Plant Dynamics Evaluation of a Monju Ex-Vessel Fuel Storage System During a Station Blackout

2012 ◽  
Author(s):  
Takero Mori ◽  
Masutake Sotsu ◽  
Kei Honda ◽  
Satoshi Suzuki ◽  
Hiroaki Ohira
Keyword(s):  
Structural Integrity ◽  
Cooling System ◽  
Natural Circulation ◽  
Heat Removal ◽  
Spent Fuel ◽  
Decay Heat ◽  
Plant Dynamics ◽  
Fuel Storage ◽  
Cooling Coils ◽  
Cooling Loops

The prototype fast breeder reactor “Monju” has an ex-vessel fuel storage system (EVSS) which consists mainly of an ex-vessel fuel storage tank (EVST) and an EVST sodium cooling system. EVST uses natural circulation of sodium for decay heat removal. Natural circulation in the EVST is generated by the decay heat from the spent fuel assemblies and the cooling of the cooling coils installed in the EVST. The EVST sodium cooling system consists of three independent loops. In each loop, sodium is circulated by electromagnetic pumps and the heat is removed by an air cooler with blowers. This system has the ability to remove the maximum decay heat using two loops, and thus, it uses two of the three loops for normal operation. During a station blackout (SBO), the pumps and blowers are stopped. However, the three air coolers are installed about 13.5 m higher than the cooling coils, and therefore, the EVST sodium cooling system potentially retains some cooling ability because of natural circulation. In this study, an analysis and evaluation of the plant dynamics for the spent fuel and the EVSS structural integrity during an SBO were performed. The ultimate heat sink for the EVST sodium cooling system is the atmosphere, and the air coolers have an exhaust stack for efficient natural circulation caused by the chimney effect. However, the EVST sodium cooling system loses pressure and the heat transfer characteristics change if the flow rate is low. It was, therefore, necessary to confirm the temperature and flow rate behavior of EVSS in this analysis. In the present calculations, the plant dynamics analysis program “Super-COPD” was used. The factors affecting the cooling ability were investigated and analytical cases were determined. In one case, the two operated loops were switched to natural circulation after an SBO. The number of cooling loops was then changed from two to three by having an operator open the vane and dampers of the standby loop. In this case, sodium temperature in the EVST increased to approximately 320°C. When the number of cooling loops was not changed and natural circulation occurred in only two loops, the sodium temperature in the EVST increased to approximately 450°C. In both cases, however, the structural integrity of the EVSS was maintained. These analytical results, therefore, help clarify the number of necessary cooling loops for efficient decay heat removal and sodium temperature behavior in an SBO.

Analysis of a Convection Loop for GFR Post-LOCA Decay Heat Removal

2004 ◽  
Author(s):  
Wesley C. Williams ◽  
Pavel Hejzlar ◽  
Pradip Saha
Keyword(s):  
Heat Transfer ◽  
Steady State ◽  
Flow Regime ◽  
Natural Circulation ◽  
Heat Removal ◽  
Computer Code ◽  
Steady State Analysis ◽  
Decay Heat ◽  
Cooling Loops ◽  
State Analysis

A computer code (LOCA-COLA) has been developed at MIT for steady state analysis of convective heat transfer loops. In this work, it is used to investigate an external convection loop for decay heat removal of a post-LOCA GFR. The major finding is that natural circulation cooling of the GFR is feasible under certain circumstances. Both helium and CO2 cooled system components are found to operate in the mixed convection regime, the effects of which are noticeable as heat transfer enhancement or degradation. It is found that CO2 outperforms helium under identical natural circulation conditions. Decay heat removal is found to have a quadratic dependence on pressure in the laminar flow regime and linear dependence in the turbulent flow regime. Other parametric studies have been performed as well. In conclusion, convection cooling loops are a credible means for GFR decay heat removal and LOCA-COLA is an effective tool for steady state analysis of cooling loops.

scholarly journals CFD Analysis of the ESFR Reactor Pit Cooling System in Case of Sodium Leakage

2021 ◽  
Author(s):  
Aleksander Grah ◽  
Haileyesus Tsige-Tamirat ◽  
Joel Guidez ◽  
Antoine Gerschenfeld ◽  
Konstantin Mikityuk ◽  
...  
Keyword(s):  
Vessel Wall ◽  
Cooling System ◽  
Heat Removal ◽  
Strong Line ◽  
Cooling Systems ◽  
Decay Heat ◽  
Cooling Loops ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

Abstract The Decay Heat Removal System (DHRS) for the ESFR Concept consists of three cooling systems, which provide highly reliable, redundant and diversified decay heat removal function. Two of the systems provide strong line of defense, whereas the third system provides a weak line of defense. This third DHR system, DHRS-3, involves separate oil and water cooling loops integrated in the reactor pit, which is installed instead of the safety vessel. It is hoped that the proposed DHR concept enables a robust demonstration of the practical elimination. For its confirmation, detailed numerical analysis is needed as a basis for further investigation. Supporting this approach, the current CFD computation provides a preliminary thermal analysis of the capability of the oil cooling system in the reactor to be used for residual heat removal pit in case of an emergency. For the evaluation, different heat flux values are assumed at the vessel wall to examine the range of the resulting temperatures. The temperature of the main vessel wall should remain below 800°C. Furthermore, a sodium leakage at 500°C into the reactor pit is assumed. The concrete structure should remain below 70°C.

