ML20216G515
| ML20216G515 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 02/28/1998 |
| From: | Chi L, Du L, Kumar G GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20216G503 | List: |
| References | |
| NEDC-32788, NUDOCS 9803200012 | |
| Download: ML20216G515 (46) | |
Text
{{#Wiki_filter:! \\ s LJ GE Nucle:r Energy NEDC-32788 Class 3 DRF H21-01895 February 1998 SAFETY REVIEW FOR ENRICO FERMI ENERGY CENTER UNIT 2 SAFETY / RELIEF VALVE SETPOINT TOLERANCE RELAXATION ANALYSES v v' s Lichao Du, Senior Engineer l l l Approved by: Approved by: G. V. Kumar, Project Manager 1,. (hi, Manager Plant Performance Improvement 9803200012 980312 PDR ADOCK 05000341 i P PDR
1 NEDC-32788 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 customer and GE, as identified in the purchase order for this report and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than the customer or for any purpose other than that for which it is intended, is not authorized; and 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. GE proprietary information have been intentionally deleted from this document. -- -and-4Iheir locations are denoted by blank spaces in the text -and bars in the margins. l l ( l
i-NEDC-32788 l l TABLE OF CONTENTS Ease
SUMMARY
S1 ( l.0 INTRODUCTION 1-1 1.1 Purpose 1-1
1.2 Background
1-l' 1.3 Present Performance Requirements 1-2 l 1.4 Froposed Performance Requirements Changes 1-2 2.0 ANALYSIS OVERVIEW 2-1 3.0 VESSELJVERPRESSURE PROTECTION ANALYSIS 3-1 3.1 Overpressure Protection Analysis Assumptions 3-1 3.2 Overpressure Protection Analysis Results 3-2 l 3.3 Conclusions 3-2 l 4.0 THERMAL LIMITS 4-1 l 4.1 Analyses and Results 4-1 4.2 Conclusions 4-2 l 5.0 ECCS/LOCA PERFORMANCE EVALUATION 5-1 5.1 Limiting Break LOCA 5-1 5.2 Small Breax LOCA 5-1 5.3 Steamline Break Outside the Containment 5-2 5.4 Conclusions 5-2 6.0 HIGH PRESSURE SYSTEM PERFORMANCE 6-1 6.1 Impact of SRV Changes on High Pressure 6-1 System Performance l 6.2 Results and Conclusions for High Pressure Coolant 6-1 l Injection System Performance Evaluation i 6.3 Results and Conclusions for Reactor Core Isolation 6-3 Cooling System Performance Evaluation 6.4 Conclusions for Standby Liquid Control 6-4 System Performance Evaluation 7.0 CONTAINMENT EVALUATION 7-1 7.1 Containment Pressure and Temperature 7-1 7.2 Safety / Relief Valve Dynamic Loads 7-1 7.3 Conclusions 7-2 l iiv e
NEDC-32788 TABLE OF CONTENTS (Continued) Eage 8.0 ATWS MITIGATION CAPABILITY 8-1 8.1 Overview 8-1 8.2 ATWS Analysis Method 8-1 8.3 Inputs and Assumptions 8-2 8.4 ATWS Analysis Results 8-2 8.5 Conclusions 8-4 9.0 PIPING EVALUATION 9-1 9.1 Introduction 9-1 9.2 Methods and Assumptions 9-1 9.3 Results and Conclusions 9-1 10.0 TECHNICAL SPECIFICATIONS 10-1
11.0 CONCLUSION
S 11-1
12.0 REFERENCES
12-1 HW l 1
NEDC-32788 TABLES Table lide Page 1-1 Comparison of Present to Proposed Performance 1-4' Requirements 2-1 Analyses Presented in This Report 2-2 3-1 MSIV Closure with Flux Scram Event Analysis Results 3-3 4-1 The Limiting Thermal Limit Case 4-3 6-1 HPCI System Performance Comparison 6-4 6-2 RCIC System Performance Comparison 6-5 6-3 Standby Liquid Control System Performance Comparison 6-6 Two-Pump Performance Evaluation 8-1 initial Operating Conditions for ATWS Analysis 8-5 8-2 ATWS MSIV Closure Transient Response 8-5 9-1 Comparison of SRV Setpoints 9-2
NEDC-32788 FIGURES - Figure Iitig g 3-1 MSIV Closure Flux Scram Event 3-4 Flow, Credit for 11 out of15 SRVs. 4-1 Plant Response to FW Controller Failure 4-4 8-1 MSIV Closure ATWS Event 8-6 8-2 MSIV Closure ATWS Event 8-7 I I inv l
NEDC-32788 i
SUMMARY
This report provides the technicaljustification to support the relaxation of the safety / relief valve (SRV) opening setpoint tolerances for the Fermi Energy Center, Unit 2, from the current value { i1% to a new value of i3%. The technical justification addresses the safety-related issues ascociated with the proposed changes in the following areas: Vesseloverpressurizution - Analysis for the most limiting pressurization event e i indicates that vessel pressure remains within the ASME Upset Code limit of 1375 psig. The analysis was performed taking credit for only 11 out of 15 SRVs. Fuel thermal limits during anticipated operational occurrences - The event was analyzed assuming the SRV nominal opening