• Title/Summary/Keyword: MCNP

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A Criticality Analysis of the GBC-32 Dry Storage Cask with Hanbit Nuclear Power Plant Unit 3 Fuel Assemblies from the Viewpoint of Burnup Credit

  • Yun, Hyungju;Kim, Do-Yeon;Park, Kwangheon;Hong, Ser Gi
    • Nuclear Engineering and Technology
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    • v.48 no.3
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    • pp.624-634
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    • 2016
  • Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that $k_{eff}$ values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

Prismatic-core advanced high temperature reactor and thermal energy storage coupled system - A preliminary design

  • Alameri, Saeed A.;King, Jeffrey C.;Alkaabi, Ahmed K.;Addad, Yacine
    • Nuclear Engineering and Technology
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    • v.52 no.2
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    • pp.248-257
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    • 2020
  • This study presents an initial design for a novel system consisting in a coupled nuclear reactor and a phase change material-based thermal energy storage (TES) component, which acts as a buffer and regulator of heat transfer between the primary and secondary loops. The goal of this concept is to enhance the capacity factor of nuclear power plants (NPPs) in the case of high integration of renewable energy sources into the electric grid. Hence, this system could support in elevating the economics of NPPs in current competitive markets, especially with subsidized solar and wind energy sources, and relatively low oil and gas prices. Furthermore, utilizing a prismatic-core advanced high temperature reactor (PAHTR) cooled by a molten salt with a high melting point, have the potential in increasing the system efficiency due to its high operating temperature, and providing the baseline requirements for coupling other process heat applications. The present research studies the neutronics and thermal hydraulics (TH) of the PAHTR as well as TH calculations for the TES which consists of 300 blocks with a total heat storage capacity of 150 MWd. SERPENT Monte Carlo and MCNP5 codes carried out the neutronics analysis of the PAHTR which is sized to have a 5-year refueling cycle and rated power of 300 MWth. The PAHTR has 10 metric tons of heavy metal with 19.75 wt% enriched UO2 TRISO fuel, a hot clean excess reactivity and shutdown margin of $33.70 and -$115.68; respectively, negative temperature feedback coefficients, and an axial flux peaking factor of 1.68. Star-CCM + code predicted the correct convective heat transfer coefficient variations for both the reactor and the storage. TH analysis results show that the flow in the primary loop (in the reactor and TES) remains in the developing mixed convection regime while it reaches a fully developed flow in the secondary loop.

Monte Carlo Simulation of Phytosanitary Irradiation Treatment for Mangosteen Using MRI-based Geometry

  • Oh, Se-Yeol;Kim, Jongsoon;Kwon, Soon-Hong;Chung, Sung-Won;Kwon, Soon-Goo;Park, Jong-Min;Choi, Won-Sik
    • Journal of Biosystems Engineering
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    • v.39 no.3
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    • pp.205-214
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    • 2014
  • Purpose: Phytosanitary irradiation treatment can effectively control regulated pests while maintaining produce quality. The objective of this study was to establish the best irradiation treatment for mangosteen, a popular tropical fruit, using a Monte Carlo simulation. Methods: Magnetic resonance image (MRI) data were used to generate a 3-D geometry to simulate dose distributions in a mangosteen using a radiation transport code (MCNP5). Microsoft Excel with visual basic application (VBA) was used to divide the image data into seed, flesh, and rind. Radiation energies used for the simulation were 10 MeV (high-energy) and 1.35 MeV (low-energy) for the electron beam, 5 MeV for X-rays, and 1.25 MeV for gamma rays from Co-60. Results: At 5 MeV X-rays and 1.25 MeV gamma rays, all areas (seeds, flesh, and rind) were irradiated ranging from 0.3 ~ 0.7 kGy. The average doses decreased as the number of fruit increased. For a 10 MeV electron beam, the dose distribution was biased: the dose for the rind where the electrons entered was $0.45{\pm}0.03$ kGy and the other side was $0.24 {\pm}0.10$ kGy. Use of an electron kinetic energy absorber improved the dose distribution in mangosteens. For the 1.35 MeV electron beam, the dose was shown only in the rind on the irradiated side; no significant dose was found in the flesh or seeds. One rotation of the fruit while in front of the beam improved the dose distribution around the entire rind. Conclusion: These results are invaluable for determining the ideal irradiation conditions for phytosanitary irradiation treatment of tropical fruit.

