• Title/Summary/Keyword: MCNP simulations

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Conceptual design of neutron measurement system for input accountancy in pyroprocessing

  • Lee, Chaehun;Seo, Hee;Menlove, Spencer H.;Menlove, Howard O.
    • Nuclear Engineering and Technology
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    • v.52 no.5
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    • pp.1022-1028
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    • 2020
  • One of the possible options for spent-fuel management in Korea is pyroprocessing, which is a process for electrochemical recycling of spent nuclear fuel. Nuclear material accountancy is considered to be a safeguards measure of fundamental importance, for the purposes of which, the amount of nuclear material in the input and output materials should be measured as accurately as possible by means of chemical analysis and/or non-destructive assay. In the present study, a neutron measurement system based on the fast-neutron energy multiplication (FNEM) and passive neutron albedo reactivity (PNAR) techniques was designed for nuclear material accountancy of a spent-fuel assembly (i.e., the input accountancy of a pyroprocessing facility). Various parameters including inter-detector distance, source-to-detector distance, neutron-reflector material, the structure of a cadmium sleeve around the close detectors, and an air cavity in the moderator were investigated by MCNP6 Monte Carlo simulations in order to maximize its performance. Then, the detector responses with the optimized geometry were estimated for the fresh-fuel assemblies with different 235U enrichments and a spent-fuel assembly. It was found that the measurement technique investigated here has the potential to measure changes in neutron multiplication and, in turn, amount of fissile material.

Experimental validation of a nuclear forensics methodology for source reactor-type discrimination of chemically separated plutonium

  • Osborn, Jeremy M.;Glennon, Kevin J.;Kitcher, Evans D.;Burns, Jonathan D.;Folden, Charles M. III;Chirayath, Sunil S.
    • Nuclear Engineering and Technology
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    • v.51 no.2
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    • pp.384-393
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    • 2019
  • An experimental validation of a nuclear forensics methodology for the source reactor-type discrimination of separated weapons-useable plutonium is presented. The methodology uses measured values of intra-element isotope ratios of plutonium and fission product contaminants. MCNP radiation transport codes were used for various reactor core modeling and fuel burnup simulations. A reactor-dependent library of intra-element isotope ratio values as a function of burnup and time since irradiation was created from the simulation results. The experimental validation of the methodology was achieved by performing two low-burnup experimental irradiations, resulting in distinct fuel samples containing sub-milligram quantities of weapons-useable plutonium. The irradiated samples were subjected to gamma and mass spectrometry to measure several intra-element isotope ratios. For each reactor in the library, a maximum likelihood calculation was utilized to compare the measured and simulated intra-element isotope ratio values, producing a likelihood value which is proportional to the probability of observing the measured ratio values, given a particular reactor in the library. The measured intra-element isotope ratio values of both irradiated samples and its comparison with the simulation predictions using maximum likelihood analyses are presented. The analyses validate the nuclear forensics methodology developed.

Computational design and characterization of a subcritical reactor assembly with TRIGA fuel

  • Asuncion-Astronomo, Alvie;Stancar, Ziga;Goricanec, Tanja;Snoj, Luka
    • Nuclear Engineering and Technology
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    • v.51 no.2
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    • pp.337-344
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    • 2019
  • The TRIGA fuel of the Philippine Research Reactor-1 (PRR-1) will be used in a subcritical reactor assembly (SRA) to strengthen and advance nuclear science and engineering expertise in the Philippines. SRA offers a versatile and safe training and research facility since it can produce neutrons through nuclear fission reaction without achieving criticality. In this work, we used a geometrically detailed model of the PRR-1 TRIGA fuel to design a subcritical reactor assembly and calculate physical parameters of different fuel configurations. Based on extensive neutron transport simulations an SRA configuration is proposed, comprising 44 TRIGA fuel rods arranged in a $7{\times}7$ square lattice. This configuration is found to have a maximum $k_{eff}$ value of $0.95001{\pm}0.00009$ at 4 cm pitch. The SRA is characterized by calculating the 3-dimensional neutron flux distribution and neutron spectrum. The effective delayed neutron fraction and mean neutron generation time of the system are calculated to be $748pcm{\pm}7pcm$ and $41{\mu}s$, respectively. Results obtained from this work will be the basis of the core design for the subcritical reactor facility that will be established in the Philippines.

