Sodium Cooled Fast Bread Reactor HCDA Codes Development

2017 ◽  
Author(s):  
Shi Tai ◽  
Zhang Dong-hui ◽  
Hu Wen-jun
Keyword(s):  
Liquid Metal ◽  
Fast Reactor ◽  
Power Distribution ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Radioactive Material ◽  
Program Model ◽  
Primary Method ◽  
The World ◽  
Gen Iv

Liquid metal fast reactor is one of the Gen IV nuclear power, it is necessary to analyze hypothetical core disruptive accident (HCDA) of FBR to ensure that the system can prevent the radioactive material from leaking out. The modified Bethe-Tait model is the primary method to analyze hypothetical core disruptive accident in the world. In order to better analyze the nuclear reactor hypothetical core disruptive accident in China, an improved B-T model is used. At present, on the basis of the improved B-T model, power distribution of the CEFR add to the progress. The results of comparison between the program and SUREX program in France show that the program model can simulate the nuclear reactor hypothetical core disruptive accident in China.

Heat Transfer Correlation for Supercritical Carbon Dioxide Flowing in Vertical Bare Tubes

2013 ◽  
Author(s):  
Sahil Gupta ◽  
Eugene Saltanov ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Error Analysis ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Large Set ◽  
Transfer Coefficients ◽  
Data Points ◽  
Institute Of Technology ◽  
Using Data ◽  
Gen Iv

Canada among many other countries is in pursuit of developing next generation (Generation IV) nuclear-reactor concepts. One of the main objectives of Generation-IV concepts is to achieve high thermal efficiencies (45–50%). It has been proposed to make use of SuperCritical Fluids (SCFs) as the heat-transfer medium in such Gen IV reactor design concepts such as SuperCritical Water-cooled Reactor (SCWR). An important aspect towards development of SCF applications in novel Gen IV Nuclear Power Plant (NPP) designs is to understand the thermodynamic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. To calculate forced convection HTCs for simple geometries, a number of empirical 1-D correlations have been proposed using dimensional analysis. These 1-D HTC correlations are developed by applying data-fitting techniques to a model equation with dimensionless terms and can be used for rudimentary calculations. Using similar statistical techniques three correlations were proposed by Gupta et al. [1] for Heat Transfer (HT) in SCCO2. These SCCO2 correlations were developed at the University of Ontario Institute of Technology (Canada) by using a large set of experimental SCCO2 data (∼4,000 data-points) obtained at the Chalk River Laboratories (CRL) AECL. These correlations predict HTC values with an accuracy of ±30% and wall temperatures with an accuracy of ±20% for the analyzed dataset. Since these correlations were developed using data from a single source - CRL (AECL), they can be limited in their range of applicability. To investigate the tangible applicability of these SCCO2 correlations it was imperative to perform a thorough error analysis by checking their results against a set of independent SCCO2 tube data. In this paper SCCO2 data are compiled from various sources and within various experimental flow conditions. HTC and wall-temperature values for these data points are calculated using updated correlations presented in [1] and compared to the experimental values. Error analysis is then shown for these datasets to obtain a sense of the applicability of these updated SCCO2 correlations.

scholarly journals Thermo-hydraulic Simulation of AP1000 Nuclear Reactor Fuel Assembly

2021 ◽  
Vol 31 (1) ◽  
pp. 60-71
Author(s):  
Leonardo Acosta Martínez ◽  
Carlos Rafael García Hernández ◽  
Jesus Rosales García ◽  
Annie Ortiz Puentes
Keyword(s):  
Fuel Assembly ◽  
Power Distribution ◽  
Capital Investment ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Geometric Model ◽  
Modular Construction ◽  
Construction Time ◽  
Hydraulic Simulation ◽  
In Series

One of the challenges of future nuclear power is the development of safer and more efficient nuclear reactor designs. The AP1000 reactor based on the PWR concept of generation III + has several advantages, which can be summarized as: a modular construction, which facilitates its manufacture in series reducing the total construction time, simplification of the different systems, reduction of the initial capital investment and improvement of safety through the implementation of passive emergency systems. Being a novel design it is important to study the thermohydraulic behavior of the core applying the most modern tools. To determine the thermohydraulic behavior of a typical AP1000 fuel assembly, a computational model based on CFD was developed. A coupled neutronic-thermohydraulic calculation was performed, allowing to obtain the axial power distribution in the typical fuel assembly. The geometric model built used the certified dimensions for this type of installation that appear in the corresponding manuals. The thermohydraulic study used the CFD-based program ANSYS-CFX, considering an eighth of the fuel assembly. The neutronic calculation was performed with the program MCNPX version 2.6e. The work shows the results that illustrate the behavior of the temperature and the heat transfer in different zones of the fuel assembly. The results obtained agree with the data reported in the literature, which allowed the verification of the consistency of the developed model.

