ML18066A341: Difference between revisions
StriderTol (talk | contribs) (Created page by program invented by StriderTol) |
StriderTol (talk | contribs) (Created page by program invented by StriderTol) |
||
| Line 32: | Line 32: | ||
.................................................................................................... | .................................................................................................... | ||
1-1 1.1 Cycle14Changes | 1-1 1.1 Cycle14Changes | ||
........................................ | ........................................ | ||
: ........................................ | : ........................................ | ||
1-1 1 . 1 . 1 Plant Operating Conditions | 1-1 1 . 1 . 1 Plant Operating Conditions | ||
| Line 42: | Line 42: | ||
1-3 1 . 1 . 5 Fuel Rod Mechanical Design ....................................................... | 1-3 1 . 1 . 5 Fuel Rod Mechanical Design ....................................................... | ||
1 -3 1.1.6 Neutronics Design Changes ........................................................ | 1 -3 1.1.6 Neutronics Design Changes ........................................................ | ||
1-3 1 . 1 . 7 Uncertainties and Response Times ................................ | 1-3 1 . 1 . 7 Uncertainties and Response Times ................................ | ||
: ............. | : ............. | ||
1-3 1 . 2 Analytical Considerations | 1-3 1 . 2 Analytical Considerations | ||
| Line 56: | Line 56: | ||
2-2 2. 7 TM/LP and T;niet LCO Verification | 2-2 2. 7 TM/LP and T;niet LCO Verification | ||
............................................................. | ............................................................. | ||
2-2 Disposition And Analysis Of Plant Events ..................................... | 2-2 Disposition And Analysis Of Plant Events ..................................... | ||
: .................... | : .................... | ||
3-1 3.1 Thermal Hydraulic Compatibility | 3-1 3.1 Thermal Hydraulic Compatibility | ||
| Line 107: | Line 107: | ||
3-24 1 5. 3 .4 Reactor Coolant Pump Shaft Break ........................................... | 3-24 1 5. 3 .4 Reactor Coolant Pump Shaft Break ........................................... | ||
3-25 15.4 Reactivity and Power Distribution Anomalies | 3-25 15.4 Reactivity and Power Distribution Anomalies | ||
.............................. | .............................. | ||
: .......... | : .......... | ||
3-25 1 5 .4. 1 Uncontrolled Control Assembly Withdrawal from a Subcritical or Low Power Startup Condition | 3-25 1 5 .4. 1 Uncontrolled Control Assembly Withdrawal from a Subcritical or Low Power Startup Condition | ||
| Line 199: | Line 199: | ||
The minimum, maximum and nominal flow rates and the startup delay times for the. AFWs were added. The SIT water volume was increased from 2,000 ft 3 to 2,011 ft 2* Siemens Power Corporation | The minimum, maximum and nominal flow rates and the startup delay times for the. AFWs were added. The SIT water volume was increased from 2,000 ft 3 to 2,011 ft 2* Siemens Power Corporation | ||
-Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 1 .1 .4 Thermal-Hydraulic Characteristics EMF-98-013 Revision 0 Page 1-3 1 . The lower tie plate de-sign changed for Reload R. These tie plates have a FUELGUARDTM debris-resistant insert. 2. Addition of an all inconel High Thermal Performance (HTPJ bottom spacer. 3. New pressure drop measurements were performed with the control blades fully inserted. 1 . 1 . 5 Fuel Rod Mechanical Design A new fuel design will be introduced for Reload R, incorporating the following changes: a. 0.009-inch increase in the pellet OD. b. 0.152-inch increase in the pellet length. c. 0.4% reduction in the pellet dish volume. d. 0.009-inch increase in the cladding ID. e. 0.8-inch increase in the fuel rod active length. f. 0. 111-inch increase in the effective plenum length. g. Initial rod internal pressure of the gadolinia bearing rods has increased. | -Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 1 .1 .4 Thermal-Hydraulic Characteristics EMF-98-013 Revision 0 Page 1-3 1 . The lower tie plate de-sign changed for Reload R. These tie plates have a FUELGUARDTM debris-resistant insert. 2. Addition of an all inconel High Thermal Performance (HTPJ bottom spacer. 3. New pressure drop measurements were performed with the control blades fully inserted. 1 . 1 . 5 Fuel Rod Mechanical Design A new fuel design will be introduced for Reload R, incorporating the following changes: a. 0.009-inch increase in the pellet OD. b. 0.152-inch increase in the pellet length. c. 0.4% reduction in the pellet dish volume. d. 0.009-inch increase in the cladding ID. e. 0.8-inch increase in the fuel rod active length. f. 0. 111-inch increase in the effective plenum length. g. Initial rod internal pressure of the gadolinia bearing rods has increased. | ||
1 .1 .6 Neutronics Design Changes 1 . The enrichment pattern for the Reload R assemblies was modified and it produces different local power distributions. | 1 .1 .6 Neutronics Design Changes 1 . The enrichment pattern for the Reload R assemblies was modified and it produces different local power distributions. | ||
: 2. The U0 2 enrichment in the gadolinia bearing fuel rods increased. | : 2. The U0 2 enrichment in the gadolinia bearing fuel rods increased. | ||
: 3. The maximum gadolinia concentration in the gadolinia-bearing fuel rods increased in Cycle 14 from 6 wt% to 8 wt%. 4. The U 238 capture-fission ratios for Cycle 14 increased slightly, resulting in an increase in the decay heat. 5. The 50% and 100% power differential rod worths at EOC conditions increased. | : 3. The maximum gadolinia concentration in the gadolinia-bearing fuel rods increased in Cycle 14 from 6 wt% to 8 wt%. 4. The U 238 capture-fission ratios for Cycle 14 increased slightly, resulting in an increase in the decay heat. 5. The 50% and 100% power differential rod worths at EOC conditions increased. | ||
: 6. The moderator density and Doppler reactivity tables changed for Cycle 14. 7. The boron endpoint concentrations for Cyc;=le 14 changed for all conditions. | : 6. The moderator density and Doppler reactivity tables changed for Cycle 14. 7. The boron endpoint concentrations for Cyc;=le 14 changed for all conditions. | ||
1 .1 . 7 Uncertainties and Response Times 1 . Cycle 14 does not contain a full core of fresh detectors. | 1 .1 . 7 Uncertainties and Response Times 1 . Cycle 14 does not contain a full core of fresh detectors. | ||
: 2. Delay times for the HPSI and LPSI pumps were increased from 27 sec to 40 sec. 3. RTD delay times were increased from 12 to 15 sec. Siemens Power Corporation | : 2. Delay times for the HPSI and LPSI pumps were increased from 27 sec to 40 sec. 3. RTD delay times were increased from 12 to 15 sec. Siemens Power Corporation | ||
-Nuclear Division | -Nuclear Division | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 1-4 4. The trip delays for the low reactor coolant flow, high pressurizer pressure , low pressurizer pressure, low steam generator pressure and TM/LP and VHP trips were increased by 0.2 sec. 1 . 2 Analytical Considerations The Technical Specification radial peaking factor limits for Reload R, an of 1 . 76 and an F; of 2.04, were unchanged from those for Reloads 0, P and Q. Radial peaking factors for reloads prior to Reload 0 were unchanged for Cycle 14 (References 3 and 4). _The Technical Specification radial peaking limits for all fuel loaded in Cycle 14 are given in Table 15.0.3.1. | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 1-4 4. The trip delays for the low reactor coolant flow, high pressurizer pressure , low pressurizer pressure, low steam generator pressure and TM/LP and VHP trips were increased by 0.2 sec. 1 . 2 Analytical Considerations The Technical Specification radial peaking factor limits for Reload R, an of 1 . 76 and an F; of 2.04, were unchanged from those for Reloads 0, P and Q. Radial peaking factors for reloads prior to Reload 0 were unchanged for Cycle 14 (References 3 and 4). _The Technical Specification radial peaking limits for all fuel loaded in Cycle 14 are given in Table 15.0.3.1. | ||
The minimum departure from nucleate boiling ratio (MDNBR) for each event made use of the HTP departure from nucleate boiling correlation (Reference 5). A 2% mixed core DNB penalty {Reference | The minimum departure from nucleate boiling ratio (MDNBR) for each event made use of the HTP departure from nucleate boiling correlation (Reference 5). A 2% mixed core DNB penalty {Reference | ||
: 22) was applied in Cycle 14. The impact of the changes in the power distributions, both radial and axial, resulting from the modified enrichment distribution and the Cycle 14 loading plan on MDNBRs was considered. . The design of the gadolinia rods for Cycle 14 utilized five axial zones: two axial enriched blankets (2.15 wt% U 235), two axial regions with no gadolinia (4.02 wt% U 235) and a central gadolinia-bearing zone with 3.99 wt% U 235 and a maximum gadolinia concentration of 8 wt%. The U0 2 fuel rod design for Cycle 14 also made use of enriched axial blankets at the top and bottom. The active fuel length of the fuel rod increased and the fuel pellet diameter also increased for Cycle 14. The effects of these designs characteristics were considered in the generation of the neutronics parameters used in the Cycle 14 event review. The fuel centerline melt {FCM) limit was re-calculated for Cycle 14 because of the higher U 235 enrichments in the gadolinia-bearing rods and the mechanical design changes . e">! _____ ----""-------=-- | : 22) was applied in Cycle 14. The impact of the changes in the power distributions, both radial and axial, resulting from the modified enrichment distribution and the Cycle 14 loading plan on MDNBRs was considered. . The design of the gadolinia rods for Cycle 14 utilized five axial zones: two axial enriched blankets (2.15 wt% U 235), two axial regions with no gadolinia (4.02 wt% U 235) and a central gadolinia-bearing zone with 3.99 wt% U 235 and a maximum gadolinia concentration of 8 wt%. The U0 2 fuel rod design for Cycle 14 also made use of enriched axial blankets at the top and bottom. The active fuel length of the fuel rod increased and the fuel pellet diameter also increased for Cycle 14. The effects of these designs characteristics were considered in the generation of the neutronics parameters used in the Cycle 14 event review. The fuel centerline melt {FCM) limit was re-calculated for Cycle 14 because of the higher U 235 enrichments in the gadolinia-bearing rods and the mechanical design changes . e">! _____ ----""-------=-- | ||
****-*--- | ****-*--- | ||
| Line 215: | Line 215: | ||
The LHGR limit on U0 2 rods for Cycle 14 was determined to be 20.93 kW/ft. This limit precludes FCM on both U0 2 and gadolinia rods. 2.2 Steamline Break Analysis The Reference 3 analysis of the Main Steamline Break (MSLB) assumed that the break in the main steamline occurred inside containment. | The LHGR limit on U0 2 rods for Cycle 14 was determined to be 20.93 kW/ft. This limit precludes FCM on both U0 2 and gadolinia rods. 2.2 Steamline Break Analysis The Reference 3 analysis of the Main Steamline Break (MSLB) assumed that the break in the main steamline occurred inside containment. | ||
The case analyzed was a HFP case with a loss of offsite power and a subsequent turbine generator assisted coastdown of the primary coolant pumps (PCPs) which was identified based on a different methodology. | The case analyzed was a HFP case with a loss of offsite power and a subsequent turbine generator assisted coastdown of the primary coolant pumps (PCPs) which was identified based on a different methodology. | ||
The reanalysis of the MSLB was performed for the following reasons: 1. To extend the analysis from a single, limiting case to the standard set of cases typically evaluated for an MSLB event. 2. To evaluate breaks both inside and outside containment. | The reanalysis of the MSLB was performed for the following reasons: 1. To extend the analysis from a single, limiting case to the standard set of cases typically evaluated for an MSLB event. 2. To evaluate breaks both inside and outside containment. | ||
: 3. To evaluate the impact of tripping the PCPs at break initiation for the loss of offsite power case, instead of modeling a time delay to simulate the turbine generator assisted PCP coastdown. | : 3. To evaluate the impact of tripping the PCPs at break initiation for the loss of offsite power case, instead of modeling a time delay to simulate the turbine generator assisted PCP coastdown. | ||
: 4. To evaluate for the HZP cases, the impact of the initial valve position (open or closed) of the MSIVs and of the MSIV bypass valves. 5. To upgrade the ANF-RELAP model to be consistent with the Cycle 14 plant configuration and conditions. | : 4. To evaluate for the HZP cases, the impact of the initial valve position (open or closed) of the MSIVs and of the MSIV bypass valves. 5. To upgrade the ANF-RELAP model to be consistent with the Cycle 14 plant configuration and conditions. | ||
The results of this analysis are presented in Reference | The results of this analysis are presented in Reference | ||
: 23. 2. 3 Control Rod Ejection Analysis The control rod ejection analysis was performed for Cycle 14 to remove conservatism from the system analysis. | : 23. 2. 3 Control Rod Ejection Analysis The control rod ejection analysis was performed for Cycle 14 to remove conservatism from the system analysis. | ||
The ANF-RELAP model for this analysis was made consistent with the Cycle 14 plant configurations and operating conditions. | The ANF-RELAP model for this analysis was made consistent with the Cycle 14 plant configurations and operating conditions. | ||
| Line 224: | Line 224: | ||
-Nuclear Division | -Nuclear Division | ||
* * | * * | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 2-2 The system transient analysis and fuel failure calculations were performed assuming no loss of off site power. The results are presented in Reference | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 2-2 The system transient analysis and fuel failure calculations were performed assuming no loss of off site power. The results are presented in Reference | ||
: 24. 2.4 LOCA Analysis Based on all of the Cycle 14 changes which affect LBLOCA, the LBLOCA event was reanalyzed 1211* These changes include Items 4 and 5 in Section 1.1.1, Items 2, 3, 5, 8, and 11 in SectiOn 1.1.3, all items in Section 1.1.4 and 1.1.5, Items 1, 2, 3, 4, and 6 in Section 1. 1 . 6 and Item 2 of Section 1 . 1 . 7. The LBLOCA analysis also addressed the increased gad rod enrichment. | : 24. 2.4 LOCA Analysis Based on all of the Cycle 14 changes which affect LBLOCA, the LBLOCA event was reanalyzed 1211* These changes include Items 4 and 5 in Section 1.1.1, Items 2, 3, 5, 8, and 11 in SectiOn 1.1.3, all items in Section 1.1.4 and 1.1.5, Items 1, 2, 3, 4, and 6 in Section 1. 1 . 6 and Item 2 of Section 1 . 1 . 7. The LBLOCA analysis also addressed the increased gad rod enrichment. | ||
| Line 230: | Line 230: | ||
Review A summary of the event review for Palisades Cycle 14 is given in Table 2. 1. This table lists each SRP Chapter 15 event, indicates whether that event is re-analyzed for Cycle 14, provides a reference to the bounding event or analysis of record for events not analyzed, and provides a cross reference between the SRP Chapter 15 event numbers and the Palisades Updated FSAR (Reference 9). 2. 6 MDNBR, Fuel Centerline Melt and System Pressure Results The LBLOCA, MSLB, and Control Rod Ejection events were analyzed for Cycle 14 with the results reported in References 21, 23, and 24 respectively. | Review A summary of the event review for Palisades Cycle 14 is given in Table 2. 1. This table lists each SRP Chapter 15 event, indicates whether that event is re-analyzed for Cycle 14, provides a reference to the bounding event or analysis of record for events not analyzed, and provides a cross reference between the SRP Chapter 15 event numbers and the Palisades Updated FSAR (Reference 9). 2. 6 MDNBR, Fuel Centerline Melt and System Pressure Results The LBLOCA, MSLB, and Control Rod Ejection events were analyzed for Cycle 14 with the results reported in References 21, 23, and 24 respectively. | ||
The dispositions and analyses reported herein confirm, for the remaining SRP Chapter 15 events, that event acceptance criteria (Section 15.0.10) are met for the Cycle 14 operation and support operation with up to 1 5 % steam generator tube plugging in either or both steam generators at a rated thermal power of 2,530 MWt. A summary of the MDNBR, FCM and system pressure results, along with the acceptance criteria, is given in Table 2.2. The thermal margins for Cycle 14 are based on the full core two-pass Cycle 14 XCOBRA-lltC model and the HTP DNBR correlation with a 2% mixed penalty applied. 2. 7 TM/LP and T.., LCO Verification A setpoint analysis was performed which verified that the TM/LP and T;ntet LCO setpoints are acceptable for Cycle 14 . Siemens Power Corporation | The dispositions and analyses reported herein confirm, for the remaining SRP Chapter 15 events, that event acceptance criteria (Section 15.0.10) are met for the Cycle 14 operation and support operation with up to 1 5 % steam generator tube plugging in either or both steam generators at a rated thermal power of 2,530 MWt. A summary of the MDNBR, FCM and system pressure results, along with the acceptance criteria, is given in Table 2.2. The thermal margins for Cycle 14 are based on the full core two-pass Cycle 14 XCOBRA-lltC model and the HTP DNBR correlation with a 2% mixed penalty applied. 2. 7 TM/LP and T.., LCO Verification A setpoint analysis was performed which verified that the TM/LP and T;ntet LCO setpoints are acceptable for Cycle 14 . Siemens Power Corporation | ||
* Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events :? 1'.{. .:{. :::; ..... :{. EMF-98-013 Revision 0 Page 2-3 Table 2.1 Disposition of Events Summary for Palisades Cycle 14 SRP Bounding Updated Event Event or FSAR Designation Event Name DisEosition Reference Designation 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 Decrease in Feedwater Temperature Bounded 15.1.3 a 15.1.2 Increase in Feedwater Flow a 1) Power Operation Bounded 15.1.3 2) Startup Bounded 15.1.3 15.1.3 Increase in Steam Flow Bounded Ref. 10 14.10 Analyzed MDNBR 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve 1) Power Operation Bounded 15.1.3 2) Startup Bounded 15.4.1 15.1.5 Steam System Piping Failures Inside Analyzed Ref. 23 14.14 and Outside of Containment 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.1 Loss of External Load 1) Primary Over-pressurization Bounded Ref. 11 14.12 2) Secondary Over-pressurization Analyzed 14.12 3) MDNBR Analyzed 14.12 15.2.2 Turbine Trip Bounded 15.2.1 15.2.3 Loss of Condenser Vacuum Bounded 15.2.1 15.2.4 Closure of the Main Steam Isolation Bounded 15.2.1 Valves (MSIV) 15.2.5 Steam Pressure Regulator Failure Not Applicable; BWR Event 15.2.6 Loss of Non-Emergency A.C. Power to the Station Auxiliaries | * Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events :? 1'.{. .:{. :::; ..... :{. EMF-98-013 Revision 0 Page 2-3 Table 2.1 Disposition of Events Summary for Palisades Cycle 14 SRP Bounding Updated Event Event or FSAR Designation Event Name DisEosition Reference Designation 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 Decrease in Feedwater Temperature Bounded 15.1.3 a 15.1.2 Increase in Feedwater Flow a 1) Power Operation Bounded 15.1.3 2) Startup Bounded 15.1.3 15.1.3 Increase in Steam Flow Bounded Ref. 10 14.10 Analyzed MDNBR 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve 1) Power Operation Bounded 15.1.3 2) Startup Bounded 15.4.1 15.1.5 Steam System Piping Failures Inside Analyzed Ref. 23 14.14 and Outside of Containment 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.1 Loss of External Load 1) Primary Over-pressurization Bounded Ref. 11 14.12 2) Secondary Over-pressurization Analyzed 14.12 3) MDNBR Analyzed 14.12 15.2.2 Turbine Trip Bounded 15.2.1 15.2.3 Loss of Condenser Vacuum Bounded 15.2.1 15.2.4 Closure of the Main Steam Isolation Bounded 15.2.1 Valves (MSIV) 15.2.5 Steam Pressure Regulator Failure Not Applicable; BWR Event 15.2.6 Loss of Non-Emergency A.C. Power to the Station Auxiliaries | ||
: 1) DNB Bounded 15.3.1 2) Long Term cooling Bounded 15.2.7 15.2.7 Loss of Normal Feedwater Flow 1) Maximum PCS pressure Bounded 15.2.1 14.13 2) Maximum Primary to secondary Boun.cled Ref. 12 14.13 pressure difference | : 1) DNB Bounded 15.3.1 2) Long Term cooling Bounded 15.2.7 15.2.7 Loss of Normal Feedwater Flow 1) Maximum PCS pressure Bounded 15.2.1 14.13 2) Maximum Primary to secondary Boun.cled Ref. 12 14.13 pressure difference | ||
: 3) Minimum steam generator inventory Bounded Ref. 12 Deleted from the FSAR Siemens Power Corporation | : 3) Minimum steam generator inventory Bounded Ref. 12 Deleted from the FSAR Siemens Power Corporation | ||
-Nuclear Division | -Nuclear Division | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 2-4 Table 2.1 Disposition of Events Summary for Palisades Cycle 14 (continued) | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 2-4 Table 2.1 Disposition of Events Summary for Palisades Cycle 14 (continued) | ||
SRP Bounding Updated Event Event or FSAR Designation Event Name Disposition Reference Designation 15.2.8 Feedwater Syi>tem Pipe Breaks Inside and Outside Containment. | SRP Bounding Updated Event Event or FSAR Designation Event Name Disposition Reference Designation 15.2.8 Feedwater Syi>tem Pipe Breaks Inside and Outside Containment. | ||
: 1) Cooldown Bounded 15.1.5 2) Heatup Bounded 15.2.1 3) Long Term Cooling Bounded 15.2.7 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW 15.3.1 Loss of Forced Reactor Bounded Ref. 3 14.7 Coolant Flow Analyzed MDNBR 15.3.2 Flow Controller Malfunction Not Applicable 14.7 15.3.3 Reactor Coolant Pump Rotor Bounded Ref. 3 14.7 Seizure Analyzed MDNBR 15.3.4 Reactor Coolant Pump Shaft Bounded 15.3.3 14.7 Break 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled Control Bank Calculated MDNBR and maximum 14.2.1 Withdrawal from a fuel temperature Subcritical or Low Power Condition 15.4.2 Uncontrolled Control Bank Calculated MDNBR and maximum 14.2.2 Withdrawal at Power LHGR Operation Conditions 15.4.3 Control Rod Misoperation | : 1) Cooldown Bounded 15.1.5 2) Heatup Bounded 15.2.1 3) Long Term Cooling Bounded 15.2.7 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW 15.3.1 Loss of Forced Reactor Bounded Ref. 3 14.7 Coolant Flow Analyzed MDNBR 15.3.2 Flow Controller Malfunction Not Applicable 14.7 15.3.3 Reactor Coolant Pump Rotor Bounded Ref. 3 14.7 Seizure Analyzed MDNBR 15.3.4 Reactor Coolant Pump Shaft Bounded 15.3.3 14.7 Break 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled Control Bank Calculated MDNBR and maximum 14.2.1 Withdrawal from a fuel temperature Subcritical or Low Power Condition 15.4.2 Uncontrolled Control Bank Calculated MDNBR and maximum 14.2.2 Withdrawal at Power LHGR Operation Conditions 15.4.3 Control Rod Misoperation | ||
: 1) Dropped Control Analyzed MDNBR Ref. 3 14.4 Bank/Rod 2) Dropped Part-Length Bounded 15.4.3(1) 14.6 Control Rod 3) Malpositioning of the Not Applicable Part-Length Control __ Group 4) Statically Misaligned Bounded 15.4.3(1) 14.6 Control | : 1) Dropped Control Analyzed MDNBR Ref. 3 14.4 Bank/Rod 2) Dropped Part-Length Bounded 15.4.3(1) 14.6 Control Rod 3) Malpositioning of the Not Applicable Part-Length Control __ Group 4) Statically Misaligned Bounded 15.4.3(1) 14.6 Control | ||
... ... .----** -5) Single Control Rod Analyzed MDNBR Ref, 10 14.2.3 Withdrawal and maximum LHGR 6) Core Barrel Failure Bounded 15.4.8 14.5 Siemens Power Corporation | ... ... .----** -5) Single Control Rod Analyzed MDNBR Ref, 10 14.2.3 Withdrawal and maximum LHGR 6) Core Barrel Failure Bounded 15.4.8 14.5 Siemens Power Corporation | ||
-Nuclear Division | -Nuclear Division | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 2-5 Table 2.1 Disposition of Events Summary for Palisades Cycle 14 (continued) | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 2-5 Table 2.1 Disposition of Events Summary for Palisades Cycle 14 (continued) | ||
SRP Event Designation Event Name 1 5.4.4 Startup of an Inactive Loop 15.4.5 Flow Controller Malfunction 15.4.6 CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant 1) Ra.ted and Power Operation Conditions | SRP Event Designation Event Name 1 5.4.4 Startup of an Inactive Loop 15.4.5 Flow Controller Malfunction 15.4.6 CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant 1) Ra.ted and Power Operation Conditions | ||
: 2) Reactor Critical, Hot Standby and Hot Shutdown 3) Refueling Shutdown Condition, Cold Shutdown Condition and Refueling Operation Disposition Bounded by rated power MDNBR. Bounding Event or Updated FSAR Reference Designation 14.8 Not Applicable: | : 2) Reactor Critical, Hot Standby and Hot Shutdown 3) Refueling Shutdown Condition, Cold Shutdown Condition and Refueling Operation Disposition Bounded by rated power MDNBR. Bounding Event or Updated FSAR Reference Designation 14.8 Not Applicable: | ||
No Flow Controller Bounded Ref. 26 14.3 Bounded Ref. 26 14.3 Bounded Ref. 26 14.3 15.4. 7 Inadvertent Loading and Operation Administrative procedures preclude this of a Fuel Assembly in an Improper event Position 15.4.8 Spectrum of Control Rod Ejection Analyzed Accidents 15.4.9 Spectrum of Rod Drop Accidents Not Applicable; BWR Event (BWR) 15.5 INCREASES IN REACTOR COOLANT INVENTORY 15.5.1 15.5.2 Inadvertent Operation of the ECCS During Power Operations CVCS Malfunction that Increases Reactor Coolant Inventory Siemens Power Corporation | No Flow Controller Bounded Ref. 26 14.3 Bounded Ref. 26 14.3 Bounded Ref. 26 14.3 15.4. 7 Inadvertent Loading and Operation Administrative procedures preclude this of a Fuel Assembly in an Improper event Position 15.4.8 Spectrum of Control Rod Ejection Analyzed Accidents 15.4.9 Spectrum of Rod Drop Accidents Not Applicable; BWR Event (BWR) 15.5 INCREASES IN REACTOR COOLANT INVENTORY 15.5.1 15.5.2 Inadvertent Operation of the ECCS During Power Operations CVCS Malfunction that Increases Reactor Coolant Inventory Siemens Power Corporation | ||
| Line 264: | Line 264: | ||
Event numbering and nomenclature are consistent with Chapter 1 5 of the SRP to facilitate review. This section provides information on the plant licensing basis including classification of plant conditions, classification of accident events by category, operating conditions, initial conditions, neutronics data, core and fuel design parameters. | Event numbering and nomenclature are consistent with Chapter 1 5 of the SRP to facilitate review. This section provides information on the plant licensing basis including classification of plant conditions, classification of accident events by category, operating conditions, initial conditions, neutronics data, core and fuel design parameters. | ||
Listings of systems and components available for accident mitigation, trip setpoints, time delays and component capacities are also included. | Listings of systems and components available for accident mitigation, trip setpoints, time delays and component capacities are also included. | ||
A non-LOCA system transient analysis for most SRP Chapter 15 events was performed for Cycle 9 (Reference 3). The events not considered in the Reference 3 analysis were the uncontrolled rod withdrawal from part power, which was analyzed and reported in Reference 13, and a loss of Normal Feedwater, which was analyzed and reported in Reference | A non-LOCA system transient analysis for most SRP Chapter 15 events was performed for Cycle 9 (Reference 3). The events not considered in the Reference 3 analysis were the uncontrolled rod withdrawal from part power, which was analyzed and reported in Reference 13, and a loss of Normal Feedwater, which was analyzed and reported in Reference | ||
: 12. A system analysis for uncontrolled rod withdrawal from a subcritical or low power condition was performed as a part of the support for Cycle 13 (Referer.ice 10). Subsequent to the Cycle 9 analysis, a system analysis was performed for the Loss of External Load event with two cases; one to determine the maximum primary pressure and one to minimize the DNBR (Reference 11 ). This event was analyzed for Cycle 14 to produce the highest secondary pressure. | : 12. A system analysis for uncontrolled rod withdrawal from a subcritical or low power condition was performed as a part of the support for Cycle 13 (Referer.ice 10). Subsequent to the Cycle 9 analysis, a system analysis was performed for the Loss of External Load event with two cases; one to determine the maximum primary pressure and one to minimize the DNBR (Reference 11 ). This event was analyzed for Cycle 14 to produce the highest secondary pressure. | ||
The results of this analysis are provided in this report. The changes introduced in Cycle 14 do not affect the system response for the events reported in References 3, 10, 11, 12 and 13. Thus, the system thermal-hydraulic responses for the various transients reported in References 3, 10, 11, 12, and 13 remain applicable for Cycle 14. MDNBR values for limiting AOOs and Postulated Accidents (PAs), as evaluated with the HTP DNB correlation, were calculated for Cycle 14. An XCOBRA-lllC model was developed for Cycle 14. This model incorporated the radial and axial power distributions from Cycle 14 as well as the hydraulic changes to the fuel assemblies. | The results of this analysis are provided in this report. The changes introduced in Cycle 14 do not affect the system response for the events reported in References 3, 10, 11, 12 and 13. Thus, the system thermal-hydraulic responses for the various transients reported in References 3, 10, 11, 12, and 13 remain applicable for Cycle 14. MDNBR values for limiting AOOs and Postulated Accidents (PAs), as evaluated with the HTP DNB correlation, were calculated for Cycle 14. An XCOBRA-lllC model was developed for Cycle 14. This model incorporated the radial and axial power distributions from Cycle 14 as well as the hydraulic changes to the fuel assemblies. | ||
| Line 270: | Line 270: | ||
-Nuclear Division | -Nuclear Division | ||
* | * | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-2 room habitability, DNB propagation, rod bow and FCM. The disposition of these items is given below: 3. 1 Thermal Hydraulic Compatibility The Reload R fuel to be loaded into Cycle 14 is identical in thermal-hydraulic behavior to the co-resident fuel except for the addition of an all inconel bottom spacer and a resistant lower tie plate. Hydraulic testing (Reference | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-2 room habitability, DNB propagation, rod bow and FCM. The disposition of these items is given below: 3. 1 Thermal Hydraulic Compatibility The Reload R fuel to be loaded into Cycle 14 is identical in thermal-hydraulic behavior to the co-resident fuel except for the addition of an all inconel bottom spacer and a resistant lower tie plate. Hydraulic testing (Reference | ||
: 25) showed different pressure loss coefficients for the HTP fuel design over what has previously been used. These new loss coefficients were incorporated into the Cycle 14 XCOBRA-lllC models. An XCOBRA-lllC model was developed for the detailed core model (non-LOCA transients) and for the reduced channel model (setpoints). | : 25) showed different pressure loss coefficients for the HTP fuel design over what has previously been used. These new loss coefficients were incorporated into the Cycle 14 XCOBRA-lllC models. An XCOBRA-lllC model was developed for the detailed core model (non-LOCA transients) and for the reduced channel model (setpoints). | ||
These models were used to reanalyze the MDNBR in support of operation for Cycle 14. In addition, a 2 % penalty was applied to the HTP DNBR limit to account for the mixture of hydraulic types in the core. 3. 2 DNB Propagation DNB propagation is most likely for conditions with high rod exposure and low system pressure. | These models were used to reanalyze the MDNBR in support of operation for Cycle 14. In addition, a 2 % penalty was applied to the HTP DNBR limit to account for the mixture of hydraulic types in the core. 3. 2 DNB Propagation DNB propagation is most likely for conditions with high rod exposure and low system pressure. | ||
Event 15.1.5 is the most limiting event for DNB propagation but since there were no DNB related fuel failures in the MSLB analysis there was no need to consider DNB propagation (Reference 23). 3.3 RodBow Due to a reduction in cladding thickness for the Reload R fuel, a Rod Bow analysis was performed per Reference | Event 15.1.5 is the most limiting event for DNB propagation but since there were no DNB related fuel failures in the MSLB analysis there was no need to consider DNB propagation (Reference 23). 3.3 RodBow Due to a reduction in cladding thickness for the Reload R fuel, a Rod Bow analysis was performed per Reference | ||
: 27. No DNB penalty is required until 50% gap closure, which does not occur until assembly exposures of 66 GWd/MTU. This is far above the maximum assembly burnup of 53,281 MWd/MTU projected for Cycle 14. FCM is unaffected because no rod bow penalty on Fa is required for burnups below 54,000 MWd/MTU. 3 .4 Control Room Habitability Control room habitability following a major accident can be affected by changes which increase the amount of radioactivity which could reach the control room. Consumers Energy currently performs all control room habitability and offsite dose radiological consequence analyses. | : 27. No DNB penalty is required until 50% gap closure, which does not occur until assembly exposures of 66 GWd/MTU. This is far above the maximum assembly burnup of 53,281 MWd/MTU projected for Cycle 14. FCM is unaffected because no rod bow penalty on Fa is required for burnups below 54,000 MWd/MTU. 3 .4 Control Room Habitability Control room habitability following a major accident can be affected by changes which increase the amount of radioactivity which could reach the control room. Consumers Energy currently performs all control room habitability and offsite dose radiological consequence analyses. | ||
The current analyses of record are contained in Reference | The current analyses of record are contained in Reference | ||
: 9. The re-analysis of the MSLB concludes that there would be no fuel failures therefore, the analysis of record for the control room habitability remains bounding. | : 9. The re-analysis of the MSLB concludes that there would be no fuel failures therefore, the analysis of record for the control room habitability remains bounding. | ||
Siemens Power Corporation | Siemens Power Corporation | ||
| Line 283: | Line 283: | ||
In cycles prior to Cycle 13, the U 235 enrichment of gadolinia-bearing rods was designed such that a gadolinia-bearing rod could not become limiting with respect to FCM and a conservative limit on U0 2 rods of 21 kW/ft was used as the FCM limit. Increasing the enrichment for the gadolinia-bearing fuel rods causes them to reach a higher relative power, compared to the peak U0 2 rod, during the latter portion of its first operating cycle. The thermal conductivity of a fuel pellet decreases with increasing gadolinia concentration. | In cycles prior to Cycle 13, the U 235 enrichment of gadolinia-bearing rods was designed such that a gadolinia-bearing rod could not become limiting with respect to FCM and a conservative limit on U0 2 rods of 21 kW/ft was used as the FCM limit. Increasing the enrichment for the gadolinia-bearing fuel rods causes them to reach a higher relative power, compared to the peak U0 2 rod, during the latter portion of its first operating cycle. The thermal conductivity of a fuel pellet decreases with increasing gadolinia concentration. | ||
In addition, the melt temperature decreases with increasing gadolinia concentration. | In addition, the melt temperature decreases with increasing gadolinia concentration. | ||
