the worst I overpressure transient, is provided in the Supplemental Reload Licensing Safety Evaluation and demonstrates margin tc the code allowable overpressure limit of 1375 psig. The analysis of the plant isolation transient is provided in the Supplemental Reload Licensing Safety Evaluation and demonstrates that the. safety valves will not open. The' safety / relief valve settings satisfy the Code requirements that the lowest valve set point be at or below the vessel design i pressure of 1250 psig. These settings are also sufficiently [ 'above the normal' operating pressure range to prevent unnecessary l cycling caused by minor transients. The design pressure of the shutdown cooling piping of the L Residual Heat Removal System is not exceeded with the reactor [ vessel steam dome less than 75 psig. l l I r l i 1
( PBAPS Unit 3 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. This equation will be used in the event of operation with a maximum fraction of limiting power osnsity (MFLPD) greater than the fraction of rated power (FRP), where:
FRP = fraction of rated thermal power (3293MWt). MFLPD = maximum fraction of limiting power density where the limiting power density is 13.4 KW/ft for all 8x8 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 l3293 MWt).
- 13. See Section 2.1.A.1.
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PBAPS Unit 3 NOTES FOR TABLE 3.2.C
- 1. For the startup and run positions of the Reactor Mode Selector fiitch, there shall be two operacle or tripped trip The SRM and IRM blocks need not systems be operablefi.r each function.
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 = maximum fraction of limiting power density where the limiting power density is 13.4 KW/ft for all 8x8 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).
- 3. IRM downscale is bypassed when it is on its lowest range.
- 4. This function is bypassed when the count rate is 2 100 cps.
- 5. One of the four SRM inputs may be bypassed.
- 6. This SAM function is bypassed when the IRM range switches are on range 8'or above.
- 7. The trip is bypassed when the reactor power is 5 30%.
- 8. This function is bypassed'when the mode switch is placed in Run.
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/ PBAPS 3.2 BASES (Cont'd) Presuure instrumentation is provided to close the main steam isolation valves in RUN Mode when the main steam line pressure drops below 850 psig. The Reactor Pressure Vessel thermal transient due to an inadvertent opening of the turbine bypass valves when not in the RUN Mode is less severe than the loss of feedwater analyzed in section 14.5 of the FSAR, therefore, closure of the Main Steam Isolation valves for thermal transient protection when not in RUN mode is not required. The HPCI high flow and temperature instrumentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves. Tripping logic for the high flow is a 1 out of 2 logic. Temperature is monitored at four (4) locations with four (4) temperature sensors at each location. Two (2) sensors at each location are powered by "A" DC. control bus and two (2) by "B" DC control bus. Each pair of sensors, e.g., "A" or "B" at each location are physically sepacated and the tripping of either "A" or "B" bus sensor will actuate HPCI isolation valves. The trip settings of $300% of design flow for high flow and 2000F for high temperature are such that core uncovery is prevented and fission product release is within limits. The RCIC high flow and temperature instrumentation are arranged the same as that for the HPCI. The trip setting of $300S. for high flow and 2000F for temperature are based on the same criteria as the HPCI. The Reactor Water Cleanup System high flow and temperature instrumentation are arranged similar to that for the HPCI. The trip settings are such that core uncovery is prevented and fission product ~ release is within limits. The instrumentation which initiates CSCS action is arranged in a dual bus system.. As for other vital instrumentation arranged in this fashion, the Specification preserves the effectiveness of the system even during periods when maintenance er testing is beine performed. An exception to this is when logic functional tesLing is being performed. The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to the fuel cladding integrity safety limit. The trip logic for this function is 1 out of n: e.g., any trip on one of 6 APRM's, 8 IRM's, or 4 SRM's will result in a rod block. The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or calibration. This time period is only 3% of the operating time in a month and does not significantly. increase the risk of preventing an inadvertent control rod withdrawal. ,,
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PBAPS 3.2 BASES (Cont'd) The APRM rod block function is flow biased and prevents a significant reduction in MCPR, especially during operation at reduced flow. The APRM provides gross core protection: i.e., limits the gross core power increase from withdrawal of control n normal withdrawal sequence. rods in t'e The trips are set so that MCPR is maintained greater than the fuel cladding integrity safety limit. The RBH rod block function provides local protection of the core; i.e., the prevention of Loiling transition in the local region of the core, for a single rod withdrawal error from a limiting control rod pattern. The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. A downscale indication on an APRM or IRH is an indication the instrument has failed or the instrument is not sensitive enough. In either case the instrument will not respond to changes in the control rod motion and thus, control rod motion is prevented. The downscale trips are set at 2.5 indicated on scale. The flow comparator and scram dischaige volume high level components have only one logic channel and are not required for safety. The flow comparator must be bypassed when operating with one recirculation water pump. The refueling interlocks also operate one logic channel, and are required.for safety only when the mode switch is in the refueling position. For effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate. The arrangenent of the tripping contacts is such as to provide this function when necessary and minimize spuricus operation. The trip settings given in the specification are adequate to assure the above criteria are met. The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.
