ML11286A093
| ML11286A093 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/20/2011 |
| From: | Reza M, Roberts K AREVA |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 3F1011-05 86-9159609-002 | |
| Download: ML11286A093 (157) | |
Text
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 /LICENSE NUMBER DPR-72 ENCLOSURE 1 AREVA 86-9159609-002 - CR-3 STEAM GENERATOR TUBE RUPTURE EVENT FOR EPU
SUMMARY
0402-01-FO1 (Rev. 016, 03/31/2011)
A CALCULATION
SUMMARY
SHEET (CSS)
AREVA Document No. 86 - 9159609 - 002 Safety Related: M_Yes D No Title CR-3 Steam Generator Tube Rupture Event for EPU Summary PURPOSE AND
SUMMARY
OF RESULTS:
PURPOSE:
Progress Energy Florida, Inc. has requested that AREVA NP, Inc. perform a safety analysis of the steam generator tube rupture (SGTR) event and the control room habitability (CRH) event for the Extended Power Uprate (EPU) for the Crystal River Unit 3 plant. To support this program, the analysis was performed in accordance with the analytical input summary (AIS) [1]. The analysis includes the SGTR event with no loss of offsite power (No-LOOP) and the CRH event with LOOP. For each event, an initial base analysis was performed and a later final analysis was performed. This document summarizes the thermal hydraulic results from the four analyses found in Reference [2].
SUMMARY
OF RESULTS:
The results of the SGTR event with no LOOP base case are shown in Figure 6-1 through Figure 6-23.
Sequences of events are presented in Table 6-7. A summary of results are provided in Table 6-8.
The results of the SGTR event with no LOOP final case are shown in Figure 6-24 through Figure 6-46.
Sequences of events are presented in Table 6-16. A summary of results are provided in Table 6-17.
The results of the CRH event with LOOP base case are shown in Figure 7-1 through Figure 7-31. Sequences of events are presented in Table 7-8. A summary of results are provided in Table 7-9.
The results of the CRH event with LOOP final case are shown in Figure 7-32 through Figure 7-62. Sequences of events are presented in Table 7-17. A summary of results are provided in Table 7-18.
THE DOCUMENT CONTAINS ASSUMPTIONS THAT SHALL BE THE FOLLOWING COMPUTER CODES HAVE BEEN USED INTHIS DOCUMENT: VERIFIED PRIOR TO USE CODENERSION/REV CODENERSION/REV E YS YES RELAP-5/MOD2/REV 27 *- O Page 1 of 156
A AR EVA 0402-01 -FO1 (Rev. 016, 03/31/2011)
Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Review Method: M Design Review (Detailed Check)
FD1 Alternate Calculation Signature Block Pages/Sections Prepared/Reviewed/Approved All Revision 002 changes All Revision 002 changes All Revision 002 changes Reviewer Independence Note: P/R/A designates Preparer (P), Reviewer (R), Approver (A);
LP/LR designates Lead Preparer (LP), Lead Reviewer (LR)
Project Manager Approval of Customer References (N/A if not applicable)
Name Title (printed or typed) (printed or typed) Signature Date Ken L. Greenwood b, Project Manager 0 VTw(A.5 5e-0 fiP 1________ _________
Mentoring Information (not required per 0402-01)
Name Title Mentor to:
(printed or typed) (printed or typed) (P/R) Signature Date N/A Page 2
A 0402-01-FOl (Rev. 016, 03/31/2011)
AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Record of Revision Revision Pages/Sections/Paragraphs No. Changed Brief Description / Change Authorization 002 Complete Issue This document is 156 pages in its entirety. Pages numbered 1-156. Removed proprietary information.
Sections 6.1.3, 6.2.3, 7.1.3, 7.2.3, Editorial changes to address customer comments.
Table 7-2, and Table 7-11 Figure 6-45 Corrected figure.
001 Complete Issue This document is 156 pages in its entirety. Pages numbered 1-156.
Section 1.0 Updated to include base and final cases for SGTR and CRH.
Section 5.0 Added additional criteria.
Section 6.0 Changed titles to indicate Non LOOP and clarified text.
Section 6.1 Changed titles to indicate Non LOOP base case.
Section 6.2 Added Non LOOP final'case.
Section 7.0 Added CRH/LOOP cases.
Section 8.0 Added References 15 and 16.
000 All Sections. Original release.
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-I +
-I 4 I I i i Page 3
A AR.EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table of Contents Page SIG NATURE BLOCK ................................................................................................................................ 2 RECO RD O F REVISIO N .......................................................................................................................... 3 LIST O F TABLES ..................................................................................................................................... 6 LIST O F FIG URES ................................................................................................................................... 8 1.0 PURPO SE ................................................................................................................................... 12 2.0 ANALYTICAL METHO DO LO GY ............................................................................................. 12 3.0 KEY ASSUM PTIONS .................................................................................................................. 12 4.0 CO MPUTER CO DE .................................................................................................................... 12 5.0 ACCEPTANCE CRITERIA .......................................................................................................... 12 6.0 NO N LOO P EVENT .................................................................................................................... 13 6 .1 Ba s e C a s e ...................................................................................................................................... 13 6.1.1 Event Description .......................................................................................................... 13 6 .1.2 Input ................................................................................................................................. 13 6 .1.3 Re s u lts ............................................................................................................................. 18 6 .2 Fina l C a s e ..................................................................................................................................... 44 6.2.1 Event Description .............................................................................................................. 44 6 .2 .2 Inp ut ................................................................................................................................. 44 6 .2 .3 Re s u lts .............................................................................................................................. 50 7.0 CO NTRO L ROO M HABITABILITY / LOO P EVENT ................................................................ 76 7 .1 Ba s e Ca se ...................................................................................................................................... 76 7.1.1 Event Description ...................................................................................................... 76 7 .1 .2 Input ................................................................................................................................. 76 7.1.3 Re s ults ............................................................................................................................. 82 7 .2 Fina l Ca s e .................................................................................................................................... 1 16 7.2.1 Event Description ........................................................................................................... 116 7.2 .2 Input ..................................................................... :................ ........................................... 1 16 Page 4
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary.
Table of Contents (continued)
Page 7 .2 .3 R e s u lts ........................................................................................................................... 12 2 8 .0 R E F E R E N C E S ......................................................................................................................... 156 Page 5
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary List of Tables Page Table 6-1: SGTR Inputs and Boundary Conditions for CR Base Case ....................................... 14 Table 6-2: Pressurizer Heater Setpoints for Banks D and E [11] - Base Case ................................. 15 Table 6-3: Makeup Line Flow versus Pressure for One Pump [12]- Base Case ............................. 16 Table 6-4: MSSV Setpoints - Base Case ........................................................................................... 16 Table 6-5: H PI Flow [13] - Base C ase ............................................................................................... 17 Table 6-6: Control Rod Insertion [10]- Base Case ............................................................................ 17 Table 6-7: SGTR Sequence of Events - Base Case ........................................................................ 19 Table 6-8: SGTR Summary of Results - Base Case ........................................................................ 20 Table 6-9: SGTR Inputs and Boundary Conditions for CR Final Case ........................................ 45 Table 6-10: Pressurizer Heater Setpoints for Banks D and E [11] - Final Case ............................... 46 Table 6-11: Makeup Line Flow versus Pressure for One Pump [12] - Final Case ........................... 47 Table 6-12: MSSV Setpoints - Final Case ........................................................................................ 47 Table 6-13: H PI Flow [13] - Final C ase ............................................................................................. 48 Table 6-14: Control Rod Insertion [10] - Final Case .......................................................................... 48 Table 6-15: ISC M C urve - Final C ase ............................................................................................... 49 Table 6-16: SGTR Sequence of Events - Final Case ..................................................................... 51 Table 6-17: SGTR Summary of Results - Final Case ..................................................................... 52 Table 7-1: CRH Inputs and Boundary Conditions for CR Base Case .......................................... 77 Table 7-2: Pressurizer Heater Parameters - Base Case ................................................................. 78 Table 7-3: Makeup Line Flow versus Pressure for One Pump [12] - Base Case ............................. 79 Table 7-4: MSSV Setpoints - Base Case ........................................................................................... 79 Table 7-5: H PI Flow [13] - Base C ase ............................................................................................... 80 Table 7-6: Control Rod Insertion [10]- Base Case ............................................................................ 80 Table 7-7: Summary of Key Operator Actions - Base Case ............................................................. 81 Table 7-8: SGTR CRH Sequence of Events - Base Case .................................................................... 83 Table 7-9: SGTR CRH Summary of Results - Base Case ................................................................. 84 Table 7-10: SGTR Inputs and Boundary Conditions for CR Final Case ........................................ 117 Table 7-11: Pressurizer Heater Parameters - Final Case ................................................................... 118 Table 7-12: Makeup Line Flow versus Pressure for One Pump [12] - Final Case .............................. 119 Table 7-13: MS SV Setpoints - Final C ase ........................................................................................... 119 T able 7-14 : H P I Flow [13 ] - Final C ase ................................................................................................ 120 Page 6
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary List of Tables (continued)
Page Table 7-15: Control Rod Insertion [10] - Final Case ............................................................................. 120 Table 7-16: Summary of Key Operator Actions - Final Case .............................................................. 121 Table 7-17: SGTR CRH Sequence of Events - Final Case ................................................................. 123 Table 7-18: SGTR CRH Summary of Results - Final Case ................................................................ 124 Page 7
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary List of Figures Page Figure 6-1: Reactor Power - Base Case .......................................................................................... 21 Figure 6-2: Steam Generator Power - Base Case ........................................................................... 22 Figure 6-3: Reactor Coolant System Temperature - Base Case .................................................... 23 Figure 6-4: Primary Pressure - Base Case ...................................................................................... 24 Figure 6-5: Indicated Pressurizer Water Level - Base Case ............................................................. 25 Figure 6-6: Steam Generator Outlet Pressure - Base Case ............................................................ 26 Figure 6-7: Steam Generator Flow Rate - Base Case ..................................................................... 27 Figure 6-8: Steam Generator Inventory - Base Case ........................................................................ 28 Figure 6-9: Total Reactor Coolant System Flow Rate - Base Case ............................... 29 Figure 6-10: Reactor Coolant Pump Flow Rate - Base Case .......................................................... 30 Figure 6-11: Pressurizer Surge Line Flow Rate - Base Case .......................................................... 31 Figure 6-12: Steam Generator Secondary Side Collapsed Level - Base Case ............................... 32 Figure 6-13: Core Average Fuel Temperature - Base Case ............................................................ 33 Figure 6-14: Reactivity - Base C ase ................................................................................................. 34 Figure 6-15: Steam Generator Ruptured Tube Flow - Base Case.................................................. 35 Figure 6-16: Makeup and Letdown Flow - Base Case ..................................................................... 36 Figure 6-17: High Pressure Injection Flow - Base Case ................................................................... 37 Figure 6-18: Pressurizer Heat - Base Case .................................................................................... 38 Figure 6-19: Turbine Bypass Flow - Base Case ............ ................................................................. 39 Figure 6-20: Pressurizer Spray - Base Case ................................................................................. 40 Figure 6-21: Main Steam Safety Valve Flow - Base Case ............................................................... 41 Figure 6-22: Subcooling Margin - Base Case ................................................................................. 42 Figure 6-23: Steam Generator Tube-to-Shell Temperature Difference - Base Case ........................ 43 Figure 6-24: Reactor Power - Final Case....................................................................................... 53 Figure 6-25: Steam Generator Power - Final Case .......................................................................... 54 Figure 6-26: Reactor Coolant System Temperature - Final Case .................................................... 55 Figure 6-27: Primary Pressure - Final Case ..................................................................................... 56 Figure 6-28: Indicated Pressurizer Water Level - Final Case .......................................................... 57 Figure 6-29: Steam Generator Outlet Pressure - Final Case ........................................................... 58 Figure 6-30: Steam Generator Flow Rate - Final Case .................................................................... 59 Figure 6-31: Steam Generator Inventory - Final Case ..................................................................... 60 Page 8
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary List of Figures (continued)
Page Figure 6-32: Total Reactor Coolant System Flow Rate - Final Case ................................................ 61 Figure 6-33: Reactor Coolant Pump Flow Rate - Final Case ........................................................... 62 Figure 6-34: Pressurizer Surge Line Flow Rate - Final Case ........................................................... 63 Figure 6-35: Steam Generator Secondary Side Collapsed Level - Final Case ................................. 64 Figure 6-36: Core Average Fuel Temperature - Final Case ............................................................. 65 Figure 6-37: R eactivity - Final C ase ................................................................................................. 66 Figure 6-38: Steam Generator Ruptured Tube Flow - Final Case ................................................. 67 Figure 6-39: Makeup and Letdown Flow - Final Case ...................................................................... 68 Figure 6-40: High Pressure Injection Flow - Final Case .................................................................... 69 Figure 6-41: Pressurizer Heat- Final Case ....................................................................................... 70 Figure 6-42: Turbine Bypass Flow - Final Case ............................................................................... 71 Figure 6-43: Pressurizer Spray - Final Case ..................................................................................... 72 Figure 6-44: Main Steam Safety Valve Flow - Final Case ............................................................... 73 Figure 6-45: Subcooling Margin - Final Case .................................................................................... 74 Figure 6-46: Steam Generator Tube-to-Shell Temperature Difference - Final Case ........................ 75 Figure 7-1: Reactor Power - Base Case ........................................................................................... 85 Figure 7-2: Steam Generator Power - Base Case ............................................................................ 86 Figure 7-3: Reactor Coolant System Temperature - Base Case ...................................................... 87 Figure 7-4: Primary Pressure - Base Case ...................................................................................... 88 Figure 7-5: Indicated Pressurizer Water Level - Base Case ............................................................. 89 Figure 7-6: Steam Generator Outlet Pressure - Base Case ............................................................ 90 Figure 7-7: Steam Generator Flow Rate - Base Case ..................................................................... 91 Figure 7-8: Steam Generator Inventory - Base Case ....................................................................... 92 Figure 7-9: Total Reactor Coolant System Flow Rate - Base Case ................................................ 93 Figure 7-10: Reactor Coolant Pump Flow Rate - Base Case .......................................................... 94 Figure 7-11: Pressurizer Surge Line Flow Rate - Base Case .................................. 95 Figure 7-12: Steam Generator Secondary Side Collapsed Level - Base Case ............................... 96 Figure 7-13: Core Average Fuel Temperature - Base Case ....... :........................ 97 Figure 7-14: R eactivity - Base C ase ................................................................................................. 98 Figure 7-15: Steam Generator Ruptured Tube Flow - Base Case ........................................................ 99 Page 9
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary List of Figures (continued)
Page Figure 7-16: Makeup and Letdown Flow - Base Case ........................................................................ 100 Figure 7-17: High Pressure Injection Flow - Base Case ..................................................................... 101 Figure 7-18: Pressurizer Heat - Base Case ........................................................................................ 102 Figure 7-19: Emergency Feedwater Flow - Base Case ...................................................................... 103 Figure 7-20: Pressurizer PORV Spray - Base Case ........................................................................... 104 Figure 7-21: Main Steam Safety Valve Flow - Base Case .................................................................. 105 Figure 7-22: Instantaneous ADV Flow - Base Case ........................................................................... 106 Figure 7-23: Subcooling Margin - Base Case ..................................................................................... 107 Figure 7-24: Steam Generator Tube-to-Shell Temperature Difference - Base Case .......................... 108 Figure 7-25: Reactor Coolant System Mass - Base Case .................................................................. 109 Figure 7-26: Flashing Fraction for Ruptured Steam Generator - Base Case ...................................... 110 Figure 7-27: Condenser Flow - Base Case ......................................................................................... 111 Figure 7-28: Core Exit Temperature - Base Case ............................................................................... 112 Figure 7-29: ADV Valve Steam - Base Case ...................................................................................... 113 Figure 7-30: Integrated ADV and MSSV Flow - Base Case ................................................................ 114 Figure 7-31: Steam Generator Secondary Side Liquid Level - Base Case ......................................... 115 Figure 7-32: R eactor P ower- Final C ase ............................................................................................ 125 Figure 7-33: Steam Generator Power - Final Case ......................................... 126 Figure 7-34: Reactor Coolant System Temperature - Final Case ....................................................... 127 Figure 7-35: Primary Pressure - Final Case ........................................................................................ 128 Figure 7-36: Indicated Pressurizer Water Level - Final Case ............................................................. 129 Figure 7-37: Steam Generator Outlet Pressure - Final Case ............................................................. 130 Figure 7-38: Steam Generator Flow Rate - Final Case ....................................................................... 131 Figure 7-39: Steam Generator Inventory - Final Case ........................................................................ 132 Figure 7-40: Total Reactor Coolant System Flow Rate - Final Case .................................................. 133 Figure 7-41: Reactor Coolant Pump Flow Rate - Final Case .............................................................. 134 Figure 7-42: Pressurizer Surge Line Flow Rate - Final Case .............................................................. 135 Figure 7-43: Steam Generator Secondary Side Collapsed Level - Final Case ................. 136 Figure 7-44: Core Average Fuel Temperature - Final Case ................................................................ 137 Figure 7-45: R eactivity - F ina l C ase .................................................................................................... 138 Page 10
A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary List of Figures (continued)
Page Figure 7-46: Steam Generator Ruptured Tube Flow - Final Case .......................... 139 Figure 7-47: Makeup and Letdown Flow - Final Case ........................................................................ 140 Figure 7-48: High Pressure Injection Flow - Final Case ...................................................................... 141 Figure 7-49: P ressurizer H eat - Final C ase ......................................................................................... 142 Figure 7-50: Emergency Feedwater Flow - Final Case ....................................................................... 143 Figure 7-51: Pressurizer PORV Spray - Final Case ............................................................................ 144 Figure 7-52: Main Steam Safety Valve Flow - Final Case .................................................................. 145 Figure 7-53: Instantaneous ADV Flow - Final Case ............................................................................ 146 Figure 7-54: Subcooling Margin - Final Case ...................................................................................... 147 Figure 7-55: Steam Generator Tube-to-Shell Temperature Difference - Final Case ........................... 148 Figure 7-56: Reactor Coolant System Mass - Final Case ............................... 149 Figure 7-57: Flashing Fraction for Ruptured Steam Generator- Final Case ...................................... 150 Figure 7-58: C ondenser Flow - Final C ase ......................................................................................... 151 Figure 7-59: Core Exit Temperature - Final Case .............................................................................. 152 Figure 7-60: ADV Valve Steam - Final Case ....................................................................................... 153 Figure 7-61: Integrated ADV and MSSV Flow- Final Case ................................................................ 154 Figure 7-62: Steam Generator Secondary Side Liquid Level - Final Case ......................................... 155 Page 11
A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary 1.0 PURPOSE Progress Energy Florida, Inc. had requested that AREVA NP, inc. perform a safety analysis of the steam generator tube rupture (SGTR) event and the control room habitability (CR1-I) event for the Extended Power Uprate (EPU) for the Crystal River Unit 3 plant. To support this program, the analysis was performed in accordance with the analytical input summary (AIS) [1]. The analysis includes the SGTR event with no loss of offsite power (No-LOOP) and the CRH event with LOOP. For each event, an initial base analysis was performed followed by a final analysis. This document summarizes the thermal hydraulic results from the four analyses found in Reference [2].