As a result of the increased power in the gadolinia-bearing fuel rods for the Cycle 14 fuel design, SPC calculated a FCM limit for the highest-powered U0 2 fuel rod in the bundle which precludes FCM in both U0 2 rods and gadolinia rods. The limit for Cycle 14 was calculated to be 20.93 kW/ft. The limit for Reload Q fuel (Cycle 13) is 21.64 kW/ft, based on the lower gadolinia concentration and the local power distribution for that design. This limit is conservative when applied to prior fuel designs for Cycles prior to Cycle 13, since the gadolinia-bearing fuel rods operated at much lower relative powers. The Linear Heat Generation Rate (LHGR) on the peak U0 2 rod required to prevent FCM for Cycle 14 in any rod was calculated based on SPC's approved setpoint methodology (Reference | As a result of the increased power in the gadolinia-bearing fuel rods for the Cycle 14 fuel design, SPC calculated a FCM limit for the highest-powered U0 2 fuel rod in the bundle which precludes FCM in both U0 2 rods and gadolinia rods. The limit for Cycle 14 was calculated to be 20.93 kW/ft. The limit for Reload Q fuel (Cycle 13) is 21.64 kW/ft, based on the lower gadolinia concentration and the local power distribution for that design. This limit is conservative when applied to prior fuel designs for Cycles prior to Cycle 13, since the gadolinia-bearing fuel rods operated at much lower relative powers. The Linear Heat Generation Rate (LHGR) on the peak U0 2 rod required to prevent FCM for Cycle 14 in any rod was calculated based on SPC's approved setpoint methodology (Reference | ||
: 17) I which describes the use of approved fuel design codes for calculating FCM powers. 1 5. 0 Accident Analyses 15.0.1 Categorization of Plant Events Plant events are placed in one of four categories. | : 17) I which describes the use of approved fuel design codes for calculating FCM powers. 1 5. 0 Accident Analyses 15.0.1 Categorization of Plant Events Plant events are placed in one of four categories. | ||
These categories, adopted by the American Nuclear Society (ANS), are described as follows: NORMAL OPERATION AND OPERATIONAL EVENTS (CONDITION I) | These categories, adopted by the American Nuclear Society (ANS), are described as follows: NORMAL OPERATION AND OPERATIONAL EVENTS (CONDITION I) | ||
| Line 295: | Line 295: | ||
* Events which are expected to occur once during the lifetime of the plant. LIMITING FAUL TS (CONDITION IV) EMF-98-013 Revision 0 Page 3-4 | * Events which are expected to occur once during the lifetime of the plant. LIMITING FAUL TS (CONDITION IV) EMF-98-013 Revision 0 Page 3-4 | ||
* Events which are not expected to occur but which are evaluated to demonstrate the adequacy of the design. Table 15.0.1 .1 lists the accident category used for each event considered in this report. This category is used in evaluating the acceptability of the results obtained from the analysis. | * Events which are not expected to occur but which are evaluated to demonstrate the adequacy of the design. Table 15.0.1 .1 lists the accident category used for each event considered in this report. This category is used in evaluating the acceptability of the results obtained from the analysis. | ||
These categories are unchanged from those established in Reference | These categories are unchanged from those established in Reference | ||
: 12. 15.0.2 Plant Characteristics and Initial Conditions Six operational conditions were considered in the disposition and analysis. | : 12. 15.0.2 Plant Characteristics and Initial Conditions Six operational conditions were considered in the disposition and analysis. | ||
These are given in Table 15.0.2.1. | These are given in Table 15.0.2.1. | ||
| Line 317: | Line 317: | ||
S 542.99 + (P -2,060) x [0.058+1.0x1 o-5 (P -2,060)) + (w -138) x [1.125 -0.0205 (w -138)) where P is pressurizer pressure ( 1,800 psia S P S 2,200 psia) and w is the primary coolant system (PCS) mass flow rate (100 Mlbm/hr S w S 150 Mlbm/hr). | S 542.99 + (P -2,060) x [0.058+1.0x1 o-5 (P -2,060)) + (w -138) x [1.125 -0.0205 (w -138)) where P is pressurizer pressure ( 1,800 psia S P S 2,200 psia) and w is the primary coolant system (PCS) mass flow rate (100 Mlbm/hr S w S 150 Mlbm/hr). | ||
The pressure range over which the T;niet LCO is confirmed is set by the low pressure trip from the TM/LP ( 1, 750 psia) compensated for uncertainties and by the high pressure trip of 2,255 psia. Both trips were compensated for uncertainties and for the control deadband of +/- 50 psi. The flow range covers all possible operational ranges, even for very high tube plugging levels in the steam generators. | The pressure range over which the T;niet LCO is confirmed is set by the low pressure trip from the TM/LP ( 1, 750 psia) compensated for uncertainties and by the high pressure trip of 2,255 psia. Both trips were compensated for uncertainties and for the control deadband of +/- 50 psi. The flow range covers all possible operational ranges, even for very high tube plugging levels in the steam generators. | ||
For primary loop flow rates greater than 150 Mlbm/hr, the inlet temperature is limited to the T;niet LCO value at 150 Mlbm/hr or to 544°F, whichever is less. The T;niet LCO is further adjusted to account for axial power shapes by the curve shown in Figure 15 .0. 7. 1 . Confirmation of the T;niet*LCO makes use of the XCOBRA-lllC computer code (Reference | For primary loop flow rates greater than 150 Mlbm/hr, the inlet temperature is limited to the T;niet LCO value at 150 Mlbm/hr or to 544°F, whichever is less. The T;niet LCO is further adjusted to account for axial power shapes by the curve shown in Figure 15 .0. 7. 1 . Confirmation of the T;niet*LCO makes use of the XCOBRA-lllC computer code (Reference | ||
: 15) to model the reactor core. The augmented radial peaking which results from the dropped control rod, was included for axial power shapes covering the ASI range of the function. | : 15) to model the reactor core. The augmented radial peaking which results from the dropped control rod, was included for axial power shapes covering the ASI range of the function. | ||
The Cycle 14 XCOBRA-lllC four channel model and the Cycle 14 axial power shapes were used to determine the power corresponding to 95/95 safety limit including a 2% mixed core DNB This analysis.was performed covering a range of pressurizer pressures, core inlet temperatures and prima.ry coolant system flow rates. Siemens Power Corporation | The Cycle 14 XCOBRA-lllC four channel model and the Cycle 14 axial power shapes were used to determine the power corresponding to 95/95 safety limit including a 2% mixed core DNB This analysis.was performed covering a range of pressurizer pressures, core inlet temperatures and prima.ry coolant system flow rates. Siemens Power Corporation | ||
-Nuclear Division | -Nuclear Division | ||
* .. * * | * .. * * | ||
* Palisades Cycle 1 4: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-7 The Cycle 14 verification analysis used the following uncertainties and transient allowances: | * Palisades Cycle 1 4: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-7 The Cycle 14 verification analysis used the following uncertainties and transient allowances: | ||
(1) a +/- 22 psia pressure measurement uncertainty, (2) a +/- 7°F inlet temperature uncertainty (Reference 12), (includes a+/- 5°F tilt allowance, and a +/- 2°F measurement uncertainty), (3) a + 0/-6% flow rate uncertainty (includes a -3% bypass flow and +/- 3% measurement uncertainty) and, (4) a -20 psi transient pressure bias. These values are the same as those used for the Cycle 13 confirmation of the T;niat LCO (Reference 10). The T;niet LCO (as defined by the functional relationship above and by Figure 15 .0. 7 .1) was confirmed to provide protection against DNB for the control rod drop event over the pressure and flow ranges established by the functional relationship and the ASI and power relationship shown in Figure 15.0.7.1. | (1) a +/- 22 psia pressure measurement uncertainty, (2) a +/- 7°F inlet temperature uncertainty (Reference 12), (includes a+/- 5°F tilt allowance, and a +/- 2°F measurement uncertainty), (3) a + 0/-6% flow rate uncertainty (includes a -3% bypass flow and +/- 3% measurement uncertainty) and, (4) a -20 psi transient pressure bias. These values are the same as those used for the Cycle 13 confirmation of the T;niat LCO (Reference 10). The T;niet LCO (as defined by the functional relationship above and by Figure 15 .0. 7 .1) was confirmed to provide protection against DNB for the control rod drop event over the pressure and flow ranges established by the functional relationship and the ASI and power relationship shown in Figure 15.0.7.1. | ||
The XCOBRA-lllC calculations demonstrated that for the maximum inlet temperatures allowed by the T;niet LCO, the DNBR was always greater than the 95/95 safety limit for the HTP correlation, with a 2% mixed core penalty over the range of pressurizer pressures, primary coolant system flow rates, axial shape indices and core power levels expected for Cycle 14 . 15.0.7.2 Thermal Margin/Low Pressure (TM/LP) Trip The function of the TM/LP trip is to protect the core against DNB during a slow heat-up or de-pressurization transient. | The XCOBRA-lllC calculations demonstrated that for the maximum inlet temperatures allowed by the T;niet LCO, the DNBR was always greater than the 95/95 safety limit for the HTP correlation, with a 2% mixed core penalty over the range of pressurizer pressures, primary coolant system flow rates, axial shape indices and core power levels expected for Cycle 14 . 15.0.7.2 Thermal Margin/Low Pressure (TM/LP) Trip The function of the TM/LP trip is to protect the core against DNB during a slow heat-up or de-pressurization transient. | ||
| Line 332: | Line 332: | ||
* loss of feedwater | * loss of feedwater | ||
* inadvertent opening of a PORV 15.0. 7 .2.1 TM/LP Uncertainties The uncertainties used in the confirmation of the TM/LP for Cycle 14 are summarized in Table 15.0.7 .2. Uncertainties in measured inputs to the TM/LP trip; inlet temperature, Siemens Power Corporation | * inadvertent opening of a PORV 15.0. 7 .2.1 TM/LP Uncertainties The uncertainties used in the confirmation of the TM/LP for Cycle 14 are summarized in Table 15.0.7 .2. Uncertainties in measured inputs to the TM/LP trip; inlet temperature, Siemens Power Corporation | ||
-Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-8 core power, pressure and axial shape index; were incorporated in the confirmation of the TM/LP. The uncertainties for instrument drift in both power and inlet temperature, calorimetric power measurement, inlet temperature measurement and primary pressure measurement were combined into an additional pressure uncertainty of 165 psi (Reference | -Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-8 core power, pressure and axial shape index; were incorporated in the confirmation of the TM/LP. The uncertainties for instrument drift in both power and inlet temperature, calorimetric power measurement, inlet temperature measurement and primary pressure measurement were combined into an additional pressure uncertainty of 165 psi (Reference | ||
: 16) and a bias of 1.5°F was added to the inlet temperature to account for thermal lag of the cold leg RTDs and the transport time from the cold-leg RTD location to the core inlet (Reference 3). The XCOBRA-lllC analysis makes use of additional uncertainties which are not related to the measured parameters. | : 16) and a bias of 1.5°F was added to the inlet temperature to account for thermal lag of the cold leg RTDs and the transport time from the cold-leg RTD location to the core inlet (Reference 3). The XCOBRA-lllC analysis makes use of additional uncertainties which are not related to the measured parameters. | ||
These include an uncertainty to account for manufacturing tolerances and in-reactor changes such as densification, a 3% measurement uncertainty for PCS flow and a reduction in the active core flow to account for a 3 % core bypass flow. 15.0. 7.2.2 TM/LP Confirmation The TM/LP trip equation is: Pvar = 2012 X QA X QR 1 + 17.0 X T;n -9493 where Pvar is the trip setpoint (based on the measured pressurizer pressure), Tin is the inlet temperature and QA and OR1 are functions designed to compensate for the change in DNBR with axial power shapes and reactor power, respectively. | These include an uncertainty to account for manufacturing tolerances and in-reactor changes such as densification, a 3% measurement uncertainty for PCS flow and a reduction in the active core flow to account for a 3 % core bypass flow. 15.0. 7.2.2 TM/LP Confirmation The TM/LP trip equation is: Pvar = 2012 X QA X QR 1 + 17.0 X T;n -9493 where Pvar is the trip setpoint (based on the measured pressurizer pressure), Tin is the inlet temperature and QA and OR1 are functions designed to compensate for the change in DNBR with axial power shapes and reactor power, respectively. | ||
| Line 341: | Line 341: | ||
-Nuclear Division | -Nuclear Division | ||
* * | * * | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.0.7.3 Variable High Power (VHP) Trip Uncertainties EMF-98-013 Revision 0 Page 3-9 The VHP trip uses an internal power program to determine the appropriate trip power and to compare it to the power signal provided by the Thermal Margin Monitor (TMM). The TMM uses the maximum of the thermal power (based on average hot leg temperatures and maximum cold leg temperatures) and neutron flux power to create the power signal. The uncertainty in the power signal created by the TMM was analyzed in Reference | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.0.7.3 Variable High Power (VHP) Trip Uncertainties EMF-98-013 Revision 0 Page 3-9 The VHP trip uses an internal power program to determine the appropriate trip power and to compare it to the power signal provided by the Thermal Margin Monitor (TMM). The TMM uses the maximum of the thermal power (based on average hot leg temperatures and maximum cold leg temperatures) and neutron flux power to create the power signal. The uncertainty in the power signal created by the TMM was analyzed in Reference | ||
: 8. The analysis showed that the use of an uncertainty of +/- 8.5% for fast transients and +/- 5.5% for slow transients was conservative. | : 8. The analysis showed that the use of an uncertainty of +/- 8.5% for fast transients and +/- 5.5% for slow transients was conservative. | ||
15.0.7.4 Increased Trip Delays For Cycle 14 the impact of increasing the delays for the High Pressurizer Pressure trip, the Low Pressurizer Pressure trip, the Low Reactor Coolant Flow trip, the TM/LP trip and the VHP trip from the values used in the analyses reported in References 3, 11, 12, and 13 was evaluated. | 15.0.7.4 Increased Trip Delays For Cycle 14 the impact of increasing the delays for the High Pressurizer Pressure trip, the Low Pressurizer Pressure trip, the Low Reactor Coolant Flow trip, the TM/LP trip and the VHP trip from the values used in the analyses reported in References 3, 11, 12, and 13 was evaluated. | ||
| Line 353: | Line 353: | ||
-Nuclear Division | -Nuclear Division | ||
* | * | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 15 Events EMF-98-013 Revision 0 Page 3-10 1 . The RPS was designed with redundancy and independence to assure that no single failure or removal from service of any component or channel of a system would result in the loss of the protection function. | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 15 Events EMF-98-013 Revision 0 Page 3-10 1 . The RPS was designed with redundancy and independence to assure that no single failure or removal from service of any component or channel of a system would result in the loss of the protection function. | ||
: 2. Each Engineered Safety Feature (ESF) was designed to perform its intended safety function assuming a failure of a single active component. | : 2. Each Engineered Safety Feature (ESF) was designed to perform its intended safety function assuming a failure of a single active component. | ||
: 3. The onsite power system and the offsite power system were designed such that each shall independently be capable of providing power for the ESF assuming a failure of a single active component in either power system. The safety analysis was structured to demonstrate that the plant systems design satisfies these single failure criteria. | : 3. The onsite power system and the offsite power system were designed such that each shall independently be capable of providing power for the ESF assuming a failure of a single active component in either power system. The safety analysis was structured to demonstrate that the plant systems design satisfies these single failure criteria. | ||
The following was assumed: 1 . The ESF required to function in an event suffers a worst single failure of an active component. | The following was assumed: 1 . The ESF required to function in an event suffers a worst single failure of an active component. | ||
: 2. Reactor trips occur at the specified setpoint within the specified delay time with a worst single active failure. 3. A concurrent loss of offsite power occurred for these PAs: Main Steamline Break, Steam Generator Tube Rupture, and LOCA. 4. A concurrent loss of offsite power occurred for the Loss of Normal Feedwater event, which is an AOO. Criteria 10, 20, 25 and 29 of Title 10 of the Code of Federal Regulations, Part 50 ( 10 CFR 50), Appendix A, require that the design and operation of the plant and the RPS assure that the SAFDLs are not exceeded during AOOs. As per the definition of .AOO in 10 CFR 50, Appendix A: "Anticipated Operational Occurrences mean those conditions of normal operation which are expected to occur one or more times during the life of the plant and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all off site power." There are two SAFDLs: ( 1) the fuel shall not experience centerline melt and (2) the DNBR shall have a minimum allowable limit such that there is a 95 % probability with a 95 % confidence interval that DNB would not occur. The following sections, numbered according to the SRP, provide a discussion of the disposition of events review and MDNBR and peak LHGR analyses performed to support Cycle 14. Siemens Power Corporation | : 2. Reactor trips occur at the specified setpoint within the specified delay time with a worst single active failure. 3. A concurrent loss of offsite power occurred for these PAs: Main Steamline Break, Steam Generator Tube Rupture, and LOCA. 4. A concurrent loss of offsite power occurred for the Loss of Normal Feedwater event, which is an AOO. Criteria 10, 20, 25 and 29 of Title 10 of the Code of Federal Regulations, Part 50 ( 10 CFR 50), Appendix A, require that the design and operation of the plant and the RPS assure that the SAFDLs are not exceeded during AOOs. As per the definition of .AOO in 10 CFR 50, Appendix A: "Anticipated Operational Occurrences mean those conditions of normal operation which are expected to occur one or more times during the life of the plant and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all off site power." There are two SAFDLs: ( 1) the fuel shall not experience centerline melt and (2) the DNBR shall have a minimum allowable limit such that there is a 95 % probability with a 95 % confidence interval that DNB would not occur. The following sections, numbered according to the SRP, provide a discussion of the disposition of events review and MDNBR and peak LHGR analyses performed to support Cycle 14. Siemens Power Corporation | ||
-Nuclear Division | -Nuclear Division | ||
| Line 383: | Line 383: | ||
-Nuclear Division | -Nuclear Division | ||
* | * | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.1.4.2 Event Disposition and Justification EMF-98-013 Revision 0 Page 3-14 The increase in steam flow due to opening a steam generator valve is less than that considered in the Increase in Steam Flow event (Event 15.1.3), and is bounded by Event 1 5. 1 . 3. The Cycle 14 changes do not affect this conclusion and Events 1 5. 1 . 3 and 15.1.4 continue to bound this event. 15.1.5 Steam System Piping Failures Inside and Outside of Containment 15.1.5.1 Event Description The steam line break event is initiated by a double-ended, guillotine break of the main steam line which leads to an uncontrolled steam release from the secondary system. The increase .in energy removal through the secondary system results in a severe overcooling of the primary system. At EOC conditions, which are characterized by the most negative MTC in the cycle, this cooldown results in a large insertion of reactivity and, potentially, a return to power. The most reactive control rod is assumed stuck in a fully withdrawn position, thereby reducing the s.hutdown margin and introducing high power-peaking factors which could lead to significant DNB and LHGR challenges with oniy a modest return to power. 15.1.5.2 Event Disposition and Justification The Main Steamline Break was re-analyzed for Cycle 14 as discussed in Section 2.2 and the results are reported in Reference | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.1.4.2 Event Disposition and Justification EMF-98-013 Revision 0 Page 3-14 The increase in steam flow due to opening a steam generator valve is less than that considered in the Increase in Steam Flow event (Event 15.1.3), and is bounded by Event 1 5. 1 . 3. The Cycle 14 changes do not affect this conclusion and Events 1 5. 1 . 3 and 15.1.4 continue to bound this event. 15.1.5 Steam System Piping Failures Inside and Outside of Containment 15.1.5.1 Event Description The steam line break event is initiated by a double-ended, guillotine break of the main steam line which leads to an uncontrolled steam release from the secondary system. The increase .in energy removal through the secondary system results in a severe overcooling of the primary system. At EOC conditions, which are characterized by the most negative MTC in the cycle, this cooldown results in a large insertion of reactivity and, potentially, a return to power. The most reactive control rod is assumed stuck in a fully withdrawn position, thereby reducing the s.hutdown margin and introducing high power-peaking factors which could lead to significant DNB and LHGR challenges with oniy a modest return to power. 15.1.5.2 Event Disposition and Justification The Main Steamline Break was re-analyzed for Cycle 14 as discussed in Section 2.2 and the results are reported in Reference | ||
: 23. 1 5. 2 Decrease in Heat Removal by the Secondary System The initiating mechanisms for Events 15.2.1, 15.2.2, 15.2.3, 15.2.4 and the heatup period of 15.2.8 are not affected by any of the Cycle 14 changes and the relative severity of these events established for Cycle 13 (Reference | : 23. 1 5. 2 Decrease in Heat Removal by the Secondary System The initiating mechanisms for Events 15.2.1, 15.2.2, 15.2.3, 15.2.4 and the heatup period of 15.2.8 are not affected by any of the Cycle 14 changes and the relative severity of these events established for Cycle 13 (Reference | ||
: 10) remain valid. The Loss of External Load, Event 15.2.1, bounds Events 15.2.2, 15.2,3, 15.2.4 and the heatup period of Event 15.2.8. 15.2.1 Loss of External Load 15.2.1.1 Event Description A Loss of External Load is initiated by either a loss of external electrical load or a turbine trip and the turbine stop valve is assumed to rapidly close (0.1 second). The plant response to this event would not change if a shorter valve closure time were assumed, Siemens Power Corporation | : 10) remain valid. The Loss of External Load, Event 15.2.1, bounds Events 15.2.2, 15.2,3, 15.2.4 and the heatup period of Event 15.2.8. 15.2.1 Loss of External Load 15.2.1.1 Event Description A Loss of External Load is initiated by either a loss of external electrical load or a turbine trip and the turbine stop valve is assumed to rapidly close (0.1 second). The plant response to this event would not change if a shorter valve closure time were assumed, Siemens Power Corporation | ||
-Nuclear Division | -Nuclear Division | ||
| Line 394: | Line 394: | ||
The primary system is protected against over-pressurization by the pressurizer safety relief valves (SRVs). Pressure relief on the secondary side is afforded by the Main Steam Safety Valves (MSSVs). Actuation of the SRVs and the MSSVs limits the magnitude of the primary system temperature and pressure increase. | The primary system is protected against over-pressurization by the pressurizer safety relief valves (SRVs). Pressure relief on the secondary side is afforded by the Main Steam Safety Valves (MSSVs). Actuation of the SRVs and the MSSVs limits the magnitude of the primary system temperature and pressure increase. | ||
With a positive MTC, corresponding to BOC, increasing primary system temperatures result in increasing core power. The increasing primary side temperatures and power reduce the margin to thermal limits (i.e., DNBR limits) and challenge the DNBR SAFDL. 15.2.1.2 Event Disposition and Justification The parameters influencing the severity of the transient include: ( 1) PCS high pressure trip setpoint, (2) SRV setpoints, (3) PCS over-pressure relief capacity, (4) Primary to secondary heat transfer, (5) MSSV setpoints, (6) Secondary side pressure, (7) MSSV relief capability, (8) Moderator reactivity coefficients, (9) Fuel assembly thermal hydraulic characteristics, and ( 10) Increased pressurizer spray flows. Initiating this event from full power bounds all other operating conditions. | With a positive MTC, corresponding to BOC, increasing primary system temperatures result in increasing core power. The increasing primary side temperatures and power reduce the margin to thermal limits (i.e., DNBR limits) and challenge the DNBR SAFDL. 15.2.1.2 Event Disposition and Justification The parameters influencing the severity of the transient include: ( 1) PCS high pressure trip setpoint, (2) SRV setpoints, (3) PCS over-pressure relief capacity, (4) Primary to secondary heat transfer, (5) MSSV setpoints, (6) Secondary side pressure, (7) MSSV relief capability, (8) Moderator reactivity coefficients, (9) Fuel assembly thermal hydraulic characteristics, and ( 10) Increased pressurizer spray flows. Initiating this event from full power bounds all other operating conditions. | ||
The Reference 11 analysis evaluated the maximum primary system over-pressure using SPC's ANF-RELAP methodology (Reference | The Reference 11 analysis evaluated the maximum primary system over-pressure using SPC's ANF-RELAP methodology (Reference | ||
: 18) and assumed all MSSVs were operational. | : 18) and assumed all MSSVs were operational. | ||
Palisades Technical Specifications allow one MSSV to be inoperable. | Palisades Technical Specifications allow one MSSV to be inoperable. | ||
A subsequent evaluation of the impact of operation with an inoperable MSSV (Reference | A subsequent evaluation of the impact of operation with an inoperable MSSV (Reference | ||
: 19) confirmed that it would not have a significant impact on the primary system pressure for this event . Secondary side pressurization was analyzed for the Loss of External Load event subsequent to the Reference 18 analysis. | : 19) confirmed that it would not have a significant impact on the primary system pressure for this event . Secondary side pressurization was analyzed for the Loss of External Load event subsequent to the Reference 18 analysis. | ||
The analysis used revised models for the Siemens Power Corporation | The analysis used revised models for the Siemens Power Corporation | ||
| Line 469: | Line 469: | ||
-Nuclear Division | -Nuclear Division | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-23 In the near term, the pump coastdown characteristics and the timing of the reactor trip, trip delays and scram bank insertion characteristics are key parameters. | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-23 In the near term, the pump coastdown characteristics and the timing of the reactor trip, trip delays and scram bank insertion characteristics are key parameters. | ||
In the longer term, natural circulation flow in the primary system removes the decay power. The primary concern with this event is the near-term challenge to the SAFDls. The event is analyzed to verify that the reactor protection system can respond fast enough to prevent penetration of the DNBR SAFDL. *-. This event is classified as a moderate frequency event (Table 15.0.1.1 ). The acceptance criteria are as described in Section 15.0.10. For this analysis, the systems challenged in this event are redundant; no single active failure in the RPS or ESF will adversely affect the consequences of the event. long term recovery is provided by the auxiliary feedwater system, as demonstrated in the analysis of Event 15.2.7 in Reference | In the longer term, natural circulation flow in the primary system removes the decay power. The primary concern with this event is the near-term challenge to the SAFDls. The event is analyzed to verify that the reactor protection system can respond fast enough to prevent penetration of the DNBR SAFDL. *-. This event is classified as a moderate frequency event (Table 15.0.1.1 ). The acceptance criteria are as described in Section 15.0.10. For this analysis, the systems challenged in this event are redundant; no single active failure in the RPS or ESF will adversely affect the consequences of the event. long term recovery is provided by the auxiliary feedwater system, as demonstrated in the analysis of Event 15.2.7 in Reference | ||
: 12. 15.3.1.2 Event Disposition The most limiting transient associated with a loss of Reactor Coolant Flow is initiated from rated power by a loss of power to all four primary coolant pumps. As the pumps ' coast down, the core flow is reduced, causing a reactor scram on low flow. As the flow coasts down, primary temperatures increase. | : 12. 15.3.1.2 Event Disposition The most limiting transient associated with a loss of Reactor Coolant Flow is initiated from rated power by a loss of power to all four primary coolant pumps. As the pumps ' coast down, the core flow is reduced, causing a reactor scram on low flow. As the flow coasts down, primary temperatures increase. | ||
The Reference 3 analysis was performed using a 0.5 sec conservatism in the low flow trip delay time which bounds the Cycle 14 increase in low flow trip delay time of 0.2 sec. This increase in temperature causes a subsequent power rise due to moderator reactivity feedback. | The Reference 3 analysis was performed using a 0.5 sec conservatism in the low flow trip delay time which bounds the Cycle 14 increase in low flow trip delay time of 0.2 sec. This increase in temperature causes a subsequent power rise due to moderator reactivity feedback. | ||
The primary challenge to DNB is from the flow rate, increased coolant temperatures and increasing core heat flux. Increased pressure resulting from the heatup of the PCS is bounded by the Loss of External Load (Event 15.2.1 ). The MDNBR analysis was performed for Cycle 14. None of the. changes for Cycle 14 will cause the system response for this event to become more limiting and the system response in the analysis of record (Reference | The primary challenge to DNB is from the flow rate, increased coolant temperatures and increasing core heat flux. Increased pressure resulting from the heatup of the PCS is bounded by the Loss of External Load (Event 15.2.1 ). The MDNBR analysis was performed for Cycle 14. None of the. changes for Cycle 14 will cause the system response for this event to become more limiting and the system response in the analysis of record (Reference | ||
: 3) remains bounding for Cycle 14. Increased Pressurizer Spray Flow -An increase in the pressurizer spray flow will reduce the primary pressure increase. | : 3) remains bounding for Cycle 14. Increased Pressurizer Spray Flow -An increase in the pressurizer spray flow will reduce the primary pressure increase. | ||
Therefore MDNBR for this event would decrease with increased spray flow. The decrease in MDNBR will be small and the result would remain above the HTP correlation limit. Siemens Power Corporation | Therefore MDNBR for this event would decrease with increased spray flow. The decrease in MDNBR will be small and the result would remain above the HTP correlation limit. Siemens Power Corporation | ||
| Line 481: | Line 481: | ||
Following the reactor trip, the heat stored in the fuel rods continues to be transferred to the primary coolant. Because of the reduced core flow, the coolant temperatures rise, the primary coolant expands and flow surges into the pressurizer creating an increasing PCS pressure. | Following the reactor trip, the heat stored in the fuel rods continues to be transferred to the primary coolant. Because of the reduced core flow, the coolant temperatures rise, the primary coolant expands and flow surges into the pressurizer creating an increasing PCS pressure. | ||
As the pressure increases, the pressurizer sprays and then the SRVs act to control the system pressure. | As the pressure increases, the pressurizer sprays and then the SRVs act to control the system pressure. | ||
The rapid reduction in core flow and the increase in coolant temperature presents a very dramatic decrease in MDNBR and the event presents a significant challenge to DNB. Fuel centerline melt is not a serious concern because of the small power increase typical of this event. PCS pressurization criteria have not been approached in prior SPC analyses of this event and are not a serious concern for this event. The Reactor Coolant Pump Rotor Seizure is an infrequent event (Table 15.0.1.1 ). The acceptance criteria for this event are presented in Section 15.0.10. For this analysis, the systems challenged in this event are redundant; no single active failure in the RPS or ESF will adversely affect the consequences of the event. The auxiliary feedwater pumps will provide cooling capability after scram, as demonstrated in Section 15.2. Tot Reference | The rapid reduction in core flow and the increase in coolant temperature presents a very dramatic decrease in MDNBR and the event presents a significant challenge to DNB. Fuel centerline melt is not a serious concern because of the small power increase typical of this event. PCS pressurization criteria have not been approached in prior SPC analyses of this event and are not a serious concern for this event. The Reactor Coolant Pump Rotor Seizure is an infrequent event (Table 15.0.1.1 ). The acceptance criteria for this event are presented in Section 15.0.10. For this analysis, the systems challenged in this event are redundant; no single active failure in the RPS or ESF will adversely affect the consequences of the event. The auxiliary feedwater pumps will provide cooling capability after scram, as demonstrated in Section 15.2. Tot Reference | ||
: 12. 15.3.3.2 Event Disposition The controlling parameters for the Reactor Coolant Pump Seizure are identical to those for the Loss of Forced Reactor Coolant Flow (Event 15.3.1) and the arguments, given in Section 15.3.1.2, for Reload Rare applicable to this event. The system response is unchanged for Cycle 14. The MDNBR analysis was performed for Cycle 14 . Siemens Power Corporation | : 12. 15.3.3.2 Event Disposition The controlling parameters for the Reactor Coolant Pump Seizure are identical to those for the Loss of Forced Reactor Coolant Flow (Event 15.3.1) and the arguments, given in Section 15.3.1.2, for Reload Rare applicable to this event. The system response is unchanged for Cycle 14. The MDNBR analysis was performed for Cycle 14 . Siemens Power Corporation | ||
-Nuclear Division | -Nuclear Division | ||
| Line 494: | Line 494: | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-26 The response of the reactor core to the withdrawal of control banks from subcritical or low power exhibits an apparent delay in responding to the reactivity insertion, followed by an extremely rapid power ramp. This behavior is caused by the fact that the doubling time for the power when the reactivity is less than about 95 ¢ is based on the time constants of the delayed neutrons and the power takes a relatively long time to respond to the reactivity insertion. | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-26 The response of the reactor core to the withdrawal of control banks from subcritical or low power exhibits an apparent delay in responding to the reactivity insertion, followed by an extremely rapid power ramp. This behavior is caused by the fact that the doubling time for the power when the reactivity is less than about 95 ¢ is based on the time constants of the delayed neutrons and the power takes a relatively long time to respond to the reactivity insertion. | ||