setpoints with a i3% tolerance, and taking credit for only 11 out of 15 SRVs. Results show that there is no impact on the fuel thermal limits due to SRV setpoint tolerance relaxation. ECCS/ Loss-of-CoolantAccident(LOCA) performance-An evaluation concluded that an increase in SRV setpoint tolerances from il% to i3% would have no impact l High pressure emergency systems performance-Analyses performed on the high pressure emergency systems indicate that these systems can provide their design functions with the new i3% SRV setpoint tolerance. Containment pressures, temperatures, and loads - Available margins, which have been verified through previous testing, will accommodate the increased loads on containment due to SRV setpoint tolerance relaxation from il% to i3%. Anticipated transients without scram (A TWS) mitigation capability - An analysis of the ATWS MSIV closure event, which is the limiting event was performed. it was found that vessel pressurization remains l S-B
i NEDC-32788 i l l l within the ASME Emergency Code limit of 1500 psig. The specified criterion for short-term overpressurization and the suppression pool temperature is also within the temperature limit based upon the limiting criterion for long-term ATWS. The evaluation 1 conservatively takes credit for only 11 out of 15 SRVs. 1 Main steam and SRVpiping stress - The steamline and SRV piping evaluations e - demonstrate that sufficient margins exist in the piping stress analyses to accommodate loads due to the SRV setpoint tolerance increase. In summary, there is no adverse impact on the safety-related systems or plant safety due to the relaxation of SRV setpoint tolerance from the current il% value to i3%. l j l r S-22 l
NEDC-32788
1.0 INTRODUCTION
1 l 1.1 PURPOSE The purpose of this report is to present the results of an evaluation of a proposed change to the setpoint tolerance requirement for the safety / relief valves (SRVs) at the Enrico Fermi Energy Ceriter, Unit 2 (Fermi-2). The proposed changes are selected to minimize the impact on plant operations from potential pressure relief system related problems due to SRV setpoint drift. Detroit Edison has requested that the changes be evaluated to support relaxing the SRV setpoint tolerance from the current 1% to i3% for the nominal SRV setpoints. The current performance requirements for the Fermi-2 SRVs are discussed in Section 1.3. Each of the present performance requirements pertinent to this analysis is identified, as well as the associated limitations and the remedial actions for exceeding the limit. The proposed performance requirement changes are discussed in Section 1.4 along with the associated limits and the analyses required to support each proposed change. A comparison of the present and proposed performance requirements is shown in Table 1-1.
1.2 BACKGROUND
The pressure relief system for the nuclear reactor pressure vessel at Fermi-2 consists of SPVs located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. The SRVs provide the following primary functions: Overpressure safety /rclief operation: SRVs open automatically against spring restraint to limit the vessel pressure excursion during a postulated pressurization transient event. Depressurization operation: The Automatic Depressurization System (ADS) functions to open selected SRVs automatically by way of a pilot air system. It is considered to be part of the Emergency Core Cooling System (ECCS) for events involving small breaks in the reactor vessel process barrier. Low-Low-Set (LLS) operation: Following initial opening and closure of SRVs, two SRVs operate in the LLS mode opening and closing at pressures lower than the other SRVs to reduce the amount of cycling on the non-LLS SRVs. l l 1-14
