Master’s thesis focuses on the history and current progress in research of molten salt reactors around the world, with an emphasis placed on the properties of molten salts and the problems associated with their use. In relation to the practical part, one chapter is devoted to the creation of input file in the MCNP software. The practical part deals with neutron activation analysis of graphite prism experiment, which is filled with powder NaCl salt. This experiment is focused on the effect of salt on neutron flux with different energy levels. The whole problem was also simulated in the MCNP environment along with the experiment. At the end of the thesis, the individual methods are compared and evaluated.

The Department of Physics and Astronomy in Uppsala is planning to construct a source-based neutron irradiation facility at Ångströms laboratory. The purpose of this facility will, as of now, mainly be for students and employees to conduct scientific experiments and to calibrate the university's neutron scintillator detectors. A vital part of the construction is the shielding of the source, for both safety reasons and practicality issues. In this project, a series of simulations were made to determine the necessary dimensions of the shielding to ensure that the radiation dose levels stayed under the stipulated 6 mSv per year. Two types of shielding materials were tested, borated and non-borated polyethylene. It was found that the borated polyethylene performed better in shielding the source, but not enough to justify the higher material cost. The optimal dimensions of the shielding for non-borated polyethylene were found to be 60x60x74cm.

The neutron plays a central role in numerous fields of physics, a fact that entails a need for methods of measuring neutron energy spectra. In this project, a technique for neutron spectrometry through measurements of neutron-induced radioactivity in activation detectors was developed and tested. The developed technique involves irradiating element samples with neutrons, measuring activation products with a gamma spectrometer, and then performing a neutron spectrum unfolding procedure. The elements indium, iron, magnesium, aluminium, zinc, titanium, and copper were used as activation detectors and irradiated with neutrons from an americium-beryllium (AmBe) neutron source. Subsequent gamma spectrometry was performed with the UGGLA high-purity germanium detector setup at Uppsala University. The GRAVEL unfolding algorithm was implemented in MATLAB and used to unfold neutron spectra based on an initial spectrum guess. The unfolded neutron spectrum agrees well with the expected AmBe spectrum, though some difference between the spectra is attributed to neutron scattering in the irradiation environment. A possible ability to find approximate neutron spectra from inaccurate initial guesses is found, but additional work is needed to understand better how the initial guess affects the result for different neutron sources. Because activation detectors do not require electrical power when measuring neutrons, can be made sensitive to a wide range of neutron energies, and do not detect other types of radiation, future applications could find the developed neutron spectrometry method practical.

This master’s thesis provides a comprehensive overview of the conventional neutron sources from the perspective of reactions which lead to the production of neutrons, advantages, disadvantages, properties and their possible utilization. In the relation to the assembly of the laboratory neutron source and the unique experimental stand “Candle” basic methods of the neutron field analysis are outlined and two of them, the neutron activation analysis and the calculation using the MCNP software code are discussed in depth to apply and compare these methods. The experimental part deals with the realization of neutron activation analysis from its design itself, through gamma spectrometry to the cadmium ratio calculation. In compliance with the measurements, a calculation with MCNP code was run and both methods were evaluated and compared. The computation is complemented with the analysis of radiation situation on the borders of the supervised area, which is compared to the legal limit.