When the reactivity insertion is rapid enough that the delayed neutron population cannot bring the power up to significant levels and cause a trip before the reactor approaches prompt critical, a very rapid power excursion can result. The power excursion is first mitigated by the Doppler feedback as the fuel temperature begins to rise and ultimately terminated by the RPS. Because the power increase in this event is very rapid, fuel rod temperature and surface heat flux lag behind the neutron power. The power transient (as well as the control rod withdrawal) can be terminated by one of the following RPS trip functions: 1 . Non-safety grade high rate-of-change of power trip, 1x10" 4% to 1 5 % power. No credit is taken for this trip by SPC. 2. VHP trip 3. TM/LP trip 4. High pressurizer pressure trip 5. High rate-of-change of power alarms which initiate Rod Withdrawal Prohibit Action. No credit is taken for this trip by SPC. | When the reactivity insertion is rapid enough that the delayed neutron population cannot bring the power up to significant levels and cause a trip before the reactor approaches prompt critical, a very rapid power excursion can result. The power excursion is first mitigated by the Doppler feedback as the fuel temperature begins to rise and ultimately terminated by the RPS. Because the power increase in this event is very rapid, fuel rod temperature and surface heat flux lag behind the neutron power. The power transient (as well as the control rod withdrawal) can be terminated by one of the following RPS trip functions: 1 . Non-safety grade high rate-of-change of power trip, 1x10" 4% to 1 5 % power. No credit is taken for this trip by SPC. 2. VHP trip 3. TM/LP trip 4. High pressurizer pressure trip 5. High rate-of-change of power alarms which initiate Rod Withdrawal Prohibit Action. No credit is taken for this trip by SPC. | ||
* 6. If the reactor is in hot shutdown or below and one PCS pump has been out of service for more than 12 hours, circuit breakers 42-01 and 42-02 are open (Reference | * 6. If the reactor is in hot shutdown or below and one PCS pump has been out of service for more than 12 hours, circuit breakers 42-01 and 42-02 are open (Reference | ||
: 14) which prevents a rod or bank withdrawal. | : 14) which prevents a rod or bank withdrawal. | ||
Of these trips, the minimum power setting of the VHP trip is the only RPS function SPC relies on to terminate this event. 15.4.1.2 Event Disposition and Justification This event can be initiated from one of two operating conditions: | Of these trips, the minimum power setting of the VHP trip is the only RPS function SPC relies on to terminate this event. 15.4.1.2 Event Disposition and Justification This event can be initiated from one of two operating conditions: | ||
Hot Standby (Condition | Hot Standby (Condition | ||
: 2) or Hot Shutdown (Condition 3). SPC has also evaluated the event for a transition within the Operating Conditions. | : 2) or Hot Shutdown (Condition 3). SPC has also evaluated the event for a transition within the Operating Conditions. | ||
Condition 2 -In this Operating Condition, the reactor has a fission power less than 2% of rated with one or more of the control banks inserted up to, and including, the Hot Zero Siemens Power Corporation | Condition 2 -In this Operating Condition, the reactor has a fission power less than 2% of rated with one or more of the control banks inserted up to, and including, the Hot Zero Siemens Power Corporation | ||
| Line 508: | Line 508: | ||
This transition would result in the event being initiated at HZP with the control banks at the HZP PDIL and the reactor just subcritical. | This transition would result in the event being initiated at HZP with the control banks at the HZP PDIL and the reactor just subcritical. | ||
Reactor power could be as low as 10-9 of rated. SPC evaluated this case to provide a bounding analysis for all bank withdrawals in this category. | Reactor power could be as low as 10-9 of rated. SPC evaluated this case to provide a bounding analysis for all bank withdrawals in this category. | ||
The analysis of this event consisted of a PTSPWR (Reference | The analysis of this event consisted of a PTSPWR (Reference | ||
: 2) run, initiated from 10*9 of rated power. From this PTSPWR run, the conditions at the point of MDNBR were used as boundary conditions for an XCOBRA-lllC (Reference | : 2) run, initiated from 10*9 of rated power. From this PTSPWR run, the conditions at the point of MDNBR were used as boundary conditions for an XCOBRA-lllC (Reference | ||
: 15) analysis of the MDNBR. They were also used to calculate the fuel centerline temperature. | : 15) analysis of the MDNBR. They were also used to calculate the fuel centerline temperature. | ||
This event is described in Reference | This event is described in Reference | ||
: 10. The changes in Cycle 14 that impact this event are the reduction at the shutdown margin with less then four pumps running and the increase in the VHP trip delay time. Reduced SOM with Less than Four RCPs Operating | : 10. The changes in Cycle 14 that impact this event are the reduction at the shutdown margin with less then four pumps running and the increase in the VHP trip delay time. Reduced SOM with Less than Four RCPs Operating | ||
-The SOM was reduced from 3.75% to 3.5% dp with less than four RCPs operating. | -The SOM was reduced from 3.75% to 3.5% dp with less than four RCPs operating. | ||
| Line 557: | Line 557: | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-33 Reactor protection for this event during hot shutdown, refueling shutdown, cold shutdown, and refueling operating conditions is provided by Technical Specification Shutdown Margin requirements. | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-33 Reactor protection for this event during hot shutdown, refueling shutdown, cold shutdown, and refueling operating conditions is provided by Technical Specification Shutdown Margin requirements. | ||
For the reactor critical and hot standby operating conditions, reactor protection is provided by the VHP trip and a non-safety grade high rate-of-change of power trip. For the rated power and power operating conditions, reactor protection is afforded by the VHP and TM/LP trips. The probability of a circumferential rupture of the core support barrel is nearly the same as that of a major rupture of the primary system piping and this event is classified as a Limiting Fault event (Table 15.0.1.1) and the acceptance criteria are given in Section 15.0.10. 15.4.3.2 Event Disposition and Justification A. Dropped Control Rod or Bank A dropped bank will not produce as great a challenge to the DNB SAFDL as a dropped rod because core power will decrease to a low power level following the drop of the lowest worth bank. In response to this power decrease, the VHP trip setpoint will reset at the low power condition. | For the reactor critical and hot standby operating conditions, reactor protection is provided by the VHP trip and a non-safety grade high rate-of-change of power trip. For the rated power and power operating conditions, reactor protection is afforded by the VHP and TM/LP trips. The probability of a circumferential rupture of the core support barrel is nearly the same as that of a major rupture of the primary system piping and this event is classified as a Limiting Fault event (Table 15.0.1.1) and the acceptance criteria are given in Section 15.0.10. 15.4.3.2 Event Disposition and Justification A. Dropped Control Rod or Bank A dropped bank will not produce as great a challenge to the DNB SAFDL as a dropped rod because core power will decrease to a low power level following the drop of the lowest worth bank. In response to this power decrease, the VHP trip setpoint will reset at the low power condition. | ||
This prevents a significant return to power and even though the peaking for the dropped bank is quite high, the MDNBR for the dropped bank will be less. than that of a dropped rod. The system analysis performed for the dropped bank (Reference | This prevents a significant return to power and even though the peaking for the dropped bank is quite high, the MDNBR for the dropped bank will be less. than that of a dropped rod. The system analysis performed for the dropped bank (Reference | ||
: 3) conservatively disregarded the VHP trip in the dropped bank analysis. | : 3) conservatively disregarded the VHP trip in the dropped bank analysis. | ||
The MDNBR presented in Table 2.2 for the dropped bank is therefore overly conservative and in fact with. the VHP trip available, the dropped bank transient would have terminated before core conditions reached a level such that the MDNBR for the dropped bank would be less than that of a dropped rod. The MDNBR analysis performed for the Dropped Rod event in Cycle 14 therefore remains bounding of the dropped bank event. B. Dropped Part-Length Control Rod A dropped part-length. | The MDNBR presented in Table 2.2 for the dropped bank is therefore overly conservative and in fact with. the VHP trip available, the dropped bank transient would have terminated before core conditions reached a level such that the MDNBR for the dropped bank would be less than that of a dropped rod. The MDNBR analysis performed for the Dropped Rod event in Cycle 14 therefore remains bounding of the dropped bank event. B. Dropped Part-Length Control Rod A dropped part-length. | ||
| Line 572: | Line 572: | ||
15.4.6 CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant 15.4.6.1 Event Description Dilution of the boron concentration can occur when primary grade water is added to the PCS via the Chemical Volume and Control System (CVCS) during cold shutdown or refueling shutdown conditions. | 15.4.6 CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant 15.4.6.1 Event Description Dilution of the boron concentration can occur when primary grade water is added to the PCS via the Chemical Volume and Control System (CVCS) during cold shutdown or refueling shutdown conditions. | ||
Boron dilution at power operation is used in the Uncontrolled Bank Withdrawal analysis (Event 15.4.2) to set the slowest reactivity insertion rates so that Event 15.4.2 bounds this event. The dilution of primary system boron adds positive reactivity to the core. This event can lead to an erosion of shutdown margin for subcritical initial conditions, or a slow power excursion at power operation. | Boron dilution at power operation is used in the Uncontrolled Bank Withdrawal analysis (Event 15.4.2) to set the slowest reactivity insertion rates so that Event 15.4.2 bounds this event. The dilution of primary system boron adds positive reactivity to the core. This event can lead to an erosion of shutdown margin for subcritical initial conditions, or a slow power excursion at power operation. | ||
Following operator detection of the boron dilution, operator action must be taken to terminate the dilution and to restore the required shutdown margin. A minimum of 15 minutes (Operating Conditions 1 through 5) or 30 minutes (Operating Condition | Following operator detection of the boron dilution, operator action must be taken to terminate the dilution and to restore the required shutdown margin. A minimum of 15 minutes (Operating Conditions 1 through 5) or 30 minutes (Operating Condition | ||
: 6) is allowed for the operator to both identify and terminate the Siemens Power Corporation | : 6) is allowed for the operator to both identify and terminate the Siemens Power Corporation | ||
-Nuclear Division | -Nuclear Division | ||
| Line 610: | Line 610: | ||
Assuming a reactor trip on TM/LP or Safety Injection Signal (SIS), no fuel failures would be predicted to occur and the radiological consequences are limited, since the source term can have no higher activity than the maximum primary coolant activity level allowed by the Technical Specifications . Siemens Power Corporation | Assuming a reactor trip on TM/LP or Safety Injection Signal (SIS), no fuel failures would be predicted to occur and the radiological consequences are limited, since the source term can have no higher activity than the maximum primary coolant activity level allowed by the Technical Specifications . Siemens Power Corporation | ||
-Nuclear Division | -Nuclear Division | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.6.2.2 Event Disposition and Justification EMF-98-013 Revision 0 Page 3-40 The changes associated with Cycle 14 operation affect neither the initiating faults leading to the pipe rupture nor the primary coolant activity level. Therefore, this event remains bounded by the analysis reported in Reference | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.6.2.2 Event Disposition and Justification EMF-98-013 Revision 0 Page 3-40 The changes associated with Cycle 14 operation affect neither the initiating faults leading to the pipe rupture nor the primary coolant activity level. Therefore, this event remains bounded by the analysis reported in Reference | ||
: 9. 15.6.3 Radiological Consequences of Steam Generator Tube Failure 15.6.3.1 Event Description This event is initiated by the mechanical failure of a steam generator tube, which causes coolant to flow from the primary system to the secondary system. This flow can deplete the primary coolant inventory thus reducing the PCS pressure, which degrades the margin to the DNB SAFDL. The tube failure also results in release of fission products from the PCS coolant to the steam side of the steam generator, which is outside of containment. | : 9. 15.6.3 Radiological Consequences of Steam Generator Tube Failure 15.6.3.1 Event Description This event is initiated by the mechanical failure of a steam generator tube, which causes coolant to flow from the primary system to the secondary system. This flow can deplete the primary coolant inventory thus reducing the PCS pressure, which degrades the margin to the DNB SAFDL. The tube failure also results in release of fission products from the PCS coolant to the steam side of the steam generator, which is outside of containment. | ||
The MSSVs are assumed to open, and the fission products are released directly to the environment. | The MSSVs are assumed to open, and the fission products are released directly to the environment. | ||
| Line 624: | Line 624: | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-41 15.6.5 Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary 15.6.5.1 Event Description A LOCA is initiated by a failure of PCS pressure boundary. | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-41 15.6.5 Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary 15.6.5.1 Event Description A LOCA is initiated by a failure of PCS pressure boundary. | ||
The event may be characterized, based on the initiators; as either SBLOCAs, involving breaks in small auxiliary piping, or LBLOCAs, involving complete ruptures of the PCS piping. The limiting features of the analyses for these events are the peak clad temperature (PCT) and the time at elevated temperature, both of which influence the extent of localized and core-wide zircaloy oxidation reactions. | The event may be characterized, based on the initiators; as either SBLOCAs, involving breaks in small auxiliary piping, or LBLOCAs, involving complete ruptures of the PCS piping. The limiting features of the analyses for these events are the peak clad temperature (PCT) and the time at elevated temperature, both of which influence the extent of localized and core-wide zircaloy oxidation reactions. | ||
The SBLOCA analysis for Palisades is not performed by SPC. 15.6.5.2 Event Disposition and Justification The controlling parameters for the transient are the following: ( 1 ) the initial fuel stored energy, (2) the decay heat, (3) the radial and axial power profiles, (4) the fuel rod-to-PCS coolant heat transfer versus time, and (5) the operating conditions for the ECCS systems. The changes for Cycle 14 affect (1), (2), (3), (4) and (5). The LBLOCA event was analyzed for Cycle 14 and the results are reported in Reference 21 . 1 5. 7 Radioactive Releases from a Subsystem or Component | The SBLOCA analysis for Palisades is not performed by SPC. 15.6.5.2 Event Disposition and Justification The controlling parameters for the transient are the following: ( 1 ) the initial fuel stored energy, (2) the decay heat, (3) the radial and axial power profiles, (4) the fuel rod-to-PCS coolant heat transfer versus time, and (5) the operating conditions for the ECCS systems. The changes for Cycle 14 affect (1), (2), (3), (4) and (5). The LBLOCA event was analyzed for Cycle 14 and the results are reported in Reference 21 . 1 5. 7 Radioactive Releases from a Subsystem or Component | ||
: 15. 7 .1 Waste Gas System Failure This event is not sensitive to the changes described in Section 1 .0. 15.7.2 Radioactive Liquid Waste System Leak or Failure (Release to Atmosphere) | : 15. 7 .1 Waste Gas System Failure This event is not sensitive to the changes described in Section 1 .0. 15.7.2 Radioactive Liquid Waste System Leak or Failure (Release to Atmosphere) | ||
This event is not sensitive to the changes described in Section 1.0. 15.7 .3 Postulated Radioactive Releases Due to Liquid-Containing Tank Failures This event is not sensitive to the changes described in Section 1.0. 15. 7 .4 Radiological Consequences of Fuel Handling Accident 15.7.4.1 Event Description A fuel handling accident occurs when a fuel assembly is damaged during refueling operations such that fuel rods are ruptured, resulting in a release of radioactivity. | This event is not sensitive to the changes described in Section 1.0. 15.7 .3 Postulated Radioactive Releases Due to Liquid-Containing Tank Failures This event is not sensitive to the changes described in Section 1.0. 15. 7 .4 Radiological Consequences of Fuel Handling Accident 15.7.4.1 Event Description A fuel handling accident occurs when a fuel assembly is damaged during refueling operations such that fuel rods are ruptured, resulting in a release of radioactivity. | ||
| Line 632: | Line 632: | ||
* * | * * | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-42 included in the previous analysis of the radiological consequences of the Fuel Handling Accident. | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-42 included in the previous analysis of the radiological consequences of the Fuel Handling Accident. | ||
15.7.4.2 Event Disposition and Justification The inventory of radioactive fission products is determined by the exposure and power level of the assemblies or fuel rods. The change in the Reload R fuel design will not affect the reference analysis (Reference | 15.7.4.2 Event Disposition and Justification The inventory of radioactive fission products is determined by the exposure and power level of the assemblies or fuel rods. The change in the Reload R fuel design will not affect the reference analysis (Reference | ||
: 9) performed by Consumers Energy. 1 5. 7. 5 Spent Fuel Cask Drop Accidents 15.7.5.1 Event Description A spent fuel cask drop accident can result in the damage of an irradiated fuel assembly and the subsequent release of radioactivity. | : 9) performed by Consumers Energy. 1 5. 7. 5 Spent Fuel Cask Drop Accidents 15.7.5.1 Event Description A spent fuel cask drop accident can result in the damage of an irradiated fuel assembly and the subsequent release of radioactivity. | ||
The inventory of fission products is determined by the exposure and power level of the damaged assemblies. | The inventory of fission products is determined by the exposure and power level of the damaged assemblies. | ||
15.7.5.2 Event Disposition and Justification The inventory of radioactive fission products is determined by the exposure and power level of the assemblies or fuel rods. The power and peaking factors tor Cycle 14 are unchanged from Cycle 13. The mechanical design change made to the Reload R design will not impact the reference analysis (Reference | 15.7.5.2 Event Disposition and Justification The inventory of radioactive fission products is determined by the exposure and power level of the assemblies or fuel rods. The power and peaking factors tor Cycle 14 are unchanged from Cycle 13. The mechanical design change made to the Reload R design will not impact the reference analysis (Reference | ||
: 9) for this event . Siemens Power Corporation | : 9) for this event . Siemens Power Corporation | ||
-Nuclear Division | -Nuclear Division | ||
| Line 642: | Line 642: | ||
-Nuclear Division Categorya Moderate (AOO). Moderate (AOO) Moderate (AOO) Moderate (AOO) Limiting Fault (PA) Moderate (A'OO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Limiting Fault (PAI Moderate (AOO) Moderate (AOO) | -Nuclear Division Categorya Moderate (AOO). Moderate (AOO) Moderate (AOO) Moderate (AOO) Limiting Fault (PA) Moderate (A'OO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Limiting Fault (PAI Moderate (AOO) Moderate (AOO) | ||
* * | * * | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-44 Table 15.0.1.1 Accident Category Used for Each Analyzed Event (continued) a SRP Event Designation Event Name 15.3.3 Reactor Coolant Pump Rotor Seizure 15.3.4 Reactor Coolant Pump Shaft Break 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled Control Rod Bank Withdrawal from a Subcritical or Low Power Condition 15.4.2 Uncontrolled Control Rod Bank Withdrawal at Power Operation Conditions 1 5.4.3 Control Rod Misoperation | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-44 Table 15.0.1.1 Accident Category Used for Each Analyzed Event (continued) a SRP Event Designation Event Name 15.3.3 Reactor Coolant Pump Rotor Seizure 15.3.4 Reactor Coolant Pump Shaft Break 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled Control Rod Bank Withdrawal from a Subcritical or Low Power Condition 15.4.2 Uncontrolled Control Rod Bank Withdrawal at Power Operation Conditions 1 5.4.3 Control Rod Misoperation | ||
: 1) Dropped Control Bank/Rod 2) Dropped Part-Length Control Rod 3) Malpositioning of the Part-Length Control Group 4) Statically Misaligned Control Rod/Bank 5) Single Control Rod Withdrawal | : 1) Dropped Control Bank/Rod 2) Dropped Part-Length Control Rod 3) Malpositioning of the Part-Length Control Group 4) Statically Misaligned Control Rod/Bank 5) Single Control Rod Withdrawal | ||
: 6) Core Barrel Failure 15.4.4 Startup of an Inactive Loop 15.4.5 Flow Controller Malfunction 15.4.6 CVCS Malfunction that Results in a Decrease in the .Boron Concentration in the Reactor Coolant 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position 15.4.8 Spectrum of Control Rod Ejection Accidents Anticipated Operational Occurrence (AOO) Postulated Accident (PAI Siemens Power Corporation | : 6) Core Barrel Failure 15.4.4 Startup of an Inactive Loop 15.4.5 Flow Controller Malfunction 15.4.6 CVCS Malfunction that Results in a Decrease in the .Boron Concentration in the Reactor Coolant 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position 15.4.8 Spectrum of Control Rod Ejection Accidents Anticipated Operational Occurrence (AOO) Postulated Accident (PAI Siemens Power Corporation | ||
-Nuclear Division Category* | -Nuclear Division Category* | ||
| Line 659: | Line 659: | ||
Siemens Power Corporation | Siemens Power Corporation | ||
-Nuclear Division | -Nuclear Division | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan EMF-98-013 Revision 0 Page 3-48 Chapter 1 5 Events b d e Table 15.0.2.2 Nominal Plant Operating Conditions Core Thermal Power (MWt) 2,530 Pump Thermal Power, total (MWt) 15 System Pressure (psia) 2,060 Vessel Coolant Flow Rate (Mlbm/hr) 144.6 8 Active Coolant Flow Rate (Mlbm/hr) 140.3b Core Inlet Coolant Temperature | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan EMF-98-013 Revision 0 Page 3-48 Chapter 1 5 Events b d e Table 15.0.2.2 Nominal Plant Operating Conditions Core Thermal Power (MWt) 2,530 Pump Thermal Power, total (MWt) 15 System Pressure (psia) 2,060 Vessel Coolant Flow Rate (Mlbm/hr) 144.6 8 Active Coolant Flow Rate (Mlbm/hr) 140.3b Core Inlet Coolant Temperature | ||
(°F) 537°fC Steam Generator Pressure (psia) 770d Steam Flow Rate (Mlbm/hr) 10.982 Feedwater Temperature | (°F) 537°fC Steam Generator Pressure (psia) 770d Steam Flow Rate (Mlbm/hr) 10.982 Feedwater Temperature | ||
(°F) 435 Number of Active Tubes per Steam 6,986 8 Generator Total Number of Assemblies 204 The Technical Specification minimum flow corresponding to a T;n1et of 532°F is 140. 7 Mlbm/hr. Assumes a 3 % core bypass flow. Maximum allowed inlet temperature at power is 544°F and at HZP T;niet is 532°F. The T;niet LCO sets a separate limit which may be more restrictive. | (°F) 435 Number of Active Tubes per Steam 6,986 8 Generator Total Number of Assemblies 204 The Technical Specification minimum flow corresponding to a T;n1et of 532°F is 140. 7 Mlbm/hr. Assumes a 3 % core bypass flow. Maximum allowed inlet temperature at power is 544°F and at HZP T;niet is 532°F. The T;niet LCO sets a separate limit which may be more restrictive. | ||
This value is for full power. Nominal Steam Generator pressure at HZP is 900 psia . Corresponds to 15% total tube plugging. | This value is for full power. Nominal Steam Generator pressure at HZP is 900 psia . Corresponds to 15% total tube plugging. | ||
| Line 751: | Line 751: | ||
* | * | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 4. References EMF-98-013 Revision 0 Page 4-1 1 . Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG-0800, LWR Edition, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, July 1981 . 2. Advanced Nuclear Fuels Corporation Methodology for Pressurized Water Reactors: | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 4. References EMF-98-013 Revision 0 Page 4-1 1 . Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG-0800, LWR Edition, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, July 1981 . 2. Advanced Nuclear Fuels Corporation Methodology for Pressurized Water Reactors: | ||
Analysis of Chapter 15 Events, ANF-84-73 Revision 4 Appendix B (P)(A) and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, July 1990. 3. Palisades Cycle 9: Analysis of Standard Review Plan Chapter 15 Events, ANF-90-078, Advanced Nuclear Fuels Corporation, September 1990. 4. Palisades Cycle 10: Disposition and Analysis of Standard Review Plan Chapter 15 Events, EMF-91-176, Siemens Nuclear Power Corporation, October 1991. 5. HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, EMF-92-153(P)(A) and Supplement 1, Siemens Power Corporation, March 1994 . 6. Palisades Cycle 14 Principal Plant Parameters, EMF-97-051, RevisiOIJ 1, Siemens Power Corporation. | Analysis of Chapter 15 Events, ANF-84-73 Revision 4 Appendix B (P)(A) and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, July 1990. 3. Palisades Cycle 9: Analysis of Standard Review Plan Chapter 15 Events, ANF-90-078, Advanced Nuclear Fuels Corporation, September 1990. 4. Palisades Cycle 10: Disposition and Analysis of Standard Review Plan Chapter 15 Events, EMF-91-176, Siemens Nuclear Power Corporation, October 1991. 5. HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, EMF-92-153(P)(A) and Supplement 1, Siemens Power Corporation, March 1994 . 6. Palisades Cycle 14 Principal Plant Parameters, EMF-97-051, RevisiOIJ 1, Siemens Power Corporation. | ||
: 7. Not used. 8. Palisades Thermal Margin Monitor Analysis, EMF-95-033(P), Revision 1 , Siemens Power Corporation, April 1995. 9. Palisades Final Safety Analysis Report, Updated Version (through Revision 18), Consumers Power Company. 10. Palisades Cycle 13: Disposition and Analysis of Standard Review Plan Chapter 15 Events, Siemens Power Corporation, November 1996. 11. Palisades loss of load Analysis, EMF-93-086 (P), Siemens Power Corporation | : 7. Not used. 8. Palisades Thermal Margin Monitor Analysis, EMF-95-033(P), Revision 1 , Siemens Power Corporation, April 1995. 9. Palisades Final Safety Analysis Report, Updated Version (through Revision 18), Consumers Power Company. 10. Palisades Cycle 13: Disposition and Analysis of Standard Review Plan Chapter 15 Events, Siemens Power Corporation, November 1996. 11. Palisades loss of load Analysis, EMF-93-086 (P), Siemens Power Corporation | ||
-Nuclear Division, April 1993. 1 2. Palisades Modified Reactor Protection System Report: Analysis of Chapter 15 Events, ANF-87-150(NP), Volume 2, Advanced Nuclear Fuels Corporation, June 1988. 13. Review and Analysis of SRP Chapter 15 Events for Palisades with a 15% Variable High Power Trip Reset, ANF-90-181, Advanced Nuclear Fuels Corporation, November 1990. 14. Palisades Plant Technical Specifications, Consumers Power Company, Appendix A to License No. DPR-20. Siemens Power Corporation | -Nuclear Division, April 1993. 1 2. Palisades Modified Reactor Protection System Report: Analysis of Chapter 15 Events, ANF-87-150(NP), Volume 2, Advanced Nuclear Fuels Corporation, June 1988. 13. Review and Analysis of SRP Chapter 15 Events for Palisades with a 15% Variable High Power Trip Reset, ANF-90-181, Advanced Nuclear Fuels Corporation, November 1990. 14. Palisades Plant Technical Specifications, Consumers Power Company, Appendix A to License No. DPR-20. Siemens Power Corporation | ||
| Line 766: | Line 766: | ||
* | * | ||
* Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events Distribution Controlled Distribution Siemens Power Corporation | * Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events Distribution Controlled Distribution Siemens Power Corporation | ||
-Nuclear Division Richland D. M. Brown, 22 R. C. Gottula, 36 J. S. Holm, 26 J. D. Martin, 31 W. T. Nutt, 36 K. C. Segard, 36 R. I. Wescott, 38/Consumeis | -Nuclear Division Richland D. M. Brown, 22 R. C. Gottula, 36 J. S. Holm, 26 J. D. Martin, 31 W. T. Nutt, 36 K. C. Segard, 36 R. I. Wescott, 38/Consumeis | ||
{ 10) EMF-98-013 Revision 0 | { 10) EMF-98-013 Revision 0 | ||
* ATTACHMENT NO. 2 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 Excerpts from the Palisades Cycle 14 ABB/CE SBLOCA Analysis 3 Pages J | * ATTACHMENT NO. 2 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 Excerpts from the Palisades Cycle 14 ABB/CE SBLOCA Analysis 3 Pages J | ||
| Line 775: | Line 775: | ||
F. Cohen 3/i;}<jg Management Approval J.M. Cleary 3f 1i.,hs I 2. Package Contents (this section may be completed after Management approval): | F. Cohen 3/i;}<jg Management Approval J.M. Cleary 3f 1i.,hs I 2. Package Contents (this section may be completed after Management approval): | ||
Total page count, including body, appendices, attachments, etc.: 68 List associated CD-ROM disk Volume Numbers and path names: 0 None Note: CD-ROM are stored as separate Quality Records CD-ROM Volume Path Names (to lowest directory which uniquely applies to this document) | Total page count, including body, appendices, attachments, etc.: 68 List associated CD-ROM disk Volume Numbers and path names: 0 None Note: CD-ROM are stored as separate Quality Records CD-ROM Volume Path Names (to lowest directory which uniquely applies to this document) | ||
Numbers a_pal_fe\0001 rOO\out Total number of sheets of microfiche: | Numbers a_pal_fe\0001 rOO\out Total number of sheets of microfiche: | ||
[8J None Number of sheets: __ _ Other attachments: | [8J None Number of sheets: __ _ Other attachments: | ||
: 3. Distribution: | : 3. Distribution: | ||
QR(2), JM Cleary jl 1111 , .. ,.,., [8J None Specify: _______________ | QR(2), JM Cleary jl 1111 , .. ,.,., [8J None Specify: _______________ | ||
_ Proprietary Information Combustion Engineering, Inc. | _ Proprietary Information Combustion Engineering, Inc. | ||
* * | * * | ||
* Calculation A-PAL-FE-0001, Rev. 00 Page 8 3. Design Inputs The results and conclusions of this analysis, Section 5, are based upon the following design inputs. 3.1 Plant Design Data The analysis of record for the Palisades SBLOCA ECCS performance analysis is Reference | * Calculation A-PAL-FE-0001, Rev. 00 Page 8 3. Design Inputs The results and conclusions of this analysis, Section 5, are based upon the following design inputs. 3.1 Plant Design Data The analysis of record for the Palisades SBLOCA ECCS performance analysis is Reference | ||
: 4. The changes for Cycle 14 Batch R fuel are documented in Reference | : 4. The changes for Cycle 14 Batch R fuel are documented in Reference | ||
: 3. The changes are shown below: * ** *** 3.2 1. Assembly loss coefficient 15.332* 2. Length of active fuel 11.05 ft** 3. Fuel pellet radius 0.015 ft*** 4. Cladding thickness 0.0020833 ft*** 5. Rod internal gas pressure (cold) 321 psia 6. PCS flow 341400 gpm 7. Core flow 331200 gpm Includes core support plate and fuel alignment plate. More details for this value are provided in Reference | : 3. The changes are shown below: * ** *** 3.2 1. Assembly loss coefficient 15.332* 2. Length of active fuel 11.05 ft** 3. Fuel pellet radius 0.015 ft*** 4. Cladding thickness 0.0020833 ft*** 5. Rod internal gas pressure (cold) 321 psia 6. PCS flow 341400 gpm 7. Core flow 331200 gpm Includes core support plate and fuel alignment plate. More details for this value are provided in Reference | ||
: 16. Top of the fuel will be the same as the previous cycle. Extra length is added to the bottom. Cladding thickness is reduced by increasing the cladding ID. Cladding OD remains the same. Fuel pellet radius is increased so that the gap between the cladding and fuel remains the same. Computer Codes This recorded calculation uses the computer codes listed below. The computer code certificates are included in the references listed below. 1. CEFLASH-4AS jl It It ,.,1*** Version: Computer: | : 16. Top of the fuel will be the same as the previous cycle. Extra length is added to the bottom. Cladding thickness is reduced by increasing the cladding ID. Cladding OD remains the same. Fuel pellet radius is increased so that the gap between the cladding and fuel remains the same. Computer Codes This recorded calculation uses the computer codes listed below. The computer code certificates are included in the references listed below. 1. CEFLASH-4AS jl It It ,.,1*** Version: Computer: | ||
| Line 791: | Line 791: | ||
F4S.l.2 HP/9000 HP-UX OS PA-RISC Reference 8 Reference 9 Proprietary Information Combustion Engineering, Inc. | F4S.l.2 HP/9000 HP-UX OS PA-RISC Reference 8 Reference 9 Proprietary Information Combustion Engineering, Inc. | ||
* | * | ||
* L Calculation A-PAL-FE-0001, Rev. 00 Page 45 5. Results and Conclusions The results and conclusions of this analysis are based upon the analysis presented in Section 4. 5.1 Summary of Results A summary of the PARCH/EM results for PCT for this analysis is provided in the ti.owing table: Reference 4 Current Analysis (Pages 209 and 211) Break Size (ft 2) at PD 0.10 0.10 Peak Clad Temperature (PCT) (°F) 1991.9 2025.8 Time to PCT (seconds) 1606.2 1520.2 PCT Node 20 20 Peak Oxidation | * L Calculation A-PAL-FE-0001, Rev. 00 Page 45 5. Results and Conclusions The results and conclusions of this analysis are based upon the analysis presented in Section 4. 5.1 Summary of Results A summary of the PARCH/EM results for PCT for this analysis is provided in the ti.owing table: Reference 4 Current Analysis (Pages 209 and 211) Break Size (ft 2) at PD 0.10 0.10 Peak Clad Temperature (PCT) (°F) 1991.9 2025.8 Time to PCT (seconds) 1606.2 1520.2 PCT Node 20 20 Peak Oxidation | ||
(%) 4.79 5.8 Peak Oxidation Node 19 19 Core Wide Oxidation(%) | (%) 4.79 5.8 Peak Oxidation Node 19 19 Core Wide Oxidation(%) | ||
< 0.71 < 0.842 PLHGR (Kw/ft) 15.8 15.8 Detailed results of the analysis are contained in the output of the computer cases. The cases (inputs and outputs) are archived on the ABB CE CD-ROM archiving system at the path shown on the cover page. Table 5-1 summarizes important parameters used in the Palisades small break LOCA ECCS performance analysis. | < 0.71 < 0.842 PLHGR (Kw/ft) 15.8 15.8 Detailed results of the analysis are contained in the output of the computer cases. The cases (inputs and outputs) are archived on the ABB CE CD-ROM archiving system at the path shown on the cover page. Table 5-1 summarizes important parameters used in the Palisades small break LOCA ECCS performance analysis. | ||
Revision as of 19:25, 25 April 2019
| ML18066A341 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 04/22/1998 |
| From: | SEGARD K C SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML18066A340 | List: |
| References | |
| EMF-98-013, EMF-98-013-R00, EMF-98-13, EMF-98-13-R, NUDOCS 9812080079 | |
| Download: ML18066A341 (102) | |
Text
- ATTACHMENT NO. 1 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 Siemens Power Corporation Report EMF-98-013 Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 15 Events May, 1998 ----------9012000079 981202 i,, ,' PDR ADOCK 05000255:
p PDR -'-"'
SIEMENS EMF-98-013 Revision 0 Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events May 1998 Siemens Power Corporation Nuclear Division
- ISSUED IN SPC-ND ON-LINE Siemens Power Corporation
-Nuclear Division DOCUME'tfI-$,STEr . DATE: . Lf' T ¥' Prepared:
eel Issue Date: Palisades Cycle 14: Disposition and Analysis of Standard Review Plan K. C. Segard, Engine PWR Safety Analys* . Chapter 15 Events Date ( I EMF-98-013 Revision 0
- Customer Disclaimer Important Notice Regarding Contents and Use of This Document Please Read Carefully Siemens Power Corporation's warranties and representations concerning the subject matter of this document are those set forth in the agreement between Siemens Power Corporation and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly provided in such agreement, neither Siemens Power Corporation nor any person acting on its behalf: a. makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method or process disclosed in this document will not infringe privately owned rights; or b. assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this document.
