ML13037A201
| ML13037A201 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/31/2013 |
| From: | GE-Hitachi Global Laser Enrichment |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| DRF Section 0000-0156-7958 R0, L-MT-13-019 NEDO-33800, Rev 0 | |
| Download: ML13037A201 (36) | |
Text
L-MT-13-019 ENCLOSURE 3 NEDO-33800, MONTICELLO NUCLEAR GENERATING PLANT AUTOMATIC DEPRESSURIZATION SYSTEM BYPASS TIMER EXTENDED POWER UPRATE, REVISION 0 35 pages follow
0 HITACHI GE Hitachi Nuclear Energy NEDO-33800 Revision 0 DRF Section 0000-0156-7958 RO January 2013 GEH Non-Proprietary Information - Class I (Public)
Monticello Nuclear Generating Plant Automatic Depressurization System Bypass Timer Extended Power Uprate Copyright 2013 GE-Hitachi Nuclear Energy Americas LLC All Rights Reserved
NEDO-33800 Revision 0 NON-PROPRIETARY NOTICE This is a non-proprietary version of the document NEDC-33800P Revision 0, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here ((
I].
IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The design, engineering, and other information contained in this document is furnished for the purpose of supporting Monticello Nuclear Generating Plant license amendment request for an extended power uprate in proceedings before the U. S. Nuclear Regulatory Commission. The only undertakings of GEH with respect to information in this document are contained in the contract between GEH and Entergy, and nothing contained in this document shall be construed as changing that contract. The use of this information by anyone for any purpose other than that for which it is intended is not authorized; and with respect to any unauthorized use, GEH makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.
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NEDO-33800 Revision 0 TABLE OF CONTENTS 1.0 Introduction.........................................................................................................................
1 2.0 Description of M odels.....................................................................................................
1 3.0 A nalysis Procedure..........................................................................................................
1 4.0 Input to Analysis............................................................................................................
2 5.0 Analysis...............................................................................................................................
3 5.1 A nalysis - OLTP - RW CU.........................................................................................
4 5.2 A nalysis - EPU - RW CU............................................................................................
5 5.3 A nalysis - EPU - STM O.............................................................................................
6 5.4 A nalysis - Increased Core Flow.................................................................................
6 6.0 Results.................................................................................................................................
7 6.1 Results - EPU - RW CU..............................................................................................
7 6.2 Results - EPU - STM O................................................................................................
8 7.0 Conclusion..........................................................................................................................
9 8.0 References.........................................................................................................................
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NEDO-33800 Revision 0 List of Tables Table 1 Analysis Plant Operational Parameters.................................................................
2 Table 2 RWCU Line Break - PCT Results........................................................................
7 Table 3 RWCU Line Break - MELLLA+ Core Flow and Gate Valve Closure -
Sequence of E vents..........................................................................................
.......... 7 Table 4 STMO Line Breaks - PCT Results........................................................................
8 Table 5 STMO Line Break - Rated Core Flow - Sequence of Events..............................
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NEDO-33800 Revision 0 List of Figures Figure 1 ADS Actuation Logic.........................................................................................
10 Figure 2-a Water Level in Hot and Average Channels.............................................................
11 Figure 2-b Water Level in Plenum and Jet Pump..................................................................
12 Figure 2-c Water Level in Downcomer (Saturated (7) and Subcooled (6))......................... 13 Figure 2-d Reactor Vessel Pressure........................................................................................
14 Figure 2-e Peak Clad Temperature........................................................................................
15 Figure 2-f E C C S F low..............................................................................................................
16 Figure 3-a Water Level in Hot and Average Channels........................................................
17 Figure 3-b Water Level in Plenum and Jet Pump.................................................................
18 Figure 3-c Water Level in Downcomer (Saturated (7) and Subcooled (6))......................... 19 Figure 3-d Reactor Vessel Pressure.....................................................................................
