A simulational study of the indirect-geometry neutron spectrometer BIFROST at the European Spallation Source, from neutron source position to detector position.

The European Spallation Source (ESS) is intended to become the most powerful spallation neutron source in the world and the flagship of neutron science in upcoming decades. The exceptionally high neutron flux will provide unique opportunities for scientific experiments but also set high requirements for the detectors. One of the most challenging aspects is the rate capability and in particular the peak instantaneous rate capability, i.e. the number of neutrons hitting the detector per channel or cm2 at the peak of the neutron pulse. The primary purpose of this paper is to estimate the incident rates that are anticipated for the BIFROST instrument planned for ESS, and also to demonstrate the use of powerful simulation tools for the correct interpretation of neutron transport in crystalline materials. A full simulation model of the instrument from source to detector position, implemented with the use of multiple simulation software packages, is presented. For a single detector tube, instantaneous incident rates with a maximum of 1.7 GHz for a Bragg peak from a single crystal and 0.3 MHz for a vanadium sample are found. This paper also includes the first application of a new pyrolytic graphite model and a comparison of different simulation tools to highlight their strengths and weaknesses.

Experimental study of concrete activation compared to MCNP simulations for safety of neutron sources.

The neutron activation of shielding materials and the generated decay gamma radiation are well-known issues in terms of occupational exposure. Though the trace elements of shielding concretes can be dominant sources of the produced activity in such cases, their concentrations are often missing from the input data of shielding-related Monte Carlo simulations. For this reason, three concrete types were studied, that were considered in the European Spallation Source (ESS) ERIC. Their composition - including the trace elements - were determined via XRF, PGAA and NAA techniques. Realistic input data were developed for these materials, containing the parent elements of all the dominant radioisotopes, and were validated against measured data of neutron irradiation experiments.

Tritium internal dose estimation from measurements with liquid scintillators.

Tritium may exist in several chemical and physical forms in workplaces, common occurrences are in vapor or liquid form (as tritiated water) and in organic form (e.g. thymidine) which can get into the body by inhalation or by ingestion. For internal dose assessment it is usually assumed that urine samples for tritium analysis are obtained after the tritium concentration inside the body has reached equilibrium following intake. Comparison was carried out for two types of vials, two efficiency calculation methods and two available liquid scintillation devices to highlight the errors of the measurements. The results were used for dose estimation with MONDAL-3 software. It has been shown that concerning the accuracy of the final internal dose assessment, the uncertainties of the assumptions used in the dose assessment (for example the date and route of intake, the physical and chemical form) can be more influential than the errors of the measured data. Therefore, the improvement of the experimental accuracy alone is not the proper way to improve the accuracy of the internal dose estimation.

Neutron activation and prompt gamma intensity in Ar/CO2-filled neutron detectors at the European Spallation Source.

Monte Carlo simulations using MCNP6.1 were performed to study the effect of neutron activation in Ar/CO2 neutron detector counting gas. A general MCNP model was built and validated with simple analytical calculations. Simulations and calculations agree that only the 40Ar activation can have a considerable effect. It was shown that neither the prompt gamma intensity from the 40Ar neutron capture nor the produced 41Ar activity have an impact in terms of gamma dose rate around the detector and background level.

CALIBRATION OF A WHOLE BODY COUNTER FOR 241Am WITH THE LLNL CHEST PHANTOM.

An occupational incorporation event occurred at the Radioactive Waste Treatment and Disposal Facility in December 2013 at Püspökszilágy, Hungary. Internal contamination due to (241)Am was discovered by a regular routine whole body counting measurement at the Centre for Energy Research, Hungarian Academy of Sciences. After that, a whole body counter was calibrated for an organ counting geometry. For preliminary calibration, a home-made MIX-D chest phantom was applied simulating uniform lung activity distribution by (241)Am point sources located in different positions within the lung volume of the phantom. In order to carry out a more precise calibration, a Lawrence Livermore National Laboratory (LLNL) chest phantom was provided by the International Atomic Energy Agency. For counting efficiency over the lungs, values of 0.46±0.19 and 0.55±0.07 cps kBq(-1) were obtained for the MIX-D and the LLNL phantom, respectively; thus, the results are in good agreement.

Radionuclide migration modeling through the soil-plant system as adapted for Hungarian environment.

The migration of radionuclides released as fallout through the food-chain to humans was modelled using the MODELMAKER software. In the established dynamic environmental transfer model ETM-2002 with compartmental structure, the principal pathways of vegetable contamination were studied specifically for the Hungarian environment. These pathways were: direct deposition on plant surface, root uptake and deposition after resuspension from the soil surface. As result of the modeling the variation of activity-concentration with time was obtained in the compartments. The validation of the model was done by comparing the calculated results with those obtained in field experiments. A sensitivity analysis of the input parameters was also carried out and the parameters were categorized by their sensitivity index (SI). According to this study, the most sensitive parameters are the daily human intake of vegetable, the distribution coefficient, the transfer factor from soil to plant and the weathering half-time. The most probable distribution types for the parameter values were also determined based on Monte Carlo simulations.

Stochastic dose/risk assessment uncertainty and the concept of potential exposures.

Assessments of future potential doses and risks (e.g. in connection with radioactive waste disposal) can be associated with substantial uncertainties. These uncertainties are often incorporated into assessments using stochastic models, and give rise to doses characterised in probabilistic terms. Questions about how doses should be related to risks arise even in the case of deterministic dose assessments. However, results of a probabilistic dose assessment may lead to additional issues when interpreting them or attempting to convert them to risk for regulatory purposes. Some problems are shown to originate from the difference between the meaning of a probabilistic dose value and the definition of potential exposure. Others relate to the appropriate choice of, or approach to compliance with, a regulatory limit. A possible method for setting limits on probabilistic (potential) doses or risks is proposed, based on an accepted technique for determining detection limits.

Automated separation process for radioanalytical purposes at nuclear power plants.

Chemical separation processes have been developed to remove the matrix components and thus to determine fission products, especially radioiodine nuclides, in the primary coolant of WWER-type nuclear reactors. Special procedures have been elaborated to enrich long-lived nuclides in waste waters to be released and to separate and enrich caesium isotopes in the environment. All processes are based mainly on ion-exchange separations using amorphous zirconium phosphate. Automated equipment was constructed to meet the demands of the plant personnel for serial analysis.