Artículos de revistas: "Fission products Analysis"

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Petrovski, A. M., T. N. Korbut, E. A. Rudak y M. O. Kravchenko. "Accounting of the vver-1200 overload influence for fission products activities calculating". Proceedings of the National Academy of Sciences of Belarus, Physical-Technical Series 64, n.º 4 (11 de enero de 2020): 491–96. http://dx.doi.org/10.29235/1561-8358-2019-64-4-491-496.

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Current work is aimed at the analysis of the fission products decay influence during fuel reloading, when calculating the accumulated fission products activity for the VVER-1200 reactor fuel campaign. The Bateman problem solution based technique was used for calculations, within the framework of the two fissile nuclides approximation. The fission products producing process for the VVER-1200 reactor stationary campaign is considered, taking into account the reactor shutdown periods for refueling and without taking them into account (instant reload approximation). It was shown, that the instant reload approximation for fission products activity calculations gives the similar accurate result, as calculations with taking into account the shutdown periods. The results can be used to significantly simplify the calculations of fission product activity accumulation in nuclear power reactors.

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2

Zhou, Tao, Peng Xu, Tian Qi, Xuemeng Qin, Juan Chen y Zhongguang Fu. "Calculation and Analysis of the Source Term of the Reactor Core Based on Multivariate Analysis of Variance". Science and Technology of Nuclear Installations 2021 (3 de junio de 2021): 1–8. http://dx.doi.org/10.1155/2021/8810668.

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The calculation of the core source term is affected by various factors, such as fuel consumption, enrichment, specific power, and operation mode. The activity of lanthanides, fission products, and the photon source strength were calculated using the ORIGEN program. The weights of each factor were calculated by multivariate analysis of variance. The results show that the radioactivity of actinides and fission products increased with the increase in fuel consumption. As enrichment increased, the radioactivity of fission products and actinides decreased. The radioactivity of fission products and actinides increased linearly with the change in specific power, with a correlation coefficient of 1. The changes in fuel consumption and enrichment have little effect on low-energy photons, but significantly affected high-energy photons. The change in specific power has little effect on the photon generation of different energy groups. The operation mode has little effect on the radioactivity of the nucleus and fission products. Multivariate analysis of variance shows that specific power is the most influential factor, followed by enrichment; the least influential factor is fuel consumption.

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3

Auxier, John D., Jacob A. Jordan, S. Adam Stratz, Shayan Shahbazi, Daniel E. Hanson, Derek Cressy y Howard L. Hall. "Thermodynamic analysis of volatile organometallic fission products". Journal of Radioanalytical and Nuclear Chemistry 307, n.º 3 (17 de diciembre de 2015): 1621–27. http://dx.doi.org/10.1007/s10967-015-4653-9.

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4

Dietz, N. L. y D. D. Keiser. "TEM Analysis of Corrosion Products From a Radioactive Stainless Steel-based Alloy". Microscopy and Microanalysis 6, S2 (agosto de 2000): 368–69. http://dx.doi.org/10.1017/s1431927600034334.

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Argonne National Laboratory has developed an electrometallurgical treatment process for metallic spent nuclear fuel from the Experimental Breeder Reactor-II. This process stabilizes metallic sodium and separates usable uranium from fission products and transuranic elements that are contained in the fuel. The fission products and other waste constituents are placed into two waste forms: a ceramic waste form that contains the transuranic elements and active fission products such as Cs, Sr, I and the rare earth elements, and a metal alloy waste form composed primarily of stainless steel (SS), from claddings hulls and reactor hardware, and ∼15 wt.% Zr (from the U-Zr and U-Pu-Zr alloy fuels). The metal waste form (MWF) also contains noble metal fission products (Tc, Nb, Ru, Rh, Te, Ag, Pd, Mo) and minor amounts of actinides. Both waste forms are intended for eventual disposal in a geologic repository.

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5

Kilim, S., E. Strugalska-Gola, M. Szuta, S. Tyutyunnikov, O. Dalkhjav, V. I. Stegailov, I. A. Kryachko et al. "Am-241 incineration measurements with activation method in the QUINTA neutron field". EPJ Web of Conferences 204 (2019): 04004. http://dx.doi.org/10.1051/epjconf/201920404004.

