ML16120A381
| ML16120A381 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 04/18/2016 |
| From: | Entergy Nuclear Operations |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16120A302 | List:
|
| References | |
| PNP 2016-015 | |
| Download: ML16120A381 (52) | |
Text
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.2Page 4.2-1 of 4.2-4 4.2DESIGN BASIS 4.2.1PERFORMANCE OBJECTIVES AND PARAMETERS FOR NORMALCONDITIONSThe Primary Coolant System is designed to operate at a power level of2,650 MWt. The present licensing limit is, however, 2,565.4 MWt core powerplus 15 MWt for the primary coolant pump heat input for a total PrimaryCoolant System output of 2,580.4 MWt. The principal parameters for thePrimary Coolant System are listed in Table 4-1. The design parameters foreach of the major components are given under the individual componentdiscussion later in this section. The Primary Coolant System is a CP CoDesign Class 1 system per Section 5.2. The applicable stress and seismiccriteria are given in Section 5.10. The primary system components andcontrols are also designed for cyclic transient conditions as listed inSubsection 4.2.2.
4.2.2DESIGN CYCLIC LOADSThe following design cyclic transients which include conservative estimates ofthe operational requirements for the components listed in Table 4-2 wereused in the fatigue analysis required by the applicable code:
1.500 heatup and cooldown cycles during the system 40-year design lifeat a heating and cooling rate of 100°F/h. The pressurizer is designedfor a cooldown rate of 200°F/h.
2.15,000 power change cycles over the range of 10% to 100% of fullload with a ramp load change of 5% of full load per minute increasingor decreasing. (The number of cycles for the normal power changehas been reduced to 2,000 due to CRDM nozzle repairs during 2004refueling outage.)
3.15,000 cycles of 10% of full load step power changes increasing from10% to 90% of full power and decreasing from 100% to 20% of fullpower. (The number of fast power changes and normal step powerchanges has been reduced to 2,000 due to CRDM nozzle repairsduring 2004 refueling outage.)
4.10 cycles of hydrostatic testing the primary system at 3,110 psig and ata temperature at least 60°F above the Nil Ductility TransitionTemperature (NDTT) of the component having the highest NDTT. Therelationship between the allowable temperature and maximum primaryto secondary pressure differential is shown in Figure 6-b of Reference 43.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.2Page 4.2-2 of 4.2-4 5.320 cycles of leak testing at 2,485 psig and at a temperature at least 60°F greater than the NDTT of the component having the highestNDTT. Steam generator level instrumentation will indicate at least lowwater level for the duration of the test and the primary to secondarydifferential pressure will not exceed 2,100 psia.
6.350,000 cycles of normal operating pressure variations of
+/- 50 psi atoperating temperature.7.500 reactor trips from 100% power. The 500 reactor trips from100% power are not explicitly addressed in the reactor vessel analytical report or in the associated specification. It is anticipated thatthe reactor trips were not addressed due to the fact that they are lesslimiting than other analyzed transients and would have a minimalimpact on the fatigue usage factor which is well below the limit of 1.0(Reference 13).In addition to the above list of normal design transients, the followingabnormal transients were also considered when arriving at a satisfactoryusage factor as defined in the ASME Boiler and Pressure Vessel Code.