A Nodal Dynamic Model for Control System Simulation of the HTR-PM Reactor Core

2010 ◽  
Author(s):  
Zhe Dong ◽  
Xiaojin Huang ◽  
Liangju Zhang
Keyword(s):  
Control System ◽  
High Temperature ◽  
Power Distribution ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Pebble Bed ◽  
Nodal Model ◽  
High Temperature Gas

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.

Research Concept of Decommissioning Knowledge Management for the Fugen NPP

2018 ◽  
Author(s):  
Yasuyoshi Taruta ◽  
Satoshi Yanagihara ◽  
Yukihiro Iguchi ◽  
Koichi Kitamura ◽  
Masashi Tezuka ◽  
...  
Keyword(s):  
Knowledge Management ◽  
Semantic Web ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Good Method ◽  
Nuclear Facilities ◽  
Energy Agency ◽  
The World ◽  
Research Concept ◽  
Construction Operation

In 2002, the International Atomic Energy Agency (IAEA) mentioned the strengthening of nuclear knowledge, technology and application. This background has that there are aging of nuclear facility and nuclear power plant staffs. In addition, it would be difficult to succession a nuclear knowledge, technology, and skills. For example, undergraduate departments of nuclear energy and science are decreasing. The IAEA discussing those situations and pointed out the importance of a nuclear knowledge management. The nuclear knowledge management (NKM) is developing a database science as management on nuclear knowledge and information. In recent years, the IAEA has also advanced knowledge taxonomies on nuclear accidents as one of a nuclear knowledge management. In Japan, this achievements of nuclear knowledge taxonomy was using in the organization of information on accidents in Fukushima. A few studies are attempts to appropriately arrange and utilize huge amounts of information. Even in nuclear facilities in Japan, it is pointed out a veteran or expert staff retirement and loss of knowledge and skill caused by this retirement. This problem is common issue in the world. Then, we created a prototype database system to utilize past documentation of knowledge and information. The database made from semantic web technology. The semantic web is a method of preparing a frame of categorized knowledge and linking information related to it. The target is a nuclear reactor of ATR Fugen that is decommissioning from 2008. Until now, cases of decommissioning completion are 17 cases in the world. One case of JPDR in Japan. It is not enough to understand a good method of decommissioning. In general, the decommissioning project requires many information related to dismantling and decontamination. Particular, past information is important to know a past contamination situation and so on. This study focus on an access method for past data and information. However, we need to pay attention to other side of decommissioning project. Because of history of operating reactor has different tasks that are design, construction, operation and decommissioning. It is not appropriate to use the collected information as it is. For that reason, we will continue our research on the points pointed out above.

Power Distribution and Fuel Centerline Temperature in a Pressure-Tube Supercritical Water-Cooled Reactor (PT SCWR)

2012 ◽  
Author(s):  
W. Peiman ◽  
Eu. Saltanov ◽  
L. Grande ◽  
I. Pioro ◽  
B. Rouben ◽  
...  
Keyword(s):  
Heat Flux ◽  
Supercritical Water ◽  
Power Distribution ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Thermal Power ◽  
Pressure Tube ◽  
Thermochemical Processes ◽  
Flux Profile ◽  
Centerline Temperature