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Am-241 sample was irradiated in spallation neutrons produced in ADS setup QUINTA at the JINR in Dubna. The energy was 660 MeV in the proton beam. The incineration study method was based on gamma-ray spectrometry. During the analysis of the spectra, several fission products were identified. Fission product activities yielded the number of fissions. Nevertheless, the lines are assumed to belong to the neutron capture product covered by parasitic Np-238 decay lines. The Np-238 lines as a result of neutron capture by Np-237 made impossible to determine the number of captures in Am-241.

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6

Hernandez Solis, Augusto, Alexey Stankovskiy, Luca Fiorito y Gert Van den Eynde. "Depletion uncertainty analysis to the MYRRHA fuel assembly model". EPJ Web of Conferences 239 (2020): 12001. http://dx.doi.org/10.1051/epjconf/202023912001.

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In this work, the objective is to perform an uncertainty analysis on a MYRRHA -Rev.1.6 irradiation cycle study, being applied to a depletion scenario of a single fresh fuel assembly while assuming reflective boundary conditions. Such analysis is statistically based on the application of Wilk’s method of building tolerance limits after 100 depletion calculations were performed with the SERPENT2 code. Due to the computational burden of such type of simulations, this propagation of nuclear data covariances study (allowed by the fast computational performance of SERPENT2) was done at constant power, constant flux and, in a final exercise, at constant power with the addition of fission yield uncertainties (all of these cases employed ENDF/B-VII.1 data). It was observed that while depleting at constant power, the statistical variation of key fission products such as 148Nd is almost not present because of the normalization factor applied to the flux. In contrast, the irradiation at constant flux reveals dependence on burnup. Finally, the added fission yield uncertainties make clear the fact that they directly impact the degree of final uncertainty computed for fission products exemplified by 148Nd and 135Xe important for burnup estimation and reactor operation, respectively.

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7

Taylor, Zack, Benjamin Collins y Ivan Maldonado. "MATRIX EXPONENTIAL METHODS FOR PARALLEL COMPUTING OF ISOTOPIC DEPLETION AND SPECIES TRANSPORT FOR MOLTEN SALT REACTOR ANALYSIS". EPJ Web of Conferences 247 (2021): 06047. http://dx.doi.org/10.1051/epjconf/202124706047.

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Matrix exponential methods have long been utilized for isotopic depletion in nuclear fuel calculations. In this paper we discuss the development of such methods in addition to species transport for liquid fueled molten salt reactors (MSRs). Conventional nuclear reactors work with fixed fuel assemblies in which fission products and fissile material do not transport throughout the core. Liquid fueled molten salt reactors work in a much different way, allowing for material to transport throughout the primary reactor loop. Because of this, fission product transport must be taken into account. The set of partial differential equations that apply are discretized into systems of first order ordinary differential equations (ODEs). The exact solution to the set of ODEs is herein being estimated using the matrix exponential method known as the Chebychev Rational Approximation Method (CRAM).

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8

Stempniewicz, M. M., L. Winters y S. A. Caspersson. "Analysis of dust and fission products in a pebble bed NGNP". Nuclear Engineering and Design 251 (octubre de 2012): 433–42. http://dx.doi.org/10.1016/j.nucengdes.2011.09.049.

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9

Thomas, L. E. y R. J. Guenther. "AEM analysis of condensed-phase xenon in UO2 spent fuel". Proceedings, annual meeting, Electron Microscopy Society of America 46 (1988): 512–13. http://dx.doi.org/10.1017/s0424820100104625.