1.200 cycles of loss of turbine load from 100% power 2.200 cycles of total loss of reactor coolant flow when at 100% power 4.2.3DESIGN SERVICE LIFE CONSIDERATIONSThe major Primary Coolant System components are designed considering a40-year service life. In order to achieve this, the strict quality controlassurance standards as outlined in Subsections 4.5.4 and 4.5.5 werefollowed.Component design has also considered environmental protection, adherenceto established operating procedures and irradiation effects on the material.The reactor vessel is the only component of the Primary Coolant Systemwhich is exposed to a significant level of neutron irradiation. The irradiationsurveillance program is outlined in Subsection 4.5.3. To compensate for anyincrease in the NDTT shift caused by irradiation, the Plant operatingprocedures for the pressure-temperature relationship during heatup andcooldown will be periodically revised to stay within the stress limits.The design of the Primary Coolant System components allows for adequateinspection techniques to be applied over the lifetime of the Plant. All reactorinternals are designed to be removable for inspection and to allow reactorvessel internal inspection. Insulation panels are removable for externalinspection of selected highly stressed areas.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.2Page 4.2-3 of 4.2-4 4.2.4CODES ADHERED TO AND COMPONENT CLASSIFICATIONThe original design, fabrication, construction, inspection, testing andclassification of all reactor coolant system components are in accordance withthe ASME Boiler and Pressure Vessel Code,Section III, 1965 edition,including all addenda through Winter 1965 (ASME B&PV Code,Section III,1965, W65a), and the Code for Pressure Piping, ASA B31.1, 1955(Reference 32). The replacement steam generators installed during 1990meet ASME Code Section III 1977 edition.The codes adhered to and component classifications are listed in Table 4-2.
Replacement parts and components will satisfy the requirements of theoriginal plant construction code in a manner that is consistent with10CFR50.55a, and the rules and requirements specified in ASME B&PVCode,Section XI, "Rules for Inservice Inspection of Nuclear Power PlantComponents", Article IWA-4 000." 4.2.5SAFETY CONSIDERATIONS OF DESIGN PARAMETERSDesign PressureAfter establishing the normal operating pressure conditions of the PrimaryCoolant System, a minimum design pressure was determined which exceedsthe normal operating pressure and anticipated operating transient pressure changes.Major considerations employed in the determination of this selected minimumdesign pressure include: normal operating pressure, instrumentation andcontrol response, reactor core thermal lag, coolant transport time, systempressure drop, and safety and relief valve characteristics. The designpressures for the individual reactor coolant system components are listed intheir respective component description sections.Design TemperatureThe design temperature was selected to exceed the normal operatingtemperature and anticipated operating transient temperature changes foreach primary coolant component. The design temperatures for the primarysystem components are listed in their respective component descriptionsections.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.2Page 4.2-4 of 4.2-4 Design LoadsThe Primary Coolant System was designed to the criteria for loadcombination and stresses as defined for a CP Co Design Class 1 system inSection 5.10. These criteria assure the integrity of the Primary CoolantSystem to withstand the load imposed by the design basis accidentsimultaneously with the load imposed by the maximum seismic disturbancewithout loss of safety function.
4.2.6PRIMARY COOLANT SYSTEM ASYMMETRIC LOADSPursuant to industry and NRC concerns for the potential effects ofasymmetric loads on the Primary Coolant System components and supports,Consumers Power (in 1978) contracted with Combustion Engineering for astudy to evaluate these concerns. A generic plant evaluation (seeReferences 33 and 34)(Section 14.17.3) was completed by CombustionEngineering for Calvert Cliffs 1 and 2, Palisades, Millstone 2 and Fort Calhoun.A further evaluation (see reference 35) was performed by CombustionEngineering to show that a flaw in the Primary Coolant System will result in adetectable leak before a large guillotine break would occur. The analysis wasreviewed by the NRC in an SER dated October 27, 1989 (Reference 36).The SER concluded that, with the exception of concerns regarding seismicgrid design, Palisades reactor system would withstand the effects ofasymmetric LOCA loads and that the reactor could be brought to a coldshutdown condition safely.