SPECIFIC FEATURES OF APPLICATION OF CADMIUM AS A NEUTRON-ABSORBING SCREEN WHILE DOPING SILICON IN A NUCLEAR REACTOR

The gradient of the neutron field in a nuclear reactor and the requirements for the permissible spread of the specific electrical resistance over the volume of the silicon ingot makes it necessary to develop an irradiation device. This is especially true for large silicon ingots. One of the options for reducing the gradient of the neutron flux along the height of the ingot is the use of neutron-absorbing screens in the design of the irradiation device. At the WWR-K reactor, cadmium with a natural isotopic composition is used as a neutron-absorbing screen material. The paper presents the results of a study of an irradiation device with a cadmium screen. The effect of a cadmium screen on the neutron-physical characteristics of an irradiation device for silicon doping is shown.

Status of development and planning activities on CANS in Korea

This report reviews the overall status of the development and planning activities of compact accelerator-based neutron sources in Korea. For the last decade, the demand for the technology development and application of CANS has significantly increased, and becomes widely accepted by the science, engineering and industry sectors. Since the first technical workshop focused on CANS under the support of the Korea Nuclear Society in fall 2016, there have been numerous efforts to launch projects by several groups. Although unsuccessful, two CANS projects were newly launched in 2020. One is the 30-MeV cyclotron-based neutron source for industrial neutron imaging at the Korea Atomic Energy Research Institute (KAERI), and the other is the BNCT technology development at the Korea Institute of Radiological & Medical Sciences. A project proposal for an expansion of the proton LINAC facility at KAERI to 200 MeV for semiconductor irradiation testing through the produced neutron field is now almost complete and will be submitted to the government funding agency for review. The CANS project for BNCT based on the proton LINAC developed by the Dawonsys consortium is briefly described. The new neutron source based on electron LINAC is prepared by the Pohang Light Source laboratory, and the initial consideration and application targets are also described. A new strategic plan for national R&D on radiation technology and the enforcement of its infrastructure is still under way, and a more systematic approach to the development and application of neutron sources will be implemented through the strategic planning.

Characterization of an ²⁴¹Am-Be neutron irradiation facility at Institute for Nuclear Science and Technology

An automated panoramic irradiator with a 241Am-Be neutron source of 5 Ci is installed in a bunker-type medium room at the Institute for Nuclear Science and Technology (INST) for calibration of neutron devices. Bonner Sphere Spectrometer (BSS) formed by 6 spheres plus bare detector, with cylindrical, almost point like, 6LiI(Eu) scintillator and 2 different spectral unfolding FRUIT and BUNKIUT codes are used to characterize the neutron field in different measurement points along the irradiation bench. The neutron field is also simulated by MCNP5 software and compared with measurements performed by the BSS. The paper shows the main results obtained in terms of neutron spectra at fixed distances from the source as well as their neutron fluence rate (totaland direct) and ambient dose equivalent rate. These values measured by the BSS with two unfolding FRUIT and BUNKIUT codes are in good agreement with that of simulated by MCNP5 within 10%.

Simulation of the Super Critical Water Loop Using ATHLET Code During an Abnormal Scenario

Supercritical water (SCW) has advantages like high thermal efficiency and can operate at high temperature and pressure. At the same time, however, these properties bring up related issues, such as material compatibility and corrosion resistance. In an effort to fully investigate the operating conditions, and solutions to these issues, test facilities are being built by many research organizations. One such organization, the Research Center Řež (CVR) located in the Czech Republic, has developed an experimental supercritical water loop (SCWL). The purpose of this loop is to provide experimental data from material testing in various conditions, including operating under the neutron field. This will be achieved by inserting a test channel into the existing experimental reactor light water reactor 15 (LVR-15), which will require a license from the state nuclear regulator (State Office for Nuclear Safety (SUJB)). Part of the licensing documentation is the safety analysis, which combines results from developed models using the thermohydraulic code ATHLET 3.1 A patch 1, as well as the experimental out of pile data. Among the postulated scenarios, an abnormal sequence (labeled A2—Loss of power in the loop) was analyzed in order to provide a preliminary benchmark. This scenario is similar to the postulated in-pile A2 and it was used for the benchmark activity. The aim of this paper is to present this activity including the adopted assumptions in the model. In particular, the paper presents, how these assumptions influenced the results indicating the discrepancies obtained in the first part of the transient.

Neutron Field Shaping Using Graphite for Reaction Rate Measurements

Neutron field shaping is the suitable method for validation of cross section in various energy regions. By increasing the share of neutrons of a certain energy interval and decreasing the share of other, a reaction becomes more sensitive to selected neutrons. As a result, reaction cross section can be validated in selected energy regions more precisely. The shaping can be carried out by both neutron filters which are materials with high absorption in some energy region, or by diffusion material changing the shape of neutron spectra by means of slowing down process. In the presented experiments, the neutron field of the light reactor 0 (LR-0) research reactor was shaped by both using graphite blocks inserted into the core and Cd cladding for increasing the epithermal reaction rate share in total reaction rates. The calculations were carried out with the Monte Carlo N-Particle Transport Code 6 (MCNP6) code and the most recent nuclear data libraries. The results in the pure graphite neutron field are in good agreement; in case of Cd cladding, significant discrepancies were reported. In case of the 23Na(n,γ)24Na reaction, overestimation by about 14% was reached in International Reactor Dosimetry and Fusion File (IRDFF-II), results in other libraries are comparable. In case of 58Fe(n,γ)59Fe, the overestimation as high as 18% is reported in IRDFF-II. For 64Zn(n,γ)65Zn reasonable agreement was reached in evaluated nuclear data file (ENDF/B-VIII), where discrepancies in pure graphite neutron field or in case of Cd cladding are about 10–15%.