• Journal of Radiation Protection and Research
• /
• v.42 no.3
• /
• pp.141-145
• /
• 2017

Background: The concrete walls inside the vaults of cyclotron facilities are activated by neutrons emitted by the targets during radioisotope production. Reducing the amount of radioactive waste created in such facilities is very important in case they are decommissioned. Thus, we proposed a strategy of reducing the neutron activation of the concrete walls in cyclotrons during operation. Materials and Methods: A polyethylene plate and B-doped Al sheet (30 wt% of B and 2.5 mm in thickness) were placed in front of the wall in the cyclotron room of a radioisotope production facility for pharmaceutical use. The target was Xe gas, and a Cu block was utilized for proton dumping. The irradiation time, proton energy, and beam current were 8 hours, 30 MeV, and $125{\mu}A$, respectively. To determine a suitable thickness for the polyethylene plate set in front of the B-doped Al sheet, the neutron-reducing effects achieved by inserting such sheets at several depths within polyethylene plate stacks were evaluated. The neutron fluence was monitored using an activation detector and 20-g on de Au foil samples with and without 0.5-mm-thick Cd foil. Each Au foil sample was pasted onto the center of a polyethylene plate and B-doped Al sheet, and the absolute activity of one Au foil sample was measured as a standard using a Ge detector. The resulting relative activities were obtained by calculating the ratio of the photostimulated luminescence of each foil sample to that of the standard Au foil. Results and Discussion: When the combination of a 4-cm-thick polyethylene plate and B-doped Al sheet was employed, the thermal neutron rate was reduced by 78%. Conclusion: The combination of a 4-cm-thick polyethylene plate and B-doped Al sheet effectively reduced the neutron activation of the investigated concrete wall.

• Nuclear Engineering and Technology
• /
• v.49 no.8
• /
• pp.1600-1609
• /
• 2017

The accelerator-driven subcritical system (ADS) is driven by a neutron source from spallation reactions introduced by the injected proton beam. Part of the neutron source has energy as high as a few hundred MeV to a few GeV. The effects of high-energy source neutrons ($E_n$ > 20 MeV) are usually approximated by energy cut-off treatment in practical core calculations, which can overestimate the predicted proton beam current in the ADS design. This article intends to quantize this effect and propose a way to solve this problem. To evaluate the effects of high-energy neutrons in the subcritical core, two models are established aiming to cover the features of current experimental facilities and industrial-scale ADS in the future. The results show that high-energy neutrons with $E_n$ > 20 MeV are of small fraction (2.6%) in the neutron source, but their contribution to the source efficiency is about 23% for the large scale ADS. Based on this, a neutron source efficiency correction factor is proposed. Tests show that the new correction method works well in the ADS calculation. This method can effectively improve the accuracy of the prediction of the proton beam current.

• Nuclear Engineering and Technology
• /
• v.50 no.1
• /
• pp.151-158
• /
• 2018

The neutron powder diffractometer (NPD) is used to study a variety of technologically important and scientifically driven materials such as superconductors, multiferroics, catalysts, alloys, ceramics, cements, colossal magnetoresistance perovskites, magnets, thermoelectrics, zeolites, pharmaceuticals, etc. Monte Carlo-based codes are powerful tools to evaluate the neutronic behavior of the NPD. In the present study, MCNPX 2.6.0 and Vitess 3.3a codes were applied to simulate NPD facilities, which could be equipped with different optic devices such as pyrolytic graphite or neutron chopper. So, the Monte Carlo-based codes were used to simulate the NPD facility of the 5 MW Tehran Research Reactor. The simulation results were compared to the experimental data. The theoretical results showed good conformity to experimental data, which indicates acceptable performance of the Vitess 3.3a code in the neutron optic section of calculations. Another extracted result of this work shows that application of neutron chopper instead of monochromator could be efficient to keep neutron flux intensity higher than $10^6n/s/cm^2$ at sample position.

• Journal of Radiation Protection and Research
• /
• v.43 no.1
• /
• pp.20-28
• /
• 2018

Background: Heat energy produced in nuclear reactors and nuclear fuel cycle facilities interactions modifies the physical properties of the shielding materials containing water content. Therefore, in the present paper, effect of the heat on shielding effectiveness of the concretes is investigated for gamma and neutron. The mass attenuation coefficients, effective atomic numbers, fast neutron removal cross-section and exposure buildup factors. Materials and Methods: The mass attenuation coefficients, effective atomic numbers, fast neutron removal cross-section and exposure buildup factors of ordinary and heavy concretes were investigated using NIST data of XCOM program and Geometric Progression method. Results and Discussion: The improvement in shielding effectiveness for photon and reduction in fast neutron for ordinary concrete was observed. The change in the neutron shielding effectiveness was insignificant. Conclusion: The present investigation on interaction of gamma and neutron radiation would be very useful for assessment of shielding efficiency of the concrete used in high temperature applications such as reactors.