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.4Page 4.4-1 of 4.4-4 4.4SYSTEM DESIGN EVALUATION 4.4.1DESIGN MARGINThe Primary Coolant System is structurally designed for operation at2,500 psia and 650°F (pressurizer 700°F). Operation of the system at2,100 psia nominal and 600°F will result in material stresses of 85% of designvalues. Detailed structural analyses have been performed by the componentvendors and reviewed independently by Combustion Engineering for allportions of the system. Welding materials used have physical propertiessuperior to the materials which they join. Inspection procedures and testsspecified and independently reviewed by Combustion Engineering werecarried out to assure that pressure-containing components have themaximum integrity obtainable with present code-approved inspectiontechniques. A detailed discussion of quality control inspections is found inChapter 15, Quality Assurance Program and Section 4.5.The Primary Coolant System is equipped with four primary coolant pumps. Inconjunction with protective instrumentation (see Chapter 7), this redundancyensures adequate core cooling. The primary coolant pumps are connected totwo 4,160-volt buses and receive power from either the main generator orfrom the offsite power system. Upon loss of all 4,160-volt power sources,short-term pumping power will be provided by the coastdown of the turbinegenerator. Under these conditions, primary flow during coastdown will beequal to or greater than that following the simultaneous loss of two pumps.The utilization of the rotating inertia of the turbine generator allows primarycoolant flow to be maintained at a much higher rate after loss of power thanwould be the case if only the pump flywheels were used to provide coastdownflow. The utilization of the turbine generator inertia allows the pump flywheelsto be designed for the two pump loss-of-flow accident. The most severesingle electrical failure analyzed, however, is the loss of offsite power and noturbine coastdown. The results are within core design limits and arepresented in Section 14.7. Following coastdown, natural circulation cooldown(see Reference 8) is established by releasing energy from the steamgenerators by steaming to atmosphere or the condenser. Makeup water isintroduced from the Auxiliary Feedwater System to the steam generators. ThePrimary Coolant System is maintained at a sufficient pressure during thisperiod (natural circulation cooldown) to avoid detrimental steam voidformation in the system; particularly in the reactor vessel upper head region.If steam voiding does occur, procedural guidance is provided to minimize andcontrol the voiding.The Plant is designed to operate at reduced power with one or two pumps outof service; however, the Plant Technical Specifications requires all fourpumps to be in service for continuous Plant operation.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.4Page 4.4-2 of 4.4-4 4.4.2PREVENTION OF BRITTLE FRACTUREBrittle fracture will not occur if the peak stresses do not exceed the yieldstress in the brittle fracture range. The establishment oftemperature-pressure limitations for operation below NDT temperature +60°Fis based on not exceeding yield for the peak stresses. This is accomplishedby the following:
1.Performing a complete and thorough stress analysis to establish stressdistribution taking into account all geometric shapes and surface stressconcentrations.
2.Establishing material properties suitable for the application byadequate specification and testing during all stages of procurementand fabrication. Periodic monitoring of the material is done during itsservice life to determine the shift in properties due to environment.
3.Postweld heat treatment to reduce the effect of residual stresseswhose magnitudes are unknown.
4.Establishing a safe limit (see Reference 9) of yield stress divided by astrength reduction factor in determining the operating pressure to allowfor defects which may be undetected by the available techniques ofnondestructive testing. It should be noted that the highest localstresses, other than those from such undetected defects, occur on thesurface of the material. All surfaces are examined by nondestructivemethods and defects and flaws so found which exceed allowable limitsare repaired.Additional areas of conservatism are the following:
1.Minimum specified values have been used for material propertiesrather than values from mill test reports of actual production material.
2.The increase in yield strength due to irradiation is not considered.The establishment of operating conditions based on the stress limitationsbelow will avoid exceeding the yield point stress in the brittle fracture rangeand hence avoid brittle fracture.Stress limitations are used to establish pressure-temperature operatingcurves for the Plant. The pressure-temperature operating curves considerPlant heatup and cooldown in both critical and noncritical reactor conditions.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.4Page 4.4-3 of 4.4-4The pressure-temperature limits are established as follows:
1.A predicted reference temperature (RTNDT) temperature shift for therequired fast neutron (E
³ 1 MeV) fluence is determined(see Reference 10). Where NDT temperature data are not available todetermine the RT NDT, Reference 11 is employed.