-Two air ejector off-gas monitors are provided and when their trip point is reached, cause an isolation of the air ejector off-gas line. Isolation is initiated when both instruments reach their high trip point or one has an upscale F
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( PBAPS Unit 3 LIMITING CONDITIONS FOR OPERATION , SURVEILLANCE REQUIREMENTS 3.3.B Control Rods (Cont'd) 4.3.B Control Rods (Cont'd)
- 4. Control rods shall not be 4. Prior to control rod with-withdrawn for startup or drawal for startup or during refueling unless at least refueling, verify that at two source range channels least two source range channelg have an observed count have an observed count rate rate equal to or greater of at least three counts per than three counts per second.
second.
- 5. During operation with 5. When a limiting control rod limiting control rod pat- pattern exists, an instru-terns, as determined by the -
ment functional test of the designated qualified person- RBM shall be performed nel, either: prior to withdrawal of the designated rod (s).
- a. Both RMB channels shall be operable, or
- b. Control rod withdrawal shall be blocked, or
- c. The operating power level shall be limited so that the MCPR will remain above the fuel cladding integrity safety limit assuming a single error that results in complete withdrawal of a single operable control rod.
C. Scram Insertion Times C. Scram Insertion Times
- 1. The average scram inser- 1. After each refueling outage, tion time, based on the and prior to synchronizing deenergization of the scram the main turbine generator pilot valve solenoids as initially following restart time zero, of all operable of the plant, all operable control rods in.the reactor fully withdrawn insequence power operation condition rods shall be scram time shall be no greater thans tested during operational hydrostatic testing or during startup from the fully withi drawn position with the nucleag
% Inserted from Avg. Scram Inser- system pressure above 800 psig.
Fully Withdrawn tion Times (sec) 5 0.375 20 .0.90 50 2.0 90 3.5 e
-103-
P PBAPS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.3.C (Cont'd) 4.3.C (Cont'd) I After exceeding 30 percent power all previously untested operable control rods shall be tested as described above prior to exceeding 40 percent power.
- 2. Whenever such scram time measurements are made (such as when a scram occurs and the scram insertion time recorders
- 2. The average of the scram are operable) an evaluation ,
insertion times for the . shall be made to provide three fastest control reasonable assurance that rods of all groups of four proper control rod drive control rods in a two-by-two performance is being maintaine: array shall be no greater than:
% Inserted From Avg. Scram Inser-Fully Withdrawn tion Times (Sec,1 5 0.39G 20 0.954 50 2.120 90 3.8
- 3. The maximum scram insertion time for 90% insertion of any operable control rod shall not exceed 7.00 seconds.
F
-104-
/ PBAPS Unit 3 3.3 and 4.3 BASES (Cont'd) C. Scram Insertion Times The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than the fuel cladding integrity safety limit. Analysis of the limiting power transients shows that the negative reactivity rates resulting from the scram with the average response to all drives as given in the above Specification, provide the required protection. The numerical values assigned to the specified scram performance are based on the analysis of data from other BWR's with control rod drives the same as those on Peach Bottom. The occurrence of scram times within the limits, but significantly longer than the average, should be viewed as an indication of a systematic problem with control rod drives especially if the number of drives exhibiting such scram times
. exceeds one control rod of a (5x5) twenty-five control array.
In the analytical treatment of the transients, which are assumed to scram on high neutron flux, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of negative reactivity insertion. This is adequate and conservative when compared to the typical time delay of about 210 milliseconds estimated from scram test results. The 290 milliseconds used in the analyses consists of 90 milliseconds for sensor.and circuit delay and 200 milliseconds to start of control rod motion. In addition the control rod drop accident has been analyzed in NEDO-10527'and its supplements 1 & 2 for the scram times given in Specification-3.3.C.