2.0 ANALYTICAL METHODOLOGY The thermal-hydraulic analysis of the double-ended rupture of a single steam generator tube at CR-3 at the full power level with nominal operating conditions was performed with the RELAP5/MOD2-B&W computer program [3]. The RELAP5/MOD2-B&W CR-3 plant model was obtained from [4] and was modified as required to analyze the SGTR event. Then general non-LOCA RELAP5/MOD2-B&W analysis modeling methods and restrictions outlined in [5] were followed for the CR-3 SGTR analysis.
At Progress Energy Florida, Inc.'s request, the CR-3 plant model was based on an uprated power level of 3014MWt and 5 percent steam generator tube plugging, with 100% of the tube plugging placed in the wetted region. Control variables were added to the RELAP5/MOD2-B&W input deck to allow the calculation of activity releases during the SGTR event. Additionally, the leak rate calculated by RELAP5 varied according to the primary to secondary pressure differential.
3.0 KEY ASSUMPTIONS A key assumption is any assumption or limitation that must be verified prior to using the results and/or conclusions of a calculation for a safety-related task. There are no key assumptions in the present calculation.
4.0 COMPUTER CODE The SGTR analysis was performed using the RELAP5/MOD2-B&W computer code [3]. This code has been approved by the NRC for use in both LOCA [6] and non-LOCA safety analysis [5]. The code allows modeling of both the primary and secondary systems and the fill systems, such as high pressure injection (HPI) and emergency feedwater (EFW). No code interfaces are required for this analysis.
5.0 ACCEPTANCE CRITERIA The acceptance criteria for the evaluation of this accident as stated in the updated FSAR are:
a) The RELAP5 system calculation for the SGTR will not determine the dose consequences but will supply system information so doses can be calculated. The dose acceptance criteria are defined in the EPU dose AIS.
b) The event must not result in additional tube failures and further degradation of the integrity of the reactor coolant pressure boundary caused by the effects of temperature gradients (thermally induced tube loadings).
The stress limits, -90/+140 'F, are provided in Reference [7] and temperatures need to be within stated limits.
Note that in this calculation the tube-to-shell temperature difference is calculated by subtracting the tube temperature from the shell temperature which is opposite of Reference [7].
Page 12
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Additional acceptance criteria for the evaluation of this accident are:
c) There must be no ROTSG overfill for either ROTSG until the RCS temperature is below 500 F and RCS pressure is below 1000 psig for the CR1H event.
6.0 NON LOOP EVENT The purpose of the steam generator tube rupture transient analysis is to calculate the total steaming and leak rate to the atmosphere until the ruptured steam generator is isolated. As stated in Section 1.0, an initial base analysis was performed followed by a final analysis.
6.1 Base Case 6.1.1 Event Description The double-ended rupture of a steam generator tube is assumed to occur at full power. Primary system pressure decreased until a reactor trip occurs on low reactor coolant pressure. The turbine trips as a result of the reactor trip and the turbine stop valves (TSVs) close rapidly. Closure of the TSVs causes the main steamline pressure to increase and open the turbine bypass valves (TBVs), atmospheric dump valves (ADVs), and the main steam safety valves (MSSVs). Post-trip secondary pressure control was accomplished via the main steam safety valves.
Allowing the pressure to be controlled by the MSSV lift and reseat pressure maximized dose release during this event. The lift setpoints of the valves were nominal minus three percent (Table 6-4) to allow the earlier lift of the MSSVs and to produce the maximum amount of steam release. No credit was taken for relief through the ADVs immediately following reactor trip. After the reactor trip, the RCS pressure continues to decrease until high pressure injection (HPI) is actuated. At 40 minutes [8], the operator began a controlled reduction of pressure using the pressurizer (PZR) spray. Immediately after ESAS actuation [8], the operator begins to throttle HPI to control pressurizer level and subcooling margin. At 45 minutes after reactor trip the operator begins an approximately 100
°F/hr cooldown of the RCS using the affected and unaffected SG turbine bypass valves (TBVs). The affected SG is isolated after the RCS temperature is reduced to 500 'F and 1000 psig [9].
6.1.2 Input The inputs and boundary conditions necessary for the analysis of the CR-3 SGTR event are included in Table 6-1.
through Table 6-6.
Page 13
A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 6-1: SGTR Inputs and Boundary Conditions for CR Base Case Parameter Value RCS Conditions Core Power 1.004*3014 MWt = 3026.1 MWt Pump Power 4.03 MWt per pump Decay Heat 1.0*ANS 1971 w/ B&W Heavy Actinides Primary Side T,,vc 582 OF RCS Pressure @ Hot Leg Tap 2170 psia Total RCS Mass Flow Rate 374,880 gpm Make-Up Flow Table 6-3 Pressurizer Parameters Indicated Pressurizer Level 220 inches Heaters / Sprays modeled (See Reference [1] Section Pressurizer Sprays and Heaters 6.1.5 for heater Capacities)
PORV not modeled Main Feedwater System Parameters MFW Temperature 460 OF Initial MFW Flow per SG
- 1800 lbm/sec per SG MFW pump run back to 3 % flow over 0 seconds on MFW Flow Transient Response reactor trip and control to SG low-level setpoint of 32.5 inches Steam Generator Parameters Steam Generator Tube Plugging 5%
Steam Generator Inventory 65%-85% OR Turbine and Main Steam Parameters TBV Actuation Signal Reactor trip and 1010 psig TBV Stroke Time 5 seconds TBV Capacity 837,000 lbm/hr/valve (1 TBV modeled per SG)
MSSV Setpoints and Capacity Table 6-4 Turbine Trip Reactor Trip + 0.50 sec delay Turbine Stop Valve Closure Turbine Trip + 0.50 sec stroke MSIV Isolation (EFIC) Setpoint not modeled MSIV Stroke Time not modeled Atmospheric Dump Valve Capacity not modeled Page 14
A Document No. 86-9159609-002 AR EVA CR-3 Steam Generator Tube Rupture Event for EPU Summary ADV Opening Setpoint not modeled ADV Stroke Time not modeled RPS Parameters Low Pressure Reactor Trip 1893.95psia Reactor Trip Delay 0.61seconds ESAS Low Pressure Trip 1714 psia ESAS Delay Time 0 seconds ECCS Parameters BWST Temperature 100 OF HPI Flow Table 6-5 Reactivity Control Parameters Control Rod Drop Time 1.4 seconds to 2/3 insertion Control Rod Insertion vs. Time Table 6-6 Delayed Neutron Fraction 0.007 (REA)
Prompt Neutron Generation Time 24.8 ptsec Doppler Coefficient -2.0x10-5 Ak/k/°F Moderator Temperature Coefficient 0.0x 10-4 Ak/k/°F
- The MFW flow rate can be adjusted during initialization to achieve a steady-state.
Table 6-2: Pressurizer Heater Setpoints for Banks D and E [11] - Base Case Bank "On" Pressure "Off" Pressure (psia) (psia)
D 2134.7 2154.7 E 2119.7 2139.7 Page 15
A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 6-3: Makeup Line Flow versus Pressure for One Pump [12] - Base Case RCS Pressure MU Line Flow RCS Pressure MU Line Flow (psia) (gpm) (psia) (Ibm/s)1 900 298 900 41.08 1100 284 1100 39.18 1300 270 1300 37.27 1500 255 1500 35.22 1700 238 1700 32.89 1900 220 1900 30.42 2100 200 2100 27.67 2300 177 2300 24.50 2500 150 2500 20.78 2700 117 2700 16.22
- 1. The MU line flow was converted to Ibm/s using a density for the conditions of the RCS pressure and a reasonable temperature of 120 'F. Slight variations in the pressure and temperature will have a negligible impact on the density of the makeup flow.
Table 6-4: MSSV Setpoints - Base Case Bank Number of Valves Nominal Setpoint Lift Pressure1 Reseat Pressure 2 (psig) (psig) (psig) 1 2 1050 1018.5 967.5 2 2 1070 1037.9 986.0 3 2 1090 1057.3 1004.4 4 23 1100 1067 1013.6 Notes:
- 1. Lift pressure represents 3% lift tolerance from nominal setpoint.
- 2. Reseat pressure represents 5% blowdown from lift pressure.
- 3. Per Reference [1], MSSV with a capacity of 583,574 Ibm/hr has a nominal setpoint of 1100 psig.
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A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 6-5: HPI Flow [13] - Base Case RCS Pressure (psia) Flow (gpm) 15 904.3 615 815.9 1215 712.1 2515 373.3 Table 6-6: Control Rod Insertion [10] - Base Case Time (sec) Percent Reactivity Inserted (%)
0.0 0.00 0.2 0.58 0.3 0.99 0.4 1.83 0.6 5.29 0.8 12.33 1.0 21.41 1.2 33.09 1.4 50.75 1.6 72.96 1.8 91.30 2.0 99.26 2.2 99.99 2.3 100.00 2.4 100.00 Page 17
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary 6.1.3 Results The results of the SGTR analysis base case are shown in Figure 6-1 through Figure 6-23. Sequences of events are presented in Table 6-7. A summary of results are provided in Table 6-8.
The initiating event was a double-ended rupture of a steam generator tube just below the lower face of the upper tube sheet. The primary system pressure decreased due to the loss of inventory (Figure 6-4). At approximately 17 minutes the RCS pressure begins to drop quickly (Figure 6-4) due to the PZR heaters tripping on low PZR level (Figure 6-18) and at 23 minutes the PZR empties out (Figure 6-5). Reactor trip occurred at 33.656 minutes on low reactor coolant pressure (Figure 6-4). The turbine tripped as a result of the reactor trip and the turbine stop valves closed within one second. Closure of the TSVs caused the main steam line pressure to increase and open the turbine bypass valves and the main steam safety valves. Within the next two minutes, the MSSVs reseated and the TBVs began automatically controlling secondary pressure near 1024.7 psia (Figure 6-6). SG levels were maintained at a level of 3.3698 feet above the upper face of the lower tube sheet (UFLTS) with main feedwater (Figure 6-12).
Following reactor trip, the low RCS pressure ESAS signal was reached within 20 seconds and HPI from two pumps was initiated (Figure 6-17). The HPI flow more than offset the tube leak and minimum pressurizer liquid level and RCS pressure were quickly regained (Figure 6-4 and Figure 6-5).
Forty-five minutes after the reactor trip, the TBVs on both steam generators were taken tinder manual control by the operator to reduce steam pressure (Figure 6-6) and begin the cooldown of the RCS (Figure 6-3). The core exit subcooling margin never reached 110 'F (Figure 6-22). Therefore, no operator action was needed to control the subcooling margin between 110 'F and 90 'F (Figure 6-20). As the primary system pressure decreased, the mass flow rate through the ruptured tube also decreased. At approximately 112 minutes, the core exit temperature reached 500 'F. At this time the affected SG was considered isolated and the analysis was terminated. The MSSV lift duration is not expected to exceed 100 seconds during the remainder of the cooldown. The total SG ruptured tube flow is not expected to exceed 40 lbm/sec during the remainder of the cooldown.
The SG tube-to-shell temperature differences (Figure 6-23) from this analysis are approximately -10/+30 'F, which is well within the acceptable limit of -90/+140 'F (Section 5.0).
Larger tensile temperature differences would occur as the operator continues to cool down the RCS, but these values would be bounded by existing normal cooldown calculations. Consequently, additional loss of reactor coolant boundary integrity would not occur due to resultant temperature gradients during the SGTR transient, thus the acceptance criterion for the event is met.
No fuel failures are expected to occur during the SGTR event.
Page 18
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 6-7: SGTR Sequence of Events - Base Case EVENT Time, seconds (minutes)
Reactor Trip on low RCS pressure 2019.3 (33.66)
Control Rods Insert and Turbine Trip 2020.0 (33.67)
SG A: 2021.1 (33.69)
'MSSVs first open SG B: 2021.2 (33.69)
ESAS low RCS pressure Signal 2032.6 (33.88)
HPI flow from two pumps started 2032.6 (33.88) 2 SG A: 4733.5 (78.89)
MSSVs last close SGB: 2100.4 (35.01)
Operator begins RCS cooldown using TBVs on both 4719.4 (78.66)
SGs.
Thot reaches 500 'F. Analysis terminated. 6746.1 (112.44)
Notes:
Page 19
A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 6-8: SGTR Summary of Results - Base Case Parameter Results Integrated tube leak 230,128 Ibm Integrated tube leak 37,625.6 gal Integrated total MSSV release 67,022.3 ibm Integrated SG A MSSV release 36,038.4 ibm Integrated SG B MSSV release 30,983.8 lbm Integrated total TBV release 399,014 ibm Integrated SG A TBV release 256,298 ibm Integrated SG B TBV release 142,716 lbm SG A: 92.6 seconds MSSV lift duration SG B: 72.4 seconds Minimum subcooling margin 17.63 'F @ 2037 seconds Maximum tube-to-shell temperature difference SG A: 27.6 'F @ 6746 seconds (shell temp - tube temp) SG B: 29.6 'F@ 6746 seconds Calculated RCS cooldown rate (TEMPF 4719.35 to 6746.1 seconds) 95.439 °F/hr SGTR event duration 6746.1 seconds Page 20
Figure 6-1: Reactor Power - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 3200
-,D Reactor t
2800 2400 2000 1600 Q
-. -.-. . . ..-
-.- . ...- . .-
. . ..-
..- . .-.-
..-
- - .-.-
- .-. .-.-
- . .-.-
.- - -.-.--
.-. .-.-. .-.-
.-
- - .-.- . .-.-
-.-.
