MAXIMISING DISCHARGE BURNUP IN AN OPEN CYCLE MOLTEN SALT REACTOR

This paper discusses work done to find an estimate of the maximum achievable discharge burnup in an open cycle molten salt reactor (MSR). An in-development deterministic code (WIMS11) is used to create a model of a simple generic MSR, and the methodology employed is discussed. Some experimentation is done with regards to the internal set-up of the ‘unit cells’ within the core, which shows there is a strong link between this geometry and the achievable burnup. Work is done to quantify the effects of removing volatile fission products and implementing a two-batch refuelling scheme. Finally, an optimization process is carried out whereby the optimal proportion of graphite moderator within the core is found which balances power across the regions while maximising discharge burnup. Two fuels are tested, one which carries only 235U and 238U, and another which also carries 232Th. It is found that the maximum achievable discharge burnup is approximately 155 MWd/kg, which is considerably higher than modern PWRs, despite a lower enrichment and only two batches of fuel being used.

Promising Neutron Irradiation Applications at the High Temperature Engineering Test Reactor

The high temperature gas-cooled reactor (HTGR) has advantages for irradiation applications such as large space available for irradiation at reflector region and high thermal neutron spectrum with the graphite moderator. High temperature engineering test reactor (HTTR), a prismatic type of the HTGR, has been constructed to establish and upgrade the basic technologies for the HTGRs. Many irradiation regions are reserved in the HTTR to be served as a potential tool for an irradiation test reactor in order to promote innovative basic researches such as materials, fusion reactor technology, and radiation chemistry. This study shows the overview of some possible irradiation applications at the HTTRs including neutron transmutation doping silicon (NTD-Si) and Iodine-125 (125I) productions. The HTTR has possibility to produce about 40 tons of doped Si-particles per year for fabrication of spherical silicon solar cell. Besides, the HTTR could also produce about 1.8 × 105 GBq/yr of 125I isotope, comparing to 3.0 × 103 GBq of total 125I supplied in Japan in 2016.

Characteristic Confirmation Test by Using HTTR and Investigation of Absorbing Thermal Load Fluctuation

The characteristic confirmation test has been demonstrating by using High Temperature engineering Test Reactor (HTTR). The nuclear heat supply performance test, which is one of the characteristic confirmation test is planned to be carried out after restarting of HTTR. Towards the realization of the industrial utilization of a High-Temperature Gas-cooled Reactor (HTGR) cogeneration system as an extension of a nuclear plant, it is important to ensure the reactor safety in the case that thermal-load of the heat application system is fluctuated or lost. The preliminary analysis for the thermal load fluctuation test, which is one of the nuclear heat supply performance test has been investigated. In the analysis, the reactor outlet temperature can continue to be stable against the reactor inlet temperature changing by the thermal fluctuation. It means that HTGR have the capability of absorbing the thermal fluctuation. This paper focuses on the investigation of the mechanism of absorbing the thermal fluctuation. With the reactor inlet temperature increasing, the graphite moderator reactivity keeps negative though the fuel reactivity becomes active. The large negative graphite moderator reactivity enhances the capability of the absorbing thermal fluctuation. In addition, in the middle of the core, the graphite moderator reactivity insertion trend is inverted. This trend is unique to HTGR because of the large temperature difference between top and bottom of HTGR core.

Determining Reactor Graphite Lifespan from Thermal Properties Degradation

This work assesses real graphite lifespan distribution of pressurized-tube reactors (RBMK) with graphite moderator. The changes in heat exchange and heat transfer conditions were taken into consideration, which changes are induced by the thermal properties’ degradation and the changes in the form of structural elements. Besides, some features associated with the differences in the neutron flux spectrum and differences in characteristics of gamma radiation field in the fuel and in the channels of control and protection system (CPS) were all taken into account.

Modelling Porosity in Quasi-Brittle Reactor Core Graphite

Some designs of nuclear reactors involve a graphite moderator within their core. Different forms of graphite have been adopted in the UK gas-cooled reactors but all have a complex structure of filler particles, matrix and pores. Changes occur in the graphite during service and in particular, porosity increases from that found in the virgin material. As part of a structural assessment, it is important to analyse the effects of this change in porosity. Software has been developed to represent the microstructure of pile grade A (PGA) and Gilsocarbon graphite with a range of porosities, to support finite element determination of material properties. The models are three dimensional geometric and voxel models based on the observed microstructures of these different graphites. Creating a sequence of model specimens with increasing porosities while holding other parameters constant, provides a representative microstructure to test the effect of increasing porosity on mechanical and physical properties.