NEDC-32788 1.3 PRESENT PERFORMANCE REQUIREMENTS 1.3.1 SRV Setpoint Tolerances From Reference 1, the current SRV and LLS setpoint configuration and the associated nominal opening setpoint values for Fermi-2 are as follows: Safety /ReliefValves: 5 with a 1135 psig setpoint 5 with a 1145 psig setpoint 5 with a 1155 psig setpoint Low-Low-Set Valves I with a 1017 psig setpoint I with a 1047 psig setpoint A narrow 11 % tolerance band on the SRV nominal setpoints was originally adopted to develop the Technical Specifications due to the acceptance criterion originally defined by the American Society of Mechanical Engineers (ASME), Section III, NB-7000. Section 3/4.4.2 of the current Technical Specifications states that the allowable setpoint variance for each SRV shall be i1%. The original Technical Specification criterion used was for consistency with the NB-7000 criterion and consistency with the SRV setpoint upper limit used in the original 1 overpressure protection analysis. The ASME has since revised the criterion for demonstrating acceptable in-service SRV operational readiness from i1% to i3 % [ Reference 2]. Consequently, a valve opening setpoint variance greater than +1% of the nominal setpoint causes an increase in valve surveillance testing costs, adding to the number of reportable events and j l diverting utility work force resources. Although the il% tolerance is specified in the Technical Specifications and has been used in plant safety evaluations, it does not represent the limiting setpoint required to ensure plant safety. Several BWRs have experienced SRV setpoint drift more than the Technical Specification limitations [ Reference 7 ]. In each case, safety evaluations were performed on a plant-specific and cycle-specific basis, demonstrating that setpoint drift did j not compromise plant safety. The need for such safety evaluations may be minimized by increasing the setpoint tolerance for the SRVs as specified in the Technical Specifications for l Fermi-2. I l 1.4 PROPOSED PERFORMANCE REQUIREMENT CHANGES This section discusses the effect of each set of the proposed performance requirement changes and the analyses necessary to support the changes. The present and proposed SRV performance requirement changes are shown in Table 1-1. l 1-24 l
i i NEDC-32788 The ASME has specified the acceptance criterion for relief valves in-service setpoint tolerance from *1% to *3%, as discussed in Reference 2. Consequently, as long as the maximum valve opening pressure remains below the nominal +3% limit, the plant will satisfy the code requirement and the valves can be considered capable of performing their pressure relief function. Prior to placing the refurbished valves into service, the valve opening setpoints must be adjusted to be within i1% of their nominal setting. Resetting the valves to a i1% setpoint tolerance ensures that the new refurbished valves will have the same performance characteristic prior to service. l I j l l.14
I i NEDC-32788 l L Table 1-1 COMPARISON OF PRESENT TO PROPOSED l PERFORMANCE REQUIREMENTS l-Performance Reauirement Present I.imit Pronosed I.imit
- 1. Opening pressure up to which the SRVs il%
13 % are capable of perfonning their intended function (i.e., operable). J l
- 2. Opening pressure up to which licensing basis 11%
13 % other than reload analyses have been performed.
- 3. Opening pressure up to which the cycle-specific
+3% 13 % reload licensing basis analyses have been l performed.
- 4. Tolerance beyond which additional valve testing il%
13 % is required as demonstrated by surveillance testing.
- 5. Tolerance on the as-left SRV setting prior to the 11%
il% valve being returned to service. i l
- 6. Number of SRVs being taken credit for 11 11 in these evaluations and analyses.
1 l l l i 1 i 1.44 l l
i NEDC-32788 L 1 l-l. 2.0 . ANALYSIS OVERVIEW 1 l I l This section identifies the analyses that may be affected by the proposed changes in SRV setpoint tolerance requirement. The cycle-dependent analyses performed in the following i sections assume that the plant operating parameters and the core design are consistent with the Fermi-2 Cycle 6 (Reload 5) licensing calculations [ References 1 & 3]. J ) Based on a review of the UFSAR, the following safety and regulatory aspects are identified as. 1 potentially being affected as a result of the SRV setpoint tolerance increase to i3%: H Each applicable safety and regulatory aspect identified in the above listed items was reviewed to ) determine the acceptability ofincreasing the SRV opening setpoint tolerance to 3% and with credit for 11 of 15 installed SRVs. For each analysis, the most limiting operating mode permitted by Fermi-2 operating license was assumed. i f Discussion of the individual analyses and evaluations is presented in the following sections of l this report and the conclusions are summarized in Table 2-1. l 1' I i l 2-12 l )
