Modeling and Experiments of a Passive Decay Heat Removal System for Advanced Nuclear Reactors

2017 ◽  
Author(s):  
Andrea Bersano ◽  
Mario De Salve ◽  
Cristina Bertani ◽  
Nicolò Falcone ◽  
Bruno Panella
Keyword(s):  
Scaling Laws ◽  
Nuclear Reactors ◽  
Natural Circulation ◽  
Heat Removal ◽  
Experimental Results ◽  
Passive Systems ◽  
Decay Heat ◽  
Safety Systems ◽  
Heat Removal System ◽  
Removal System

Within the field of research and development of innovative nuclear reactors, in particular Generation IV reactors and Small Modular Reactors (SMR), the design and the improvement of safety systems play a crucial role. Among all the safety systems high attention is dedicated to passive systems that do not need external energy to operate, with a very high reliability also in the case of station blackout, and which are largely used in evolutionary technology reactors. The aim of this work is the experimental and numerical analysis of a passive system that operates in natural circulation in order to study the mechanism and the efficiency of heat removal. The final goal is the development of a methodology that can be used to study this class of systems and to assess the thermal-hydraulic code RELAP5 for these specific applications. Starting from a commercial size system, which is the decay heat removal system of the experimental lead cooled reactor ALFRED, an experimental facility has been designed, built and tested with the aim of studying natural circulation in passive systems for nuclear applications. The facility has been simulated and optimized using the thermal-hydraulic code RELAP5-3D. During the experimental tests, temperatures and pressures are measured and the experimental results are compared with the ones predicted by the code. The results show that the system operates effectively, removing the given thermal power. The code can predict well the experimental results but high attention must be dedicated to the modeling of components where non-condensable gases are present (condenser pool and surrounding ambient). This facility will be also used to validate the scaling laws among systems that operate in natural circulation.

S-CO2 Turbine Design for Decay Heat Removal System of Sodium Cooled Fast Reactor

2016 ◽  
Author(s):  
Seong Kuk Cho ◽  
Jekyoung Lee ◽  
Jeong Ik Lee ◽  
Jae Eun Cha
Keyword(s):  
Experimental Data ◽  
Fast Reactor ◽  
Design Methodology ◽  
Nuclear Reactors ◽  
Heat Removal ◽  
Design Code ◽  
Decay Heat ◽  
Turbine Design ◽  
Heat Removal System ◽  
Removal System

A Sodium-cooled Fast Reactor (SFR) has receiving attention as one of the promising next generation nuclear reactors because it can recycle the spent nuclear fuel produced from the current commercial nuclear reactors and accomplish higher thermal efficiency than the current commercial nuclear reactors. However, after shutdown of the nuclear reactor core, the accumulated fission products of the SFR also decay and release heat via radiation within the reactor. To remove this residual heat, a decay heat removal system (DHRS) with supercritical CO2 (S-CO2) as the working fluid is suggested with a turbocharger system which achieves passive operational capability. However, for designing this system an improved S-CO2 turbine design methodology should be suggested because the existing methodology for designing the S-CO2 Brayton cycle has focused only on the compressor design near the critical point. To develop a S-CO2 turbine design methodology, the non-dimensional number based design and the 1D mean line design method were modified and suggested. The design methodology was implemented into the developed code and the code results were compared with existing turbine experimental data. The data were collected under air and S-CO2 environment. The developed code in this research showed a reasonable agreement with the experimental data. Finally using the design code, the turbocharger design for the suggested DHRS and prediction of the off design performance were carried out. As further works, more effort will be put it to expand the S-CO2 turbine test data for validating the design code and methodology.

Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor

2015 ◽  
Vol 52 (9) ◽  
pp. 1102-1121 ◽  
Author(s):  
Osamu Watanabe ◽  
Kazuhiro Oyama ◽  
Junji Endo ◽  
Norihiro Doda ◽  
Ayako Ono ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Evaluation Methodology ◽  
Decay Heat ◽  
Heat Removal System ◽  
Removal System

scholarly journals CFD Analysis of the Passive Decay Heat Removal System of an LBE-Cooled Fast Reactor

2018 ◽  
Vol 2018 ◽  
pp. 1-11
Author(s):  
Jiarun Mao ◽  
Lei Song ◽  
Yuhao Liu ◽  
Jiming Lin ◽  
Shanfang Huang ◽  
...  
Keyword(s):  
Heat Exchange ◽  
Fast Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Engineering Project ◽  
Decay Heat ◽  
The Core ◽  
Main Vessel ◽  
Heat Removal System ◽  
Removal System

This paper presents capacity of the passive decay heat removal system (DHRS) operated under the natural circulation conditions to remove decay heat inside the main vessel of the Lead-bismuth eutectic cooled Fast Reactor (LFR). The motivation of this research is to improve the inherent safety of the LFR based on the China Accelerator Driven System (ADS) engineering project. Usually the plant is damaged due to the failure of the main pumps and the main heat exchangers under the Station Blackout (SBO). To prevent this accident, we proposed the DHRS based on the diathermic oil cooling for the LFR. The behavior of the DHRS and the plant was simulated using the CFD code STAR CCM+ using LFR with DHRS. The purpose of this analysis is to evaluate the heat exchange capacity of the DHRS and is to provide the reference for structural improvement and experimental design. The results show that the stable natural circulations are established in both the main vessel and the DHRS. During the decay process, the heat exchange power is above the core decay heat power. In addition, in-core decay heat and heat storage inside the main vessel are efficiently removed. All the thermal-hydraulics parameters are within a safe range. Moreover, the highest temperature occurs at the upper surface of the core. A swirl occurs at the corner of the lateral core surface and some improvements should be considered. And the natural circulation driving force can be further increased by reducing the loop resistance or increasing the natural circulation height based on the present design scenario to enhance the heat exchange effect.