.- -.-.-
-.-.
.-.--.-. .-.-
-.-.-
-.-.
.-.-
-.-. -.-.-
.-.-
-.-. -.-.
.-.- .-.- . ..-.-.
. . .-. -. - -
1200 800
--
--. ~~ ~~ - - --------------
- - - -' -------------- -------------- o o
o 400 -- -- -- ----
00 ON 0
0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 ON Time (s) r'Q UATL UJWW -pOOI
Figure 6-2: Steam Generator Power - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 1600
--e SG A
-- < SG B 1400 ------------- ---------------------------- --------------- -------------
1200 I----------- ------------ -------------
1000 800 ----- --------
Q 0
600 400 200 ------------------------
00
--
--
---
-- --
0 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s)
UATL UJWW -p00 2
Figure 6-3: Reactor Coolant System Temperature - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 650
-0 Tave
-- , TH-1
-A TH-2 625 ---------------
-- -------- -- E TC-1A
-- o TC-1B TC-2A TC-2B 600 ---------------
575 ----- --------
I----- -------- -------------
W
--- ------- . ..- - -- - .-- . . . .
.-.--
550 ----- -------
525 ------------- ----- -------
_,. . . -. .--. .--
. .-. . .--..
. . . . .-. .
E H/
500 -------------- - ------------- ------------- ------------------
475 -------------- ---------- I
450 ------------- ------------- ------------ -------------
425 ------------- ------------- ---------- -------------
00 400 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s)
UATL UJWW -pO03
Figure 6-4: Primary Pressure - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 2300
---< SG B, Hot Leg P-,T Tap
---------
2200 -------------- --------
---
2100 1--- ----- -------------- -- ............
------------- ---------
2000 ------------- ------------
--------
-1
------ ---------
1900 1 --------------
x
:--
Cn~
1800 I -------------
-------
----- -- -- -- ------ ------ -- --- ---------
I---------
1700 ----------------- - --------- --------------- -------- ---- ------- ------- ---------
1600 1 ------- ---- ---------------------
--------- I--------
1500 ------------- ------------- ------------- ---------
1400 1 ------------- --- ---------- ----------------------
00 1300 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s)
UATL UJWW -pO04
Figure 6-5: Indicated Pressurizer Water Level - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 360
-- o Pressurizer ]
300 C.)
I 180 ------- - - - -- - -
O9 N ON 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s) k)
005 UATL UJWW -p
Figure 6-6: Steam Generator Outlet Pressure - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 1200
-- SG B MSL B-2 1100 1000 -----------
- --------------------- -----I
- , I*._"* * ,
900 -- -----------------
800 C
700 ------ -- - --
- ------ --------- - -
600 500 I0 400 0 500 1000 1500 2000 2500 3000 '3500 4000 '4500 5000 5500 6000 6500 7000 Time (s) 00 6 UATL UJWW _p
Figure 6-7: Steam Generator Flow Rate - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 9000
-- 4
-- 0 SG A MS
-A SG B MS I 800, ....... -- SGBMFW
.............
1600 ---------------
1400 1200 E 1000 800 600 400 ----------- ------------- ------------- -------------------------- ------------ ------ ------------
-- - - - - - - - .......... -------------
200 ------- - -------- - I --------------------------I-------------
00 E3
.1 0 0 500 1000 1500 2000 2500 3000 3500 4000 4500 - 5000 5500 6000 6500 7000 Time (s) 7 UATL UJWW -p00
Figure 6-8: Steam Generator Inventory - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 80000 72000 ----------------
64000 56000 S. 48000 I-0 40000 0.' 32000 i(
24000 16000 8000 ON 0 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 (I C
Time (s) t0 8
UATL UJWW _pOO
Figure 6-9: Total Reactor Coolant System Flow Rate - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 400000
-0 RCS I 395000 ------------------
390000 -- - -----
-------------------------------------------------- --------
--- -- --------------------------------
385000 ------------ ---------------------
-------------
380000 -------------
- --- ------------------- ---
---
C,)
375000( Eý ---- --- ------------
---------------------------
U 370000 ------ ------------------
- ----------
0 365000 ------------------
360000 355000 00 350000 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s)
UATL UJWW _p009
Figure 6-10: Reactor Coolant Pump Flow Rate - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 11000 10800 10600 10400 10200 10000 U
9800 C4) 9600 9400 9200 00 9000 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s)
UATL UJWW -pOlo
Figure 6-11: Pressurizer Surge Line Flow Rate - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 240 ________ -.- 0 Surge Line=
240 S120---
S 80.. . . . . ..
C>
2
.. -
-- . . -. - . -
. ---
. . . -. - . -. --
. . -. --
. .-
. -. --
. . -. - .--- - - - -- -- - - -- -- - - --.- - - - - --.- - - - -- - - - - - -- -i-i- - - - -
410 -- - -- - ---- - - - -- .... .... ....--...
--. ... ..... ....... .. . . . . . . . :. . . . .---
. ---
. .. .
20 0 500---------- 1000---------------
1500--- 2000------ 2- 3- 3- 4 -0 4
-.............
. ...... ... T ime.(s)
NO
-160 ______
- 120 ---- -- - - - - -- - ---- - - -- -- -- -- - - ----- - - - -- - - -- - -- ---- - ---- -- - - -- --
0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 '
Time (s)
UATL UJWW _pO I
Figure 6-12: Steam Generator Secondary Side Collapsed Level - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 15
-- SG A
-SG B 14 ------------- ------- ------------- ------------
13 -- -------------------------------------- ------------
12 --------------------------------------- - ------
b-I1I ------------- ------------- --------- -----------
10 ------------ ----------- ------------ ------------7 ------------------------- ------- -------
9 --------------------------------------- - -- - --- ---------- --- --- ---- ------
8 ------------- ------------- -------------------------- ------------------------- -------------
rU-7 ------------- ------------- I-------------- ------------ - ------------------------ ------------
6 --------------------------- --------------------------- -------------------------
U 5 ------------- -------------- I------ ------------
I
4 ------------- ------------- ------- ----
-------------
3 ------------- ------------- ----------- ----- - - ------ - --- -------------
2 - ---------- ------------------ -- - ------ ----------------- -- - ----------
00
---------------------------------------- -- -----------
0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s)
UATL UJWW -p01 2
Figure 6-13: Core Average Fuel Temperature - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 1600 )-0 Vol-Wtd 15O00&'
1400 ---- ---- ---- ------- ---- ---------- ---
1200()G-----------------------------
13 0 00 --- -------- ------------------.; .-
.-.--- --------
- .-.- ------- ------- -.---
------------- .-- -.--
---
.-- --- --------- -
-------- --- -
.-- -------------- ---------
.-- ------------
12 0 0 -------------. ..--
-- ..--
. I.--
--..-.--
. . -----.----.
.-------------. .I..--
. -------------. -------. - ..--. ----.
I.-- .-- . . . . . .. . . . . .. .. . . . . .
1000 - -- --- - - -- - - - - - - -- - - - - - -- - - - - - - - - - - - - -- - - - - - --- - - - - - - - -- - - - -
8000 -----------.-----.------.-.----... I.------------
.........
.. ..-- I.------.-----.---------.-.-.----.-.--------.---------------.-.----- .---.------------
.-------.
4.00 ---- - - - -- - - - - - - - - - - - - -- - - - -- -
400-- ------ ------------- ---------
30000 10001
- 0. 500 1000. 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 700001.
Time (s)
UATL UJWW _p013
Figure 6-14: Reactivity - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 1.0 1-0 Reactivity]
0.5 0.0 I -* *
-0.5 .......... ------------- - - - - -- - - - - - - - - - - - - - - - -- - - - - - - -
-1.0 -------------
-
-1.5 ...... ----
-2.0 -------------
-2.5 ---------- ------
-------------
00
-3.0 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s)
UATL UJWW -p01 4
Figure 6-15: Steam Generator Ruptured Tube Flow - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 40
--0 Tube
-- Tube Sheet
-- -Tota-35 --------------
:-- - - -- - - - - -- -- --- ----------- .........
30 1-------------
------
25 H~g 0
C/)
-o 20 -----------------
-- - -- - - - - -- ---- ------
. .. .. .. .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . .
I-)
15 I -----------------------------
----------------- ------------------------------------
I-------- ----------- ----------------
10 -------------
--------------
L------------- -------------
---------------------------------- -------------
5 I--------------
------------- --- -------- -- ---------- ---------------------I------ ---------- - -
Go ON U
500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 ON Time (s)
L11 0
0 UATL UJWW _p015
Figure 6-16: Makeup and Letdown Flow - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 40 -- 0 Makeup Flow
-- Letdown Flow 35 200......-.-.---7.-.-..-.-.-.-.-.-.-.-.--.....-.-.-.-.-.-.-........-.-.-.-.-.-... . ........-.-.-.-..-.-.-.-.-.-..-.-..................-
- 5. --- --- --- -------- - --- -I-- -- - - - - -- - -- - - --- - -- - --- - - - --
-10 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s)
UATL UJWW _p01 6
Figure 6-17: High Pressure Injection Flow - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 100 75 . .. ..
50 00 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s)
UATL UJWW _pO17
Figure 6-18: Pressurizer Heat - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 2.0
-Op 1.9 ------------- ------ ---
------------- ------------------------ ---------------------------
------------- ------------- .......... ----------- - ------------------------------------------ ------------- ------------
1.8
,1.7 ---------------------------- .......... -- -- - ---------------- ------------- --------------------------- --------
------------ -------------- ---------- ---------------------------------- .......... -----------------------------------
1.6 1.5 I-------------- ---------- ------------------ ------- ------------- ------------I-------------
--------- -- ------ ------------------- I-------------- ------------
1.4 1.3 ------------- ----------------------------------------------------- ------------- --------
1.2 --------------- ------------------ --------- ------------------- ----------------------------------------
---------- ---------
1.1 ---------------
I-------------- ---------------- ----------........... -------- ----------------------------
1.0 0.9 --------------- --- -- - ------------ ------------- I-------------- -------------I--------------
N
--------------------- ----------------
0.8 0.7 ------------------------------ ------------------- ------- --- - ----------
0.6 ------------------------------------------------------ ----------------------------
------------- ........... ---------- ----- ------------
0.5 0.4 - ----- - ---------- ............. I ý---------------- ---- ---------------
- ----------------- ----- -------............ --- -- ----- --- --------
0.3 0.2 ----------------------- -------------- r --------------------------I-- ------------
00 0.1 --------------------------- --------------- --------------------------I-------------
0.0 Eý am a 6 E? Eý 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s) 00 UATL UJWW _pO18
Figure 6-19: Turbine Bypass Flow - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 300
-- Total 250 I--------------
------------- I-------
200 -------------
......
....... ----.......... .......................----.....
Un 150 ------------- ------ ---- - ----- ------
-- - - - - - -- - - - - --- - - - -- - - - -
0ý Cal 100 --------------------------- --------------
50 ---------------
-------------
............. -------------
00
-50 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s)
-p 0 19 UATL UJWW
Figure 6-20: Pressurizer Spray - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 25
-- 0 Spray Flow 20 ------------
U
. 15 C410 N
5 00 IC 0 - -~ -
- 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s)
UATL UJWW -p020
Figure 6-21: Main Steam Safety Valve Flow - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 4500000
Total 4000000 ------------ --- -------
- - - - - ------ ------I ------ ----- --. . -. .- .-. * ------------..
..
3500000 - ------------
3000000 ------------
2500000 1 2000000 1 -------------
Cd2
. . . . . . . . . . . . . ... . . .. . . . .. . .. . . . . . . .. . . . . ... .. . . . . . . . . . . .- ..--
----
1500000 1 .......... -------------
1000000 500000 ---------- -------------
00 0!A 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s) 0 UATL UJWW -p 0 2 1
Figure 6-22: Subcooling Margin - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Case Available SGTR,Margin Offsite Base Available
- Power 140 Subcooling of Thermal Figure 6-22: Analysis Hydraulic LER9Margin ]
120 ----------
. .- . . . . . . . . . . . . . . - . --. . -. -. .- .--
. . -. - . -. .-
- - I 100 ------ ---- - ------------- ------------------------------------------------------
---------- --
-- - -
-- - -- - - -- - -- - -- - -- - -- - -- - -
- o 80 ------------------------------------ ------------- ------- ----------------- ------------- --- ----------
--------------
. . .. . . . . . . . . . --
. . - . -. . - . -. . - - - - - - - - ---------------- ...
..
© 60 ------------------------ ......... ------------ -------------- ------------- -------------
40 ----- ------- ------------- ------------------- I------------- ------
20 ---------------------------
------------------------------------------------------
00 0
0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s)
UATL UJWW -p022
Figure 6-23: Steam Generator Tube-to-Shell Temperature Difference - Base Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 50
-o SG A
-x SG A, Wetted Region
-- SG B
------- . . .SG B, Wetted Region 40 ----------- --------
30
. ....
- - - - - -- - - - - - ---- - - - - - -- - - - - --- -- - - - - -- - - - - ---- -----------------------
20 -------------------------- ------------- -------------
HI 10 --------------------------- - -- - ------ ------------------------------
0 ------------- --------------------------- --- - --- - -- --- ---
i
-i -10, -----------. ........ -- - ---- -----------
H.
-20 ------------- ------------- I-------------- ---- ----------------------
-30 ------------- ------------- ------------
-40 ------------------------------------------ --------- -------------
00
-50 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 6000 6500 7000 Time (s)
UATL UJWW -p023
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary 6.2 Final Case 6.2.1 Event Description The double-ended rupture of a steam generator tube is assumed to occur at full power. Primary system pressure decreased until a reactor trip occurs on low reactor coolant pressure. The turbine trips as a result of the reactor trip and the turbine stop valves (TSVs) close rapidly. Closure of the TSVs causes the main steamline pressure to increase and open the turbine bypass valves (TBVs), atmospheric dump valves (ADVs), and the main steam safety valves (MSSVs). Post-trip secondary pressure control was accomplished via the main steam safety valves.
Allowing the pressure to be controlled by the MSSV lift and reseat pressure maximized dose release during this event. The lift setpoints of the valves were nominal minus three percent (Table 6-12) to allow the earlier lift of the MSSVs and to produce the maximum amount of steam release. No credit was taken for relief through the ADVs immediately following reactor trip. After the reactor trip, the RCS pressure continues to decrease until high pressure injection (HPI) is actuated. At 47 minutes [8] after the reactor trip, the operator began a controlled reduction of pressure using the pressurizer (PZR) spray and will continue so long as adequate core exit subcooling margin is maintained. The subcooling margin limit is defined in Table 6-15. Immediately after ESAS actuation
[8], the operator begins to throttle HPI to control pressurizer level and subcooling margin. At 45 minutes after reactor trip the operator begins an approximately 100 °F/hr cooldown of the RCS using the affected and unaffected SG turbine bypass valves (TBVs). The affected SG is isolated and the analysis was terminated after the RCS temperature is reduced to 500 'F and 1000 psig [9].
6.2.2 Input The inputs and boundary conditions necessary for the analysis of the CR-3 SGTR event are included in Table 6-9 through Table 6-15.