The information contained herein is for the sole use of the Customer.
In order to avoid impairment of rights of Siemens Power Corporation in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term) of such information until so authorized in writing by Siemens Power Corporation or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless expressly provided in the Agreement.
No rights or licenses in or to any patents are implied by the furnishing of this document *
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan EMF-98-013 Revision 0 Page i Chapter 1 5 Events Contents 1. lntroduction
....................................................................................................
1-1 1.1 Cycle14Changes
........................................
- ........................................
1-1 1 . 1 . 1 Plant Operating Conditions
.........................................................
1-1 1.1.2 Plant Control System .................................................................
1-2 1.1.3 Plant Safety Systems ................................................................
1-2 1 .1 .4 Thermal-Hydraulic Characteristics
...............................................
1-3 1 . 1 . 5 Fuel Rod Mechanical Design .......................................................
1 -3 1.1.6 Neutronics Design Changes ........................................................
1-3 1 . 1 . 7 Uncertainties and Response Times ................................
- .............
1-3 1 . 2 Analytical Considerations
.......................................................................
1-4 2. Summary and Conclusions
................................................................................
2-1 3. 2.1 Fuel Centerline Melt ..............................................................................
2-1 2.2 Steamline Break Analysis .......................................................................
2-1 2.3 Control Rod Ejection Analysis .................................................................
2-1 2.4 LOCA Analysis .....................................................................................
2-2 2.5 Event Review .......................................................................................
2-2 2.6 MDNBR, Fuel Centerline Melt and System Pressure Results .......................
2-2 2. 7 TM/LP and T;niet LCO Verification
.............................................................
2-2 Disposition And Analysis Of Plant Events .....................................
- ....................
3-1 3.1 Thermal Hydraulic Compatibility
.............................................................
3-2 3.2 DNB Propagation
..................................................................................
3-2 3.3 Rod Bow ..............................................................................................
3-2 3.4 Control Room Habitability
......................................................................
3-2 3.5 Fuel Centerline Melt ................................................................................
3-3 15.0 Accident Analyses ................................................................................
3-3 15.0.1 Categorization of Plant Events ...................................................
3-3 15.0.2 Plant Characteristics and Initial Conditions
.................................
3-4 15.0.3 Power Distribution
...*...............................................................
3-4 15.0.4 Range of Plant Operating Parameters and States .........................
3-5 15.0.5 Reactivity Coefficients Used in the Safety Analysis .....................
3-5 15.0.6 Scram Insertion Characteristics
.................................................
3-5 15.0.7 Reactor Protection System Trip Setpoints and Time Delays ..........
3-6 15.0.8 Component Capacities and Setpoints
.........................................
3-9 15.0.9 Plant Systems and Components Available for Mitigation of Accident Effects ......................................................................
3-9 15.0.10 Plant Licensing Basis and Single Failure Criteria ..........................
3-9 15.1 Increase in Heat Removal by the Secondary System ...............................
3-11 15.1.1 15.1.2 15.1.3 15.1 .4 15.1.5 15.2 15.2.1 Decrease in Feedwater_Temperature
........................................
3-11 Increase in Feed water Flow ....................................................
3-11 Increase in Steam Flow ..........................................................
3-1 2 Inadvertent Opening of a Steam Generator Relief or Safety Valve .........................................................................
3-1 3 Steam System Piping Failures Inside and Outside of Containment
.........................................................................
3-14 Decrease in Heat Removal by the Secondary System ................
3-14 Loss of External Load ..............................................................
3-1 4 Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan EMF-98-013 Revision 0 Page ii Chapter 1 5 Events 15.2.2 Turbine Trip ...........................................................................
3-17 15.2.3 Loss of Condenser Vacuum ......................................................
3-18 15.2.4 Closure of the Main Steam Isolation Valves (MSIV) ....................
3-18 15.2.5 Steam Pressure Regulator Failure ..............................................
3-19 Loss of Non-Emergency A.C. Power to the Station Auxiliaries
...... 3-19 15.2.7 Loss of Normal Feedwater Flow ...............................................
3-20 15.2.8 Feedwater System Pipe Breaks Inside and Outside Containment
.. 3-21 1 5 .3 Decrease in Reactor Coolant System Flow .............................................
3-22 15.3.1 Loss of Forced Reactor Coolant Flow ........................................
3-22 15.3.2 Flow Controller Malfunction
.....................................................
3-24 15.3.3 Reactor Coolant Pump Rotor Seizure .........................................
3-24 1 5. 3 .4 Reactor Coolant Pump Shaft Break ...........................................
3-25 15.4 Reactivity and Power Distribution Anomalies
..............................
- ..........
3-25 1 5 .4. 1 Uncontrolled Control Assembly Withdrawal from a Subcritical or Low Power Startup Condition
.*.................................................
3-25 15.4.2 Uncontrolled Control Bank Withdrawal at Power ........................
3-28 15.4.3 Control Rod Misoperation
.........................................................
3-30 15.4.4 Startup of an Inactive Loop ......................................................
3-35 15.4.5 Flow Controller Malfunction
.....................................................
3-35 15.4.6 CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant ......................
1 .................
3-35 1 5 .4. 7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position ....................................................................
3-36 15.4.8 Spectrum of Control Rod Ejection Accidents
..............................
3-37 15.4.9 Spectrum of Rod Drop Accidents (BWR) ....................................
3-37 15.5 Increases in Reactor Coolant System Inventory
......................................
3-37 1 5. 5. 1 Inadvertent Operation of the ECCS that Increases Reactor Coolant Inventory
...................................................................
3-37 15.5.2 CVCS Malfunction that Increases Reactor Coolant Inventory
....... 3-38 1 5. 6 Decreases in Reactor Coolant ...............................................................
3-38 15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve .. 3-38 15.6.2 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment
...................................
3-39 15.6.3 Radiological Consequences of Steam Generator Tube Failure ....... 3-40 15.6.4 Radiological Consequences of a Main Steam Line Failure Outside Containment (BWR) ...........................*.........................
3-40 15.6.5 Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary ....... 3-41 1 5. 7 Radioactive Releases from a Subsystem or Component
...........................
3-41 1 5. 7. 1 Waste Gas System Failure .......................................................
3-41 15. 7 .2 Radioactive Liquid Waste System Leak or Failure (Release to Atmosphere)
..........................................................................
3-41 1 5. 7. 3 Postulated Radioactive Releases Due to Liquid-Containing Tank Failures ..........................................................................
3-41 15.7.4 Radiological Consequences of Fuel Handling Accident .................
3-41 15.7.5 Spent Fuel Cask Drop Accidents
...............................................
3-42 4. References
.....................................................................................................
4-1
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan EMF-98-013 Revision 0 Page iii Chapter 1 5 Events 2.1 2.2 15.0.1.1 15.0.2.1 15.0.2.2 15.0.2.3 15.0.3.1 15.0.4.1 15.0.5.1.
15.0.7.1 15.0.7.2 15.0.8.1 15.0.9.1 15.2.1.1 15.4.3.1 15.0.3.1 15.0.3.2 15.0.6.1 15.0.7.1 15.0.7.2 15.0.7.3 15.2.1.1 15.2.1.2 15.2.1.3 15.2.1.4 15 .. 2.1.5 Tables Disposition of Events Summary for Palisades Cycle 14 ..............................
2-3 Summary of Analysis Result ....................................................................
2-7 Accident Category Used for Each Analyzed Event ...................................
3-43 Plant Operational Conditions
.................................................................
3-4 7 Nominal Plant Operating Conditions
.......................................................
3-48 Nominal Reload R Fuel Design Parameters
..............................................
3-49 Core Power Distribution
......................................................................
- .. 3-50 Range of Key Initial Condition Operating Parameters
................................
3-51 Reactivity Parameters
...........................................................................
3-5 2 Trip Setpoints for Operation at 2530 MWt .............................................
3-53 TM/LP Uncertainties
...... : ......................................................................
3-54 Component Capacities and Setpoints
.....................................................
3-55 Overview of Plant Systems and Equipment Available for Transient and Accident Conditions
.............................................................................
3-56 Sequence of Events for the Loss of External Load Secondary Side Pressurization
......................................................................................
3-60 Summary of Results for Control Rod Misoperation Events ........................
3-61 Figures Limiting Axial Power Shape (100% Power) ....................... , .....................
3-62 Limiting Axial Power Shape (90% Power) ...............................................
3-63 Integrated Scram Worth With Most Reactive Rod Stuck Out ....................
3-64 TrNLEr Limiting Condition of Operation
......................................................
3-65 Axial Shape Function for TM/LP Trip ......................................................
3-66 Radial Function for TM/LP .....................................................................
3-67 Loss of External Load Secondary Side Pressurization
-Pressurizer Pressure ..............................................................................
- ..............
3-68 Loss of External Load Secondary Side Pressurization
-Core Inlet and Steam Generator Primary side Exit Temperatures
....................................
3-69 Loss of External Load Secondary Side Pressurization-Secondary Steam Dome w 2wwSide Pressure .........................................................
3-70 Loss of External Load Secondary Side Pressurization-MSSV Flow Rate for Loop 1 ................................................................................................
3-71 Loss of External Load Secondary Side Pressurization-MSSV Flow Rate for Loop 2 ................................................................................................
3-72 Siemens Flower Corporation_-
Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 1 . Introduction EMF-98-013 Revision 0 Page 1-1 This report documents the results of the disposition and analysis of the FSAR Section 14 events, which are the same events described in Chapter 15 of the Standard Re.view Plan (SRP), in support of Palisades Cycle 14 operation with reload fuel supplied by Siemens Power Corporation
-Nuclear Division (SPC). The events were dis positioned in accordance with SPC's approved methodology (Reference 18). Averification of the TM/LP and Tinlet LCO function was performed for Cycle 14 along with an evaluation of the MDNBR and maximum LHGR for the applicable SRP Chapter 1 5 events. Section 2.0 presents a summary of results of this disposition and analysis.
Section 3.0 presents the conditions employed in the analysis of events and discussion of the event disposition and MDNBR results for the SRP Chapter 15 events. The events are numbered in accordance with the SRP to facilitate review. 1 . 1 Cycle 14 Changes The changes for Cycle 14 that impact the disposition of the Chapter 1 5 events and the verification of the setpoints, MDNBR and fuel centerline melt (FCM) analyses consist of the following: 1 . 1 . 1 Plant Operating Conditions The changes described below were implemented in the Plant Parameters Document {PPD) (Reference 6). 1. Maximum nominal T;n1et of 544°F was added. 2. The nominal steam generator flow decreased from 11.57 Mlbm/hr to 10.982 Mlbm/hr. 3. The nominal HZP steam generator pressure of 900 psia was added. 4. The HFP and HZP steam generator masses were changed. a. The HFP liquid mass increased from 120,385 lbm to 133,593 lbm and the vapor mass decreased from 8, 779 lbm to 8,545 lbm. b. The HZP liquid mass of 203,783 lbm and the HZP vapor mass of 6,976 lbm were added. 5. The maximum steam generator tube plugging was assumed to be 1 5 % for analysis purposes.
Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan EMF-98-013 Revision 0 Page 1-2 Chapter 1 5 Events 6. The shutdown margin to be supported with less than four reactor pumps operating was reduced from 3.75% to 3.5%. 1 . 1 . 2 Plant Control System The changes described below were implemented in the PPD. 1 . The delay time for the opening or closing of the normal feedwater regulating valves incre3sed from 20.5 sec to 22 sec. 2. The letdown capacity increased to 120 gpm. 3. The minimum charging pump capacity of 33 gpm. 4. The maximum pressurizer spray flow rate increased from 280 gpm to 500 gpm. 1 . 1 .3 Plant Safety Systems The changes described below were implemented in the PPD. 1. 2. 3. 4. 5. 6. 7. 8. 9. 10. 11 . The pressurizer pressure for actuation of the SI decreased from an allowable value of 1,593 psig to an allowable value of 1,590 psig . The LPSI pump performance includes allowances for degradation over the range of operating pressures.
A reactor trip setpoint on a containment high-pressure of S 3.7 psig was added. The low steam generator level reactor trip has been changed to an allowable value of 18.77% of the narrow range reading. The low steam generator level trip for AFW actuation has been changed to an allowable value of 25.9% of the narrow range reading. SI actuation on high containment pressure was given an allowable range of 3. 7 psig to 4.3 psig. The MSSV capacities were made consistent with design capacities.
The MSSV opening pressure and flow data are now staged by bank. The steam assisted MSIV closure time of 2 sec, consisting of one second for signal processing delay and one second for valve closure. The single flow rate value (2, 125 gpm) for the containment sprays was replaced with a table of values that are based on the failure assumptions.
The minimum, maximum and nominal flow rates and the startup delay times for the. AFWs were added. The SIT water volume was increased from 2,000 ft 3 to 2,011 ft 2* Siemens Power Corporation
-Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 1 .1 .4 Thermal-Hydraulic Characteristics EMF-98-013 Revision 0 Page 1-3 1 . The lower tie plate de-sign changed for Reload R. These tie plates have a FUELGUARDTM debris-resistant insert. 2. Addition of an all inconel High Thermal Performance (HTPJ bottom spacer. 3. New pressure drop measurements were performed with the control blades fully inserted. 1 . 1 . 5 Fuel Rod Mechanical Design A new fuel design will be introduced for Reload R, incorporating the following changes: a. 0.009-inch increase in the pellet OD. b. 0.152-inch increase in the pellet length. c. 0.4% reduction in the pellet dish volume. d. 0.009-inch increase in the cladding ID. e. 0.8-inch increase in the fuel rod active length. f. 0. 111-inch increase in the effective plenum length. g. Initial rod internal pressure of the gadolinia bearing rods has increased.
1 .1 .6 Neutronics Design Changes 1 . The enrichment pattern for the Reload R assemblies was modified and it produces different local power distributions.
- 2. The U0 2 enrichment in the gadolinia bearing fuel rods increased.
- 3. The maximum gadolinia concentration in the gadolinia-bearing fuel rods increased in Cycle 14 from 6 wt% to 8 wt%. 4. The U 238 capture-fission ratios for Cycle 14 increased slightly, resulting in an increase in the decay heat. 5. The 50% and 100% power differential rod worths at EOC conditions increased.
- 6. The moderator density and Doppler reactivity tables changed for Cycle 14. 7. The boron endpoint concentrations for Cyc;=le 14 changed for all conditions.
1 .1 . 7 Uncertainties and Response Times 1 . Cycle 14 does not contain a full core of fresh detectors.
- 2. Delay times for the HPSI and LPSI pumps were increased from 27 sec to 40 sec. 3. RTD delay times were increased from 12 to 15 sec. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 1-4 4. The trip delays for the low reactor coolant flow, high pressurizer pressure , low pressurizer pressure, low steam generator pressure and TM/LP and VHP trips were increased by 0.2 sec. 1 . 2 Analytical Considerations The Technical Specification radial peaking factor limits for Reload R, an of 1 . 76 and an F; of 2.04, were unchanged from those for Reloads 0, P and Q. Radial peaking factors for reloads prior to Reload 0 were unchanged for Cycle 14 (References 3 and 4). _The Technical Specification radial peaking limits for all fuel loaded in Cycle 14 are given in Table 15.0.3.1.
The minimum departure from nucleate boiling ratio (MDNBR) for each event made use of the HTP departure from nucleate boiling correlation (Reference 5). A 2% mixed core DNB penalty {Reference
- 22) was applied in Cycle 14. The impact of the changes in the power distributions, both radial and axial, resulting from the modified enrichment distribution and the Cycle 14 loading plan on MDNBRs was considered. . The design of the gadolinia rods for Cycle 14 utilized five axial zones: two axial enriched blankets (2.15 wt% U 235), two axial regions with no gadolinia (4.02 wt% U 235) and a central gadolinia-bearing zone with 3.99 wt% U 235 and a maximum gadolinia concentration of 8 wt%. The U0 2 fuel rod design for Cycle 14 also made use of enriched axial blankets at the top and bottom. The active fuel length of the fuel rod increased and the fuel pellet diameter also increased for Cycle 14. The effects of these designs characteristics were considered in the generation of the neutronics parameters used in the Cycle 14 event review. The fuel centerline melt {FCM) limit was re-calculated for Cycle 14 because of the higher U 235 enrichments in the gadolinia-bearing rods and the mechanical design changes . e">! _____ ----""-------=--
- -*---
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 2. Summary and Conclusions EMF-98-013 Revision 0 Page 2-1 The disposition of SRP Chapter 15 events for Palisades, Cycle 14, the setpoint verification, FCM and MDNBR analyses consider the impact of several changes in fuel design and plant operation.
The discussion of these analyses with regard to the changes in Section 1.1 are summarized below. 2. 1 Fuel Centerline Melt To account for the new fuel rod mechanical design introduced into Cycle 14, the maximum allowed LHGR on a U0 2 rod to preclude FCM was calculated.
The LHGR limit on U0 2 rods for Cycle 14 was determined to be 20.93 kW/ft. This limit precludes FCM on both U0 2 and gadolinia rods. 2.2 Steamline Break Analysis The Reference 3 analysis of the Main Steamline Break (MSLB) assumed that the break in the main steamline occurred inside containment.
The case analyzed was a HFP case with a loss of offsite power and a subsequent turbine generator assisted coastdown of the primary coolant pumps (PCPs) which was identified based on a different methodology.
The reanalysis of the MSLB was performed for the following reasons: 1. To extend the analysis from a single, limiting case to the standard set of cases typically evaluated for an MSLB event. 2. To evaluate breaks both inside and outside containment.
- 3. To evaluate the impact of tripping the PCPs at break initiation for the loss of offsite power case, instead of modeling a time delay to simulate the turbine generator assisted PCP coastdown.
- 4. To evaluate for the HZP cases, the impact of the initial valve position (open or closed) of the MSIVs and of the MSIV bypass valves. 5. To upgrade the ANF-RELAP model to be consistent with the Cycle 14 plant configuration and conditions.
The results of this analysis are presented in Reference
- 23. 2. 3 Control Rod Ejection Analysis The control rod ejection analysis was performed for Cycle 14 to remove conservatism from the system analysis.
The ANF-RELAP model for this analysis was made consistent with the Cycle 14 plant configurations and operating conditions.
Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 2-2 The system transient analysis and fuel failure calculations were performed assuming no loss of off site power. The results are presented in Reference
- 24. 2.4 LOCA Analysis Based on all of the Cycle 14 changes which affect LBLOCA, the LBLOCA event was reanalyzed 1211* These changes include Items 4 and 5 in Section 1.1.1, Items 2, 3, 5, 8, and 11 in SectiOn 1.1.3, all items in Section 1.1.4 and 1.1.5, Items 1, 2, 3, 4, and 6 in Section 1. 1 . 6 and Item 2 of Section 1 . 1 . 7. The LBLOCA analysis also addressed the increased gad rod enrichment.
2.5 Event
Review A summary of the event review for Palisades Cycle 14 is given in Table 2. 1. This table lists each SRP Chapter 15 event, indicates whether that event is re-analyzed for Cycle 14, provides a reference to the bounding event or analysis of record for events not analyzed, and provides a cross reference between the SRP Chapter 15 event numbers and the Palisades Updated FSAR (Reference 9). 2. 6 MDNBR, Fuel Centerline Melt and System Pressure Results The LBLOCA, MSLB, and Control Rod Ejection events were analyzed for Cycle 14 with the results reported in References 21, 23, and 24 respectively.
The dispositions and analyses reported herein confirm, for the remaining SRP Chapter 15 events, that event acceptance criteria (Section 15.0.10) are met for the Cycle 14 operation and support operation with up to 1 5 % steam generator tube plugging in either or both steam generators at a rated thermal power of 2,530 MWt. A summary of the MDNBR, FCM and system pressure results, along with the acceptance criteria, is given in Table 2.2. The thermal margins for Cycle 14 are based on the full core two-pass Cycle 14 XCOBRA-lltC model and the HTP DNBR correlation with a 2% mixed penalty applied. 2. 7 TM/LP and T.., LCO Verification A setpoint analysis was performed which verified that the TM/LP and T;ntet LCO setpoints are acceptable for Cycle 14 . Siemens Power Corporation
- Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events :? 1'.{. .:{. :::; ..... :{. EMF-98-013 Revision 0 Page 2-3 Table 2.1 Disposition of Events Summary for Palisades Cycle 14 SRP Bounding Updated Event Event or FSAR Designation Event Name DisEosition Reference Designation 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 Decrease in Feedwater Temperature Bounded 15.1.3 a 15.1.2 Increase in Feedwater Flow a 1) Power Operation Bounded 15.1.3 2) Startup Bounded 15.1.3 15.1.3 Increase in Steam Flow Bounded Ref. 10 14.10 Analyzed MDNBR 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve 1) Power Operation Bounded 15.1.3 2) Startup Bounded 15.4.1 15.1.5 Steam System Piping Failures Inside Analyzed Ref. 23 14.14 and Outside of Containment 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.1 Loss of External Load 1) Primary Over-pressurization Bounded Ref. 11 14.12 2) Secondary Over-pressurization Analyzed 14.12 3) MDNBR Analyzed 14.12 15.2.2 Turbine Trip Bounded 15.2.1 15.2.3 Loss of Condenser Vacuum Bounded 15.2.1 15.2.4 Closure of the Main Steam Isolation Bounded 15.2.1 Valves (MSIV) 15.2.5 Steam Pressure Regulator Failure Not Applicable; BWR Event 15.2.6 Loss of Non-Emergency A.C. Power to the Station Auxiliaries
- 1) DNB Bounded 15.3.1 2) Long Term cooling Bounded 15.2.7 15.2.7 Loss of Normal Feedwater Flow 1) Maximum PCS pressure Bounded 15.2.1 14.13 2) Maximum Primary to secondary Boun.cled Ref. 12 14.13 pressure difference
- 3) Minimum steam generator inventory Bounded Ref. 12 Deleted from the FSAR Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 2-4 Table 2.1 Disposition of Events Summary for Palisades Cycle 14 (continued)
SRP Bounding Updated Event Event or FSAR Designation Event Name Disposition Reference Designation 15.2.8 Feedwater Syi>tem Pipe Breaks Inside and Outside Containment.
