• Journal of Radiation Protection and Research
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• 제14권1호
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• pp.46-55
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• 1989

사용후 핵연료 용기에 대한 중성자 차폐 해석을 위하여 각분할법 코드인 ONEDANT 및 XSDRN과 몬테칼로 코드인 MCNP를 사용하였다. ORIGEN-S로 부터 결정된 선원항이 ONEDANT및 XSDRN에 각각 이용되었고, MCNP에 입력되는 선원항으로는 ONEDANT와 XSDRN으로 부터 계산된 중성자 스펙트럼을 사용하였으며, 중성자 에너지군은 27군과 10군으로 하였다. 감손 우라늄을 중성자 차폐 물질로 사용하였을 경우, MCNP의 계산 결과에 대하여 ONEDANT의 계산결과는 10%, XSDRN은 20% 이내에서 접근하였다. 또한, MCNP의 계산 결과에 의하면, 고려한 중성자 차폐물질의 성능은 감손 우라늄, 철, 그리고 납의 순으로 좋은 것으로 나타났다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
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• pp.76-85
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• 2005

The conceptual study of Accelerator Driven System (ADS) had lasted for about five years and ended in 1999 in China. As one project of 'the major state basic research program (973)' in energy domain, which is sponsored by the China Ministry of Science and Technology (MOST), a five years program of basic research for ADS physics and related technology has been launched since 2000 and passed national review last month. CIAE (China Institute of Atomic Energy), IHEP (Institute of High Energy Physics), PKU-IHIP (Institute of Heavy Ion Physics in Peking University) and other institutions are jointly carrying on the research. The research activities are focused on HPPA physics and technology, reactor physics of external source driven sub-critical assembly, nuclear data base and material study. For HPPA, a high current injector consisting of an ECR ion source, LEBT and a RFQ accelerating structure of 3.5MeV has been built. In reactor physics study, a series of neutron multiplication experimental study has been carried out and is being carrying on. The VENUS facility has been constructed as the basic experimental platform for the neutronics study in ADS blanket. It's a zero power sub-critical neutron multiplying assembly driven by external neutron produced by a pulsed neutron generator. The theoretical, experimental and simulation study on nuclear data, material properties and nuclear fuel circulation related to ADS is carrying on to provide the database for ADS system analysis. The main results on ADS related researches will be reported.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2852-2859
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• 2020

This paper deals with the simulation-based design optimization and experimental validation of the characteristics of an in-core silver Self-Powered Neutron Detector (SPND). Optimized dimensions of the SPND are determined by combining Monte Carlo simulations and analytical methods. As a first step, the Monte Carlo transport code MCNPX is used to follow the trajectory and fate of the neutrons emitted from an external source. This simulation is able to seamlessly integrate various phenomena, including neutron slowing-down and shielding effects. Then, the expected number of beta particles and their energy spectrum following a neutron capture reaction in the silver emitter are fetched from the TENDEL database using the JANIS software interface and integrated with the data from the first step to yield the origin and spectrum of the source electrons. Eventually, the MCNPX transport code is used for the Monte Carlo calculation of the ballistic current of beta particles in the various regions of the SPND. Then, the output current and the maximum insulator thickness to avoid breakdown are determined. The optimum design of the SPND is then manufactured and experimental tests are conducted. The calculated design parameters of this detector have been found in good agreement with the obtained experimental results.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1425-1430
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• 2024

This study develops superheated emulsion detectors that are both sensitive to fast neutrons, and thermal neutrons owing to the exergonic ⁶₃Li(n, α)³₁H capture reaction caused by the ⁶Li-containing compound dispersed throughout the gel-like medium. The experimental research was conducted on two SEDs. One detector was an ordinary Freon-12 detector and the other was a Freon-12 detector containing 3.4 % (by weight) LiCl. In order to investigate the sensitivity of lithium-containing SEDs to thermal neutrons, two types of SEDs were simultaneously exposed to various flux levels of thermal neutrons from ²⁴¹Am-Be neutron source inside a cylindrical tank filled with water. A Boron-lined proportional counter was used to estimate the thermal neutron flux and the relevant MCNP code was developed for flux and dose calculations in the prepared set-up around ²⁴¹Am-Be source. The results demonstrate that there is a proportional relationship between the variations of SED response and the change in thermal neutron flux and dose. Also, the sensitivity of SED was estimated.

• 전자공학회논문지 IE
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• 제45권2호
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• pp.6-12
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• 2008

본 연구에서는 아스팔트 함량의 중요성을 인식하여 현재 법적 규제 면제치인 100[${\mu}Ci$]이하의 방사성 동위원소를 이용한 아스팔트 함량측정기의 개발을 목표로 하였다. 이를 위하여 중성자선원의 위치에 따라 아스팔트 함량에 변화를 주는 3가지의 분야로 나누었다. 먼저, 아스팔트 혼합물과 중성자 선원과의 간격을 줄일 경우, 반사체 설치의 경우, 이력수를 변화시켰을 경우로 나누어서 컴퓨터 시뮬레이션을 통하여 살펴보았으며, 만족스러운 오차범위 결과를 얻어 아스팔트 함량측정기기의 개발을 위한 설계 자료로 활용하고자하였다.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.304-317
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• 2016