NEDC-32788 Table 2-1 ANALYSES PRESENTED IN THIS REPORT AND ACCEPTANCE CRITERIA Item Section Acceptance Criteria Vessel Overpressurization 3.0 Acceptable for ASME Code limit of 1375 psig ( and Technical Specification Steam Dome Pressure Limit of 1325 psig ) Thermal Limits 4.0 Acceptable for anticipated operational occurrences (MCPR & MAPLHGR j limits ) ECCS/LOCA Performance 5.0 Acceptable - PCT < 2200 F i l High Pressure System Evaluation 6.0 Acceptable for injection against +3% above lowest SRV setpoint pressure Containment Evaluations 7.0 Acceptable - within SRV discharge load margins. ATWS Mitigation 8.0 Acceptable for ASME criterion of l max. vessel pressure limit of 1500 psig, and Suppression Pool Temperature Limits Main Steam and SRV Piping Analysis 9.0 Acceptable-for highest group setpoint pressure +3% Effects on Technical Specifications 10.0 Acceptable - Changes 11% to i3% tolerance for LCO and retains il% tolerance for return to service. l I l 2-22 l ) J
NEDC-32788 l 3.0 VESSEL OVERPRESSURE PROTECTION ANALYSIS One of the design requirements of the SRVs is to limit the pressure rise to less than 110% of the [ design pressure or 1375 psig for events with a frequency of occurrence in the ASME Upset (Service Level B) category (i.e., anticipated operational occurrences or " transients"). The limiting overpressure event for Fermi-2 is the Main Steamline Isolation Valve (MSIV) Closure l with Flux Scram event (Reference 3] This event assumes the failure of the MSIV position switches to initiate a scram. The reactor is shut down by the second scram signal, the high neutron flux scram caused by the vessel pressurization and the resultant collapse of moderator voids within the reactor core. The current Technical Specifications basis of the vessel overpressure protection analyses for Fermi-2 takes credit for just 1I out of 15 SRVs and with the SRV setpoints at 1% above nominal. To justify the increase in setpoint tolerance, the analyses are performed with the assumption that all the SRV opening setpoints have drifted above their nominal trip setpoint by 3%. 3.1 OVERPRESSURE PROTECTION ANALYSIS ASSUMPTIONS The following assumptions and initial conditions were used in analyzing the MSIV closure with flux scram event for Fermi-2: O O O 9 l 3-14
NEDC-32788 l 3.2 OVERPRESSURE PROTECTION ANALYSIS RESULTS Figure 3-1 shows the reactor response results for the MSIV Flux Scram event, based on the Fermi-2 Cycle 6 loading and with the increased drift in SRV opening setpoints. J I l i Therefore, regarding the vessel overpressurization requirement, Fermi-2 satisfies the ASME limit with SRV configuration at a +3% tolerance setting and with credit for any 11 out of 15 SRVs. Table 3-1 shows the resultant peak vessel pressure for the MSIV Closure Flux Scram event analyzed and Figure 3-1 shows the time histories of key parameters during this transient event.
3.3 CONCLUSION
The analysis shows that Fermi-2 limiting overpressure event of MSIV Closure with Flux Scram, along with credit for 11 of 15 SRVs and with relaxation of SRV setpoint tolerance from il% to i3%, would not violate the ASME required peak vessel pressure limit of 1375 psig (and Technical Specification Steam Dome Pressure Limit of 1325 psig). 1 1 3-24 l l
NEDC-32788 Table 3-1 MSIV CLOSURE WITH FLUX SCRAM EVENT l i i l .I I 1 l l l 3-14 i
NEDC-32788 i I t I l l l Figure 3-1 MSIV Closure Flux Scram Event 3-44 I
NEDC-32788 4.0 THERMAL LIMITS The effect of transients on fuel thermal limits is one of the important considerations for relaxation of the SRV setpoint tolerance. The minimum critical power ratio (MCPR) is the most significant thermal limit for this evaluation. A review of the Fermi-2 analysis identified the transients that have the greatest potential effect on fuel thermal limits. 4.1 ANALYSIS AND RESULTS Reference 3 provides the results of transient analyses performed to support Fermi-2 Cycle 6 plant operation. l Therefore, the MCPR will not be affected by SRVs opening at nominal setpoint +3% or -3%. Hence, it can be concluded that extending the setpoint tolerance from il% to i3% does not affect the fuel thermal margins or limits. l 4-14
NEDC-32788 the i3% SRV setpoint tolerance relaxation will not have any impact on the MCPR transient event.