SuperCritical Water-cooled nuclear Reactor (SCWR) designs are one of six nuclear-reactor concepts being developed under the Generation IV International Forum (GIF) initiative. A generic pressure-tube SCWR consists of distributed fuel channels with coolant inlet and outlet temperatures of 350 and 625°C at 25 MPa, respectively. Such reactor coolant outlet conditions allow for high thermal efficiencies of SCW Nuclear Power Plant (NPP) of about 45–50%. In addition to high thermal efficiencies, SCWR designs provide the means for co-generation of hydrogen through thermochemical processes such as the Cu–Cl cycle. The main objective of this paper is to determine the power distribution inside the core of an SCWR by using a lattice code - DRAGON and a diffusion code - DONJON. As a result of these calculations, heat-flux profiles in all fuel channels were determined. Consequently, the heat-flux profile in a channel with the maximum thermal power was used as an input into a thermal-hydraulic code, which was developed in MATLAB in order to calculate a fuel centerline temperature for UO2 and UC nuclear fuels. Results of an analysis showed that the fuel centerline temperature of UC was significantly lower than that of UO2. This paper also studies effects of energy groups on multi-group diffusion calculations and proposes nine energy groups for further neutronic studies related to SCWRs.

Void Reactivity Effect Research on Liquid Metal Cooled Fast Reactor

2017 ◽  
Author(s):  
Yang Lyu ◽  
Xiao Liang
Keyword(s):  
Liquid Metal ◽  
Power Systems ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Reactor Core ◽  
Nuclear Industry ◽  
Fourth Generation ◽  
Reactivity Effect ◽  
The Core ◽  
Reactivity Coefficient

In the fourth generation of advanced nuclear power systems, the liquid metal cooled fast reactor plays a more and more important role, such as SFR, LFR and ADS system with LBE coolant. Void reactivity effect means bubbles produced in the core area will induce the change of reactivity. And this reactivity will affect the safety of the reactor. Through investigation and comparison of several liquid metal cooled fast reactors in the nuclear industry, this paper studies bubbles in different positions and partial voiding of the active zone inside the core and fuel assemblies with Monte Carlo core physics calculation method and then concludes the main influencing factors of void reactivity coefficient. The results can provide reference for the follow-up research and development of new type liquid metal fast reactor core design.

Determination of Specific Heat of Eutectic Indium – Bismuth-Tin Liquid Metal Alloys as a Test Material for Liquid Metal - Cooled Applications

2013 ◽  
Vol 420 ◽  
pp. 185-193 ◽  
Author(s):  
Adam Lipchitz ◽  
Glenn Harvel ◽  
Takeyoshi Sunagawa
Keyword(s):  
Liquid Metal ◽  
Thermodynamic Properties ◽  
Specific Heat ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Metal Flow ◽  
Weight Percent ◽  
Test Material ◽  
Plant Operations ◽  
High Level

Currently, Russia, India, China, France, South Korea, and Japan are actively pursuing liquid metal cooled applications such as liquid cooled metal nuclear reactor concepts. The liquid metal coolants being considered for these designs are sodium, lead and lead-bismuth eutectic; these designs utilize reactive and toxic materials at temperatures up to 1073 K for nuclear power plant operations and other similar applications. To simulate these systems with the actual coolant material requires a high level of safety systems. Use of these materials in university experimental laboratory settings is difficult due to the safety hazards and that lead (Pb) is a designated substance requiring special permission to use. Therefore, a less toxic and less reactive liquid metal that can be used to simulate liquid metal cooled flows will allow for a greater number of investigations and experimentation of liquid metal flow with regards to the field of thermal hydraulics. Good candidates for a liquid metal experimental fluid are alloys from the indium-bismuth-tin system such as Fields metal, which by weight percent is 51% indium, 32.5% bismuth and 16.5% tin and possesses a melting temperature of 333 K. However, the thermodynamic properties of Fields metal and similar alloys in their liquid state are not well described in literature. This work experimentally measures the specific heat of the eutectic alloys of theindium-bismuth-tin tertiary system using a differential scanning calorimeter technique and analyzes the results to determine if the thermodynamic properties of the system have sufficient scaling for experimental modeling applications. The results verify the melting temperatures of the alloys and establish a relationship between temperature and specific heat.

scholarly journals International and Indian Civil Liability Regime for Nuclear Damage - Operator's Liability v. Supplier's Liability

2011 ◽  
pp. 87-97
Author(s):  
Sidhant Chandalia
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Nuclear Energy ◽  
Nuclear Reactor ◽  
Electricity Supply ◽  
Catastrophic Failure ◽  
Nuclear Accidents ◽  
The World ◽  
The Cost ◽  
Nuclear Damage

