Seismic Response of the High-Temperature Engineering Test Reactor Core Bottom Structure

1992 ◽  
Vol 99 (2) ◽  
pp. 169-176 ◽  
Author(s):  
Tatsuo Iyoku ◽  
Yoshiyuki Inagaki ◽  
Shusaku Shiozawa ◽  
Masatoshi Futakawa ◽  
Toshiyo Miki
Keyword(s):  
High Temperature ◽  
Seismic Response ◽  
Reactor Core ◽  
Test Reactor

Seismic Study of High-Temperature Engineering Test Reactor Core Graphite Structures

1992 ◽  
Vol 99 (2) ◽  
pp. 158-168 ◽  
Author(s):  
Tatsuo Iyoku ◽  
Yoshiyuki Inagaki ◽  
Shusaku Shiozawa ◽  
Isoharu Nishiguchi
Keyword(s):  
High Temperature ◽  
Reactor Core ◽  
Test Reactor

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.

Improvement of Exchanging Method of Neutron Startup Source of High Temperature Engineering Test Reactor

2016 ◽  
Author(s):  
Hiroaki Sawahata ◽  
Yosuke Shimazaki ◽  
Etsuo Ishitsuka ◽  
Kazunori Yamazaki ◽  
Yoshinori Yanagida ◽  
...  
Keyword(s):  
High Temperature ◽  
Radiation Dose ◽  
Neutron Source ◽  
Reactor Core ◽  
Handling Performance ◽  
Technical Issues ◽  
Neutron Leakage ◽  
Poly Ethylene ◽  
Test Reactor ◽  
Transportation Container

In the HTTR, 252Cf is loaded in the reactor core as a neutron startup source and changed at the frequency. In this exchange work, there were two technical issues; slightly higher radiation exposure of workers by neutron leakage and reliability of neutron source transportation container in handling. To reduce the radiation dose by the neutron leakage, detailed numerical evaluations using PHITS code were carried out, the effective shielding method for fuel handling machine was proposed. Easily removable poly-ethylene blocks and particles were used as the neutron shieling, and installed in the cooling paths of the fuel handling machine. As a result, the collective effective dose by neutron was reduced from about 700man-μSv to about 300man-μSv. As to the neutron source transportation container, the handling performance was improved and the handling work was safely accomplished by downsizing.

Thermal Hydraulic Studies of a Fluoride Salt Cooled High Temperature Test Reactor With Different CFD Methods

2014 ◽  
Author(s):  
Chenglong Wang ◽  
Yao Xiao ◽  
Jianjun Zhou ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
...  
Keyword(s):  
High Temperature ◽  
Channel Model ◽  
Single Channel ◽  
Fluoride Salt ◽  
Applied Physics ◽  
Local Flow ◽  
Single Channel Analysis ◽  
Test Reactor ◽  
Safety Operation ◽  
Temperature Test

The Fluoride-salt-cooled High temperature Reactor (FHR) is new reactor concept-about a decade old which is mainly on going in China and U.S. The preliminary thermal-hydraulic studies of the Fluoride salt cooled High temperature Test Reactor (FHTR) is necessary for the development of the FHR technology. In this paper, the thermal-hydraulics of FHTR (also called TMSR-SF) designed by Shanghai Instituted of Applied Physics (SINAP) is studied in different power modes. The temperature distributions of the coolant and the fuel pebble are obtained using a steady-state thermal-hydraulic analysis code for FHR. The comprehensive local flow and heat transfer are investigated by computational fluid dynamics (CFD) for the locations where may have the maximum pebble temperature based on the results from single channel analysis. The profiles of temperature, velocity, pressure and Nu of the coolant on the surface of the pebble as well as the temperature distribution of a fuel pebble are obtained and analyzed. Numerical results showed that the results of 3-D simulation are in reasonable agreement with that of single channel model and also illustrated safety operation of the preliminary designed TMSR-SF in different power mode.

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