2.The ASME Code (see References 12 and 54) prescribes themethodologyused for obtaining the allowable loadings for any ferriticpressure-retaining materials in ASME Class 1 components. Thismethodologyis based upon the principles of linear elastic fracturemechanics and involves a reference stress intensity factor prediction which is a lower bound of static, dynamic and crack arrest criticalvalues. The reference stress intensity factor is a function of coolanttemperature as well as temperature gradients through the reactorvessel wall. The calculated reference stress intensity factor mustexceed that produced by pressure membrane stress in the vessel wallplus that produced by vessel wall thermal gradient stress. In theinequality associated with stress intensity comparison, a safety factorof 2 is applied to operating pressure membrane stress and a safetyfactor of 1.5 for hydrotest membrane stress.
3.The allowable pressure for a given operating temperature and heatupor cooldown rate is determined from the stress intensity equality (seeReference 11) with the appropriate safety factors included. A moredetailed discussion of limits may be found in the basis discussion ofthe Technical Specifications LCO 3.4.3 (Reference 23).
4.Minimum primary coolant temperature for criticality is given in thePalisades Technical Specifications. This is calculated from hydrotestpressure. In addition, a minimum of 40°F temperature margin (seeReference 14) is included for all pressure-temperature curves for thereactor in the critical condition with respect to those in the noncritical condition.The pressure-temperature limits are based upon the hypothetical (1/4)T reference flaw which is assumed to exist on either the inner or outer vesselwall surface (see Reference s 12 and 54). The calculations yield continuousstress limitation curves.During the 2004 Refueling Outage, inspections of the reactor head requiredby NRC Order EA-03-009 (Reference 52), resulted in the need to repair twoof the CRD nozzles. The repair of these two nozzles required a change to thepressure-temperature cooldown curves and a corresponding amendment tothe operating license. This change in the pressure-temperature limits isbased on not exceeding the fracture toughness limits of the J-groove remnantmaterials of the repaired nozzles.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.4Page 4.4-4 of 4.4-4Pressurized thermal shock concerns noted in Reference 21 have beenaddressed by the addition of a 200°F subcooling curve to thepressure-temperature curves in the emergency operating procedures. Thiscurve was developed per Reference 22 and has been adjusted for normalinstrument inaccuracies. This curve supercedes the maximum cooldowncurve whenever the PCS has experienced an uncontrolled cooldown(specifics of which are specified in the emergency operating procedures).
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.2Page 4.2-1 of 4.2-4 4.2DESIGN BASIS 4.2.1PERFORMANCE OBJECTIVES AND PARAMETERS FOR NORMALCONDITIONSThe Primary Coolant System is designed to operate at a power level of2,650 MWt. The present licensing limit is, however, 2,565.4 MWt core powerplus 15 MWt for the primary coolant pump heat input for a total PrimaryCoolant System output of 2,580.4 MWt. The principal parameters for thePrimary Coolant System are listed in Table 4-1. The design parameters foreach of the major components are given under the individual componentdiscussion later in this section. The Primary Coolant System is a CP CoDesign Class 1 system per Section 5.2. The applicable stress and seismiccriteria are given in Section 5.10. The primary system components andcontrols are also designed for cyclic transient conditions as listed inSubsection 4.2.2.
4.2.2DESIGN CYCLIC LOADSThe following design cyclic transients which include conservative estimates ofthe operational requirements for the components listed in Table 4-2 wereused in the fatigue analysis required by the applicable code:
1.500 heatup and cooldown cycles during the system 40-year design lifeat a heating and cooling rate of 100°F/h. The pressurizer is designedfor a cooldown rate of 200°F/h.
2.15,000 power change cycles over the range of 10% to 100% of fullload with a ramp load change of 5% of full load per minute increasingor decreasing. (The number of cycles for the normal power changehas been reduced to 2,000 due to CRDM nozzle repairs during 2004refueling outage.)
3.15,000 cycles of 10% of full load step power changes increasing from10% to 90% of full power and decreasing from 100% to 20% of fullpower. (The number of fast power changes and normal step powerchanges has been reduced to 2,000 due to CRDM nozzle repairsduring 2004 refueling outage.)