Comprehensive validation of silicon cross sections

  • Czakoj, Tomas;Kostal, Michal;Simon, Jan;Soltes, Jaroslav;Marecek, Martin;Capote, Roberto
    • Nuclear Engineering and Technology
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    • v.52 no.12
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    • pp.2717-2724
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    • 2020
  • Silicon, especially silicon in the form of SiO2, is a major component of rocks. Final spent fuel storages, which are being designed, are located in suitable rock formations in the Earth's crust. Reduction of the uncertainty of silicon neutron scattering and capture is needed; improved silicon evaluations have been recently produced by the ORNL/IAEA collaboration within the INDEN project. This paper deals with the nuclear data validation of that evaluation performed at the LR-0 reactor by means of critical experiments and measurement of reaction rates. Large amounts of silicon were used both as pure crystalline silicon and SiO2 sand. The critical moderator level was measured for various core configurations. Reaction rates were determined in the largest core configuration. Simulations of the experimental setup were performed using the MCNP6.2 code. The obtained results show the improvement in silicon cross-sections in the INDEN evaluations compared to existing evaluations in major libraries. The new Thermal Scattering Law for SiO2 published in ENDF/B-VIII.0 additionally reduces the discrepancy between calculation and experiments. However, an unphysical peak is visible in the neutron spectrum in SiO2 obtained by calculation with the new Thermal Scattering Law.

CURRENT RESEARCH ON ACCELERATOR-BASED BORON NEUTRON CAPTURE THERAPY IN KOREA

  • Kim, Jong-Kyung;Kim, Kyung-O
    • Nuclear Engineering and Technology
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    • v.41 no.4
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    • pp.531-544
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    • 2009
  • This paper is intended to provide key issues and current research outcomes on accelerator-based Boron Neutron Capture Therapy (BNCT). Accelerator-based neutron sources are efficient to provide epithermal neutron beams for BNCT; hence, much research, worldwide, has focused on the development of components crucial for its realization: neutron-producing targets and cooling equipment, beam-shaping assemblies, and treatment planning systems. Proton beams of 2.5 MeV incident on lithium target results in high yield of neutrons at relatively low energies. Cooling equipment based on submerged jet impingement and micro-channels provide for viable heat removal options. Insofar as beam-shaping assemblies are concerned, moderators containing fluorine or magnesium have the best performance in terms of neutron accumulation in the epithermal energy range during the slowing-down from the high energies. NCT_Plan and SERA systems, which are popular dose distribution analysis tools for BNCT, contain all the required features (i.e., image reconstruction, dose calculations, etc.). However, detailed studies of these systems remain to be done for accurate dose evaluation. Advanced research centered on accelerator-based BNCT is active in Korea as evidenced by the latest research at Hanyang University. There, a new target system and a beam-shaping assembly have been constructed. The performance of these components has been evaluated through comparisons of experimental measurements with simulations. In addition, a new patient-specific treatment planning system, BTPS, has been developed to calculate the deposited dose and radiation flux in human tissue. It is based on MCNPX, and it facilitates BNCT efficient planning based via a user-friendly Graphical User Interface (GUI).

A feasibility study on photo-production of 99mTc with the nuclear resonance fluorescence

  • Ju, Kwangho;Lee, Jiyoung;ur Rehman, Haseeb;Kim, Yonghee
    • Nuclear Engineering and Technology
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    • v.51 no.1
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    • pp.176-189
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    • 2019
  • This paper presents a feasibility study for producing the medical isotope $^{99m}Tc$ using the hazardous and currently wasted radioisotope $^{99}Tc$. This can be achieved with the nuclear resonance fluorescence (NRF) phenomenon, which has recently been made applicable due to high-intensity laser Compton scattering (LCS) photons. In this work, 21 NRF energy states of $^{99}Tc$ have been identified as potential contributors to the photo-production of $^{99m}Tc$ and their NRF cross-sections are evaluated by using the single particle estimate model and the ENSDF data library. The evaluated cross sections are scaled using known measurement data for improved accuracy. The maximum LCS photon energy is adjusted in a way to cover all the significant excited states that may contribute to $^{99m}Tc$ generation. An energy recovery LINAC system is considered as the LCS photon source and the LCS gamma spectrum is optimized by adjusting the electron energy to maximize $^{99m}Tc$ photo-production. The NRF reaction rate for $^{99m}Tc$ is first optimized without considering the photon attenuations such as photo-atomic interactions and self-shielding due to the NRF resonance itself. The change in energy spectrum and intensity due to the photo-atomic reactions has been quantified using the MCNP6 code and then the NRF self-shielding effect was considered to obtain the spectrums that include all the attenuation factors. Simulations show that when a $^{99}Tc$ target is irradiated at an intensity of the order $10^{17}{\gamma}/s$ for 30 h, 2.01 Ci of $^{99m}Tc$ can be produced.

