Summary of Tritium Source Term Study in 10 MW High Temperature Gas-Cooled Test Reactor

2020 ◽  
Vol 76 (4) ◽  
pp. 513-525
Author(s):  
X. Liu ◽  
W. Peng ◽  
F. Xie ◽  
J. Cao ◽  
Y. Dong ◽  
...  
Keyword(s):  
High Temperature ◽  
Source Term ◽  
Term Study ◽  
Test Reactor ◽  
High Temperature Gas

Characteristic Tests on Transition Core of HTR-10

2018 ◽  
Author(s):  
Liqiang Wei ◽  
Dongmei Ding ◽  
Ling Liu ◽  
Yucheng Wang ◽  
Xiaoming Chen ◽  
...  
Keyword(s):  
High Temperature ◽  
Control Rod ◽  
Safety Requirements ◽  
Operation Characteristic ◽  
Power Operation ◽  
Test Reactor ◽  
High Temperature Gas

After a long-term shutdown, the 10MW high temperature gas-cooled test reactor (HTR-10) was restarted, and the operation & safety characteristics of the HTR-10 transition core are tested and verified. A series of the characteristic tests have been implemented, such as the value calibrating test of the control rod and boron absorber ball, the disturbance characteristic of helium circulator, the start-stop characteristic and the stable power operation characteristic, which indicated the characteristics of the reactor transition core meet the design and safety requirements.

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.

Study on Simulation, Control and Online Assistance Integrated System of 10MW High Temperature Gas-Cooled Test Reactor

2004 ◽  
Author(s):  
Shaojie Luo ◽  
Lei Shi ◽  
Shutang Zhu
Keyword(s):  
Control System ◽  
High Temperature ◽  
System Design ◽  
Reactor Core ◽  
Control Flow ◽  
Integrated System ◽  
Control System Design ◽  
Simulation Control ◽  
Test Reactor ◽  
High Temperature Gas

In order to provide a convenient tool for engineering designed, safety analysis, operator training and control system design of the high temperature gas-cooled test reactor (HTR), an integrated system for simulation, control and online assistance of the HTR-10 has been designed and is still under development by the Institute of Nuclear Energy Technology (INET) of Tsinghua University in China. The whole system is based on a network environment and includes three subsystems: the simulation subsystem (SIMUSUB), the visualized control designed subsystem (VCDSUB) and the online assistance subsystem (OASUB). The SIMUSUB consists of four parts: the simulation calculating server (SCS), the main control client (MCC), the data disposal client (DDC) and the results graphic display client (RGDC), all of which can communicate each other via network. The SIMUSUB is intended to analyze and calculate the physical processes of the reactor core, the main loop system and the stream generator, etc., as well as to simulate the normal operation and transient accidents, and the result data can be graphically displayed through the RGDC dynamically. The VCDSUB provides a platform for control system modeling where the control flow systems can be automatically generated and graphically simulated. Based on the data from the field bus, the OASUB provides some of the reactor core parameter, which are difficult to measure. This whole system can be used as an educational tool to understand the design and operational characteristics of the HTR-10, and can also provide online supports for operators in the main control room, or as a convenient powerful tool for the control system design.

Development of the Prediction Technology of Cable Disconnection of In-Core Neutron Detector for the Future High-Temperature Gas-Cooled Reactors

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
Yosuke Shimazaki ◽  
Hiroaki Sawahata ◽  
Taiki Kawamoto ◽  
Hisashi Suzuki ◽  
Masanori Shinohara ◽  
...  
Keyword(s):  
High Temperature ◽  
Neutron Detector ◽  
Time Domain Reflectometry ◽  
Technical Issue ◽  
Electrical Method ◽  
Inspection Method ◽  
Wide Range ◽  
Test Reactor ◽  
The Time Domain ◽  
High Temperature Gas

Maintenance technologies for the reactor system have been developed by using the High Temperature engineering Test Reactor (HTTR). One of the important purposes of development is to accumulate the experiences and data to satisfy the availability of operation up to 90% by shortening the duration of the periodical maintenance for future high-temperature gas-cooled reactors (HTGRs) by shifting from time-based maintenance to condition-based maintenance. The technical issue of the maintenance of the in-core neutron detector, wide range monitor (WRM), is to predict the malfunction caused by cable disconnection to plan the replacement schedule. This is because it is difficult to observe directly inside the WRM in detail. The electrical inspection method was proposed to detect and predict the cable disconnection of the WRM by remote monitoring from outside the reactor using the time-domain reflectometry (TDR). The disconnection position, which was specified by the electrical method, was identified by nondestructive and destructive inspection. The accumulated data are expected to be contributed for advanced maintenance of future HTGRs.

Guidance for Developing Fuel Design Limit of High Temperature Gas-Cooled Reactor

2020 ◽  
Author(s):  
Hiroyuki Sato ◽  
Takeshi Aoki ◽  
Hirofumi Ohashi
Keyword(s):  
High Temperature ◽  
Development Process ◽  
Evaluation Criteria ◽  
Research Association ◽  
Safety Requirements ◽  
Fuel Design ◽  
Physical Barriers ◽  
Test Reactor ◽  
And Control ◽  
High Temperature Gas

Abstract The present study aims to propose a guidance that facilitates to determine fuel design limits of commercial HTGR on the basis of licensing experience through the high temperature engineering test reactor (HTTR) construction. The guidance consists of a set of figure of merits (FOMs) and a process to determine their evaluation criteria. The FOMs are firstly identified to satisfy safety requirements and a basic concept of safety guides established in a special committee under the Atomic Energy Society of Japan with the support of the Research Association of High Temperature Gas Cooled Reactor Plant. The development process for the evaluation criteria takes into account not only the top-level regulatory criteria but also design dependent constraints including the performance of fission product containment in physical barriers other than fuel, fuel qualification criteria, design specifications of an instrumentation and control system. As a result, a comprehensive and transparent procedure for designers of prismatic-type commercial HTGR has been developed. This paper provides details on the development procedure for fuel design limit. methods to derive the limits are also presented.

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 Improvement of Analysis Technology for High Temperature Gas-Cooled Reactor by Using Data Obtained in High Temperature Engineering Test Reactor

2008 ◽  
Vol 2 (1) ◽  
pp. 83-91
Author(s):  
Shigeaki NAKAGAWA ◽  
Daisuke TOCHIO ◽  
Kuniyoshi TAKAMATSU ◽  
Minoru GOTO ◽  
Tetsuaki TAKEDA
Keyword(s):  
High Temperature ◽  
Using Data ◽  
Test Reactor ◽  
High Temperature Gas
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