4.10 cycles of hydrostatic testing the primary system at 3,110 psig and ata temperature at least 60°F above the Nil Ductility TransitionTemperature (NDTT) of the component having the highest NDTT. Therelationship between the allowable temperature and maximum primaryto secondary pressure differential is shown in Figure 6-b of Reference 43.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.2Page 4.2-2 of 4.2-4 5.320 cycles of leak testing at 2,485 psig and at a temperature at least 60°F greater than the NDTT of the component having the highestNDTT. Steam generator level instrumentation will indicate at least lowwater level for the duration of the test and the primary to secondarydifferential pressure will not exceed 2,100 psia.
6.350,000 cycles of normal operating pressure variations of
+/- 50 psi atoperating temperature.7.500 reactor trips from 100% power. The 500 reactor trips from100% power are not explicitly addressed in the reactor vessel analytical report or in the associated specification. It is anticipated thatthe reactor trips were not addressed due to the fact that they are lesslimiting than other analyzed transients and would have a minimalimpact on the fatigue usage factor which is well below the limit of 1.0(Reference 13).In addition to the above list of normal design transients, the followingabnormal transients were also considered when arriving at a satisfactoryusage factor as defined in the ASME Boiler and Pressure Vessel Code.
1.200 cycles of loss of turbine load from 100% power 2.200 cycles of total loss of reactor coolant flow when at 100% power 4.2.3DESIGN SERVICE LIFE CONSIDERATIONSThe major Primary Coolant System components are designed considering a40-year service life. In order to achieve this, the strict quality controlassurance standards as outlined in Subsections 4.5.4 and 4.5.5 werefollowed.Component design has also considered environmental protection, adherenceto established operating procedures and irradiation effects on the material.The reactor vessel is the only component of the Primary Coolant Systemwhich is exposed to a significant level of neutron irradiation. The irradiationsurveillance program is outlined in Subsection 4.5.3. To compensate for anyincrease in the NDTT shift caused by irradiation, the Plant operatingprocedures for the pressure-temperature relationship during heatup andcooldown will be periodically revised to stay within the stress limits.The design of the Primary Coolant System components allows for adequateinspection techniques to be applied over the lifetime of the Plant. All reactorinternals are designed to be removable for inspection and to allow reactorvessel internal inspection. Insulation panels are removable for externalinspection of selected highly stressed areas.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.2Page 4.2-3 of 4.2-4 4.2.4CODES ADHERED TO AND COMPONENT CLASSIFICATIONThe original design, fabrication, construction, inspection, testing andclassification of all reactor coolant system components are in accordance withthe ASME Boiler and Pressure Vessel Code,Section III, 1965 edition,including all addenda through Winter 1965 (ASME B&PV Code,Section III,1965, W65a), and the Code for Pressure Piping, ASA B31.1, 1955(Reference 32). The replacement steam generators installed during 1990meet ASME Code Section III 1977 edition.The codes adhered to and component classifications are listed in Table 4-2.
Replacement parts and components will satisfy the requirements of theoriginal plant construction code in a manner that is consistent with10CFR50.55a, and the rules and requirements specified in ASME B&PVCode,Section XI, "Rules for Inservice Inspection of Nuclear Power PlantComponents", Article IWA-4 000." 4.2.5SAFETY CONSIDERATIONS OF DESIGN PARAMETERSDesign PressureAfter establishing the normal operating pressure conditions of the PrimaryCoolant System, a minimum design pressure was determined which exceedsthe normal operating pressure and anticipated operating transient pressure changes.Major considerations employed in the determination of this selected minimumdesign pressure include: normal operating pressure, instrumentation andcontrol response, reactor core thermal lag, coolant transport time, systempressure drop, and safety and relief valve characteristics. The designpressures for the individual reactor coolant system components are listed intheir respective component description sections.Design TemperatureThe design temperature was selected to exceed the normal operatingtemperature and anticipated operating transient temperature changes foreach primary coolant component. The design temperatures for the primarysystem components are listed in their respective component descriptionsections.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.2Page 4.2-4 of 4.2-4 Design LoadsThe Primary Coolant System was designed to the criteria for loadcombination and stresses as defined for a CP Co Design Class 1 system inSection 5.10. These criteria assure the integrity of the Primary CoolantSystem to withstand the load imposed by the design basis accidentsimultaneously with the load imposed by the maximum seismic disturbancewithout loss of safety function.