Towards a better understanding of detection properties of different types of plastic scintillator crystals using physical detector and MCNPX code

  • Ayberk Yilmaz;Hatice Yilmaz Alan;Lidya Amon Susam;Baki Akkus;Ghada ALMisned;Taha Batuhan Ilhan;H.O. Tekin
    • Nuclear Engineering and Technology
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    • v.54 no.12
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    • pp.4671-4678
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    • 2022
  • The purpose of this comprehensive research is to observe the impact of scintillator crystal type on entire detection process. For this aim, MCNPX (version 2.6.0) is used for designing of a physical plastic scintillation detector available in our laboratory. The modelled detector structure is validated using previous studies in the literature. Next, different types of plastic scintillation crystals were assessed in the same geometry. Several fundamental detector properties are determined for six different plastic scintillation crystals. Additionally, the deposited energy quantities were computed using the MCNPX code. Although six scintillation crystals have comparable compositions, the findings clearly indicate that the crystal composed of PVT 80% + PPO 20% has superior counting and detecting characteristics when compared to the other crystals investigated. Moreover, it is observed that the highest deposited energy amount, which is a result of the highest collision number in the crystal volume, corresponds to a PVT 80% + PPO 20% crystal. Despite the fact that plastic detector crystals have similar chemical structures, this study found that performing advanced Monte Carlo simulations on the detection discrepancies within the structures can aid in the development of the most effective spectroscopy procedures by ensuring maximum efficiency prior to and during use.

A Study on Non-proportionality of Phoswich Detector Using Monte Carlo Simulation (몬테칼로 전산모사를 이용한 Phoswich 계측기의 비선형성 연구)

  • Kim, Jae-Cheon;Kim, Jong-Kyung;Kim, Soon-Young;Kim, Yong-Kyun;Lee, Woo-Gyo
    • Journal of Radiation Protection and Research
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    • v.29 no.4
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    • pp.263-268
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    • 2004
  • Using the Monte Carlo simulation, a study on the lion-proportionality of the prototype phoswich detector with $2'{\times}2'$ CSI(Tl) and plastic scintillator, which was made by KAERI, has been carried. The defector response functions (DRFs) calculated by simulations were compared with the experimental measurement on the $^{137}Cs\;and\;^{60}Co$. To precisely simulate the DRF for the phoswich, the CSI(Tl) non-proportionality was calculated using the electron response and the simplified electron cascade sequence for treating the photoelectric absorption event. The resulting DRFs of $^{137}Cs\;and\;^{60}Co$ sources obtained by simulations were compared with experiments for verification. For $^{137}Cs$, gamma-ray responses simulated by MCNP5 are generally good agreement with the measured ones. But the DRF of $^{60}Co$ does not match well with the results of experiment in the energy region below second peak due to the coincidence effect of two gamma-rays (1.17 MeV and 1.33 MeV). Through the analysis of the non-proportionality of CsI(Tl) in the prototype phoswich, the improved DRFs considering non-proportionality were produced and the simulation results were verified using the experimental measurements. However, to more precisely reproduce the DRF for the phoswich, further studies in relation to the electron channeling effect and the Doppler broadening effect of a scintillator are still needed as well as considering that effect of the transfer contribution.