-Surveillance requirement 4.3.C was originally written and used as a diagnostic surveillance technique during pre-operational and startup testing of Dresden 2 & 3 for the early discovery and identification of significant changes in drive scram performance following major changes in plant operation. The reason for the application of this surveillance was the unpredicatable and degraded scram performance of drives at Dresden 2. The cause of the slower scram performances has been conclusively F
--111-t
PBAPS Unit 3 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.I Average Planar LHGR 4.5.I Averace 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 i of average planar exposure shall not exposure shall be checked daily I exceed the limiting value shown in during reactor operation at Figure 3.5.1.C, D, F, G, H, I&J >25% rated thermal power. as applicable. 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 pre-scribed limits. If the APLHGR is note 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 hours unless APLHGR is returned to within limits during this period. Surveil-lance and corresponding action shall continue until reactor operation is within the prescribed limits. 3.5.J Local LHGR 4.5.J Local LHGR l l During power operation, the linear The LHGR as a function of core heat generation rate (LHGR) of height shall be checked daily l any rod in any fuel assembly at during reactor operation at any axial location.shall not exceed >25% rated thermal power. design LHGR. LHGR$LHGRd l LHGRd = Design LHGR 13.4 kW/ft for all.8x8 fuel l 1 l
-133a-
/
PBAPS Unit 3 ! LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS j 3.5.J Local LHGR (Cont'd) L 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) 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 hours unless LHGR is returned to within limits , during this. period. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits. 3.5.K Minimum Critical Power 4.5.K Minimum Critical Power Ratio (MCPR) Ratio (MCPR)
- 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 g25% rated thermal power.
for each type of fuel shall be equal to or greater than the value 2. Verification of the given.in Specification 3.5.K.2 or . Applicability of 3.5.K.2.a 3.5.K.3 times Kf, where Kf is as Operating Limit MCPR Values I shown-in Figure'3.5.1.E. If at shall be performed every 120 any time during operation it operating days by scram time is determined by normal testing 19 or more control surveillance that the limiting rods on a rotating basis and
- .value for MCPR is being- performing the following
exceeded, action shall be i initiated within one,(1) hour a. .The average scram time to restore MCPR-to within prescribed to the 20% insertion limits. If'the MCPR is not gsition shall be: returned to within prescribed limits within five (5) hours, Tave 5 [h-
. reactor power shall be decreased b. The average scram time
-at a rate which'would bring the to the 20% insertion reactor to the cold shutdown . position is determined condition within 36 hours unless .as follows:
MCPR is: returned to within limits during this period. Surveillance-and corresponding. action shall continue until' reactor operation: Tave=n I i=1 NiD is within the prescribed m. limits. . I: Ni i=1. where: n'=-number of
, . surveillance = tests performed.
"to date in the cycle.
-133b-
( PBAPS LIMITING CONDITIONS FOR OPERATION SURVE1LLANCE REOUIREMENTS 3.5.K. Minimum Critical Power 4.5.K. M'91 mum 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,
- a. If requirement 4.5.K.2 is met:
The Operating Limit MCPR values r~- are as given in Tables 3.5.K.1 /i = average scram time to and 3.5.K.2. the 20% insertion position of all rods measured in
- b. If requirement 4.5.K.2 is not -
the ith surveillance test. met: The Operating Limit MCPR c. The adjusted analysis mean values as a function of r scram time (rB) is calculated are as given in Figures as follows: 3.5.K.1 and 3.5.K.2. j/ 7 Ns
}gjp
~/ B = n +1.65, CF~
Ni i=1 Where: Where: l~= Eave 'EB , = mean of the distribution 0.90 -l'B for average scram insertio3 time to the 20% position = 0.710 sec
- 3. The Operating Limit MCPR values N2 = 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 C7" = standard deviation of the performed distribution for average scram insertion time to the 20% position = 0.053.
E
-133c-
( PBAPS Unit 3 Table 3.5.K.2 OPERATING LIMIT MCPR VALUES FOR VARIOUS CORE EXPOSURES
- MCPR Operating Limit
~ Fuel Type For Ircremental Cycle Core Average Exposure BOC to 2000 MWD /t 2000 MWD /t before EOC Before EOC To EOC 8x8 1.24 1.27 PTA &P 8X8R 1,25 '
1.30 8x8R 1.24 1.27
+
I
- If. requirement 4.5.K.2 is met.
i r
-133d-
PBAPS Unit 3 l l Table 3.5.K.3 l ? i i OPERATING LIMIT MCPR VALUES ! FOR VARIOUS CORE EXPOSURES
- MCPR Operating Limit Fuel Type For Incremental Cycle Core Averace Exposure BOC to 2000 MWD /t 2000 MWD /t before EOC Before EOC To EOC 8x8 1.33
- 1.39 PTA &P 8X8R 1.36 1.42 8x8R 1.'33 1.39 p
- If requirement 4.5.K.2 is not met.