20 Figure 3-e Peak Clad Temperature........................................................................................
21 Figure 3-f E C C S Flow..............................................................................................................
22 Figure 4-a Water Level in Hot and Average Channels........................................................
23 Figure 4-b Water Level in Plenum and Jet Pump.................................................................
24 Figure 4-c Water Level in Downcomer (Saturated (7) and Subcooled (6))......................... 25 Figure 4-d Reactor Vessel Pressure.....................................................................................
26 Figure 4-e Peak Clad Temperature........................................................................................
27 Figure 4-f EC C S Flow..............................................................................................................
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NEDO-33800 Revision 0 ACRONYMS Term Definition ADS Automatic Depressurization System CFR Code of Federal Regulations CS Core Spray DW Drywell ECCS Emergency Core Cooling System EPU Extended Power Uprate GEH GE-Hitachi Nuclear Energy Americas LLC HPCI High Pressure Coolant Injection LOCA Loss-of-Coolant Accident LPCI Low Pressure Coolant Injection MELLLA+
Maximum Extended Load Line Limit Analysis Plus MSIV Main Steam Isolation Valve OLTP Original Licensed Thermal Power PCT Peak Cladding Temperature RCIC Reactor Core Isolation Cooling RPV Reactor Pressure Vessel RWCU Reactor Water Cleanup RX Reactor SRV Safety Relief Valve STMO Main Steam Line Break Outside Containment vi
NEDO-33800 Revision 0
1.0 INTRODUCTION
A combined modification of the automatic depressurization system (ADS) and emergency core coolant system (ECCS) pump start logic was implemented at Monticello to reduce the dependence on operator action for certain events, which may require rapid reactor pressure vessel depressurization. The modification includes eliminating the high drywell pressure permissive of the ADS logic and adding a bypass timer to the low reactor pressure permissive of the low pressure ECCS pump start logic. The modification provides automatic ADS initiation for line breaks outside containment. An analysis was performed at original licensed thermal power (OLTP) to determine the ADS bypass timer setting in Reference 1.
The purpose of this document is to report an ECCS-Loss-of-coolant accident (LOCA) reanalysis of the base analysis at OLTP in Reference 1 for Monticello using SAFER/GESTR-LOCA and then extend the ECCS-LOCA reanalysis for extended power uprate (EPU) conditions and maximum extended load line limit analysis plus (MELLLA+) conditions. Specifically, using the limiting case as defined in Reference 1 along with the latest ECCS-LOCA analysis input and calculation procedures, results are provided for the GE14 fuel analyzed at EPU.
2.0 DESCRIPTION
OF MODELS The ECCS-LOCA results are generated using the standard evaluation model, which consists of computer codes, LAMB, TASC, SAFER and GESTR-LOCA. Reference 2 provides further description. Only SAFER and GESTR-LOCA are run since boiling transition times calculated by LAMB and TASC are not required for this analysis.
3.0 ANALYSIS PROCEDURE The Monticello specific SAFE/REFLOOD-LOCA analysis in Reference 1 evaluated LOCA events outside of containment for the peak clad temperature (PCT) with Appendix K assumptions. The reactor water cleanup line break outside containment (RWCU) ((
)) and the main steam line break outside containment (STMO) ((
were evaluated in Reference 1 at OLTP.
((
)) was re-evaluated with the current methodology of SAFER/GESTR-LOCA to benchmark with the RWCU break case in Reference 1. The RWCU and STMO breaks were both evaluated at EPU with the current methodology.
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NEDO-33800 Revision 0 4.0 INPUT TO ANALYSIS The Monticello heat balance conditions used in the analysis for EPU with rated core flow and MELLLA+ core flow are listed in Table 1.
