ML062130032
| ML062130032 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities  |
| Issue date: | 05/31/2006 |
| From: | General Electric Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| DRF 0000-0042-5222, GE-NE-0000-0053-8435-RINP, Rev 1 | |
| Download: ML062130032 (64) | |
Text
ATTACHMENT 5 GE-NE-0000-0053-8435-Ri NP, "Dresden 2 & 3 and Quad Cities I & 2 Safety Valve Setpoint Tolerance Relaxation," May 2006
GE Nuclear Energy GE-NE-0000-0053-8435-RINP Revision I DRF 0000-0042-5222 Class I May 2006 Dresden 2 & 3 and Quad Cities 1 & 2 Safety Valve Setpoint Tolerance Relaxation
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION INFORMATION NOTICE This is a non-proprietary version of the document GE-NE-0000-0053-8435-RINP, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here [
IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between the company receiving this document and GE. Nothing contained in this document shall be construed as changing the applicable contract. The use of this information by anyone other than a customer authorized by GE to have this document, or for any purpose other than that for which it is intended, is not authorized. With respect to any unauthorized use, GE makes no representation or warranty, and assumes no liability as to the completeness, accuracy or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.
Copyright, General Electric Company, 2006.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION TABLE OF CONTENTS ABSTRACT ............................................................................. ...................................... vii 1 INTRODUCTION ................ ............... ............................ ........................... 0.................... 1-I 1.1 PURPO SE ................................................... .................................................. 1-1 1.2 OVERALL EVALUATION APPROACH .................................................................. 1-1 1.3
SUMMARY
AND CONCLUSIONS .......................................................................... 1-2 2 VESSEL OVERPRESSURE/ANTICIPATED OPERATIONAL OCCURRENCE EVALUATION ......................................................................................................... 2-1 2.1 ANALYSIS OVERVIEW ........................................................................................... 2-1 2.2 OVERPRESSURE ANALYSIS .................................................................................. 2-2 2.3 REVIEW OF CHAPTER 15 EVENTS ....................................................................... 2-3 2.4 REVIEW OF EQUIPMENT OUT OF SERVICE OPTIONS ..................................... 2-6 2.5 MARGIN TO SPRING SAFETY VALVE SETPOINT ............................................. 2-7 2.6 ISOLATION CONDENSER AND LOSS OF FEEDWATER FLOW ....................... 2-7 3 ATWS EVALUATION . ................................................................................................ 3-1 3.1 ANALYSIS OVERVIEW .......................................................................................... 3-1 3.2 ANALY SIS INPUTS ................................................................................................... 3-2 3.3 ANALY SIS RESULTS ............................................................................................... 3-8 3.4 CO N CLUSIO N S ........................................................................................................ 3-29 4 ECCS/LOCA EVALUATION ....o......... o.... o............ ............... ... ... ........ ..... o.. ................ 4-1 4.1 ECCS/LOCA PERFORMANCE EVALUATION ...................................................... 4-1 5 CONTAINMENT RESPONSE AND LOADS ANALYSIS ......... 5-1 5.1 CONTAINMENT PRESSURE AND TEMPERATURE FOR DBA LOCA .............. 5-1 5.2 SMALL STEAM LINE BREAKS ............................................................................... 5-1 5.3 IB A AN D SB A ............................................................................................................ 5-2 5.4 NUREG-0783 LOCAL SUPPRESSION POOL TEMPERATURE ........................... 5-2 iii
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 5.5 DBA LOCA HYDRODYNAMIC LOADS ................................................................. 5-3 5.6 SRV DYNAM IC LOAD S ........................................................................................... 5-3 5.7 RIPD EVALU ATION ................................................................................................. 5-3 6 HIGH PRESSURE SYSTEMS PERFORMANCE .................... 6-1 6.1 HIGH PRESSURE COOLANT INJECTION ............................................................. 6-1 6.2 REACTOR CORE ISOLATION COOLING .............................................................. 6-2 6.3 STEAM BYPASS PRESSURE CONTROL SYSTEM .............................................. 6-3 6.3.1 D escription................................................................................................... 6-3 6.3.2 Inputs andAssumptions.............................................................................. 6-3 6.3.3 Evaluation.................................................................................................... 6-4 6.3.4 Conclusion ............................................................................................... 6-5 6.4 STANDBY LIQUID CONTROL SYSTEM ............................................................... 6-5 6.5 SAFE SHUTDOWN MAKUP PUMP SYSTEM ........................................................ 6-6 6.6 ISOLATION CONDENSER SYSTEM ...................................................................... 6-6 7 APPENDIX R ANALYSIS ...... ..... . ....... ... ... ... .................................................. . ...... 7-1 7.1 VESSEL INVENTORY ASSESSMENT .................................................................... 7-1 7.2 CONTAINMENT RESPONSE ASSESSMENT ......................................................... 7-1 8 VESSEL THERMAL CYCLE ASSESSMENT ............................................ .... 8-1 8.1 ANALYSIS OVERVIEW ........................................................................................... 8-1 8.2 INPUTS AND ASSUMPTIONS ................................................................................. 8-1 8.3 AN ALYSIS RESULTS ............................................................................................... 8-1
8.4 CONCLUSION
S AND RECOMMENDATIONS ...................................................... 8-1 9 OPERATING MODES AND EOOS REVIEW . ............ .......................... 9-1 10 EMERGENT EPU ISSUES REVIEW . ............................................. 10-1 11 REFERENCES .... . ... ... . ..... ... ....... ............ ............... ............... .............. ... ... ... ..... 11-1 iv
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION LIST OF TABLES TABLE TITLE PAGE 1-1 Summary of Analyses Presented in this Report 1-3 2-1 Overpressure Results with 3% Setpoint Tolerance 2-2 2-2 Chapter 15 Event Descriptions 2-3 3-1 Summary of ATWS Key Input Parameters 3-3 3-2 Axial Power Shapes 3-4 3-3 Key Equipment Parameters 3-5 3-4 Summary of Key ODYN Parameters for ATWS Calculation 3-8 3-5 Summary of Peak Suppression Pool Temperature, 3-9 Containment Pressure and Integrated SRV Flow 3-6 Sequence of Events for MSIVC at BOC 3-10 3-7 Sequence of Events for MSIVC at EOC 3-11 3-8 Sequence of Events for PRFO at BOC 3-12 3-9 Sequence of Events for PRFO at EOC 3-13 3-10 Acceptance Criteria Results 3-14 3-11 Peak Pressures for Other System Evaluations 3-15 v
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION LIST OF FIGURES FIGURE TITLE PAGE 3-1 MSIVC - BOC - GE14 Fuel 3-16 3-2 MSIVC - BOC - GE14 Fuel 3-17 3-3 MSIVC - BOC - GE14 Fuel 3-18 3-4 MSIVC - EOC - GEl4 Fuel 3-19 3-5 MSIVC - EOC - GE14 Fuel 3-20 3-6 MSIVC - EOC - GE14 Fuel 3-21 3-7 PRFO - BOC - GE14 Fuel 3-22 3-8 PRFO - BOC - GE14 Fuel 3-23 3-9 PRFO - BOC - GE14 Fuel 3-24 3-10 PRFO - EOC - GE14 Fuel 3-25 3-11 PRFO - EOC - GE14 Fuel 3-26 3-12 PRFO - EOC - GE14 Fuel 3-27 3-13 Containment Response MSIVC EOC 3-28 3-14 Containment Response PRFO EOC 3-28 vi
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION ABSTRACT This report summarizes the analysis results that support the operation of Dresden Units 2 and 3, and Quad Cities Units 1 and 2, with a setpoint tolerance increase from 1% to 3% for the Safety function of the Target Rock Dual Mode Safety Relief Valves and the Dresser Spring Safety Valves.
This report specifically addresses several analyses/subject areas that are sensitive to the valve setpoint tolerances. Other subjects that are insensitive to the valve setpoint tolerance change are not addressed in this report.
Several requirements were identified in this report in order to implement the setpoint tolerance increase. These requirements are summarized in the introduction section of the report.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 1 INTRODUCTION 1.1 ' PURPOSE Reference I presents a generic evaluation of the effects of increasing the setpoint tolerance of the Safety Relief Valves and identifies specific areas that should be evaluated on a plant specific basis. This report provides the results of the plant specific evaluations performed to assess the impact of the setpoint tolerance increase. These evaluations support the operation of Dresden Units 2 and 3 and Quad Cities Units 1 and 2 with an increase in the setpoint tolerance for the safety function of the Target Rock Dual Mode Safety Relief Valves (SRV) and the Dresser Spring Safety Valves (SSV) from 1% to 3%. The increase in setpoint tolerance includes both an increase in the upper limit of the setpoint tolerance as well as a decrease in the lower limit of the setpoint tolerance. The upper limit is defined as +3% and the lower limit is defined as -3%.
1.2 OVERALL EVALUATION APPROACH The impact of the SRV setpoint tolerance increase on the following subjects is addressed in this report:
- Vessel Overpressure
- UFSAR Chapter 15 Events
- Anticipated Transients Without Scram (ATWS)
- LOCA
- Containment Response and Loads
- High Pressure Systems Performance
- Appendix R Fire Protection
- Vessel Thermal Cycle
- Operating Mode and Equipment Out Of Service (EOOS) Review
- Extended Power Uprate (EPU) Evaluation Review These subjects are affected by the increase in valve setpoints associated with the setpoint tolerance change from 1 to 3%.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 1.3
SUMMARY
AND CONCLUSIONS A summary of the results of the evaluations for each of the subjects of concern is provided in Table 1-1. The evaluation determined that the impact of the setpoint tolerance increase on the following subjects are acceptable: 1) Vessel Overpressure, 2) USAR Chapter 15 Events, 3)
ATWS Analysis, 4) ECCS/LOCA Performance, 5) Containment Response and Loads Assessment, 6) High Pressure Systems Performance, 7) Appendix R Fire Protection, 8) Vessel Thermal Cycle, 9) Plant Operating Modes and EOOS, and 10) EPU Project Impact. These specific subjects were addressed in detail as described in this report.