- 1) Cooldown Bounded 15.1.5 2) Heatup Bounded 15.2.1 3) Long Term Cooling Bounded 15.2.7 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW 15.3.1 Loss of Forced Reactor Bounded Ref. 3 14.7 Coolant Flow Analyzed MDNBR 15.3.2 Flow Controller Malfunction Not Applicable 14.7 15.3.3 Reactor Coolant Pump Rotor Bounded Ref. 3 14.7 Seizure Analyzed MDNBR 15.3.4 Reactor Coolant Pump Shaft Bounded 15.3.3 14.7 Break 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled Control Bank Calculated MDNBR and maximum 14.2.1 Withdrawal from a fuel temperature Subcritical or Low Power Condition 15.4.2 Uncontrolled Control Bank Calculated MDNBR and maximum 14.2.2 Withdrawal at Power LHGR Operation Conditions 15.4.3 Control Rod Misoperation
- 1) Dropped Control Analyzed MDNBR Ref. 3 14.4 Bank/Rod 2) Dropped Part-Length Bounded 15.4.3(1) 14.6 Control Rod 3) Malpositioning of the Not Applicable Part-Length Control __ Group 4) Statically Misaligned Bounded 15.4.3(1) 14.6 Control
... ... .----** -5) Single Control Rod Analyzed MDNBR Ref, 10 14.2.3 Withdrawal and maximum LHGR 6) Core Barrel Failure Bounded 15.4.8 14.5 Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 2-5 Table 2.1 Disposition of Events Summary for Palisades Cycle 14 (continued)
SRP Event Designation Event Name 1 5.4.4 Startup of an Inactive Loop 15.4.5 Flow Controller Malfunction 15.4.6 CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant 1) Ra.ted and Power Operation Conditions
- 2) Reactor Critical, Hot Standby and Hot Shutdown 3) Refueling Shutdown Condition, Cold Shutdown Condition and Refueling Operation Disposition Bounded by rated power MDNBR. Bounding Event or Updated FSAR Reference Designation 14.8 Not Applicable:
No Flow Controller Bounded Ref. 26 14.3 Bounded Ref. 26 14.3 Bounded Ref. 26 14.3 15.4. 7 Inadvertent Loading and Operation Administrative procedures preclude this of a Fuel Assembly in an Improper event Position 15.4.8 Spectrum of Control Rod Ejection Analyzed Accidents 15.4.9 Spectrum of Rod Drop Accidents Not Applicable; BWR Event (BWR) 15.5 INCREASES IN REACTOR COOLANT INVENTORY 15.5.1 15.5.2 Inadvertent Operation of the ECCS During Power Operations CVCS Malfunction that Increases Reactor Coolant Inventory Siemens Power Corporation
- Nuclear Division Precluded by System Pressure Overpressure:
Bounded Reactivity:
Bounded 15.2.1 15.4.6 14.16
- -Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 2-6 Table 2.1 Disposition of Events Summary for Palisades Cycle 14 {continued) b SAP Event Designation Event Name DisQosition 15.6 DECREASES IN REACTOR COOLANT INVENTORY 15.6.1 Inadvertent Opening of a PWR Analyzed MDNBR Pressurizer Pressure ReHef Valve 15.6.2 Radiological Consequences of the Bounded failure of Small Lines Carrying Primary Coolant Outside of Containment 15.6.3 Radiological Consequences of Steam Note a Generator Tube Failure 1) MDNBR Bounded 2) Radiological Consequences Bounded 15.6.4 Radiological Consequences of a Main Not Applicable; Steam Line Failure Outside BWR Event Containment 15.6.5 Loss of Coolant Accidents Resulting LBLOCA from a Spectrum of Postulated Piping Analyzedb Breaks Within the Reactor Coolant Pressure Boundary 15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.1 Waste Gas System Failure Bounded 15.7.2 Radioactive Liquid Waste System Bounded Leak or Failure (Release to Atmosphere) 15.7.3 Postulated Radioactive Releases due Bounded to Liquid-Containing Tank Failures 15.7.4 Radiological Consequences of Fuel Bounded Handling Accidents 15.7.5 Spent Fuel Cask Drop Accidents Bounded The Steam Generator Tube Rupture event is not analyzed by SPC. The SBLOCA event is not analyzed by SPC. Siemens Power Car oration -Nuclear Division Bounding Updated Event or FSAR Reference Designation Ref. 3 15.6.5 14.23 15.6.1 14.15 Ref. 9 14.17 Ref. 9 14.21 Ref. 9 14.20 Ref. 9 14.20 Ref. 9 14.19 Ref. 9 11
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan EMF-98-013 Revision 0 Page 2-7 Chapter 1 5 Events b c d
- g Table 2.2 Summary of Analysis Result Pea kb Maximum Event MDNBR" LHR (kW/ft) Pressurizer Pressure (Qsial 1 5.1.3 Increase in Steam Flow 1.89 16.99 2,033.1 15.1.5 Steam System Piping Failures Inside 1.31 20.8 Note c and Outside of Containment 15.2.1 Loss of External Load 2,614.Sd 15.2.7 Loss of Normal Feedwater 2,271.9 15.3.1 Loss of Forced Reactor Coolant 1.46 15.59 2, 127.8 Flow 15.3.3 Reactor Coolant Pump Rotor Seizure 1.41 15.98 2, 145.2 15.4.1 Uncontrolled Control Bank 4.47 Note e 2,161.0 Withdrawal from Subcritical or Low Power 15.4.2 Uncontrolled Control Bank 1.70 17.56 2,267.1 Withdrawal at Power 1 5.4.3 Control Rod Misoperation
- Dropped Rod 1.62 17.59 2,010.0
- Dropped Bank' 1.50 18.88 2,010.0
- Single Rod 1.27 19.67 2,267.1 15.4.6 CVCS Malfunction Resulting in Adequacy of Shutdown Margin is demonstrated Decreased Boron Concentration 15.4.8 Control Rod Ejection 1.63 Note g 2,217.2 15.6.1 Inadvertent Opening of a PWR 1.83 16.34 2,110.1 Pressurizer Pressure Relief Valve MDNBRs are based on the HTP correlation (95/95 limit = 1.164 including a 2% mixed core penalty).
The FCM limit for Cycle 14 was calculated to be 20.93 kW/ft for U0 2 rods. This value precludes melting of any fuel rod (with or without gadolinia).
This is depressurization event and the pressurizer pressures drops below 1500 psia. The maximum secondary side pressure for the secondary side overpressurization case is 1,063.3 psia . The peak fuel centerline temperature was evaluated for Cycle 14 for this event and found to be 1326°F, which is well below the minimum melt temperature for any of the fuel rods in Reload R. The transient simulation for the dropped bank did not take credit for the VHP trip and the calculated DNBR is conservative.
The peak fuel centerline temperature was evaluated for Cycle 14 for this event and found to be 2873°F, which is well below the maximum melt temperature of 4621°F for any of the fuel rods in Reload R. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 2-8 Table 2.2 Summary of Analysis Results (continued)
Event Dis12osition/Results 15.6.3 Radiological Consequences of Steam Generator CPCo evaluates this event Tube Failure 15.6.5 Loss of Coolant Accidents Resulting from a See Reference 21 Spectrum of Postulated Piping Breaks Within the Reactor Coolant* Pressure Boundary 15. 7.1 Waste Gas System Failure CPCo evaluates this event 15.7.3 Postulated Radioactive Releases due to Liquid-CPCo evaluates this event Containing Tank Failures 15.7.4 Radiological Consequences of Fuel Handling CPCo evaluates this event Accidents 15.7.5 Spent Fuel Cask Drop Accidents CPCo evaluates this event Siemens Power Corooration
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 3. Disposition And Analysis Of Plant Events EMF-98-013 Revision 0 Page 3-1 This section provides the disposition of the Chapter 1 5 events of the SRP and analyses performed to support Cycle 14 operation.
Event numbering and nomenclature are consistent with Chapter 1 5 of the SRP to facilitate review. This section provides information on the plant licensing basis including classification of plant conditions, classification of accident events by category, operating conditions, initial conditions, neutronics data, core and fuel design parameters.
Listings of systems and components available for accident mitigation, trip setpoints, time delays and component capacities are also included.
A non-LOCA system transient analysis for most SRP Chapter 15 events was performed for Cycle 9 (Reference 3). The events not considered in the Reference 3 analysis were the uncontrolled rod withdrawal from part power, which was analyzed and reported in Reference 13, and a loss of Normal Feedwater, which was analyzed and reported in Reference
- 12. A system analysis for uncontrolled rod withdrawal from a subcritical or low power condition was performed as a part of the support for Cycle 13 (Referer.ice 10). Subsequent to the Cycle 9 analysis, a system analysis was performed for the Loss of External Load event with two cases; one to determine the maximum primary pressure and one to minimize the DNBR (Reference 11 ). This event was analyzed for Cycle 14 to produce the highest secondary pressure.
The results of this analysis are provided in this report. The changes introduced in Cycle 14 do not affect the system response for the events reported in References 3, 10, 11, 12 and 13. Thus, the system thermal-hydraulic responses for the various transients reported in References 3, 10, 11, 12, and 13 remain applicable for Cycle 14. MDNBR values for limiting AOOs and Postulated Accidents (PAs), as evaluated with the HTP DNB correlation, were calculated for Cycle 14. An XCOBRA-lllC model was developed for Cycle 14. This model incorporated the radial and axial power distributions from Cycle 14 as well as the hydraulic changes to the fuel assemblies.
This model was applied to all DNB event analyses for Cycle 14. Additional evaluations were performed for Cycle 14 that do not explicitly fall under the SRP Chapter 1 5 event review. These include; thermal-hydraulic compatibility, control Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-2 room habitability, DNB propagation, rod bow and FCM. The disposition of these items is given below: 3. 1 Thermal Hydraulic Compatibility The Reload R fuel to be loaded into Cycle 14 is identical in thermal-hydraulic behavior to the co-resident fuel except for the addition of an all inconel bottom spacer and a resistant lower tie plate. Hydraulic testing (Reference
- 25) showed different pressure loss coefficients for the HTP fuel design over what has previously been used. These new loss coefficients were incorporated into the Cycle 14 XCOBRA-lllC models. An XCOBRA-lllC model was developed for the detailed core model (non-LOCA transients) and for the reduced channel model (setpoints).
These models were used to reanalyze the MDNBR in support of operation for Cycle 14. In addition, a 2 % penalty was applied to the HTP DNBR limit to account for the mixture of hydraulic types in the core. 3. 2 DNB Propagation DNB propagation is most likely for conditions with high rod exposure and low system pressure.
Event 15.1.5 is the most limiting event for DNB propagation but since there were no DNB related fuel failures in the MSLB analysis there was no need to consider DNB propagation (Reference 23). 3.3 RodBow Due to a reduction in cladding thickness for the Reload R fuel, a Rod Bow analysis was performed per Reference
- 27. No DNB penalty is required until 50% gap closure, which does not occur until assembly exposures of 66 GWd/MTU. This is far above the maximum assembly burnup of 53,281 MWd/MTU projected for Cycle 14. FCM is unaffected because no rod bow penalty on Fa is required for burnups below 54,000 MWd/MTU. 3 .4 Control Room Habitability Control room habitability following a major accident can be affected by changes which increase the amount of radioactivity which could reach the control room. Consumers Energy currently performs all control room habitability and offsite dose radiological consequence analyses.
The current analyses of record are contained in Reference
- 9. The re-analysis of the MSLB concludes that there would be no fuel failures therefore, the analysis of record for the control room habitability remains bounding.
Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 3.5 Fuel Centerline Melt EMF-98-013 Revision 0 Page 3-3 The U 235 enrichment of the central zone of the gadolinia-bearing rods was increased for Cycle 14 in order to improve fuel cycle economics.
In cycles prior to Cycle 13, the U 235 enrichment of gadolinia-bearing rods was designed such that a gadolinia-bearing rod could not become limiting with respect to FCM and a conservative limit on U0 2 rods of 21 kW/ft was used as the FCM limit. Increasing the enrichment for the gadolinia-bearing fuel rods causes them to reach a higher relative power, compared to the peak U0 2 rod, during the latter portion of its first operating cycle. The thermal conductivity of a fuel pellet decreases with increasing gadolinia concentration.
In addition, the melt temperature decreases with increasing gadolinia concentration.
As a result of the increased power in the gadolinia-bearing fuel rods for the Cycle 14 fuel design, SPC calculated a FCM limit for the highest-powered U0 2 fuel rod in the bundle which precludes FCM in both U0 2 rods and gadolinia rods. The limit for Cycle 14 was calculated to be 20.93 kW/ft. The limit for Reload Q fuel (Cycle 13) is 21.64 kW/ft, based on the lower gadolinia concentration and the local power distribution for that design. This limit is conservative when applied to prior fuel designs for Cycles prior to Cycle 13, since the gadolinia-bearing fuel rods operated at much lower relative powers. The Linear Heat Generation Rate (LHGR) on the peak U0 2 rod required to prevent FCM for Cycle 14 in any rod was calculated based on SPC's approved setpoint methodology (Reference
- 17) I which describes the use of approved fuel design codes for calculating FCM powers. 1 5. 0 Accident Analyses 15.0.1 Categorization of Plant Events Plant events are placed in one of four categories.
These categories, adopted by the American Nuclear Society (ANS), are described as follows: NORMAL OPERATION AND OPERATIONAL EVENTS (CONDITION I)
- Events which are expected to occur frequently in the course of power operation, refueling, maintenance, or plant maneuvering.
FAUL TS OF MODERATE FREQUENCY (CONDITION II)
- Events which are expected to occur on a frequency of once per year during plant operation.
Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events INFREQUENT FAUL TS (CONDITION Ill)
- Events which are expected to occur once during the lifetime of the plant. LIMITING FAUL TS (CONDITION IV) EMF-98-013 Revision 0 Page 3-4
- Events which are not expected to occur but which are evaluated to demonstrate the adequacy of the design. Table 15.0.1 .1 lists the accident category used for each event considered in this report. This category is used in evaluating the acceptability of the results obtained from the analysis.
These categories are unchanged from those established in Reference
- 12. 15.0.2 Plant Characteristics and Initial Conditions Six operational conditions were considered in the disposition and analysis.
These are given in Table 15.0.2.1.
These operational conditions were considered in establishing the subevents associated with each event initiator.
A set of initial conditions was established for the events analyzed that are consistent with the requirements for each condition of operation.
The conditions of plant operation are unchanged for Cycle 14. The nominal plant rated operating conditions are presented in Table 15.0.2.2 and principal fuel design characteristics in Table 15.0.2.3.
The uncertainties listed below were used in the accident analyses in Reference 3, 10, 11, 12, and 13. Core Power Primary Coolant Temperature Primary Coolant Pressure Primary Coolant Flow 15.0.3 Power Distribution
+/-2% +/- 5°F +/- 50 8 psi +/-3% The radial and axial power peaking factors used in the analysis are presented in Table 15.0.3.1.
The analyses for the inlet temperature (T mretl Limiting Condition of Operation (LCO) and for the TM/LP trip utilize a large number of axial power distributions and associated Axial Shape Indexes (ASls)
- This value represents the control uncertainties rather than the trip uncertainties.
Trip uncertainties are+/- 22 psi. Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-5 Figures 15.0.3.1 and 15.0.3.2 show the DNB-limiting axial shape used to analyze events initiated from 100% power and those initiated from 90% power, respectively.
The only events initiated from Condition 1 (power operation) that are limiting when initiated at less. than full power are the bank and rod withdrawal events, which are initiated from 90% rated power. Bounding radial and axial power peaking factors are used to set the LCOs in the Technical Specifications (Reference 14), which protect against DNB during normal operation and all AOOs. (Some events analyzed result in transient redistribution of the radial power peaking factors. Transient radial power redistribution is treated as described in Section 15.4.3.2 of Reference 2.) 15.0.4 Range of Plant Operating Parameters and States Table 15.0.4.1 presents the range of key plant operating parameters considered in the Reference 3 and Reference 11 transient analyses.
This range is unchanged for Cycle 14 . 15.0.5 Reactivity Coefficients Used in the Safety Analysis Table 15.0.5.1 presents the reactivity coefficients used in the transient analyses in References 3, 10, 11, 12, and 13. These analyses conservatively support the Technical Specification moderator temperature coefficient (MTC) of < +0.5 x 10*4 The nominal full-power Cycle 14 burnup is 13,810 MWd/MTU. The safety analysis is, however, applicable to a full-power, end-of-Cycle 14 exposure of 14,4 70 MWd/MTU that accounts for a coastdown at the end of the cycle. 15.0.6 Scram Insertion Characteristics The insertion worth of 2.0% and a control rod drop time of 2.5 seconds (to 90% insertion) have been supported by the analyses for Cycle 14. Figure 15.0.6.1 presents the negative reactivity insertion curve used for the transient analyses in References 3, 10, 11, 12, and 13. The curve does not include the trip channel delay time, but does include the clutch release time. The time in Figure 15.0.6.1 is measured from the point at which the clutch receives a release signal. The insertion worth corresponds to having the most reactive control rod stuck out . Siemens Power Corporation
-Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.0.7 Reactor Protection System Trip Setpoints and Time Delays EMF-98-013 Revision 0 Page 3-6 The applicability of the inlet temperature LCO, (T;niet LCO), and the TM/LP trip functions were confirmed for Cycle 14 operation.
The results of the Cycle 14 verification analyses are presented in the following sections.
15.0.7.1 Inlet Temperature Limiting Condition of Operation The T;niet LCO provides protection against penetrating DNB during AOO transients that do not have a subsequent reactor trip, by limiting the measured inlet temperature based on the system pressures and flows and by limiting the operating power to compensate for axial power shapes. The most limiting AOO not protected by a reactor trip is the inadvertent drop of a full-length control rod. The confirmation consists of demonstrating the ability of the T;niet LCO to provide DNB protection for this transient.
The T;niet LCO is given as:
S 542.99 + (P -2,060) x [0.058+1.0x1 o-5 (P -2,060)) + (w -138) x [1.125 -0.0205 (w -138)) where P is pressurizer pressure ( 1,800 psia S P S 2,200 psia) and w is the primary coolant system (PCS) mass flow rate (100 Mlbm/hr S w S 150 Mlbm/hr).
The pressure range over which the T;niet LCO is confirmed is set by the low pressure trip from the TM/LP ( 1, 750 psia) compensated for uncertainties and by the high pressure trip of 2,255 psia. Both trips were compensated for uncertainties and for the control deadband of +/- 50 psi. The flow range covers all possible operational ranges, even for very high tube plugging levels in the steam generators.
For primary loop flow rates greater than 150 Mlbm/hr, the inlet temperature is limited to the T;niet LCO value at 150 Mlbm/hr or to 544°F, whichever is less. The T;niet LCO is further adjusted to account for axial power shapes by the curve shown in Figure 15 .0. 7. 1 . Confirmation of the T;niet*LCO makes use of the XCOBRA-lllC computer code (Reference
- 15) to model the reactor core. The augmented radial peaking which results from the dropped control rod, was included for axial power shapes covering the ASI range of the function.
The Cycle 14 XCOBRA-lllC four channel model and the Cycle 14 axial power shapes were used to determine the power corresponding to 95/95 safety limit including a 2% mixed core DNB This analysis.was performed covering a range of pressurizer pressures, core inlet temperatures and prima.ry coolant system flow rates. Siemens Power Corporation
-Nuclear Division
- .. * *
- Palisades Cycle 1 4: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-7 The Cycle 14 verification analysis used the following uncertainties and transient allowances:
(1) a +/- 22 psia pressure measurement uncertainty, (2) a +/- 7°F inlet temperature uncertainty (Reference 12), (includes a+/- 5°F tilt allowance, and a +/- 2°F measurement uncertainty), (3) a + 0/-6% flow rate uncertainty (includes a -3% bypass flow and +/- 3% measurement uncertainty) and, (4) a -20 psi transient pressure bias. These values are the same as those used for the Cycle 13 confirmation of the T;niat LCO (Reference 10). The T;niet LCO (as defined by the functional relationship above and by Figure 15 .0. 7 .1) was confirmed to provide protection against DNB for the control rod drop event over the pressure and flow ranges established by the functional relationship and the ASI and power relationship shown in Figure 15.0.7.1.
The XCOBRA-lllC calculations demonstrated that for the maximum inlet temperatures allowed by the T;niet LCO, the DNBR was always greater than the 95/95 safety limit for the HTP correlation, with a 2% mixed core penalty over the range of pressurizer pressures, primary coolant system flow rates, axial shape indices and core power levels expected for Cycle 14 . 15.0.7.2 Thermal Margin/Low Pressure (TM/LP) Trip The function of the TM/LP trip is to protect the core against DNB during a slow heat-up or de-pressurization transient.
In order to perform this function, the TM/LP trip was designed to scram before the reactor reaches a. condition which would result in DNB or before the exit temperature exceeds the saturation temperature.
The reactor core is protected against violating the SAFDL on DNB during rapid transients by the VHP trip, the low flow trip and the high pressurizer pressure trip. Slow transients generally involve either a slow heatup of the primary coolant system caused by a power mismatch between the primary and secondary systems or a slow depressurization of the primary system with or without a slow power ramp. Transients which are protected by the TM/LP are as follows:
- uncontrolled control rod withdrawal from power
- inadvertent boron dilution
- excess load
- loss of feedwater
- inadvertent opening of a PORV 15.0. 7 .2.1 TM/LP Uncertainties The uncertainties used in the confirmation of the TM/LP for Cycle 14 are summarized in Table 15.0.7 .2. Uncertainties in measured inputs to the TM/LP trip; inlet temperature, Siemens Power Corporation
-Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-8 core power, pressure and axial shape index; were incorporated in the confirmation of the TM/LP. The uncertainties for instrument drift in both power and inlet temperature, calorimetric power measurement, inlet temperature measurement and primary pressure measurement were combined into an additional pressure uncertainty of 165 psi (Reference
- 16) and a bias of 1.5°F was added to the inlet temperature to account for thermal lag of the cold leg RTDs and the transport time from the cold-leg RTD location to the core inlet (Reference 3). The XCOBRA-lllC analysis makes use of additional uncertainties which are not related to the measured parameters.
These include an uncertainty to account for manufacturing tolerances and in-reactor changes such as densification, a 3% measurement uncertainty for PCS flow and a reduction in the active core flow to account for a 3 % core bypass flow. 15.0. 7.2.2 TM/LP Confirmation The TM/LP trip equation is: Pvar = 2012 X QA X QR 1 + 17.0 X T;n -9493 where Pvar is the trip setpoint (based on the measured pressurizer pressure), Tin is the inlet temperature and QA and OR1 are functions designed to compensate for the change in DNBR with axial power shapes and reactor power, respectively.
These functions are shown in Figures 15.0.7.2 and 15.0.7.3.
The TM/LP trip equation was developed to protect against DNB and hot leg saturation.
Since nothing has changed for Cycle 14 which would impact the ability of the trip to protect against hot leg saturation, it was not necessary to confirm this function of the TM/LP. The ability of the TM/LP trip to protect against DNB was confirmed for Cycle 14 over a pressure range from 1 , 750 psia to 2,250 psia and for a minimum measured primary coolant flow rate of 140. 7 Mlbm/hr. The pressures corresponding to the correlation limit, as adjusted by the 2 % mixed core penalty, and the TM/LP pressures (including all uncertainties) were compared to confirm that the TM/LP trip pressure was always greater than the MDNBR pressure.
Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.0.7.3 Variable High Power (VHP) Trip Uncertainties EMF-98-013 Revision 0 Page 3-9 The VHP trip uses an internal power program to determine the appropriate trip power and to compare it to the power signal provided by the Thermal Margin Monitor (TMM). The TMM uses the maximum of the thermal power (based on average hot leg temperatures and maximum cold leg temperatures) and neutron flux power to create the power signal. The uncertainty in the power signal created by the TMM was analyzed in Reference
- 8. The analysis showed that the use of an uncertainty of +/- 8.5% for fast transients and +/- 5.5% for slow transients was conservative.
15.0.7.4 Increased Trip Delays For Cycle 14 the impact of increasing the delays for the High Pressurizer Pressure trip, the Low Pressurizer Pressure trip, the Low Reactor Coolant Flow trip, the TM/LP trip and the VHP trip from the values used in the analyses reported in References 3, 11, 12, and 13 was evaluated.
The impact on each of the SRP Chapter 15 events was considered.
The impact of increasing trip delays was considered for the limiting, or bounding, event in each SRP category.
It was concluded that an increase in delay times of 0.2 seconds would not have a significant impact on the safety analysis and would not require additional analysis.
' 15.0.8 Component Capacities and Setpoints Table 15.0.8.1 presents the component setpoints and capacities supported by this analysis.
These are unchanged from those used in analyses reported in References 3, 10, 11,12,and13.
15.0.9 Plant Systems and Components Available for Mitigation of Accident Effects Table 15.0.9.1 is a summary of trip functions, engineered safety features, and other equipment available for mitigation of accident effects. These are listed for all SRP Chapter 1 5 events and are unchanged from those used in the analyses reported in the References 3, 10, 11, 12, and 13. 15 .0.10 Plant licensing Basis and Single Failure Criteria The licensing basis for Palisades is set forth in the Final Safety Analysis Report (Reference 9). The single failure criteria are established by the plant licensing basis. Palisade's licensing basis has the following single failure criteria:
Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 15 Events EMF-98-013 Revision 0 Page 3-10 1 . The RPS was designed with redundancy and independence to assure that no single failure or removal from service of any component or channel of a system would result in the loss of the protection function.
- 2. Each Engineered Safety Feature (ESF) was designed to perform its intended safety function assuming a failure of a single active component.
- 3. The onsite power system and the offsite power system were designed such that each shall independently be capable of providing power for the ESF assuming a failure of a single active component in either power system. The safety analysis was structured to demonstrate that the plant systems design satisfies these single failure criteria.
The following was assumed: 1 . The ESF required to function in an event suffers a worst single failure of an active component.
- 2. Reactor trips occur at the specified setpoint within the specified delay time with a worst single active failure. 3. A concurrent loss of offsite power occurred for these PAs: Main Steamline Break, Steam Generator Tube Rupture, and LOCA. 4. A concurrent loss of offsite power occurred for the Loss of Normal Feedwater event, which is an AOO. Criteria 10, 20, 25 and 29 of Title 10 of the Code of Federal Regulations, Part 50 ( 10 CFR 50), Appendix A, require that the design and operation of the plant and the RPS assure that the SAFDLs are not exceeded during AOOs. As per the definition of .AOO in 10 CFR 50, Appendix A: "Anticipated Operational Occurrences mean those conditions of normal operation which are expected to occur one or more times during the life of the plant and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all off site power." There are two SAFDLs: ( 1) the fuel shall not experience centerline melt and (2) the DNBR shall have a minimum allowable limit such that there is a 95 % probability with a 95 % confidence interval that DNB would not occur. The following sections, numbered according to the SRP, provide a discussion of the disposition of events review and MDNBR and peak LHGR analyses performed to support Cycle 14. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 1 5. 1 Increase in Heat Removal by the Secondary System EMF-98-013 Revision O Page The system initiators and responses for Events 15.1.1, 15.1.2, 15.1.3, and 15.1.4 are not affected by the Cycle 14 changes, described in Section 1 .0. The relative PCS cooldown rate and severity of each of these events remains unchanged from the event disposition for Cycle 13. For this category of events, the Increase in Steam Flow event (15.1.3) bounds Events 15.1.1, 15.1.2 and 15.1.4. 15.1.1 Decrease in Feedwater Temperature 1 5. 1 . 1 . 1 Event Description . A decrease in feedwater temperature event may result from the loss of one or more of the feedwater heaters. This loss can be caused by the loss of extraction steam flow from the turbine generator or by an accidental opening of a feedwater heater bypass line. The event results in a decrease of the secondary side enthalpy leading to an increase in the primary to secondary side heat transfer rate in the steam generators.
The steam generator outlet temperature on the primary side decreases causing the core inlet to also decrease.
With a negative Moderator Temperature Coefficient (MTC) the reduced core inlet temperature results in an increase in the core power and a decrease in thermal margin. 15.1.1.2 Event Disposition and Justification As noted in Section 1 5 .1 , this event is bounded by the Increase in Steam Flow event (Event 15.1.3). The event initiator and system responses for this event are less severe than those for Event 15.1.3. 15.1.2 Increase in Feedwater Flow 15.1.2.1 Event Description The Increase in Feedwater Flow event is initiated by a failure in the feedwater system. The failure may be the result of: (1) complete opening of a feedwater regulating valve, (2) over-speed of the feedwater pumps with the feedwater valve in the manual position, (3) inadvertent startup of the second feedwater pump at low power, (4) startup of the auxiliary feedwater system, or (5) inadvertent opening of the feedwater control valve bypass line. Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-12 The event results in an increase in the primary to secondary heat transfer rate in the steam generators due to increased feedwater flow. The steam generator outlet temperature on the primary side decreases causing the core inlet temperature to also decrease.
With a negative MTC the reduced core inlet temperature results in an increase in the core power and a decrease in thermal margin. 15.1.2.2 Event Disposition and Justification As noted in Section 15.1, this event is bounded by the Increase in Steam Flow -event (Event 15.1.3). The event initiator and system responses for this event are less severe than those for Event 15 .1 .3. 15.1.3 Increase in Steam Flow 15.1.3.1 Event Description The increase in steam flow event is initiated by an increase in steam demand. The increased steam demand may be initiated by operator action or by a regulating valve malfunction.
The event initiator is modeled as a step increase in steam flow. A step increase in steam flow can be caused by a rapid opening of the turbine control valves, atmospheric dump valves or the turbine bypass valve to the condenser.
The feedwater regulating valves open to increase the feedwater flow in an attempt to match the increased steam demand and maintain steam generator water level. In response to the increased steam flow, the secondary system pressure decreases, resulting in an increase in the primary to secondary heat transfer rate in the steam generators.
The primary side steam generator outlet temperature decreases due to the enhanced heat removal. The average temperature of the PCS decreases due to the mismatch between the power being removed by the steam generators and the power being generated in the core and the primary system fluid contracts, resulting in an outsurge of fluid from the pressurizer.
The pressurizer level and pressure decrease as fluid surges out of the pressurizer.
If the MTC is negative, the reactor core power would increase as the moderator temperature decreases.
TM/LP and VHP trips are available to prevent the violation of the SAFDLs. Depending on the magnitude of the increase in steam demand, a reactor trip may or may not be activated.
If no trip occurs, the reactor system will reach a new steady-state condition at a power level, greater than the initial power level, which matches the increased heat removal rate. The final steady-state condition which is achieved depends upon the Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 *Page 3-13 magnitude of the MTC.
- If the MTC is positive, the reactor power will decrease as the core average coolant temperature decreases, and this event will not produce a challenge to the SAFDLs. This is a moderate frequency event (see Table 15.0.1 .1) and the acceptance criteria are described in Section 15.0.10. Single failure criteria for this event are given in Section 15.0.10. For this analysis, the systems challenged in this event are redundant and no single active failure in the RPS or ESF can adversely impact the consequences of the event. 15.1.3.2 Event Disposition and Justification The Cycle 14 changes that could affect this event are: ( 1 ) the secondary side liquid inventory, (2) the feedwater regulating valve opening time, and (3) the assembly thermal hydrau lie characteristics.
Increased Liquid Inventory
-The liquid inventory on the secondary side would be expected to affect this event in two ways: It could affect the heat transfer coefficient.s from the primary to secondary and it could affect the steam generator level dynamic response.
Since Event 15.1.3 is a quasi-static event and both of these effects are dynamic, neither would affect the outcome of the Increase in Steam Flow event. Feedwater Regulating Valve Opening Time -The increase in the feedwater valve opening time will not impact Event 15.1.3. Since the event is slow, minor changes in dynamics are unimportant.