l t
+
Y
-133e-
PBAPS Unit 3 3.5 BASES (Cont'd.) H. Enaineerino Safeguards Compartments Cooling and Ventilation One unit cooler in each pump compartment is capable of providing adequate ventilation flow and cooling. Engineering analyses indicate 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. V?ntilation 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 P.pecification assures that the peak cladding temperature fo11 ewing 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) follcwing 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 factor. The limiting value for APLHGR is shown in Figure 3.5.1.C, D, F, G, H, I, and J. The calculational procedure used to establish the APLHGR shown on Figures 3.5.1.C, D, Fj G, H, I, and J is based on a loss-of-coolant accident analysis. The analysic was performed using General Electric (GE) calculationci models which are consistent g with the requirements of Appendix .R to 10 CFR Part 50. A complete discussion of each code employed in the analysis ir 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 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. F
-140-
/
PBAPS Unit 3 3.5.I BASES (Cont'd.) J. Local LHGR This specification assures that the linear heat generation rate in any 8X8 fuel rod is less than the design linear heat generation. The maximum LHGR shall be checked daily during reactor operation at >25% power to determine if fuel burnup, or control rod movement has caused changes in power distribution. For LHGR to be at the design LHGR below 25% rated thermal power, the peak local LHGR must be a factor of approximately ten (10) greater than the average-LHGR which is precluded by a considerable margin when employing any permissible control rod pattern. K. Minimum Critical Power Ratio (MCPR) Operatino Limit MCPR The required operating limit MCPR's at steady _ state operating conditions are derived'from the established fuel cladding integrity Safety Limit MCPR and analyses of the abnormal operational transients presented in Supplemental Reload Licensing Analysis and ReferenceL7. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety-Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.1. To assure that the fuel cladding integrity Safety Limit is not violated during any anticipated abnormal operational-transient, the most limiting transients have been analyzed to determine which result in the largest reduction in critical power ratio (CPR).- The transients evaluated are as described in reference 7. F
-140a-
[ PBAPS Unit 3 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. Two codes are used to analyze the rod withdrawal error transient. The first code simulates the three dimensional BWR core nuclear and thermal-hydraulic characteristics. Using this code a limiting control rod pattern is determined; the following assumptions are included in this determination: (1) The core is operating at full power in the xenon-free condition. (2) The highest worth control rod is assumed to be fully inserted. (3) The analysis is performed for the most reactive point in the cycle. (4) The control rods are assumed to be the worst possible pattern without exceeding thermal limits. (5) A bundle in the vicinity of the highest worth control rod is assumed to be operating at the maximum allowable linear heat generation rate. l (6) A bundle in the vicinity of the highest worth control rod is assumed to be operating at the minimum. allowable critical power ratio.
'The three-dimensional BWR code then simulates the core response to the control rod withdrawal error. The second code calculates the Rod Block Monitor. response to the rod withdrawal error. This i code simulates the Rod Block Monitor under selected failure conditions (LPRM) for the core response (calculated by the 3 -
dimensional BWR simulation code) for the control rod withdrawal. The analysis of the rod withdrawal error for Peach Bottom Unit 3 considers the continuous withdrawal of the maximum worth control i rod at its maximum drive speed from the reactor which is operating with the limiting control rod pattern as discussed above. i I l
- + -140b-
/
PBAPS Unit 3 3.5.K BASES (Cont'd.) A brief summary of the analytical method used to determine the nuclear characteristics is given in Section 3 of Reference 7. Analysis of the abnormal operational transients is presented in Section 5.2 of Reference 7. Input data and operating conditions used in this analysis are shown in Table 5-8 of Reference 7 and in the Supplemental Reload Licensing Analysis. L. Average 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 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 (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 incore locations is obtained and the power distribution, APLHGR, LHGR and MCPR calculated. Corrective action is inittated 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.
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 l 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.
-140c-
J PBAPS Unit 3 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 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 Gen'ral e 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. General Electric Company Analytica] Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566 (Draft), August 1974.
5.- General Electric Refill Reflood Calculation (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 Boiling Water Reactor Generic Reload Fuel
[ Application. NEDO-24011-P-A. l l
- 8. Loss-of-Coolant Accident Analysis For Peach Bottom Atomic Power Etation Unit 3, NEDO-24082, December 1977.