Emergency Decay Heat Removal in a GEN-IV Gas-Cooled Fast Reactor

2006 ◽  
Author(s):  
Lap Y. Cheng ◽  
Hans Ludewig ◽  
Jae Jo
Keyword(s):  
Fast Reactor ◽  
Cooling System ◽  
Natural Circulation ◽  
Heat Removal ◽  
Forced Flow ◽  
Decay Heat ◽  
Conversion Unit ◽  
Station Blackout ◽  
Combination Approach ◽  
Gen Iv

A series of transient analyses using the system code RELAP5-3d has been performed to confirm the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400MWt GEN-IV gas-cooled fast reactor. The accident sequence of interest is a station blackout simultaneous with a small break (10 sq.inch/0.645m2) in the reactor vessel. The analyses cover the three phases of decay heat removal in a depressurization accident: (1) forced flow cooling by the power conversion unit (PCU) coast down, (2) active forced flow cooling by a battery powered blower, and (3) passive cooling by natural circulation. The blower is part of an emergency cooling system (ECS) that by design is to sustain passive decay heat removal via natural circulation cooling 24 hours after shutdown. The RELAP5 model includes the helium-cooled reactor, the ECS (primary and secondary side), the PCU with all the rotating machinery (turbine and compressors) and the heat transfer components (recuperator, pre-cooler and inter-cooler), and the guard containment that surrounds the reactor and the PCU. The transient analysis has demonstrated the effectiveness of passive decay heat removal by natural circulation cooling when the guard containment pressure is maintained at or above 800kPa.

The Choose of Decay Heat Removal Systems of SFR

2018 ◽  
Author(s):  
Nina Yue ◽  
Rong Cai ◽  
Yun Wang ◽  
Suizheng Qiu ◽  
Dalin Zhang
Keyword(s):  
Fast Reactor ◽  
Large Scale ◽  
Cooling System ◽  
Natural Circulation ◽  
Heat Removal ◽  
Decay Heat ◽  
Station Blackout ◽  
The One ◽  
Safety Systems ◽  
Heat Removal System

A sodium-cooled fast reactor is a significant candidate for future power reactor systems. Decay heat removal is an essential function of reactor safety systems The decay heat removal system should have the capacity to remove the decay heat with natural circulation in any accident. There are three types of decay heat removal systems, namely direct reactor auxiliary cooling system, primary reactor auxiliary cooling system, and intermediate reactor auxiliary cooling system. The one dimensional systems analysis code THACS was applied to conduct transient analyses of a sodium-cooled fast reactor, and the capabilities of three types of decay heat removal systems against a station blackout accident were compared. The results indicate that these three types of decay heat removal systems can remove the residual heat effectively. For large-scale sodium-cooled fast reactor, the capabilities of primary reactor auxiliary cooling system and intermediate reactor auxiliary cooling system were better, because the cold sodium from the penetrating heat exchanger in these two auxiliary cooling systems could directly flow into the core assemblies.

scholarly journals Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

2008 ◽  
Vol 2008 ◽  
pp. 1-11 ◽  
Author(s):  
Avinash J. Gaikwad ◽  
P. K. Vijayan ◽  
Sharad Bhartya ◽  
Kannan Iyer ◽  
Rajesh Kumar ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Transport ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Heat Removal ◽  
Transport Systems ◽  
Critical Examination ◽  
Decay Heat ◽  
Station Blackout

Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT), SGs, and PDHRs) under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections). On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

Performance of Sodium Natural Circulation Loop for Passvie Decay Heat Removal

2008 ◽  
Author(s):  
Hae-Yong Jeong ◽  
Kwi-Seok Ha ◽  
Won-Pyo Chang ◽  
Yong-Bum Lee ◽  
Dohee Hahn ◽  
...  
Keyword(s):  
Fast Reactor ◽  
System Analysis ◽  
Natural Circulation ◽  
Heat Removal ◽  
Circulation Loop ◽  
Ambient Atmosphere ◽  
Steady State Condition ◽  
Decay Heat ◽  
Natural Circulation Loop ◽  
Test Loop

The Korea Atomic Energy Research Institute (KAERI) is developing a Generation IV sodium-cooled fast reactor design equipped with a passive decay heat removal circuit (PDRC), which is a unique safety system in the design. The performance of the PDRC system is quite important for the safety in a simple system transient and also in an accident condition. In those situations, the heat generated in the core is transported to the ambient atmosphere by natural circulation of the PDRC loop. It is essential to investigate the performance of its heat removal capability through experiments for various operational conditions. Before the main experiments, KAERI is performing numerical studies for an evaluation of the performance of the PDRC system. First, the formation of a stable natural circulation is numerically simulated in a sodium test loop. Further, the performance of its heat removal at a steady state condition and at a transient condition is evaluated with the real design configuration in the KALIMER-600. The MARS-LMR code, which is developed for the system analysis of a liquid metal-cooled fast reactor, is applied to the analysis. In the present study, it is validated that the performance of natural circulation loop is enough to achieve the required passive heat removal for the PDRC. The most optimized modeling methodology is also searched for using various modeling approaches.

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