DESIGN OPTIMIZATION OF RADIATION SHIELDING STRUCTURE FOR LEAD SLOWING-DOWN SPECTROMETER SYSTEM

  • KIM, JEONG DONG;AHN, SANGJOON;LEE, YONG DEOK;PARK, CHANG JE
    • Nuclear Engineering and Technology
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    • v.47 no.3
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    • pp.380-387
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    • 2015
  • A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as $^{235}U$, $^{239}Pu$, $^{241}Pu$, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux ($>10^{12}n/cm^2{\cdot}s$) neutron source comprised of a high-energy (30 MeV)/high-current (~2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (< $0.06{\mu}Sv/h$), a few shielding materials [high-density polyethylene (HDPE)eBorax, $B_4C$, and $Li_2CO_3$] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

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).

Reevaluation of Photon Activation Yields of 11C, 13N, and 15O for the Estimation of Activity in Gas and Water Induced by the Operation of Electron Accelerators for Medical Use

  • Masumoto, Kazuyoshi;Matsumura, Hiroshi;Kosako, Kazuaki;Bessho, Kotaro;Toyoda, Akihiro
    • Journal of Radiation Protection and Research
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    • v.41 no.3
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    • pp.286-290
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    • 2016
  • Background: Activation of air and water in the electron linear accelerator for medical use has not been considered severely. By the new Japanese regulation for protection of radiation hazard, it became indispensable to evaluate of activation of air and water in the accelerator room. The measurement of induced activity in air and water components in the electron energy region of 10 to 20 MeV is very difficult, because this energy region is close to the threshold energy region of photonuclear reactions. Then, we measured the photonuclear reaction yields of $^{13}N$, $^{15}O$, and $^{11}C$ by using the electron linear accelerator. Obtained data were compared with the data calculated by the Monte Carlo method. Materials and Methods: An activation experiment was performed at the Research Center for Electron Photon Science, Tohoku University. Highly purified $SiO_2$, $Si_3N_4$, and carbon disks were irradiated for 10 minutes by bremsstrahlung converted by a tungsten plate. Induced activity from C, N, and O was obtained. Monte Carlo calculation was performed using MCNP5 and AERY (DCHAIN-SP) to simulate the experimental condition. Cross section data were adopted the KAERI dataset. Results and Discussion: In our experiment in hospital, calculated values were not agreed with experimental values. It might be three possible reasons as the cause of this deference, such as irradiation energy, calculation procedure and cross section data. Obtained data of this work, calculated and experimental values were good agreement with each other within one order. In this work, we used KAERI dataset of photonuclear reaction instead of JENDL. Therefore, it was found that the photonuclear cross section data of light elements are most important for yield calculation in these reactions. Conclusion: Further improvement for calculation using a new dataset JENDL/PD-2015 and considering electron energy spreading will be needed.

Quantitative Evaluation of Radiation Dose Rates for Depleted Uranium in PRIDE Facility

  • Cho, Il Je;Sim, Jee Hyung;Kim, Yong Soo
    • Journal of Radiation Protection and Research
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    • v.41 no.4
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    • pp.378-383
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    • 2016
  • Background: Radiation dose rates in PRIDE facility is evaluated quantitatively for assessing radiation safety of workers because of large amounts of depleted uranium being handled in PRIDE facility. Even if direct radiation from depleted uranium is very low and will not expose a worker to significant amounts of external radiation. Materials and Methods: ORIGEN-ARP code was used for calculating the neutron and gamma source term being generated from depleted uranium (DU), and the MCNP5 code was used for calculating the neutron and gamma fluxes and dose rates. Results and Discussion: The neutron and gamma fluxes and dose rates due to DU on spherical surface of 30 cm radius were calculated with the variation of DU mass and density. In this calculation, an imaginary case in which DU density is zero was added to check the self-shielding effect of DU. In this case, the DU sphere was modeled as a point. In case of DU mixed with molten salt of 50-250 g, the neutron and gamma fluxes were calculated respectively. It was found that the molten salt contents in DU had little effect on the neutron and the gamma fluxes. The neutron and the gamma fluxes, under the respective conditions of 1 and 5 kg mass of DU, and 5 and $19.1g{\cdot}cm^{-3}$ density of DU, were calculated with the molten salt (LiCl+KCl) of 50 g fixed, and compared with the source term. As the results, similar tendency was found in neutron and gamma fluxes with the variation of DU mass and density when compared with source spectra, except their magnitudes. Conclusion: In the case of the DU mass over 5 kg, the dose rate was shown to be higher than the environmental dose rate. From these results, it is concluded that if a worker would do an experiment with DU having over 5 kg of mass, the worker should be careful in order not to be exposed to the radiation.

