Search Results (7)

The point-kernel (PK) technique has a long history in applied radiation shielding, originating from the early days of digital computers. The PK technique applied to gamma-ray attenuation is one of many successful applications, based on the linear superposition principle applied to distributed radiation sources. Mathematically speaking, the distributed source will produce a detector response equivalent to the numerical integration of the radiation received from an equivalent number of point sources. In this treatment, there is no interference between individual point sources, while inherent limitations of the PK method are its inability to simulate gamma scattering in shields and the usage of simple boundary conditions. The PK method generally works for gamma-ray shielding with corrective B-factor for scattering and only specifically for fast neutron attenuation in a hydrogenous medium with the definition of cross section removal. This paper presents theoretical and programming aspects of the PK program developed for a distributed source of photons (line, disc, plane, sphere, slab volume, etc.) and slab shields. The derived flux solutions go beyond classical textbooks as they include the analytical integration of Taylor B-factor, obtaining a closed form readily suitable for programming. The specific computational modules are unified with a graphical user interface (GUI), assisting users with input/output data and visualization, developed for the fast radiological characterization of simple shielding problems. Numerical results of the selected PK test cases are presented and verified with the CADIS hybrid shielding methodology of the MAVRIC/SCALE6.1.3 code package from the ORNL. Full article

The UK government and many international experts have pointed out that nuclear energy has an important role to play in the transition towards a decarbonised energy system since it is the only freely manageable very low-carbon energy technology with 24/7 availability to complement renewables. Besides current investments in light water reactor technologies, we need innovation for improved fuel usage and reduced waste creation, like that offered by iMAGINE, for the required broad success of nuclear technologies. To allow for quick progress in innovative technologies like iMAGINE and their regulation, a timely investment into urgently needed experimental infrastructure and expertise development will be required to assure the availability of capacities and capabilities. The initial steps to start the development of such a new reactor physics experimental facility to investigate molten salt fast reactor technology are discussed, and a stepwise approach for the development of the experimental facility is described. The down selection for the choice for a diverse control and shutdown system is described through manipulating the reflector (control) and splitting the core (shutdown). The developed innovative core design of having the two core parts in two different rooms opens completely new opportunities and will allow for the manifestation of the request for separated operational and experimental crews, as nowadays requested by regulators into the built environment. The proposed physical separation of safety-relevant operational systems from the experimental room should on the one hand help to ease the access to the facility for visiting experimental specialists. On the other hand, the location of all safety-relevant systems in a now separated access-controlled area for the operational team will limit the risk of misuse through third party access. The planned experimental programme is described with the major steps as follows: core criticality experiments, followed by experiments to determine the neutron flux, neutron spectrum and power distribution as well as experiments to understand the effect of changes in reactivity and flux as a function of salt density, temperature and composition change. Full article

The lifespan of core graphite under neutron irradiation in a commercial molten salt reactor (MSR) has an important influence on its economy. Flattening the fast neutron flux (≥0.05 MeV) distribution in the core is the main method to extend the graphite irradiation lifespan. In this paper, the effects of the key parameters of MSRs on fast neutron flux distribution, including volume fraction (VF) of fuel salt, pitch of hexagonal fuel assembly, core zoning, and layout of control rod assemblies, were studied. The fast neutron flux distribution in a regular hexagon fuel assembly was first analyzed by varying VF and pitch. It was demonstrated that changing VF is more effective in reducing the fast neutron flux in both global and local graphite blocks. Flattening the fast neutron flux distribution of a commercial MSR core was then carried out by zoning the core into two regions under different VFs. Considering both the fast neutron flux distribution and burnup depth, an optimized core was obtained. The fast neutron flux distribution of the optimized core was further flattened by the rational arrangement of control rod channels. The calculation results show that the final optimized core could reduce the maximum fast neutron flux of the graphite blocks by about 30% and result in a more negative temperature reactivity coefficient, while slightly decreasing the burnup and maintaining a fully acceptable core temperature distribution. Full article