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Release of the abundant fission gases xenon and krypton in UO2 reactor fuels is a limiting factor in normal performance of fuel rods and a concern in possible accidents involving transient overheating of the fuel. Consequently, a knowledge of the fission gas behavior in fuel is of great interest. Although fission gases in fuel are widely believed to exist as gas bubbles or atoms in solution in the UO2, we have obtained evidence by analytical electron microscopy that the xenon and krypton can also exist as a condensed phase, i.e. as a liquid or solid at high internal pressures in the UO2. This finding is likely to be important in modeling fission gas release.In a typical light-water power reactor (LWR), operating temperatures vary from about 650K at the edge of a fuel pellet to about 1400K at peak-power axial regions. Samples prepared from different radial locations in peak-power sections of low gas-release LWR fuels ATM-101 and ATM-103 were examined in a 200 KV AEM to determine how the gas and solid fission products varied with local fuel operating temperature.

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10

Chebboubi, A., S. Julien-Laferrière, J. Nicholson, G. Kessedjian, O. Serot, A. Blanc, D. Bernard et al. "Measurements of Fission Products Yields with the LOHENGRIN mass spectrometer at ILL". EPJ Web of Conferences 242 (2020): 01001. http://dx.doi.org/10.1051/epjconf/202024201001.

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The CEA in collaboration with ILL and LPSC has developed a measurement program on symmetric and heavy mass fission product distributions. The combination of measurements with ionisation chamber and Ge detectors is necessary to describe precisely the heavy fission product region in mass and charge. Recently, new measurements of fission yields and kinetic energy distributions, for different fissioning systems (233,235 U(nth, f),241 Am(2nth, f) and 239,241 Pu(nth, f), were performed with recoil spectrometer LOHENGRIN. The focus has been done on the self-normalization of the data to provide new absolute measurements, independently of any libraries along with the experimental covariance matrix. To reach precise measurements, a new experimental procedure was developed along with a new analysis method.

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11

Kilim, Stanisław, Elżbieta Strugalska-Gola, Marcin Szuta, Marcin Bielewicz, Sergej I. Tyutyunnikov, Walter I. Furman, Jindra Adam y Vladimir I. Stegailov. "Np-237 incineration study in various beams in ADS setup QUINTA". Nukleonika 63, n.º 1 (1 de marzo de 2018): 17–22. http://dx.doi.org/10.1515/nuka-2018-0003.

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Abstract Neptunium-237 samples were irradiated in a spallation neutron field produced in accelerator-driven system (ADS) setup QUINTA. Five experiments were carried out on the accelerators at the JINR in Dubna - one in carbon (C6+), three in deuteron, and one in a proton beam. The energy in carbon was 24 GeV, in deuteron 2, 4 and 8 GeV, respectively, and 660 MeV in the proton beam. The incineration study method was based on gamma-ray spectrometry. During the analysis of the spectra several fission products and one actinide were identified. Fission product activities yielded the number of fissions. The actinide (Np-238), a result of neutron capture by Np-237, yielded the number of captures. The main goal of this work was to find out if and how the incineration rate depended on parameters of the accelerator beam.

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12

Wang, Yizhen, Menglei Cui, Jiong Guo, Jinlin Niu, Yingjie Wu, Baokun Liu y Fu Li. "Lognormal-Based Sampling for Fission Product Yields Uncertainty Propagation in Pebble-Bed HTGR". Science and Technology of Nuclear Installations 2020 (25 de septiembre de 2020): 1–21. http://dx.doi.org/10.1155/2020/8014521.

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Uncertainty analyses of fission product yields are indispensable in evaluating reactor burnup and decay heat calculation credibility. Compared with neutron cross section, there are fewer uncertainty analyses conducted and it has been a controversial topic by lack of properly estimated covariance matrix as well as adequate sampling method. Specifically, the conventional normal-based sampling method in sampling large uncertainty independent fission yields could inevitably generate nonphysical negative samples. Cutting off these samples would introduce bias into uncertainty results. Here, we evaluate thermal neutron-induced U-235 independent fission yields covariance matrix by the Bayesian updating method, and then we use lognormal-based sampling method to overcome the negative fission yields samples issue. Fission yields uncertainty contribution to effective multiplication factor and several fission products’ atomic densities at equilibrium core of pebble-bed HTGR are quantified and investigated. The results show that the lognormal-based sampling method could prevent generating negative yields samples and maintain the skewness of fission yields distribution. Compared with the zero cut-off normal-based sampling method, the lognormal-based sampling method evaluates the uncertainty of effective multiplication factor and atomic densities are larger. This implies that zero cut-off effect in the normal-based sampling method would underestimate the fission yields uncertainty contribution. Therefore, adopting the lognormal-based sampling method is crucial for providing reliable uncertainty analysis results in fission product yields uncertainty analysis.

