Research activities on nuclear reactor physics and thermal-hydraulics in Japan after Fukushima-Daiichi accident

2017 ◽  
Vol 55 (6) ◽  
pp. 575-598 ◽  
Author(s):  
Shuichiro Miwa ◽  
Yasunori Yamamoto ◽  
Go Chiba
Keyword(s):  
Nuclear Reactor ◽  
Thermal Hydraulics ◽  
Reactor Physics ◽  
Research Activities ◽  
Fukushima Daiichi

scholarly journals CTF and FLOCAL Thermal Hydraulics Validations and Verifications within a Multiscale and Multiphysics Software Development

Energies ◽  
2021 ◽  
Vol 14 (5) ◽  
pp. 1220
Author(s):  
Sebastian Davies ◽  
Ulrich Rohde ◽  
Dzianis Litskevich ◽  
Bruno Merk ◽  
Paul Bryce ◽  
...  
Keyword(s):  
Software Development ◽  
Turbulent Mixing ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Journal Article ◽  
Thermal Hydraulics ◽  
Reactor Physics ◽  
Wide Range ◽  
The Uk ◽  
Simulation Codes

Simulation codes allow one to reduce the high conservativism in nuclear reactor design improving the reliability and sustainability associated with nuclear power. Full-core coupled reactor physics at the rod level are not provided by most simulation codes. This has led in the UK to the development of a multiscale and multiphysics software development focused on LWRS. In terms of the thermal hydraulics, simulation codes suitable for this multiscale and multiphysics software development include the subchannel code CTF and the thermal hydraulics module FLOCAL of the nodal code DYN3D. In this journal article, CTF and FLOCAL thermal hydraulics validations and verifications within the multiscale and multiphysics software development have been performed to evaluate the accuracy and methodology available to obtain thermal hydraulics at the rod level in both simulation codes. These validations and verifications have proved that CTF is a highly accurate subchannel code for thermal hydraulics. In addition, these verifications have proved that CTF provides a wide range of crossflow and turbulent mixing methods, while FLOCAL in general provides the simplified no-crossflow method as the rest of the methods were only tested during its implementation into DYN3D.

scholarly journals SIMULATOR ASSISTED ENGINEERING – APPLICATIONS IN NUCLEAR ENGINEERING EDUCATION AT KHALIFA UNIVERSITY

2021 ◽  
Vol 247 ◽  
pp. 14003
Author(s):  
Ahmed K. Alkaabi ◽  
Mohamed Ali ◽  
Ho Joon Yoon ◽  
Oussama Ashy
Keyword(s):  
Student Performance ◽  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Operating Conditions ◽  
Nuclear Engineering ◽  
Thermal Hydraulics ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Plant Behavior ◽  
Control Functions

The Generic Pressurized Water Reactor (GPWR) simulator has been used in the Nuclear I&C Laboratory at Khalifa University (KU) since 2013 to improve student performance in nuclear engineering that is a multidisciplinary field involving nuclear reactor physics, thermodynamics, fluid mechanics, thermal hydraulics, radiation, etc. The simulator, developed by Western Service Corporation, has been integrated as a teaching and educational tool in different Engineering Programs at KU (Mechanical and Nuclear engineering). This lab is used in an undergraduate course where students apply the knowledge taught from different courses such as nuclear systems, fuel cycle, thermal hydraulics, safety principle, and control functions through a virtual operating NPP simulator. This real-time, full scope and high fidelity simulator allows to perform different operating conditions such as plant startups, shutdowns, and load maneuvers; as well as normal and abnormal plant transients, and critical scenarios and accidents. Since its installation in the Nuclear I&C Laboratory at KU in 2013, thirty students have benefited from this learning simulator. The main skills and learning outcomes expected to be achieved by students through the using of this tool are (i) ability to describe different NPP components and understand different process occurring in different subsystems, (ii) explain and apply safety principles and protective protocols, and (iii) analyze and interpret the plant behavior during transient operations and when severe accidents happen.

Full Implicit Integrate Solution to the Coupled Neutronics/Thermal-Hydraulics Problems

2013 ◽  
Author(s):  
Han Zhang ◽  
Fu Li
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Thermal Hydraulics ◽  
Field Equations ◽  
Krylov Method ◽  
Tightly Coupled ◽  
Physical Insight ◽  
Second Order Convergence ◽  
The Individual ◽  
Integrate Solution

The traditional solution of the coupled neutronics/ thermal-hydraulics problems has typically been performed by solving the individual field separately and then transferring information between each other. In this paper, full implicit integrate solution to the coupled neutronics/ thermal-hydraulic problem is investigated. There are two advantages compared with the traditional method, which are high temporal accuracy and stability. The five equations of single-phase flow, the solid heat conduction and the neutronics are employed as a simplified model of a nuclear reactor core. All these field equations are solved together in a tightly coupled, nonlinear fashion. Firstly, Newton-based method is employed to solve nonlinear systems due to its local second-order convergence rate. And then the Krylov iterative method is used to solve the linear systems which are from the Newton linearization. The two procedures above are the so-called Newton-Krylov method. Furthermore, in order to improve the performance of the Krylov method, physics-based preconditioner is employed, which is constructed by the physical insight. Finally, several Newton-Krylov solution approaches are carried out to compare the performance of the coupled neutronics / thermal-hydraulic equations.

