• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.153-164
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• 2014

Currently, there are many programs which are now being developed or already developed to predict radionuclide and corrosion product at the stage of designing NPP. However, since there are not many developments in evaluating quantity of activation corrosion products occurring when disassembling a nuclear power plant there exist some difficulties in calculating accurately. In order to evaluate activation products inventory for the research of effect of neutron activation in the reactor vessel, component of nuclear reactor and adjacent structures, it should be evaluated by using operation history of nuclear reactor, material composition of structure and average neutron flux in every field representing fixed structure of nuclear reactor. In this study, CORA, PACTOLE, CRUDSIM, CREAT and ACE codes are analyzed to predict the quantity of radionuclide and corrosion product of primary reactor which is used at the stage of designing. As a future study, the accuracy in calculating the quantity of product corrosion can be increase by finding out the possibility of use and improvement for evaluation of the decontamination.

• Journal of the Korean Chemical Society
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• v.8 no.2
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• pp.47-56
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• 1964

When a developed paper partition chromatogram was irradiated by means of the pneumatic tube system of the Korean research reactor (neutron flux: 1.5 ${\times}10^{12}n/cm^2$sec.) the qualitative confirmation of the developed spot on the chromatogram was possible. In the case of an organic halogen compounds (chloro-acid, chloro-ester, iodide, and fluoride) the spot analysis was possible by the present method whereas the same spot could not give the distinct coloring with a common coloring reagents. Filterpaper thickness calibration and activity calibration induced by irradiation of the components of the filter paper, which were a source of erraneous interpretation of the spot, were searched and an average filterpaper calibration method and filter paper activity were improvised to obtain a good qualitative analysis of the spot. Finally the use and applicability of this method for the analysis identification of an organic halogen compound were evaluated. As the filter paper phase an ordinary phase (Whatmann #1, filter paper) and reversed phase (liquid paraffin impregnated) were used.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1253-1260
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• 2022

This paper reports the control method for the core power of the China initiative Accelerator Driven System (CiADS) facility. In the CiADS facility, an intense external neutron source provided by a proton accelerator coupled to a spallation target is used to drive a sub-critical reactor. Without any control rod inside the sub-critical reactor, the core power is controlled by adjusting the proton beam intensity. In order to continuously change the beam intensity, an adjustable aperture is considered to be used at the Low Energy Beam Transport (LEBT) line of the accelerator. The aperture size is adjusted based on the Proportional Integral Derivative (PID) controllers, by comparing either the setting beam intensity or the setting core power with the measured value. To evaluate the proposed control method, a CiADS core model is built based on the point reactor kinetics model with six delayed neutron groups. The simulations based on the CiADS core model have indicated that the core power can be controlled stably by adjusting the aperture size. The response time in the adjustment of the core power depends mainly on the adjustment time of the beam intensity.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3732-3753
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• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.

• Journal of Energy Engineering
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• v.25 no.2
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• pp.52-64
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• 2016

The production of nuclear energy from thorium which is non-fissile material was a main issue until the middle of 1970's, because of the thorium's abundance as energy resources, its capability of breeding fissile material U233, and the reduction of long-lived actinides. However, to use thorium as nuclear fuel, some obstacles such as the necessities of external neutron source and long-term neutron irradiation for effective breeding, and the production of high radioactive isotopes in the course of thorium breeding cycle should be overcome. The difficulties to resolve these cons of thorium cycle became the reason of interruption of the related researches in the middle of 1970's. But in the 21st century, the change of societal perspective regarding nuclear energy and the appearance of accelerator-driven nuclear reactor shift those cons into pros and rehabilitate the study of thorium. The high activity of thorium cycle turned out to be a good option as higher resistance and easier detectibility of nuclear proliferation and the employment of subcritical accelerator-driven reactor as external neutron sources is considered to enhance the nuclear safety. In this study we compare the thorium cycle with the currently-used uranium cycle and analyze the technical status and perspective of thorium researches which use accelerator-driven reactors.

• Proceedings of the Korean Vacuum Society Conference
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• 2013.02a
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• pp.537-537
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• 2013

Cyclotron-accelerated ion beams are used for various researches, such as nuclear physics, nuclear chemistry, biotechnology, and material sciences including radio-isotope production. Recently considerable applications are asked to the cyclotron development undertaken to meet user requirements of various ions'energies, intensities, and their pulsed beams. For instance, a cocktail beam acceleration technique rapidly changing the ion species and energies was developed to irradiating integrated circuit chips. Also a chopping system in a cyclotron injection line is considered for producing a pulsed ion beam with a relatively long period compared with that generated by the resonance frequency. For the research in neutron time-of-flight measurement, a single-pulsed beam with a repetition interval of the order of mili-seconds or longer is necessary to have a good resolution and to remove background events. In this paper a feasibility of pulsed beam with an external ion source is simulated by adopting a combination system of a chopper accompanying with a bunching stage in the injection line and an additional chopper after the exit of the cyclotron in order to produce beam pulses with a range of $1{\mu}s{\sim}1ms$ periods from a resonance RF cycle. The pulseperiod will be adjusted by chopping the number of beam bunches from the injected pulses in the injection line. However, the longer pulses will have reduced number of beam pulses and sacrificed beam currents. Because the beam users need an intense single pulsed beam, a careful tuning of the acceleration phase and a high-intense external ion source are necessary to achieve an intense single-pulsed beam from the cyclotron. It is essential to strictly match the acceleration phase of injected beams in the central region of the cyclotron to improve its efficiency. An effect of space charge at each pulse from the ion source will be also considered.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.19-26
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• 2020