Molten salt reactors have gained substantial interest in the last years due to their flexibility and their potential for simplified closed fuel cycle operation for massive expansion in low-carbon electricity production, which will be required for a future net-zero society. The importance of a zero-power reactor for the process of developing a new, innovative rector concept, such as that required for the molten salt fast reactor based on iMAGINE technology, which operates directly on spent nuclear fuel, is described here. It is based on historical developments as well as the current demand for experimental results and key factors that are relevant to the success of the next step in the development process of all innovative reactor types. In the systematic modelling and simulation of a zero-power molten salt reactor, the radius and the feedback effects are studied for a eutectic based system, while a heavy metal rich chloride-based system are studied depending on the uranium enrichment accompanied with the effects on neutron flux spectrum and spatial distribution. These results are used to support the relevant decision for the narrowing down of the configurations supported by considerations on cost and proliferation for the follow up 3-D analysis. The results provide for the first time a systematic modelling and simulation approach for a new reactor physics experiment for an advanced technology. The expected core volumes for these configurations have been studied using multi-group and continuous energy Monte-Carlo simulations identifying the 35% enriched systems as the most attractive. This finally leads to the choice of heavy metal rich compositions with 35% enrichment as the reference system for future studies of the next steps in the zero power reactor investigation. An alternative could be the eutectic system in the case the increased core diameter is manageable. The inter-comparison of the different applied codes and approaches available in the SCALE package has delivered a very good agreement between the results, creating trust into the developed and used models and methods. Full article

Among all the possible occurring reactivity effects of a fast reactor, the situations whereby the control rod was inserted, or the coolant was voided could lead to strong anisotropy of neutron flux distribution, therefore the angular dependence on neutron flux should be considered during the few-group cross-sections generation. Therefore, the purpose of this paper is to compare the influence whether the angular dependence on neutron flux is considered in the calculation of few-group cross sections for the reactivity effect calculation. In the study, the 1-D SN finite difference neutron transport equation solver was implemented in the TULIP of SARAX code system so that the high-order neutron flux could be obtained. Meanwhile, the improved Tone’s method was also applied. The numerical results were obtained based on three experimental FR cores, the JOYO MK-I core, ZPPR-9 core, and ZPPR-10B core. Both control rod worth and sodium void reactivity were calculated and compared with the measurement data. By summarizing and comparing the results of 46 cases, significant differences were found between different consideration of the neutronic analysis. The consideration of angular dependence on neutron flux distribution in the few-group cross-sections generation was beneficial to the neutronic design analysis of FR, especially for the reactivity effect calculation. Full article

The Neutron Radiography Reactor at Idaho National Laboratory (INL) has two beamlines extending radially outward from the east and north faces of the reactor core. The control rod withdrawal procedure has recently been altered, potentially changing power distribution of the reactor and thus the properties of the neutron beams, calling for characterization of the neutron beams. The characterization of the East Radiography Station involved experiments used to measure the following characteristics: Neutron flux, neutron beam uniformity, cadmium ratio, image quality, and the neutron energy spectrum. The ERS is a Category-I neutron radiography facility signifying it has the highest possible rank a radiography station can achieve. The thermal equivalent neutron flux was measured using gold foil activation and determined to be 9.61 × 10⁶ ± 2.47 × 10⁵ n/cm²-s with a relatively uniform profile across the image plane. The cadmium ratio measurement was performed using bare and cadmium-covered gold foils and measured to be 2.05 ± 2.9%, indicating large epithermal and fast neutron content in the beam. The neutron energy spectrum was measured using foil activation coupled with unfolding algorithms provided by the software package Unfolding with MAXED and GRAVEL (UMG). The Monte-Carlo N-Particle (MCNP6) transport code was used to assist with the unfolding process. UMG, MCNP6, and measured foil activities were used to determine a neutron energy spectrum which was implemented into the MCNP6 model of the east neutron beam to contribute to future studies. Full article

CANDLE (Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy producing reactor) reactors have been intensively researched in the last decades [1–6]. Research shows that this kind of reactor is highly economical, safe and efficiently saves resources, thus extending large scale fission nuclear energy utilization for thousands of years, benefitting the whole of society. For many developing countries with a large population and high energy demands, such as China and India, middle (1000 MWth) and large (2000 MWth) CANDLE fast reactors are obviously more suitable than small reactors [2]. In this paper, the middle and large CANDLE reactors are investigated with U-Pu and combined ThU-UPu fuel cycles, aiming to utilize the abundant thorium resources and optimize the radial power distribution. To achieve these design purposes, the present designs were utilized, simply dividing the core into two fuel regions in the radial direction. The less active fuel, such as thorium or natural uranium, was loaded in the inner core region and the fuel with low-level enrichment, e.g. 2.0% enriched uranium, was loaded in the outer core region. By this simple core configuration and fuel setting, rather than using a complicated method, we can obtain the desired middle and large CANDLE fast cores with reasonable core geometry and thermal hydraulic parameters that perform safely and economically; as is to be expected from CANDLE. To assist in understanding the CANDLE reactor’s attributes, analysis and discussion of the calculation results achieved are provided. Full article