Table 1 Analysis Plant Operational Parameters Parameter Unit Nominal Appendix K Core Thermal Power MWt 2004.0 2044.1
% of Extended Power Uprate 100.0 102.0 Rated Core Flow Mlbm/hr 57.6 57.6
% of Rated Core Flow 100.0 100.0 MELLLA+ Core Flow Mlbm/lhr 46.1 46.1
% of Rated Core Flow 80.0 80.0 Vessel Steam Dome Pressure psia 1025.0 1040.0 Feedwater Temperature OF 400.8 402.9 2
NEDO-33800 Revision 0 5.0 ANALYSIS The ADS functions as a backup to the operation of the high pressure injection systems, e.g.
feedwater, HPCI, reactor core isolation cooling (RCIC) by means of selected safety relief valves.
These systems when augmented with the ADS provide protection against excessive fuel cladding heatup for hypothetical loss of coolant events over a range of steam or liquid line breaks inside the drywell. The ADS depressurizes the vessel, permitting the operation of the low pressure injection systems of low pressure coolant injection (LPCI) and core spray.
Monticello USAR Section 6.2.5.1 indicates that Monticello implemented Option 2B in response to NUREG-0737, Item II.K.3.18, "ADS Logic Modifications." Option 2B eliminated the high drywell pressure permissive from the ADS logic and added a manual switch allowing the operator to prevent (inhibit) automatic ADS initiation. In addition, the Option 2B modification added a bypass timer for the low reactor vessel pressure permissive for the low pressure ECCS pump start logic. The ADS is currently activated automatically upon coincident signals of low water level in the reactor pressure vessel (RPV) and low pressure emergency core coolant system (ECCS) pump running. The complete start sequence for the ADS actuation logic is shown in Figure 1.
Reference I presented an analysis that determined the setting for the bypass timer at OLTP. The bypass timer and associated Option 2B modification extend ADS operation for pipe breaks outside containment and eliminates the requirement for operator action for those loss of coolant accident or other postulated events that do not result in a high drywell pressure trip but may require depressurization of the reactor vessel to maintain adequate core cooling. Reference I identified the limiting event for determining the bypass timer setting ((
)) This basis remains valid for current plant configuration.
The goal of the bypass timer setting calculation in Reference 1 was to determine a timer setting that would limit the PCT significantly below the 10 Code of Federal Regulations (CFR) 50.46 limiting PCT of 2200'F (approximately 1500'F) for the limiting event to assure adequate core cooling for transient or accident events that may require depressurization. ((
)) with the PCT of approximately 1500'F, which is well below the 10 CFR 50.46 acceptance limit of 2200'F.
This report presents the analysis for EPU power that determines the time from event initiation to the time of PCT at 2200'F followed by ADS operation. The evaluation applies to events that do not result in a release to the drywell but may require depressurization of the reactor pressure vessel to maintain adequate core cooling. The limiting events that result in the most inventory loss without any release of steam to the drywell are the RWCU and STMO breaks. The RWCU break is reanalyzed at OLTP in Section 5.1 and then at EPU in Section 5.2. The STMO break is analyzed at EPU in Section 5.3. The 10 CFR 50.46 notifications up to and including Notification 2011-03 have been incorporated into the analysis.
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NEDO-33800 Revision 0 5.1 ANALYSIS - OLTP - RWCU Reference I describes the modification of the ADS logic and ECCS pump start logic to assure adequate core cooling for the events requiring rapid depressurization. Adequate core cooling was assured if the PCT for an event was less than 15007F. Although the maximum PCT imposed by 10 CFR 50.46 is 2200'F, a lower limit was established for the bypass timer analysis at OLTP.
The lower limit assured the line break event outside containment would not be the limiting event for the plant safety analyses.
The limiting events were the large recirculation line breaks inside containment, which cannot be isolated and therefore remain as the limiting events. Thus, the goal for the analysis in Reference I at OLTP was to determine a setting for the bypass timer which would limit the PCT significantly below 2200'F (approximately 1500'F) for the limiting line break event outside containment.
