The Experts below are selected from a list of 5031 Experts worldwide ranked by ideXlab platform

Dong-seong Sohn - One of the best experts on this subject based on the ideXlab platform.

  • Fuel PERFORMANCE CODE COSMOS FOR ANALYSIS OF LWR UO2AND MOX Fuel
    Nuclear Engineering and Technology, 2011
    Co-Authors: Byung-ho Lee, Yang-hyun Koo, Jin-sik Cheon, Young-wook Tahk, Dong-seong Sohn
    Abstract:

    The paper briefs a Fuel performance code, COSMOS, which can be utilized for an analysis of the thermal behavior and fission gas release of Fuel, up to a high burnup. Of particular concern are the models for the Fuel thermal conductivity, the fission gas release, and the cladding corrosion and creep in UO2 Fuel. In addition, the code was developed so as to consider the inhomogeneity of MOX Fuel, which requires restructuring the thermal conductivity and fission gas release models. These improvements enhanced COSMOS’s precision for predicting the in-pile behavior of MOX Fuel. The COSMOS code also extends its applicability to the instrumented Fuel test in a research reactor. The various in-pile test results were analyzed and compared with the code’s prediction. The database consists of the UO2 irradiation test up to an ultra-high burnup, power ramp test of MOX Fuel, and instrumented MOX Fuel test in a research reactor after base irradiation in a commercial reactor. The comparisons demonstrated that the COSMOS code predicted the in-pile behaviors well, such as the Fuel temperature, rod internal pressure, fission gas release, and cladding properties of MOX and UO2 Fuel. This sufficient accuracy reveals that the COSMOS can be utilized by both Fuel vendors for Fuel design, and license organizations for an understanding of Fuel in-pile behaviors.

  • irradiation test of MOX Fuel in the halden reactor and the analysis of measured data with the Fuel performance code cosmos
    Nuclear Engineering and Technology, 2005
    Co-Authors: Wolfgang Wiesenack, Byung-ho Lee, Dong-seong Sohn
    Abstract:

    The burning-out of excess plutonium from the reprocessing of spent nuclear Fuel and from the dismantlement of nuclear weapons is recently emphasized due to the difficulties in securing the final repository for the spent Fuel and the necessity to consume the ex-weapons plutonium. An irradiation test in the Halden reactor was launched by the OECD Halden Reactor Project (HRP) to investigate the in-pile behavior of plutonium-embedded Fuel as a form of mixed oxide (MOX) and of inert matrix Fuel (IMF). The first cycle of irradiation was successfully accomplished with good integrity of test Fuel rods and without any undesirable fault of instrumentations. The test results revealed that the MOX Fuel is more stable under irradiation environments than IMF. In addition, MOX Fuel shows lower thermal resistance due to its better thermal conductivity than IMF. The on-line measured in-pile performance data of attrition milled MOX Fuel are used in the analysis of the in-pile performance of the Fuel with the Fuel performance code, COSMOS. The COSMOS code has been developed for the analysis of MOX Fuel as well as Fuel up to high burnup and showed good capability to analyze the in-reactor behavior of MOX Fuel even with different instrumentation.

  • A Thermal Conductivity Model for LWR MOX Fuel and Its Verification Using In-pile Data
    Nuclear Engineering and Technology, 2002
    Co-Authors: Byung-ho Lee, Yang-hyun Koo, Jin-sik Cheon, Hyung-kook Joo, Dong-seong Sohn
    Abstract:

    The MOX Fuel for LWR is fabricated either by direct mechanical blending of UO and PuO or by two stage mixing. Hence Pu-rich particles, whose Pu concentrations are higher than pellet average one and whose size distribution depends on a specific fabrication method, are inevitably dispersed in MOX pellet. Due to the inhomogeneous microstructure of MOX Fuel, the thermal conductivity of LWR MOX Fuel scatters from 80 to 100 % of UO Fuel. This paper describes a mechanistic thermal conductivity model for MOX Fuel by considering this inhomogeneous microstructure and presents an explanation for the wide scattering of measured MOX Fuel＇s thermal conductivity. The developed model has been incorporated into a KAERI＇s Fuel performance code, COSMOS, and then evaluated using the measured in-pile data for MOX Fuel. The database used for verification consists of homogeneous MOX Fuel at beginning-of-life and inhomogeneous MOX Fuel at high turnup. The COSMOS code predicts the thermal behavior of MOX Fuel well except for the irradiation test accompanying substantial fission gas release. The over-prediction with substantial fission gas release seems to suggest the need for the introduction of a recovery factor to a term that considers the burnup effect on thermal conductivity.