Small nuclear reactor features higher power capacity, longer operation life than conventional power sources. It could be an ideal alternative of existing power source applied for special equipment for terrestrial or underwater missions. In this paper, a 25kWe heat pipe cooled reactor power source applied for multiple use is preliminary designed. Based on the design, a thermal-hydraulic analysis code for heat pipe cooled reactor is developed to analyze steady and transient performance of the designed nuclear reactor. For reactor design, UN fuel with 65% enrichment and potassium heat pipes are adopted in the reactor core. Tungsten and LiH are adopted as radiation shield on both sides of the reactor core. The reactor is controlled by 6 control drums with B₄C neutron absorbers. Thermoelectric generator (TEG) converts fission heat into electricity. Cooling water removes waste heat out of the reactor. The thermal-hydraulic characteristics of heat pipes are simulated using thermal resistance network method. Thermal parameters of steady and transient conditions, such as the temperature distribution of every key components are obtained. Then the postulated reactor accidents for heat pipe cooled reactor, including power variation, single heat pipe failure and cooling channel blockage, are analyzed and evaluated. Results show that all the designed parameters satisfy the safety requirements. This work could provide reference to the design and application of the heat pipe cooled nuclear power source.

• Journal of radiological science and technology
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• v.37 no.3
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• pp.223-231
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• 2014

Medical linear accelerator is widely used in radiation treatment field, and high energy photons, above 10 MV nominal accelerator voltage, are commonly utilized for the radiation treatment. However, these high energy photons lead the photo-nuclear reaction and the generation of photo-neutrons are accompanied. Thus, these problematic factors are issued in the view of radiation protection. Therefore, linear accelerator and radiation treatment room are simulated from MCNPX program in this study. The measurement points of interest are selected and analyzed, and the resulting effects derived from the properties of photo-neutron are evaluated. Therefore, we realized that the number of generating photo-neutrons was decreased by depending on the distance from the source. No matter what the nominal energy is set, the rates thermal neutrons to fast neutrons are marginal. It is founded that the amount of the thermal neutrons were decreased by depending on the distance from the source.

• The Bulletin of The Korean Astronomical Society
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• v.36 no.2
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• pp.147.2-147.2
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• 2011

We have initiated a long-term identification campaign of supernova remnant candidates in X-ray regime. In the short-listed unidentified sources from the ROSAT All Sky Survey, we have chosen the brightest candidate, G308.3-1.4, as our pilot target for a dedicated investigation with Chandra X-ray Observatory. Our observation has revealed an incomplete shell-like X-ray structure which well-correlated with the radio feature. Together with the spectral properties of a shocked heated plasma, we confirm that G308.3-1.4 is indeed a supernova remnant. A bright X-ray point source which locates close to the remnant center is also uncovered in this observation. Its spectral behavior conform with those observed in a rare class of neutron stars. The properties of its optical/infrared counterpart suggests the evidence for a late-type companion star. Interestingly, possible excesses in B-band and H-alpha have been found which indicate this can be an accretion-powered system. With the further support from the putative periodicity of ~1.4 hrs, this source can possibly provide the direct evidence of a binary system survived in a supernova explosion for the first time.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.783-796
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• 2014

The coarse mesh finite difference (CMFD) method is applied to the discontinuous finite element method based discrete ordinate calculation for source convergence acceleration. The three-dimensional (3-D) DFEM-Sn code FEDONA is developed for general geometry applications as a framework for the CMFD implementation. Detailed methods for applying the CMFD acceleration are established, such as the method to acquire the coarse mesh flux and current by combining unstructured tetrahedron elements to rectangular coarse mesh geometry, and the alternating calculation method to exchange the updated flux information between the CMFD and DFEM-Sn. The partial current based CMFD (p-CMFD) is also implemented for comparison of the acceleration performance. The modified p-CMFD method is proposed to correct the weakness of the original p-CMFD formulation. The performance of CMFD acceleration is examined first for simple two-dimensional multigroup problems to investigate the effect of the problem and coarse mesh sizes. It is shown that smaller coarse meshes are more effective in the CMFD acceleration and the modified p-CMFD has similar effectiveness as the standard CMFD. The effectiveness of CMFD acceleration is then assessed for three-dimensional benchmark problems such as the IAEA (International Atomic Energy Agency) and C5G7MOX problems. It is demonstrated that a sufficiently converged solution is obtained within 7 outer iterations which would require 175 iterations with the normal DFEM-Sn calculations for the IAEA problem. It is claimed that the CMFD accelerated DFEM-Sn method can be effectively used in the practical eigenvalue calculations involving general geometries.