A Study on Development of a PIN Semiconductor Detector for Measuring Individual Dose (개인 선량 측정용 PIN 반도체 검출기 개발에 관한 연구)

  • Lee, B.J.;Lee, W.N.;Khang, B.O.;Chang, S.Y.;Rho, S.R.;Chae, H.S.
    • Journal of Radiation Protection and Research
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    • v.28 no.2
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    • pp.87-95
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    • 2003
  • The fabrication process and the structure of PIN semiconductor detectors have been designed optimally by simulation for doping concentration and width of p+ layer, impurities re-contribution due to annealing and the current distribution due to guard ring at the sliced edges. The characteristics to radiation response has been also simulated in terms of Monte Carlo Method. The device has been fabricated on n type, $400\;{\Omega}cm$, orientation <100>, Floating-Zone silicon wafer using the simulation results. The leakage current density of $0.7nA/cm^2/100{\mu}m$ is achieved by this process. The good linearity of radiation response to Cs-137 was kept within the exposure ranges between 5 mR/h and 25 R/h. This proposed process could be applied for fabricating a PIN semiconductor detector for measuring individual dose.

Characterization of a CLYC Detector and Validation of the Monte Carlo Simulation by Measurement Experiments

  • Kim, Hyun Suk;Smith, Martin B.;Koslowsky, Martin R.;Kwak, Sung-Woo;Ye, Sung-Joon;Kim, Geehyun
    • Journal of Radiation Protection and Research
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    • v.42 no.1
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    • pp.48-55
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    • 2017
  • Background: Simultaneous detection of neutrons and gamma rays have become much more practicable, by taking advantage of good gamma-ray discrimination properties using pulse shape discrimination (PSD) technique. Recently, we introduced a commercial CLYC system in Korea, and performed an initial characterization and simulation studies for the CLYC detector system to provide references for the future implementation of the dual-mode scintillator system in various studies and applications. Materials and Methods: We evaluated a CLYC detector with 95% $^6Li$ enrichment using various gamma-ray sources and a $^{252}Cf$ neutron source, with validation of our Monte Carlo simulation results via measurement experiments. Absolute full-energy peak efficiency values were calculated for gamma-ray sources and neutron source using MCNP6 and compared with measurement experiments of the calibration sources. In addition, behavioral characteristics of neutrons were validated by comparing simulations and experiments on neutron moderation with various polyethylene (PE) moderator thicknesses. Results and Discussion: Both results showed good agreements in overall characteristics of the gamma and neutron detection efficiencies, with consistent ~20% discrepancy. Furthermore, moderation of neutrons emitted from $^{252}Cf$ showed similarities between the simulation and the experiment, in terms of their relative ratios depending on the thickness of the PE moderator. Conclusion: A CLYC detector system was characterized for its energy resolution and detection efficiency, and Monte Carlo simulations on the detector system was validated experimentally. Validation of the simulation results in overall trend of the CLYC detector behavior will provide the fundamental basis and validity of follow-up Monte Carlo simulation studies for the development of our dual-particle imager using a rotational modulation collimator.

A Concise Design for the Irradiation of U-10Zr Metallic Fuel at a Very Low Burnup

  • Guo, Haibing;Zhou, Wei;Sun, Yong;Qian, Dazhi;Ma, Jimin;Leng, Jun;Huo, Heyong;Wang, Shaohua
    • Nuclear Engineering and Technology
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    • v.49 no.4
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    • pp.734-743
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    • 2017
  • In order to investigate the swelling behavior and fuel-cladding interaction mechanism of U-10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel-cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal-hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.