Following the main steam line isolation valve (MSIV) closure and the RWCU line isolation, the reactor pressure increases up to the setpoint to activate the safety relief valves (SRV). Without feedwater, HPCI and RCIC, the SRV actuations gradually reduce the reactor water inventory down below top of active fuel, which causes the PCT to rise to 15007F. The ADS initiates at this point to depressurize the reactor vessel. The low pressure ECCS pumps then are able to start and inject into the reactor vessel to provide core cooling.
Reference I Section 3.0 indicates that ADS actuation occurred ((
)) with the PCT of approximately 1500'F, which is well below the 10 CFR 50.46 acceptance limit of 22001F. The analysis in Reference I was performed using the SAFE/REFLOOD-LOCA methodology. The RWCU line break outside containment at OLTP analysis in Reference 1 was reanalyzed with the current methodology of SAFER/GESTR-LOCA. This provides a benchmark for the RWCU line break outside containment at EPU power in Section 5.2 and STMO line break outside containment at EPU power in Section 5.3.
The analysis method described in Reference 1 Section 2.3 regarding the reactor power, initial water level, RWCU isolation valve closing time, loss of feedwater flow, unavailability of HPCI and RCIC, availability of LPCI and core spray, decay heat and limiting break area was used for the reanalysis of the RWCU line break outside containment at OLTP using the SAFER/GESTR-LOCA methodology. The reanalysis determined that with ADS actuation ((
))
the PCT is 1500'F, which is consistent with Reference 1. The resulting plots for the reanalysis of the RWCU line break outside containment at OLTP are provided in Figure 2.
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NEDO-33800 Revision 0 5.2 ANALYSIS - EPU - RWCU The RWCU line break outside the containment was analyzed at EPU power. This event can result in a significant amount of inventory loss before the break isolates and result in the reactor remaining at high pressure after the break isolates. The event and resultant PCT are evaluated in accordance with the initial conditions and assumptions listed below.
The reactor is operating at 102 % of rated power. This maximizes the fuel cladding heatup and conforms with the requirements of 10 CFR 50 Appendix K. The initial heat balance values defined in Reference 3 for EPU power were used. A small feedwater temperature increase will slightly decrease the time of PCT at 2200'F.
The initial reactor water level is at the scram level, i.e. the low water level. This minimizes the time for ADS initiation.
Feedwater flow is assumed lost within one second after the break occurs. This assumption minimizes the amount of inventory available to the reactor.
The decay heat value is 120 % of the 1971 ANS standard which conforms to the requirements of 10 CFR 50 Appendix K.
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NEDO-33800 Revision 0 5.3 ANALYSIS - EPU - STMO The STMO line break outside the containment was analyzed at EPU power. This event can result in a significant amount of inventory loss and result in the reactor remaining at high pressure after the break isolates. The event and resultant PCT are evaluated in accordance with the initial conditions and assumptions listed above regarding the heat balance, feedwater flow, high pressure systems, low pressure systems and decay heat. The initial reactor water level is at the normal water level.
5.4 ANALYSIS - INCREASED CORE FLOW The input changes for increased core flow are the heat balance, pressure drop conditions and boiling transition times. Increased core flow conditions are characterized by reduced downcomer subcooling that results from the increased core flow. The reduced downcomer subcooling results in a lower initial break flow, later dryout time and a later core uncovery. The PCT results for increased core flow conditions are usually lower than at rated core flow conditions. Therefore, the RWCU and STMO breaks were not evaluated with increased core flow conditions.
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NEDO-33800 Revision 0 6.0 RESULTS Analysis results for RWCU line break are in Section 6.1 and STMO line break are in Section 6.2.