MDNBR Analysis -The MDNBR for this event was re-analyzed for Cycle 14. 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve 15.1.4.1 Event Description This event is initiated by an increase in steam flow caused by the inadvertent opening of a steam generator relief or safety valve. The increase in steam flow rate causes a mismatch between the heat generation rate on the primary side and the heat removal rate on the secondary side. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.1.4.2 Event Disposition and Justification EMF-98-013 Revision 0 Page 3-14 The increase in steam flow due to opening a steam generator valve is less than that considered in the Increase in Steam Flow event (Event 15.1.3), and is bounded by Event 1 5. 1 . 3. The Cycle 14 changes do not affect this conclusion and Events 1 5. 1 . 3 and 15.1.4 continue to bound this event. 15.1.5 Steam System Piping Failures Inside and Outside of Containment 15.1.5.1 Event Description The steam line break event is initiated by a double-ended, guillotine break of the main steam line which leads to an uncontrolled steam release from the secondary system. The increase .in energy removal through the secondary system results in a severe overcooling of the primary system. At EOC conditions, which are characterized by the most negative MTC in the cycle, this cooldown results in a large insertion of reactivity and, potentially, a return to power. The most reactive control rod is assumed stuck in a fully withdrawn position, thereby reducing the s.hutdown margin and introducing high power-peaking factors which could lead to significant DNB and LHGR challenges with oniy a modest return to power. 15.1.5.2 Event Disposition and Justification The Main Steamline Break was re-analyzed for Cycle 14 as discussed in Section 2.2 and the results are reported in Reference
- 23. 1 5. 2 Decrease in Heat Removal by the Secondary System The initiating mechanisms for Events 15.2.1, 15.2.2, 15.2.3, 15.2.4 and the heatup period of 15.2.8 are not affected by any of the Cycle 14 changes and the relative severity of these events established for Cycle 13 (Reference
- 10) remain valid. The Loss of External Load, Event 15.2.1, bounds Events 15.2.2, 15.2,3, 15.2.4 and the heatup period of Event 15.2.8. 15.2.1 Loss of External Load 15.2.1.1 Event Description A Loss of External Load is initiated by either a loss of external electrical load or a turbine trip and the turbine stop valve is assumed to rapidly close (0.1 second). The plant response to this event would not change if a shorter valve closure time were assumed, Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-15 since 0.1 seconds is essentially instantaneous closure. Normally a reactor trip would occur on a turbine trip, however, to calculate a conservative system response, the reactor trip on turbine trip is disabled.
The steam dump system (atmospheric dump valves -ADVs) is assumed to be inoperable.
These assumptions allow the Loss of External Load event to bound the consequences of Event 15.2.2 (Turbine Trip -steam dump system unavailable) and Event 15.2.4 (Closure of both MSIVs -valve closure time is comparable to the turbine stop valve). The Loss of External Load event challenges the acceptance criteria for both primary and secondary system pressurization and for DNBR. The event results in an increase in the primary system temperatures due to an increase in the secondary side temperature.
As the primary system temperatures increase, the coolant expands into the pressurizer causing an increase in the pressurizer pressure.
The primary system is protected against over-pressurization by the pressurizer safety relief valves (SRVs). Pressure relief on the secondary side is afforded by the Main Steam Safety Valves (MSSVs). Actuation of the SRVs and the MSSVs limits the magnitude of the primary system temperature and pressure increase.
With a positive MTC, corresponding to BOC, increasing primary system temperatures result in increasing core power. The increasing primary side temperatures and power reduce the margin to thermal limits (i.e., DNBR limits) and challenge the DNBR SAFDL. 15.2.1.2 Event Disposition and Justification The parameters influencing the severity of the transient include: ( 1) PCS high pressure trip setpoint, (2) SRV setpoints, (3) PCS over-pressure relief capacity, (4) Primary to secondary heat transfer, (5) MSSV setpoints, (6) Secondary side pressure, (7) MSSV relief capability, (8) Moderator reactivity coefficients, (9) Fuel assembly thermal hydraulic characteristics, and ( 10) Increased pressurizer spray flows. Initiating this event from full power bounds all other operating conditions.
The Reference 11 analysis evaluated the maximum primary system over-pressure using SPC's ANF-RELAP methodology (Reference
- 18) and assumed all MSSVs were operational.
Palisades Technical Specifications allow one MSSV to be inoperable.
A subsequent evaluation of the impact of operation with an inoperable MSSV (Reference
- 19) confirmed that it would not have a significant impact on the primary system pressure for this event . Secondary side pressurization was analyzed for the Loss of External Load event subsequent to the Reference 18 analysis.
The analysis used revised models for the Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-16 MSSVs and modeled the piping between the steamline and the MSSV itself. This analysis was performed with the same operability conditions used in the LOEL analysis performed to minimize DNBR, assuming one bank of MSSVs to be inoperable.
This was done to maximize the pressure on the secondary side. With the Loss of External Load there is a concurrent turbine trip, isolating the secondary side. The atmospheric dump valves (ADVs) were also assumed to be inoperable during the event allowing secondary side pressure to increase causing the MSSVs to open. The event summary for the secondary side over-pressurization analysis is presented in Table 15.2.1.1.
Secondary side pressure increased during the LOEL with a maximum pressure at 13.7 seconds. The behavior of the primary and secondary systems in this event are shown in Figures 15.2.1.1 through 15.2.1.5.
The Cycle 14 changes that could affect this event are the change in the steam generator liquid inventories, the assumed steam generator plugging level, changes to the MSSV capacities, and increased trip delay time. The impact of each of the changes is discussed below: Increased Liquid Inventory
-The initial liquid inventory on the secondary side could affect . the heat transfer coefficients from primary to secondary.
The changes that affect the secondary side dynamics that might affect the primary side response would be minute when compared to the tube plugging assumptions.
Steam Generator Tube Plugging Level -The tube plugging level assumed in the primary side transient analysis was a conservatively high value of 25% (Reference 11 ). The secondary side pressurization case assumed a tube plugging level of 0%. Both analyses therefore bound the tube plugging level assumptions of 15% in the appropriate direction.
MSSV Capacities
-The table of capacities given in Reference 6 were used in the secondary side over-pressurization re-analysis of Event 15.2.1. These values are greater than those used in the Reference 11 analysis.
Increased steam flow would tend to slightly reduce the secondary side pressure once the MSSVs open. However, since the Reference 11 analysis ignored the entry losses for the MSSVs and understated the secondary side pressure slightly, the Reference 11 analysis remains bounding for the primary side pressurization case. Increased Trip Delays -The secondary side pressurization re-analysis of the Loss of External Load event did not incorporate the 0.2 sec increase in the high pressure trip delay in the system analysis.
Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-17 A subsequent disposition showed that delaying the high pressure trip by 1 .0 second would produce an increase of about 3 psi in the secondary side pressurization.
Since the increase in the high pressure trip delay is only 0.2 seconds, the increase in secondary side pressure.
would therefore not exceed 0.6 psi and is of no real consequence to this event. The trip which actuates to protect this event is the High Pressurizer Pressure Trip, which is set at 2,255 psia and has an uncertainty of +/- 22 psi. The pressurizer pressure reaches a maximum of 2,580 psia 10.9 seconds after the event is initiated.
The relief valve setpoint is 2,575 psia and it has a 3% accumulation.
The liquid insurge is not enough to open the relief valve completely.
During most of the transient, the pressurizer sees an insurge of about 15 ft 3 per second. Using the specific volume for steam at 700 °F and 2,580 psia, the mass of steam which must be relieved is about 61 lbm/second.
This is significantly less than t.he relief capacity of the safety valve (127.8 lbm/second).
This implies that had the reactor not tripped, the relief valve would still have maintained the PCS pressure at or near the setpoint.
The capacity of the relief valve is sufficient to prevent overpressurization, even if the reactor trip is delayed . ' The total volume of the pressurizer is 1,507 ft 3* The maximum liquid volume reached in the Reference 11 analysis is about 1, 190 ft 3 , leaving about 307 ft 3 of steam volume, which would have taken about 20 seconds to fill at the average insurge rate. Delaying the trip does not produce a significant challenge to filling the pressurizer with liquid. Increased Pressurizer Spray Flow -For the primary overpressurization analysis the pressurizer sprays are disabled and an increase in the spray flow rate has no impact. Increasing the pressurizer spray flow for the steam-side analysis delays the high pressure trip by about one second resulting in a 3 psi increase in the steam-side peak pressure.
This is a negligible increase and the steam-side pressure remains below the maximum allowed steam-side pressure.
15.2.2 Turbine Trip 15.2.2.1 Event Description This event is initiated by the turbine tripping which results in the rapid Closure of the turbine stop valves. A reactor trip occurs on a turbine trip and the steam dump system mitigates the consequences of this event. The primary system is protected against pressurization by the pressurizer SRVs. Pressure relief on the secondary side is afforded by the MSSVs. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.2.2.2 Event Disposition and Justification EMF-98-013 Revision 0 Page 3-18 The Loss of External Load (Event 15.2.1) was evaluated in such a manner that it bounds the consequences of this event. 15.2.3 Loss of Condenser Vacuum 15.2.3.1 Event Description This event is initiated by a reduction in either the circulating water flow or by an increase in the circulating water temperature, which then reduces the heat loss from the condenser and increases the back pressure.
This condition can result in a turbine trip and it will also result in no steam bypass to the condenser.
The primary system is protected against over-pressurization by the SRVs. Pressure relief on the secondary side is afforded by the MSSVs. 15.2.3.2 Event Disposition and Justification At rated power and power operating conditions, the assumptions made for the Loss of . External Load event (Event 15.2.1) are conservative relative to the Loss of Condenser Vacuum. From any initial power below rated power, the operator would have sufficient time to control the primary and secondary system temperatures.
These conclusions will not change for Cycle 14. 15.2.4 Closure of the Main Steam Isolation Valves (MSIV) 15.2.4.1 Event Description Closure of the Main Steam Isolation Valves is initiated by the loss of control air to the MSIV operator.
These valves are swinging check valves designed to fail in the closed position.
The inadvertent closure of the MSIVs drastically reduces the steam load. 15.2.4.2 Event Disposition and Justification Closure time for the MSIVs is less than 5 seconds. A MSIV closure event would progress in a fashion similar to a Loss of External Load (Event 15.2.1) except for the slower valve closure time. The consequences of Event 15.2.1 would bound those for Event 15.2.4
- because of the more rapid valve closure time. This disposition is unchanged for Cycle 14. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter l 5 Events 15.2.5 Steam Pressure Regulator Failure EMF-98-013 Revision 0 Page 3-19 Palisades does not have steam pressure regulators and the Steam Pressure Regulator Failure event was not considered in this analysis.
15.2.6 Loss of Non-Emergency A.C. Power to the Station Auxiliaries 15.2.6.1 Event Description A Loss of Non-Emergency A.C. Power to the Station Auxiliaries may be caused .by a complete loss of the offsite grid together with a turbine generator trip or by a failure in the onsite A.C. Power distribution system. The loss of A.C. power results in the loss of power to the primary coolant pumps and condensate pumps which, in turn, results in the loss of the main feedwater pumps. The combination of the decrease in primary coolant flow rate, the cessation of main feedwater flow and trip of the turbine generator compounds the event consequences.
The decrease of both primary coolant flow and main feedwater decreases the primary to secondary system heat transfer rate in the steam generators resulting in the heatup of ,the primary system coolant. The increase in primary system coolant temperature increases the overpressurization potential and increases the threat of penetrating DNB with a positive MTC. 15.2.6.2 Event Disposition and Justification The event is most limiting when initiated from full power conditions.
At full power operation the stored heat in the fuel rods is maximized and the margin to DNB is minimized.
This event is separated into two distinct phases: the and the long-term.
The near-term phase is characterized by the loss of power resulting in the coastdown of the primary coolant pumps, the coastdown of the main feedwater pumps and the trip of the turbine generator.
The coastdown of the primary coolant pumps causes a rapid reduction in thermal margin. The trip of the reactor and the subsequent insertion of control rods terminates the challenge to the DNB limits. The near-term phase of the event is similar to that of a Loss of Forced Reactor Coolant Flow (Event 15.3.1 ). The near-term consequences of this event are addressed in the analysis of Event 15.3.1 . Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-20 The long-term consequences of a Loss of A.C. Power event are determined by the heat removal capacity of the auxiliary feedwater system. The long-term portion is similar to the Loss of Normal Feedwater Flow event (Event 15.2.7). The long-term effects are bounded by the Loss of Normal Feedwater Flow event. The changes for Cycle 14 do not alter this disposition.
1 5. 2. 7 Loss of Normal Feedwater Flow 15.2.7.1 Event Description A Loss of Normal Feedwater Flow is initiated by the trip of the main feedwater pumps or a malfunction in the feedwater control valves. The loss of main feedwater flow decreases the amount of subcooling in the secondary side downcomer which diminishes the primary to secondary system heat transfer in the steam generator and leads to an increase in the PCS temperature.
As the PCS temperature increases, the coolant expands into the pressurizer which increases the pressure by compressing the steam volume. The opening of the MSSVs controls the heatup of the primary side. The long-term cooling ' of the primary system is governed by the heat removal capacity of the AFW flow. The AFW pumps are automatically started upon a steam generator low liquid level signal. 15.2.7.2 Event Disposition and Justification A Loss of Normal Feedwater Flow makes the most credible challenge to acceptance criteria for rated power and power operating conditions, since more stored heat is contained in the fuel than at lower powers. For the initial PCS heat-up phase of the transient, both the DNB and the primary system over-pressurization acceptance criteria would be challenged.
The DNB challenge is maximized when it is assumed that offsite power is lost causing the primary coolant pumps to coast down. After the reactor trip system is activated, the core power would be drastically reduced mitigating the challenge to DNB. The Loss of Forced Reactor Coolant Flow (Event 15.3.1) bounds the short term DNB consequences of the Loss of Normal Feedwater.
The Loss of External Load (Event 15.2.1) bounds the short-term PCS pressurization challenge.
For the longer term phase of the transient, the slow PCS heatup threatens to both over-pressurize the primary systerri (by filling the pressurizer with liquid) and dry out the steam generators.
The parameters influencing the severity of the long term phase of the Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-21 transient include: ( 1) decay heat generation, (2) secondary MSSV settings, (3) primary coolant pump operation, (4) AFW flow rate, and (5) steam generator secondary side mass at the time of reactor trip. The changes made for Cycle 14 that could impact the long term heat-up for this event are the AFW flow rate and the decay heat. AFW Operation
-In Reference 12, it was shown that with a flow rate of 300 gpm, the auxiliary feedwater system would prevent dry-out of the steam generators.
The AFW pumps are capable of delivering a fairly large quantity of water to the steam generators.
Testing showed that P-8A can deliver 420 gpm when directed to a single steam generator at pressure of 890 psia. The AFW system controls the flow to each steam generator at 165 gpm with a flow controller uncertainty of 22 gpm. Treating the controller errors as random, the total flow rate would be 330 +/- 16 gpm, which is greater than the 300 gpm assumed for the Reference 12 analysis.
The Reference 12 analysis remains bounding.
Decay Heat -The Reference 12 analysis used asymptotic actinides.
This corresponds to a U 238 capture to fission ratio of 1 .0. The decay heat assumptions from the Reference 1 2 analysis continue to bound the Cycle 14 values. Increased Pressurizer Spray Flow -In the analysis of the Loss of Normal Feedwater Flow event, the pressurizer sprays are disabled and an increase in the spray flow rate has no impact. 15.2.8 Feedwater System Pipe Breaks Inside and Outside Containment 15.2.8.1 Event Description This event is initiated by a main feedwater system pipe rupturing downstream of the feed line check valve causing a blowdown of the affected steam generator.
If the rupture is assumed to occur upstream of the check valve, the event will behave much like the Loss of Normal Feedwater Flow (Event 15.2.7). Since the AFW flow is injected into the steam generators via a piping network separate from the main feedwater system, delivery of AFW will not be interrupted by the pipe rupture. The event results in a primary system cooldown followed by a heatup of the PCS. Initially, the event results in a cooldown of the primary side coolant due to the energy removal during the blowdown stage of the event. The eventual depletion of secondary side Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-22 inventory and lack of main feedwater causes the primary system to heat up much like a Loss of Normal Feedwater Flow. 15.2.8.2 Event Disposition and Justification During power operation, the cool-down consequences of this event are bounded by the Main Steam Line Break (Event 15.1.5) and the heat-up consequences are bounded by the Loss of External Load (Event 15.2.1 ). The long-term cooling requirements are bounded by the Loss of Normal Feedwater event analysis (Event 15.2. 7). Feedline pipe breaks from' non-power operating conditions are bounded by Event 15.1.5. The Cycle 14 system changes that could impact the disposition of this event impacts all of these events in a similar manner. Also the configuration and operational conditions for Cycle 14 remain unchanged from previous cycles and the Feedline Break event remains bounded by Events 15.1.5, 15.2.1and15.2.7.
15.3 Decrease in Reactor Coolant System Flow 15.3.1 Loss of Forced*Reactor Coolant Flow 15.3.1.1 Event Description This event is initiated by the simultaneous loss of electric power to all of the reactor primary coolant pumps. As primary coolant pumps begin to coast down, the reactor coolant flow decreases.
If the reactor is at power when this event occurs, the primary coolant temperature in the reactor core will rise rapidly. This will result in a rapid reduction in DNB margin, and could result in DNB if the reactor is not tripped promptly.
The primary coolant expands, since the heat removal at the steam generator does not increase enough to offset the increased heating in the core, this causes an insurge into the pressurizer, a reduction of the pressurizer gas volume and a rapid increase in primary coolant system pressure.
The primary system over-pressurization will be mitigated by the action of the SRVs and the reduction in core power following reactor trip. Reactor trip signals are provided from low PCS flow. The MDNBR is determined by the primary coolant flow decay and the initial increase and then rapid decrease in core power and heat flux. The ratio of core heat flux to flow initially increases, peaks, and then falls rapidly after the reactor trip. MDNBR occurs near the peak in the heat flux-to-flow and is strongly affected by the magnitude of the peak. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-23 In the near term, the pump coastdown characteristics and the timing of the reactor trip, trip delays and scram bank insertion characteristics are key parameters.
In the longer term, natural circulation flow in the primary system removes the decay power. The primary concern with this event is the near-term challenge to the SAFDls. The event is analyzed to verify that the reactor protection system can respond fast enough to prevent penetration of the DNBR SAFDL. *-. This event is classified as a moderate frequency event (Table 15.0.1.1 ). The acceptance criteria are as described in Section 15.0.10. For this analysis, the systems challenged in this event are redundant; no single active failure in the RPS or ESF will adversely affect the consequences of the event. long term recovery is provided by the auxiliary feedwater system, as demonstrated in the analysis of Event 15.2.7 in Reference
- 12. 15.3.1.2 Event Disposition The most limiting transient associated with a loss of Reactor Coolant Flow is initiated from rated power by a loss of power to all four primary coolant pumps. As the pumps ' coast down, the core flow is reduced, causing a reactor scram on low flow. As the flow coasts down, primary temperatures increase.
The Reference 3 analysis was performed using a 0.5 sec conservatism in the low flow trip delay time which bounds the Cycle 14 increase in low flow trip delay time of 0.2 sec. This increase in temperature causes a subsequent power rise due to moderator reactivity feedback.
The primary challenge to DNB is from the flow rate, increased coolant temperatures and increasing core heat flux. Increased pressure resulting from the heatup of the PCS is bounded by the Loss of External Load (Event 15.2.1 ). The MDNBR analysis was performed for Cycle 14. None of the. changes for Cycle 14 will cause the system response for this event to become more limiting and the system response in the analysis of record (Reference
- 3) remains bounding for Cycle 14. Increased Pressurizer Spray Flow -An increase in the pressurizer spray flow will reduce the primary pressure increase.
Therefore MDNBR for this event would decrease with increased spray flow. The decrease in MDNBR will be small and the result would remain above the HTP correlation limit. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.3.2 Flow Controller Malfunction EMF-98-013 Revision 0 Page 3-24 There are no flow controllers on the PCS at Palisades and this event is not credible.
15.3.3 Reactor Coolant Pump Rotor Seizure 15.3.3.1 Event Description The Reactor Coolant Pump Rotor Seizure is initiated when a primary coolant pump rotor stops rotating instantaneously.
Flow through the affected loop is rapidly reduced, causing a reactor trip due to a low primary loop flow signal. The flow continues to decrease in the affected loop until it reverses.
Following the reactor trip, the heat stored in the fuel rods continues to be transferred to the primary coolant. Because of the reduced core flow, the coolant temperatures rise, the primary coolant expands and flow surges into the pressurizer creating an increasing PCS pressure.
As the pressure increases, the pressurizer sprays and then the SRVs act to control the system pressure.
The rapid reduction in core flow and the increase in coolant temperature presents a very dramatic decrease in MDNBR and the event presents a significant challenge to DNB. Fuel centerline melt is not a serious concern because of the small power increase typical of this event. PCS pressurization criteria have not been approached in prior SPC analyses of this event and are not a serious concern for this event. The Reactor Coolant Pump Rotor Seizure is an infrequent event (Table 15.0.1.1 ). The acceptance criteria for this event are presented in Section 15.0.10. For this analysis, the systems challenged in this event are redundant; no single active failure in the RPS or ESF will adversely affect the consequences of the event. The auxiliary feedwater pumps will provide cooling capability after scram, as demonstrated in Section 15.2. Tot Reference
- 12. 15.3.3.2 Event Disposition The controlling parameters for the Reactor Coolant Pump Seizure are identical to those for the Loss of Forced Reactor Coolant Flow (Event 15.3.1) and the arguments, given in Section 15.3.1.2, for Reload Rare applicable to this event. The system response is unchanged for Cycle 14. The MDNBR analysis was performed for Cycle 14 . Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.3.4 Reactor Coolant Pump Shaft Break 15.3.4.1 Event Description EMF-98-013 Revision 0 Page 3-25 This event is initiated by a failure of a PCS pump shaft resulting in a free-wheeling impeller.
The impact of a Reactor Coolant Pump Shaft Break is a loss of pumping power from the affected pump and a reduction in the PCS flow rate. The flow reduction due to the seizure of a pump rotor is more severe than that for a shaft break; however, the potential for flow reversal is greater if the pump shaft breaks. The event is terl".'inated by the low reactor coolant flow trip. 15.3.4.2 Event Disposition The event is most limiting at rated power conditions because of a minimum margin to DNBR limits. The initial flow reduction for this event is bounded by that for the Reactor Coolant Pump Rotor Seizure (Event 15.3.3). The changes for Cycle 14 will not affect the flow reduction.
The potential for greater reverse flow due to a shaft break is accounted for in the seized rotor analysis by decreasing the rotor inertia to zero at the time of predicted flow reversal.
The consequences of the Reactor Coolant Pump SHaft Break are bounded by Event 15.3.3. 1 5 .4 Reactivity and Power Distribution Anomalies 1 5 .4.1 Uncontrolled Control Assembly Withdrawal from a Subcritical or Low Power Startup Condition 15.4.1.1 Event Description This event is initiated by the uncontrolled withdrawal of a control bank, which inserts positive reactivity and produces a power excursion.
The event initiator could be a malfunction in the reactor control or control rod control systems. The consequences of a single bank withdrawal from reactor critical, hot standby and hot shutdown (subcritical) operating conditions are considered in this event category; the consequences at power operating conditions are considered in Event 15.4.2. The control. rods are connected by electrical circuits in pre-selected configurations (banks). The electrical circuits which implement these configurations also prevent the control rods from being withdrawn except with their respective banks. Power is supplied to the banks in such a way that no more than two banks can be withdrawn at the same time and only in the proper withdrawal sequence.
Siemens Power Corporation*-
Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-26 The response of the reactor core to the withdrawal of control banks from subcritical or low power exhibits an apparent delay in responding to the reactivity insertion, followed by an extremely rapid power ramp. This behavior is caused by the fact that the doubling time for the power when the reactivity is less than about 95 ¢ is based on the time constants of the delayed neutrons and the power takes a relatively long time to respond to the reactivity insertion.
When the reactivity insertion is rapid enough that the delayed neutron population cannot bring the power up to significant levels and cause a trip before the reactor approaches prompt critical, a very rapid power excursion can result. The power excursion is first mitigated by the Doppler feedback as the fuel temperature begins to rise and ultimately terminated by the RPS. Because the power increase in this event is very rapid, fuel rod temperature and surface heat flux lag behind the neutron power. The power transient (as well as the control rod withdrawal) can be terminated by one of the following RPS trip functions: 1 . Non-safety grade high rate-of-change of power trip, 1x10" 4% to 1 5 % power. No credit is taken for this trip by SPC. 2. VHP trip 3. TM/LP trip 4. High pressurizer pressure trip 5. High rate-of-change of power alarms which initiate Rod Withdrawal Prohibit Action. No credit is taken for this trip by SPC.
- 6. If the reactor is in hot shutdown or below and one PCS pump has been out of service for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, circuit breakers 42-01 and 42-02 are open (Reference
- 14) which prevents a rod or bank withdrawal.
Of these trips, the minimum power setting of the VHP trip is the only RPS function SPC relies on to terminate this event. 15.4.1.2 Event Disposition and Justification This event can be initiated from one of two operating conditions:
Hot Standby (Condition
- 2) or Hot Shutdown (Condition 3). SPC has also evaluated the event for a transition within the Operating Conditions.
Condition 2 -In this Operating Condition, the reactor has a fission power less than 2% of rated with one or more of the control banks inserted up to, and including, the Hot Zero Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-27 Power (HZP) Power Dependent Insertion Limit (PDIL) for the reactor. This is the most limiting Operating Condition from which to initiate the event. Condition 3 -When the reactor is in Hot Shutdown, all banks are fully inserted and the reactor is shut down by at least 2% dp. The boron concentration is maintained high enough to keep the reactor subcritical until the banks have been withdrawn to .the HZP PDIL. If the shutdown banks, A and B, and regulating banks 1, 2, 3 and 4 are each worth less than 2 % dp the reactor cannot be brought critical by the withdrawal of a bank. If the worth of the bank withdrawals which are made to reach the HZP PDIL is greater than any one bank, the reactor cannot be brought critical by the withdrawal of a bank. This latter condition is true for Cycle 14. Transitions
-Based on the above comments, initiating this event from Hot Standby will bound all other Operating Conditions.
The transition from Hot Shutdown to Hot Standby would not produce a credible challenge since the reactor core must remain highly borated until the transition is made (Palisades Nuclear Plant General Operating Procedure 3, Revision 14, "Hot Shutdown to Critical in Hot Standby").
The transition from critical ' operation in Condition 2 to subcritical is such that it is possible to have the reactor just subcritical with the banks at the HZP PDIL (Palisades Nuclear Plant General Operating Procedure 8, Revision 13, "Plant Shutdown to Hot Standby/Shutdown:
Power Reduction").
This transition would result in the event being initiated at HZP with the control banks at the HZP PDIL and the reactor just subcritical.
Reactor power could be as low as 10-9 of rated. SPC evaluated this case to provide a bounding analysis for all bank withdrawals in this category.
The analysis of this event consisted of a PTSPWR (Reference
- 2) run, initiated from 10*9 of rated power. From this PTSPWR run, the conditions at the point of MDNBR were used as boundary conditions for an XCOBRA-lllC (Reference
- 15) analysis of the MDNBR. They were also used to calculate the fuel centerline temperature.
This event is described in Reference
- 10. The changes in Cycle 14 that impact this event are the reduction at the shutdown margin with less then four pumps running and the increase in the VHP trip delay time. Reduced SOM with Less than Four RCPs Operating
-The SOM was reduced from 3.75% to 3.5% dp with less than four RCPs operating.
The maximum bank worth for Cycle 14 is Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 15 Events EMF-98-013 Revision 0 Page 3-28 1. 757% which is much less than the SOM. The change in SOM will have no impact on this event. Delay to VHP Trip -The initial mitigation for the event comes from the Doppler feedback.
For the analysis of record, the Doppler feedback provides some mitigation but the power rise is terminated by the VHP trip. Extrapolating the power for the analysis of record 0.5 seconds beyond the peak would result in a maximum heat flux which is about 3% higher. The maximum heat flux in the analysis is equivalent to about 35% power. The increase iii the fuel temperature would be of a similar amount, since both the heat flux and the fuel temperature are responding with about the same lag. Since the MDNBR for this event is nearly 5 and the maximum fuel temperature is only slightly greater than 1, 100°F, the increased power would still not result in any challenge to the fuel SAFDLs. Increased Pressurizer Spray Flow -An increase in the pressurizer spray flow will reduce the primary pressure increase.
Therefore MDNBR for this event would decrease with increased spray flow. The decrease in MDNBR will be small and the result would remain above the HTP correlation limit. MDNBR Analysis -The MDNBR was calculated using the pressure, heat flux and flow corresponding to the point of MDNBR identified in the transient analysis as boundary conditions for XCOBRA-lllC.
Axial power shapes and corresponding radial distributions were calculated for a series of bank positions starting at the HZP PDIL and ending with the bank fully withdrawn.
The MDNBR was calculated for each of these power distributions to bound the possible conditions which might occur as a result of a bank withdrawal.
The MDNBR was recalculated for this event in Cycle 14. 15.4.2 Uncontrolled Control Bank Withdrawal at Power 15.4.2.1 Event Description As with Event 1 5 .4. 1 , this event is initiated by an uncontrolled withdrawal of a control bank. This withdrawal adds positive reactivity to the core which leads to power and temperature excursions.
Event 15.4.2 considers the consequences of control bank withdrawals at rated and operating initial power levels. The reactor protection trip system is designed to preclude penetration of the SAFDLs for this event. This analysis is structured such that the TM/LP, High Pressurizer Pressure, and VHP trips must provide the protection of the SAFDLs. Siemens Power Corporation
-Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-29 The TM/LP trip function is designed to protect against DNB for slow heatup and depressurization transients.
Principal DNB parameters such as power, core inlet temperature and core axial power distribution are measured.
The function decreases the margin to the trip pressure when process variables indicate a decrease in operating margin. This function is base.d on the core protection boundaries.
Operation within these boundaries assures protection of the SAFDLs. A broad range of reactivity insertion rates and initial operating conditions are possible.
The range of reactivity insertion is from very slow, as would be associated with a gradual boron dilution, and bounded on the fast end of the range by the withdrawal of the control bank with the highest worth at the highest rate. The objectives of the analysis are to confirm the adequacy of the transient allowances in the TM/LP trip to protect against DNB for slow withdrawal rates and to confirm that the VHP and High Pressurizer Pressure setpoints protect the fast withdrawals.
The analysis considers a spectrum of reactivity insertion rates and initial power levels. Since neutronic feedback, as a function of cycle exposure and design, can influence the results, these . effects are also included in the analysis.
This event is classified as a moderate frequency event (Table 15.0.1.1 ). The acceptance criteria are as described in Section 15.0.10. The single failure criteria are given in Section 15.0.10. The safety systems challenged in this event are redundant and no single active failure will adversely affect the consequences of the event. 15.4.2.2 Event Disposition and Justification References 3 and 1 3 reported system responses over a range of values for reactivity insertion rate, moderator and Doppler reactivity coefficients and initial power. The changes in the Cycle 14 that could affect this event are the increased trip delay, the MSSV flow rates, and the increased differential rod worth. Increased Trip Delay -A change in the trip delay for the TM/LP would result i.n a degradation of DNB margin. The power for the limiting withdrawal rate in the analysis of record was ramping upward at a rate of about 0.5%/second.
The inlet temperature had increased by 2°F in the 25 seconds of the transient, which corresponds to a temperature ramp rate of about 0.08°F/second.
A.n increase in the trip delay of 0.2 seconds would result in an increase in power of 0.1 % and an increase in the inlet temperature of 0.016°F. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-30 These changes in power and inlet temperature would not change the MDNBR by more than a few tenths of a percent. MSSV Flow Rates -Bank withdrawals from part power have the potential to produce a significant over-pressure transient on the secondary side of the steam generator.