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Water Hammer ◽  
Compression Process ◽  
Flammable Gas ◽  
Piping Systems ◽  
Fukushima Daiichi ◽  
Accident Conditions

Water hammers, or fluid transients, compress flammable gasses to their autognition temperatures in piping systems to cause fires or explosions. While this statement may be true for many industrial systems, the focus of this research are reactor coolant water systems (RCW) in nuclear power plants, which generate flammable gasses during normal operations and during accident conditions, such as loss of coolant accidents (LOCA’s) or reactor meltdowns. When combustion occurs, the gas will either burn (deflagrate) or explode, depending on the system geometry and the quantity of the flammable gas and oxygen. If there is sufficient oxygen inside the pipe during the compression process, an explosion can ignite immediately. If there is insufficient oxygen to initiate combustion inside the pipe, the flammable gas can only ignite if released to air, an oxygen rich environment. This presentation considers the fundamentals of gas compression and causes of ignition in nuclear reactor systems. In addition to these ignition mechanisms, specific applications are briefly considered. Those applications include a hydrogen fire following the Three Mile Island meltdown, hydrogen explosions following Fukushima Daiichi explosions, and on-going fires and explosions in U.S nuclear power plants. Novel conclusions are presented here as follows. 1. A hydrogen fire was ignited by water hammer at Three Mile Island. 2. Hydrogen explosions were ignited by water hammer at Fukushima Daiichi. 3. Piping damages in U.S. commercial nuclear reactor systems have occurred since reactors were first built. These damages were not caused by water hammer alone, but were caused by water hammer compression of flammable hydrogen and resultant deflagration or detonation inside of the piping.

Coupled Analysis of Thermal Hydraulics and Neutronics for a Molten Salt Reactor

2017 ◽  
Author(s):  
Jian Ge ◽  
Dalin Zhang ◽  
Wenxi Tian ◽  
Suizheng Qiu ◽  
G. H. Su
Keyword(s):  
Molten Salt ◽  
Nuclear Reactor ◽  
Coupled Analysis ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
Delayed Neutron ◽  
Analysis Model ◽  
Liquid Salt ◽  
Energy Equations ◽  
Volume Method

As one of the six selected optional innovative nuclear reactor in the generation IV International Forum (GIF), the Molten Salt Reactor (MSR) adopts liquid salt as nuclear fuel and coolant, which makes the characteristics of thermal hydraulics and neutronics strongly intertwined. Coupling analysis of neutronics and thermal hydraulics has received considerable attention in recent years. In this paper, a new coupling method is introduced based on the Finite Volume Method (FVM), which is widely used in the Computational Fluid Dynamics (CFD) methodology. Neutron diffusion equations and delayed neutron precursors balance equations are discretized and solved by the commercial CFD package FLUENT, along with continuity, momentum and energy equations simultaneously. A Temporal And Spatial Neutronics Analysis Model (TASNAM) is developed using the User Defined Functions (UDF) and User Defined Scalar (UDS) in FLUENT. A neutronics benchmark is adopted to demonstrate the solution capability for neutronics problems using the method above. Furthermore, a steady state coupled analysis of neutronics and thermal hydraulics for the Molten Salt Advanced Reactor Transmuter (MOSART) is performed. Two groups of neutrons and six groups of delayed neutron precursors are adopted. Distributions of the liquid salt velocity, temperature, neutron flux and delayed neutron precursors in the core are obtained and analyzed. This work can provide some valuable information for the design and research of MSRs.

Overview on Fusion Nuclear Technology Experimental Testing

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Slavomir Entler ◽  
Jan Kysela
Keyword(s):  
Eutectic Alloy ◽  
Nuclear Reactor ◽  
Experimental Testing ◽  
High Heat ◽  
Neutron Field ◽  
Research Centre ◽  
Nuclear Technology ◽  
High Heat Flux ◽  
Research Activities ◽  
Experimental Facilities

Research Centre Rez in the Czech Republic has carried out a number of research and development activities on the nuclear technology of the fusion reactor International Thermonuclear Experimental Reactor (ITER). These contributions have led to the development of numerous experimental facilities. The initial experimental research related to ITER was focused on the technology of the LiPb eutectic alloy, and a production unit and technological channel were constructed. At a later time, material tests were performed in the neutron field of the LVR-15 research nuclear reactor. Interactions of EUROFER 97 and the LiPb eutectic alloy were examined in in-pile and out-pile tests, and the technology of the LiPb was developed. First wall mock-ups were in-pile and out-pile tested under high heat flux (HHF) cycle loads. At present, a full-size mock-up of the ITER Test Blanket System (TBS) and an HHF testing complex are constructed. This paper provides an overview of the research activities and experimental facilities.

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