6.1 RESULTS - EPU - RWCU The RWCU results at EPU power regarding the time for the PCT to reach 2200'F are listed in Table 2. The time for PCT to reach 2200'F is the total time from the start of the LOCA event and includes the time delay from the break occurrence to the ADS bypass timer initiation, the ADS bypass timer, the ECCS pump start time and the ADS actuation timer. The difference in time of PCT to reach 2200'F between rated core flow and MELLLA+ core flow is very small ((
)) The limiting RWCU line break scenario at EPU power is MELLLA+ core flow and gate valve closure.
Table 2 RWCU Line Break - PCT Results Line Core Flow RWCU Time of PCT at 22001F Time of PCT at Break Isolation Valve (seconds) 22001F (minutes)
Closure The sequence of events for the limiting RWCU line break scenario at EPU power of MELLLA+
core flow and gate valve closure is listed in Table 3 and the plots are in Figure 3.
Table 3 RWCU Line Break - MELLLA+ Core Flow and Gate Valve Closure -
Sequence of Events Time Time Description (seconds) (minutes) 1___
1]
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NEDO-33800 Revision 0 6.2 RESULTS - EPU - STMO The STMO results at EPU power regarding the time for the PCT to reach 2200'F are listed in Table 4. The time for PCT to reach 2200'F is the total time from the start of the LOCA event and includes the delay time from the break occurrence to the ADS bypass timer initiation, the ADS bypass timer, the ECCS pump start time and the ADS actuation timer. ((
)) The limiting STMO line break scenario at EPU power is rated core flow.
Table 4 STMO Line Breaks - PCT Results Core Flow Time of PCT at 2200IF Time of PCT at 2200'F (seconds)
(minutes)
The sequence of events for the limiting STMO line break scenario at EPU power of rated core flow is listed in Table 5 and the plots are in Figure 4.
Table 5 STMO Line Break-Rated Core Flow - Sequence of Events Time Time Description (seconds)
(minutes)
_ 4.
I.
+
+
+
4-4-8
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7.0 CONCLUSION
Plant specific analyses were performed at EPU power for the RWCU and STMO line break events outside containment to determine the time for the PCT to reach 22000F. ((
))
The time for PCT to reach 2200°F is the total time from the start of the LOCA event and includes the time delay from the break occurrence to the ADS bypass timer initiation, the ADS bypass timer, the ECCS pump start time and the ADS actuation timer.
The ADS bypass timer setting can be determined based on the limiting time for the PCT to reach 2200'F from the results provided in Table 2 and Table 4. The setting provides sufficient margin to assure that the events outside containment can be mitigated and do not become limiting events for plant safety analyses.
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NEDO-33800 Revision 0 Figure 1 ADS Actuation Logic 10
NEDO-33800 Revision 0 Figure 2-a Water Level in Hot and Average Channels OLTP, RWCU Line Break Outside Containment, HPCI Single Failure GE14 Fuel, (2 CS + 4 LPCI + 3 ADS) Available, Appendix K Assumptions 11
NEDO-33800 Revision 0 Figure 2-b Water Level in Plenum and Jet Pump OLTP, RWCU Line Break Outside Containment, HPCI Single Failure GE14 Fuel, (2 CS + 4 LPCI + 3 ADS) Available, Appendix K Assumptions 12
NEDO-33800 Revision 0 Figure 2-c Water Level in Downcomer (Saturated (7) and Subcooled (6))
OLTP, RWCU Line Break Outside Containment, HPCI Single Failure GE14 Fuel, (2 CS + 4 LPCI + 3 ADS) Available, Appendix K Assumptions 13