Because of the VHP trip and the nominal power mismatch that can occur for post-scram conditions, the secondary side pressure challenge from Event 15.2.1 remains bounding of this event. Increased .Differential Rod Worth -Because the limiting condition for this event is at an intermediate withdrawal rate and at BOC, the increased EOC differential rod worth does not result in an increased challenge to the SAFDLs. The changes for Cycle 14 do not affect the system response for these transients.
The boundary conditions for the most limiting event identified in these analyses remained unchanged for Cycle 14. The MDNBR and peak LHGR were calculated for Cycle 14. The MDNBR calculations were performed using a 90% axial power shape, and the peak LHGR was found to be below the FCM limit of 20.93 kW/ft for Cycle 14. Increased Pressurizer Spray Flow -An increase in the pressurizer spray flow will reduce the primary pressure increase.
Therefore MDNBR for this event would decrease with increased spray flow. The decrease in MDNBR will be small and the result would remain above the HTP correlation limit. MDNBR Analysis -The MDNBR for this event was re-analyzed for Cycle 14. 15.4.3 Control Rod Misoperation Control Rod Misoperation encompasses transient and steady state configurations resulting from several different event initiators.
The specific events analyzed under this event category are: A. Dropped Control Rod or Bank B. Dropped Part-Length Control Rod C. Malpositioning of the Part-Length Control Group D. Statically Misaligned Control Rod or Bank E. Single Control Rod Withdrawal F. Core Barrel Failure Siemens Power Corporation
-Nuclear Division
- .. *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events l5.4.3.1 Event Description A. Dropped Control Rod or Bank EMF-98-013 Revision 0 Page 3-31 The dropped rod and dropped bank events are initiated by control rod drive mechanism de-energizing and releasing the rod or bank or by a malfunction associated with a control bank. The dropped rod events are classified as Moderate Frequency events (Table 15.0.1.1 ). Acceptance criteria are given in Section 15.0.10. In these events, the reactor power initially drops in response to the insertion of negative reactivity.
This results in reduction of the moderator temperature due to a mismatch between core power being generated and secondary system load demand. The core power redistributes both radially and axially in response to the local power effect of the dropped rod or bank. For EOC condition, the reactor power begins to increase due to the positive reactivity insertion from the negative MTC and reduced moderator temperature.
If the initial power decrease is more than 20% or so, the VHP trip will reset and the reactor will trip when it tries to return to power. If the reactor does not trip, the moderator temperature decreases only enough to return the core to the initial power level. The rod . and bank drop events challenge the DNBR SAFDL because of the combination of power and increased radial peaking. This challenge to the SAFDL is protected by the Tinlet LCO function for rod drops which do not result in a reactor trip. For bank drops the VHP trip will intervene.
B. Dropped Part-Length Control Rod This event is similar to the Dropped Control Rod discussed above but involves a part-length control rod. C. Malpositioning of the Part-Length Control Group This event is similar to the Statically Misaligned Bank discussed below but involves part-length control rods. D. Statically Misaligned Control Rod or Bank Static misalignment of a rod or a bank occurs when a malfunction of the control rod drive mechanism causes a control rod to be out of alignment with its bank, either higher or lower than the other control rods in the same bank, or when a bank is out of alignment with the Power-Dependent Insertion Limit (POil). During this event, the reactor can be at steady-state rated full power or part-power conditions with enhanced power peaking. This Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-32 event is classified as a Moderate Frequency event (Table 15.0.1.1 ). Acceptance criteria are given in Section 15.0.10. In the evaluation of the static rod misalignment event, a control bank is inserted but one of the rods is assumed to remain in a withdrawn state. This results in a local increase of the radial power peaking factor and a corresponding reduction in the DNB margin. The most severe misalignment occurs at full power operation, with one bank inserted beyond its PDIL and one of the control rods is fully withdrawn.
The radial power redistribution consequences of a reverse misalignment, i.e., one rod is inserted while the bank remains withdrawn, are essentially the same as the dropped rod event. The evaluation of the bank misalignment event assumes one bank is inserted or withdrawn beyond the PDIL. This event is most limiting at full power. E. Single Control Rod Withdrawal The control rod withdrawal event is initiated by an electrical or mechanical failure in the control system that causes the inadvertent withdrawal of a single control rod. This causes . an insertion of positive reactivity which results in a power excursion transient.
The movement of a single control rod out of sequence with the rest of the bank results in a local increase in the radial power peaking factor. The combination of increasing power and increasing local power peaking degrades the MDNBR and results in a challenge to DNB margin. The system response is essentially the same as that occurring in the Uncontrolled Bank Withdrawal at Power (Event 15.4.2). This event is classified as an Infrequent event (Table 15.0.1.1 ). Acceptance criteria are given in Section 15.0.10. F. Core Barrel Failure This event is initiated by the circumferential rupture of the core support barrel. The core stop supports serve to support the barrel and the reactor core by transmitting all loads directly to the vessel. The clearance between the core barrel and the supports is approximately one-half inch at operating temperatures.
The worst possible axial location of the barrel rupture is at the mid-plane of the vessel nozzle penetrations.
This forms a direct flow path between the inlet and exit nozzles in parallel with the path that goes through the core. The core sustains a small reactivity transient induced by the motion of the core relative to the inserted rod bank(s). Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-33 Reactor protection for this event during hot shutdown, refueling shutdown, cold shutdown, and refueling operating conditions is provided by Technical Specification Shutdown Margin requirements.
For the reactor critical and hot standby operating conditions, reactor protection is provided by the VHP trip and a non-safety grade high rate-of-change of power trip. For the rated power and power operating conditions, reactor protection is afforded by the VHP and TM/LP trips. The probability of a circumferential rupture of the core support barrel is nearly the same as that of a major rupture of the primary system piping and this event is classified as a Limiting Fault event (Table 15.0.1.1) and the acceptance criteria are given in Section 15.0.10. 15.4.3.2 Event Disposition and Justification A. Dropped Control Rod or Bank A dropped bank will not produce as great a challenge to the DNB SAFDL as a dropped rod because core power will decrease to a low power level following the drop of the lowest worth bank. In response to this power decrease, the VHP trip setpoint will reset at the low power condition.
This prevents a significant return to power and even though the peaking for the dropped bank is quite high, the MDNBR for the dropped bank will be less. than that of a dropped rod. The system analysis performed for the dropped bank (Reference
- 3) conservatively disregarded the VHP trip in the dropped bank analysis.
The MDNBR presented in Table 2.2 for the dropped bank is therefore overly conservative and in fact with. the VHP trip available, the dropped bank transient would have terminated before core conditions reached a level such that the MDNBR for the dropped bank would be less than that of a dropped rod. The MDNBR analysis performed for the Dropped Rod event in Cycle 14 therefore remains bounding of the dropped bank event. B. Dropped Part-Length Control Rod A dropped part-length.
control rod will not be as severe as a dropped full-length control rod and is therefore bounded by the dropped full-length control rod event. C. Malpositioning of the Part-Length Control Group Use of part-length control rods is not allowed during power operation at Palisades.
The part-length control rods are maintained in a fully withdrawn state and this event is not credible.
Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events D. Statically Misaligned Control Rod/Bank EMF-98-013 Revision 0 Page 3-34 The radial peaking augmentation factor for this event is smaller than that for a Dropped Control Rod event. Therefore, this event is not as severe and is bounded by the Dropped Control Rod. E. Single Control Rod Withdrawal The withdrawal of a single control rod results in a reactivity insertion and a localized increase in radial peaking. The degradation of core conditions characteristic of *a reactivity insertion transient, combined with an increase in local peaking, poses a challenge to DNBR limits. The MDNBR was calculated for Cycle 14 using the 90% limiting axial power shape in Figure 15.0.3.2 and the system response for the most limiting withdrawal rate in Event 15.4.2. The calculated peak LHGR increased for Cycle i 4 due to an increase of 4% in local peaking, but remained below the 20.93 kW/ft limit for centerline melt. Since no fuel failures were predicted for this event, the acceptance criteria are met. Since the system response for Event 15.4.2 is used for this event, the disposition of the increase in the VHP trip delay time and the reduction in SOM with fewer than four pumps operating for event 1 5 .4.2 is also applicable to this event. Since the peak power obtained during a low-power reactivity insertion increases with increasing withdrawal rate, the system response for the withdrawal of a single control rod is bounded by the results for a withdrawal of a control bank from Condition 2 (Event 15.4.1 ). Condition 3 does not present a credible challenge since the worth of the most reactive control rod is less than the required shutdown margin for Hot Shutdown conditions and the reactor can not be made critical by the withdrawal of any single control rod. The consequences of the withdrawal of a single control rod while in Operating Conditions 2 or 3 (Table 15.0.2.1) are either bounded or do not challenge the acceptance criteria.
F. Core Barrel Failure A core barrel failure results in a small reactivity insertion and small increase in bypass flow. The event is not credible during hot shutdown, refueling shutdown, cold shutdown and refueling operation due to the Technical Specification shutdown margin requirements.
The event initiated from rated power bounds initiation from a lower power or from Hot Standby Operating Conditions.
Siemens Power Corporation
-Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-35 The core barrel failure event is bounded by the consequences of Control Rod Ejection (Event 15.4.8). Specifically, the reactivity insertion rate and radial power redistribution for the Control Rod Ejection are much worse than those-corresponding to a core barrel failure. Both of these events are classified as Limiting Faults (Table 15.0.1.1 ). 15.4.4 Startup of an Inactive Loop 15.4.4.1 Event Description This event is initiated by the startup of an inactive primary coolant pump while operating with less than four pumps. The startup of the inactive pump can lead to an introduction of colder primary coolant into the reactor core. The lower coolant temperature, together with a negative MTC, can cause an increase in core power and a degradation of DNB margin. Sufficient protection is available to limit the consequences of this event. 15.4.4.2 Event Disposition and Justification Plant Startup procedures do not allow operation with powers greater than 10-4 % of rated power when less than four PCPs are operating.
15.4.5 Flow Controller Malfunction There are no flow controllers on the PCS at Palisades and this event is not credible.
15.4.6 CVCS Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant 15.4.6.1 Event Description Dilution of the boron concentration can occur when primary grade water is added to the PCS via the Chemical Volume and Control System (CVCS) during cold shutdown or refueling shutdown conditions.
Boron dilution at power operation is used in the Uncontrolled Bank Withdrawal analysis (Event 15.4.2) to set the slowest reactivity insertion rates so that Event 15.4.2 bounds this event. The dilution of primary system boron adds positive reactivity to the core. This event can lead to an erosion of shutdown margin for subcritical initial conditions, or a slow power excursion at power operation.
Following operator detection of the boron dilution, operator action must be taken to terminate the dilution and to restore the required shutdown margin. A minimum of 15 minutes (Operating Conditions 1 through 5) or 30 minutes (Operating Condition
- 6) is allowed for the operator to both identify and terminate the Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-36 reduction of the boron concentration.
Analysis of the boron dilution must demonstrate that the shutdown margin required by the Technical Specification is sufficient to allow at least 1 5 minutes for Operating Conditions 1 through 5 and 30 minutes for Operating Condition 6 before the reactor becomes critical.
In the event of a boron dilution during Cold Shutdown, Hot Shutdown or Hot Standby, the following indications and alarm functions are available to alert the operator:
- Volume control tank level indication and high/low alarms;
- Letdown diverter valve position indication;
- Charging flow indication; and
- Wide range logarithmic nuclear instrumentation.
A boron dilution at power operating conditions behaves in a manner similar to a slow uncontrolled withdrawal of a control bank (Event 15.4.2). 15.4.6.2 Event Disposition and Justification The parameters affecting the boron dilution time-to-criticality include: ( 1) the volume of the PCS coolant, (2) the PCS charging flow rate, (3) the PCS charging boron concentration, (4) the PCS boron concentration at event initiation versus operating mode, and (5) the PCS critical boron concentration versus operating mode. The changes introduced in Cycle 14 do not impact (1 ), (2) or (3), but do impact (4) and (5). The initial boron concentrations and critical boron concentrations for Cycle 14 are higher than those for Cycle 12. The time to reach criticality for any subevent is driven by the ratio of the critical concentration to the initial concentration.
As the ratio of critical to initial boron concentration decreases, the time to critical during the boron dilution increases.
The operating conditions show a decrease in the ratio of the boron concentrations relative to Cycle 12. The analysis for Cycle 1 2 therefore bounds Cycle 14. 1 5 .4. 7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position 15.4.7.1 Event Description An inadvertent loading of a fuel assembly in an improper position can result in an alteration of the power distribution in the core which can adversely affect thermal margin. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.4.7 .2 Event Disposition and Justification EMF-98-013 Revision 0 Page 3-37 The event is precluded due to the administrative controls and procedures, including startup testing, that ensure a properly loaded core. The Cycle 14 changes do not alter this disposition.
15.4.8 Spectrum of Control Rod Ejection Accidents 15.4.8.1 Event Description This event is initiated by a failure in the pressure housing for the control rod drive mechanism, which causes a rapid ejection of the affected control rod. The ejection of the control rod inserts positive reactivity causing an increase in core power. The resultant core thermal power excursion is limited primarily by the Doppler reactivity effect due to the increased fuel temperatures and is terminated by the VHP trip. Because of the increase in core power, this event challenges acceptance criteria for deposited enthalpy, consequences of radiological releases and PCS pressurization . 15.4.8.2 Event Disposition and Justification This event was analyzed for Cycle 14 (Reference 24). 15.4.9 Spectrum of Rod Drop Accidents (BWR) This event pertains to BWRs and is not applicable to Palisades.
1 5. 5 Increases in Reactor Coolant System Inventory 15.5.1 Inadvertent Operation of the ECCS that Increases Reactor Coolant Inventory 1 5. 5. 1 . 1 Event Description This event is caused by an inadvertent actuation of the ECCS that results in an increase in the primary system inventory.
The primary challenge is to the primary system over-pressurization criterion.
15.5.1.2 Event Disposition and Justification As long as the PCS pressure is maintained above the shutoff heads for the pumps, no flow can be initiated from the HPSI and LPSI. The shutoff head for the LPSI pump is so low that injection by this pump is not credible.
Siemens Power Corporation
-Nuclear Division
- Palisades Cycle* 14: Disposition and Analysis of Standard Review Plan Chapter 15 Events EMF-98-013 Revision 0 Page .3-38 Cycle 14 changes do not result in modifications to the plant configuration which would affect this event and Event 15.2.1 remains bounding.
15.5.2 CVCS Malfunction that Increases Reactor Coolant Inventory 15.5.2.1 Event Description A malfunction in the CVCS could result in the inadvertent operation of the charging system pumps. If the letdown system is not operating, this C?ln lead to an increase in the PCS inventory and, potentially, an over-pressurization of the primary system, possibfy accompanied by a dilution of the primary system boron concentration.
--15.5.2.2 Event Disposition and Justification The PCS over-pressurization for this event is controlled by the charging system flow rate capacity and the relief capacity of the primary safety valves. The mass flow (steam discharge) capacity of the three safety valves is significantly greater than the inlet mass flow of the three charging pumps. Therefore, there is sufficient discharge capacity to prevent the primary system from being over-pressurized.
The changes will not affect this disposition.
The potential boron dilution consequence is bounded by Event 15.4.6. 1 5. 6 Decreases in Reactor Coolant 15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve 15.6.1.1 Event Description This event is initiated by the inadvertent opening of a pressurizer SRV or PORV, which results in the blowdown of PCS. The primary system pressure decreases rapidly until the pressurizer liquid is depleted.
The PCS is then stabilized at a pressure corresponding to saturation of the hot leg. Reactor scram occurs on TM/LP well before the pressurizer liquid is depleted during the full power case, thus terminating any challenge to the SAFDLs. This event is primarily considered a de-pressurization event and loss-of-inventory event. At BOC, there can be a positive moderator density coefficient at full power and the thermal margin can be eroded by increased power, increased coolant inlet temperatures and decreased pressure.
In the long term, the event can become a non-limiting variant of the SBLOCA (Event 15.6.5). This accident is classified as a Moderate Frequency event (Table 15.0.1.1 ). The TM/LP trip affords protection against violation of the acceptance criteria (Table 15.0.1 .1) for this Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-39 event. The systems challenged are redundant; no single active failure in the RPS or ESF will adversely affect the consequences of the event. 15.6.1.2 Event Disposition and Justification The event is principally of concern in the short term because of the potential challenge to the DNB SAFDL due to de-pressurization before scram. The de-pressurization has little effect on core power or primary temperatures.
For non-power Operating Conditions, the stored energy in the PCS coolant is less than that for Power Operation and final reactor power is limited to levels low enough that no challenge to DNB is possible.
At full power, this event is a de-pressurization event in which power, inlet temperature and flow remain essentially the same. The parameters controlling the severity of this transient are the PORV flow rate and the TM/LP trip setpoint.
The nominal PORV flow rate for Cycle 14 remains bounded by the Reference 3 analysis and the 0.5 second conservatism on the TM/LP trip delay time bounds a 0.2 second increase in the trip delay for Cycle 14 . In the long term, the PCS will continue to lose coolant through the open valve. Since the escaping inventory is steam, the flow rates are significantly less than the break flows experienced in the SBLOCA event. The event remains far less severe in the long term than the SBLOCA for Cycle 14. The MDNBR analysis was performed for Cycle 14. 15.6.2 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment 15.6.2.1 Event Description This event is initiated by the rupture of a small line carrying primary coolant outside of containment.
This rupture leads to a depletion of primary system coolant and a release of contaminated liquid. The charging and HPSI systems provide sufficient coolant to maintain the PCS inventory.
Assuming a reactor trip on TM/LP or Safety Injection Signal (SIS), no fuel failures would be predicted to occur and the radiological consequences are limited, since the source term can have no higher activity than the maximum primary coolant activity level allowed by the Technical Specifications . Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 15.6.2.2 Event Disposition and Justification EMF-98-013 Revision 0 Page 3-40 The changes associated with Cycle 14 operation affect neither the initiating faults leading to the pipe rupture nor the primary coolant activity level. Therefore, this event remains bounded by the analysis reported in Reference
- 9. 15.6.3 Radiological Consequences of Steam Generator Tube Failure 15.6.3.1 Event Description This event is initiated by the mechanical failure of a steam generator tube, which causes coolant to flow from the primary system to the secondary system. This flow can deplete the primary coolant inventory thus reducing the PCS pressure, which degrades the margin to the DNB SAFDL. The tube failure also results in release of fission products from the PCS coolant to the steam side of the steam generator, which is outside of containment.
The MSSVs are assumed to open, and the fission products are released directly to the environment.
The controlling factors for this event are the tube break size the activity in the PCS coolant, the allowed activity on the secondary side and the secondary side decontamination factors. The latter are strongly affected by uncovering o'f steam generator tubes. 15.6.3.2 Event Disposition and Justification This event is controlled by the system response of the plant. The AFW operation has a
- potential effect on the system response and the magnitude of this radiological source-term.
- AFW Operation
-The AFW pumps are capable of delivering a fairly large quantity of water to the steam generators.
Testing has shown that P-8A will provide 420 gpm when directed to a single steam generator at pressure of 890 psia. The system controls the flow to each steam generator at 165 gpm with a flow controller uncertainty of 22 gpm. Treating the controller errors as random, the total flow rate is 330 +/- 16 gpm. The AFW flow for Cycle 1 3 will be greater than 300 gpm and there is no increased likelihood of uncovery.
Cycle 14 remains bounded by existing analyses.
15.6.4 Radiological Consequences of a Main Steam Line Failure Outside Containment (BWR) This event pertains to BWRs and is not applicable to Palisades.
Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-41 15.6.5 Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary 15.6.5.1 Event Description A LOCA is initiated by a failure of PCS pressure boundary.
The event may be characterized, based on the initiators; as either SBLOCAs, involving breaks in small auxiliary piping, or LBLOCAs, involving complete ruptures of the PCS piping. The limiting features of the analyses for these events are the peak clad temperature (PCT) and the time at elevated temperature, both of which influence the extent of localized and core-wide zircaloy oxidation reactions.
The SBLOCA analysis for Palisades is not performed by SPC. 15.6.5.2 Event Disposition and Justification The controlling parameters for the transient are the following: ( 1 ) the initial fuel stored energy, (2) the decay heat, (3) the radial and axial power profiles, (4) the fuel rod-to-PCS coolant heat transfer versus time, and (5) the operating conditions for the ECCS systems. The changes for Cycle 14 affect (1), (2), (3), (4) and (5). The LBLOCA event was analyzed for Cycle 14 and the results are reported in Reference 21 . 1 5. 7 Radioactive Releases from a Subsystem or Component
- 15. 7 .1 Waste Gas System Failure This event is not sensitive to the changes described in Section 1 .0. 15.7.2 Radioactive Liquid Waste System Leak or Failure (Release to Atmosphere)
This event is not sensitive to the changes described in Section 1.0. 15.7 .3 Postulated Radioactive Releases Due to Liquid-Containing Tank Failures This event is not sensitive to the changes described in Section 1.0. 15. 7 .4 Radiological Consequences of Fuel Handling Accident 15.7.4.1 Event Description A fuel handling accident occurs when a fuel assembly is damaged during refueling operations such that fuel rods are ruptured, resulting in a release of radioactivity.
The inventory of radioactive fission products is determined by the exposure and power level of the assemblies.
The power and peaking factors for Cycle 14 remain unchanged from Cycle 13. Increases in the gap fractions of 1-131 and Kr-85 for high burnup fuel have been Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-42 included in the previous analysis of the radiological consequences of the Fuel Handling Accident.
15.7.4.2 Event Disposition and Justification The inventory of radioactive fission products is determined by the exposure and power level of the assemblies or fuel rods. The change in the Reload R fuel design will not affect the reference analysis (Reference
- 9) performed by Consumers Energy. 1 5. 7. 5 Spent Fuel Cask Drop Accidents 15.7.5.1 Event Description A spent fuel cask drop accident can result in the damage of an irradiated fuel assembly and the subsequent release of radioactivity.
The inventory of fission products is determined by the exposure and power level of the damaged assemblies.
15.7.5.2 Event Disposition and Justification The inventory of radioactive fission products is determined by the exposure and power level of the assemblies or fuel rods. The power and peaking factors tor Cycle 14 are unchanged from Cycle 13. The mechanical design change made to the Reload R design will not impact the reference analysis (Reference
- 9) for this event . Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-43 Table 15.0.1.1 Accident Category Used for Each Analyzed Event SRP Event Designation Event Name 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 Decrease in Feedwater Temperature 15.1.2 Increase in Feedwater Flow 15.1.3 Increase in Steam Flow 15.1 .4 Inadvertent Opening of a Steam Generator Relief or Safety Valve 15.1.5 Steam System Piping Failures Inside and Outside of Containment 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.1 Loss of External Load 15.2.2 Turbine Trip 15.2.3 Loss of Condenser Vacuum 15.2.4 Closure of the Main Steam Isolation Valves (MSIV) 15.2.5 Steam Pressure Regulator Failure 15.2.6 Loss of Non-emergency A.C. Power to the Station Auxiliaries 15.2.7 Loss of Normal Feedwater Flow 15.2.8 Feedwater System Pipe Breaks Inside and Outside Containment 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW 15.3.1 Loss of Forced Reactor Coolant Flow 15.3.2 Flow Controller Malfunction Anticipated Operational Occurrence (AOO) Postulated Accident (PA) Siemens Power Corporation
-Nuclear Division Categorya Moderate (AOO). Moderate (AOO) Moderate (AOO) Moderate (AOO) Limiting Fault (PA) Moderate (A'OO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Limiting Fault (PAI Moderate (AOO) Moderate (AOO)
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-44 Table 15.0.1.1 Accident Category Used for Each Analyzed Event (continued) a SRP Event Designation Event Name 15.3.3 Reactor Coolant Pump Rotor Seizure 15.3.4 Reactor Coolant Pump Shaft Break 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled Control Rod Bank Withdrawal from a Subcritical or Low Power Condition 15.4.2 Uncontrolled Control Rod Bank Withdrawal at Power Operation Conditions 1 5.4.3 Control Rod Misoperation
- 1) Dropped Control Bank/Rod 2) Dropped Part-Length Control Rod 3) Malpositioning of the Part-Length Control Group 4) Statically Misaligned Control Rod/Bank 5) Single Control Rod Withdrawal
- 6) Core Barrel Failure 15.4.4 Startup of an Inactive Loop 15.4.5 Flow Controller Malfunction 15.4.6 CVCS Malfunction that Results in a Decrease in the .Boron Concentration in the Reactor Coolant 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position 15.4.8 Spectrum of Control Rod Ejection Accidents Anticipated Operational Occurrence (AOO) Postulated Accident (PAI Siemens Power Corporation
-Nuclear Division Category*
Infrequent (PA) Limiting Fault (PA) Moderate (AOO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Moderate (AOO) Infrequent (PA) Limiting Fault (PA) Moderate (AOO) Moderate (AOO) Moderate (AOO) Infrequent (PA) Limiting Fault (PAI
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-45 Table 15.0.1.1 Accident Category Used for Each Analyzed Event (continued) a SRP Event Designation Event Name 15.4.9 Spectrum of Rod Drop Accidents (BWR) 15.5 INCREASES IN REACTOR COOLANT SYSTEM INVENTORY 15.5.1 Inadvertent Operation of the ECCS that Increases Reactor Coolant Inventory 15.5.2 CVCS Malfunction that Increases Reactor Coolant Inventory 15.6 DECREASES IN REACTOR COOLANT INVENTORY 15.6.1 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve 15.6.2 Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside of Containment 15.6.3 Radiological Consequences of Steam Generator Tube Failure 15.6.4 Radiological Consequences of a Main* Steam Line Failure Outside Containment 15.6.5 Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary Anticipated Operational Occurrence (AOC) Postulated Accident (PA) Siemens Power Corporation
-Nuclear Division Categorya Not Applicable Moderate (AOO) Moderate (AOO) Moderate Infrequent (PA) Limiting Fault (PA) Not Applicable Limiting Fault (PA)
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-46 Table 15.0.1.1 Accident Category Used for Each Analyzed Event (continued)
SRP Event Designation Event Name Category 15.7 RADIOACTIVE RELEASES FROM A SUBSYSTEM OR COMPONENT 15.7 .1 Waste Gas System Failure Note a 15.7.2 Radioactive Liquid Waste System Note a Leak or Failure (Release to Atmosphere) 15.7.3 Postulated Radioactive Releases Due Infrequent (PA) to Liquid-Containing Tank Failures 15. 7.4 Radiological Consequences of Fuel Limiting Fault (PA) Handling Accident 15. 7.5 Spent Fuel Cask Drop Accidents Infrequent (PA) This event has been deleted from the SRP but is part of the Palisades licensing basis. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter l5 Events Operating Condition 1 -Power Operation 2 -Hot Standby 3 -Hot Shutdown 4 -Refueling Shutdown Condition 5 -Cold Shutdown Condition 6 -Refueling Operation Table 15.0.2.1 Plant Operational Conditions Reactivity Pow era Critical > 2%b Any Withdrawn Rode < 2 % Shutdown Margind < 1 0" 4% 2% Lip Shutdown margin of at 0 least 5 % Lip with all control rods withdrawn keff S 0.98 with all control 0 rods in the core and the highest worth control rod fully withdrawn Any operation involving 0 movement of core components when the vessel head is unbolted or removed Percent based on 2,530 MWt. Does not include decay heat. EMF-98-013 Revision 0 Page 3-47 Average Core Temperature
> 525°F > 525°F > 525°F < 210°F < 210°F b When the fission power from the reactor core corresponds to 2,530 MWt, the reactor is at rated power. c d In this Operating Condition the reactor may be critical or subcritical.
Reactor critical is defined as having a fission power of at least > 10-4% of 2,530 MWth. The shutdown margin is the amount of reactivity which, if added to the reactor while it is subcritical, would just make it critica1 or; if it is critical, the amount the reactor would be subcritical if all control banks (with the exception of the single most worthy rod) were inserted instantaneously.
Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan EMF-98-013 Revision 0 Page 3-48 Chapter 1 5 Events b d e Table 15.0.2.2 Nominal Plant Operating Conditions Core Thermal Power (MWt) 2,530 Pump Thermal Power, total (MWt) 15 System Pressure (psia) 2,060 Vessel Coolant Flow Rate (Mlbm/hr) 144.6 8 Active Coolant Flow Rate (Mlbm/hr) 140.3b Core Inlet Coolant Temperature
(°F) 537°fC Steam Generator Pressure (psia) 770d Steam Flow Rate (Mlbm/hr) 10.982 Feedwater Temperature
(°F) 435 Number of Active Tubes per Steam 6,986 8 Generator Total Number of Assemblies 204 The Technical Specification minimum flow corresponding to a T;n1et of 532°F is 140. 7 Mlbm/hr. Assumes a 3 % core bypass flow. Maximum allowed inlet temperature at power is 544°F and at HZP T;niet is 532°F. The T;niet LCO sets a separate limit which may be more restrictive.
This value is for full power. Nominal Steam Generator pressure at HZP is 900 psia . Corresponds to 15% total tube plugging.
Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events Table 15.0.2.3 Nominal Reload R Fuel Design Parameters Total Number of Reload R Fuel Assemblies 60 Fuel Assembly Design Type 15 x 15 Fueled Rods per Assembly 216 Instrument Tubes per Assembly 1 Guide Bars per Assembly 8 Assembly Pitch (inches) 8.485 8 Rod Pitch (inches) 0.550 Fuel Pellet Outside Diameter (inches) 0.36 Clad Inside Diameter (inches) 0.367 Clad Outside Diameter (inches) 0.417 Active Fuel Length (inches) 132.6 Number of Spacers 10 EMF-98-013 Revision 0 Page 3-49 This is the average pitch. It averages the 0.32" gap for the control blades to produce an effective pitch. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events Table 15.0.3.1 Core Power Distribution b Radial Peaking Factor: * *
- Reload M 0 Reload Nb Reloads 0 through R Axial Peaking Factor Engineering Tolerance Uncertainty Fraction of Power Deposited in the Fuel Assembly ( F:) 1.57 1.66 1.76 See Figure 15.0.3.1 1.03 0.974 Reference 3 addresses the radial peaking limits for Reload M. Reference 4 addresses the radial peaking limits for Reload N. Siemens Power Corporation
-Nuclear Division EMF-98-013 Revision 0 Page 3-50 Peak Rod ( F;) . 1.92 1.92 2.04
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-51 Table 15.0.4.1 Range of Key Initial Condition Operating Parameters Core thermal power Average coolant temperature (Power operation)
Primary coolant system pressure Pressurizer water level Feedwater flow and temperature Subcritical to 2,580.6 MWt 8 Nominal +/- 5°F 2,060 psia +/- 50 psi Programmed
+/- 5 % of level span Range consistent with power level a Includes an additional 2% to account for the calorimetric error. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events Table 15.0.5.1 Reactivity Parameters BOC Item Bounding Moderator Temperature Coefficient, 10-4 Llp/°F 0.5 Doppler Temperature Coefficient, 10-5 Llp/°F -1.09 Moderator Pressure Coefficient, 1 o-a Llp/psi -1.0 Delayed Neutron Fraction 0.0075 U 238 Atoms Consumed per Total Atoms Fissioned 0.60 8 Value used in Reference 3 is 0.54. Siemens Power Corporation
-Nuclear Division EOC EMF-98-013 Revision 0 Page 3-52 Bounding -3.5 -1.76 7.0 0.0045 0.70
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan EMF-98-013 Revision 0 Page 3-53 Chapter 1 5 Events b c d Table 15.0.7.1 Trip Setpoints for Operation at 2530 MWt Trip Setpoint Values for Analysis Delay Timea Low Reactor Coolant Flow 93% of Technical Specification 0.8 sec flow High Pressurizer Pressure 2,277 psia 0.8 Low Pressurizer Pressure 1,750 psia 0.8 Low Steam Generator Pressure 485 psia 0.8 Low Steam Generator Levelb 24.88 feet 0.8 Thermal Margin/Low Pressurec P = f(TH,Td 0.8 Variable High Power Trip < 23.5d above power with 0.6 a 115% maximum and a 36.86% minimum The delay times cited in this table were used in the transient analysis.
