Results of the AGR-2 TRISO fuel performance demonstration irradiation experiment in the Advanced Test Reactor

2021 ◽  
Vol 150 ◽  
pp. 107833
Author(s):  
Paul A. Demkowicz ◽  
Blaise P. Collin ◽  
David A. Petti ◽  
Grant L. Hawkes ◽  
James W. Sterbentz ◽  
...  
Author(s):  
Shohei Ueta ◽  
Jun Aihara ◽  
Masaki Honda ◽  
Noboru Furihata ◽  
Kazuhiro Sawa

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].


Author(s):  
Jian Li ◽  
Ding She ◽  
Lei Shi ◽  
Jing Zhao

Tristructural isotropic (TRISO) fuel particles are chosen as the major fuel type of High temperature gas cooled reactor (HTGR). The TRISO coated particle also acts as the first barrier for radioactivity retention. The performance of the TRISO coated particle has a significant influence on the safety of HTGR. A set of fuel performance analysis codes have been developed during the past decades. The main functions of these codes are conducting stress calculation and failure probability prediction. PANAMA is a widely used German version fuel performance analysis code, which simulates the mechanical performance of TRISO coated particle under normal and accident conditions. In this code, only a simple pressure vessel model is considered, which is insufficient in stress analysis and fuel failure rate prediction. Nowadays, efforts have been done to update the fuel performance model utilized in PANAMA code, and a new TRISO fuel performance analysis code, FFAT, is under developed. This paper describes the newly updated TRISO fuel performance model and presents some first results based on the updated model.


2010 ◽  
Vol 405 (1) ◽  
pp. 74-82 ◽  
Author(s):  
Jeffrey J. Powers ◽  
Brian D. Wirth

Author(s):  
Yao Xiao ◽  
Lin-wen Hu ◽  
Suizheng Qiu ◽  
Dalin Zhang ◽  
Guanghui Su ◽  
...  

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.


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