Core Design Studies of a Fast Reactor and Accelerator-Driven System for Two-Stage Fast Spectrum Fuel Cycle Option

Lead-bismuth eutectic (LBE) cooled fast reactor, one of the six types of reactors in Gen-IV, has very good inherent safety and significant advantages in reducing and burning nuclear wastes, enhancing economy. Also LBE cooled accelerator driven system (ADS) has been a very innovative and potential waste burner. COBRA-EN is a mature, stable and widely-used sub-channel analysis code for light water cooled reactor but it couldn’t be applied in Pb-Bi-cooled reactor directly. Some modifications were made for COBRA-EN in the present work, then the code was named COBRA-PB and was suitable for the sub-channel analysis of Pb-Bi-cooled reactor. The modified code was verified and validated with CFX and experimental results. There was a good agreement between the two results. Then sub-channel analysis of Pb-Bi-cooled reactor was done with the modified code.

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The knowledge management on the design of a generation IV sodium fast reactor project at CEA. The case and methodology applied on the Astrid project

EPJ Nuclear Sciences & Technologies ◽

10.1051/epjn/2020016 ◽

2020 ◽

Vol 6 ◽

pp. 53

Author(s):

Gilles Rodriguez ◽

Philippe Amphoux ◽

David Plancq ◽

Edwige Richebois ◽

Frédéric Varaine ◽

...

Keyword(s):

Knowledge Management ◽

Fast Reactor ◽

Fuel Cycle ◽

Project Team ◽

Basic Design ◽

Design Phase ◽

Design Studies ◽

Know How ◽

Sodium Fast Reactor ◽

One Year

From 2010 to 2019, the French Alternative Energies and Atomic Commission (CEA) associated with industrial partners realized the Basic Design of a prototype Sodium Fast Reactor. This project was called ASTRID (ASTRID for Advanced Sodium Technological Reactor for Industrial Demonstration). ASTRID design studies were financed through governmental funds until the end of the basic design. These funds covered also the design studies for the core manufacturing workshop, the refurbishment or construction of large test loops. One year before the term of this Basic Design phase (in 2018), industrial partners, CEA and the French State conducted a review of fast neutrons reactors and fuel cycle strategy. The review which is now translated into the Multiannual Energy Program concluded that the perspective of industrial deployment of Fast Reactors is more distant. Yet it has been concluded to keep this option open, requiring to maintain competences, and to progress on technological barriers and further develop know-how. The strategy for complete closure of nuclear fuel cycle is maintained as a long-term sustainability objective (in the second half of the 21st century). Therefore, as a direct consequence of this decision, the ASTRID project stopped at the end of 2019 at its Basic Design phase. Quickly the question raised on the Knowledge Management (KM) and Know-How capitalization of the huge amount of studies and results realized during ten years (around 23 000 technical documents). Moreover the challenge was to realize this KM process in less than one year, before the ASTRID project team definitive split. The paper is presenting an innovative KM methodology which has been created and specifically performed on the ASTRID project. It is based on a series of interviews and video recordings, all transformed into some New KM tools called “MOOK” (MOOK for Management of Organized Online Knowledge). All these MOOKs considered as “data rich contents” are then inter-connected and linked by the ASTRID Product Breakdown Structure to some fundamental documents, for a comprehensive and quick mapping of the project. They finally form an efficient KM tool recorded in a PLM Software (PLM for Product Lifecycle Management). Thus the ASTRID project team has realized a high level and easy-to-use “GPS” (Global Positioning System) tool to keep the ASTRID history, context, knowledge and know-how for years. This KM methodology can be easily adapted to other nuclear projects and needs.

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Neutronics Design of a Small LBE Cooled Accelerator Driven System CIADS

Volume 2: Plant Systems, Construction, Structures and Components; Next Generation Reactors and Advanced Reactors ◽

10.1115/icone21-15994 ◽

2013 ◽

Cited By ~ 1

Author(s):

Xinzhe Wang ◽

Hong Yu ◽

Jian Zhang ◽

Shixi Wang ◽

Yun Hu ◽

...