Numerical Calculations of IASCC Test Worker Exposure using Process Simulations (공정 시뮬레이션을 이용한 조사유기응력부식균열 시험 작업자 피폭량의 전산 해석에 관한 연구)

  • Chang, Kyu-Ho;Kim, Hae-Woong;Kim, Chang-Kyu;Park, Kwang-Soo;Kwak, Dae-In
    • Journal of the Korean Society of Radiology
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    • v.15 no.6
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    • pp.803-811
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    • 2021
  • In this study, the exposure amount of IASCC test worker was evaluated by applying the process simulation technology. Using DELMIA Version 5, a commercial process simulation code, IASCC test facility, hot cells, and workers were prepared, and IASCC test activities were implemented, and the cumulative exposure of workers passing through the dose-distributed space could be evaluated through user coding. In order to simulate behavior of workers, human manikins with a degree of freedom of 200 or more imitating the human musculoskeletal system were applied. In order to calculate the worker's exposure, the coordinates, start time, and retention period for each posture were extracted by accessing the sub-information of the human manikin task, and the cumulative exposure was calculated by multiplying the spatial dose value by the posture retention time. The spatial dose for the exposure evaluation was calculated using MCNP6 Version 1.0, and the calculated spatial dose was embedded into the process simulation domain. As a result of comparing and analyzing the results of exposure evaluation by process simulation and typical exposure evaluation, the annual exposure to daily test work in the regular entrance was predicted at similar levels, 0.388 mSv/year and 1.334 mSv/year, respectively. Exposure assessment was also performed on special tasks performed in areas with high spatial doses, and tasks with high exposure could be easily identified, and work improvement plans could be derived intuitively through human manikin posture and spatial dose visualization of the tasks.

Evaluation and Verification of the Attenuation Rate of Lead Sheets by Tube Voltage for Reference to Radiation Shielding Facilities (방사선 방어시설 구축 시 활용 가능한 관전압별 납 시트 차폐율 성능평가 및 실측 검증)

  • Ki-Yoon Lee;Kyung-Hwan Jung;Dong-Hee Han;Jang-Oh Kim;Man-Seok Han;Jong-Won Gil;Cheol-Ha Baek
    • Journal of the Korean Society of Radiology
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    • v.17 no.4
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    • pp.489-495
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    • 2023
  • Radiation shielding facilities are constructed in locations where diagnostic radiation generators are installed, with the aim of preventing exposure for patients and radiation workers. The purpose of this study is seek to compare and validate the trend of attenuation thickness of lead, the primary material in these radiation shielding facilities, at different maximum tube voltages by Monte Carlo simulations and measurement. We employed the Monte Carlo N-Particle 6 simulation code. Within this simulation, we set a lead shielding arrangement, where the distance between the source and the lead sheet was set at 100 cm and the field of view was set at 10 × 10 cm2. Additionally, we varied the tube voltages to encompass 80, 100, 120, and 140 kVp. We calculated energy spectra for each respective tube voltage and applied them in the simulations. Lead thicknesses corresponding to attenuation rates of 50, 70, 90, and 95% were determined for tube voltages of 80, 100, 120, and 140 kVp. For 80 kVp, the calculated thicknesses for these attenuation rates were 0.03, 0.08, 0.21, and 0.33 mm, respectively. For 100 kVp, the values were 0.05, 0.12, 0.30, and 0.50 mm. Similarly, for 120 kVp, they were 0.06, 0.14, 0.38, and 0.56 mm. Lastly, at 140 kVp, the corresponding thicknesses were 0.08, 0.16, 0.42, and 0.61 mm. Measurements were conducted to validate the calculated lead thicknesses. The radiation generator employed was the GE Healthcare Discovery XR 656, and the dosimeter used was the IBA MagicMax. The experimental results showed that at 80 kVp, the attenuation rates for different thicknesses were 43.56, 70.33, 89.85, and 93.05%, respectively. Similarly, at 100 kVp, the rates were 52.49, 72.26, 86.31, and 92.17%. For 120 kVp, the attenuation rates were 48.26, 71.18, 87.30, and 91.56%. Lastly, at 140 kVp, they were measured 50.45, 68.75, 89.95, and 91.65%. Upon comparing the simulation and experimental results, it was confirmed that the differences between the two values were within an average of approximately 3%. These research findings serve to validate the reliability of Monte Carlo simulations and could be employed as fundamental data for future radiation shielding facility construction.