Based on the results of the different analyses described in this report, several areas require further evaluation for implementation of the setpoint tolerance increase. The subjects that require additional evaluation are identified in Table 1-I and will be addressed by Exelon before the implementation of the setpoint tolerance increase.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Table 1-1: Summary of Analyses Presented in this Report Subject Section J Result Vessel Overpressure, Transient Analysis 2.0 Acceptable' and Spring Safety Valve Margin ATWS Analysis 3.0 Acceptable' ECCS/LOCA Evaluation 4.0 Acceptable Containment Response and Loads 5.0 Acceptable 2 Analysis High Pressure Systems Performance 6.0 Acceptable 3.4 Appendix R Analysis 7.0 Acceptable Vessel Thermal Cycle Assessment 8.0 Acceptable Operating Modes and Equipment Out of 9.0 Acceptable Service Review Emergent Extended Power Uprate Issues 10.0 Acceptable Review
- 2. SRV Dynamic Loads will be assessed by Exelon to ensure the requirements described in section 5 are met.
- 3. MOV operation will be assessed by Exelon to ensure the requirements described in section 6 are met
- 4. The Standby Liquid Control System performance will be assessed by Exelon to ensure the requirements described in section 6 are met.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 2 VESSEL OVERPRESSURE/ANTICIPATED OPERATIONAL OCCURRENCE EVALUATION 2.1 ANALYSIS OVERVIEW Reference I presents a generic evaluation of the effect of increasing the setpoint tolerance to
+/- 3% for safety valves in the pressure relief system. This section presents the results of the plant specific evaluations associated with the increase of the setpoint tolerance of the safety valves from +/- 1 to +/- 3%. In this section Safety Valves (SV) are defined as valves that are qualified for use in the ASME overpressure protection analysis and include the spring safety valves (SSV) and the safety function of the Target Rock Dual Mode Safety Relief Valve (DSRV). In addition to the plant specific overpressure analysis, a plant specific review of the events in Chapter 15 of the FSARs was performed to determine if any other events are impacted by the setpoint tolerance increase. This review is summarized in Table 2-2. Based on the generic evaluation in Reference I and the review of the Chapter 15 events in Table 2-2, the overpressure analysis was evaluated with the safety valves at the +3 % limit and the Loss of Feedwater Event was reviewed with the safety valve setpoints at the -3% limit. All other events were determined to be unaffected by the change in setpoint tolerance.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 2.2 OVERPRESSURE ANALYSIS The most recent MSIVF transients, for Dresden Units 2 and 3 and Quad Cities Units 1 and 2, were analyzed with a 3% safety valve setpoint tolerance. The results of these analyses are provided in Table 2-1 below. These results demonstrate that the dome pressure safety limit (1345 psig) and the peak vessel pressure limit (1375 psig) are met when analyzed with a 3%
setpoint tolerance. The Overpressure analyses were performed in accordance with the methodologies described in Reference 2.
Table 2-1: Overpressure Results with 3 % Setpoint Tolerance Plant Power Flow # SSVs #DSRVs Peak Peak Basis (o Rated) (0 Rated) Credited Credited Dome Vessel Pressure Pressure (psig) (psig)
Dresden 2 102 108 8 0 1339 1365 Cycle 20 Reload 95.3 8 0 1339 1361 Licensing Results Dresden 3 102 108 8 0 1323 1351 Cycle 19 Reload 95.3 8 0 1323 1348 Licensing Results Quad 102 108 8 1 1342 1366 Cycle 19 Reload Cities 1 95.3 8 1 1340 1362 Licensing Results
- Quad 102 108 8 1 1339 1362 Cycle 18 Reload Cities 2 95.3 8 1 1339 1360 Licensing Results *
- The Quad Cities I and Quad Cities 2 results include the effects of the Acoustical Side Branch Modification.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 2.3 REVIEW OF CHAPTER 15 EVENTS The following table describes the impact of the setpoint tolerance on the events in Chapter 15 of the FSAR.
Table 2-2: Chapter 15 Event Descriptions Increase in Heat Removal by the Reactor Coolant System Loss of FW Heater (LFWH)
Manual Flow Control (MFC) This transient results in a power increase due to increased core inlet subcooling. The increase in reactor power occurs at a moderate rate. No safety or relief valve actuation occurs during this transient. Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Auto Flow Control (AFC) MFC is more severe than AFC because AFC would limit the power increase. No safety or relief valve actuation occurs during this transient. Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Feedwater Controller Failure Maximum Demand (FWCF) This transient is similar to a Turbine Trip, however it is initiated at a higher power. This transient is analyzed on a reload specific basis for CPR as well as for pressure margin to the unpiped safety valve setpoints. ((
)) Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Increase in St*am Flow Pressure Regulator Failure Upscale This event results in a decrease in vessel pressure followed by a low pressure isolation.
)) The vessel pressure increase is bounded by the Main Steam Isolation Valve Closure with direct scram which does not result in safety valve actuation. Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Decrease in Heat Removal by the Reactor Coolant System Pressure Regulator Failure Downscale Backup pressure regulator controls pressure. This event results in a small pressure change and power perturbation. No safety actuation occurs during this transient.
Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Load Rejection With Bypass (LRWBP) Severity varies with BPV capacity and the results are bounded by the Load Rejection without Bypass event.
Without Bypass (LRNBP) This transient results in a large vessel pressurization and increase in reactor power and is analyzed on a reload specific basis for CPR as well as for pressure margin to the unpiped safety valve setpoints. ((
)) Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Turbine Trip With Bypass (TTWBP) Severity varies with BPV capacity and the results are bounded by the Turbine Trip without Bypass event.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Without Bypass (TINBP) This transient results in a large vessel pressurization and increase in reactor power and is analyzed on a reload specific basis for CPR as well as for pressure margin to the unpiped safety valve setpoints. ((
)) Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
MSIV Closure Direct Scram (MSIVD) This transient is not limiting from a CPR perspective because of thc slow steam flow shutoff rate associated with the MSIV stroke times. This transient is analyzed on a cycle specific basis to determine the pressure margin to unpiped spring safety valve setpoints.
)) Therefore, this transient is not impacted by the safety valve setpoint tolerancc change.
Flux Scram (MSIVF) This transient is analyzed on a cycle specific basis to ensure that the ASME boiler code requirements and dome pressure tech spec. safety limits are met. The peak vessel pressure increases as the Safety Valve opening setpoints are increased. This transient has been analyzed using the upper bound of the 3 % tolerance for the spring safety valve opening setpoints and the safety mode of the dual mode relief valve opening setpoints.
Single MSIV Closure This event is bounded by the MSIVD transient for peak pressure and is a non-limiting MCPR transient compared to other analyzed pressurization events. No safety valve actuation occurs during the transient Therefore, this transient is not impacted by the
_safety valve sctpoint tolerance change.
Loss of Condenser Vacuum This event is similar to a Turbine Trip event with no bypass, but there is a period of time where bypass valve flow is available. The duration of the bypass valve flow depends on the rate of loss of vacuum. Because of the limited bypass flow, the event is less severe than a turbine trip without bypass. No safety valve actuation occurs during the transient.
Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Loss ofAuxiliary Power This is a delayed turbine trip with recirculation pump trip. No safety valve actuation occurs during the transient Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Loss of Feedwater Flow (LOFW) This transient results in a low level scram followed by a low-low level isolation. The transient is not limiting from a CPR perspective and because of the time delay between the scram and the MSIV closure, this event is far from limiting from an overpressure concern. This event does not result in safety valve actuations so the increased setpoint tolerance does not result in higher peak pressures. The Bernoulli effect on the L3 setpoint is not impacted by the setpoint tolerance change because the L3 setpoint is reached before any valve actuation occurs. The effect of the increased setpoint tolerance on the initiation of flow to the isolation condenser was also evaluated.
Decrease in Reactor Coolant System Flow Rate Trip of One Pump Motor Field Breaker This event results in a pump coastdown and power decrease. No safety valve actuation occurs during the transient. Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Line Breaker This event results in a pump coastdown and power decrease. No safety valve actuation occurs during the transient. Therefore, this transient is not impacted by the safety valve sctpoint tolerance change.
Trip of All Recire Loops Drive Motors This event results in a flow coastdown and power decrease and may result in high Level Turbine Trip after a significant power decrease. No safety valve actuation occurs during the transient. Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Pump Motors This event results in a flow coastdown and power decrease and may result in high Level Turbine Trip after a significant power decrease. No safety valve actuation occurs during the transient Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Recirculation Flow Controller Malfunction Decreasing Flow This event is similar to field breaker trip and results in a power decrease. No safety valve actuation occurs during the transient Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Shaft Seizure Two Loop Operation This event results in a rapid flow decrease which causes reactor power to decrease. No safety valve actuation occurs during the transient. Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Single Loop Operation This event results in a rapid flow decrease which causes reactor power to decrease. No safety valve actuation occurs during the transient. Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Recirculation Pump Shaft Break This event results in a rapid flow decrease which causes reactor power to decrease. No safety valve actuation occurs during the transient. Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Jet Pump Malfunction The event results in very small change (decrease) to core flow which causes reactor power to decrease. No safety valve actuation occurs during the transient Therefore, this Itransient is not impacted by the safety valve setpoint tolerance change.
