scholarly journals The Evaluation of the High Temperature Gas Cooled Reactor Safety to Fulfill the Requirement of the Next Generation Nuclear

2020 ◽  
Vol 21 (2) ◽  
pp. 71
Author(s):  
Julwan Hendry Purba, ST., M.App.IT. ◽  
Arya Adhyaksa Waskita ◽  
Damianus Toersiwi Sony Tjahyani
Keyword(s):  
High Temperature ◽  
Fission Products ◽  
Nuclear Fission ◽  
Nuclear Plant ◽  
Radioactive Material ◽  
Inherent Safety ◽  
Plant Design ◽  
Next Generation ◽  
Safety Features ◽  
High Temperature Gas

High temperature gas cooled reactor (HTGR) has been considered to be the most promising option to meet energy demands in the future. It has also been selected as the next generation nuclear plant. The primary safety requirement of the next generation nuclear plant design is to limit radioactive material releases to practically eliminate the need for public evacuation or sheltering beyond the exclusion area boundary. The purpose of this study is to evaluate the safety design of HTGRs in order to fulfill the requirement of the next generation nuclear plant. To achieve this objective, inherent safety features, fundamental safety functions, and confinement functions realized into the design of HTGRs are comprehensively evaluated. It is found that design provisions of HTGRs can fulfill the intention of keeping radionuclides at their original sources. The layers of the coated fuel particles are very robust to retain nuclear fission products for all foreseeable reactivity events. There will be no possibility of radioactive materials to be released even though related safety systems and operator intervention are not involved in the recovery actions. This design has complied with the requirement of the next generation nuclear plant, which is to practically eliminate the need for public evacuation or sheltering beyond the exclusion area boundary.Keywords: High temperature gas cooled reactor, inherent safety features, fundamental safety functions, confinement functions, next generation nuclear plant

A Brief Review of Models Representing Creep of Alloy 617

2005 ◽  
Author(s):  
Robert W. Swindeman ◽  
Michael J. Swindeman ◽  
Weiju Ren
Keyword(s):  
High Temperature ◽  
Nuclear Plant ◽  
Next Generation ◽  
Alloy 617 ◽  
Temperature Range ◽  
Damage Models ◽  
The Us ◽  
Creep Models ◽  
High Temperature Gas

Alloy 617 is being considered for the construction of components to operate in the Next Generation Nuclear Plant (NGNP). Service temperatures will range from 650 to 1000°C. To meet the needs of the conceptual designers of this plant, a materials handbook is being developed that will provide information on alloy 617, as well as other materials of interest. The database for alloy 617 to be incorporated into the handbook was produced in the 1970s and 1980s, while creep and damage models were developed from the database for use in the design of high-temperature gas-cooled reactors. In the work reported here, the US database and creep models are briefly reviewed. The work reported represents progress toward a useful model of the behavior of this material in the temperature range of 650 to 1000°C.

Study on Emergency Action Levels of the Modular High Temperature Gas-Cooled Reactor

2017 ◽  
Author(s):  
Xiaochuan Zang ◽  
Tao Liu
Keyword(s):  
High Temperature ◽  
Fission Product ◽  
Operating Mode ◽  
Inherent Safety ◽  
Pressurized Water ◽  
Plant Safety ◽  
Water Reactors ◽  
Safety Features ◽  
Action Levels ◽  
High Temperature Gas

The emergency action level (EAL) scheme for a modular high temperature gas-cooled reactor (HTR) plant refers to the generic EAL development guidance for pressurized water reactors (PWR) with HTR modification due to its design issues. Based on reactor’s accidents analysis and consequence assessment, EAL scheme of HTR is established through the steps of category and classification. Four emergency classes are set for HTR consisting of U (Emergency Standby), A (Facilities Emergency), S (Site Area Emergency) and G (General Emergency). The Recognition Category of Initiating Condition (IC) and EAL contains A - Abnormal Rad Levels / Radiological Effluent, F - Fission Product Barrier, H - Hazards and Other Conditions Affecting Plant Safety, S - System Malfunction. The methodology for development of EALs for HTR on Fission Product Barrier and System Malfunction has some differences from PWR’s due to differences on operating mode, inherent safety features and system characteristics.