Page 44
A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 6-9: SGTR Inputs and Boundary Conditions for CR Final Case Parameter Value RCS Conditions Core Power 1.004*3014 MWt = 3026.1 MWt Pump Power 4.03 MWt per pump Decay Heat 1.O*ANS 1971 w/ B&W Heavy Actinides Primary Side Tave 582 OF RCS Pressure @ Hot Leg Tap 2170 psia Total RCS Mass Flow Rate 374,880 gpm Make-Up Flow Table 6-11 Pressurizer Parameters Indicated Pressurizer Level 220 inches Heaters / Sprays modeled (See Reference [1] Section Pressurizer Sprays and Heaters 6.1.5 for heater capacities)
PORV not modeled Main Feedwater System Parameters MFW Temperature 460 OF Initial MFW Flow per SG
- 1800 Ibm/sec per SG MFW pump run back to 3 % flow over 0 seconds on MFW Flow Transient Response reactor trip and control to SG low-level setpoint of 32.5 inches Steam Generator Parameters Steam Generator Tube Plugging I 5%
Steam Generator Inventoryj 65%-85% OR Turbine and Main Steam Parameters TBV Actuation Signal Reactor trip and 1010 psig TBV Stroke Time 5 seconds TBV Capacity 837,000 lbm/hr/valve (1 TBV modeled per SG)
MSSV Setpoints and Capacity Table 6-12 Turbine Trip Reactor Trip + 0.50 sec delay Turbine Stop Valve Closure Turbine Trip + 0.50 sec stroke MSIV Isolation (EFIC) Setpoint not modeled MSIV Stroke Time not modeled Atmospheric Dump Valve Capacity not modeled Page 45
A ARE VA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary ADV Opening Setpoint not modeled ADV Stroke Time not modeled RPS Parameters Low Pressure Reactor Trip 1893.95psia Reactor Trip Delay 0.61seconds ESAS Low Pressure Trip 1714 psia ESAS Delay Time 0 seconds ECCS Parameters BWST Temperature 100 OF HPI Flow Table 6-13 Reactivity Control Parameters Control Rod Drop Time 1.4 seconds to 2/3 insertion Control Rod Insertion vs. Time Table 6-14 Delayed Neutron Fraction 0.007 (REA)
Prompt Neutron Generation Time 24.8 jtsec Doppler Coefficient -2.0x10 5 Ak/k/°F Moderator Temperature Coefficient 0.0x10-4 Ak/k/`F
- The MFW flow rate can be adjusted during initialization to achieve a steady-state.
Table 6-10: Pressurizer Heater Setpoints for Banks D and E [11]- Final Case Bank "On" Pressure "Off" Pressure (psia) (psia)
D 2134.7 2154.7 E 2119.7 2139.7 Page 46
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 6-11: Makeup Line Flow versus Pressure for One Pump [12] -Final Case RCS Pressure MU Line Flow RCS Pressure MU Line Flow (psia) (gpm) (psia) (Ibm/s)l 900 298 900 41.08 1100 284 1100 39.18 1300 270 1300 37.27 1500 255 1500 35.22 1700 238 1700 32.89 1900 220 1900 30.42 2100 200 2100 27.67 2300 177 2300 24.50 2500 150 2500 20.78 2700 117 2700 16.22
- 1. The MU line flow was converted to Ibm/s using a density for the conditions of the RCS pressure and a reasonable temperature of 120 'F. Slight variations in the pressure and temperature will have a negligible impact on the density of the makeup flow.
Table 6-12: MSSV Setpoints - Final Case Bank Number of Valves Nominal Setpoint Lift Pressure' Reseat Pressure 2 (psig) (psig) (psig) 1 2 1050 1018.5 967.5 2 2 1070 1037.9 986.0 3 2 1090 1057.3 1004.4 4 23 1100 1067 1013.6 Notes:
- 1. Lift pressure represents 3% lift tolerance from nominal setpoint.
- 2. Reseat pressure represents 5% blowdown from lift pressure.
- 3. Per Reference [14] Section 10.3.4, 1 MSSV with a capacity of 583,574 Ibm/hr has a nominal setpoint of 1100 psig.
Page 47
A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 6-13: HPI Flow [13] - Final Case RCS Pressure (psia) Flow (gpm) 15 904.3 615 815.9 1215 712.1 2515 373.3 Table 6-14: Control Rod Insertion [10] - Final Case Time (sec) Percent Reactivity Inserted (%)
0.0 0.00 0.2 0.58 0.3 0.99 0.4 1.83 0.6 5.29 0.8 12.33 1.0 21.41 1.2 33.09 1.4 50.75 1.6 72.96 1.8 91.30 2.0 99.26 2.2 99.99 2.3 100.00 2.4 100.00 Page 48
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 6-15: ISCM Curve - Final Case Temperature, F Pressure, psig 197.0 67 276.9 100 339.4 200 377.7 300 407.2 400 431.6 500 479.4 750 516.0 1000 551.3 1300 580.4 1600 589.1 1700 597.4 1800 605.4 1900 613.0 2000 620.3 2100 627.3 2200 634.0 2300 640.5 2400 Page 49
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary 6.2.3 Results The results of the SGTR analysis final case are shown in Figure 6-24 through Figure 6-46. Sequences of events are presented in Table 6-16. A summary of results are provided in Table 6-17.
The initiating event was a double-ended rupture of a steam generator tube just below the lower face of the upper tube sheet. The primary system pressure decreased due to the loss of inventory (Figure 6-27). At approximately 17 minutes the RCS pressure begins to drop quickly (Figure 6-27) due to the PZR heaters tripping on low PZR level (Figure 6-41) and at 23 minutes the PZR empties out (Figure 6-28). Reactor trip occurred at 33.656 minutes on low reactor coolant pressure (Figure 6-27). The turbine tripped as a result of the reactor trip and the turbine stop valves closed within one second. Closure of the TSVs caused the main steam line pressure to increase and open the turbine bypass valves and the main steam safety valves. Within the next two minutes, the MSSVs reseated and the TBVs began automatically controlling secondary pressure near 1024.7 psia (Figure 6-29). SG levels were maintained at a level of 3.3698 feet above the upper face of the lower tube sheet (UFLTS) with main feedwater (Figure 6-35).
Following reactor trip, the low RCS pressure ESAS signal was reached within 20 seconds and HPI from two pumps was initiated (Figure 6-40). The HPI flow more than offset the tube leak and minimum pressurizer liquid level and RCS pressure were quickly regained (Figure 6-27 and Figure 6-28).
Forty-five minutes after the reactor trip, the TBVs on both steam generators were taken under manual control by the operator to reduce steam pressure (Figure 6-29) and begin the cooldown of the RCS (Figure 6-26). Forty-seven minutes after the reactor trip (i.e. 80.7 minutes after event initiation), the operator began a controlled reduction of pressure using the PZR spray which is operated based on core exit subcooling margin (SCM). The subcooling margin limit is defined in Table 6-15. The core exit subcooling margin never reached 110 'F (Figure 6-45). Therefore, no operator action was needed to control the subcooling margin between 110 'F and 90 'F (Figure 6-43). As the primary system pressure decreased, the mass flow rate through the ruptured tube also decreased. At approximately 114 minutes, the core exit temperature reached 500 'F. At approximately 117 minutes, the RCS pressure reached 1000 psig and the core exit temperature was less than 500 'F. At this time the affected SG was considered isolated and the analysis was terminated. The MSSV lift duration is not expected to exceed 100 seconds during the remainder of the cooldown. The total SG ruptured tube flow is not expected to exceed 40 lbm/sec during the remainder of the cooldown.
The SG tube-to-shell temperature differences (Figure 6-46) from this analysis are approximately -10/+30 'F, which is well within the acceptable limit of -90/+140 'F (Section 5.0).
Larger tensile temperature differences would occur as the operator continues to cool down the RCS, but these values would be bounded by existing normal cooldown calculations. Consequently, additional loss of reactor coolant boundary integrity would not occur due to resultant temperature gradients during the SGTR transient, thus the acceptance criterion for the event is met.
No fuel failures are expected to occur during the SGTR event.
Page 50
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 6-16: SGTR Sequence of Events - Final Case EVENT Time, seconds (minutes)
Reactor Trip on low RCS pressure 2019.3 (33.66)
Control Rods Insert and Turbine Trip 2020.0 (33.67)
SG A: 2021.1 (33.69)
'MSSVs first open SG B: 2021.2 (33.69)
ESAS low RCS pressure Signal 2032.6 (33.88)
HPI flow from two pumps started 2032.6 (33.88) 2 SG A: 4733.5 (78.89)
MSSVs last close SG B: 2100.4 (35.01)
Operator begins RCS cooldown using TBVs on both 4719.4 (78.66)
SGs.
Thor reaches 500 'F or less and RCS pressure reaches 7010.0 (116.83) 1000 psig or less. Analysis terminated.
Notes:
Page 51
A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 6-17: SGTR Summary of Results - Final Case Parameter Results Integrated tube leak 220,184 Ibm Integrated tube leak 36,112.1 gal Integrated total MSSV release 67,022.3 lbm Integrated SG A MSSV release 36,038.4 ibm Integrated SG B MSSV release 30,983.8 lbm Integrated total TBV release 431,715 Ibm Integrated SG A TBV release 270,081 lbm Integrated SG B TBV release 161,634 lbm MSSV lift duration (SG A) 92.6 seconds (SG B) 72.4 seconds Minimum subcooling margin 17.63 'F @ 2037 seconds Maximum tube-to-shell temperature difference SG-A: 26.0 'F @ 701 seconds (shell temp - tube temp) SG-B: 29.7 'F @ 7010 seconds Calculated RCS cooldown rate 92.129 'F/hr (TEMPF 4719.35 to 6746.1 seconds)
SGTR event duration 7010.0 seconds Page 52
Figure 6-24: Reactor Power - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 3200
--0 Reactor e E) E) 2800 --------------- ------- ---------
2400 ---------------
- ------------- --------------- -------------
2000 ---------------
-------------
1600 01 1200 -------------
800 -------------- --------------- ----------------------------- ---------- --------------- -----------
400 --------------- ----- -- ------ -------------- --------------- --------------- ---------- --------------- -----------
00 0
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
TPUV TIET -pool
Figure 6-25: Steam Generator Power - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 1600 o i -- SG A
-SG B 1400
-------
1200 ------------
1000 ------------ --------------------------
-------
- 0. 800 ------------
)
------ -------
600 ------------
- ---------
-------------
---------------
400 ------------ ........... -------------
------------- --------------- ------
----------
-------------------------
200 1 ------------ ---------------
00 0
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
TPUV TIET -pO02
Figure 6-26: Reactor Coolant System Temperature - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 650
--e Tave
- TH-1
-L TH-2 625 ------------------------------- ---------------
-E- TC-1A
-- TC-1B
-- TC-2A TC-2B 600 --------------- ------------ ---------------
575 --------------- --------------- ---------------
I H~
550 C14 525 500 475 450 425 1 00 400 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
TPUV TIET -p003
Figure 6-27: Primary Pressure - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 2400
---0--
xSGSG B, A, Hot Leg P-T Tap 2200 --------------- I ---------------
2000 -----------------------------------
-- - - - - -.-
- --------..-.- .. ---.----.. . . . . . . . .-------..
.. . -- - - - - - -- - - -- -- -
1800 --------------- -------
C)4 a4~
1600 ----------------------------------------------
1400 V-
.....--------- ---.-
...-. ----------.. .........-. ------..
1200 00 1000 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
TPUV TIET -pO04
Figure 6-28: Indicated Pressurizer Water Level - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 360
[--G Pressurizer 300 rJ) 0 240 a) 0
.. .. .. . .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .-.. . .. . . . . . . ..
180 a)
I-a) 120 N
60 0
1200 1800 2400 3000 4800 5400 6000 6600 7200 Time (s) 0 tQ 5
TPUV TIET -p00
Figure 6-29: Steam Generator Outlet Pressure - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 1200
-:SGBMSLB-2 1000 c ~900 800 0
0 700 600 500 4000 0 600 1200 1800 2400 3000 Time i0(s) 3600 4200 4800 5400 6000 6600 7200 Y con TPUV TIET _p006
Figure 6-30: Steam Generator Flow Rate - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 2000
-A SG B MS e
E) i --------------- -
SGBMFW S
1800 -------------- -------
- --- -- -----------
tt = = z 1600 - --------------- - ------------..----------------
..---------------
1400 1200 0'
1000 0
800 --------------- --------------- -----------
600 ----------- ------ ------- -----------
400 ............. --------- -- ---------
200 --------------- -----------
0 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s) CD to-TPUV TIET _p00 7
Figure 6-31: Steam Generator Inventory - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 80000
-OSGA
- SGB 7 20 00 . . . .. . . . . . . . . .. . . . . .. . . . . . . . . .. . . . . .. . . . . .. .. .. . . . . . . . .. . . . . . . .. . . . . .. .. . . . .. . .. . . . . . .. . . .. .. . . . . . .. . . . . .. . . . . .. . . . . . .. . . . . . . . . .. . . . . . . .. . . . . . .. . . . . . . .. . . . .. . .. . . . . . .. . . . . . .. . . . . . .
640000 16000 .. ..
48 0 00 .. . . . . . . . . . . .. . . . . . . . . . . . . .. .
q)0 0 0 -- - -- -- -I---- - -----.- -- --- -- --
3 2 0- ---.. ....- -- -- -- --.-- -- - -- --.---- --- --.--- - -- -- -- --- ---- - -- --- ---- -- --- ---- - ---.- -- -- -- -
40060020 1800..2400.3000.3600.4200.4800.5400.6000.6600.7200 8000 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
ON 8
TPUV TIET -p00
Figure 6-32: Total Reactor Coolant System Flow Rate - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 400000
-- RCS 395000 --------- ----- ------------------------------ --------------- ---------- --------------- ...........
390000 -------- -- -- ---------------............ .. .................. ------------------- ---- ------ --
385000 -- - - ----------
-------------- ----------
..........
E 380000 --------------- --------------- ---------------------------- --------------- -----------
375000 ------------ --------------- -----------
U 370000 --------------- ------ ---------------
Ht 365000 .......
---------
-- ---- --------- - ------ ..............
--- ---------- --------------- ---------------
360000 -------------- I ---------------
---------------
--------------- ---------------
--- -- -----------------------
---------
355000 --------- ---- -------------- ---------------
I ------------------------
--------------------
100 350000 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
TPUV TIET _pO09
Figure 6-33: Reactor Coolant Pump Flow Rate - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 11000 pR -
--- RCP-1B
-- RCP-2A 10800 ----- -
- RCP-2B 10400
- 10200 9080 0 -- - - - - ------ ----- -
----- ---- - __ _ _ _ __---------------_ __ - -- - - - - - - - - - - - -
9400
.4.0.0 .~. ~. . . . ..~ . . . ~
9~ . . .~.. . . .~. ~. . ~. - -. - - ---- - . ... .... .... ... .... .. ----
. . .. -.. .. ...----- ..............
...
9000
- 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
TPUV TIET -polo
Figure 6-34: Pressurizer Surge Line Flow Rate - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 240 I-e 200 -------------
..........
160 ----------------------------
120 ------------
...........
C/3 80 ------------
...........
40 ------------ I--------------
N
-40
-80
-120 00 I
-160 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s) I 0
TPUV TIET _p011
Figure 6-35: Steam Generator Secondary Side Collapsed Level - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 15
-- SG A
-- SG B 14 [
............................
13 I
12 1 ---- ----- -- ----------I
--..--------------..----------
11 ------------------- ------------
---------------
10, --------------- ---------------
-----------
-o 9 -------------------- ......... ---------------
--------------- --------- ---------------
8 ----------
---- ----------------------
U
-e 7 -----------
- ---------
I------------
------------
rj*)
------------ ------------------------ ---------------
--------------- ----------
6
~0 0 5 .......... - ----------- ----------- ---------------
I --------------- ---------------
4 ---------------
--------------- --------------------------
--------------
--
---------------
3 ----------
--------------- ------------- ---------
-----------
2 -----------
00
---- ---- -
0 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s) rQ TPUV TIET -p012
Figure 6-36: Core Average Fuel Temperature - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 1600
-eVol-Wtd 1500 1400 1300 --------------- ----------
1200 -----------
1100 ---------------
1000 --------------- -------- -------
ct 900 - ----------------
800 ------------- - ----------------
------- --
-- - - -- --- -- -
700 ---------------
--
600 500 ------------------
400 ------------ ------------------
300 -----------
00 200 100 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
ON TPUV TIET _pO13
Figure 6-37: Reactivity - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 1.0 I-- Reactivity 0.5
-0.5 02
-1.0
--- ---
-1.5 ....
. ...
--... ........
-----------
-2.0 ----------------
---------- ------------
-2.5 -----------
00 CN IN
-3.0 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
TPUV TIET .pOl14
Figure 6-38: Steam Generator Ruptured Tube Flow - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available
--0 Tube
-- Tube Sheet Total 35 ----------- -- --------------- -- -. -
. -- - ---------------- . .. --
. -.-- - -. . . . . .. . . . . --. -
-- ---------- . ..-. -
. -. - ..-
.. -- - - ----- . -- - - - - -
. . . . . --
. . . ------- --------------
U M¢325 -- .--.-------..--..---------.---.---.-- -- * -----
--- - ------------ --------------- -- -- ..............---
* ------------
o20 2 0 I------ -----
I- ---------- "
00 0A 0 600 1200 1800 2400, 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
TPUV TIET -p015
Figure 6-39: Makeup and Letdown Flow - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 40
-o Makeup Flow
-,Letdown Flow 35 --------------- -------------- ------------------
30 1--------------- - ------- -------------------
25 -- ------- -- -- ------ ----------------
. .-. .-. .-.- . ..-. .-.--
---------------------
U 20 1--------------- -
.----
. . ---
. . , ---..
