Licensing Considerations of a Fluoride Salt Cooled High Temperature Test Reactor

2013 ◽  
Author(s):  
Yao Xiao ◽  
Lin-wen Hu ◽  
Charles Forsberg ◽  
Suizheng Qiu ◽  
Guanghui Su
Keyword(s):  
High Temperature ◽  
Coolant Temperature ◽  
Outlet Temperature ◽  
Fluoride Salt ◽  
Applied Physics ◽  
Reference Case ◽  
Triso Fuel ◽  
Regulatory Guidelines ◽  
Test Reactor ◽  
Temperature Test

The Fluoride salt cooled High temperature Reactor (FHR) is an innovative reactor concept that uses high temperature TRISO fuel with a low-pressure liquid salt coolant. Design of Fluoride salt cooled High temperature Test Reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and United States. An FHTR based on pebble bed core design with coolant temperature 600–700 °C is being planned for construction by the Shanghai Institute of Applied Physics (SINAP). This paper provides a preliminary thermal hydraulic licensing analysis of an FHTR using SINAP’s pebble core design as a reference case. The operation limits based on criteria outlined in U.S. regulatory guidelines are evaluated. Limiting Safety System Settings (LSSS) considering uncertainties for forced convection operation are obtained. The LSSS power and coolant outlet temperature are 24.6 MW and 720 °C, respectively.

Development of a Thermal-Hydraulic Analysis Code and Transient Analysis for a FHTR

2014 ◽  
Author(s):  
Yao Xiao ◽  
Lin-wen Hu ◽  
Suizheng Qiu ◽  
Dalin Zhang ◽  
Guanghui Su ◽  
...  
Keyword(s):  
High Temperature ◽  
The United States ◽  
Coolant Temperature ◽  
Transfer Model ◽  
Fluoride Salt ◽  
Applied Physics ◽  
Reference Case ◽  
Triso Fuel ◽  
Test Reactor ◽  
Thorium Molten Salt Reactor

The Fluoride-salt-cooled High-temperature Reactor (FHR) is an advanced reactor concept that uses high temperature TRISO fuel with a low-pressure liquid salt coolant. Design of Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and the United States. An FHTR based on pebble bed core design with coolant temperature 600–700 °C is being planned for construction by the Chinese Academy of Sciences (CAS)’s Thorium Molten Salt Reactor (TMSR) Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides preliminary thermal hydraulic transient analyses of an FHTR using SINAP’s pebble core design as a reference case. A point kinetic model is calculated by developing a microcomputer code coupling with a simplified porous medium heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the FHTR by simulating basic transient conditions including the unprotected loss of flow, unprotected overcooling, and unprotected transient overpower accidents. The results show that the SINAP’s pebble core design is an inherently safe reactor design.

Development of a Thermal-Hydraulic Analysis Code and Transient Analysis for a Fluoride-Salt-Cooled High-Temperature Test Reactor

2015 ◽  
Vol 1 (1) ◽  
Author(s):  
Yao Xiao ◽  
Lin-Wen Hu ◽  
Suizheng Qiu ◽  
Dalin Zhang ◽  
Su Guanghui ◽  
...  
Keyword(s):  
High Temperature ◽  
Computer Code ◽  
The United States ◽  
Coolant Temperature ◽  
Transfer Model ◽  
Fluoride Salt ◽  
Pebble Bed ◽  
Reference Case ◽  
Test Reactor ◽  
Temperature Test

The fluoride-salt-cooled high-temperature reactor (FHR) is an advanced reactor concept that uses high-temperature tristructural isotropic (TRISO) fuel with a low-pressure liquid salt coolant. Design of the fluoride-salt-cooled high-temperature test reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and the United States. An FHTR based on pebble-bed core design with a coolant temperature of 600–700°C is being planned for construction by the Chinese Academy of Sciences’ (CAS) Thorium Molten Salt Reactor (TMSR) Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides preliminary thermal-hydraulic transient analyses of an FHTR using SINAP’s pebble-bed core design as a reference case. A point kinetic model is implemented using computer code by coupling with a simplified porous medium heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the FHTR by simulating several transient conditions including the unprotected loss of flow, unprotected overcooling, and unprotected transient overpower accidents. The results show that SINAP’s pebble-bed core is a very safe reactor design.