Nuclear energy has seen tremendous growth in the last two decades and has a considerable share in world electricity supply. No nuclear reactor can be 100 % safe. Every reactor has a small, but finite chance of catastrophic failure, as seen in Chernobyl, Three Mile Island, Fukushima and many smaller accidents around the world, including those in India. Nuclear projects are non-bankable in the sense that they cannot be insured. If they could, the matter would be simple enough. The nuclear plant and every person likely to be affected by radiation would be insured for a suitable sum, but the cost of a disaster and the lawsuits that would ensue make it virtually impossible to insure a nuclear power plant. Hence, there is a need to put an artificial compensation and liability mechanism in place to deal with nuclear accidents. The issue is not merely the amount of compensation to be paid in the event of an accident but also who would be encumber with the bill, the operators or the suppliers, and to what extent.

Comparative Study of Basic Reactor Physics of the DFR Concept Using U-Pu and TRU Fuel Salts

2017 ◽  
Author(s):  
Xiang Wang ◽  
Rafael Macian-Juan
Keyword(s):  
Steady State ◽  
Liquid Metal ◽  
Molten Salt ◽  
Fast Reactor ◽  
Flux Distribution ◽  
Molten Salt Reactor ◽  
Reactor Physics ◽  
Fuel Salt ◽  
Gen Iv ◽  
First Time

The Dual Fluid Reactor (DFR) is a molten salt fast reactor developed by the IFK1 based on the Gen-IV Molten-Salt Reactor (MSR) and the Liquid-Metal Cooled Reactor (SFR, LFR) concepts. The analysis reported focuses on the comparison between previous neutronic calculations with the default fuel salt of U-Pu mixture and new ones with a transuranium (TRU) salt fuel option under steady state conditions. They include criticality, neutron spectrum, spatial flux distribution and temperature coefficient values. Fuel based on molten TRU salts has already been considered for the MSFR and other molten salt reactor designs. Therefore, the DFR for the first time has a comparable baseline with other molten salt reactors, so that its performance with TRU salt fuel can be assessed.

Experimental Activity for the Investigation of Mixing and Thermal Stratification Phenomena in the CIRCE Pool Facility

2016 ◽  
Author(s):  
D. Martelli ◽  
M. Tarantino ◽  
I. Di Piazza
Keyword(s):  
Fast Reactor ◽  
Thermal Stratification ◽  
Nuclear Reactor ◽  
Technology Development ◽  
Safety Issue ◽  
Natural Circulation ◽  
Fundamental Aspect ◽  
Horizon 2020 ◽  
Fuel Bundle ◽  
Gen Iv

Since the Lead-cooled Fast Reactor (LFR) has been conceptualized in the frame of GEN IV International Forum (GIF), ENEA is strongly involved in the HLM technology development. In particular, several experimental campaign employing HLM loop and pool facilities (CIRCE [1], NACIE [2], HELENA [3], HERO [4]) are carried out in order to support HLM technologies development. In this frame, suitable experiments were carried out on the CIRCE pool facility refurbished with the Integral Circulation Experiment (ICE) test section in order to investigate the thermal hydraulics and the heat transfer in grid spaced Fuel Pin Bundle cooled by liquid metal providing, among the others aim, experimental data in support of codes validation for the European fast reactor development. The study of thermal stratification in large pool reactor is relevant in the design of HLM nuclear reactor especially for safety issue. Thermal stratification should induce thermomechanical stresses on the structures and in accidental scenario conditions, could opposes to the establishment of natural circulation which is a fundamental aspect for the achievements of safety goals required in the GEN-IV roadmap. In the present work, a Protected Loss of Heat Sink with Loss Of Flow (PLOHS+LOF) scenario is experimentally simulated and the mixed convection with thermal stratification phenomena is investigated during the simulated transient, as foreseen in the frame of Horizon 2020 SESAME project [5]. A full characterization of thermal stratification inside the pool is presented, and the main results gained during the run are reported. The two tests named A (20 h) and B (6 h) here reported, essentially differs for the power supplied to the fuel bundle during the full power run (800 kW and 600 kW respectively). After the transition to natural circulation conditions, the power supplied to the bundle is decreased to about 30 kW simulating the decay heat. Finally the Nusselt number for the central subchannel of the fuel bundle simulator (FPS) is evaluated and compared with values obtained from Ushakov and Mikityuk correlations [6–7].

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