  • A Unified Thermal Conductivity Model of LWR MOX Fuel Considering Its Microstructural Characteristics
    Journal of Nuclear Science and Technology, 2002
    Co-Authors: Byung-ho Lee, Yang-hyun Koo, Jin-sik Cheon, Dong-seong Sohn
    Abstract:

    LWR MOX Fuel, fabricated either by direct mechanical blending of UO2 and PuO2 powders or by two stage mixing, inevitably has Pu-rich particles dispersed in the matrix of Fuel pellet, whose Pu concentrations are higher than pellet average one and whose size distribution depends on specific fabrication method. This paper describes a mechanistic thermal conductivity model of MOX Fuel by considering this inhomogeneous microstructure and explains the wide scattering of measured MOX's thermal conductivity. The developed model has been incorporated into a KAERI's Fuel performance code, COSMOS, and then evaluated using measured data for irradiated MOX Fuel. The measured temperatures were obtained from both homogeneous MOX at beginning of life and inhomogeneous MOX at high burnup. The COSMOS predicts the thermal behavior of MOX Fuel well except for irradiation test accompanying substantial fission gas release. This over-prediction of Fuel temperature, that is, the under-prediction of thermal conductivity, in the M...

  • Modeling and parametric studies of the effect of inhomogeneity on fission gas release in LWR MOX Fuel
    Annals of Nuclear Energy, 2002
    Co-Authors: Yang-hyun Koo, Byung-ho Lee, Jin-sik Cheon, Dong-seong Sohn
    Abstract:

    Abstract To analyze the effect of an inhomogeneous mixture of an PuO 2 powder on fission gas release in MOX Fuel, a model has been developed using the assumption that gas release mechanism in Pu-rich particles is identical with that in UO 2 Fuel. A parametric study was performed to see the respective effect of the number density, size and fraction of Pu retained in the Pu-rich particles on gas release in MOX Fuel. The model shows that, for the condition of all the other remaining parameters being fixed, more gas is released in a MOX Fuel for lower number density of, smaller size of, and larger fraction of Pu retained in, the Pu-rich particles. However, there exists some condition or combination of parameters for which the effect of inhomogeneity on gas release is negligible depending on the characteristics of MOX Fuel. Comparison with measured data for OCOM MOX Fuel shows that the present model can predict the level of gas release in MOX Fuel once the release mechanism in the Pu-rich particles is known.

Takashi Namekawa - One of the best experts on this subject based on the ideXlab platform.

  • Fabrication Technology for MOX Fuel Containing AmO2 by an In-cell Remote Process
    Journal of Nuclear Science and Technology, 2004
    Co-Authors: Hiroshi Yoshimochi, Shin-ichi Koyama, Masanao Nemoto, Kenji Mondo, Takashi Namekawa
    Abstract:

    An in-cell remote fabrication technique was developed for MOX Fuel pellets containing 3 and 5% americium (Am-MOX Fuel pellet). The Fuel pellet was fabricated by means of conventional powder metallurgy. A series of Fuel pellet fabrication apparatuses were systematically installed in the alpha-gamma cell (hot cell) to protect workers from a strong γ-ray exposure from 241 Am, and were remotely controlled from a panel in the operation room outside the hot cells as much as possible. From a preliminary UO2 pellet fabrication run, ball milling of powder for 4h, pressing at 4t/cm2 and sintering at 1,700°C for 2h were determined as a good fabrication, but the ball milling time was too short for the UO2 and Am-PuO2 powders of different morphologies to be uniformly mixed. Then, the 5% Am-MOX Fuel pellet of density more than 93% T.D. which is proper to the irradiation in FBR was successfully fabricated by extending the ball milling time for more than 10 h. It was, furthermore, found that the complete cleanup of the p...

  • Thermal conductivity modeling of high burnup MOX Fuel
    Journal of Nuclear Science and Technology, 2002
    Co-Authors: Shinsuke Yamanaka, Sumiko Masuo, Ken Kurosaki, Masayoshi Uno, Kazuya Yamamoto, Takashi Namekawa
    Abstract:

    AbstractThe thermal conductivity of MOX Fuel containing FP precipitates such as oxide and metallic inclusions was evaluated by using a finite element method. (U0.8Pu0.2)O2 and (M0.94Nd0.06)O2 (M: U0.8Pu0.2) were employed as the Fuel matrix, and BaUO3 and Mo-Ru-Rh-Pd alloy as the inclusions. The dispersed phase size and concentration were determined from post irradiation examination results of about 10 at.% burnup MOX Fuel. 2.5 vol.% of the spherical inclusions with the diameter of 25 μm were dispersed in the Fuel matrix cube with a size of 400×400×400 μm. The effect of the inclusions on the thermal conductivity of the Fuel pellet was calculated to be at most several percents, which is about one order lower than that of the dissolved FPs.

