• Nuclear Engineering and Technology
• /
• 제50권5호
• /
• pp.788-793
• /
• 2018

High-density spent fuel (SF) storage racks have been installed to increase SF pool capacity. In these SF racks, neutron absorber materials were placed between fuel assemblies allowing the storage of fuel assemblies in close proximity to one another. The purpose of the neutron absorber materials is to preclude neutronic coupling between adjacent fuel assemblies and to maintain the fuel in a subcritical storage condition. METAMIC neutron absorber has been used in high-density storage racks. But, neutron absorber materials can be subject to severe conditions including long-term exposure to gamma radiation and neutron radiation. Recently, some of them have experienced degradation, such as white spots on the surface. Under these conditions, the material must continue to serve its intended function of absorbing neutrons. For the first time in Korea, this article uses a neutron attenuation test to examine the performance of METAMIC surveillance coupons. Also, scanning electron microscope analysis was carried out to verify the white spots that were detected on the surface of METAMIC. In the neutron attenuation test, there was no significant sign of boron loss in most of the METAMIC coupons, but the coupon with white spots had relatively less B-10 content than the others. In the scanning electron microscope analysis, corrosion material was detected in all METAMIC coupons. Especially, it was confirmed that the coupon with white spots contains much more corrosion material than the others.

• Nuclear Engineering and Technology
• /
• 제54권1호
• /
• pp.262-268
• /
• 2022

This paper introduces a novel concept of the pulsed neutron facility (PNF) for maximizing the production of the thermal neutrons and its application to medical use based on prompt gamma neutron activation analysis (PGNAA) using Monte Carlo simulations. The PNF consists of a compact D-T neutron generator, a graphite pile, and a detection system using Cadmium telluride (CdTe) detector arrays. The configuration of fuel pins in the graphite monolith and the design and materials for the moderating layer were studied to optimize the thermal neutron yields. Biological samples - normal and cancerous breast tissues - including chlorine, a trace element, were used to investigate the sensitivity of the characteristic γ-rays by neutron-trace material interactions and the detector responses of multiple particles. Around 90 % of neutrons emitted from a deuterium-tritium (D-T) neutron generator thermalized as they passed through the graphite stockpile. The thermal neutrons captured the chlorines in the samples, then the characteristic γ-rays with specific energy levels of 6.12, 7.80 and 8.58 MeV were emitted. Since the concentration of chlorine in the cancerous tissue is twice that in the normal tissue, the count ratio of the characteristic g-rays of the cancerous tissue over the normal tissue is approximately 2.

• Nuclear Engineering and Technology
• /
• 제56권2호
• /
• pp.402-409
• /
• 2024

High energy neutron irradiations impact on structural and electrical properties of alumina are studied with particular emphasis on real time in-situ radiation induced conductivity measurement in low flux region. Polycrystalline Al₂O₃ samples are subjected to high energy neutrons produced from D-T neutron generator and Am-Be neutron source. 14 MeV neutrons from D-T generator are chosen to study the role of fast neutron irradiation in the structural modification of samples. Real time in-situ electrical measurement is performed to investigate the change in insulation resistance of Al₂O₃ due to radiation induced conductivity at low flux regime. During neutron irradiation, a significant transient decrease in insulation resistance is observed which recovers relative higher value just after neutron exposure is switched off. XRD results of 14 MeV neutron irradiated samples suggest annealing effect. Impact of relatively low energy neutrons on the structural properties is also studied using Am-Be neutrons. In this case, clustering is observed on the sample surface after prolonged neutron exposure. The structural characterizations of pristine and irradiated Al₂O₃ samples are performed using XRD, SEM, and EDX. The results from these characterizations are analysed and interpreted in the manuscript.

• Nuclear Engineering and Technology
• /
• 제54권7호
• /
• pp.2572-2578
• /
• 2022

An approximate method for calculating the cadmium ratio of a cyclotron-based thermal neutron radiography system was developed. In this method, the Monte-Carlo code, MCNP6.2, was employed to calculate the neutron capture rates of Au-197, and the cadmium ratio was obtained by computing the ratio of neutron capture rates. From the simulation results, the computed cadmium ratio is reasonably acceptable, and the assumption of ignoring the fast neutron contribution to the cadmium ratio is valid.

• 한국방사선학회논문지
• /
• 제3권3호
• /
• pp.23-25
• /
• 2009

Gd-157 material has very large neutron capture cross section in the thermal region. So it is very useful to shield material for thermal neutrons. Futhermore, in the neutron capture experiment and calculation, the neutron absorption and scattering are very important. Especially these effects are conspicuous in the resonance energy region and below the thermal energy region. In the case of very narrow resonance, the effect of scattering is to be more considerable factor. In the present study, we obtained energy dependent neutron absorption ratios of natural indium in energy region from 0.003 to 100 keV by MCNP-4B Code. The coefficients for neutron absorption was calculated for circular type and 1 mm thickness. In the lower energy region, neutron absorption is larger than higher region, because of large capture cross section (1/v). Furthermore it seems very different neutron absorption in the large resonance energy region. These results are very useful to decide the thickness of sample and shielding materials.