Reactivity and Power Distribution Anomalies Control Rod Withdrawal Error During Startup This transient results in a power increase from very low powers. The increase in reactor power can occur at a high rate, but the neutron monitoring system is designed to limit the peak power achieved during the transient The peak powers achieved are sufficiently low such that no safety valve actuation occurs during this transient. Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
At Power This transient results in a power increase due to increased reactivity associated with the control rod withdrawal. The increase in reactor power occurs at a moderate rate. The pressure regulator maintains vessel pressure and no safety valve actuation occurs during this transient Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Startup of an Inactive Recire Loop ((
)) The pressure regulator maintains vessel pressure and no safety valve actuation occurs during this transient.
Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Flow Controller Failure - Increasing The rapid flow increase results in a power increase that occurs at a moderate rate. The Flow pressure regulator maintains vessel pressure and no safety valve actuation occurs during this transient, Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Slow Flow Runout The slow flow runout transient is not an original Chapter 15 FSAR event, ((
)) This event assumes a slow increase in recirculation flow rate in both loops from the minimum core flow to the maximum core flow. This analysis is a conservative process for evaluating flow runout events.
The slow increase in core flow causes an increase in reactor power and corresponding increase in steam flow. The pressure regulator maintains vessel pressure and no safety valve actuation occurs during this transient. Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Mislocated Fuel Assembly Accident This scenario is modeled with a 3 dimensional core simulator code. The event does not result in increased pressure or safety valve actuation. Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Misoriented Fuel Assembly This scenario is modeled with a 3 dimensional core simulator code. The event does not Accident result in increased pressure or safey valve actuation. Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Control Rod Drop Accident This results in a very rapid increase in neutron flux and a corresponding increase in fuel temperature. A reactor scram terminates the transient The pressure regulator maintains vessel pressure. No safety actuation occurs during this transient Therefore, this transient is not impacted by the safety valve setpoint tolerance change.
Increase in Coolant Inventory Inadvertent -PCI This event is analyzed on a reload specific basis for CPR and margin to unpiped SSV.
This is an event where the HPCI system is inadvertently initiated. The increased core subcooling causes power to increase. [I
)) It is possible that the inadvertent HPCI initiation could cause water level to increase to the Level 8 setpoint resulting in a turbine trip. This event is similar to the FWCF. In either case, no safety valve actuation occurs. ((
)) Therefore, this I transient is not impacted by the safety valve setpoint tolerance change.
Decrease in Reactor Coolant Inventory One RV/SV Opening Event is not limiting with respect to MCPR or fuel duty because the event results in a very small power change. The event is analyzed for the highest single valve capacity and the highest single valve capacity is not changed with the safety valve setpoint tolerance change.
Instrument Line Break These events are considered in the Loss of Coolant Analysis section of this report.
Steam Line Break Outside Containment LOCA Inside Containment Radioactive Release from a Subsystem or Component Liquid Release due to Tank Failure These events are evaluated for radiological consequences and are not affected by the Fuel Handling Accident safety and relief valve setpoint tolerance increase.
Spent Fuel Cask Drop Accident 2.4 REVIEW OF EQUIPMENT OUT OF SERVICE OPTIONS In addition to the events in Chapter 15 of the FSAR, the following equipment out-of-service options were considered when determining the impact of the setpoint tolerance change:
I. Turbine Bypass OOS
- 4. Single Recirc Loop
- 5. Power Load Unbalance OOS
- 7. Pressure Regulator OOS
- 9. MSIV Out of Service Various combinations of equipment out-of-service options are allowed as described in Reference
- 3. These flexibility options are considered when performing critical power ratio and peak vessel pressure analyses. ((
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION
)) Therefore, the increase in the safety valve setpoint tolerance does not impact the critical powers for the equipment out of service options listed above. The Turbine Bypass Out-of-Service option considers the effects of not meeting the fast response performance analyzed for the reload. It does not remove the ability of the pressure regulator to open the bypass valves in an attempt to maintain vessel pressure for slow events such as the rod withdrawal error or loss of feedwater heating where core power and steam flow may increase above the rated value.
For vessel overpressure calculations, the limiting event is the Main Steam Isolation Valve Closure with flux scram. This transient is evaluated from 102% of rated power at the high and low end of the rated power licensed core flow. The overpressure results are bounding for the equipment out-of-service options listed above.
The ADS system relies on the relief valves and is not impacted by the safety valve setpoint tolerance increase.
Therefore, the equipment out-of-service options listed above are not impacted by the valve setpoint tolerance increase.
2.5 MARGIN TO SPRING SAFETY VALVE SETPOINT
((I 2.6 ISOLATION CONDENSER AND LOSS OF FEEDWATER FLOW The loss of feedwater event relies on the reactor core isolation cooling system or the isolation condenser in order to maintain sufficient coolant inventory to ensure water level remains above Top of Active Fuel (TAF). This event was analyzed during the implementation of EPU. The results of the analysis were used to ensure that flow would initiate to the isolation condensers during the Loss of Feedwater Analysis. During the Loss of Feedwater event, only the lowest set of relief valves actuate. If additional relief valves opened, the reactor vessel pressure profile would be affected and the initiation of flow to the isolation condenser could also be affected.
The increase in setpoint tolerance only applies to the spring safety valves and the safety mode of the Target Rock Dual Mode S/RV. The relief valve setpoint tolerances remain unaffected. The increase in setpoint tolerance lowers the low end of the setpoint tolerance band for the Target Rock Dual Mode S/RV. Based on the nominal setpoint of 1135 psig for the Target Rock Dual Mode S/R, the low end of the 3 % setpoint tolerance is 1101 psig or 1115.7 psia- This is higher than the peak Reactor Pressure of 1099.9 psia in the previous LOFW analysis. Therefore, the Target Rock Dual Mode S/RV will not lift during the LOFW event with the 3% setpoint tolerance. Based on this information, the Loss of Feedwater event and associated initiation of flow to the isolation condenser is not impacted by the increase in setpoint tolerance.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 3 ATWS EVALUATION 3.1 ANALYSIS OVERVIEW This section describes the impact of the setpoint tolerance increase on the Dresden and Quad Cities ATWS analysis.
The ATWS analysis is performed in order to demonstrate that reactor integrity, containment integrity, and fuel integrity are maintained for scenarios where an automatic SCRAM fails to occur. Reactor integrity is demonstrated by ensuring that peak reactor vessel pressure is within the ASME Service Level C limit of 1500 psig. Containment integrity is demonstrated by ensuring that the peak suppression pool temperature is below the maximum bulk suppression pool temperature limit of 202TF and containment pressure is less than the containment design pressure limit of 62 psig. Fuel integrity is demonstrated by ensuring that peak cladding temperature is below the 10CFR50.46 limit of 2200°F and fuel local cladding oxidation is below the 10CFR50.46 limit of 17 % total clad thickness. Because the cladding temperature increase for ATWS is of short duration and limited magnitude, cladding oxidation is not explicitly calculated in the ATWS analysis.
)) The ATWS analysis performed during the implementation of EPU demonstrated that all acceptance criteria listed above were met.
The increased setpoint tolerance associated with the Spring Safety Valves increases the upper analytical limit of the Spring Safety Valve setpoints. This increase in setpoint tolerance alone will tend to increase the peak vessel pressure during the ATWS events as well as the subsequent pressure peaks as Spring Safety Valves cycle to assist in maintaining vessel pressure. Both the upper limit (+3%) and the lower limit (-3%) of the setpoint tolerance band were considered.
((
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION
))The MSIVC and PRFO transients were re-evaluated using the updated design inputs summarized in Table 3-3. The MSIVC and PRFO events were re-analyzed at BOC and EOC As part of the EPU ATWS analysis, the drywell temperature response was evaluated and a peak drywell airspace temperature of 31 10F was calculated. The drywell temperature of 31 10 F is well below the equipment qualification temperature limit of 340°F. The drywell temperature analysis resulted in a peak drywell airspace pressure of 34.2 psig and a corresponding peak wetwell pressure of 38.4 psig at the bottom of the wetwell. This pressure is well below the 62 psig containment design limit and demonstrates that the peak containment pressure design limit is met. The EPU drywell temperature analysis resulted in a peak containment shell temperature of 280*F which is below the shell temperature limit of 2810 F. ((
)) therefore, the peak drywell temperature analysis is not affected by an increase in the setpoint tolerance of the unpiped spring safety valves. Additionally, the electromatic relief valve delay times have been reduced from the values used during the EPU analysis. This will not impact the drywell temperature evaluation significantly, but would tend to increase the percentage of steam flow to the suppression pool. Therefore the drywell temperature evaluation performed in conjunction with the ATWS analysis during the implementation of EPU remains bounding for the increased setpoint tolerance. The suppression pool temperatures were evaluated as part of the analysis for the increased setpoint tolerance.
3.2 ANALYSIS INPUTS Table 3-1 summarizes the initial conditions assumed for the ATWS event. These conditions are consistent with the initial conditions assumed for the ATWS analysis performed for the implementation of EPU.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Table 3-1: Summary of ATWS Key Input Parameters Parameter Value Dome Pressure, psia 1020 Rated Core Flow, Mlbm/hr 98.0 Core flow, Mlbm / % of Rated 93.4/95 Rated Power, MWt 2957 Power, MWt / % of Rated 2957/100 Steam Flow, Rated, Mlbm/hr 11.71 Feedwater Temperature, *F 356 Initial Dynamic Void Reactivity Coefficient (EOC Value), //o -11.7 (BOC)
-10.3 (EOC)
Core Average Void Fraction (EOC Value), % 49.5 (BOC) 36.2 (EOC)
Initial Doppler Coefficient (EOC Value), O/°F -0.13 (BOC)
-0.14 (EOC)
Initial Suppression Pool Liquid Volume (19) 111,500 Initial Suppression Pool Temperature (0 F) 98 Initial Suppression Pool Mass, Mlbm 6.916 Initial Inventory in CST, Ibm 740,000 Initial Inventory in Condenser/Hotwell, Ibm 476,000 3-3
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Table 3-2 shows the initial axial power shapes for the beginning of cycle and end of cycle analyses. These axial power shapes are consistent with the initial axial power shapes in the ATWS analysis performed for the implementation of EPU. The ATWS analysis results are based on GEI4 fuel. These analyses are applicable to the current Dresden and Quad Cities cores with GE14 reloads. A small amount of Legacy fuel remains in some cores, however GE14 fuel is the dominant fuel type.
