In this paper is described the concept of a 20 MW microreactor cooled by molten salts with the registered trademark of Energy Well. Energy Well reactor is designed to supply either electricity or heat, in remote areas or into populated areas. The description includes context, main design features, safety systems, thermohydraulic analysis in RELAPS 3D, logistic, and transport of the reactor.

A new special structural nickel-based alloy for use in molten fluoride salt environments in molten salt reactors (MSR) began to be developed by the COMTES FHT Company in the Czech Republic as early as 2001. The outcome of this development was the MoNiCr alloy, an alternative to Hastelloy-N. The present study was carried out on two experimentally-manufactured nickel alloys: MoNiCr and HN80MTY. Its purpose was to activate recrystallization processes in the as-cast microstructure of these alloys. In addition, experiments were performed to find a temperature which produces complete recrystallization. Static and dynamic recrystallization was studied in both alloys as well. An important aspect was to determine the lowest amount of deformation which still ensures complete recrystallization and provides as uniform recrystallized grains as possible. Such microstructure is well-suited for subsequent forming operations. Specimen microstructures were characterized using light microscopy (LM), scanning electron microscopy (SEM), energy-dispersive X-ray spectroscopy (EDX) and electron backscatter diffraction (EBSD). Furthermore, corrosion tests of the MoNiCr alloy were conducted.

This paper describes the measurement of 55Mn(n,2n) and 127I(n,2n) reaction rates in a well-defined reactor field in a special core of LR-0 reactor. The reaction rates were derived using gamma-spectrometry by measuring gamma activities of irradiated MnO2 and NaI samples at a high purity germanium (HPGe) detector. The spectral average cross section (SACS) in 235U prompt fission neutron spectrum (PFNS) was experimentally determined to be 0.2393 ± 0.015 × 10−3 b for 55Mn and 1.2087 ± 0.052 × 10−3 b for 127I. These obtained results were compared with calculations by MCNP6 code using ENDF/B VII.1, ENDF/B VII, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND- 2010, CENDL-3.1, and IRDFF nuclear data libraries. In a case of 55Mn, a good agreement with ENDF/B VII.1, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND, and CENDL 3.1 nuclear data libraries was found, where C/E−1 is 0.1%, while IRDFF underestimated by about 15.8%. In the case of 127I, more significant discrepancies were found, where JENDL 3.3 and JENDL 4 overestimate the result by about 31.3%.