Improvement of neutron startup source handling work by developing new transportation container for high-temperature engineering test reactor (HTTR)

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.

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THERMAL STUDIES ABOUT THE HIGH TEMPERATURE ENGINEERING TEST REACTOR

10.26678/abcm.encit2016.cit2016-0693 ◽

2016 ◽

Author(s):

Maria Elizabeth Scari ◽

Antonella Lombardi Costa

Keyword(s):

High Temperature ◽

Thermal Studies ◽

Test Reactor

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

Volume 5: Innovative Nuclear Power Plant Design and New Technology Application; Student Paper Competition ◽

10.1115/icone22-30219 ◽

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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