Table 3-2: Axial Power Shapes Node Location BOC (2957 EOC (2957 (From Bottom of Active Fuel) MW/95% ow) Mw Flow) 1 0.37 0.14 2 1.28 0.43 3 1.60 0.50 4 1.68 0.56 5 1.66 0.65 6 1.60 0.76 7 1.53 0.88 8 1.46 1.00 9 1.40 1.11 10 1.34 1.20 11 1.28 1.27 12 1.21 1.33 13 1.14 1.37 14 1.05 1.40 15 0.88 1.28 16 0.81 1.32 17 0.77 1.40 is 0.70 1.44 19 0.63 1.45 20 0.55 1.41 21 0.46 1.29 22 0.36 1.08 23 0.15 0.47 24 0.08 0.26 3-4
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Table 3-3 summarizes key equipment parameters and input values used in the ATWS analysis.
For comparison, Table 3-3 also shows the input values used in the ATWS analysis performed for the implementation of EPU. In addition to the design inputs summarized in Table 3-3, the replacement steam dryer parameters were incorporated into the ATWS analysis as well as the Acoustic Side Branch Modifications. The inclusion of the replacement steam dryer and acoustic side branch modifications is conservative for the ATWS analyses presented in this section. The updated dryer parameters were based on a steam dryer D/P of 0.10 psid and a dryer weight of 100,200 Ibm consistent with Reference 4 and the Acoustic Side Branch Modifications were based on a SRV inlet piping pressure drop of 11 psid for a flow rate of 644,543 Ibm/hr.
Table 3-3: Key Equipment Parameters Parameter Original Re-Analysis EPU Analysis Nominal Closure Time of MSIV, sec 4.0 4.0 Relief Valve System Capacity, % NBR Steam Flow at 1120 18.4/4 18.4/4 psig / No. of Valves Relief Valve Nominal Opening Setpoint Range, psig 1112, 1112, 1115, 1115, (Note 1) 1135, 1135 1135, 1135 Relief Valve Closing Setpoint, % of Opening Setpoint 96 93.2 Relief Valve Time Delay On Opening Signal, sec 1.85 0.677 Relief Valve Opening Stroke Time, sec 0.25 0.25 Relief Valve Closure Time Delay, sec 4.0 4.0 Relief Valve Closure Stroke Time, sec 10.0 10.0 Opening Delay for the 2 lowest setpoint relief valves on 10.0 15.0 subsequent valve cycling, sec.
Safety/Relief Valve System Capacity, % NBR Steam Flow at 5.3/1 5.3 / 1 1125 psig / No. of Valves Safety/Relief Valve Nominal Opening Setpoint, psig 1135 1135 (Note 2) 3-5
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Parameter Original Re-Analysis EPU
_Analysis Safety/Relief Valve Closing Setpoint, % of Opening Setpoint 96 93.2 Safety/Relief Valve Time Delay On Opening Signal, sec 0.4 0.4 Safety/Relief Valve Opening Stroke Time, sec 0.15 0.25 Safety/Relief Valve Closure Time Delay, sec 0.4 0.4 Safety/Relief Valve Closure Stroke Time, sec 10.0 10.0 Safety Valve System Capacity, % NBR Steam Flow at 1240 44.1 / 8 44.1 /8 psig /No. of Valves Safety Valve Nominal Opening Setpoint, psig 1240, 1240, 1240, 1240, 1250, 1250, 1250, 1250, 1260, 1260, 1260, 1260, 1260, 1260 1260, 1260 Safety Valve Setpoint Tolerance, % 1 3 Safety Valve Closing Setpoint, % of Opening Setpoint 96 96 Safety Valve Opening Stroke Time, sec 0.3 0.3 Safety Valve Closure Stroke Time, sec 0.3 0.3 Recirc Pump Trip Logic Delay and Time Constant, sec 0.60 0.60 SLCS Injection Location: Lower Plenum Standpipe Yes Yes Number of SLCS Pumps 2 2 SLCS Injection Rate per Pump, gpm 40 40 Nominal Boron-10 Enrichment, % 19.8 19.8 Sodium Pentaborate Concentration, % 14 14 Boron Injection Initiation Temperature (BIIT), OF 110 110 SLCS Liquid Transport Time, sec 60 60 3-6
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Parameter Original Re-Analysis EPU Analysis SLCS Liquid Solution Enthalpy, Btu/lbm 78 78 Time to Inject Hot Shutdown Boron Weight, sec 1073 1138 LPCI Flow Rate, gpm 5000 5000 Enthalpy of the HPCI Flow, Btu/lbm 103 103 ATWS High Pressure Setpoint, psig 1250 1250 Low Pressure Isolation Setpoint, psig 825 785 Number of RHR Loops 2 2 Number of RHR Loops for LOOP event I (Note 3) 1 RHR Service Water Temperature 'F 98 98 RHR Heat Exchanger K-Factor per Loop in Containment 343 343 Cooling Mode, Btu/sec- 0F RHR Heat Exchanger K-Factor per Loop during Loss of 343 (Note 3) 343(Note 4)
Offsite Power, Btu/sec- "F Notes:
((
- 3. EPU analyses assumed that one RHR heat exchanger was available with K-Factor of 343 Btu/Sec-oF. If suppression pool temperature limit was exceeded, the number of RHR loops available becomes 2 with a reduced K Factor of 252 Btu/Sec-°F.
- 4. The Re-Analysis for the SRV Tolerance Change assumes only one RHR heat exchanger is available with a K factor of 343 Btu/Sec-oF.
3-7
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 3.3 ANALYSIS RESULTS The ATWS analysis yielded similar results to previous ATWS analyses. The ODYN results from this analysis are summarized in Table 3-4 below. The suppression pool temperature, suppression pool airspace pressure and integrated valve flows are shown in Table 3-5. A sequence of key events was developed for each of the transients analyzed. These are provided in Tables 3-6 through 3-9. Table 3-10 shows the ATWS acceptance criteria and the applicable limiting results. Plots of key ODYN outputs were generated for each of the transient analyzed and these are provided in Figures 3-1 through 3-12. Finally, plots of suppression pool temperature and suppression pool airspace pressure verses time are provided for the MSIVC and PRFO transients at end of cycle in Figures 3-13 through 3-14.
Table 3-4: Summary of Key ODYN Parameters for ATWS Calculation Event Power (MWt) Exposure Peak Neutron Peak Heat Flux Peak Vessel
/Flow (%) Flux (%) ( Press. (Psig)
MSIVC MSIVC PRFO PRFO ]
Notes: Values in the parentheses represent the time of peak values in seconds The peak neutron and heat fluxes are normalized to the respective initial power of the individual cases.
3-8
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Table3-5: Summary of Peak Suppression Pool Temperature, Containment Pressure and Integrated SRV Flow Event Power Exposure Peak Peak Integrated SSV (MWt) Suppression Suppression and RV* Flow
/Flow (%) Pool Pool Airspace at Hot Temperature, OF Pressure, psig Shutdown, Ibm MSIVC ((
MSIVC PRFO PRFO Note:Values in the parentheses represent the time of peak values in seconds.
Values in the brackets represent the hot shutdown time in seconds. The hot shutdown in ODYN ATWS evaluation is defined as neutron flux less than 0.1% for more than 100 seconds.