_
15 --------------- ------- ---------------
0
-- --------------------------- -- - ------
101 5 -----------------
-------- - -------------
-5 ---------- --- -----------
00 ON IN
-10 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
ON 00 TPUV TIET -p01 6
Figure 6-40: High Pressure Injection Flow - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 120 S--O HPI 10 0 ------.- -.-- --------.----.-.-- ---. ----- -.-.-- -----.--
-- -----.-.-.-- .-------- ------ ------
LL 20 .. . . . .. . . . .. . . . .. . . . .. . . . .. . . . . . . . . .. . . . -- ----.
ON 40d rT,)
20 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s) 17 TPUV TIET _pO
Figure 6-41: Pressurizer Heat - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 2.0 S--0 1.9 --------------
1.8 --------------
1.7 --- -------------
1.6 1.5 -------------
1.4 -------------
1.3 1.2 -------------
---
1.1 -------------
1.0 -------------
ct ----------
0.9 1 ----------
0.8 .... ---------- -------------
0.7 ----------
- -------------
-
0.6 1 --------- - -------------
0.5 1 ----------- -------------
0.4 0.3 00 0.2 0.1 ----------
-
ON 0.0 L.
0 600 1200 1800 2400 3000 3600 4200 4800 7200 Time (s)
'-.A TPUV TIET _pO 18
Figure 6-42: Turbine Bypass Flow - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 300 11
Total 250 - ------------------------------ --------------- ------------
.....
........................... ................
200 ------------------------- ------------- -----------------
U 0/
--- --------- ----------- - ------------ ------------
150
................
100 - ------------ ------- --------------- ----- ---- ------- ........ ............................
50 % --------.... -------- ------ ----------
------------ ------------ -
--- ------------
00 LA
-50 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
TPUV TIET _p0 19
Figure 6-43: Pressurizer Spray - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 25
-e Spray Flow 20 ----
- -----
U 15 C/2 Cn 10 5
00 00 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
--..
TPUV TIET -p0 2 O
Figure 6-44: Main Steam Safety Valve Flow - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 4500000
-A Total 4000000 3500000 3000000 2500000 2000000 C/)
1500000 1000000 500000 00 A
- P~ ON 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s) 21 TPUV TIET -p0
Figure 6-45: Subcooling Margin - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 140 l
-eMarginý ISCM Sat Curve Margin 12 0 . .. . . . .. .. . . . . -- - - -- - ----.- -----------..
-- - - - - ----.- -----.-----.-
.- . .-.-
L.-
.-. .------------.. -----------..
--.- ---.- .. . . . . . . . . . .. . . . . . . .. . . . . . . .. . .
.---.-.-.-.-.---..
0 60 .... .
0 0¢3 20 . . .. . . .. . . . .. .i - -*. . .. . . . .. . . . .. . . . ----------..
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
A,.
TPUV TIET -p022
Figure 6-46: Steam Generator Tube-to-Shell Temperature Difference - Final Case Thermal Hydraulic Analysis of SGTR, Offsite Power Available 50
-OSGA
---xSG A, Wetted Region
-- ASGB
-- SG B, Wetted Region 40 30 20
-- -- - -- -- -- -
----------
,0 10 ---
---
---
--- ---- --- --- --- ---
0
-10
-.o H..
-20
-30
-40 00
-50 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 Time (s)
"-.4
(.*
TPUV TIET -p023
A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary 7.0 CONTROL ROOM HABITABILITY/ LOOP EVENT The purpose of the steam generator tube rupture control room habitability transient analysis is to calculate the total steaming and leak rate to the atmosphere until the ruptured steam generator is isolated. As stated in Section 1.0, an initial base analysis was performed followed by a final analysis.
7.1 Base Case 7.1.1 Event Description In the CR1- event, the double-ended rupture of a steam generator tube is assumed to occur at full power. Primary system pressure decreases until a reactor trip occurs on low reactor coolant pressure. The turbine trips as a result of the reactor trip. Loss of offsite power (LOOP) is assumed to occur coincident with reactor trip. Consequently, reactor coolant pumps and main feedwater pumps trip. Emergency feedwater actuates automatically to raise steam generator liquid levels to fifty percent on the EFIC high range (natural circulation setpoint). The operator performs normal post-trip verifications and then at 30 minutes after reactor trip begins a cooldown of the reactor coolant system at a cooldown limit of 240°F/hr [9] using the atmospheric dump valves ensuring that the bad ROTSG level is maintained less than overfill. The operator finds that the ADV on the good ROTSG is not available and the cooldown commences with only the bad ROTSG initially. The PORV is not used for the CRH analysis. The RCS is depressurized using PZR tHPV's at 25 minutes after reactor trip as a conservative measure since it takes longer to initiate the depressurization and occurs at a lower rate than the PORV. After an additional 25 minutes following cooldown initiation (55 minutes after reactor trip), the ADV on the unaffected OTSG is opened as well. The reactor coolant system is limited to a cooldown rate of 100°F/hr at temperatures below 500'F and above 280'F [15]. The reactor coolant system cooldown continues, by steaming the unaffected steam generator. Reactor coolant system pressure is maintained near the minimum subcooling margin by throttling high pressure safety injection (which started automatically on low reactor coolant pressure ESAS or was started manually by the operator). The HPI throttling is performed as necessary to keep the pressurizer level in an acceptable range with acceptable RCS subcooling margin beginning immediately after ESAS actuation [8]. The transient ends when the RCS temperature is < 500'F, and the RCS pressure is < 1000 psig. At this time the ruptured steam generator is considered isolated, since the ADV, MSIV's, blowdown, and FW/EFW feeds on the affected steam generator will be isolated.
Note that the operator action times in the CRH base case are lower than the times historically assumed in the SGTR event. The final CRH case described in Section 7.2 uses later operator action times.
7.1.2 Input The inputs and boundary conditions necessary for the analysis of the CR-3 SGTR event are included in Table 7-1 through Table 7-7.
Page 76
A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 7-1: CRH Inputs and Boundary Conditions for CR Base Case Parameter Value RCS Conditions Core Power 1.004*3014 MWt = 3026.1 MWt Pump Power 4.03 MWt per pump Decay Heat 1.O*ANS 1971 w/ B&W Heavy Actinides Primary Side Tave 582 OF RCS Pressure @ Hot Leg Tap 2170 psia Total RCS Mass Flow Rate 374,880 gpm Make-Up Flow Table 7-3 Pressurizer Parameters Indicated Pressurizer Level 220 inches Pressurizer Sprays and Heaters Sprays not modeled / Heaters (See Reference [1]
Section 6.1.5 for heater Capacities)
HPV Flow Diameter 0.375 inches [16]
PORV not modeled Main Feedwater System Parameters MFW Temperature 460 OF Initial MFW Flow per SG
Initiation LOOP/ Loss of 4 pumps Delay 40 s Flow 660 gpm Temperature 140°F EFIC Fill Rate 6.29 in/min Steam Generator Parameters Steam Generator Tube Plugging 5%
Steam Generator Inventory 65%-85% OR Turbine and Main Steam Parameters TBV Actuation Signal not modeled TBV Stroke Time not modeled Page 77
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary TBV Capacity not modeled MSSV Setpoints and Capacity Table 7-4 Turbine Trip Reactor Trip + 0.50 sec delay Turbine Stop Valve Closure Turbine Trip + 0.50 sec stroke MSIV Isolation (EFIC) Setpoint 649 psig MSIV Stroke Time 0 seconds Atmospheric Dump Valve Capacity 620,000 lbm/hr/valve ADV Opening Setpoint 1035 psig ADV Stroke Time **20 seconds RPS Parameters Low Pressure Reactor Trip 1893.95psia Reactor Trip Delay 0.61 seconds ESAS Low Pressure Trip 1714 psia ESAS Delay Time 0 seconds ECCS Parameters BWST Temperature 100 OF HPI Flow Table 7-5 Reactivity Control Parameters Control Rod Drop Time 1.4 seconds to 2/3 insertion Control Rod Insertion vs. Time Table 7-6 Delayed Neutron Fraction 0.007 (REA)
Prompt Neutron Generation Time 24.8 [tsec Doppler Coefficient -2.0x 10-5 Ak/k/°F Moderator Temperature Coefficient 0.0x 10- 4 Ak/k/°F Operator Actions Key Operator Action Times Table 7-7 The MFW flow rate can be adjusted during initialization to achieve a steady-state.
- The expected stroke time is 25 to 30 seconds open. The event is insensitive to changes in ADV stroke time.
Table 7-2: Pressurizer Heater Parameters - Base Case Bank "On" Power (kW)
Emergency Diesel Backup With LOOP 252 The total heater power is assumed until the LOOP. Following LOOP, the emergency diesel backed heaters totaling 252 kW are assumed to be available. Modeling the emergency diesel backed heaters following LOOP is conservative, since the heaters would not be available unless manually aligned (Reference [1], page 12).
Page 78
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 7-3: Makeup Line Flow versus Pressure for One Pump [12] - Base Case RCS Pressure MU Line Flow RCS Pressure MU Line Flow (psia) (gpm) (psia) (Ibm/s)1 900 298 900 41.08 1100 284 1100 39.18 1300 270 1300 37.27 1500 255 1500 35.22 1700 238 1700 32.89 1900 220 1900 30.42 2100 200 2100 27.67 2300 177 2300 24.50 2500 150 2500 20.78 2700 117 2700 16.22
- 1. The MU line flow was converted to lbm/s using a density for the conditions of the RCS pressure and a reasonable temperature of 120 'F. Slight variations in the pressure and temperature will have a negligible impact on the density of the makeup flow.
Table 7-4: MSSV Setpoints - Base Case Bank Number of Valves Nominal Setpoint Lift Pressure1 Reseat Pressure 2 (psig) (psig) (psig) 1 2 1050 1018.5 967.5 2 2 1070 1037.9 986.0 3 2 1090 1057.3 1004.4 4 23 1100 1067 1013.6 Notes:
- 1. Lift pressure represents 3% lift tolerance from nominal setpoint.
- 2. Reseat pressure represents 5% blowdown from lift pressure.
- 3. Per Reference [14] Section 10.3.4, 1 MSSV with a capacity of 583,574 ibm/hr has a nominal setpoint of 1100 psig.
Page 79
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 7-5: HPI Flow [13] - Base Case RCS Pressure (psia) Flow (gpm) 15 904.3 615 815.9 1215 712.1 2515 373.3 Table 7-6: Control Rod Insertion [10] - Base Case Time (sec) Percent Reactivity Inserted (%)
0.0 0.00 0.2 0.58 0.3 0.99 0.4 1.83 0.6 5.29 0.8 12.33 1.0 21.41 1.2 33.09 1.4 50.75 1.6 72.96 1.8 91.30 2.0 99.26 2.2 99.99 2.3 100.00 2.4 100.00 Page 80
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 7-7: Summary of Key Operator Actions - Base Case Action Time After Reactor Trip PZR HPV open. 25 min Begin RCS cooldown (ADV on affected SG open). 30 min ADV on unaffected SG open. 55 min Page 81
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary 7.1.3 Results The results of the SGTR CRH analysis are shown in Figure 7-1 through Figure 7-31. The sequence of events is presented in Table 7-8. A summary of results are provided in Table 7-9.
The initiating event was a double-ended rupture of a steam generator tube just below the lower face of the upper tube sheet. The primary system pressure decreased due to the loss of inventory (Figure 7-4). Reactor trip occurred at 33.92 minutes on low reactor coolant pressure (Figure 7-4). The turbine tripped as a result of the reactor trip and the turbine stop valves closed within one second (Figure 7-6). At the time of reactor trip a loss of offsite power (LOOP) occurs. At the time of LOOP the reactor coolant pumps trip off (Figure 7-10), a loss of main feedwater occurs (Figure 7-7), loss of the condenser occurs (Figure 7-27), and EFW actuation occurs 40 seconds later (Figure 7-19). Closure of the TSVs caused the main steam line pressure to increase and open the main steam safety valves (Figures Figure 7-21 and Figure 7-30).
After LOOP, the emergency feedwater actuated and controlled the SG levels to fifty percent on the EFIC high range (Figure 7-12 and Figure 7-31). Following reactor trip, the low RCS pressure ESAS signal was reached and the makeup and letdown were isolated (Figure 7-16). Additionally, the HPI from two pumps was initiated (Figure 7-17). The HPI flow more than offset the tube leak and minimum pressurizer liquid level and RCS pressure were quickly regained (Figure 7-4 and Figure 7-5).
Twenty-five minutes after the reactor trip, control room operators begin reactor coolant system depressurization by opening the PZR HPV. Thirty minutes after reactor trip, the control room operators begin cooldown by opening the ADV on each SG. At this point the operators observed that the ADV on the unaffected OTSG will not open. An equipment operator is dispatched to open the ADV on the unaffected SG. The operator commences the cooldown of the reactor coolant system using the ADV on the ruptured SG (Figure 7-3, Figure 7-6, and Figure 7-28). The subcooling margin (SCM) did not reach 110 F (Figure 7-23). The PORV opening to control subcooling margin was not modeled (Figure 7-20). Shortly after beginning cooldown, the MSSVs on the affected SG close and remain closed (Figure 7-30). The MSSVs on the intact SG close at approximately 72 minutes (Figure 7-30)..
As the primary system pressure decreased, the mass flow rate through the ruptured tube also decreased (Figure 7-15). Figure 7-28 demonstrates that the core exit temperature compared reasonable well to the temperature targeted based on the cooldown rates. At approximately 85 minutes, the core outlet temperatures reached 500 F.
It should be noted that the core exit temperature achieved 500 F before the second ADV on the intact generator could be opened and the RCS pressure reached 1000 psig. The analysis was extended until the ADV on the intact generator was opened and the RCS pressure reached 1000 psig.
The SG tube-to-shell temperature differences can be seen in Figure 7-24. Section 5.0 states the acceptable SG tube-to-shell temperature differences are -90/+ 140 F.
Larger tensile temperature differences would occur as the operator continues to cool down the RCS, but these values would be bounded by existing normal cooldown calculations. Consequently, additional loss of reactor coolant boundary integrity would not occur due to resultant temperature gradients during the SGTR transient, thus the acceptance criterion for the event is met.
No fuel failures are expected to occur during the CRH event.
Page 82
A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 7-8: SGTR CRH Sequence of Events - Base Case EVENT Time, seconds (minutes)
Reactor Trip on low RCS pressure, followed by 2035.3 (33.9) coincident LOOP Control Rods Insert and Turbine Trip 2035.9 (33.9)
'MSSVs first open SG A: 2037.7 (34.0)
SG B: 2037.9 (34.0)
ESAS low RCS pressure Signal 2090.5 (34.8)
HPI flow from two pumps started 2090.6 (34.8)
Operators open the Pressurizer HPV valve. 3535.3 (58.9) 2 SG A: 3818.9 (63.7)
MSSVs last close SG B: 4316.5 (71.9)
Operator begins RCS cooldown using ADVs.
ADV on intact SG fails to open remotely. 3835.3 (63.9)
Cooldown proceeds using ADV on faulted SG.
Tcore outlet reaches < 500 F 5083.4 (84.7)
ADV on intact SG opened manually. 5335.3 (88.9)
RCS pressure < 1000 psig. Analysis terminated. 6392.2 (106.5)
Notes:
Page 83
A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 7-9: SGTR CRH Summary of Results - Base Case Parameter Results Integrated tube leak 223,233 ibm Integrated tube leak 36,368 gal Integrated total MSSV release 129,581 ibm Integrated SG A MSSV release 81,300 Ibm Integrated SG B MSSV release 48,281 Ibm Integrated total ADV release 185,930 lbm Integrated SG A ADV release 154,514 Ibm Integrated SG B ADV release 31,416 Ibm MSSV lift duration SG A: 1781.2 seconds (last close - first open, see Table 7-8) SG B: 2278.6 seconds Minimum subcooling margin 21.05 F @ 2019 seconds Maximum tube-to-shell temperature difference 80.01 F @ 6080 seconds (shell temp - tube temp)
SGTR CRH event duration 6392.2 seconds Page 84
Figure 7-1: Reactor Power - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 3200 [e *Reactor 2 800 -----------. ---.