Analyses for Cycles 1 3 and 14 have confirmed that exceeding the values given in the table by 0.2 seconds will not impact the safety analysis.
Height above low tap for level measurement.
The TM/LP trip setpoint is based on pressurizer pressure (P) setpoint, varying as a function of the maximum cold leg temperature (Tc), the measured power, and the measured axial shape index. Used for fast transient.
For slow transient it is 20.5% above power. Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events Table 15.0. 7 .2 TM/LP Uncertainties Instrument Drift (Power, T;n 101) Calorimetric Power T;n1et measurement Pressure Measurement RTD Measurement Engineering Tolerances Primary Coolant Flow Bypass Inlet Temperature Bias to Account for RTD and Transport Time Delays Axial Shape Index Siemens Power Corporation
-Nuclear Division EMF-98-013 Revision 0 Page 3-54 +/- 165 psi +/- 3% +/- 3% -3% 1.5°F +/- 0.06 Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-55 Table 15.0.8.1 Component Capacities and Setpoints Pressurizer Safety Valves (3) Pressurizer Power Operated Relief Valves (2) Steam Line Relief Valves (24) Turbine Stop and* Control Valves Steam Dump Valves and Turbine Bypass Pressurizer Backup Heaters Pressurizer Proportional Heaters Pressurizer Spraysd Letdown Orifice Valves CVCS Makeup System Normal Feedwater system b At 2,575 psia At 2,400 psia At 3% accumulation Response Time 2 sec 0.1 sec 3.0 sec Nominal Setpoint Total Setpoint Uncertainty Capacity 2,500 psia 3% 191 lbm/sec 2,540 psia 230,000 8 2,580 psia lbm/hr/valve 2,400 psia 22 psia 271 .lbm/sec 487,BOOb lbm/hr/valve Group A: 985 psig 3% 511 ,563 lbm/hrc Group B: 1,005 psig 521,802 lbm/hrc Group C: 1,025 psig 532,041 lbm/hrc Turbine trip then Tavg program Always on 1,'350 kW Full on-1,985 psia 50 psia 150 kW Full off-2,060 psia 50 psia Full on-2,060 psia 50 psia 500 gpm Full off-2,060 psia 50 psia (1.5 gpm continuous flow) Level controller 120 gpm Level controller 133 gpm Feedwater controller 1.098 x 10 7 lbm/hr d Pressurizer sprays are turned on or off when the pressurizer pressures passes through the 2060 psia setpoint.
If pressurizer pressures increases (exceeds 2060 psia) sprays come on. Likewise when pressure decreases passing through the setpoint sprays turn off. Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-56 Table 15.0.9.1 Overview of Plant Systems and Equipment Available for Transient and Accident Conditions SRP No. Event Reactor Trip Functions 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1.1 Feedwater System Malfunctions 15.1.2 15.1.3 Increase in Steam Flow 15.1.4 -Inadvertent Opening of a Steam Generator Relief or Safety Valve 15.1.5 Steam System Piping Failure VHPTrip TM/LP Trip Low Steam Generator Pressure Trip Safety Injection Actuation Signal Low Steam Generator Pressure Trip TM/LP Trip VHP Trip Sat ety Injection Actuation Signal Low Steam Generator Pressure Trip TM/LP Trip VHPTrip Safety Injection Actuation Signal Low Steam Generator Pressure Trip TM/LP Trip VHP Trip Safety Injection Actuation Signal High Containment Pressure Siemens Power Corporation
-Nuclear Division Other Signals/Equipment Steam Generator Water Level Signals Feedwater Isolation Valves Main Steam Line Isolation Valves Turbine Trip on Reactor Trip Chemical and Volume Control System (CVCS) Steam Generator Water Level Signals Main Steam Line Isolation Valves Turbine Trip on Reactor Trip Atmospheric Steam Dump Controller Steam Bypass to Condenser Controller Auxiliary Feedwater System eves Steam Generator Water Level Signals Main Steam Line Isolation Valves Turbine Trip on Reactor Trip Atmospheric Steam Dump Controller Steam Bypass to Condenser Controller Auxiliary Feedwater System eves Steam Generator Water Level Signals _ Main Steam Line Isolation Valves Turbine Trip on Reactor Trip Atmospheric Steam Dump Controller Steam Bypass to Condenser Controller Auxiliary Feedwater System Containment Spray Containment Isolation Containment Air Coolers eves
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 15 Events EMF-98-013 Revision 0 Page 3-57
- Table 15.0.9.1 Overview of Plant Systems and Equipment Available for Transient and Accid.ent Conditions (continued)
SRP No. Event Reactor Trie Functions Other Signals/Eguiement 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.1 Loss of External Load/Turbine High Pressurizer Pressure Trip Steam Generator Water Level Signals 15.2.2 Trip/Loss of Condenser VHP Trip Turbine Trip on Reactor Trip 15.2.3 Vacuum/MSIV Closure TM/LP Trip Atmospheric Steam Dump Controller 15.2.4 Low Steam Generator Water Level Trip Steam Bypass to Condenser Controller Steam Generator Safety Valves Pressurizer Safety Valves Pressurizer Sprays 15.2.6 Loss of Nonemergency Low Reactor Coolant Flow Trip Steam Generator Water Level Signals AC Power to the High Pressurizer Pressure Trip Steam Generator Safety Valves Station Auxiliaries TM/LP Trip Pressurizer Safety Valves Low Steam Generator Water Level Trip Auxiliary Feedwater System 15.2.7 Loss of Nonnal Low Steam Generator Water Level Trip Steam Generator Water Level Signals Feedwater Flow High Pressurizer Pressure Trip Steam Generator Spfety Valves TM/LP Trip Pressurizer Safety Valves Auxiliary Feedwater System Pressurizer Sprays and Level Control 15.2.8 Feedwater System Pipe Break High Pressurizer Pressure Trip Steam Generator Water Level Signals TM/LP Trip Steam Generator Safety Valves Low Steam Generator Water Level Trip Pressurizer Safety Valves Low Steam Generator Pressure Trip Auxiliary Feedwater System Pressurizer Sprays and Level Control 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE 15.3.1 Loss of Forced Reactor Coolant Low Reactor Coolant Flow Trip Atmospheric Steam Dump Controller Flow TM/LP Trip Steam Bypass to Condenser Controller High Pressurizer Pressure Trip Steam Generator Safety Valves Pressurizer Safety Valves 15.3.3 Reactor Coolant Pump Rotor Low Reactor Coolant Flow Trip Atmospheric Steam Dump Controller 15.3.4 Seizure/Shaft Break High Pressurizer Pressure Trip Steam Bypass to Condenser Controller Steam Generator Safety Valves Pressurizer Safety Valves Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-58 Table 15.0.9.1 Overview of Plant Systems and Equipment Available for Transient and Accident Conditions (continued)
SRP No. Event Reactor Trie Functions Other Signals/Eguiement 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1 Uncontrolled Control Rod Bank* TM/LP Trip Non-safety Grade High Rate-of-Change of Withdrawal from a Subcritical or VHP Trip Power Trip Low Power Startup Condition High Pressurizer Pressure Trip High Rate-of-Change of Power Alarms, which initiate Rod Withdrawal Prohibit Action 15.4.2 Uncontrolled Control Rod Bank VHPTrip Pressurizer Safety Valves Withdrawal at Power Operation TM/LP Trip Steam Generator Safety Valves Conditions High Pressurizer Pressure Trip Pressurizer Spray and Level Control Control Rod and Bank Deviation Alarms 15.4.3 Control Rod Misoperation TM/LP Trip Pressurizer Safety Valves Low Steam Generator Water Level Trip Steam Generator Safety Valves
- Safety Injection Actuation Pressurizer Spray and Level Control Signal Control Rod and Deviation Alarms eves 15.4.4 Startup of an Inactive VHP Trip Administrative Procedures for Startup of Loop TM/LP Trip an Idle Pump. Plant Operation with less than all four primary coolant pumps is not permitted by Technical Specifications except for very short periods of time and at reduced power levels (Tech Spec Table 2.3.1). 15.4.6 Chemical Volume and Control VHP Trip Non-safety Grade High Rate-of *Change of System (CVCSI Malfunction that TM/LP Trip Power Trip Results High Pressurizer Pressure Trip Administrative Procedures in a Decrease in the Boron Sufficient Operator Response Time Concentration in the Reactor Coolant 15.4.7 Inadvertent Loading and Operation Technical Specification measurement of a Fuel Assembly in an Improper requirements and administrative Position procedures preclude occurrence 15.4.8 Spectrum of Control Rod Ejection VHP Trip Non-safety Grade High Rate-of-Change of Accidents TM/LP Trip Power Trip Long Term, Safety Injection eves Actuation Signal Siemens Power Corporation
-Nuclear Division
.... * *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-59 Table 15.0.9.1 Overview of Plant Systems and Equipment Available for Transient and Accident Conditions (continued)
SRP No. Event Reactor Trip Functions Other Signals/Equipment 15.5 INCREASE IN REACTOR COOLANT INVENTORY 15.5.1 15.5.2 Inadvertent Operation of the ECCS/CVCS Malfunction that Increases Reactor Coolant Inventory VHP Trip TM/LP Trip High Pressurizer Pressure Trip Non*safety Grade High Rate*of-Change of Power Trip Pressurizer Safety Valves Overpressurization Mitigation System (Operating Conditions 4 through 61 15.6 DECREASE IN REACTOR COOLANT INVENTORY 15.6.1 15.6.3 15.6.5 Inadvertent Opening of a PWR Pressurizer TM/LP Trip Safety Injection System Pressurizer Heaters eves Pressure Relief Safety Injection Actuation Signal Valve Steam Generator Tube Failure Loss of Coolant Accidents Resulting from a Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary TM/LP Trip Safety Injection Actuation Signal Steam Generator Safety Valves Main Steam Line Isolation Valves Atmospheric Steam Dump Controller Steam Bypass to Condenser Controller Auxiliary Feedwater System eves No credit taken for a reactor trip ECCS by the RPS due to the rapid Auxiliary Feedwater System depletion of the moderator which shuts Containment Isolation down the r.eactor core almost Containment Spray and Air Cooler irrmediately, followed by ECCS injection which contains sufficient boron to maintain the reactor core in a subcritical configuration . Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-60 Table 15.2.1.1 Sequence of Events for the Loss of External Load Secondary Side Pressurization Event Time Value Event Initiation
0.0 Turbine
Trip 0.0 Loss of Main Feedwater
0.1 Reactor
Scram on High Pressurizer Pressure 7.6 Peak Reactor Power 8.1 108% MSSVs Open Loop 1: Bank 1 3.2 Bank 2 7.4 Bank 3 Loop 2: Bank 1 3.3 Bank 2 7.4 Bank 3 Pressurizer PORVs Open 9.0 Peak Pressurizer Pressure 9.0 2377.8 psia Peak Secondary side Pressure 13.7 1063.25 psia Event Terminates 20.0 Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 3-61 Table 15.4.3.1 Summary of Results for Control Rod Misoperation Events Maximum Event (Power) Condition 8 MDNBR LHR (kW/ft) Dropped Control Rod ( 100%) 1.62 17.59 Dropped Control Bank ( 100%) 1 1.50 18.88 Statically Misaligned Control Rod 1 Bounded (Dropped (100%) Control Rod-100%) Statically Misaligned Bank (50%) 1 Bounded (Dropped Control Rod-100%) Statically Misaligned Bank (65%) 1 Bounded (Dropped Control Rod-100%) Single Rod Withdrawal (91.5%) 1 1.27 19.67 Single Rod Withdrawal (50%) 1 Bounded (Single Rod Withdrawal from Rated Power) Single Rod Withdrawal (0%) 2 Bounded (Event 15.4.1) Single Rod Withdrawal (0%) 3 Subcritical Core Barrel Failure ( 100%) 1 Bounded (Event 15.4.8) a These operating modes are defined in Table 15.0.2.1.
Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events ..... 1.0 Q,) 3:: 0 a.. 15 *x < .75 ""O Q,) .!::! 15 E ..... 0 :z ,5 .o .1 A 3 Fraction of Active Fuel Height Figure 15.0.3.1 Limiting Axial Power Shape (100% Power) Siemens Power Corporation
-Nuclear Division .9 EMF-98-013 Revision 0 Page 3-62 1.0
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events .... 1.0 Cl) 31:: 0 a.. c *x < .75 ""O Cl) c E .... 0 .5 z .o .1 A .5 3 Fraction of Active Fuel Height Figure 15.0.3.2 Limiting Axial Power Shape (90% Power) Siemens Power Corporation
-Nuclear Division .9 EMF-98-013 Revision 0 Page 3-63 1.0 Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 1.0 .+J *-> .+J .75 u a Q) 0::: E a I... .5 u (/) -a Q) N a .25 E I... 0 z .0 .0 .5 1.0 1.5 2.0 2.5 Time (seconds)
Note: Time measured from the point at which the control rod drive clutch receives the signal to release the control rods. EMF-98-013 Revision 0 Page 3-64 .3.0 Figure 15.0.6.1 Integrated Scram Worth With Most Reactive Rod Stuck Out Siemens Power Corporation
-Nuclear Division Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 1.2 1.1 (-0.080, 1.080 I.. 1.0 (-0.080, 1.000) G) ..... ::> 0 0.9 a. ,, 0.8 G) .... c 0:: 0.7 'I-0 0.6 c 0 -0.5 .... 0 c I.. 0.4 u.. 0.3 0.2 -0.6-0.5 -0.4 -0.3 -0.2 -0.1 -o.o 0.1 Axial Shape Index 0.2 0.3 Figure 15.0. 7 .1 TiNLET Limiting Condition of Operation Siemens Power Corporation
-Nuclear Division EMF-98-013 Revision 0 Page 3-65 0.-400, 1 .080) 0.4 0.5
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1"5 Events 1A c 0 0 1,3 c :I LL. a.. 1.2 EMF-98-013 Revision 0 Page 3-66 1t
-.7 -.8 -/J -,4 -.3 -.2 -.1 ,1 .2 .5 Axial Shq>e Index (ASI) Figure 15.0. 7 .2 Axial Shape Function for TM/LP Trip Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 'I"'" a::: 0 1 * .4 1.2 1.0 .8 .6 . .4 .o .25 .5 .75 1.0 Power {Fraction of Rated) Figure 15.0.7.3 Radial Function for TM/LP Siemens Power Corporation
-Nuclear Division {1.2, 1.2) 1.25 EMF-98-013 Revision 0 Page 3-67 1.5
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 2-400.0 2300,0 ............
2200.0 c *-"' a. Q) 2100.0 L. :::::J "' "' Q) L. a.. 2000,0 1900.0 1800.0 .o 2.0 4,0 6,0 ao 10.0 120 Time (sec) 14,0 16.0 EMF-98-013 Revision 0 Page 3-68 18,0 20.0 Figure 15.2.1.1 Loss of External Load Secondary Side Pressurization
-Pressurizer Pressure Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 570.0 565.0 Q) 560.0 L... ::J ...... 0 L... Q) a. E Q) I-555.0 550.0 545.0 .o 2.0 -4.0 6.0 8.0 1 o.o 12.0 Time (sec) EMF-98-013 Revision 0 Page 3-69 Core Inlet T emperoture SG-1 Exit Temper,oture SG-2 Exit Temperature 14.0 16.0 18.0 20.0 Figure 15.2.1.2 Loss of External Load Secondary Side Pressurization
-Core Inlet and Steam Generator Primary Side Exit Temperatures Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 1100.0 1050.0 '"" 1000.0 0 *-(/) a. '--' Q) 950.0 L.. ::J (/) (/) Q) L.. a.. 900.0 850.0 800.0 .o 2.0 '4.0 6.0 ao 10.0 120 Time (sec) 1-4.0 16.0 EMF-98-013 Revision 0 Page 3-70 18.0 20.C Figure 15.2.1.3 Loss of External Load Secondary Side Pressurization-Secondary Steam Dome Side Pressure Siemens Power Corporation
-Nuclear Division ---------
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 600.0 EMF-98-013 Revision 0 Page 3-71
500.0 ,,.......
0 Q) 400.0 C1J ' E ..a .;::::., 300.0 Q) +' c 0::: == 200.0 ..Q LL.. 100.0 .o .o 2.0 4.0 6,0 ! I I I I I I I I I I I a.o 1 o.o 12.0 Time (sec) MSSV Baik 1 ---MSSV Bcrik 2 MSSV Bcrik 3 14.0 16.0 18.0 20.0 Figure 15.2.1.4 Loss of External Load Secondary Side Pressurization-MSSV Flow Rate for Loop 1 Siemens Power Corporation
-Nuclear Division
- -Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 600.0 500.0 ,,....., 0 G> -400.0 rn ' E .a v 300.0 G> 0 a:::: 3: 200.0 0 LL. 100.0 .o EMF-98-013 Revision 0 Page 3-72 MSsV Bank 1 MSsV Bank 2 MSsV Bank 3 .o 2.0 4.0 8.0 10.0 12.0 14.0 16.0 1 B.O 20.0 Time (sec) Figure 15.2.1.5 Loss of External Load Secondary Side Pressurization-MSSV Flow Rate for Loop 2 Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events 4. References EMF-98-013 Revision 0 Page 4-1 1 . Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG-0800, LWR Edition, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, July 1981 . 2. Advanced Nuclear Fuels Corporation Methodology for Pressurized Water Reactors:
Analysis of Chapter 15 Events, ANF-84-73 Revision 4 Appendix B (P)(A) and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, July 1990. 3. Palisades Cycle 9: Analysis of Standard Review Plan Chapter 15 Events, ANF-90-078, Advanced Nuclear Fuels Corporation, September 1990. 4. Palisades Cycle 10: Disposition and Analysis of Standard Review Plan Chapter 15 Events, EMF-91-176, Siemens Nuclear Power Corporation, October 1991. 5. HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, EMF-92-153(P)(A) and Supplement 1, Siemens Power Corporation, March 1994 . 6. Palisades Cycle 14 Principal Plant Parameters, EMF-97-051, RevisiOIJ 1, Siemens Power Corporation.
- 7. Not used. 8. Palisades Thermal Margin Monitor Analysis, EMF-95-033(P), Revision 1 , Siemens Power Corporation, April 1995. 9. Palisades Final Safety Analysis Report, Updated Version (through Revision 18), Consumers Power Company. 10. Palisades Cycle 13: Disposition and Analysis of Standard Review Plan Chapter 15 Events, Siemens Power Corporation, November 1996. 11. Palisades loss of load Analysis, EMF-93-086 (P), Siemens Power Corporation
-Nuclear Division, April 1993. 1 2. Palisades Modified Reactor Protection System Report: Analysis of Chapter 15 Events, ANF-87-150(NP), Volume 2, Advanced Nuclear Fuels Corporation, June 1988. 13. Review and Analysis of SRP Chapter 15 Events for Palisades with a 15% Variable High Power Trip Reset, ANF-90-181, Advanced Nuclear Fuels Corporation, November 1990. 14. Palisades Plant Technical Specifications, Consumers Power Company, Appendix A to License No. DPR-20. Siemens Power Corporation
-Nuclear Division
- *
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events EMF-98-013 Revision 0 Page 4-2 15. XCOBRA-11/C:
A Computer Code to Determine the Distribution of Coolant During Steady-State and Transient Core Operation, XN-NF-75-21 (A), Revision 2, Exxon Nuclear Company, January 1986. 16. Determination of Palisades Thermal Margin/Low Pressure Trip Coefficients, Combustion Engineering, Inc., September 1971. 17. RODEX2: Fuel Rod Thermal Mechanical Response Evaluation Model, XN-NF-81-58(P)(A), Revision 2 and Supplements 1 and 2, Siemens Power Corporation, March 1 984 .. 18. ANF-RELAP Methodology for Pressurized Water Reactors:
Analysis of Non-LOCA Chapter 15 Events, ANF-89-151 (P)(A), Siemens Nuclear Power Corporation, April 1992. 19. Letter, H. G. Shaw to R. J. Gerling, "Impact of an Inoperable Secondary Valve on the Palisades Loss of Load Analysis," HGS:268:93, July 26, 1993. 20. Not used. 21 . Palisades large Break LOCA/ECCS Analysis, EMF-98-026, Revision 0, Siemens Power Corporation
-Nuclear Division, April 1998. 22. Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations, XN-NF-82-21 (P)(A), Revision 1, Exxon Nuclear, September 1983. 23. Main Steamline Break Analysis for Palisades, EMF-98-012, Revision 0, Siemens Power Corporation, May 1998. 24. Palisades Control Rod Ejection Analysis, EMF-98-021, Revision 0, Siemens Power Corporation
-Nuclear Division, April 1998. 25. Single Phase Hydraulic Flow Test of SPC Palisades HTP Fuel Assembly, EMF-97-003(P), Siemens Power Corporation
-Nuclear Division, September 1997. 26. Palisades Cycle 12: Disposition and Analysis of Standard Review Plan Chapter 15 Events, EMF-95-022, Siemens Power Corporation
-Nuclear Division, April 1 995. 27. Computational Procedure for Evaluating Fuel Rod Bowing, XN-75-32(P)(A) and Supplements 1, 2, 3, and 4, October 1983 . Siemens Power Corporation
-Nuclear Division
- Palisades Cycle 14: Disposition and Analysis of Standard Review Plan Chapter 1 5 Events Distribution Controlled Distribution Siemens Power Corporation
-Nuclear Division Richland D. M. Brown, 22 R. C. Gottula, 36 J. S. Holm, 26 J. D. Martin, 31 W. T. Nutt, 36 K. C. Segard, 36 R. I. Wescott, 38/Consumeis
{ 10) EMF-98-013 Revision 0
- ATTACHMENT NO. 2 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 Excerpts from the Palisades Cycle 14 ABB/CE SBLOCA Analysis 3 Pages J
- Design Analysis Title Page Title: SBLOCA Evaluation of Palisades Cycle 14 Fuel Design Changes Document Number: A-PAL-FE-0001 Revision Number: 00 Quality Class: [8J QC-I (Safety-Related) 0 QC-2 (Not Safety-Related) 0 QC-3 (Not Safety-Related)
I. Approval of Completed Analysis This Design Analysis is complete and verified.
Management authorizes the use of its results. Printed Name Signature Date Cognizant Engineer(s)
Naugab E. Lee iE -0'.i.4,r I Mentor [8J None Independent Reviewer(s)
F. Cohen 3/i;}<jg Management Approval J.M. Cleary 3f 1i.,hs I 2. Package Contents (this section may be completed after Management approval):
Total page count, including body, appendices, attachments, etc.: 68 List associated CD-ROM disk Volume Numbers and path names: 0 None Note: CD-ROM are stored as separate Quality Records CD-ROM Volume Path Names (to lowest directory which uniquely applies to this document)
Numbers a_pal_fe\0001 rOO\out Total number of sheets of microfiche:
[8J None Number of sheets: __ _ Other attachments:
- 3. Distribution:
QR(2), JM Cleary jl 1111 , .. ,.,., [8J None Specify: _______________
_ Proprietary Information Combustion Engineering, Inc.
- *
- Calculation A-PAL-FE-0001, Rev. 00 Page 8 3. Design Inputs The results and conclusions of this analysis, Section 5, are based upon the following design inputs. 3.1 Plant Design Data The analysis of record for the Palisades SBLOCA ECCS performance analysis is Reference
- 4. The changes for Cycle 14 Batch R fuel are documented in Reference
- 3. The changes are shown below: * ** *** 3.2 1. Assembly loss coefficient 15.332* 2. Length of active fuel 11.05 ft** 3. Fuel pellet radius 0.015 ft*** 4. Cladding thickness 0.0020833 ft*** 5. Rod internal gas pressure (cold) 321 psia 6. PCS flow 341400 gpm 7. Core flow 331200 gpm Includes core support plate and fuel alignment plate. More details for this value are provided in Reference
- 16. Top of the fuel will be the same as the previous cycle. Extra length is added to the bottom. Cladding thickness is reduced by increasing the cladding ID. Cladding OD remains the same. Fuel pellet radius is increased so that the gap between the cladding and fuel remains the same. Computer Codes This recorded calculation uses the computer codes listed below. The computer code certificates are included in the references listed below. 1. CEFLASH-4AS jl It It ,.,1*** Version: Computer:
Reference:
Code Certificate:
F4S.l.2 HP/9000 HP-UX OS PA-RISC Reference 8 Reference 9 Proprietary Information Combustion Engineering, Inc.
- L Calculation A-PAL-FE-0001, Rev. 00 Page 45 5. Results and Conclusions The results and conclusions of this analysis are based upon the analysis presented in Section 4. 5.1 Summary of Results A summary of the PARCH/EM results for PCT for this analysis is provided in the ti.owing table: Reference 4 Current Analysis (Pages 209 and 211) Break Size (ft 2) at PD 0.10 0.10 Peak Clad Temperature (PCT) (°F) 1991.9 2025.8 Time to PCT (seconds) 1606.2 1520.2 PCT Node 20 20 Peak Oxidation
(%) 4.79 5.8 Peak Oxidation Node 19 19 Core Wide Oxidation(%)
< 0.71 < 0.842 PLHGR (Kw/ft) 15.8 15.8 Detailed results of the analysis are contained in the output of the computer cases. The cases (inputs and outputs) are archived on the ABB CE CD-ROM archiving system at the path shown on the cover page. Table 5-1 summarizes important parameters used in the Palisades small break LOCA ECCS performance analysis.
Table 5.2 summarizes the results of the limiting break small break LOCA of the Palisades ECCS performance analysis.
Tables 5-1 and 5-2 correspond to Tables 5.0-3 and 5.0-4 of Reference 4 respectively with proper changes for the parameters and results from this analysis.
It must be noted that this CEFLASH-4AS analysis models the whole.core as the Batch R type fuel. This is an acceptable approach because CEFLASH-4AS models the whole core in the average sense and the mixed core effect is not important for the slow small break LOCA transients.
Jl 1111 lll919 Proprietary Information Combustion Engineering, Inc.
ATTACHMENT No. 3 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 Consumers Energy Letter GEJ*97*17 "New PCS and Core Flow Assumptions" 3 Pages
- Consumers Eney A CMS Energy Company Bob Wescott Palisades Nuclear Plant 217J30 Blue Star Memorial Highway Covert. Ml 49043 Siemens Power Corporation
-Nuclear Division P.O. Box 130 Richland, WA 99352-0130
SUBJECT:
New PCS and Core Flow Assumptions October 24, 1997 This memo provides new PCS and core flowrates to be used in all future safety analyses.
When all FSAR Chapter 14 events have been reanalyzed with these newflowrates, we will submit a technical specification change request. The development of these new flowrates is included as an attachment.
- PCS flow rate technical specification limit PCS flow rate analytical limit Core flow rate analytical limit 352,000 gpm 341,400 gpm 331,200 gpm If you have any questions, please call me at 616-764-2408.-
GE Jarka -Palisades GEJ97*17 cc: .. GABaustian
- Sandy Phillips Bill Nutt (SPC) TA Meyers Attachment to GEJ97*17 INPUTS: PCS flow rate limit at HZP = 140.7 M-lbm/hr Ref: T.S. 3.1.1.c Hot Zero Power (HZP) conditions:
2060 psia & 532 °F Maximum core inlet temperature allowed by the Tinlet LCO equation:
543.64 ° F HOT Full Power (HFP) conditions for the safety analyses:
2060 psia & 543.64 ° F (Values consistent with EMF-91-176, Table 15.0.2-2 and EMF-97-051, Table 2.1) coolant specific volume (ft3/lbm) 2060 psia & 532 °F = 0.020894 2060 psia & 543.64 °F = 0.021213 1967 ASME Steam Tables (PC Program) Verification of HFP PCS flow rate used by Siemens for Palisades safety analysis -calculations HFP flow rate= HZP flowrate
- specific volume at HZP specific volume at HFP HFP flow rate= 140.7 M-lbm/hr
- 0.020894/0.021213
= 138.584 M-lbm/hr Accounting for 3% flow rate measurement uncertainty HFP flow rate = 138.6
- 0.97 = 134.4 M-lbm/hr (Value used in safety analysis)
REVISED PCS FLOW RATE LIMIT
- Flow Rate (gpm) = Flow Rate (lbm/hr) *specific volume (ft 3/lbm)
- 7.4805 gal/ft 3 * (1 hr/60 min) HZP Condition!?_:.*
Minimum PCS pressure = 2060 -50 = 2010 psia (Pressurizer pressure)
The 22 psi pressure measurement uncertainty was not used in this calculation since 1) the nominal operating pressure is 2060 psia, 2) there are several PCS pressure indications available in the control room and 3) the operators will generally use the lowest believable PCS pressure indication when they are purposely maintaining the PCS pressure less than 2030 psi a.
- Attachment to GEJ97*17 (continued)
Core inlet pressure will be higher than the pressurizer pressure due to core and vessel outlet pressure losses and elevation head between the hot leg and the pressurizer steam volume. The reactor core and vessel outlet pressure losses are given in PTR-4 as 11 and 11.5 psi respectively.
For a nominal pressurizer water level of 57%, the elevation head to the hot leg centerline is about 23 feet (641 -618). The average coolant temperature in the surge line is approximated as being the average between the pressurizer saturation temperature of 637 F and hot leg temperature of 593 F . Specific volume for the surge line at 201 O psia and 615 Fis 0.024123 ft3/lbm. Using this data, the corresponding pressure head for the elevation head of 23 feet is 6.6 psi. Therefore, the core inlet pressure will be 29.1 psi higher than the PCS pressure.
Since volumetric flow rate increases as pressure decreases, a pressure differential of 25 psi will be used in the following calculations.
Minimum core inlet pressure = 2010 psi a + 25 psi = 2035 psia Specific volume at 2035 psia and 532 F = 0.020900 ft3/lbm Flow Rate= 140.7 M-lbm/hr
- 0.0209
- 7.4805 / 60 = 366,623 gpm HFP Conditions:
Since there are several cold leg temperature indications available in the control room and the TMM monitors the highest cold leg temperature, the appropriate value to be used in the following calculation is 543.64 F rounded up to 544 F. The minimum core inlet pressure is bounded by the above calculation of 2035 psia. Specific volume at 2035 psia and 544 F = 0.021220 ft3/lbm Flow Rate= 138.584 M-lbm/hr
- 0.02122
- 7.4805 / 60 = 366,638 gpm To allow sufficient margin in future PCS flow rate measurements to verify technical specification requirements, the 366,638 gpm is reduced by approximately 4% to 352,000 gpm. The analytical limit for the converted PCS flow rate is equal to the minimum PCS flow rate reduced by 3 % for measurement uncertainty.
The allowed value and the analytical limit are the same siac;e _the PCS flow_ rate is a steady state input value to the safety analysis.
Analytical limit for PCS flow rate= 352,000 * .97 = 341,400 gpm The analytical limit for core flow rate is equal to the analytical limit for PCS flow rate . reduced by 3 % for the core bypass flow. Analytical limit for core flow rate= (352,000 * .97) * .97 = 331,200 gpm