3-9
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Table 3-6: Sequence of Events for MSIVC at BOC Event Time (s)
MSIV Isolation Initiates MSIVs Closed Peak Neutron Flux (( ]
Opening of the First Relief Valve High Pressure ATWS Setpoint Recirculation Pumps Tripped Peak Heat Flux Occurs (( ]
Peak Vessel Pressure (( ))
BUT Reached Feedwater Reduction Initiated SLCS Pumps Start Boron Solution Reaches Lower Plenum HSBW Injected and Water Level Ramped up Peak Suppression Pool Temperature (( ))
Water Level Restored to Normal Band Hot Shutdown Achieved (Neutron flux below 0.1% for more ))
than 100 seconds) 3-10
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Table 3-7: Sequence of Events for MSIVC at EOC Event Time (s)
MSIV Isolation Initiates U MSIVs Closed Peak Neutron Flux (( ))
Opening of the First Relief Valve High Pressure ATWS Setpoint Recirculation Pumps Tripped Peak Heat Flux Occurs (( ))
Peak Vessel Pressure (( ))
BHT Reached Feedwater Reduction Initiated SLCS Pumps Start Boron Solution Reaches Lower Plenum HSBW Injected and Water Level Ramped up Peak Suppression Pool Temperature (( ]
Water Level Restored to Normal Band Hot Shutdown Achieved (Neutron flux below 0.1% for more ))
than 100 seconds) 3-11
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Table 3-8: Sequence of Events for PRFO at BOC Event Time (s)
Turbine Control and Bypass Valves Start Open MSIV Closure Initiated by Low Steamline Pressure Peak Neutron Flux (( ))
MSIVs Closed Opening of the First Relief Valve High Pressure ATWS Setpoint Tripped Recirculation Pumps Tripped Peak Heat Flux Occurs (( ))
Peak Vessel Pressure (( ))
BHT Reached Feedwater Reduction Initiated SLCS Pumps Start Boron Solution Reaches Lower Plenum HSBW Injected and Water Level Ramped up Peak Suppression Pool Temperature ([ ]
Water Level Restored to Normal Band Hot Shutdown Achieved (Neutron flux below 0.1% for more than 100 seconds) 3-12
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Table 3-9: Sequence of Events for PRFO at EOC Event Time (s)
Turbine Control and Bypass Valves Start to Open MSIV Closure Initiated by Low Steamline Pressure MSIVs Closed Peak Neutron Flux(( ]
Opening of the First Relief Valve High Pressure ATWS Setpoint Tripped Recirculation Pumps Tripped Peak Heat Flux Occurs (( ))
Peak Vessel Pressure (( ))
BUT Reached Feedwater Reduction Initiated SLCS Pumps Start Boron Solution Reaches Lower Plenum HSBW Injected and Water Level Ramped up Peak Suppression Pool Temperature (( ]
Water Level Restored to Normal Band Hot Shutdown Achieved (Neutron flux below 0.1% for more than 100 seconds) 3-13
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Table 3-10: Acceptance Criteria Results Acceptance Allowed Value Limiting Result ATWS Event and Criteria Conditions Peak vessel pressure 1500 1478 (( ))
(psig)
Peak cladding 2200 Not Calculated (1) N/A temperature (-F)
Peak suppression 202 191 [
pool temperature (OF) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
Peak containment 62 38.4(2) ]
pressure (psig)
Notes:
(1) Not Calculated based on the significant margin to the allowable value for the EPU ATWS analysis.
(2) The peak containment pressure is based on the SHEX calculated pressure at the bottom of the wetwell from the EPU drywell temperature analysis.
3-14
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Table 3-11: Peak Pressures for Other System Evaluations Parameter Value Elevation Comments Lower Plenum 1301 psig (Pressure at 152 The lower plenum pressure for all Pressure inches above vessel 0) transients was reviewed and compared to the initiation time of the SLCS pumps.
1301 psig is the highest lower plenum pressure that occurs after the initiation of the SLCS pumps. This pressure is based on an elevation of 152 inches above vessel
- 0. In addition to the PRFO and MSIVC transients, the LOOP transient was considered for the evaluation of the lower plenum pressure. The LOOP transient resulted in the limiting lower plenum pressure during the time when SLCS was operating. However, it is noted that there was less than 5 psi difference between the peak lower plenum pressure for all events.
Downcomer 1469 psig - (Pressure at 309 The pressure in the downcomer is Pressure GEl4 inches above vessel 0) calculated by ODYN. These pressures represent the peak pressure in the downcomer for all ATWS transients. The enthalpy of the fluid in the downcomer varies around the time of peak downcomer pressure from 535 BTU/Ibm to 570 BTU/lbm. The peak pressure value is based on the PRFO transient at BOC.
3-15
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Figure 3-1: MSIVC - BOC - GE14 Fuel r
))
3-16
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Figure 3-2: MSIVC - BOC - GE14 Fuel
[
))
3-17
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Figure 3-3: MSIVC - BOC - GE14 Fuel
((I I]
3-18
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Figure 3-4: MSIVC - EOC - GE14 Fuel
((I 1]
3-19
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Figure 3-5: MSIVC - EOC - GE14 Fuel
((
11 3-20
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Figure 3-6: MSIVC - EOC - GE14 Fuel
))
3-21
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Figure 3-7: PRFO - BOC - GE14 Fuel
((:
))
3-22
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Figure 3-8: PRFO - BOC - GE14 Fuel
))
3-23
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Figure 3-9: PRFO - BOC - GE14 Fuel
((
))
3-24
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Figure 3-10: PRFO - EOC - GE14 Fuel
((:
))
3-25
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Figure 3-11: PRFO - EOC- GE14 Fuel 1]
3-26
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Figure 3-12: PRFO - EOC - GE14 Fuel
[ff 1]
3-27
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Figure 3-13: Containment Response MSIVC EOC
((
))
Figure 3-14: Containment Response PRFO EOC
))
3-28
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION
3.4 CONCLUSION
S The ATWS evaluation incorporating the 3% setpoint tolerance confirms that all ATWS acceptance criteria are met. Therefore, based on the current Dresden and Quad Cities core loadings, the implementation of this increased tolerance at the Dresden and Quad Cities units is acceptable. The ATWS evaluations are based on an 80 gallon per minute Standby Liquid Control System injection rate with a 14% weight concentration of Sodium Pentaborate solution containing naturally enriched Boron. The peak lower plenum pressure during the operation of the Standby Liquid Control System is 1301 psig. The Standby Liquid Control system is required to attain an equivalent Boron injection rate with a lower plenum pressure up to 1301 psig in order for these analyses to remain valid for the increased setpoint tolerance. The Standby Liquid Control System performance is addressed in Section 6.4. The following recommendation remains applicable:
It is necessary for HPCI suction to be from the CST when suppression pool temperature exceeds the HPCI qualification limit even if suppression pool water level exceeds the automatic high level alarm setpoint 3-29
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 4 ECCS/LOCA EVALUATION 4.1 ECCS/LOCA PERFORMANCE EVALUATION Reference 5 provides the results of the Loss-of-Coolant-Accident Analysis (LOCA) performed by GE Nuclear Energy for Dresden and Quad Cities Station (D/Q). The analysis was performed using the SAFER/GESTR-LOCA application methodology approved by the Nuclear Regulatory Commission (NRC).
The impact of safety valve setpoint relaxation on the ECCS-LOCA performance for BWR 2-6 plants has been evaluated on a generic basis in the BWROG report approved by the NRC
[Reference 1, NEDC-31753P]. The ECCS conclusions contained in Reference I apply to Reference 5.
)) As such, plant-specific evaluations of ECCS performance and the impact of safety valve set point relaxation on LOCA Licensing Basis PCT are not required.
4-1
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 5 CONTAINMENT RESPONSE AND LOADS ANALYSIS This section presents the results of the various containment related evaluations in support of the safety relief valve setpoint tolerance increase from 1% to 3% for Dresden Units 2 & 3 and Quad Cities Units I & 2.
5.1 CONTAINMENT PRESSURE AND TEMPERATURE FOR DBA LOCA The effects on the peak containment pressure and temperature response for the short-term DBA LOCA event and on the peak suppression pool temperature and wetwell pressure for the long-term DBA LOCA from implementation of EPU and replacement dryer Reference 4 for Quad Cities were considered. Relaxation of the SSV and SRV safety valve setpoint tolerance has no effect on the DBA LOCA event because the vessel depressurizes without any ERV, SRV or SSV actuations. Therefore, there is no impact on the DBA LOCA containment pressure and temperature and on the DBA LOCA suppression pool temperature and wetwell pressure from implementation of EPU. The inputs of containment pressure and suppression pool temperature to the available NPSH analysis from implementation of EPU are also unaffected. The same conclusions as above are applicable for the future dryer replacement at Dresden.
5.2 SMALL STEAM LINE BREAKS Small steam line break (SLB) spectrum (0.01, 0.10, 0.30 and 0.75 ft2 breaks) was evaluated for EPU implementation to determine the drywell temperature for generating the EQ curve. The larger SLBs that produce the most limiting peak drywell temperature are however, large enough to maintain the initial vessel pressure below the ERV, SRV and SSV setpoints and also large enough to depressurize through the break without requiring ERV, SRV or SSV actuation.
Therefore, an increase in SSV or SRV safety valve opening setpoint has no effect on the larger SLBs that do not have ERV, SRV or SSV actuation. For smaller SLBs events, the ERV and SRVs can actuate, however, the reactor pressure is maintained below the SSV setpoints. The drywell temperature response for smaller SLBs that require ERV and SRV actuation may be slightly affected. For these breaks, the peak drywell temperature is well below that of the larger limiting SLB. Furthermore, the peak drywell temperature for the smaller SLBs occurs later in the event at the time the drywell sprays are actuated. Since this time occurs after many ERV and SRV actuations the peak temperature is controlled by the integrated steam flow to the drywell which is not affected by the change in the SSV and SRV safety valve setpoint tolerance. The long-term drywell temperature, after the sprays are initiated, is controlled by the break steam mass flow to the drywell and the spray temperature. The drywell spray temperature is controlled by the suppression pool temperature that is mainly governed by energy transferred to the suppression pool through the ERV and SRV. The rate of ERV and SRV energy transfer to the suppression pool is controlled by the vessel depressurization rate (assumed at 100 0F/hr), the initial vessel liquid inventory, and decay heat. These factors are not affected by the changes to the SRVs. The break steam flow to the drywell is controlled by the vessel pressure response, which is determined by the assumed vessel depressurization rate of 100*F/hr. This parameter is also unaffected by the change in the SSV and SRV safety valve setpoint tolerance. Since the 5-1
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION steam break flow and drywell spray temperature response for the smaller SLBs are not impacted by the SRV changes, the drywell temperature response for the smaller SLBs is also not impacted.
Therefore, an increase in SSV and SRV safety valve setpoint tolerance has no impact on the drywell temperature response and the EQ curve from EPU implementation remains valid. In addition, the same conclusions as above are applicable for the dryer replacement in Reference 4 for Quad Cities and future dryer replacement for Dresden.