An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan), a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcrofte-Walton type accelerator, which generates the external neutron source by deuteriu-metritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

• Journal of Radiation Protection and Research
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• 제28권4호
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• pp.327-335
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• 2003

The commercially available neutron survey meter, the REM500, which uses a tissue equivalent proportional counter (TEPC) and the self-constructed TEPC were used to determine the microdosimetric quantities of several neutron calibration fields at Korea Atomic Energy Research Institute (KAERI). Microdosimetric spectra, absorbed dose, dose equivalent as well as quality factor were derived and compared with several neutron fields which were produced by using the shadow objects to make neutron scattered and being used as a kind of realistic neutron calibration fields at KAERI. The response of REM500 as a function of mean energy was evaluated with these neutron fields using the counts measured and the predetermined reference value. The response of the self-made TEPC and the REM500 was compared using one of the neutron calibration filelds of a $^{252}Cf$ source. The reference quantities of scattered neutron calibration fields were determined using a Bonner Sphere (BS). The value of frequency-mean lineal energy, dose-mean lineal energy and quality factor of two $^{252}Cf$ sources (unmoderated and $D_2O$ moderated) were determined to check the differences in the reference neutron fields between KAERI and Pacific Northwest National Laboratory (PNNL, USA) and the results were in good agreement within 1%. It means that there is no big difference in dosimetric quantifies of neutron calibration fields of two laboratories.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.647-653
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• 2020

Stainless steel is used commonly in nuclear applications for shielding radiation, so in this study, three different types of new stainless steel samples were designed and developed. New stainless steel compound ratios were determined by using Monte Carlo Simulation program Geant 4 code. In the sample production, iron (Fe), nickel (Ni), chromium (Cr), silicium (Si), sulphur (S), carbon (C), molybdenum (Mo), manganese (Mn), wolfram (W), rhenium (Re), titanium (Ti) and vanadium (V), powder materials were used with powder metallurgy method. Total macroscopic cross sections, mean free path and transmission number were calculated for the fast neutron radiation shielding by using (Geant 4) code. In addition to neutron shielding, the gamma absorption parameters such as mass attenuation coefficients (MACs) and half value layer (HVL) were calculated using Win-XCOM software. Sulfuric acid abrasion and compressive strength tests were carried out and all samples showed good resistance to acid wear and pressure force. The neutron equivalent dose was measured using an average 4.5 MeV energy fast neutron source. Results were compared to 316LN type stainless steel, which commonly used in shielding radiation. New stainless steel samples were found to absorb neutron better than 316LN stainless steel at both low and high temperatures.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.681-688
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• 2020

The telegrapher's theory was used to develop a new formulation for the neutron noise equation. Telegrapher's equation is supposed to demonstrate a more realistic approximation for neutron transport phenomena, especially in comparison to the diffusion theory. The physics behind such equation implies that the signal propagation speed is finite, instead of the infinite as in the case of ordinary diffusion. This paper presents the theory and results of the development of a new method for calculation of the neutron noise using the telegrapher's equation as its basis. In order to investigate the differences and strengths of the new method against the diffusion based neutron noise, a comparison was done between the behaviors of two methods. The neutron noise based on SN transport considered as a precision measuring point. The Green's function technique was used to calculate the neutron noise based on telegrapher's and diffusion methods as well as the transport. The amplitude and phase of Green's function associated with the properties of the medium and frequency of the noise source were obtained and their behavior was compared to the results of the transport. It was observed, the differences in some cases might be considerable. The effective speed of propagation for the noise perturbations were evaluated accordingly, resulting in considerable deviations in some cases.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1593-1603
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• 2023

In deterministic and Monte Carlo transport codes, b-delayed emission is included using a group structure where all of the precursors are grouped together in 6 groups or families, but given the increase in computational power, nowadays there is no reason to keep this structure. Furthermore, there have been recent efforts to compile and evaluate all the available b-delayed neutron emission data and to measure new and improved data on individual precursors. In order to be able to perform a transient Monte Carlo simulation, data from individual precursors needs to be implemented in a transport code. This work is the first step towards the development of a tool to explore the effect of individual precursors in a fissile system. In concrete, individual precursor data is included by expanding the capabilities of the open source Monte Carlo code OpenMC. In the modified code - named Time Dependent OpenMC or OpenMC(TD)- time dependency related to β-delayed neutron emission was handled by using forced decay of precursors and combing of the particle population. The data for continuous energy neutron cross-sections was taken from JEFF-3.1.1 library. Regarding the data needed to include the individual precursors, cumulative yields were taken from JEFF-3.1.1 and delayed neutron emission probabilities and delayed neutron spectra were taken from ENDF-B/VIII.0. OpenMC(TD) was tested in a monoenergetic system, an energy dependent unmoderated system where the precursors were taken individually or in a group structure, and in a light-water moderated energy dependent system, using 6-groups, 50 and 40 individual precursors. Neutron flux as a function of time was obtained for each of the systems studied. These results show the potential of OpenMC(TD) as a tool to study the impact of individual precursor data on fissile systems, thus motivating further research to simulate more complex fissile systems.