--..
.. I --.-
. --.-.--. -----.-.- ------------.--- ------. --------- .. . .. .. . . . .. .. . .. ... .. .. . .. .. . .. .. .. . .. .. . .. .. . .. .. .
1200 ----------- - ----------- ----- -- -
4600 .................... .. ............
................ ......................................................................... ...............................................
8000 0
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s) 00 USSA -pOO
Figure 7-2: Steam Generator Power - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 1600
--eSGA
,-- SG B 1400
. . . . . . . . . . . --
. . . -. -
..-
. -. . -
. -
..-. .. . . .. . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . ----------------..
1200
............................... ----..........................................
.... . -,- .......---- ..............- ,.......................... .................
1000 800 I- 600
.0 400 200 0
-200 00
-400 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s) 00 USSA _p 0 0 2
Figure 7-3: Reactor Coolant System Temperature - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 650
-0 Tave
--- TH-1 TH-2 625 ------------------
----
- --------------
--- ------------- TC-1A
--- TC-1 B
-< TC-2A TC-2B 600 ----------------
-------- ----------------------
E) G 0 G !E) 575 ---------------- -------
------------------------ ---------- ......
550 -----------------
---------------- -
525 --------------
r-4 500 -----------------
475 450 425 00 tJ1 400 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s) 00 03 USSA _p 0
Figure 7-4: Primary Pressure - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 2600 _____
-2OSG A, Hot Leg P-T Tap SG B, Hot Leg P-T Tap 1400 2 000 - - - - -- --------- -- -- - -- -- - I--- -- - -- -- - -- - - -- - -- -- - -- - - -- - -- -- - -- -
- 1. 12000 0
2200 600- 2400--
1800---------
1200----------------- 300-60-400400540-00-60
- 1800 1000 00 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 :
(s) '
00Time 0 04 USSA _p
Figure 7-5: Indicated Pressurizer Water Level - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 360
-- e Pressurizer 300 ------- ----------------
cj3 0..
240 ------------------------------- ------- ----------------
0,.
0,,
I 180 ----- ---------- ---------------------- ----------
---------------------------------- ........ ---------------
0l.
0 120 ------------ ---- ------------
- - - - - -- - - -
------------------------- ------- ----------
-- ----------------
N 60 --------------- -------- -------- ------------ ------------- ------- ---------
0 tý CQ 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s)
G0
\,c USSA _p 0 0 5
Figure 7-6: Steam Generator Outlet Pressure - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 1200 SG A MSL A-1
-A SG B MSL B-1 --------------
. * , . .- * .
900 800 700
...................................................................
CIO 600 C/)
500 400 300
200 100 ------------- 00 ON 0 ON 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 0 Time (s) 0 0
USSA _p0 0 6
Figure 7-7: Steam Generator Outlet Pressure - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 2000
-oSG A MS
-- SG B MS 1800 e -----------
!e-,,- -------------
..... .... .... .... .... -----
-- S G B MFW ... ... ... ... ... ... ... ...
- -
¥ 1 /- - - 1,_ Ir )1 !
1600 ---------------
1400 ----------- -
1200 ---------------
................... - -..
.-.-. ------.---....----.......... ----...........
Q 1000 1.... -------------
800 0
(J~ 600 -------------
400 -------------
200 -------------
00 ON 0 ___________ -. 7 ~-.--,.- - - - - ~~ ~- ~
0 600 1200 1800 2400- 3000 3600 4200 4800 5400 6000 6600 ON Time (s)
USSA _p 0 0 7
Figure 7-8: Steam Generator Inventory - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 100000 90000 80000 70000 S
0 60000 0
50000 40000 30000 00 20000 6600 tO 00 8 USSA _p
Figure 7-9: Total Reactor Coolant System Flow Rate - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 400000
--R CS
---------
360000 ----------------
320000 - --------------------------
280000 --------------- :-----------------
- ----------------- -- --------------
E --------------------.-.-------------------- ........
Cl' --------------------- ------------------ ---------------------------
240000 200000 I ----------------
---------
U 160000 --
- ---------- -----------------
120000 ---------------- ------------------------------------ -----------------
. . . ---------.-. . . . ... . . . . . . . . . .
..
80000 ----------------
40000 -
-----------------
00 0
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s)
USSA _p 0 0 9
Figure 7-10: Reactor Coolant Pump Flow Rate - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 10000 G x 7-ýEl iG x
-- RCP-iA
-< RCP-1B
-A RCP-2A 9000 --------------
-----------------
RCP-2B 8000 ------------------
7000 6000 5000 4000 U 3000 rý4 2000 1- ----------------
-----------------
------- ---- ----- ---------
1000 -- ----------------------------- -------
--------------
0 ---------------- -------
-----
-10001 00
-2000 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6*600 Time (s)
USSA -polo
Figure 7-11: Pressurizer Surge Line Flow Rate - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 240 [-e Surge Line 200 160
"* 120 120 ---------------------------------..... .......................................................
----------..........................-------------
.....................
80 N
0 -----------
O4 ----------
I.-- ----------------..... .........-- ----------- ---------.---------------.---.-- -----------.--
00
-120 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s)
USSA -pO11
Figure 7-12: Steam Generator Secondary Side Collapsed Level - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 30
-SG A
-- xSG B
. .. .. ...
. . . . . . .. . . . .. . . . . . . . .. . . . . .. . . . .. . . . . . .
28
.---------
. . . . . . . . . .-- .
26 -----------
.. . . . . . . .. . . . . . . . . . . .
. . . . ---
.---- . . . .---
. . .---
. - ------- -----------
24 22 C) 20 -------------------------------- - -------- -------------------------
---- -- - -----------
18 -- --- ------------------------------- --- ---- ------------------------------ - -- -- -- -----------
------ ----------------- -
16 ----- --- ----------------------- - -- -----------------
U
-- -- -- --- I ------------- -- -- ---- - ----------- ------------ --------- - -- - . . ... . .. . . . . . . . . . . . . . . .. . . . . . . . .
14 ---------------------
. . . . . .. ... . . . . . . . . . . . .. . . . . .. . . . -. . .. . . .
12 -------- ---- - ---- ----------------------------- --- --- -- ---- - --------------------------------- ------
0ý 10 - -----------I ------------- --- ------ - -- ------- - -- --------------------- --- -----
8 6
4 00 2
0 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s) 12 USSA -p 0
Figure 7-13: Core Average Fuel Temperature - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 1600 I--0 Vol-Wtd
. .-.- ------- . --------------. ----------- .-.-.-.--
-----------.--
.-.-- ..
1500 -.-------------.
----------. . . . . . -. --. -.--- .- ------ .- -- .-- ----------- --
.-- ---------
.- .-
-.- .-------.- ----- .- --
I -tut)
1300
................................ .................. ................................
1200
"-
1100 ------------------ ----
--- --- ---------- ------- --- --- --- -- --- --- --- -- - --- --- --- --- --- --- --- --- -- ---
1000 900 ---------------- - - -- - - - - -- -
-
800 -- - - - - - -- - - - - - - - -- --------
--
1U
...- .............. ...
.. . . --- --------- !.. ......
- --------...
700 -- - - - -- -- - - - - - - -- - ----------- -
. ..... . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . --- . . . .- .- . . -- . .- .- . . -. -. -----------------------------
600 - - - - - -- -- - - - - - - -- -
-- ---
- ---
--- --- --- - -- --- --- -- - - --- --- --- --- --- ---- --- --- -- --- ---
. . . .. .. . . . . . . . . . . . . ------ . . . . -----
. . . ----.
. . . . . . . .. .. . . . . . . -. . . . . . . . . . . . . ..--
500 --- ---------
400 - - - - - -- - - - - -
300 00 200 (N 100 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 C)
Time (s)
USSA _p013
Figure 7-14: Reactivity - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 1.0
--e Reactivity 0.5 -------------
0.0( 0 G q) (3 E) ie
-0.5 -----
-1.0 --------
. -- - -----------
...........
U
-1.5 ------
------------------------------ ------------------------------- --------------
-2.0 ---------------------------- -------------------
-2.5 00 (A
-3.0 0 600 1200 1800 2400 3000 36500 4200 4800 5400 6000 6600 Time (s) 00c 00 0 0
-p 0 14 USSA
Figure 7-15: Steam Generator Ruptured Tube Flow - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 48
-- Tube
-- x Tube Sheet Total 4--
-o - - - -- - ...
3--
...
..... . . . . .... ... .. ...... ... . ... . . . . . . . . . . . .
.6 ------------------------ ---.
¢00 0600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s)
-p 0 15 USSA
Figure 7-16: Makeup and Letdown Flow - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 40
--e Makeup Flow
-x Letdown Flow 35 .. .. .. .... .. . . . ... . .
- -------- -------- -------------------------..-----
30 ---- -------
25 ---------------- ------- .. . . .. . . . . . . . . . . . . . . . . . . . . .
------ -------- ------- ------ - .......................
U 20 15 -- -- ---- -- ------------------- ---------------- -------- ---- --- --- . . ---
. . . ------..
.. .
C 10 5 -------------- ------------
00 X X :)< )( : )( : X
-10 L-0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s) 0 16 USSA _p
Figure 7-17: High Pressure Injection Flow - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 120 I--0HPI 90 U
60 0
30 00 0 t * -. __________
600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s)
_p0 17 USSA
Figure 7-18: Pressurizer Heat - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 2.0 I-0 PZR Heaters 1.9 ------------------ ---------------
. ...
..
1.8 ------------------ ----------- - -
1.7
-----------------
i--i_ -- - - - -- -- ---
1.6 ---------------- ---------- ---------- -- -- ...........
1.5 ----------
----------
1.4 ----------- --------------- ----------
- ----------------
1.3 ---------- ------------- - -----------
1.2 ------- ---------- ---------- -------- - ----- -- - - - -- - - ---
1.1 --------------- ----------
1.0 --------------
..........----------
0.9 ------------ ---------- ----------
N 0.8 ........... -- - - - -- - - ---
---
0.7 ------------------
0.6 ---------- ----------
0.5 ----------
0.4 ----------
0.3 ------- ----------
0.2 ---------- ----------
0o 0.1 -- ---------- ..........
0.0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s)
USSA _pO1 8
Figure 7-19: Emergency Feedwater Flow - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 200 -- 0 SG A
-- SGB Total 160 2i 140 C,.)
1606 ---- -----.. .........-.--
------.-.-.---.--
.- -----------
--.--- ----
------
20-- ---- --------------------- --------
80 -
20 .. . . . . .. .. . . .. . . . . . - . '..
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s) 0 19 USSA _p
Figure 7-20: Pressurizer PORV Spray - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 20
[-oPORV Flow J
.. . .I 18 ------------ -----------
.................
16 -----------
14 ........... ------
U 12 ------ ---------- ---------------------------
-
.......... ---------- . . . . . . . .. . . . . . . . . . . -
. . . ..-
10 ------------
0 8 ------ ---------- ----------------------------
6 -----------------
4 ---------- ----------- ...........
2 ---------- -----------------
00 A
IJ 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s)
USSA _p 0 2 0
Figure 7-21: Main Steam Safety Valve Flow - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 5000000
-- SGA
-- tSGoB ATotal
--
4500000 4000000 3500000 ---------------
3000000 ------------
2500000 ---------------
2000000 ---------------
1500000 -----------
0 1000000 h 500000 00 Un Ii ~. - - ~ - - ~ - - p~
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s)
USSA _p0 2 1
Figure 7-22: Instantaneous ADV Flow - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 250
-- (DSG A
-~SGB
-Total 200 U
150 100 cds 50 00 0 600 1200 1800 2400 3000 3600 4200 4800 5 6600 Time (s)
C\
USSA _p 0 2 2
Figure 7-23: Subcooling Margin - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 160
-0 Margin 140 ------------------
-- ---------------------------------------------- -----------------
120 --------------- I ----------------
-----------------
-----------------
100 1-- ---------------------------- ------------------------------- -----------------
-7o ----- --------- I ---------
......------------------------ A-..
80 ----------
60 ----- -----------------------------
------------------ --------- ------
40( ---------------------------------
-- -------- -----------------
20 -----------------
- -----------------
--- --- ...... -- ---- ------:------ -----------------
00 0
0 600 1200 1800 2400 3000 36 600 4200 4800 5400 6000 6600 Time (s) 23 USSA _p 0
Figure 7-24: Steam Generator Tube-to-Shell Temperature Difference - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 125 _'SG A
-0 SG A
---< SG A, Wetted Region
-- A SG B
-- E SG B, Wetted Re ion 75 ------------------------.-. --.-.-
.-.-- -----.-
-.- -------------.--.- --.-
.- -. -.- --- .......
-~ 50
. . .. . 0-
.. -.-. ....
- - -- - -- - -- .. . . .. . .
25 . -----------
.. ------------ .. .. .- .-.-----.--.-.
.......- --------.------------.----..
.- ..
.. . . . . . . .. . . . . . ...
4--0
---- - - - -- - -
Z3o
-50 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 ,
Time (s) 00 USSA _p 0 2 4
Figure 7-25: Reactor Coolant System Mass - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 650000 i-__eROS 630000
. -.
6300000 --.-. --.-.-.
-.
--.- --------------------------- ---- ---------------. --------. -------.----.--.
-..-- -. -.-.--
--. .-
5900000 ---------------------------.
--.- -- .--- .------------.---.--------------
.......-----.--- .------.----------.------.---------
.-.------.-.-----.------- ----------------
-.-------.-----------
St 590000 570000 0
4950000 CIOo 00 450000 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s)
'~0 USSA _p 0 2 5
Figure 7-26: Flashing Fraction for Ruptured Steam Generator - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 10.00
-0 Ruptured SG 8.00 ---.------.---------- --------
..........-
..--.- --.-- .. . --- ---.--..---.--.--...
--.--..
.. .. -.. . . .. . .. .. . .. . .. . .. . .. . ..
.
. --------- . .. .. .. .. .. ..
.
rJ9 4.0000 -- ----.---------------.-------------------------------.--.--.------------.-----------------. ---------.------.----.
.-----------------. --------.---
-o
-2.00..................
0o
-6 .0 0 . . . . .. .... ......... . . . . . . . . . .. . . . .. . .. . .. . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . .
0600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 ON*
Time (s)
Cz USSA _p0 26
Figure 7-27: Condenser Flow - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 3900
-e Condenser Flow I I
3400 2900
. . . ...
. . . . . . . . . . . . .. --.. . . .. . .
2400
. . .. . . . . . . . . . . . . .. . . . . . . .. . .
1900
. . . .. . . . . . . . . . . . . . . . . . . . . . . . .
U 1400
. .----------..
.. . .--..
. . . .. .
900 400 00
, r-,
-100 L 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s) 0 27 USSA _p
Figure 7-28: Core Exit Temperature - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 650
-- o Core.
-- < Targe 625 600 575 550 525 -------------
U 500 -------------
U 475 450 425 00 400 0 600 1200 1800 2400 3000 3600 4200 Time (s) 02 8 USSA _p
Figure 7-29: ADV Valve Steam - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 1.10
--o SG A
--- SG B 1.00 0.90 0.80
-/-
0.70 ----------
S C2 0.60 0.50 --------- -
0.40 --------- -
0.30 0.20 --
0.10 00 ON Lit 0.00 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 ON Time (s)
USSA _p02 9
Figure 7-30: Integrated ADV and MSSV Flow - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 200000
-~MS 180000 SSVs SG B .........................
-.-.-------. ..
--.-.----.--.. .. .. .. ..
160000 140000 120000
©-
. . .. . .... --
.- .-.--
- -- . . . .
-.--.- .
ct 100000 80000 60000 -
40000 20000 00 0 i AFý i I i , ýý , I I - Iý 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s)
USSA _p 030
Figure 7-31: Steam Generator Secondary Side Liquid Level - Base Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 100
-- SG A
- SG B 90 -----------------------I------ --------- ------
80 ............ -------------------- ------ . .. . . . ..
..