5.3 IMA AND SBA The impact on intermediate and small break accidents (i.e., IBA and SBA) from EPU implementation was also evaluated. The containment pressure and temperature response for the IBA (a liquid line break of 0.1 ft2) and the SBA (a steam line break of 0.01 ft2), were originally evaluated as part of the Mark I Containment Program and documented in the Plant Unique Load Definition reports (PULD - References 6 and 7). The results for the [BA and SBA documented in References 6 and 7 are based on endpoint type calculations which are controlled by the amount of initial stored energy in the primary system and decay heat. There is no increase in the initial primary system stored energy or decay heat due to an increase in the SRV safety valve setpoint tolerance. Therefore, there is no impact on the IBA and SBA event results presented in References 6 and 7. Additionally, for the SBA the References 6 and 7 drywell temperature response is taken to be bounding, constant value of 340'F. This bounding drywell temperature value would not change due to an increase in SRV setpoint tolerance.
The EPU IBA and SBA analyses were performed using the GE SHEX containment code but with assumptions which are consistent with the References 6 and 7 analyses. ((
)) Because of the assumption on vessel depressurization used for the IBA and SBA analyses, the changes to the SRV safety valve setpoint tolerance would have no impact on the results of the EPU calculations for the IBA and SBA.
In addition, the same conclusions as above are applicable for the dryer replacement in Reference 4 for Quad Cities and future dryer replacement for Dresden.
5.4 NUREG-0783 LOCAL SUPPRESSION POOL TEMPERATURE Quad Cities and Dresden have quenchers on the ERV and SRV discharge lines which, per Reference 9 ensures stable condensation at elevated local suppression pool near saturation. The NRC Safety Evaluation for Reference 9 (see Reference 8), conditionally approved elimination of the NUREG-0783 local pool temperature limits based on the Reference 9 evaluation. Per Reference 8, the local pool temperature limits of NUREG-0783, and associated evaluations, can only be eliminated if plants have pump suction inlets below the elevation of the quencher. This condition was imposed to address NRC concerns regarding steam ingestion of SRV steam into pump suction inlets at high local suppression pool temperature. An evaluation in 2001 determined that steam ingestion into the ECCS suction strainers will not occur for the Dresden and Quad Cities plants with the existing SRVs. This evaluation used the SRV flow capacity for 5-2
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION the existing Target Rock SRV. Using the parameters for the Target Rock SRV provided as input by Exelon, and considering a 3% tolerance on the SRV safety valve opening setpoint pressure,
((
)) Therefore, the conclusions from the 2001 evaluation remain valid in that steam ingestion is not predicted.
In addition, the same conclusions as above are applicable for the dryer replacement in Reference 4 for Quad Cities and future dryer replacement for Dresden.
5.5 DBA LOCA HYDRODYNAMIC LOADS The DBA LOCA hydrodynamic loads, such as pool swell, vent thrust, condensation oscillation and chugging are dependent on the containment pressure and temperature response during the DBA LOCA. Because the containment DBA LOCA pressure and temperature response are not affected by an increase in the SSV and SRV safety valve setpoint tolerance, the DBA LOCA hydrodynamic loads are also unaffected.
In addition, the same conclusions as above are applicable for the dryer replacement in Reference 4 for Quad Cities and future dryer replacement for Dresden.
5.6 SRV DYNAMIC LOADS The SRV safety valve setpoint tolerance increase has no effect on the ERVs since they are not the ASME code safety valves.
The SRV discharge loads are determined by the following controlling parameters:
- SRV discharge line (SRVDL) and containment geometry
- Water leg length in the SRVDL at the time of SRV opening
- SRV flow capacity and SRV opening pressure Since a relaxed SRV setpoint tolerance can increase the SRV safety valve opening pressure, the SRV discharge dynamic loads are expected to increase. Exelon will need to evaluate the SRV discharge dynamic loads with the SRV setpoint tolerance increase.
5.7 RIPD EVALUATION During normal operation, there is no SRV actuation. Therefore, the SRV setpoint tolerance change have no effect on the reactor internal pressure differences (RIPDs) at normal conditions in Reference 4 and as shown for EPU.
For upset conditions, any event in which SRVs will actuate would have a faster depressurization due to increased SRV flow as a result of SRV setpoint tolerance change, causing higher DPs across the reactor internals. ((
5-3
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION
)) Therefore, the RIPD results at upset conditions in Reference 4 and for EPU remain valid for the SRV setpoint tolerance relaxation.
The limiting emergency event used for RIPD is an inadvertent actuation of all ADS valves.
Increased SRV flow capacity as a result of SRV setpoint tolerance change would have a faster depressurization and thus would result in higher DPs for reactor internals. ((
)) Thus, the RIPD results at emergency conditions in Reference 4 and for EPU are still applicable for the SRV setpoint tolerance increase.
The limiting faulted event for RIPD is an instantaneous circumferential break of one main steam line, for which SRV does not actuate. Therefore, the SRV setpoint tolerance relaxation has no effect on the RIPD results at faulted conditions in Reference 4 and for EPU.
As part of RIPD, the analyses for acoustic and flow-induced loads on jet pump, core shroud and shroud support due to recirculation line break are not affected by SRV setpoint tolerance relaxation because the SRVs will not actuate during the event. Therefore, the SRV setpoint tolerance increase does not impact the acoustic and flow-induced load analyses for EPU.
5-4
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 6 HIGH PRESSURE SYSTEMS PERFORMANCE This section summarizes the evaluation of high pressure systems, as well as the performance of systems such as pressure control and piping.
6.1 HIGH PRESSURE COOLANT INJECTION The purpose of the High pressure Coolant Injection (HPCI) systems at Quad Cities Units 1 and 2 and Dresden Units 2 and 3 is to provide high pressure emergency cooling water to the reactor to prevent excessive peak fuel clad temperature (PCT) following small line breaks that do not result in rapid depressurization. It operates to perform this function in conjunction with the Core Spray (CS) or Low Pressure Coolant Injection (LPCI) systems, and with credit for operation of the Automatic Depressurization System (ADS). The HPCI system also functions as a backup to the Reactor Core Isolation Cooling (RCIC) system at Quad Cities, or the Isolation Condenser (IC) at Dresden, in case of a failure of those systems following a transient event. To achieve this purpose, the HPCI system is designed to supply makeup water to the reactor at a capacity of 5600 gpm over a reactor pressure range of 1135 psia to 165 psia.
The maximum reactor operating pressure for rated makeup flow for the HPCI system at both Dresden and Quad Cities is based on the upper analytical limit (UAL) of the lowest group of relief valves (RVs), on condition that that this group contains a sufficient number of valves to provide the long-term relief function, even allowing for another independent failure within the RVs.
Dresden and Quad Cities both use two reactor relief valves at the lowest RV group setpoint. It was confirmed during EPU that operation of only one RV is needed for the long-term pressure relief function. Thus, the maximum reactor pressure for HPCI system water makeup operation is based on the upper analytical setpoint for the lowest group of RVs. For EPU, this corresponds to a pressure of 1115 psig.
These RVs are not within the group of valves that are receiving a setpoint tolerance increase.
Therefore, the RV setpoints are not changing and there is no effect on the HPCI system maximum reactor injection pressure due to the SRV setpoint tolerance increase.
The HPCI system steam supply line contains break detection instrumentation designed to detect high steam flow, indicative of a break in that line. The isolation setpoints for this instrumentation are based on a differential pressure across the flow sensing device. Because the reactor vessel pressure for HPCI system operation remains the same, there will be no increase in rated steam flow to the turbine, and therefore, no effect on the break detection instrumentation or the trip setpoints.
The HPCI steam line containment isolation motor-operated valves (MOVs) are normally open with the system in standby. At Quad Cities and Dresden they are evaluated to be capable of closing against a differential pressure of approximately 1147 psid. This closing differential pressure is based on the current SRV nominal setpoint of 1135 psig and a 1% setpoint tolerance.
A change to a 3% setpoint tolerance will increase the upper analytical limit to 1169.1 psig.
6-1
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Therefore, Exelon will assure that the HPCI steam line MOVs are evaluated to operate acceptably with a reactor vessel pressure of 1169.1 psig prior to implementation.
The HPCI system injection valve is normally closed and is signaled to open during a system initiation. Since the -PCI system is designed for injection based on the RV setpoint, which is not changing, the injection valve is not affected by the SRV setpoint tolerance change.
It is concluded that the SRV setpoint tolerance increase will have no effect on the capability of the HPCI system to provide makeup water to the reactor vessel. The SRV setpoint tolerance change will affect both the Quad Cities and Dresden HPCI steam line MOVs with respect to the maximum closure differential pressure. The maximum closure pressure will increase to 1169.1 psig.
6.2 REACTOR CORE ISOLATION COOLING The purpose of the RCIC system is to provide cooling water to the Quad Cities Unit 1 or Unit 2 reactor in the event that the reactor becomes isolated from the main condenser simultaneously with a loss of the feedwater system. To achieve this purpose, the RCIC system is designed to supply makeup water to the reactor at a capacity of 400 gpm over a reactor pressure range of 1135 psia to 165 psia.
The maximum reactor operating pressure for water makeup for the RCIC system at the Quad Cities plant is based on the upper analytical limit of the lowest group of relief valves (RVs),
providing that this group includes a sufficient number of valves to provide the long-term relief function and there are allowances for another independent failure within the RVs.
The Quad Cities plant uses two RVs in the lowest group of reactor relief valves. For EPU, it was determined that operation of only one RV is needed for the long-term pressure relief function.
Thus, the maximum reactor pressure for RCIC system water makeup operation is based on the upper analytical setpoint for the lowest group of RVs. For EPU, this corresponds to a pressure of 1115 psig.
These RVs are not within the group of valves that are receiving a setpoint tolerance increase.