• Journal of Radiation Industry
• /
• v.17 no.4
• /
• pp.443-449
• /
• 2023

KAERI(Korea Atomic Energy Research Institute) is conducting research and development of large-scale radiation generators and the latest radiation measuring instruments. In particular, research and development of security screening equipment using an electron beam accelerator and a neutron generator is in progress recently. Globally, 20 ft containers are used to transport imports and exports, and electron beam accelerators are radiation sources to measure the shape of the material inside the container during customs inspections in each country. KAERI is developing a device that can use an electron beam accelerator and a neutron generator sequentially to grasp the shape of various materials as well as the location of the internal target material. In this study, when using the neutron generator, the radiation dose and the degree of activation by neutron for the facility and surrounding environment, facility equipment were simulated using MCNP and FISPACT code. As a result, the shielding structures inside and outside the radiation control area were satisfactory to the reference level established conservatively based on the Korean Nuclear Act.

• Nuclear Engineering and Technology
• /
• v.52 no.8
• /
• pp.1784-1791
• /
• 2020

The 14 MeV neutron generator facility is being developed by the Institute for Plasma Research India to conduct the lab scale experiments related to Indian breeding blanket system for ITER and DEMO. It will also be utilized for material testing, shielding experiments and development of fusion diagnostics. Occupational radiation exposure control is necessary for the all kind of nuclear facilities to get the operational licensing from governing authorities and nuclear regulatory bodies. In the same way, the radiation exposure for the 14 MeV neutron generator facility at the occupational worker area and accessible zones for general workers should be under the permissible limit of AERB India. The generator is designed for the yield of 10¹² n/s. The shielding assessment has been made to estimate the radiation dose during the operational time of the neutron generator. The facility has many utilities and constraints like ventilation ducts, accessible doors, accessibility of neutron generator components and to conduct the experiments which make the shielding assessment challenging to provide proper safety for occupational workers and the general public. The neutron and gamma dose rates have been estimated using the MCNP radiation transport code and ENDF -VII nuclear data libraries. The ICRP-74 fluence to dose conversion coefficients has been used for the assessment. The annual radiation exposure has been assessed by considering 500 h per year operational time. The provision of local shield near to neutron generator has been also evaluated to reduce the annual radiation doses. The comprehensive results of radiation shielding capability of neutron generator building and local shield design have been presented in the paper along with detailed maps of radiation field.

• Journal of Radiation Industry
• /
• v.18 no.2
• /
• pp.133-140
• /
• 2024

The Korean 2^nd basic plan for management of high-level radioactive waste presented a plan to manage spent nuclear fuel through dry storage facilities in NPP on-site. For the construction and operation of the facility, it is necessary to develop the monitoring system of the integrity of spent nuclear fuel before operation. NUREG-1536 recommends that the theoretical cask array, typically in the 2×10 array, should account for shadowing effect among the dry storage casks. The objective of this study was to evaluate neutron flux accounting for shadowing effect among dry storage casks. The neutron release rate was evaluated using ORIGEN based on the design basis fuel condition. And the simulation of dry storage casks and evaluation of the shadowing effect were performed using MCNP. Shadowing effect of other dry storage casks was the highest at the center of the dry storage facility of the 2×10 array compared with the outside of the cask. The shadowing effect of neutron flux on the surface among the metal casks was approximately 18% at point 1, 23% at point 2, and 43% at point 3. For the concrete casks, the shadowing effect of neutron flux on the surface was approximately 46% at point 1, 51% at point 2, and 52% at point 3. This means that correction is necessary to monitor the integrity of spent nuclear fuel in each dry storage cask through evaluation of shadowing effect. The results of this study will be used for comparative analysis of neutron measurement data from spent nuclear fuels in dry storage cask. Additionally, the neutron flux evaluation procedure used in this study could be used as the basic data of safety assessment of dry storage cask and development of safety guide.

• Proceedings of the KSEEG Conference
• /
• 2001.04a
• /
• pp.376-377
• /
• 2001

• Journal of the Korean Society of Radiology
• /
• v.2 no.3
• /
• pp.27-31
• /
• 2008

The radiation protection measures for the photoneutrons are one of the most important issue of radiation safety in high energy X-ray facilities. When the photoneutrons are released from the facility, the general public as well as occupational workers are exposed to unexpected radiations by neutron skyshine effect. In this study, the photoneutron inventory are calculated using monte carlo mothed, and the neutron skyshine dose rate is assessed using the inventory. A 9MeV X-ray cargo inspection facility is considered as a reference facility.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2004-2010
• /
• 2022

Quantification of sensitive material is of vital importance when it comes to the movement of nuclear fuel throughout its life cycle. Within the electrorefiner vessel of electrochemical separation facilities, the task of quantifying plutonium by neutron analysis is especially challenging due to it being in a constant mixture with curium. It is for this reason that current neutron multiplicity methods would prove ineffective as a safeguards measure. An alternative means of plutonium verification is investigated that utilizes the (𝛼,n) signature that comes as a result of the eutectic salt within the electrorefiner. This is done by utilizing the multiplicity variable a and breaking it down into its constituent components: spontaneous fission neutrons and (𝛼,n) yield. From there, the (𝛼,n) signature is related to the plutonium content of the fuel.