I l r
-140d-
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-42h-
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- 11421-
I e. PBAPS Unit 3 3.6.A & 4.6.A. Bases (Cont'd) The vessel pressurization temperatutes at any time period can be determined from the thermal power output of the plant and its relation to the neutron fluence and from Figure 3.6.1, 3.6.2, or 3.6.3 in conjunction with Figure 3.6.4. Note: Figure 3.6.3 includes an additional 400F margin required by 10 CFR 50 Appendix G. Neutron flux wires and samples of vessel material are installed in the reactor vessel adjacent to the vessel wall at the core midplane level. The wires and samples will be removed and tested to experimentally verify the values used for Figure 3.6.4. As described in paragraph 4.2.5 of the Safety Analysis report,
-detailed stress analyses have been made on the reactor vessel for both steady state and transient conditions with respect to caterial fatigue. The results of these transients are compared to allowable stress limits.- Requiring the coolant temperature in an idle recirculation loop to be within 500F of the operating loop temperature before a recirculation pump is-started assures that the changes in coolant temperature at the reactor vessel nozzles and bottom head region are acceptable.
The plant safety analyses (Ref: NEDE-24011-P-A) states that all MSIV valve ~ closure - Flux scram is the event which satisfies the
'ASME Boiler and Pressure Code requirements for protectinn from the consequences of pressure in excess of the vessel design
- pressure. - The reactor vessel pressure code limit of 1375 psig, given in Subsection 4.2 of the FSAR,-is well above the peak pressure produced by the above overpressure event.
F
-152a-
i PBAPS Unit 3 3.6.D & 4.6.D BASES Safety and Relief Valves The safety / relief and safety valves are required to be operable above-the pressure (122 psig) at which the core spray system is not designed to deliver full flow. The pressure relief system for each unit at the Pench Bottom APS has been sized to meet two design bases. First, the total capacity of the safety / relief and the safety valves has been established to meet the overpressure protection criteria of the ASME code. Second, the distribution of this required capacity-between safety / relief valves and safety
- l. valves has been set to meet design basis 4.4.4.1 of subsection i 4.4 of the FSAR which states that the nuclear system safety / relief valves shall prevent opening of the safety valves during normal plant isolations.and load rejections.
The details of the analysis which shows compliance with the ASME code requirements is presented in subsection 4.4 of the FSAR and the Reactor Vessel Overpressure Protection Summary Technical j Report presented in Appendix K of the FSAR. L i Eleven safety / relief valves and two safety valves have been installed on Peach Bottom Unit 3 with a total capacity of 79.51% of rated steam flow. The analysis of the worst overpressure
' transient demonstrates margin to the code _ allowable overpressure limit of11375 psig.
To meet the-power generation / design basis, the total pressure
- relief _ system _ capacity of 79.51% has been divided-into 65.96%
safety / relief (11 valves) and 13.55% safety-(2 valves). The
- j. analysis of the plant isolation transient shows-that the 11
- safety / relief valves limit pressure at the~ safety valves below
.the setting of the safety valves. Therefore, the safety valves
- - will not open.
! -Experience in safety / relief and safety-valve. operation shows that-
- a. testing of 50 per cent of-the valves per year is adequate to detect failure:or deteriorations. The safety / relief and safety 4 valves are benchtested every second P
, -157-
PBAPS Unit 3 5.0 MAJOR DESIGN FEATURES 5.1 SITE FEATURES The site is located partly in Peach Bottom Township, York County, partly in Drumore Township, Lancaster County, and partly in Fulton Township, Lancaster County, in southeastern Pennsylvania
-on the westerly shore of Conowingo Pond at the mouth of Rock Run Creek. It is about 38 miles north-northeast of Baltimore, Maryland, and 63 miles west-southwest of Philadelphia, Pennsylvania. Figures 2.2.1 through 2.2.4 of the FSAR show the site location with respect to surrounding communities.
5.2 REACTOR A. The core shall consist of not more than 764 fuel assemblies. 7 x 7 fuel assemblies shall contain 49 fuel rods and 8 x 8 fuel assemblies shall contain 62 or 63 fuel rods. B. -One Pressurized Test Assembly may be inserted in the Core for up to four full fuel cycles. C. The reactor core shall contain 185 cruciform-shaped control rods. The control material shall be boron carbide powder (B.C) compacted to approximately 70% of the theoretical density. D. One Fast Scram Control Rod Drive may be utilized during operation. 5.3 REACTOR VESSEL The reactor vessel shall be as described in Table 4.2.2 of the FSAR. The applicable design codes shall be as described in Table 4.2.1 of the FSAR. 5.4 CONTAINMENT A. The principal design parameters for the primary containment shall be as given in Table 5.2.1 of the FSAR. The applicable design codes-shall be as described in Appendix M of the FSAR. B. The secondary containment shall be as described in Section 5.3 of the'FSAR.
'C. Penetrations to the primary' containment and piping passing through'such penetrations shall be designed in accordance with standards set forth in SectionH5.2.3.4 of the FSAR.
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-241-}}