4.2 CONCLUSION
S The analysis shows that the thermal limits of the limiting anticipated operational occurrence (or l transient event), the Feedwater Controller Failure, with turbine bypass system and moisture separator reheater out of service, and reduced feedwater temperature for BOC to EEOC i exposures, would not be affected by the relaxation of SRV setpoint tolerance from i1% to i3%. l Further, other transient events remain non-limiting and bounded by the above event. t i l I 4-24 l j i l l
f NEDC-32788 f l Tsble4-1 . THE LIMITING THERMAL MARGIN CASE 1 l t I i l l l. l i I 4*14 1
1 NEDC-32788 q l 1 Figure 4-1 Plant Response to FW Controller Failure 4-d4 l
NEDC-32788 5.0 ECCS/LOCA PERFORMANCE EVALUATION The Fermi-2 LOCA analysis [ Reference 4] has been reviewed to determine the effect of an increase in the SRV opening pressures and the effect of the number of SRVs inoperable on the Emergency Core Cooling System (ECCS) performance. The SRV opening pressures were assumed to drin up to 3% above the nominal trip setpoint. The ECCS is designed to provide adequate core cooling during a postulated LOCA by limiting the calculated PCT to below the requirements of 10CFR50.46 (i.e., to less than 2200*F). A change in the SRV opening pressures can only affect the pipe break events for which SRV actuation occurs. An assessment of these events was performed to determine the effect of valve actuation at +3% over the setpoint. The intent of this assessment was to demonstrate that the limiting break (i.e., the one yielding the highest PCT) remains unaffected by these changes and that the effect on the other " breaks", which have a lower PCT, would not cause them to become the limiting break. The following postulated pipe break scenarios were considered: 5.1 LIMITING BREAK LOCA Based on a review of the results of the Reference 4 analyses, the limiting break for Fermi-2 is the design basis accident (DBA) recirculation line break. For this type of event, the reactor vessel i depressurizes very rapidly through the break itself. Because of the rapid vessel depressurization, SRV actuation will not occur. Therefore, an increase in the SRV opening setpoint does not have any adverse impact on the limiting break analysis results. 5.2 SMALL BREAK LOCA For a postulated small break LOCA, the increased SRV opening pressure due to +3% setpoint drin will not have a significant effect on the overall system response. The timing of the initial l 5-B l w
i NEDC-32788 I l l ( Based on the above discussion, the small break LOCA will remain non-limiting. l 5.3 STEAMLINE BREAK OUTSIDE THE CONTAINMENT In this event, a double-ended guillotine break of one main steamline occurs outside the containment. The reactor vessel is completely isolated upon closure of the MSIVs. As a result, the break is isolated and the reactor vessel pressure increases due to the decay heat after the reactor trip. I Therefore, the steamline break LOCA event will remain non-limiting with the proposed SRV setpoint tolerance relaxation.
5.4 CONCLUSION
Since the Fermi-2 ECCS/LOCA analysis results are insensitive to SRV opening pressure increases, the relaxation of the SRV setpoint tolerance has no adverse impact on DBA LOCA analysis. Further, other non-limiting breaks which have lower PCTs will still remain bounded by DBA LOCA case. l l 23 l 2 i
NEDC-32788 6.0 HIGH PRESSURE SYSTEM PERFORMANCE The purpose of this section is to evaluate the impact of the proposed safety / relief valve (SRV) opening setpoint tolerance relaxation from il% to 13% on the performance of the high pressure systems at the Fermi-2 power plant. The following systems are included in the evaluation: High Pressure Coolant Injection (HPCI) Reactor Core Isolation Cooling (RCIC) Standby Liquid Control System (SLCS) 6.1 IMPACT ON IIIGII PRESSURE SYSTEM PERFORMANCE The most significant impact of the SRV setpoint tolerance relaxation program on the HPCI, RCIC and SLC Systems is the increased reactor pressure specified for system operation. Tables 6-1 through 6-3, respectively, show the performance comparisons for the llPCI, RCIC and SLC Systems. 6.2 RESULTS AND CONCLUSIONS FOR IIPCI The llPCI System was found to have the capability to deliver the required flow of 5000 gpm at the increased reactor pressure resulting from relaxation of the SRV setpoint tolerance. 'Ihis relaxation does not cause the operating pressure or temperature of those system components directly impacted by the increased reactor pressure, including the valves, to exceed their original design values. l 6-F