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13

Klunder, Gregory L., John E. Andrews, Patrick M. Grant, Brian D. Andresen y Richard E. Russo. "Analysis of Fission Products Using Capillary Electrophoresis with On-Line Radioactivity Detection". Analytical Chemistry 69, n.º 15 (agosto de 1997): 2988–93. http://dx.doi.org/10.1021/ac970042e.

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14

Wasim, M. "Interferences in instrumental neutron activation analysis by threshold reactions and uranium fission for miniature neutron source reactor". ract 101, n.º 9 (septiembre de 2013): 601–6. http://dx.doi.org/10.1524/ract.2013.2064.

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Summary Miniature neutron source reactors (MNSR) are known for their stable neutron flux characteristics and are mostly employed for neutron activation analysis (NAA). Interfering reactions are sometimes observed in instrumental neutron activation analysis (INAA). Failure to correct for these interferences produces significant systematic positive errors. This paper provides correction factors for the interferences caused by the threshold reactions and fission products of 235U. These factors were calculated by using the experimentally determined thermal, epithermal and fast neutron flux and epithermal neutron flux shape factor and the nuclear data from the literature using the Høgdahl convention. Correction factors were calculated for (n, p) and (n, α) reactions for the most commonly observed radionuclides in INAA. Similarly, correction factors for uranium fission were calculated for 9 elements (Ce, Ba, La, Mo, Nd, Pd, Ru, Sm and Zr). The correction factors were validated by analyzing different materials. A comparison of uranium fission factors with those published in the literature showed a good agreement except for 97Zr, 99Mo and 131Ba which is due to difference in the flux characteristics. In general, these factors can be used with confidence.

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15

Rochman, Dimitri Alexandre y Eric Bauge. "Fission yields and cross sections: correlated or not?" EPJ Nuclear Sciences & Technologies 7 (2021): 5. http://dx.doi.org/10.1051/epjn/2021005.

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Cross sections and fission yields can be correlated, depending on the selection of integral experimental data. To support this statement, this work presents the use of experimental isotopic compositions (both for actinides and fission products) from a sample irradiated in a reactor, to construct correlations between various cross sections and fission yields. This study is therefore complementing previous analysis demonstrating that different types of nuclear data can be correlated, based on experimental integral data.

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16

Khamdeev, M. I. y E. A. Erin. "Plasma parameters in atomic-emission spectral analysis of phosphate concentrates of the fission products". Industrial laboratory. Diagnostics of materials 85, n.º 2 (1 de marzo de 2019): 17–22. http://dx.doi.org/10.26896/1028-6861-2019-85-2-17-22.

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Physical parameters of electric arc plasma as well as their time dependences are calculated when analyzing phosphate precipitates of the fission products of irradiated nuclear fuel. Phosphate concentrates of the fission products are known for their complex chemical composition and high thermal and chemical stability. Hence, direct atomic emission spectral analysis of phosphate powders without transferring them into solutions is advisable. Different conditions of sample preparation and synthesis of the reference materials determine the different chemical forms of the elements to be determined. This, in turn, affects the kinetics of their evaporation in the electrode crate and excitation processes in the plasma. The known mechanisms of those processes cannot always be transferred to specific conditions of the given method of analysis thus entailing the necessity of studying the effect of the samples chemical composition on the results of determination, proper choice of spectroscopic carriers, detailed study of spectra excitation processes in spectral analysis, and analysis of the physical parameters of the electric arc plasma. We used the lines Zn I 307.206 nm and Zn I 307.589 nm to measure the effective temperature of the central hot sections of the arc in a range of4500 - 6500 K. NaCl, BaCl2 and NaCl + T1C1 were studied to reduce the effect of the sample elemental composition on excitation conditions of the spectra and their stabilization as a spectroscopic carrier. In control experiments we used carrier-free samples. The coincidence of the values of the plasma physical parameters within the measurement error not exceeding 20%, as well as the identity of the nature of the kinetic curves for samples of phosphate precipitates and synthetic reference materials prove their correctness. The result of the study substantiate correctness of the direct atomic-emission spectral procedure in analysis of phosphate concentrates of fission when using synthetic reference materials.