A. N. Shmelev - One of the best experts on this subject based on the ideXlab platform.

  • Radiation protection potential of MOX-Fuel doped with 231Pa and Cs radioisotopes.
    Radiation protection dosimetry, 2005
    Co-Authors: E F Kryuchkov, V. B. Glebov, V. A. Apse, A. N. Shmelev
    Abstract:

    The paper addresses the problem of MOX-Fuel self-protection during full cycle of MOX-Fuel management. Under conditions of the closed LWR cycle the proliferation-resistance levels were evaluated for fresh and spent MOX-Fuel with 2 3 1 Pa and Cs feed. As it follows from the paper results, combination of these two admixtures being doped into MOX-Fuel is able to enhance the inherent radiation barrier and to weaken shortcomings of both proliferation deterrents.

  • Radiation protection of MOX-Fuel by doping with 231Pa and 232U*
    Progress in Nuclear Energy, 2005
    Co-Authors: E F Kryuchkov, V. B. Glebov, V. A. Apse, A. N. Shmelev
    Abstract:

    The paper addresses the problem of MOX-Fuel self-protection at the "Spent Fuel Standard" level and more during full cycle of MOX-Fuel management. Under conditions of the closed LWR cycle the proliferation-resistance levels were evaluated for fresh and spent MOX-Fuel doped with Pa-231 and U-232. According to the evaluations, it was derived that cyclic regime with Pa-U feeding at the level of 5% HM makes it possible to achieve the proliferation resistance of fresh MOX-Fuel at the level high enough to protect it against short-term unauthorized actions (receiving the lethal dose for about 10 minutes in the vicinity of MOX-Fuel assembly) and also against the acts of terrorism (receiving the shock lethal dose for a few minutes). (C) 2005 Elsevier Ltd. All rights reserved.

L. J. Jardine - One of the best experts on this subject based on the ideXlab platform.

  • Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia
    2002
    Co-Authors: K. G. Kudinov, A.a. Tretyakov, Y. P. Sorokin, V. V. Bondin, L. F. Manakova, L. J. Jardine
    Abstract:

    In Russia, mixed oxide (MOX) Fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX Fuel and Fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on production capacity of 2.75 tons of weapons plutonium per year to produce the resulting Fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-Fuel production. Making MOX-Fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimatesmore » of MOX Fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX Fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-Fuel production waste is incineration or calcination, alkali sintering, and dissolution of sintered products in nitric acid. Insoluble residues are then mixed with vitrifying components and Pu sludges, vitrified, and sent for storage and disposal.« less

  • Estimate of the Sources of Plutonium-Containing Wastes Generated from MOX Fuel Production in Russia
    2001
    Co-Authors: K. G. Kudinov, A.a. Tretyakov, Y. P. Sorokin, V. V. Bondin, L. F. Manakova, L. J. Jardine
    Abstract:

    In Russia, mixed oxide (MOX) Fuel is produced in a pilot facility ''Paket'' at ''MAYAK'' Production Association. The Mining-Chemical Combine (MCC) has developed plans to design and build a dedicated industrial-scale plant to produce MOX Fuel and Fuel assemblies (FA) for VVER-1000 water reactors and the BN-600 fast-breeder reactor, which is pending an official Russian Federation (RF) site-selection decision. The design output of the plant is based on production capacity of 2.75 tons of weapons plutonium per year to produce the resulting Fuel assemblies: 1.25 tons for the BN-600 reactor FAs and the remaining 1.5 tons for VVER-1000 FAs. It is likely the quantity of BN-600 FAs will be reduced in actual practice. The process of nuclear disarmament frees a significant amount of weapons plutonium for other uses, which, if unutilized, represents a constant general threat. In France, Great Britain, Belgium, Russia, and Japan, reactor-grade plutonium is used in MOX-Fuel production. Making MOX-Fuel for CANDU (Canada) and pressurized water reactors (PWR) (Europe) is under consideration Russia. If this latter production is added, as many as 5 tons of Pu per year might be processed into new FAs in Russia. Many years of work and experience are represented in the estimates of MOX Fuel production wastes derived in this report. Prior engineering studies and sludge treatment investigations and comparisons have determined how best to treat Pu sludges and MOX Fuel wastes. Based upon analyses of the production processes established by these efforts, we can estimate that there will be approximately 1200 kg of residual wastes subject to immobilization per MT of plutonium processed, of which approximately 6 to 7 kg is Pu in the residuals per MT of Pu processed. The wastes are various and complicated in composition. Because organic wastes constitute both the major portion of total waste and of the Pu to be immobilized, the recommended treatment of MOX-Fuel production waste is incineration or calcination, alkali sintering, and dissolution of sintered products in nitric acid. Insoluble residues are then mixed with vitrifying components and Pu sludges, vitrified, and sent for storage and disposal.