-------------------------------- ------
70 0
--------------- ------
60
----------------- --------- ------------
50 40 ------------- ---------------------------------
Q)
(A
---------------: -----------
30 20 ------------- ............L---------
, -------
10 00 0 Uh 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 Time (s)
USSA _p 0 3 1
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary 7.2 Final Case 7.2.1 Event Description In the CRH event, the double-ended rupture of a steam generator tube is assumed to occur at full power. Primary system pressure decreases until a reactor trip occurs on low reactor coolant pressure. The turbine trips as a result of the reactor trip. Loss of offsite power (LOOP) is assumed to occur coincident with reactor trip. Consequently, reactor coolant pumps and main feedwater pumps trip. Emergency feedwater actuates automatically to raise steam generator liquid levels to fifty percent on the EFIC high range (natural circulation setpoint). The operator performs normal post-trip verifications and then at 45 minutes after reactor trip begins a cooldown of the reactor coolant system at a cooldown limit of 240°F/hr [9] using the atmospheric dump valves ensuring that the bad ROTSG level is maintained less than overfill. The operator finds that the ADV on the good ROTSG is not available and the cooldown commences with only the bad ROTSG initially. The PORV is not used for the CR1 analysis. The RCS is depressurized using PZR HPV's at 47 minutes after reactor trip as a conservative measure since it takes longer to initiate the depressurization and occurs at a lower rate than the PORV. After an additional 25 minutes following cooldown initiation (70 minutes after reactor trip), the ADV on the unaffected OTSG is opened as well. The reactor coolant system is limited to a cooldown rate of 100°F/hr at temperatures below 500'F and above 280'F [15]. The reactor coolant system cooldown continues by steaming the unaffected steam generator. Reactor coolant system pressure is maintained near the minimum subcooling margin by throttling high pressure safety injection (which started automatically on low reactor coolant pressure ESAS or was started manually by the operator). The HPI throttling is performed as necessary to keep the pressurizer level in an acceptable range with acceptable RCS subcooling margin beginning immediately after ESAS actuation [8]. The transient ends when the RCS temperature is < 500'F, and the RCS pressure is < 1000 psig. At this time the ruptured steam generator is considered isolated, since the ADV, MSIV's, blowdown, and FW/EFW feeds on the affected steam generator will be isolated.
7.2.2 Input The inputs and boundary conditions necessary for the analysis of the CR-3 SGTR event are included in Table 7-10 through Table 7-16.
Page 116
A Document No. 86-9159609-002 AREVA CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 7-10: SGTR Inputs and Boundary Conditions for CR Final Case Parameter Value RCS Conditions Core Power 1.004*3014 MWt = 3026.1 MWt Pump Power 4.03 MWt per pump Decay Heat 1.0*ANS 1971 w/B&W Heavy Actinides Primary Side Tave 582 OF RCS Pressure @ Hot Leg Tap 2170 psia Total RCS Mass Flow Rate 374,880 gpm Make-Up Flow Table 7-12 Pressurizer Parameters Indicated Pressurizer Level 220 inches Pressurizer Sprays and Heaters Sprays not modeled / Heaters (See Reference [1]
Section 6.1.5 for heater Capacities)
HPV Flow Diameter 0.375 inches [16]
PORV not modeled Main Feedwater System Parameters MFW Temperature 460 OF Initial MFW Flow per SG
Initiation LOOP/ Loss of 4 pumps Delay 40 s Flow 660 gpm Temperature 140°F EFIC Fill Rate 6.02 in/min Steam Generator Parameters Steam Generator Tube Plugging 5%
Steam Generator Inventory 65%-85% OR Turbine and Main Steam Parameters TBV Actuation Signal not modeled TBV Stroke Time not modeled Page 117
A AREVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary TBV Capacity not modeled MSSV Setpoints and Capacity Table 7-13 Turbine Trip Reactor Trip + 0.50 sec delay Turbine Stop Valve Closure Turbine Trip + 0.50 sec stroke MSIV Isolation (EFIC) Setpoint 649 psig MSIV Stroke Time 0 seconds Atmospheric Dump Valve Capacity 620,000 lbm/hr/valve ADV Opening Setpoint 1035 psig ADV Stroke Time **20 seconds RPS Parameters Low Pressure Reactor Trip 1893.95psia Reactor Trip Delay 0.61 seconds ESAS Low Pressure Trip 1714 psia ESAS Delay Time 0 seconds ECCS Parameters BWST Temperature 100 OF HPI Flow Table 7-14 Reactivity Control Parameters Control Rod Drop Time 1.4 seconds to 2/3 insertion Control Rod Insertion vs. Time Table 7-15 Delayed Neutron Fraction 0.007 (REA)
Prompt Neutron Generation Time 24.8 [tsec Doppler Coefficient -2.0x10 5 Ak/k/°F Moderator Temperature Coefficient 0.0x10 4 Ak/k/°F Operator Actions Key Operator Action Times Table 7-16 The MFW flow rate can be adjusted during initialization to achieve a steady-state.
- The expected stroke time is 25 to 30 seconds open. The event is insensitive to changes in ADV stroke time.
Table 7-11: Pressurizer Heater Parameters - Final Case Bank "O~n" Power (kW)
Emergency Diesel Backup With LOOP 252 The total heater power is assumed until the LOOP. Following LOOP, the emergency diesel backed heaters totaling 252 kW are assumed to be available. Modeling the emergency diesel backed heaters following LOOP is conservative, since the heaters would not be available unless manually aligned (Reference [1], page 12).
Page 118
A AR9VA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 7-12: Makeup Line Flow Versus Pressure for One Pump [12] - Final Case RCS Pressure MU Line Flow RCS Pressure MU Line Flow (psia) (gpm) (psia) (Ibm/s)1 900 298 900 41.08 1100 284 1100 39.18 1300 270 1300 37.27 1500 255 1500 35.22 1700 238 1700 32.89 1900 220 1900 30.42 2100 200 2100 27.67 2300 177 2300 24.50 2500 150 2500 20.78 2700 117 2700 16.22
- 1. The MU line flow was converted to lbm/s using a density for the conditions of the RCS pressure and a reasonable temperature of 120 'F. Slight variations in the pressure and temperature will have a negligible impact on the density of the makeup flow.
Table 7-13: MSSV Setpoints - Final Case Bank Number of Valves Nominal Setpoint Lift Pressure1 Reseat Pressure 2 (psig) (psig) (psig) 1 2 1050 1018.5 967.5 2 2 1070 1037.9 986.0 3 2 1090 1057.3 1004.4 4 23 1100 1067 1013.6 Notes:
- 1. Lift pressure represents 3% lift tolerance from nominal setpoint.
- 2. Reseat pressure represents 5% blowdown from lift pressure.
- 3. Per Reference [14] Section 10.3.4, 1 MSSV with a capacity of 583,574 lbm/hr has a nominal setpoint of 1100 psig.
Page 119
A AR.EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 7-14: HPI Flow [13] - Final Case RCS Pressure (psia) Flow (gpm) 15 904.3 615 815.9 1215 712.1 2515 373.3 Table 7-15: Control Rod Insertion [10] - Final Case Time (sec) Percent Reactivity Inserted (%)
0.0 0.00 0.2 0.58 0.3 0.99 0.4 1.83 0.6 5.29 0.8 12.33 1.0 21.41 1.2 33.09 1.4 50.75 1.6 72.96 1.8 91.30 2.0 99.26 2.2 99.99 2.3 100.00 2.4 100.00 Page 120
S'A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 7-16: Summary of Key Operator Actions - Final Case Action Time After Reactor Trip PZR HPV open. 47 min Begin RCS cooldown (ADV on affected SG open). 45 min ADV on unaffected SG open. 70 min Page 121
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary 7.2.3 Results The results of the SGTR CRH analysis are shown in Figure 7-32 through Figure 7-62. The sequence of events is presented in Table 7-17. A summary of results are provided in Table 7-18.
The initiating event was a double-ended rupture of a steam generator tube just below the lower face of the upper tube sheet. The primary system pressure decreased due to the loss of inventory (Figure 7-35). Reactor trip occurred at 33.92 minutes on low reactor coolant pressure (Figure 7-35). The turbine tripped as a result of the reactor trip and the turbine stop valves closed within one second (Figure 7-37). At the time of reactor trip a loss of offsite power (LOOP) occurs. At the time of LOOP the reactor coolant pumps trip off (Figure 7-41), a loss of main feedwater occurs (Figure 7-38), loss of the condenser occurs (Figure 7-58), and EFW actuation occurs 40 seconds later (Figure 7-50). Closure of the TSVs caused the main steam line pressure to increase and open the main steam safety valves (Figure 7-52 and Figure 7-61).
After LOOP, the emergency feedwater actuated and controlled the SG levels to fifty percent on the EFIC high range (Figure 7-43 and Figure 7-62). Following reactor trip, the low RCS pressure ESAS signal was reached and the makeup and letdown were isolated (Figure 7-47). Additionally, the HPI from two pumps was initiated (Figure 7-48). The HPI flow more than offset the tube leak and minimum pressurizer liquid level and RCS pressure were quickly regained (Figure 7-35 and Figure 7-36).
Forty-seven minutes after the reactor trip, control room operators begin reactor coolant system depressurization by opening the PZR HPV. Forty-five minutes after reactor trip, the control room operators begin cooldown by opening the ADV on each SG. At this point the operators observed that the ADV on the unaffected OTSG will not open. An equipment operator is dispatched to open the ADV on the unaffected SG. The operator commences the cooldown of the reactor coolant system using the ADV on the ruptured SG (Figure 7-34, Figure 7-37, and Figure 7-59). The subcooling margin (SCM) did not reach 110 F (Figure 7-54). The PORV opening to control subcooling margin was not modeled (Figure 7-51). Shortly after beginning cooldown, the MSSVs on the affected SG close and remain closed (Figure 7-61). The MSSVs on the intact SG close at approximately 84 minutes (Figure 7-61).
As the primary system pressure decreased, the mass flow rate through the ruptured tube also decreased (Figure 7-46). Figure 7-59 demonstrates that the core exit temperature compared reasonable well to the temperature targeted based on the cooldown rates. At approximately 98 minutes, the core outlet temperatures reached 500 F.
It should be noted that the core exit temperature achieved 500 F before the second ADV on the intact generator could be opened and the RCS pressure reached 1000 psig. The analysis was extended until the ADV on the intact generator was opened and the RCS pressure reached 1000 psig.
The SG tube-to-shell temperature differences can be seen in Figure 7-55. Section 5.0 states the acceptable SG tube-to-shell temperature differences are -90/+140 F. Larger tensile temperature differences would occur as the operator continues to cool down the RCS, but these values would be bounded by existing normal cooldown calculations. Consequently, additional loss of reactor coolant boundary integrity would not occur due to resultant temperature gradients during the SGTR transient, thus the acceptance criterion for the event is met.
The analysis confirmed that less than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of EFW injection is needed during the CRH analysis before the ruptured steam generator can be isolated as stated in Section 6.2.4 of Reference [I]. The ruptured steam generator was isolated at approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 8 minutes and the analysis was terminated at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 10 minutes.
No fuel failures are expected to occur during the SGTR event.
Page 122
ýA AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 7-17: SGTR CRH Sequence of Events - Final Case EVENT Time, seconds (minutes)
Reactor Trip on low RCS pressure, followed by 2035.3(33.9) coincident LOOP Control Rods Insert and Turbine Trip 2035.9 (33.9) lMSSVs first open SG A: 2037.7 (34.0)
SG B: 2037.9 (34.0)
ESAS low RCS pressure Signal 2091.0 (34.9)
HPI flow from two pumps started 2091.0 (34.9) 2 MSSVs last close SG A: 4733.0 (78.9)
SG B: 5005.8 (83.4)
Operators open the Pressurizer HPV valve. 4855.3 (80.9)
Operator begins RCS cooldown using ADVs.
ADV on intact SG fails to open remotely. 4735.3 (78.9)
Cooldown proceeds using ADV on faulted SG.
Tcore outlet reaches < 500 F 5858.4 (97.6)
ADV on intact SG opened manually. 6235.3 (103.9)
RCS pressure < 1000 psig 7697.0 (128.3)
Analysis terminated 7800.0 (130.0)
Notes:
Page 123
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary Table 7-18: SGTR CRH Summary of Results - Final Case Parameter Results Integrated tube leak 285,382 ibm Integrated tube leak 46,030 gal Integrated total MSSV release 164,691 ibm Integrated SG A MSSV release 108,094 ibm Integrated SG B MSSV release 56,597 ibm Integrated total ADV release 207,612 ibm Integrated SG A ADV release 165,358 ibm Integrated SG B ADV release 42,254 Ibm MSSV lift duration SG A: 2695.3 seconds (last close - first open, see Table 14-1) SG B: 2967.9 seconds Minimum subcooling margin 21.05 F @ 2019.4 seconds Maximum tube-to-shell temperature difference 83.46 F @ 7800.0 seconds (shell temp - tube temp)
SGTR CRH event duration 7800.00 seconds Page 124
Figure 7-32: Reactor Power - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 3200
- Reactor
¢) () L'* (J I U ,'U -
2800 2400 2000 1600 0
... . .. .. . . . . . . . ..
1200 800 400 --------------------- - ----- ---
00
---
~
--------- -------- -------
n 0
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 ON Time (s)
TAAI TCVL -pool
Figure 7-33: Steam Generator Power - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 1600 1400 1200 1000 800 I-600 C
0 400 200 0
-200 00 CN
-400 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s) t-)
TAAI TCVL _p002
Figure 7-34: Reactor Coolant System Temperature - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 650 - Tave
- TH-1 6O00 575 HH-450' 425 500 ---
400 ',
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 :
Time (s) ,
6---------
TAAI TCVL _p003
Figure 7-35: Primary Pressure - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 2600
-SG B,Hot Leg P-T Tap 2 4 00 -------------- -- -. -- . . --- - -- - - -------------- -- -. .-.-
. . . -. - -- - - - ------ -- --- ----- ----- - -- ---- .... ----
22 000 -------------.-. -----------.. ..-
-.
--.. -. -- ..-----
... - .. ... .. .. . .. ... ..... .. .. . .... .
d 1 1 2 00-
... .. ...... ... .. ..... ..... ......
...... .......... .....--.....
- ---. ..... ..... ..... ..... ...... ..... ..... ..... ...... . .. . . ..... .....
I400 00 1000 ,
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s) 11 00 TAAI TCVL _p004
Figure 7-36: Indicated Pressurizer Water Level - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 360 Pressurizer 300 ------------
-------------
C/ý 240 -----
--
C0 180 ---- ------- - ------------------- ------------------------ ------------
- ----------
........
7N 120 ----------- ------ -------- ---------- ---------
--- ---------------
I--------------
60 -----------------------
---------------
-----------
00 00 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s)
TAAI TCVL _pO05
Figure 7-37: Steam Generator Outlet Pressure - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 1200 SG A MSL A-1 SG A MSL A-2
SG B MSL B-1 1100 ---------------
1000 W 11 V -V ..... ----..... ...
900
--- - kp! " 9ý- - -
------------- -------
800 ---------- -------------- :-------
¢O --- - ------
700 600 ....... ....
0 . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . ... . . . . . . . .
500 ---------- ---- ------------ ---
0 400 ---------- ---------------------
300 ---------- ---------------------
200 ---------- ---------------------
100 ----------
----- 00 0
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s)
TAAI TCVL _p006
Figure 7-38: Steam Generator Outlet Pressure - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 2000
-- SG A MS
-xSG A MFW
-- SG B MS 1800 ------------------------- ---------- ---- SG B MFW
><,a2'>k :/,: X/,,u; 1600
.. .. .. . .. .. . . . . . . . . . . . . . . . . . .
rd~
1400 1200 0
1000 0
8001 0
S 600 1 400 200 1 00 0 _..- '- , /, .* . 3) q , .0 E "..Z ^.Z .. : .*}. .. 'Z.;} *Z* ".
- Z - *. *i"[;
.Z. L* "7 0 600 1200 1800 2400 3000 3600 4200 4800 -- 5400 6000 6600 7200 7800 Time (s)
TAAI TCVL _p007
Figure 7-39: Steam Generator Inventory - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 100000
-oSG A SG B 90000 1-------------- ..............
--------------
-------------------
------------ ---------- --------------------------------------- -------------
80000 --------------
.............