scholarly journals An experimental study on natural circulation decay heat removal system for a loop type fast reactor

2016 ◽  
Vol 53 (9) ◽  
pp. 1385-1396 ◽  
Author(s):  
Ayako Ono ◽  
Hideki Kamide ◽  
Jun Kobayashi ◽  
Norihiro Doda ◽  
Osamu Watanabe
Keyword(s):  
Experimental Study ◽  
Fast Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Decay Heat ◽  
Heat Removal System ◽  
Removal System

scholarly journals Experimental Research on Passive Residual Heat Removal System of Chinese Advanced PWR

2014 ◽  
Vol 2014 ◽  
pp. 1-8 ◽  
Author(s):  
Zhuo Wenbin ◽  
Huang Yanping ◽  
Xiao Zejun ◽  
Peng Chuanxin ◽  
Lu Sansan
Keyword(s):  
Steady State ◽  
Scaling Laws ◽  
Heat Removal ◽  
Experimental Results ◽  
Test Facility ◽  
Semiempirical Model ◽  
Pressurized Water ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

Passive residual heat removal system (PRHRS) for the secondary loop is one of the important features for Chinese advance pressurized water reactor (CAPWR). To prove the safety characteristics of CAPWR, serials of experiments have been done on special designed PRHRS test facility in the former stage. The test facility was built up following the scaling laws to preserve the similarity to CAPWR. A total of more than 300 tests have been performed on the test facility, including 90% steady state cases and 10% transient cases. A semiempirical model was generated for passive heat removal functions based on the experimental results of steady state cases. The dynamic capability characteristics and reliability of passive safety system for CAPWR were evidently proved by transient cases. A new simulation code, MISAP2.0, has been developed and calibrated by experimental results. It will be applied in future design evaluation and optimization works.

Preliminary Safety Analysis of the IRIS Reactor

2002 ◽  
Author(s):  
M. E. Ricotti ◽  
A. Cammi ◽  
A. Cioncolini ◽  
A. Cipollaro ◽  
F. Oriolo ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Feed Water ◽  
Deterministic Analysis ◽  
Loss Of Coolant Accident ◽  
Safety Features ◽  
Safety Systems ◽  
Heat Removal System ◽  
Reactor Pressure ◽  
Removal System

A deterministic analysis of the IRIS safety features has been carried out by means of the best-estimate code RELAP (ver. RELAP5 mod3.2). First, the main system components were modeled and tested separately, namely: the Reactor Pressure Vessel (RPV), the modular helical-coil Steam Generators (SG) and the Passive (natural circulation) Emergency Heat Removal System (PEHRS). Then, a preliminary set of accident transients for the whole primary and safety systems was investigated. Since the project was in a conceptual phase, the reported analyses must be considered preliminary. In fact, neither the reactor components, nor the safety systems and the reactor signal logics were completely defined at that time. Three “conventional” design basis accidents have been preliminary evaluated: a Loss Of primary Flow Accident, a Loss Of Coolant Accident and a Loss Of Feed Water accident. The results show the effectiveness of the safety systems also in LOCA conditions; the core remains covered for the required grace period. This provides the basis to move forward to the preliminary design.

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.

Study on a Small Power Reactor With Compact Pressure Vessel and Natural Circulation

2014 ◽  
Author(s):  
Kenya Takiwaki ◽  
Shungo Sakurai ◽  
Yutaka Takeuchi ◽  
Yasushi Yamamoto
Keyword(s):  
Power Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Cooling Water ◽  
Heat Removal ◽  
Small Power ◽  
Control Rod ◽  
Decay Heat ◽  
Heat Removal System ◽  
Removal System

There is movement which is developing the small reactor for the small electricity grid in place of a big power reactor which requires the high capital cost. This paper introduces a small power reactor whose purpose is to achieve high economic competitiveness and advanced safety. In order to attain high economic competitiveness, it is designed to be small and simple and uses natural circulation and high pressure. A steam generator is integrated into the reactor pressure vessel (RPV), thus dispensing with a primary system and preventing radiation leakage from the reactor core. The small core is designed to have a high power density (100 MW/m3, almost twice that of a conventional boiling water reactor). The concept of a 300 MWt (100 MWe) core design is established by introducing a boiling heat transfer system. By boiling cooling water, the cooling-water circulating flow quantity in a reactor core is enlarged. By increasing a flow, the minimum critical power ratio is improved, which is an important core characteristic. Furthermore, using a burnable poison (Gd2O3), the excess reactivity of a reactor core is reduced and excess reactivity is controlled only by the control rod. Moreover, the maximum linear power density is improved and the critical power ratio is minimized by optimizing the burnable poison arrangement and the control rod pattern. In order to attain high safety, our small reactor has an advanced decay heat removal system that can cool the core without external support. This decay heat removal system is part of the secondary cooling system and combined with a cooling tower. As a result, the quantity of cooling water stored in the decay heat removal system is reduced, and longtime decay heat removal is possible by small equipment.

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