• Nuclear Engineering and Technology
• /
• 제52권3호
• /
• pp.477-484
• /
• 2020

The control rod drive mechanism needs to be debugged after reactor fresh fuel loading. It is of great importance to monitor the subcriticality of this process accurately. A modified method was applied to the subcriticality monitoring process, in which only a single control rod cluster was fully withdrawn from the core. In order to correct the error in the results obtained by Neutron Source Multiplication Method, which is based on one point reactor model, Monte Carlo neutron transport code was employed to calculate the fission neutron distribution, the iterated fission probability and the neutron flux in the neutron detector. This article analyzed the effect of a coarse mesh and a fine mesh to tally fission neutron distributions, the iterated fission probability distributions and to calculate correction factors. The subcriticality before and after modification is compared with the subcriticality calculated by MCNP code. The modified results turn out to be closer to calculation. It's feasible to implement the modified NSM method in large local reactivity addition process using Monte Carlo code based on 3D model.

• Journal of Radiation Protection and Research
• /
• 제26권3호
• /
• pp.167-175
• /
• 2001

Semiconductor diode detectors coated with neutron reactive material are presently under investigation for various uses, such as remote sensing of thermal neutrons, fast neutron counting, and thermal neutron radiography. Theory indicates that single-coated devices can yield thermal neutron efficiencies from 4% to 11 %, which is supported by experimental evidence. Radiation endurance measurements indicate that the devices function well up to a limiting thermal neutron fluence of $10^{13}/cm^2$, beyond which noticeable degradation occurs. Thermal neutron contrast images of step wedges and simple phantoms, taken with dual in-line pixel devices, show promise for thermal neutron imaging detectors.

• 분석과학
• /
• 제16권5호
• /
• pp.426-429
• /
• 2003

Neutron induced prompt gamma-ray spectroscopy (NIPS) system equipped with a $^{252}Cf$ neutron source and a n-type coaxial HPGe detector was installed for the quantitative analysis of aqueous samples in KAERI, Korea. Since the thermal neutron flux for the $^{252}Cf$ neutron source is relatively low compared to that for the reactor, the use of a thermal neutron reflector in the NIPS system may lead to improved results. The enhancement by using various reflectors was carried out by comparing the Cl peak with or without a cadmium plate between sample and the $^{252}Cf$ source. The use of pyrolitic graphite as a reflector provided a good result.

• Journal of Plant Biology
• /
• 제7권4호
• /
• pp.9-14
• /
• 1964

According to the results measured the respiratory quotient of Saccharomyces cerevisiae with neutron radiation by manometric direct method, the respiratory quotient of them was stimulated at the dose(7$\times$106N/$\textrm{cm}^2$/sec) of neutron radiation for 60 seconds, and was inhibited in each group irradiated at the high dose (7$\times$108N/$\textrm{cm}^2$/sec) of neutron. Its physiological effects influenced on neutron had relations with respiratory quotient, reproductive rate and fermentation in the curve of normal logarithmic phase. The multiple reactions which appeared in yeast, indicated that a great deal of physiological function were closely correlated with the irradiated dosage of neutron. The kinds of free amino acid in yeast irradiated with neutron were different from those of unirradiated yeast. The activityof dehydrogenase system accelerated the metabolic function of yeast irradiated at some low dose of neutron. By this results, it may demonstrate that the fact which the phenomena obtained in the stimulation of neutron possess its character for several generation, is dependent on the theory of mutation. Subsequently, it seemed reasonable certain dominant type of microorganisms.

• Nuclear Engineering and Technology
• /
• 제49권5호
• /
• pp.920-927
• /
• 2017

With ever-advancing computer technology, the Monte Carlo (MC) neutron transport calculation is expanding its application area to nuclear reactor transient analysis. Dynamic MC (DMC) neutron tracking for transient analysis requires efficient algorithms for delayed neutron generation, neutron population control, and initial condition modeling. In this paper, a new MC steady-state simulation method based on time-dependent MC neutron tracking is proposed for steady-state initial condition modeling; during this process, prompt neutron sources and delayed neutron precursors for the DMC transient simulation can easily be sampled. The DMC method, including the proposed time-dependent DMC steady-state simulation method, has been implemented in McCARD and applied for two-dimensional core kinetics problems in the time-dependent neutron transport benchmark C5G7-TD. The McCARD DMC calculation results show good agreement with results of a deterministic transport analysis code, nTRACER.