---------- ------------------------------------------- ----- ------
70000 --------------------- -- - ---- ---I---------
--------------- ------- -
Q 60000 --------------
- --- ----- ------------------- ---------- -----------------------------------------------------------
0 50000 ---------- --------------
- --------
---------- -------------------------------------------I----------- ---
40000 - --- ------- ..............
----------- ------
-- ---------------------------- ---------- ------ ----------------------------- ----------- ---
30000 -
---------------
......--------------------- ......
I -- ----------- ---------- ------------------------------------------------- ---------
---
00 20000 I 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time i TAAI TCVL _p008
Figure 7-40: Total Reactor Coolant System Flow Rate - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 400000 I-0 360000 320000 280000 240000 200000 U
160000 120000 80000 40000 ---------------
00 0 L 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s)
TAAI TCVL _p 0 0 9
Figure 7-41: Reactor Coolant Pump Flow Rate - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 10000 ______R -
____________ -ROP-1 A S---x RCP-1 B
-ARCP-2A
-4 RCP-2B 8 00 0 ------------------. -- --.------------.- . .. - --.-.- -.. --------.. - -- - --- ... .
. . . . . . . . -.-.--.--.-.---.. . . . .. . . . . . . .
--.
7 00 0 . . . . . . . . .-- . - . --.. -. . . . . . . . . . . .. . . . . - -- - -. -. -. -. - ..-- .- .-.. .. .- .--
-- . . .- .-. -- -.. .- .--. . .. -- .- .- .--. . .- .-. -- . . .- .- -.. -.- .-. -. . . .. . . . . .
-.. .. .. .. .
600 0 -- . .-.--
--- ---------.--.--
.-- -.- ..-- ---------.- --.- ----.- ------.-
-.- -------------.-.-.- -------.- -----------.- -.- .I.-
.-------------.
5 0 0 00 . . . .. . . . . . .. . . . . . . . .. . . . .. . . . . .. . . . .. .. . . . .. . . . . . . .. . . . ..... . . . . . .. . . . . ..... . . .. . . . . . . .. . . .. . . . .. . . .. . . . . . .. . . . . . . . . . . . . .. .... . . . . . . . . .. . . . .......
400 0 .. . . . . . .. . . .. . . . . . . . . .
0 00 31 .. . ... . . .. . . .. . . .. . .
3--
0 0--- ---
---
-- -- - ---- ----- --- ------------ ------------- ----- --------------- --------
U
-2000 _____o____
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 78000" Time -- (s) 4--
TAAI TCVL -pOlo
Figure 7-42: Pressurizer Surge Line Flow Rate - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 240 SSurge Line 200 ---------------
--------------
I----------------------------- -------
CI 160 1----------- --------------
120 ----------- - -------------
- -------------
--
-------------- -------------- -------
I --------------
----------------------------I-----------------------------
---- ----.-- ---- ---
-.
I-
---
---.-. ---- ---
80 1 -------------- -------------- -------------- -------
40 --------------
- -- ---- I- -------- - ---------------I--------------- ---------- -------
. .. ...... . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . .. . . . . . . .
0( -------------
-40 -------------- -------
-----------------------
-80 -----------------------------
I-------- -------------- --------
00 ON Lu
-120 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 ON Time (s)
TAAI TCVL _pO 1
Figure 7-43: Steam Generator Secondary Side Collapsed Level - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 30
-- 0SGA 28-x SG B 24: :: :
- - -----------------------
26 - --------- ----------- ------------ ----------- I--------- --------
- - 22---
03 20---
.18------------ ------
0C/.) 16---
U
.- 14------
rd/3
.
¢ 12 4- -- -------- ---------------
©0 00 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s)
TAAITCVL _p012
Figure 7-44: Core Average Fuel Temperature - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 1600
- VoI-Wtd 1500
-. .- ----------- . -------. ---------------- --------------.
. .. .. ... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -.
......
1300 1200 1100 03 U 1000 --------- ---- - -.------.-- -- .-- --- ---. ---- . . . . . . . . . . . . . .
1> 900 I-t..
. -.-------------. -. -------------.-.- ----------
800 ----------
H 700 ----- ------- ----------
----..... -.......
----..... ----- --------------
07 600 ---------------
--------------------
500 1--------------- I~
-- - - --- -- - -
--
400 1---- - -- ---
300 1---------------
00 200 ---------------
1 (VO 100 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s)
TAAI TCVL _p013
Figure 7-45: Reactivity - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 1.0
-0 Reactivitty 0.5 1 4 . .. . . . . . . . -
. ..-
__ --
--------------
0.0 e e --------------,. . ------
--------------
-0.5 ------------ --------------- ----------- ---------- ------ ------------
-1.0 --------------------------------- --- ...... -- ---------- -------------- ------ ---------------------------
-1.5 ------------- - ------- ------------
......
- -------------
-2.0 ------------
-------------- -------------- ------
I
-2.5 ---------- --- --------------------------- -- ---------- -------------- ------ --------------
-------------
00
-34 0 tJI
-30 5400 6000 6600 7200 7800 0 600 1200 1800 2400 3000 3600 4200 4800 Time (s) 00 TAAI TCVL _p014
Figure 7-46: Steam Generator Ruptured Tube Flow - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 48 F--ETube I 0
0,.
0) 00 C7 C>
15 TAAI TCVL _p0
Figure 7-47: Makeup and Letdown Flow - Final Case Thermal Hydraulic Analysis of SGTR for CRH1, LOOP 40
-e Makeup Flow
-* Letdown Flow 35
"30 25 ---------------------
U 20 15 10 5
-5 --------------
00
-10 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time 0
0 TAAI TCVL _p0 1 6
Figure 7-48: High Pressure Injection Flow - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 120 Ef-ý-pý]
90 ----------- ----------------------------------------- ------------ --- ---- ----
U Q
60 ------------1--------- ----- ---------- ------------------
......
- 30 -
--- ----------
I 00 0
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s)
TAAI TCVL _pO17
Figure 7-49: Pressurizer Heat - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 2.0
-- 0 PZR Heatr
--------------.. .. ----------. - -------------
. . . . .-
1.9
-.- .--
-------------
1.8 ------------- ---------- -------------
------ . .------ . ,-- -... . . . . . . . ..
-. -------------
1.7 1.6 ----------- ---------..----------------------------- --------- -----
1.5 -----------
1.4 ------------- -------------
1.3 -----------
.. .. --------.-. . . . . ..
1.2 -------- --- ------------ ------------- -----
- - . .- .- -. .- - - - - - - - - - -- - - -
.-- . . . . ..
. . . . .. . . . . . . . . . . . . . . . . . . . . . . .
1.1
.----
. . . .---- . . . ---
. . ----
. . . .. . .
1.0 ----------- -------------
a5 0.9 -----------
0.8 ----------- - -------------
0.7 ------------- 21-
.........................................
0.6 -------------------
0.5 ------------
------------ ...................
0.4
............
0.3 0.2 -------------
00
0.1 (11 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s)
TAAI TCVL _p018
Figure 7-50: Emergency Feedwater Flow - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 200 -- 0 SG A
-eSGA-
-- SGB
-- , Total 180 ----
160 ...
U C.* 140 120 80------------------------------
CI 40- ---------- ------I-- --- ----------------- ------- ------
20 -----------------------------.-------------4------z
- -- - -- - - - -- - -- - -- - - ----- ------ -- - --------
,00 0*
0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 01ý Time (s)
TAAI TCVL p019
Figure 7-51: Pressurizer PORV Spray - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 20 -OPORV FIow C,.)
S12 10..
18----
- -- ------ --- ------ --- --- ---- - --- -- ------ --- --- --------- ------I
- ---- - ---
16 - ----------- -------- - --
- --- ---
-- ------
---
--- - ------------ ------- ------------ ----------
14 -- - - - - - - - - - - - - -- - - - - - -- - - - - - -- - - - - - - - -- - - - - - -- - - - - - - -- - - --
0~~~~~~~------- -------- , o -------- c o' o 'o o'oodo ©o3' oo'.
b4 12 --------- ------------------- --------- --------- ---- --- ------ 4200----- 480
--- -- - - - -- -- - - --- ---- 0 - - - - -660 - - - - 720
- -- - - -- -
10---------- - -- ---------
--- - -- - --- - --
-- ---- --- - ----- --- -- -- --- -- - -- - --- - -- - - --- -- - ---- -- - -- - - -- - -- -
8-- ----
~~T-s--
---(-- ------ -- - ----
--------- i----
m ---- -- ---- ---------- ---- -------- -- ------------
Time (s TAAI TCVL _p020
Figure 7-52: Main Steam Safety Valve Flow - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 5000000
-oSGA SGTB
-A- Total
...... ........... .......... . ........... ..... . ..... ..........
.. . ........ ........... . .......... ..... ..... .....
........... ........-- . ...........
4500000 4000000 3500000 3000000 2500000 2000000 1500000 1000000 j
500000 00
.01 n
0
- Q" ý 0i " Q ,i !!ý ii Ii..ý.!i L; - ~ I 1.-1 . ,-I IIPý- III I1 - " ,. III- I 600 i200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 - -- 7800 Time (s) I
_p0 2 1 TAAI TCVL
Figure 7-53: Instantaneous ADV Flow - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 250
-o SGA
-- SGB
-- Totat 200 150 ----------
ct CI 100 ------------
50 -------------- .......... ------------
LI A 00 0 600 1200 1800 2400 3000 3600 4200 4800 5400 7800 0D Time (s)
TAAI TCVL _p022
Figure 7-54: Subcooling Margin - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 160 140 120 100 80 0o 60 40 20 00 0
0 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s)
TAAI TCVL
Figure 7-55: Steam Generator Tube-to-Shell Temperature Difference - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 125
-0 SG A
--- SG A, Wetted Region
-- SG B
-- SG B, Wetted Region 100 1
----
75 I K I I 50 P
Q 25 ------------ ----------
~j) tx',
N H IFn:
10 ----------------
-. -A .--
-25 7-----
- --------- -----------
00
-50 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s) 00*
TAAITCVL _p024
Figure 7-56: Reactor Coolant System Mass - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 650000 630000 610000 590000 ...............
4 570000 cJ~
550000 C
530000 U
C 510000 490000 470000 00 450000 450000 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s)
TAAI TCVL _p025
Figure 7-57: Flashing Fraction for Ruptured Steam Generator - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 10.00 p p p p p p
-e Ruptured SG
. -------------. ..
..
8.00 ---------------------------
I
--- ----------
------------
- -- - - - - - - ----- - - -- - - - - - - - - -- - - - - - - - - - - - -
6.00 4.00 ------------- ---------------------
--------------
-------------------------- --------------- -------------- --------------
0) 2.00 7 T 7 7it --- -- -- ------------ ----
----------
- ------ ------------------------ ------------------ --------------
4-0.004 I--------------
--------------
-------------- ----------
............------------------ ----------------------------------
-2.00 ---------------
--------------------------
-- ----------
cdt
-4.00 ---------- 7-----------
--------------
P
-6.00 -------------- ----
I--------------- ----
---- P
--
-8.00 ------------- I---------------
--------------
-------
------ 00
-in* on(
-1000 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s)
TAAI TCVL _p026
Figure 7-58: Condenser Flow - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 3900
-- o Condenser Flow c) e E) E) .b 3400 ---------------------
r ------------.
-- . . . . . . . . . . . .-, -------------
2900 --------------
. .. ... ,
-- - - -- - - ---------------..
2400 -------------- I--
-------------
1900 ----------------------------
CIO 1400 -------------
U 900 -------------
400 ----------
00
-100 I 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s)
TAAI TCVL _p027
Figure 7-59: Core Exit Temperature - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 650 p p p - _____________________________ p p
-0 Core Exit
-- Target 625 ----------- I--------------- -------------
600 ------------- ---------------------------- - -------------
575 ................
-------- ----
--------------
550 -------- -------------- ------------------------------- - -- ---- --- -------------------------- --------------
CI-
----- ------------
A -- --------------
525 I-----------
-----------
U
----------- -------------- ----------- -------------- ---------
500 475 -------------- ---------- I -------------- - --------- --------------
450 -------------- ------------ I -------- ------------ -------------- --- .......
425 ------------ -------------- --------------
---------- ------ -----
00 A (~A
- UU 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 Time (s)
TAAI TCVL _p028
Figure 7-60: ADV Valve Steam - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 1.10
-- 0 SG A
-- SGB 1.00 0.90 0.80 0.70 0.60 0.50 0.40 0.30 0.20 0.10 I0 0,,
0.00 600 1200 1800 2400 3000 3600 4200 5400 6000 6600 7200 7800 Time (s) I TAAI TCVL _p0 2 9
Figure 7-61: Integrated ADV and MSSV Flow - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 200000_____
-A MSSVs SG A 18 0 0 0 0 --------------.-. .-----
-.- .-.--
. .----.------- .-- .---------
.--------
.------ .---------.
.------------ ---. El M S SS s SSG B
180000 ----- ----- -------- s-----S----------
M--------S--
-600- - - - -- - - - -- - - - - -- - - - -- --- - - - - - --- -- -
..o 120000 0
-e 100000
-* 80000 60000 i!i 40000 - - - - - - -
20000 00 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800----
Time (s) E3 00 TAAI TCVL _p030
Figure 7-62: Steam Generator Secondary Side Liquid Level - Final Case Thermal Hydraulic Analysis of SGTR for CRH, LOOP 100
-- SG A
- SG B 90 --------------
--------------
...........
E) 80 0
70 60
........... ............... ------ ------------------------------- ---
i --
----- -- ---- i
I-
-e 50 ..............I .........
V 40 ------------------ --------------------------------------
0 30 ------------------------ --------------
20 -- --------
I --------------
10 ----------
00 ON 0 0 600 1200 1800 2400 3000 3600 4200 4800 5400 6000 6600 7200 7800 ON Time (s)
TAAI TCVL _p031
A AR EVA Document No. 86-9159609-002 CR-3 Steam Generator Tube Rupture Event for EPU Summary
8.0 REFERENCES
- 1. AREVA NP Inc. Document 51-9061032-006, "CR3 EPU Steam Generator Tube Rupture AIS".
- 2. AREVA NP Inc. Document 32-9067265-003, "CR3 Steam Generator Tube Rupture Event for EPU".
- 3. AREVA NP Inc. Topical Report, "RELAP5/MOD2-B&W - An Advanced Computer Program for Light Water Reactor LOCA and Non-LOCA Transient Analysis", BAW-10164P-A, Rev. 6, June 2007.
- 4. AREVA NP Inc. Document 32-9055717-001, "CR-3 Non-LOCA Base Decks for EPU".
- 5. AREVA NP Inc. Topical Report, "RELAP5/MOD2-B&W - For Safety Analysis of B&W-Designed Pressurized Water Reactors", BAW-10193P-A, January 2000.
- 6. AREVA NP Inc. Topical Report "BWNT LOCA - BWNT Loss-of-Coolant Accident Evaluation Model for Once-Through Steam Generator Plants, BAW 10192PA, Rev. 0, June 1998.
- 7. AREVA NP Inc. Document 38-9141279-000, "CR-3 DIT ROTSG Tube-to-Shell Temperature Limits."
- 8. AREVA NP Inc. Document 51-1266316-01, "AIS For Analysis of SGTR For CR-3".
- 9.
- Crystal River Unit 3, EOP-06, "Steam Generator Tube Rupture", Revision 20.
- 10. AREVA NP Inc. Document 51-9058845-000, "CR-3 Safety Analysis Kinetics Input."
- 11. AREVA NP Inc. Document 51-5064728-00, "CR-3 SBO AIS".
- 12. AREVA NP Inc. Document 32-5006067-00, "CR-3 MU Flow vs. RCS Pressure".
- 13. AREVA NP Inc. Document 32-9145453-001, "CR-3 I-WI Flow with Modified Throttle Valves MUV 590-593".
- 14.
- Crystal River Unit 3, EOP-14, "Emergency Operating Procedure Enclosures", Revision 26.
- 15.
- Crystal River Unit 3, "Enhanced Design Basis Document for the Reactor Coolant System," Revision 11, DBD6 1.
- These documents are maintained and controlled in a retrievable form by B&W Canada or Progress Energy Florida. Per AREVA NP Inc. procedures, use of these references is allowed in safety-grade calculations with the approval of the project manager as denoted on page 2 of this document.
Page 156