Spontaneous Neutron Source Research Articles

Development of a new Monte-Carlo neutron transport code for radiation shielding

A continuous energy Monte-Carlo method-based neutron transport code for fixed source in 3D solid geometry has been developed. The main objective is to integrate this code with recently developed Monte-Carlo gamma shielding code MCBLD, so as to build an in-house code for coupled neutron-gamma shielding study. Also it is intended to use of the new ENDF data library (ENDF/B-VIII) in this code. This paper gives an overview of the code in terms of processing neutron interaction data and numerical scheme to track scattered neutrons. Present capability of the code is demonstrated by presenting fluence due to mono-energetic neutron source as well as widely used spontaneous neutron source from 252Cf isotope, placed at the centre of different size iron spheres. Results are also generated from general purpose open-source Monte-Carlo code FLUKA. Overall good agreement between the two validates the numerical implementation of angle and energy dependent neutron cross-section data in this code.

241Am/Be source optimum geometry for DSRS management-based Monte Carlo simulations

The geometry form related to 241Am/Be spontaneous neutron sources has been investigated to evaluate the most optimizing geometry for disused sealed radioactive source (DSRS) disposal. Nine source geometries were assessed: point, disk, sphere 1, sphere 2, sphere 3, cylinder 1, cylinder 2, rectangle 1, and rectangle 2. The most radiological optimizing geometries were found to be the disk, followed by the point source, sphere 1, and cylinder 1, while the rectangle or parallelepiped 2 source forms were the worst cases. Neutron and photon generated fluxes were computed, and the parallelepiped 2 source geometry generated the lower value of flux inversely to the effective dose rate computed in the decision-making area, where it was found to be the highest value. The obtained results were in agreement with the as low as reasonably achievable principle for the exposure rate optimization. The obtained exposure dose rates were found to be <∼2 µSv/h (maximum value on horizontal calculation) and 1.75 µSv/h (maximum value on vertical computation), which are lower than the 2.5 µSv/h acceptable limit for the public area. Whenever possible, the DSRSs should be pressed into a disk form (or a cylindrical form with a small height) before embedding them into the Am1 P60 capsule for disposal purposes. The dose profile for the 241Am/Be source obtained, the neutron flux, and the gamma generated from neutron absorption showed agreement with the expected experimental physical data. The supported data contribute to postulate that Monte Carlo methods are effective computational tools that can be used to select the most effective radioactive waste form for disposal purposes.

New Cf-252 neutron source shielding design based Monte Carlo simulation using material combination

Neutron shielding has been a worldwide concern for decades and appropriate methods in nuclear reactors and associated facilities for shielding have been developed. Light materials such as paraffin and water cannot be used in neutron radiography due to structural and fire concerns, and a material in the solid form is needed for effective neutron shielding. Therefore, the present study developed a new neutron shielding design for fixed industrial radiography facilities with solid structures based on material combination. Different materials were investigated to find the most appropriate combination design to shield the spontaneous neutron emitted from a Cf-252 source (with an energy in the range from several keV up to 20 MeV). The combination of iron, graphite, boron (or borate materials), and lead in this order, respectively, were found to be the most appropriate shielding structure for an open Cf-252 source used in fixed industrial radiography. As iron is characterized by a high removal cross section, its use in shielding the californium spontaneous neutron source is the key outcome of the present study. These results were confirmed with the Monte Carlo simulation-based particle and heavy ion transport code system.

Measurements and Model Calculations of Activation Reaction Rate for (n, p) Reaction on 54Fe Isotope

In China Fusion Engineering Testing Reactor research, the neutronics experiment is essential in validating the neutronics codes and tools used in a blanket. Reduction activation ferritic-martensitic steel is the main structural material of the water-cooled ceramic breeder blanket, of which the main composition is Fe. However, the activation reaction of 54 Fe(n, p) 54 Mn will bring the long half-life product of 54 Mn. Therefore, the reaction rate of the reaction channel 54 Fe(n, p) 54 Mn was investigated with accuracy in this paper. In a fusion reactor, neutron is generated not only from deuterium tritium reaction but also from the deuterium deuterium (DD) reaction. The spontaneous neutron source, 3 × 10 8 n/s 252 Cf, was used to simulate DD neutron field. After irradiated for about 5 days, the activated Fe foil was measured by a high-resolution gamma-ray spectrometer with a high-purity germanium detector. The neutron flux was calculated using a Monte Carlo transport code and Fusion Evaluated Neutron Data Library (FENDL) 3.0 Files. The neutron group cross section was processed by NJOY. The relative error between the calculated value and the experimental value is less than 7%. The error of the database EAF2010 is the smallest in this paper. The self-shielding factor has a greater impact on FENDL 3.0 database.

Thorium breeder and burner fuel cycles in reduced-moderation LWRs compared to fast reactors

In this paper, a reduced-moderation LWR (RMLWR) system is compared to a sodium-cooled fast reactor (SFR) as a reference fast system for sustained transuranic burning and self-sufficient break-even cycles using thorium (Th) as the fertile material. SFRs can achieve a higher discharge burn-up (and incineration rate – in the case of burners) than RMLWRs, which results in a highly significant reduction of ∼40% and ∼60% in fuel reprocessing and fabrication requirements for breeder and burner reactors respectively. With sufficiently reduced moderation, the repository radiotoxicity and decay heat of actinide process waste from RMLWRs is very similar to that from SFRs, assuming same fraction of process losses per reprocessed fuel inventory, for the breeder and burner cases. In all cases, fuel fabrication is challenging due to high energy gamma and spontaneous neutron sources. The gamma source is comparable in the burner cores but the spontaneous neutron source in the burner options is approximately an order of magnitude lower for the SFR after a large number of recycles. Th fuel is highly preferable to uranium fuel in RMLWRs due to a large improvement in the void coefficient, which must be kept negative. This constraint makes it difficult, if not impossible, to fuel an RMLWR burner or breeder with uranium. Similarly, the void coefficient in SFRs is significantly better with Th fuel, although it is probably not essential for this to be negative. Although the neutron spectrum and core geometry are substantially different, RMLWRs are a variant of conventional LWR technology. RMLWRs have never been built, whereas some SFRs have been built and operated. Despite this, SFRs are anticipated to have a higher capital cost, which appears to outweigh the lower reprocessing and fuel fabrication requirements, although this conclusion is sensitive to the reactor size, cost and availability factor assumptions.

Method of experimental estimation of the effective delayed neutron fraction and of the neutron generation lifetime in the IBR-2 pulsed reactor

Experimental investigations of random fluctuations in the power of the IBR-2 periodic pulsed reactor carried out in the steady-state mode of its operation and in the pulsed mode made it possible to obtain estimations of the main parameters of the kinetics – the effective fraction of delayed neutrons and the average lifetime of prompt neutrons. Functionals were measured relating the main parameters of the kinetic behaviour: the prompt neutron decay constant (steady-state mode) and the relative dispersion of pulse energy fluctuations. It was shown that the experimental values of parameters of the kinetic behaviour are close to the calculated ones. In addition, the power of the spontaneous neutron source was estimated on the basis of the analysis of neutron noise.