4.2.6PRIMARY COOLANT SYSTEM ASYMMETRIC LOADSPursuant to industry and NRC concerns for the potential effects ofasymmetric loads on the Primary Coolant System components and supports,Consumers Power (in 1978) contracted with Combustion Engineering for astudy to evaluate these concerns. A generic plant evaluation (seeReferences 33 and 34)(Section 14.17.3) was completed by CombustionEngineering for Calvert Cliffs 1 and 2, Palisades, Millstone 2 and Fort Calhoun.A further evaluation (see reference 35) was performed by CombustionEngineering to show that a flaw in the Primary Coolant System will result in adetectable leak before a large guillotine break would occur. The analysis wasreviewed by the NRC in an SER dated October 27, 1989 (Reference 36).The SER concluded that, with the exception of concerns regarding seismicgrid design, Palisades reactor system would withstand the effects ofasymmetric LOCA loads and that the reactor could be brought to a coldshutdown condition safely.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.4Page 4.4-1 of 4.4-4 4.4SYSTEM DESIGN EVALUATION 4.4.1DESIGN MARGINThe Primary Coolant System is structurally designed for operation at2,500 psia and 650°F (pressurizer 700°F). Operation of the system at2,100 psia nominal and 600°F will result in material stresses of 85% of designvalues. Detailed structural analyses have been performed by the componentvendors and reviewed independently by Combustion Engineering for allportions of the system. Welding materials used have physical propertiessuperior to the materials which they join. Inspection procedures and testsspecified and independently reviewed by Combustion Engineering werecarried out to assure that pressure-containing components have themaximum integrity obtainable with present code-approved inspectiontechniques. A detailed discussion of quality control inspections is found inChapter 15, Quality Assurance Program and Section 4.5.The Primary Coolant System is equipped with four primary coolant pumps. Inconjunction with protective instrumentation (see Chapter 7), this redundancyensures adequate core cooling. The primary coolant pumps are connected totwo 4,160-volt buses and receive power from either the main generator orfrom the offsite power system. Upon loss of all 4,160-volt power sources,short-term pumping power will be provided by the coastdown of the turbinegenerator. Under these conditions, primary flow during coastdown will beequal to or greater than that following the simultaneous loss of two pumps.The utilization of the rotating inertia of the turbine generator allows primarycoolant flow to be maintained at a much higher rate after loss of power thanwould be the case if only the pump flywheels were used to provide coastdownflow. The utilization of the turbine generator inertia allows the pump flywheelsto be designed for the two pump loss-of-flow accident. The most severesingle electrical failure analyzed, however, is the loss of offsite power and noturbine coastdown. The results are within core design limits and arepresented in Section 14.7. Following coastdown, natural circulation cooldown(see Reference 8) is established by releasing energy from the steamgenerators by steaming to atmosphere or the condenser. Makeup water isintroduced from the Auxiliary Feedwater System to the steam generators. ThePrimary Coolant System is maintained at a sufficient pressure during thisperiod (natural circulation cooldown) to avoid detrimental steam voidformation in the system; particularly in the reactor vessel upper head region.If steam voiding does occur, procedural guidance is provided to minimize andcontrol the voiding.The Plant is designed to operate at reduced power with one or two pumps outof service; however, the Plant Technical Specifications requires all fourpumps to be in service for continuous Plant operation.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.4Page 4.4-2 of 4.4-4 4.4.2PREVENTION OF BRITTLE FRACTUREBrittle fracture will not occur if the peak stresses do not exceed the yieldstress in the brittle fracture range. The establishment oftemperature-pressure limitations for operation below NDT temperature +60°Fis based on not exceeding yield for the peak stresses. This is accomplishedby the following:
1.Performing a complete and thorough stress analysis to establish stressdistribution taking into account all geometric shapes and surface stressconcentrations.