Shin-ichi Koyama - One of the best experts on this subject based on the ideXlab platform.

  • Oxidation Behavior of Am-containing MOX Fuel Pellets in Air
    Energy Procedia, 2015
    Co-Authors: Kosuke Tanaka, Hiroshi Yoshimochi, Hiroshi Obayashi, Shin-ichi Koyama
    Abstract:

    Abstract Americium-containing MOX (Am-MOX) Fuels were subjected to heating tests using thermogravimetric and differential thermal analysis measurements in a flowing gas atmosphere of dry air to investigate the effect of Am addition on oxidation behavior of MOX Fuel. Three kinds of Am-MOX Fuel pellets containing 3, 5 and 10 wt.% Am were prepared for the examination together with MOX Fuel and UO2 Fuel pellets as references. Sintered Fuel pellets were heat-treated to adjust the oxygen-to-metal ratio to 2.00 and were crushed into small pieces. The weight gain due to oxidation was monitored and the pulverization behavior of the Fuel pellets was observed. The specimens were analyzed by X-ray diffraction in order to investigate the change of the phase relation. The specimens were subjected to programmed rate heating test (6 K/min) from room temperature to 1073 K. The UO2 pellet specimen was oxidized rapidly to an O/M ratio of 2.67 (i.e. U3O8 was formed) and it was pulverized easily. The specimens of MOX Fuel and Am-MOX Fuel pellets, however, were oxidized gradually to O/M ratio around 2.3 (i.e. MO2+x and/or M4O9 were formed) and there was no pellet crumbling. Although the oxidation rate slightly decreased with increasing the Am content in the Am-MOX Fuels, the oxidation curve shapes of Am-MOX Fuels were similar to the curve of MOX Fuel. The isothermal heating test was carried out for the specimens of MOX, 3 wt.% Am-MOX and 5 wt.% Am-MOX Fuels. The kinetic analysis of the oxidation in the isothermal heating test was evaluated by the Johnson-Mehl equation.

  • effects of interaction between molten zircaloy and irradiated MOX Fuel on the fission product release behavior
    Journal of Nuclear Science and Technology, 2014
    Co-Authors: Kosuke Tanaka, Hiroshi Obayashi, Shuhei Miwa, Isamu Sato, Takashi Hirosawa, Shinichi Sekine, Masahiko Osaka, Shin-ichi Koyama
    Abstract:

    As a first step for obtaining experimental data on the effects of high-temperature chemical interaction on fission product release behavior, we focused on the dissolution of irradiated uranium plutonium mixed oxide (MOX) Fuel by molten zircaloy (Zry) and carried out a heating test under the reducing atmosphere. Pieces of an irradiated MOX Fuel pellet and cladding were subjected to the heating test at 2373 K for five minutes. The fractional release rate of cesium (specifically 137Cs) was monitored during the test and its release behavior was evaluated. The observation of microstructures and measurements of elemental distribution in the heated specimen were also performed. We demonstrated experimentally that the Fuel dissolution by molten Zry accelerated the release of Cs from the Fuel pellets.

  • Fabrication Technology for MOX Fuel Containing AmO2 by an In-cell Remote Process
    Journal of Nuclear Science and Technology, 2004
    Co-Authors: Hiroshi Yoshimochi, Shin-ichi Koyama, Masanao Nemoto, Kenji Mondo, Takashi Namekawa
    Abstract:

    An in-cell remote fabrication technique was developed for MOX Fuel pellets containing 3 and 5% americium (Am-MOX Fuel pellet). The Fuel pellet was fabricated by means of conventional powder metallurgy. A series of Fuel pellet fabrication apparatuses were systematically installed in the alpha-gamma cell (hot cell) to protect workers from a strong γ-ray exposure from 241 Am, and were remotely controlled from a panel in the operation room outside the hot cells as much as possible. From a preliminary UO2 pellet fabrication run, ball milling of powder for 4h, pressing at 4t/cm2 and sintering at 1,700°C for 2h were determined as a good fabrication, but the ball milling time was too short for the UO2 and Am-PuO2 powders of different morphologies to be uniformly mixed. Then, the 5% Am-MOX Fuel pellet of density more than 93% T.D. which is proper to the irradiation in FBR was successfully fabricated by extending the ball milling time for more than 10 h. It was, furthermore, found that the complete cleanup of the p...