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17

Sidhu, R. S., R. J. Chen, Yu A. Litvinov y Y. H. Zhang. "Revisiting the Analysis of the Isochronous Mass Measurements of Uranium Fission Fragments at the ESR". EPJ Web of Conferences 227 (2020): 02012. http://dx.doi.org/10.1051/epjconf/202022702012.

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The re-analysis of experimental data on mass measurements of ura- nium fission products obtained at the ESR in 2002 is discussed. State-of-the-art data analysis procedures developed for such measurements are employed.

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18

Barber, D. H. "Implementation of A Gibbs Energy Minimizer In A Fission-Product Release Computer Program". AECL Nuclear Review 2, n.º 1 (1 de junio de 2013): 39–48. http://dx.doi.org/10.12943/anr.2013.00005.

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SOURCE 2.0 is the Canadian computer program for calculating fractional release of fission products from the UO2 fuel matrix. In nuclear accidents, fission-product release from fuel is one of the physical steps required before radiation dose from fission products can affect the public. Fission-product release calculations are a step in the analysis path to calculating dose consequences to the public from postulated nuclear accidents. SOURCE 2.0 contains a 1997 model of fission-product vaporization by B.J. Corse et al. based on lookup tables generated with the FACT computer program. That model was tractable on computers of that day. However, the understanding of fuel thermochemistry has advanced since that time. Additionally, computational resources have significantly improved since the time of the development of the Corse model and now allow incorporation of the more-rigorous thermodynamic treatment. Combining the newer Royal Military College of Canada (RMC) thermodynamic model of irradiated uranium dioxide fuel, a new model for fission-product vaporization from the fuel surface, a commercial user-callable thermodynamics subroutine library (ChemApp), an updated nuclide list, and updated nuclear physics data, a prototype computer program based on SOURCE IST 2.0P11 has been created that performs thermodynamic calculations internally. The resulting prototype code (with updated and revised data) provides estimates of 140La releases that are in better agreement with experiments than the original code version and data. The improvement can be quantified by a reduction in the mean difference between experimental and calculated release fractions from 0.70 to 0.07. 140La is taken to be representative of “low-volatile” fission products. To ensure that the existing acceptable performance for noble gases and volatile fission products is not adversely affected by the changes, comparisons were also made for a representative noble gas, 85Kr, and a representative volatile fission-product, 134Cs. These nuclides have the largest dataset in the SOURCE 2.0 validation test suite. This improvement provides increased confidence in the safety margin for equipment qualification in Loss-of-Coolant Accidents with Loss of Emergency Core Cooling.

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19

Ngwenya, N. y E. M. N. Chirwa. "Biological removal of cationic fission products from nuclear wastewater". Water Science and Technology 63, n.º 1 (1 de enero de 2011): 124–28. http://dx.doi.org/10.2166/wst.2011.021.

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Nuclear energy is becoming a preferred energy source amidst rising concerns over the impacts of fossil fuel based energy on global warming and climate change. However, the radioactive waste generated during nuclear power generation contains harmful long-lived fission products such as strontium (Sr). In this study, cationic strontium uptake from solution by microbial cultures obtained from mine wastewater is evaluated. A high strontium removal capacity (qmax) with maximum loading of 444 mg/g biomass was achieved by a mixed sulphate reducing bacteria (SRB) culture. Sr removal in SRB was facilitated by cell surface based electrostatic interactions with the formation of weak ionic bonds, as 68% of the adsorbed Sr2+ was easily desorbed from the biomass in an ion exchange reaction with MgCl2. To a lesser extent, precipitation reactions were also found to account for the removal of Sr from aqueous solution as about 3% of the sorbed Sr was precipitated due to the presence of chemical ligands while the remainder occurred as an immobile fraction. Further analysis of the Sr-loaded SRB biomass by scanning electron microscopy (SEM) coupled to energy dispersive X-ray (EDX) confirmed extracellular Sr2+ precipitation as a result of chemical interaction. In summary, the obtained results demonstrate the prospects of using biological technologies for the remediation of industrial wastewaters contaminated by fission products.

