scholarly journals Analysis of Fission Products’ Release in Pebble-Bed High-Temperature Gas-Cooled Reactor Fuel Elements Using a Modified FRESCO II Numerical Model

2021 ◽  
Vol 2021 ◽  
pp. 1-8
Author(s):  
Chao Fang ◽  
Chuan Li ◽  
Jianzhu Cao ◽  
Ke Liu ◽  
Sheng Fang
Keyword(s):  
High Temperature ◽  
Source Term ◽  
Fission Products ◽  
Release Rates ◽  
Time Step ◽  
The Difference ◽  
Advanced Nuclear System ◽  
Fuel Elements ◽  
Term Analysis ◽  
High Temperature Gas

The radiation safety design and emergency analysis of an advanced nuclear system highly depends on the source term analysis results. In modular high-temperature gas-cooled reactors (HTGRs), the release rates of fission products (FPs) from fuel elements are the key issue of source term analysis. The FRESCO-II code has been established as a useful tool to simulate the accumulation and transport behaviors of FPs for many years. However, it has been found that the mathematical method of this code is not comprehensive, resulting in large errors for short-lived nuclides and large time step during calculations. In this study, we used the original model of TRISO particles and spherical fuel elements and provided a new method to amend the FRESCO-II code. The results show that, for long-lived radionuclides (Cs-137), the two methods are perfectly consistent with each other, while in the case of short-lived radionuclides (Cs-138), the difference can be more than 1%. Furthermore, the matrix method is used to solve the final release rates of FPs from fuel elements. The improved analysis code can also be applied to the source term analysis of other HTGRs.

scholarly journals ICP-MS analysis of fission product diffusion in graphite for high-temperature gas-cooled reactors

2015 ◽  
Author(s):  
◽  
Lukas Michael Carter
Keyword(s):  
High Temperature ◽  
Fission Product ◽  
Diffusion Coefficients ◽  
Inductively Coupled Plasma ◽  
Nuclear Reactor ◽  
Fission Products ◽  
Release Rates ◽  
Inductively Coupled ◽  
Icp Ms ◽  
High Temperature Gas

High-temperature gas-cooled reactors (HTGRs) are one of the candidates being considered for the replacement of current nuclear reactor designs. Diffusion coefficients for fission products in HTGR graphite are required for estimation of fission product release rates from such reactors. We developed a method for analysis of fission product of fission product surrogate release rates from heated graphite samples. The graphite samples were infused with fission product surrogate material, and material which diffused from the graphite samples was transported via a carbon aerosol laden He jet system to an online inductively coupled plasma mass spectrometer for quantification of the release rate. Diffusion coefficients for cesium in IG-110 and NBG-18 grade nuclear graphites are reported.

scholarly journals Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

2017 ◽  
Vol 2017 ◽  
pp. 1-6 ◽  
Author(s):  
Xuegang Liu ◽  
Xin Huang ◽  
Feng Xie ◽  
Fuming Jia ◽  
Xiaogui Feng ◽  
...  
Keyword(s):  
High Temperature ◽  
Source Term ◽  
Pressurized Water Reactor ◽  
Sample Collection ◽  
Analytical Methodology ◽  
Pressurized Water ◽  
The Core ◽  
Irradiated Graphite ◽  
Term Analysis ◽  
High Temperature Gas

The high temperature gas-cooled reactor (HTGR) has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR) is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10) in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.

A Numerical Analysis on Graphite Dust Deposition and Resuspension in HTR-10 Steam Generator

2014 ◽  
Author(s):  
Wei Peng ◽  
Tian-qi Zhang ◽  
Ya-nan Zhen ◽  
Su-yuan Yu
Keyword(s):  
High Temperature ◽  
Steam Generator ◽  
Load Capacity ◽  
Fission Products ◽  
Dust Deposition ◽  
Preliminary Calculation ◽  
Transfer Tube ◽  
High Load Capacity ◽  
Fuel Elements ◽  
High Temperature Gas

The behavior of graphite dust is important to the safety analysis of High-Temperature Gas-cooled Reactor (HTGR). The fission products released by fuel elements would enter the primary loop and combine with dust, resulting in that the dust has a high load capacity of cesium, strontium, iodine and tritium. It would bring difficulty and inconvenience to the maintenance and repair of steam generator. Therefore, the behavior of graphite dust in the steam generator is essential to the safety of High Temperature Gas-cooled Reactors. The present study focused on the deposition and resuspension of graphite dust in steam generator of HTR by numerical method. The results show that the graphite dust in steam generator deposits on the surface of heat transfer tube through turbulent deposition, thermophoretic deposition, and other depositional mechanisms, of which thermophoretic deposition is the main mechanism for the particles with the diameter of 2.2μm in the present study. The preliminary calculation result shows that about 6760mg/m2 of graphite dust tends to load on the tube surface.

scholarly journals The R&D of HTR-STAC Program Package: Source Term Analysis Codes for Pebble-Bed High-Temperature Gas-Cooled Reactor

2018 ◽  
Vol 2018 ◽  
pp. 1-9 ◽  
Author(s):  
Hongyu Chen ◽  
Chuan Li ◽  
Haoyu Xing ◽  
Chao Fang
Keyword(s):  
High Temperature ◽  
Nuclear Reactor ◽  
Source Term ◽  
Safety Analysis ◽  
Operating Conditions ◽  
Primary Circuit ◽  
Pebble Bed ◽  
Water Reactors ◽  
Term Analysis ◽  
High Temperature Gas