2.Establishing material properties suitable for the application byadequate specification and testing during all stages of procurementand fabrication. Periodic monitoring of the material is done during itsservice life to determine the shift in properties due to environment.
3.Postweld heat treatment to reduce the effect of residual stresseswhose magnitudes are unknown.
4.Establishing a safe limit (see Reference 9) of yield stress divided by astrength reduction factor in determining the operating pressure to allowfor defects which may be undetected by the available techniques ofnondestructive testing. It should be noted that the highest localstresses, other than those from such undetected defects, occur on thesurface of the material. All surfaces are examined by nondestructivemethods and defects and flaws so found which exceed allowable limitsare repaired.Additional areas of conservatism are the following:
1.Minimum specified values have been used for material propertiesrather than values from mill test reports of actual production material.
2.The increase in yield strength due to irradiation is not considered.The establishment of operating conditions based on the stress limitationsbelow will avoid exceeding the yield point stress in the brittle fracture rangeand hence avoid brittle fracture.Stress limitations are used to establish pressure-temperature operatingcurves for the Plant. The pressure-temperature operating curves considerPlant heatup and cooldown in both critical and noncritical reactor conditions.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.4Page 4.4-3 of 4.4-4The pressure-temperature limits are established as follows:
1.A predicted reference temperature (RTNDT) temperature shift for therequired fast neutron (E
³ 1 MeV) fluence is determined(see Reference 10). Where NDT temperature data are not available todetermine the RT NDT, Reference 11 is employed.
2.The ASME Code (see References 12 and 54) prescribes themethodologyused for obtaining the allowable loadings for any ferriticpressure-retaining materials in ASME Class 1 components. Thismethodologyis based upon the principles of linear elastic fracturemechanics and involves a reference stress intensity factor prediction which is a lower bound of static, dynamic and crack arrest criticalvalues. The reference stress intensity factor is a function of coolanttemperature as well as temperature gradients through the reactorvessel wall. The calculated reference stress intensity factor mustexceed that produced by pressure membrane stress in the vessel wallplus that produced by vessel wall thermal gradient stress. In theinequality associated with stress intensity comparison, a safety factorof 2 is applied to operating pressure membrane stress and a safetyfactor of 1.5 for hydrotest membrane stress.
3.The allowable pressure for a given operating temperature and heatupor cooldown rate is determined from the stress intensity equality (seeReference 11) with the appropriate safety factors included. A moredetailed discussion of limits may be found in the basis discussion ofthe Technical Specifications LCO 3.4.3 (Reference 23).
4.Minimum primary coolant temperature for criticality is given in thePalisades Technical Specifications. This is calculated from hydrotestpressure. In addition, a minimum of 40°F temperature margin (seeReference 14) is included for all pressure-temperature curves for thereactor in the critical condition with respect to those in the noncritical condition.The pressure-temperature limits are based upon the hypothetical (1/4)T reference flaw which is assumed to exist on either the inner or outer vesselwall surface (see Reference s 12 and 54). The calculations yield continuousstress limitation curves.During the 2004 Refueling Outage, inspections of the reactor head requiredby NRC Order EA-03-009 (Reference 52), resulted in the need to repair twoof the CRD nozzles. The repair of these two nozzles required a change to thepressure-temperature cooldown curves and a corresponding amendment tothe operating license. This change in the pressure-temperature limits isbased on not exceeding the fracture toughness limits of the J-groove remnantmaterials of the repaired nozzles.
FSAR CHAPTER 4 - PRIMARY COOLANT SYSTEMRevision 30SECTION 4.4Page 4.4-4 of 4.4-4Pressurized thermal shock concerns noted in Reference 21 have beenaddressed by the addition of a 200°F subcooling curve to thepressure-temperature curves in the emergency operating procedures. Thiscurve was developed per Reference 22 and has been adjusted for normalinstrument inaccuracies. This curve supercedes the maximum cooldowncurve whenever the PCS has experienced an uncontrolled cooldown(specifics of which are specified in the emergency operating procedures).