- NEDC-32788 The 11PCI System motor-operated valves that are impacted by the increased reactor and system operating pressures have been evaluated for operability at the increased operating and differential pressures resulting from SRV setpoint tolerance relaxation. The air-operated valves have also been evaluated for operability at the increased reactor pressure. The evaluations concluded that these MOVs and AOVs can function as required at their new operating pressures [ Reference 17]. For the 11PCI System, the increase in the motor-operated valve electrical loads resulting from operation at higher differential pressures has been estimated to have a negligible impact on the plant electrical distribution systems. This is based on the results of the valve operability evaluations that demonstrated that the existing valve operators and motors have the capability to meet the new operating pressure requirements. The liPCI System includes instrumentation that measures the steam flow rate in the turbine steam supply line as a means of detecting a break in the system piping. The instrumentation indicates if an isolation should occur. The IIPCI turbine is verified to be capable of operating at the increased steam supply pressures and temperatures with this program. The 11PCI pump operating parameters remain within the original pump design envelope with this program. The impact of the SRV setpoint tolerance relaxation on the remainder of the system components was determined to be negligible because of the very small increase in operating pressure and/or temperature. 6-27 l
NEDC-32788 6.3 RESULTS AND CONCLUSIONS FOR RCIC The RCIC System was found to have the capability to deliver the required flow of 600 gpm [ Reference 9] at the increased reactor pressure resulting from relaxation of the SRV setpoint tolerance. This change does not cause the operating pressure or temperature of those system components directly impacted by the increased reactor pressure, including the valves, to exceed their original design values. The RCIC turbine is verified to be capable of operating at the increased speed, steam supply pressures and temperatures with this program. The RCIC System motor-operated valves that are impacted by the increased reactor and system operating pressures have been evaluated for operability at the increased operating and ditTerential pressures resulting from SRV setpoint tolerance relaxation. The air-operated valves have also been evaluated for operability at the increased reactor pressure. The evaluations concluded that these MOVs and AOVs can function as required at their new operating pressures [ Reference 17]. For the RCIC System, the increase in the motor-operated valve electrical loads resulting from operation at higher difTerential pressures has been estimated to have a negligible impact on the plant electrical distribution systems. This is based on the results of the valve operability evaluations that demonstrated that the existing valve operators and motors have the capability to meet the new operating pressure requirements. l 6-33
NEDC-32788 I l 1 The impact of this SRV setpoint tolerance relaxation on the remainder of the system components was determined to be negligible because of the very small increase in operating pressure and/or j temperature. 6.4 RESULTS AND CONCLUSIONS FOR STANDBY LIQUlD CONTROL SYSTEM An evaluation of the SLC System performance at the increased reactor pressure resulting from the proposed relaxation of the SRV setpoint tolerance showed that the system has the capability to deliver the required flow rate of neutron absorber solution to the RPV at the higher pressure. The impact of this proposed change on the remainder of the SLC System components was determined to be negligible because the operating pressures remain well below the SLC System design pressures. No modifications or setpoint changes are required for the SLC System as a result of this change. Operation of the SLC System is dependent on the isolation of the Reactor Water Clean Up (RWCU) System. The RWCU System is isolated by closing two isolation j valves to prevent boron removal from the reactor water. These motor-operated valves have been j evaluated for operability at the increased operating and differential pressures resulting from SRV setpoint tolerance relaxation. The evaluations concluded that these MOVs can function as required at their new operating pressures. [ Reference 17]. 6-2 l
NEDC-32788 t Table 6-1 HPCI SYSTEM PERFORMANCE COMPARISON Current SRr Design Changes l l Nominal SRV Setpoint, psig i135 1135 SRV Setpoint Tolerance il% 3% j System Flow Rate, gpm 5000 5000 Pump Brake llorsepower, bhp 4400 Rated Pump Speed, rpm 4100 Pumn Characteristics Total Dynamic IIcad at rated speed, ft 2980 2980 i Minimum Design Pressure, psig 1307 1307 l D s n Ra e S m 4100 l ( Nominal Overspeed Trip Speed, rpm 5000 Steam Flow Rate, Ibm /hr 207,000 1 l l 6-57