Therefore, the RV setpoints are not changing and there is no effect on the RCIC system maximum reactor injection pressure due to the SRV setpoint tolerance increase.
The RCIC system steam line contains break detection instrumentation designed to detect high flow in the line indicative of a break in that line. The isolation setpoints for this instrumentation are based on a differential pressure across the flow sensing device. Because the reactor vessel pressure for RCIC system operation remains the same, there will be no increase in steam flow and is no effect on the break detection instrumentation and the trip setpoints.
The RCIC steam line containment isolation motor-operated valves (MOVs) are capable of closing against a differential pressure of 1147 psid (MO-1301-16) and 1146 psid (MO-1301-17).
The closing differential pressure is based on an SRV nominal setpoint of 1135 psig and a 1%
setpoint tolerance (1147 psig). The high energy line break (HELB) maximum differential pressure for the MOVs is also based on an upstream pressure of 1147 psig. The SRV has a 6-2
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION nominal setpoint of 1135 psig. A change to a 3% setpoint tolerance will increase the upper analytical setpoint to 1169.1 psig. Therefore, Exelon will assure that the RCIC steam line MOVs are evaluated to operate acceptably with a reactor vessel pressure of 1169.1 psig for normal closure and for the HELB closure analysis prior to implementation.
The RCIC system injection valve is normally closed and is signaled to open during a system initiation. Since the RCIC system is designed for injection based on the RV setpoint, and the RV setpoint tolerance is not changing, the injection valve is not affected by the SRV setpoint tolerance change.
It is concluded that the SRV setpoint tolerance increase will have no affect on the capability of the RCIC system to provide makeup water to the reactor vessel. The SRV setpoint tolerance change will affect the RCIC steam line MOVs with respect to the maximum closure differential pressure. The maximum closure pressure (reactor vessel pressure) will increase to 1169.1 psig.
6.3 STEAM BYPASS PRESSURE CONTROL SYSTEM 6.3.1 Description The purpose of this section is to evaluate the impact of the proposed main steam safety relief valve opening setpoint tolerance relaxation on the Steam Bypass Pressure Control System functionality and performance at both Dresden Units 2 & 3 and Quad Cities Units 1 & 2. This report will be summarized in an overall evaluation to support a Tech Spec change to increase the set point tolerance of the safety relief valves from 1% to 3%.
For this evaluation, each of the following was reviewed to determine affects (if any) the relaxation of Main Steam Safety Valve (MSSV) setpoint tolerances with respect to the Steam Bypass Pressure Control System function and performance.
- Safety Analysis Reports for both the Quad Cities 1 & 2 and the Dresden 2 & 3
" Extended Power Uprate (References 10 and 11); Sections 5.2, 5.2.1, 5.3.11, 5.3.13, and 7.3; also Figure 1-1 and Table 1-2.
" The most recent reload licensing analysis for Dresden Units 2 & 3 and the Quad Cities Units 1 & 2.
6.3.2 Inputs and Assumptions Based on the Safety Analysis Reports for the Dresden and Quad Cities extended power uprate projects:
" the normal reactor operating pressure is 1005 psig,
" the rated vessel steam flow is 11.71 Mlb/hr, 6-3
GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION
" the bypass capacity of each of the Dresden Units is 33.5% of rated reactor steam flow, and
" the bypass capacity of each of the Quad Cities Units is 33.3% of rated reactor steam flow.
Based on the individual Dresden and Quad Cities reload-licensing analysis reviewed:
" the normal reactor operating pressures are 1005 psig,
" the vessel steam flows are 11.71 Mlb/hr,
" the bypass capacity credited in the transient analysis (single BPVOOS) for each of the Dresden Units is 29.8%, and
" the bypass capacity credited in the transient analysis (single BPVOOS) for each of the Quad Cities Units is 29.6%.
" By definition, the safety relief valves are not expected to relieve (lift) within the normal operating range.
6.3.3 Evaluation The Steam Bypass Pressure Control System (SBPCS) is a normally operating system, which provides fast and stable responses to system disturbances related to steam pressure and flow changes and thereby controls reactor pressure within its normal operating range. SBPCS consists of the pressure regulation system, turbine control valve system and the steam bypass valve system.
The EPU evaluations for the SBPCS, summarized four (4) evaluations performed for the Steam Bypass Pressure Control System as follows:
a) EPU impact to system design basis controlling parameters. The rated steam bypass absolute flow rate does not change, but the increase in steam flow results in the reduced percentage of bypass capacity (i.e., the absolute bypass flow rate as expressed as a percentage of EPU reactor rated steamflow). The bypass capacity is sufficient to support operation of Dresden and Quad Cities at EPU conditions. Setpoint tolerance relaxation affect on this evaluation- None, since the normal operatingsteamflow rates used as an input to the originalEPUreport are the same as the flow rates in the reloadtransient analyses (11. 71 Mlbm/hr).
b) EPU impact to control room operator instrunentation, setpoint adjustments, indications, alarms, and SBPCS controls. Minimal impact on equipment. Signal ranges and adjustment capabilities are adequate to support EPU. Pressure regulator setpoint adjustment is required (decreased) to maintain 1020 psia (1005 psig) steam dome pressure to account for the increase in main steam line pressure drop. Setpoint tolerance relaxation affect on this evaluation - None, since the normal operating steam dome pressure is not changed as documented in the most recent reload transient analyses (1005 ps.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION c) Determine if bypass valve inlet pressure conditions are significantly changed due to the changes in the steam line pressure drop to the Turbine Stop Valves (TSV) and steam chest at EPU conditions. Steam passing capabilities of the bypass valves were not significantly impacted by EPU. Setpoint tolerance relaxationaffect on this evaluation -
None, since the normal operating steam flow rate and steam dome pressure are not changedas documented in the most recent reloadtransientanalysis (L 71 Mlbm/hr and 1005psig respectively), the steam linepressuredrop does not change.
d) Determine if the transient performance of the SBPCS, operating under EPU conditions, are impacted for the evaluation of major transients such as main turbine- generator trip or main generator load rejection. The transient evaluations performed for events that require SBPCS operation determined that bypass capacities were adequate for the transient analysis to remain valid at EPU conditions. Selpoint tolerancerelaxationaffect on this evaluation - The steam bypass capacity, calculatedas 33.3%for Quad Citiesand 33.5%for Dresden, was determined to be adequatefor the transientanalysis to remain validfor EPU conditions. The most recent reload transientanalysesfor both Dresden and QuadCities only takes creditfor bypass capacityof eight of the nine bypass valves to reflect a single bypass valve out of service (BPVOOS). Therefore, the bypass capacities creditedfor the transient analysis of 29.6% and 29.8% respectively, which are bound by the EPUcapacities,are determinedto be adequate.
6.3.4 Conclusion This evaluation concludes that the Steam Bypass Pressure Control System functional and performance requirements are not affected by the MSSV setpoint tolerance relaxation.
6.4 STANDBY LIQUID CONTROL SYSTEM The Standby Liquid Control System (SLCS) is designed to shut down the reactor from rated power condition to cold shutdown in a postulated event in which all or some of the control rods cannot be inserted or during a postulated ATWS event. The SLCS accomplishes this function by pumping a sodium pentaborate solution into the vessel at a prescribed boron injection rate in order to provide neutron absorption and achieve a subcritical reactor condition.
The original performance design basis for the SLCS was that it must be capable of injecting the system design rated flow into the reactor vessel using a single SLC pump at a maximum reactor pressure equal to the SRV group with the lowest setpoint operating in the relief mode. This method has been superseded by the use of the maximum reactor vessel pressure occurring during the limiting ATWS event when the SLCS is in operation in consideration of NRC Information Notice 2001-13.
Exelon will ensure that the 10CFR50.62 requirement to inject 86 GPM of 13% sodium pentaborate solution, or the equivalent, plus the ATWS specific injection requirements stated in Section 3.0 of this report are met for injection against the maximum reactor vessel pressure of 1301 psig at the SLCS sparger occurring during an ATWS event when the SLCS is in operation without opening of the SLCS relief valve.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 6.5 SAFE SHUTDOWN MAKUP PUMP SYSTEM The purpose of the safe shutdown makeup pump (SSMP) system is to provide cooling water to the Quad Cities Unit 1 or Unit 2 reactor in the event that the reactor becomes isolated from the main condenser simultaneously with a loss of the feedwater system. This system was installed as a common backup to the Quad Cities Unit I and Unit 2 RCIC systems. To achieve this purpose, the SSMP system is designed to supply makeup water to the reactor at a capacity of 400 gpm over a reactor pressure range of 1135 psia (1120 psig) to 165 psia (150 psig), the same as the RCIC system.
The system consists of a single motor-driven pump designed for a flow rate of 400 gpm at 2885 feet. The system can pump to either Quad Cities Unit I or Unit 2. The SSMP injection valves are interlocked to allow injection to only one reactor at a time.
Because this system is installed as a backup to the RCIC system, it shares the same design basis with respect to the maximum reactor vessel pressure for injection. The RCIC system is capable of injecting makeup water to the reactor vessel up to a vessel pressure of 1120 psig (1135 psia).
It has been determined that the lowest group of RVs are capable of maintaining reactor vessel pressure below the maximum design injection pressure of 1120 psig for long-term pressure reactor vessel pressure relief.
These RVs are not within the group of SRVs that are receiving a setpoint tolerance increase.
Therefore, the RV setpoints are not changing and there is no effect on the SSMP system maximum reactor injection pressure due to the SRV setpoint tolerance increase.
It is concluded that the SRV setpoint tolerance increase will have no affect on the SSWP system.