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((I Figure 2-d Reactor Vessel Pressure OLTP, RWCU Line Break Outside Containment, HPCI Single Failure GE14 Fuel, (2 CS + 4 LPCI + 3 ADS) Available, Appendix K Assumptions 14
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((I I]
Figure 2-e Peak Clad Temperature OLTP, RWCU Line Break Outside Containment, IPCI Single Failure GE14 Fuel, (2 CS + 4 LPCI + 3 ADS) Available, Appendix K Assumptions 15
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((I Figure 2-f ECCS Flow OLTP, RWCU Line Break Outside Containment, HPCI Single Failure GE14 Fuel, (2 CS + 4 LPCI + 3 ADS) Available, Appendix K Assumptions 16
NEDO-33800 Revision 0 Figure 3-a Water Level in Hot and Average Channels EPU, RWCU Line Break Outside Containment, HPCI Single Failure GE14 Fuel, (2 CS + 4 LPCI + 3 ADS) Available, Appendix K Assumptions 17
NEDO-33800 Revision 0 II Figure 3-b Water Level in Plenum and Jet Pump EPU, RWCU Line Break Outside Containment, HPCI Single Failure GE14 Fuel, (2 CS + 4 LPCI + 3 ADS) Available, Appendix K Assumptions 18
NEDO-33800 Revision 0 I]
Figure 3-c Water Level in Downcomer (Saturated (7) and Subcooled (6))
EPU, RWCU Line Break Outside Containment, HPCI Single Failure GE14 Fuel, (2 CS + 4 LPCI + 3 ADS) Available, Appendix K Assumptions 19
NEDO-33800 Revision 0 Figure 3-d Reactor Vessel Pressure EPU, RWCU Line Break Outside Containment, HPCI Single Failure GE14 Fuel, (2 CS + 4 LPCI + 3 ADS) Available, Appendix K Assumptions 20
NEDO-33800 Revision 0 Figure 3-e Peak Clad Temperature EPU, RWCU Line Break Outside Containment, HPCI Single Failure GE14 Fuel, (2 CS + 4 LPCI + 3 ADS) Available, Appendix K Assumptions 21
NEDO-33800 Revision 0 Figure 3-f ECCS Flow EPU, RWCU Line Break Outside Containment, HPCI Single Failure GE14 Fuel, (2 CS + 4 LPCI + 3 ADS) Available, Appendix K Assumptions 22
NEDO-33800 Revision 0 Figure 4-a Water Level in Hot and Average Channels EPU, Steam Line Break Outside Containment, Battery Failure GE14 Fuel, (1 CS + 2 LPCI + 3 ADS) Available, Appendix K Assumptions 23
NEDO-33800 Revision 0 I((
Figure 4-b Water Level in Plenum and Jet Pump EPU, Steam Line Break Outside Containment, Battery Failure GE14 Fuel, (1 CS + 2 LPCI + 3 ADS) Available, Appendix K Assumptions 24
NEDO-33800 Revision 0 Figure 4-c Water Level in Downcomer (Saturated (7) and Subcooled (6))
EPU, Steam Line Break Outside Containment, Battery Failure GE14 Fuel, (1 CS + 2 LPCI + 3 ADS) Available, Appendix K Assumptions 25
NEDO-33800 Revision 0 Figure 4-d Reactor Vessel Pressure EPU, Steam Line Break Outside Containment, Battery Failure GE14 Fuel, (1 CS + 2 LPCI + 3 ADS) Available, Appendix K Assumptions 26
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((I Figure 4-e Peak Clad Temperature EPU, Steam Line Break Outside Containment, Battery Failure GE14 Fuel, (1 CS + 2 LPCI + 3 ADS) Available, Appendix K Assumptions 27
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Figure 4-f ECCS Flow EPU, Steam Line Break Outside Containment, Battery Failure GE14 Fuel, (1 CS + 2 LPCI + 3 ADS) Available, Appendix K Assumptions 28
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8.0 REFERENCES
- 1. AE-79-0884, "Bypass Timer Calculation for the ADS/ECCS Modification for Monticello",
August 29, 1984.
- 2. NEDC-32514P, "Monticello SAFER/GESTR-LOCA Loss of Coolant Accident Analysis",
Revision 1, October 1997.
- 3. Design Information Transmittal (DIT) Number 1253620-1, S.J. Hammer (MNGP) to J.A. Hren (GEH), May 22, 2012.
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