Source term analysis is important in the design and safety analysis of advanced nuclear reactor and also provides a radiation safety analysis basis for Modular High-Temperature Gas-Cooled Reactor (HTR). High-Temperature Gas-Cooled Reactor-Pebble-bed Modules (HTR-PM) design by China is a typical Gen-IV and due to different safety concepts and systems, the implements of source term analysis in light water reactors are not entirely applicable to HTR-PM. To solve this problem, HTR-PM Source Term Analysis Code (HTR-STAC) has been developed and related V&V has been finished. HTR-STAC consists of five units, including LOOP (Primary Circuit Source Term Analysis Code), NORMAL (Normal Condition Airborne Source Term Analysis Code), ARCC (Accident Release Category Calculation code), CARBON (C-14 Source Term Analysis Code), and TRUM (Tritium Source Term Analysis Code). LOOP and NORMAL may be used as calculating primary circuit coolant radioactivity and the release of airborne radioactivity to the environment under normal operating conditions of HTR-PM, respectively. The code ARCC composed of several source term analysis programs in the different typical accidents scenario, including SGTR (Steam Generator Tube Rupture), LOCA (Loss of Coolant Accident), and the Transient Process, is compiled based on the results given by LOOP and NORMAL. CARBON and TRUM are developed to calculate the productions of C-14 and H-3 through a different mechanism. Furthermore, the V&V has been performed and show some positive results.

scholarly journals Study on the Implementation of Quality Assurance Aspect on High-Temperature Gas Cooled Reactor (HTGR)

2021 ◽  
Vol 2048 (1) ◽  
pp. 012022
Author(s):  
Sunarto ◽  
Sigit Santosa ◽  
Khusnul Khotimah ◽  
Sriyana
Keyword(s):  
Quality Assurance ◽  
High Temperature ◽  
Customer Requirements ◽  
Double Barrier ◽  
Quality Management Systems ◽  
Fuel Elements ◽  
Technological Analysis ◽  
Term Analysis ◽  
High Temperature Gas ◽  
Htgr Fuel

Abstract High-Temperature Gas Cooled Reactor (HTGR) Power Reactors have a layered safety system with the concept of a double barrier system. However, quality assurance is required to ensure the fulfillment of technological analysis weightings on power chamber materials, power ratings, fabrication components of High- Temperature Gas Cooled Reactor (HTGR) fuel elements, primary and secondary coolant pressures to meet customer requirements and be carried out continuously systematic and objective. This study analyzes the application of quality assurance, safety, security, the correctness of test/calibration results, increasing competitiveness, consumer protection and building trust (brand image) in the use of HTGR reactors to provide a reliable level of safety and security. The study method used is based on the literature review. The output of this study is the document of the HTGR reactor quality assurance systems to fulfill the IAEA-TECDOC-1645 requirements according to safety and standardization in frameworks design, material, fuel, and physical properties of the quality management systems. HTGR reactor has technical qualification, good performance of HTGR fuel, safety and accident analysis source term analysis, control of multi-modular HTGR and related human factor analysis, also optimizing radiation protection of HTGR

First-Principle Studies of Source Term Analysis in HTR-PM: From Framework to Preliminary Result

2016 ◽  
Author(s):  
Chao Fang ◽  
Chuan Li
Keyword(s):  
High Temperature ◽  
Binding Energy ◽  
Ab Initio ◽  
Adsorption Energy ◽  
Source Term ◽  
Fission Products ◽  
Primary Circuit ◽  
Face Centered Cubic ◽  
First Principle ◽  
Fuel Elements

High Temperature Reactor-Pebblebed Modules (HTR-PM) is a typical high-temperature gas cooled reactor (HTGR) [1]. In the HTR-PM, helium is used as the coolant to the primary circuit and the fission products released from fuel elements would be carried into circulation by helium [2]. When analyzing the source terms in HTR-PM, it is important and necessary to know the amount of nuclides adsorbed on the component materials of primary circuit [3] and furthermore, the detail mechanism of adsorption is also essential, which could not be obtained from traditional phenomenological analysis and conservative estimation. In order to solve this challenge, we established a framework with ab-initio methods. In this paper, the detail theory of ab-initio theory and the actual usage in the calculation of the adsorption energy, Fermi level, density of state and charge density difference are given firstly. And then, we show the calculated results of adsorption behaviors of radioactive fission products (Cs, Sr, Ag, I) on 2•1/4Cr1Mo and SiC, which are important structural materials for steam generator and coated particle of fuel elements for HTR-PM, respectively. It is found that Ag and I atoms prefer to be adsorbed at the square hollow site of the face-centered cubic iron cell with binding energy of about 1eV and 3eV respectively. By contrast, Cs and Sr atoms are not adsorbed on the surface of 2•1/4Cr1Mo. For the study of adsorption on SiC, it shows that all the four nuclides can be adsorbed on the surface of SiC with the binding energy of about 1∼3 eV. Finally, the adsorption rates of these nuclides are estimated by using the first-principle calculation results of adsorption energy. The adsorption rate can be used to determine the amount of adsorbed radioactive nuclides for nuclear safety evaluation of HTR-PM. These results can illustrate the micro pictures of the interaction of fission products and material, which is a new and useful way to analyze the source term in physical level.

Automatic X-ray inspection for escaped coated particles in spherical fuel elements of high temperature gas-cooled reactor

Energy ◽  
2014 ◽  
Vol 68 ◽  
pp. 385-398 ◽  
Author(s):  
Min Yang ◽  
Qi Liu ◽  
Hongsheng Zhao ◽  
Ziqiang Li ◽  
Bing Liu ◽  
...  
Keyword(s):  
High Temperature ◽  
Coated Particles ◽  
X Ray ◽  
Fuel Elements ◽  
High Temperature Gas
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