NEDC-32788 Table 6-2 i RCIC SYSTEM PERFORMANCE COMPARISON Current SRV Design Chanoes Nominal SRV Setpoint, psig 1135 1135 SRV Setpoint Tolerance 1% 3% System Flow Rate, gpm 600 600 Pump Brake Horsepower, bhp 660 Rated Pump Speed, rpm 4550 Pump Characteristics Total Dynamic Head at Rated Speed, ft 2940 3007 i Minimum Design Pressure, psig 1250 Turbine Characteristics Design Rated Speed, rpm 4550 Nominal Overspeed Trip Speed, rpm 5625 i Steam Flow Rate, Ibm /hr 26,400 1 6-{G l l
NEDC-32788 Table 6-3 STANDBY LIQUID CONTROL SYSTEM PERFORMANCE COMPARISON SINGLE PUMP OPERATION Current SRV llnign Changes SRV Nominal Setpoint Pressure, psig 1135 1135 SRV Setpoint Tolerance 1% 3% Nominal Injection Flow Rate, gpm 43 43 System Parameters i Pump Discharge Pressure, psig 1234 SLCS Relief Valve (i3% Tolerance) Nominal Setting, psig 1370 Minimum Setting, psig 1328 l 6-27 l
NEDC-32788 7.0 CONTAINMENT EVALUATION l The increase in the SRV setpoint tolerance to i3% was assessed to determine the potential impact on the containment design limits. The two primary areas of concern for the containment structures are (1) the pressure and temperature response and (2) the containment hydrodynamic loads. 7.1 CONTAINMENT PRESSURE AND TEMPERATURE The effect of the +3% increase in SRV setpoint tolerance on the peak containment pressure, the peak containment temperature, and the peak suppression pool temperature was considered for each of the limiting events identified in Reference 10. The most limiting event in terms of peak containment pressure and temperature response as well as peak suppression pool temperature is the design basis accident (DBA) LOCA, a double-ended guillotine break of the recirculation line. Relaxation of the SRV setpoint tolerance has no effect on this event because the vessel depressurizes without any SRV actuation. Therefore, there is no impact on the DBA-LOCA containment peak pressure and temperature or on the peak DBA-LOCA suppression pool temperature. Small steamline breaks can result in high drywell temperature conditions which can last for relatively long time periods because the vessel remains at high pressure for a longer period than for the DBA [ Reference 10]. For small steamline breaks with SRV actuation, the peak drywell temperature occurs relatively late in the event following many SRV actuation cycles. Since the SRVs will return to the low-low set setpoints following the first actuation, an increase in the SRV opening setpoint pressure for the first actuation will have a negligible effect on the peak drywell temperature. Therefore, an increase in the SRV setpoint tolerance to i3% will not have an impact on the peak drywell temperature for small steamline breaks. 7.2 SAFETY / RELIEF VALVE DYNAMIC LOADS When the SRV opens, pressure and thmst loads are exerted on the SRV discharge piping and the attached T-Quencher. Additionally, the expulsion of water and air into the suppression pool through the T-Quencher results in pressure loads on the submerged portion of the torus shell, and drag loads on submerged structures. These SRV discharge loads can be affected by the increased SRV flow rates associated with the higher opening setpoints. These loads, which are described in Fermi-2 Plant Unique Analysis Report (PUAR), Volume 1 Section 1-4.2 [ Reference 10],: 2e: l 7-12
NEDC-32788 Thrust loads on SRV discharge piping and T-Quencher Torus shell pressures Water jet loads and air bubble induced drag loads on submerged structures. The calculated stress on a particular structure is determined by combining SRV discharge loads with the loads from other various sources (e.g., Small Break Accident Loads, Safe Shutdown Earthquake, and dead weight). The loading combinations to be considered for Fermi-2 are prescribed in the PUAR, which also determines the limiting load combination, the actual load, and the allowable load for a given structure. l
7.3 CONCLUSION
S The increase of the SRV opening setpoint tolerance from 1% to i3% does not affect the DBA LOCA response of the containment in terms of containment pressure and temperature. The increase in the SRV opening setpoint tolerance from 1% to i3% affects only the first SRV actuation cycle; thus, the overall effect on the containment pressure and temperature is negligible. Although increasing the SRV opening setpoint tolerance from il% to i3% will increase the SRV discharge line loads, T-Quencher loads, and torus pool loads by less than 3%, the l conservatism and margins available show that the increase in the SRV setpoint tolerance from 11% to 13% is acceptable. Thus, increasing the SRV opening setpoint tolerance from il% to i3% is acceptable for containment evaluation. i 7-22 l