6.6 ISOLATION CONDENSER SYSTEM The Isolation Condenser (IC) system design basis is to provide reactor core cooling in the event that the reactor pressure vessel (RPV) becomes isolated from the main condenser by closure of the main steam isolation valves. This event concurrent with the loss of all feedwater flow (LOFW) by the loss of offsite power is the design transient for the IC system. This report evaluates the impact of SRV setpoint tolerance change on the IC system. The IC system applies to the Dresden plants only.
Automatic initiation of IC operation occurs when a high reactor pressure signal of 1068 psig exists for more than 15 seconds. The initiation setpoint and time delay are independent of the SRV setpoint and setpoint tolerance increase. The SRV has a setpoint of 1135 psig. For a 1%
setpoint tolerance, the upper analytical setpoint is 1146.4 psig. For a 3% setpoint tolerance, the upper analytical setpoint is 1169.1 psig. Both of these setpoints are above the IC high reactor pressure initiation signal of 1068 psig. This setpoint exceeds the IC initiation setpoint. Thus, the SRV upper analytical setpoint tolerance change does not affect the IC initiation.
The lower analytical setpoint for the 3% tolerance change results in an SRV setpoint of 1101 psig. For the 1% setpoint tolerance, the lower analytical setpoint is 1123.6 psig. Refer to Section 2.6 in this report for discussion of the effect on the IC system initiation.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION Those portions of the IC system interfacing directly with the reactor (the RCPB) are designed to 1250 psig and 575*F. The setpoint tolerance change will not increase the maximum reactor pressure following transient and accident events above the current limits. Consequently, the SRV setpoint tolerance increase will not impose changes to the design values for the IC system RCPB components.
The IC system motor-operated valves will not be affected by the SRV setpoint tolerance increase. The steam line isolation valves are maintained open during normal plant operation.
The condensate return line isolation valves are maintained closed and must open to allow IC system operation. The reactor operating pressure is not increased; therefore, periodic testing of these valves is not affected. Since the differential pressure across the condensate return line valve will not change (the reactor vessel dome pressure acts equally on both sides of the valve),
there is no effect on the valve opening capability due to the SRV setpoint tolerance increase.
The IC system steam line and condensate return line contain break detection instrumentation designed to detect high flow in the line indicative of a break in that line. The isolation setpoints for this instrumentation are based on a differential pressure across the flow sensing device.
Because the design flow for the IC system remains the same, there is no effect on the break detection instrumentation and the trip setpoints for the SRV setpoint tolerance increase.
It is concluded that the SRV setpoint tolerance increase will have no affect on the IC system.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 7 APPENDIX R ANALYSIS This section provides an Appendix R fire protection safety evaluation for Quad Cities and Dresden SRV setpoint tolerance increase (safety mode from 1% to 3%).
7.1 VESSEL INVENTORY ASSESSMENT Increased SRV setpoint tolerance to +3% will cause SRV actuation at higher pressure and thus result in a slight delay in the SRV actuation. Consequently, the instantaneous flow rates out of the SRVs are increased due to the higher critical flow rates in comparison to the case with SRVs at currently analyzed setpoint tolerance. However, the change in the total inventory lost from the vessel due to SRV setpoint tolerance relaxation is negligible. This is because the inventory loss is primarily dependent on the decay heat, which remains unaffected by SRV setpoint tolerance relaxation. In addition, the existing Target Rock SRV with the same capacity was only assumed for stuck open and SRV cycling in EPU evaluations. Therefore, the vessel water level responses and conclusions in the EPU evaluations are still applicable for +/-3% SRV safety valve setpoint tolerance change. Note that the inventory loss as a result of SORV during first 10 minutes is not affected by SRV setpoint relaxation because opening of SRV is caused by a fire at time initiation and closing of SORV is due to manual operator action, not by reaching SRV setpoint.
7.2 CONTAINMENT RESPONSE ASSESSMENT The suppression pool temperature is mainly governed by energy transferred to the suppression pool through the SRVs. Before depressurization, the similar energy would be transferred to the suppression pool due to a net slightly increased SRV flow as a result of the SRV safety valve setpoint tolerance increase, balanced by less SRV cyclings caused by the +/-3% SRV setpoint tolerance change. After depressurization, the rate of SRV energy transfer to the suppression pool and total energy transfer to the suppression pool are controlled by the vessel depressurization rate (assumed at 100oF/hr), the initial vessel liquid inventory, and decay heat which are unaffected.
Thus, the SRV setpoint tolerance change has no adverse impact on the suppression pool temperature, as well as containment temperature and pressure for an Appendix R fire event.
Therefore, the containment response in the EPU evaluations are still applicable.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 8 VESSEL THERMAL CYCLE ASSESSMENT 8.1 ANALYSIS OVERVIEW Reactor Pressure Vessel Thermal Cycles: Safety Relief Valve blow down is a "thermal cycle" of the RPV and is counted over the life of the plant. The design basis allowable for SRV blow down is 5 (UFSAR Table 3.9-1). The elevated set point at which the SRV can lift may impact the fatigue usage of the RPV. The number of allowable SRV events was qualitatively reviewed considering the relaxation of their set point tolerance.
8.2 INPUTS AND ASSUMPTIONS There are two transient pressure rise events and one pressure decrease event considered for the vessel thermal cycle design:
The pressure rise events are the overpressure 1250 psig event and the overpressure 1375 psig event.
The pressure decrease event is the single relief or safety valve blow down event.
The evaluation is based on a 2% increase in the opening set-point, relaxation from 1% to 3%,
and it is assumed that this is within the RPV design pressure.
8.3 ANALYSIS RESULTS There will be pressure and temperature oscillations during the overpressure events due to the SRV cycling. The temperature oscillation resulting from the SRV opening set-point change from 1% to 3% is within the design temperatures assumed for these events on the thermal cycle diagram.
The SRV opening set-point relaxation from 1%to 3%, will have limited effects on these three events. The set point increase does not have any effect on the pressure decrease event.
8.4 CONCLUSION
S AND RECOMMENDATIONS It is recommended that the SRV blow-down events be limited to the 5 cycles already reported in the USAR. It is also recommended that the plant use a fatigue monitor program to review the number of cycles and accumulated fatigue usage.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 9 OPERATING MODES AND EOOS REVIEW For the purpose of this task, the review is based on any additional requirements imposed by the SRV setpoint tolerance increase that would affect a specific option. The SRV setpoint tolerance increase is evaluated as to its impact on various analyses that involve conditions or events that actuate the SRVs. All the options include analyses that actuate the SRV and therefore impact the basis for each option. The SRV setpoint tolerance increase is concluded to not impose any additional requirements on the operation and licensing basis for Dresden and Quad Cities, therefore the setpoint tolerance increase is entirely compatible with the Operating Modes and EOOS. Note that an SRV OOS has been previously evaluated, however, that option does not meet the over pressure criteria under EPU. This conclusion remains for the setpoint tolerance increase and the SRV OOS option remains unacceptable for Dresden and Quad Cities.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 10 EMERGENT EPU ISSUES REVIEW The specific areas identified as requiring direct evaluation as a result of the SRV setpoint tolerance increase are addressed by the separate tasks included in the scope. Other tasks are concluded to not be affected on the basis that the SRV setpoint tolerance increase has no impact on normal operating conditions and/or events that do not actuate the SRVs. For this review, the technical and licensing activities corresponding to the most recent EPU project are examined with respect to the SRV setpoint tolerance increase to determine if a risk exists with respect to the Dresden and Quad Cities EPU Basis. No specific areas of concern were identified in this review.
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GE-NE-0000-0053-8435-RINP NON-PROPRIETARY INFORMATION 11 REFERENCES
- 1. NEDC-31753P, "BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report", Class III, February 1990.
- 2. NEDE-2401 1-P-A-15, "General Electric Standard Application for Reactor Fuel (GESTAR),
Class III, September 2005.
- 3. GE-NE-J 11-03912-00-01-R3, "Dresden 2 and 3 Quad Cities I and 2 Equipment Out-Of-Service and Legacy Fuel Transient Analysis" Rev 3, September 2005.
- 4. NEDC-33187, "Safety Evaluation in Support of the New Steam Dryer for Quad Cities Unit I
& 2", Revision 2, May 2005.
- 5. NEDC-32990P, "SAFER/GESTR-LOCA Loss-of-Coolant-Accident Analysis for Dresden Nuclear Station 2 and 3 and Quad Cities Nuclear Station Units I and 2," Revision 2, September 2003.
- 6. "Mark I Containment Program Plant Unique Load Definition Quad Cities Units I and 2,"
NEDO-24567, Rev.2, April 1982."
- 7. "Mark I Containment Program Plant Unique Load Definition Dresden Nuclear Power Station: Units 2 and 3," NEDO-24566, Rev. 2, April 1982.
- 8. Gary M. Holahan (NRC) to Pinelli (BWROG), "Transmittal of the Safety Evaluation of General Electric Co. Topical Report; NEDO-30832 Entitled "Elimination of Limit on BWR Suppression Pool Temperature for SRV Discharge with Quenchers" and NEDO-31695 Entitled "BWR Suppression Pool Temperature Technical Specification Limits," August 29, 1994.
- 9. NEDO-30832-A "Elimination of Limit on BWR Suppression Pool Temperature for SRV Discharge with Quenchers", May 1995.
- 10. GE Nuclear Energy, Safety Analysis Report for Dresden 2 & 3 Extended Power Uprate, NEDC-32962P, Revision 2, (Proprietary), August 2001.
- 11. GE Nuclear Energy, "Safety Analysis Report for Quad Cities I & 2 Extended Power Uprate, NEDC-32961P Revision 2" (Proprietary), August 2001.
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