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.

Analysis of the Limiting Safety System Settings of a Fluoride Salt–Cooled High-Temperature Test Reactor

2014 ◽  
Vol 187 (3) ◽  
pp. 221-234 ◽  
Author(s):  
Yao Xiao ◽  
Lin-Wen Hu ◽  
Charles Forsberg ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
...  
Keyword(s):  
High Temperature ◽  
Safety System ◽  
Fluoride Salt ◽  
Test Reactor ◽  
Temperature Test

Development of High Temperature Gas-Cooled Reactor Fuel for Extended Burnup

2010 ◽  
Author(s):  
Shohei Ueta ◽  
Jun Aihara ◽  
Masaki Honda ◽  
Noboru Furihata ◽  
Kazuhiro Sawa
Keyword(s):  
High Temperature ◽  
Coating Layer ◽  
Fission Products ◽  
Pyrolytic Carbon ◽  
Principal Function ◽  
Energy Agency ◽  
Triso Fuel ◽  
Coating Layers ◽  
Test Reactor ◽  
Fuel Particles

Current HTGRs such as the High Temperature Engineering Test Reactor (HTTR) of Japan Atomic Energy Agency (JAEA) use Tri-Isotropic (TRISO)-coated fuel particles with diameter of around 1 mm. TRISO fuel consists of a micro spherical kernel of oxide or oxycarbide fuel and coating layers of porous pyrolytic carbon (buffer), inner dense pyrolytic carbon (IPyC), silicon carbide (SiC) and outer dense pyrolytic carbon (OPyC). The principal function of these coating layers is to retain fission products within the particle. Particularly, the SiC coating layer acts as a barrier against the diffusive release of metallic fission products and provides mechanical strength for the particle [1].

Loss of Core Cooling Test With One Cooling Line Inactive in Vessel Cooling System of High-Temperature Engineering Test Reactor

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
Yusuke Fujiwara ◽  
Takahiro Nemoto ◽  
Daisuke Tochio ◽  
Masanori Shinohara ◽  
Masato Ono ◽  
...  
Keyword(s):  
Control System ◽  
High Temperature ◽  
Cooling System ◽  
Thermal Power ◽  
Reactor Power ◽  
Coolant Temperature ◽  
Water Cooling ◽  
Test Reactor ◽  
Core Cooling ◽  
Cooling Function

In the high-temperature engineering test reactor (HTTR), the vessel cooling system (VCS) which is arranged around the reactor pressure vessel (RPV) removes residual heat and decay heat from the reactor core when the forced core cooling is lost. The test of loss of forced cooling (LOFC) when one of two cooling lines in VCS lost its cooling function was carried out to simulate the partial loss of cooling function from the surface of RPV using the HTTR at the reactor thermal power of 9 MW, under the condition that the reactor power control system and the reactor inlet coolant temperature control system were isolated, and three helium gas circulators (HGCs) in the primary cooling system (PCS) were stopped. The test results showed that the reactor power immediately decreased to almost zero, which is caused by negative feedback effect of reactivity, and became stable as soon as HGCs were stopped. On the other hand, the temperature changes of permanent reflector block, RPV, and the biological shielding concrete were quite slow during the test. The temperature decrease of RPV was several degrees during the test. The numerical result showed a good agreement with the test result of temperature rise of biological shielding concrete around 1 °C by the numerical method that uses a calibrated thermal resistance by using the measured temperatures of RPV and the air outside of biological shielding concrete. The temperature increase of water cooling tube panel of VCS was calculated to be about 15 °C which is sufficiently small in the view point of property protection. It was confirmed that the sufficient cooling capacity of VCS can be maintained even in case that one of two water cooling lines of VCS loses its function.

scholarly journals Deterministic Modeling of the High Temperature Test Reactor

2010 ◽  
Author(s):  
J. Ortensi ◽  
J. J. Cogliati ◽  
M. A. Pope ◽  
R. M. Ferrer ◽  
A. M. Ougouag
Keyword(s):  
High Temperature ◽  
Deterministic Modeling ◽  
Test Reactor ◽  
Temperature Test
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