NEDC-32788 8.0 ATWS MITIGATION ANALYSIS 8.1 OVERVIEW Although evaluation of plant response for an Anticipated Transient Without Scram (ATWS) event is beyond design basis, Fermi-2 includes changed safety features to mitigate the consequences of an ATWS. The following evaluations analyze the ATWS impact on Safety Relief Valve setpoint tolerance relaxation to +3% The ATWS event which results in the maximum vessel pressure is the inadvertent MSIV closure event. This event was analyzed by taking credit for 1I out of 15 SRVs. The analysis demonstrates that the applicable overpressure protection criterion is satisfied with the SRV setpoint tolerance relaxation up to +3%. 8.2 ATWS ANALYSIS METHOD l 8-F
NEDC-32788 8.3 INPUTS AND ASSUMPTIONS 8.4 ATWS ANALYSIS RESULTS 8.4.1 Short-Term Overpressurization Event g 8-27 l
4 NEDC-32788 l i I i l i I 8.4.2 Sensitivity on Suppression Pool Temperature I i 8.4.3 Sensitivity on I) oppler and Void Reactivity i l l l 8-9
NEDC-32788 l l t
8.5 CONCLUSION
S Based on the analysis, it is concluded that changing SRV setpoint tolerance to i3% of the j-nominal setpoint would not adversely impact the vessel overpressurization criteria and the - suppression pool temperature criterion for the limiting ATWS event. t l l t I 8-9 l
NEDC-32788 Table 8-1 INITIAL OPERATING CONDITIONS FOR ATWS ANALYSIS Table 8-2 l ATWS MSIV CLOSURE TRANSIENT RESPONSES i l 8-S
NEDC-32788 i l Figure 8-1 MSIV Closure ATWS Event 8-{G l
L 1 l NEDC-32788 l I I i J Figure 8-2 MSIV Closure ATWS Event, 8-27 l
NEDC-32788 l 9.0 FERMI 2 SRV SETPOINT PIPING EVALUATION
9.1 INTRODUCTION
The consequence of the proposed SRV setpoint tolerance increase on the main steam and SRV piping was evaluated. The analysis was performed to demonstrate compliance with ASME Code requirements. 9.2 MET!!ODS AND ASSUMPTIONS l Analyses were performed in accordance with the requirements of the ASME Section III Code. Loads and cyclic duty were in accordance with the piping Design Specification. ) The SRV setpoints used for this evaluation are shown in Table 9-1. These setpoints bound the current SRV setpoints with a 13% tolerance. 9.3 RESULTS AND CONCLUSIONS Since the proposed SRV setpoint tolerance maximum pressures are lower than the setpoints used for the evaluation (Table 9-1), the piping is still acceptable. l 9-12
N NEDC-32788 Table 9-1 COMPARISON OF SRV SETPOINTS 9-22 l
r. I l NEDC-32788 10.0 TECHNICAL SPECIFICATIONS l 4 L Applicable changes to the Fermi-2 plant-specific Technical Specifications Limiting Conditions j for Operation, Surveillance Requirements, and Bases Sections 3/4.4.2 were identified. The Setpoint tolerance specified in the Limiting Conditions for Operation (LCO 3.4.2.1) is proposed to be relaxed from 1% to i3%. Additionally, the footnote to LCO 3.4.2.1 is proposed to be revised to require that, following setpoint verification testing and prior to placing' SRVs in j ervice, the setpoints must be returned to within il% of the Technical Specification-specified j s values. L 10-9 l
NEDC-32788
11.0 CONCLUSION
S Based on the evaluations and analyses performed and described in the foregoing sections of this report, it has been determined that all the proposed performance objectives listed in Table 1-1 are satisfactory for Fermi-2. The evaluations support the basis for SRV setpoint tolerance relaxation from 1% to i3% of the nominal setpoint. 1 1 l l l. 1 1 i 11-14 l
I NEDC-32788
- 12. REFERENCES l -
1. Enrico Fermi Energy Center Unit 2, " Transient Protection Parameters Verification for i Reload Licensing Analysis, Reload 5 Cycle 6"(OPL-3 Form). l 2. ANSI /ASME OM-1-1981, as referenced in Subsection IWV-3500 of the ASME Code, l Section XI,1986 Edition. 3. l 4. l 5 } i 6. 7. i 8. liPCI System Process Diagram, Doc. No. 729E626AB, Rev. 4. 9. RCIC System Process Diagram, Doc. No. 729E621 AB, Rev. 5. 10. "Enrico Fermi Atomic Power Plant Unit 2 Plant Unique Analysis Report (PUAR)", DET-04-028 Volumes 1-6, Revision 1, November 1983. I 1. " Final Test Report;In-Plant Safety Relief Valve Discharge Test; Fermi-2 Atomic Power Plant, Volume I." NUTECH Report No. DET-22-103, Revision 0, August 1987. 12. " Quick Look Test Report; In-Plant Safety / Relief Valve Discharge Test; Fermi-2 Atomic Power Plant, Volume I," NUTECH Report No. DET-22-102, Revision 0, April 1987. 13. "SRV Setpoint Tolerance Margin Analysis" DE&S Document VT5900-TR-03, October 20,1997. 14 15 Assessment of BWR Mitigation of ATWS, Volume II, NEDE-24222, December 1979. l 12-12
NEDC-32788 L i 16 1 l-17 Fermi-2 Design Calculation DC-5939, Volume I, " Evaluation of Selected MOVs and AOVs in HPCI, RCIC,and RWCU for SRV Setpoint Tolerance Change from 1135 psig il% to 1135 psig i3%." i 12-22 l 1}}