Keyword(s):

High Energy Physics ◽

Fuel Cycle ◽

Thermal Power ◽

Reactor Core ◽

High Energy ◽

Physical Parameters ◽

Design Work ◽

Driven System ◽

Energy Domain ◽

Accelerator Driven System

As an important part of advanced fuel cycle R&D, conceptual study of accelerator driven system (ADS) in China started since 1995. In 2000, China Institute of Atomic Energy (CIAE), Institute of High Energy Physics (IHEP) and other institutes started a ten-year project aiming at ADS fundamental R&D on physics and related technologies, which is one item of “Key Project of Chinese National Program for Fundamental Research and Development (973 Program)” in energy domain. In order to get a better understanding of ADS neutronics characteristic, China Fast Reactor Research Center initiates a preliminary R&D program focused on neutronics design of a small lead-bismuth eutectic cooled ADS with fast spectrum. In this program, the reactor core of a 10MW thermal power ADS called CIADS (China Initiative ADS) with MOX fuel has been studied and designed. For generally concerning, CIADS can operate in either subcritical or critical mode. Different parameters, such as target size and position, position that transmutation assemblies are placed have been studied during the design work. Results show that a half size target and one zone loading can meet the needs for a small size ADS. Moreover, some important physical parameters of CIADS, such as keff, ks, power peak factor and neutron maximum flux density are evaluated. According to the R&D work, it’s appropriate to set the ks of CIADS at 0.96∼0.98.

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Current Status of the Development of the Fast Reactor Fuel Cycle in India

Revue Générale Nucléaire ◽

10.1051/rgn/20096120 ◽

2009 ◽

pp. 120-126

Author(s):

K.V. Govindan Kutty ◽

P.R. Vasudeva Rao ◽

Baldev Raj

Keyword(s):

Fast Reactor ◽

Fuel Cycle ◽

Reactor Fuel ◽

Current Status ◽

Fast Reactor Fuel

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Development of Safety Analysis Code System of Beam Transport and Core for Accelerator Driven System

SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo ◽

10.1051/snamc/201405202 ◽

2014 ◽

Author(s):

Naoto Aizawa ◽

Tomohiko Iwasaki

Keyword(s):

Safety Analysis ◽

Code System ◽

Beam Transport ◽

Driven System ◽

Accelerator Driven System

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Analysis of the Accelerator-Driven System Fuel Assembly during the Steam Generator Tube Rupture Accident

Materials ◽

10.3390/ma14081818 ◽

2021 ◽

Vol 14 (8) ◽

pp. 1818

Author(s):

Di-Si Wang ◽

Bo Liu ◽

Sheng Yang ◽

Bin Xi ◽

Long Gu ◽

...

Keyword(s):

Fuel Assembly ◽

Steam Generator ◽

Metal Flow ◽

Maximum Temperature ◽

Steam Generator Tube ◽

Driven System ◽

Fuel Assemblies ◽

Accelerator Driven System ◽

The Impact ◽

Steam Bubble

China is developing an ADS (Accelerator-Driven System) research device named the China initiative accelerator-driven system (CiADS). When performing a safety analysis of this new proposed design, the core behavior during the steam generator tube rupture (SGTR) accident has to be investigated. The purpose of our research in this paper is to investigate the impact from different heating conditions and inlet steam contents on steam bubble and coolant temperature distributions in ADS fuel assemblies during a postulated SGTR accident by performing necessary computational fluid dynamics (CFD) simulations. In this research, the open source CFD calculation software OpenFOAM, together with the two-phase VOF (Volume of Fluid) model were used to simulate the steam bubble behavior in heavy liquid metal flow. The model was validated with experimental results published in the open literature. Based on our simulation results, it can be noticed that steam bubbles will accumulate at the periphery region of fuel assemblies, and the maximum temperature in fuel assembly will not overwhelm its working limit during the postulated SGTR accident when the steam content at assembly inlet is less than 15%.

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