Progress of the HTR-10 Measured Data Utilization

2014 ◽  
Author(s):  
Fubing Chen ◽  
Yujie Dong ◽  
Yanhua Zheng ◽  
Lei Shi ◽  
Fu Li ◽  
...  
Keyword(s):  
High Temperature ◽  
Measured Data ◽  
Inherent Safety ◽  
Data Utilization ◽  
Existing Problems ◽  
New Energy ◽  
Full Power ◽  
Power Operation ◽  
Safety Features ◽  
High Temperature Gas

The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10), which was designed, constructed and operated by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, is the first High Temperature Gas-cooled Reactor (HTGR) in China. Stepping into the commissioning phase in April, 2000, the HTR-10 attained the first criticality in December, 2000 and achieved its full power operation in January, 2003. Up to now, the HTR-10 has been successfully operated for more than ten years with different power levels. During the relatively long period of commissioning and operation, various kinds of tests were carried out on this reactor. Within the scope of this paper, the commissioning stage, the operation history and the test implementation of the HTR-10 are briefly summarized. At this stage, the HTR-10 is the only pebble bed HTGR under operation in the world, so the measured data from this reactor are extremely valuable for verifying the inherent safety features incorporated in small modular HTGRs as well as for testing the computer programs employed in the HTGR design process. With the purpose of ensuring the code credibility, validation work using the HTR-10 operation and test data has been performed for several years in INET. What is more, these data were partly shared with different countries through some collaborative research projects related to code development and assessment. In this paper, progress of the HTR-10 measured data utilization is reviewed. Meanwhile, existing problems observed from the code-to-test as well as code-to-code comparisons are pointed out. In addition, possible reasons of such problems are discussed in detail.

Design of the Gas Sampling and Analyzing System of HTR-PM

2017 ◽  
Author(s):  
Jiang Zhu ◽  
Feng Xie
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Fission Products ◽  
Structural Materials ◽  
Primary Coolant ◽  
Primary Loop ◽  
Gas Sampling ◽  
Purification System ◽  
Safety Features ◽  
High Temperature Gas

The high temperature gas-cooled reactor pebbled-bed module (HTR-PM) which is a modular high temperature gas-cooled reactor demonstration power plant, is characterized by inherent safety features and high generating efficiency. It adopts numerous graphite for structural materials in the reactor core, helium as primary coolant, and tristructural isotropic (TRISO) coated particles embedded in the graphite matrix as fuel elements. However, at high temperature the impurities in the helium can react with the graphite to cause corrosion of structural materials. Therefore, it is very necessary to monitor and control the composition and content of gaseous impurities in the primary coolant. In HTR-PM, the gas sampling and analyzing system has been designed to sample the primary helium at different positions in the helium purification system which is used to reduce the quantity of chemical impurities and remove the radioactive dust and gaseous fission products in the primary loop, and monitor the gas composition and individual concentration online. In the current paper, the composition of the gaseous impurities which need to be monitored in the primary loop of HTR-PM is presented, the design of the gas sampling positions in the helium purification system is discussed, and the main gas analyzing instruments are introduced.

Designing DOE’s NGNP: Don’t Forget High-Temperature Gas Reactor (HTGR) Experience

2010 ◽  
Author(s):  
J. K. August ◽  
J. J. Hunter
Keyword(s):  
High Temperature ◽  
Nuclear Plant ◽  
Next Generation ◽  
Hydrogen Economy ◽  
Design Features ◽  
Reactor Technology ◽  
Design Alternative ◽  
The U.S ◽  
High Temperature Gas ◽  
Fundamental Design

The U.S Department of Energy’s (DOE) Next Generation Nuclear Plant (NGNP) has three goals in its mission: deliver (1) electricity, (2) process steam and heat, and (3) generate hydrogen to support a hydrogen economy. DOE selected the high-temperature, helium-cooled graphite (moderated) reactor (HTGR) technology for its fundamental design. However, the NGNP faces many challenging design, licensing and cost hurdles. With huge risks and its associated expenses, the NGNP — a completely new reactor technology — requires an open-minded understanding, consideration and application of earlier HTGR design features. This paper suggests reconsidering some HTGR features in an updated Generation IV HTGR as a design alternative.

Method and Validation for Measurement of Effective Thermal Diffusivity and Conductivity of Pebble Bed in High Temperature Gas-Cooled Reactors

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Yongyong Wu ◽  
Cheng Ren ◽  
Rui Li ◽  
Xingtuan Yang ◽  
Jiyuan Tu ◽  
...  
Keyword(s):  
High Temperature ◽  
Thermal Diffusivity ◽  
Inverse Method ◽  
Experimental Result ◽  
Test Facility ◽  
Inherent Safety ◽  
Pebble Bed ◽  
Effective Thermal Diffusivity ◽  
Under Construction ◽  
High Temperature Gas

The effective thermal diffusivity and conductivity of pebble bed in the high temperature gas-cooled reactor (HTGR) are two vital parameters to determine the operating temperature and power in varisized reactors with the restriction of inherent safety. A high-temperature heat transfer test facility and its inverse method for processing experimental data are presented in this work. The effective thermal diffusivity as well as conductivity of pebble bed will be measured at temperature up to 1600 °C in the under-construction facility with the full-scale in radius. The inverse method gives a global optimal relationship between thermal diffusivity and temperature through those thermocouple values in the pebble bed facility, and the conductivity is obtained by conversion from diffusivity. Furthermore, the robustness and uncertainty analyses are also set forth here to illustrate the validity of the algorithm and the corresponding experiment. A brief experimental result of preliminary low-temperature test is also presented in this work.

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