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20

Fulsom, Bryan. "Bragg curve spectroscopy for improved fission fragment identification". EPJ Web of Conferences 242 (2020): 01006. http://dx.doi.org/10.1051/epjconf/202024201006.

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We report on the development of Bragg curve spectroscopy techniques to improve fission fragment identification in the measurement of independent fission product yields. The NIFFTE collaboration’s fissionTPC detector provides ionization energy and particle tracking information from neutroninduced fission targets. A joint effort between PNNL, LLNL, LANL, and the Colorado School of Mines is investigating the ionization profiles deposited by U-235, U-238, and Pu-239 fission products in this detector, with the goal of including additional stopping power information beyond a standard 2E analysis. The aim is to improve the determination of fragment atomic and mass numbers with this information, via methods such as parametric fits and machine learning techniques.

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21

Huang, Jintao, Bun Tsuchiya, Kenji Konashi y Michio Yamawaki. "Thermodynamic analysis of chemical states of fission products in uranium–zirconium hydride fuel". Journal of Nuclear Materials 294, n.º 1-2 (abril de 2001): 154–59. http://dx.doi.org/10.1016/s0022-3115(01)00446-9.

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22

Arrigo, Leah M., Jun Jiang, Zachary S. Finch, James M. Bowen, Staci M. Herman, Larry R. Greenwood, Judah I. Friese y Brienne N. Seiner. "Separation of Lanthanide Isotopes from Mixed Fission Product Samples". Separations 8, n.º 7 (20 de julio de 2021): 104. http://dx.doi.org/10.3390/separations8070104.

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The measurement of radioactive fission products from nuclear events has important implications for nuclear data production, environmental monitoring, and nuclear forensics. In a previous paper, the authors reported the optimization of an intra-group lanthanide separation using LN extraction resin from Eichrom Technologies®, Inc. and a nitric acid gradient. In this work, the method was demonstrated for the separation and quantification of multiple short-lived fission product lanthanide isotopes from a fission product sample produced from the thermal irradiation of highly enriched uranium. The separations were performed in parallel in quadruplicate with reproducible results and high decontamination factors for 153Sm, 156Eu, and 161Tb. Based on the results obtained here, the fission yields for 144Ce, 153Sm, 156Eu, and 161Tb are consistent with published fission yields. This work demonstrates the effectiveness of the separations for the intended application of short-lived lanthanide fission product analysis requiring high decontamination factors.

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23

Vogt, R., J. Randrup, P. Talou, J. T. Van Dyke y L. A. Bernstein. "Parameter Optimization and Sensitivity Studies of Spontaneous Fission with FREYA". EPJ Web of Conferences 239 (2020): 05003. http://dx.doi.org/10.1051/epjconf/202023905003.

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For many years, the state of the art for simulating fission in transport codes amounted to sampling from average distributions. However, such "average" fission models have limited capabilities. Energy is not explicitly conserved and no correlations are available because all particles are emitted independently. However, in a true fission event, the emitted particles are correlated. Recently, Monte Carlo codes generating complete fission events have been developed, thus allowing the use of event-by-event analysis techniques. Such techniques are particularly useful because the complete kinematic information is available for the fission products and the emitted neutrons and photons. It is therefore possible to extract any desired observables, including correlations. The fast event-by-event fission code FREYA (Fission Reaction Event Yield Algorithm) generates large samples of complete fission events, employing only a few physics-based parameters. A recent optimization of these parameters for the isotopes in FREYA that undergo spontaneous fission is described and results are presented. The sensitivity of neutron observables in FREYA to the input yield functions is also discussed and the correlation between the average neutron multiplicity and fragment total kinetic energy is quantified.

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24

Mukhamadeev, Ruben, Leonid Parafilo, Yury Baranaev y Albert Suvorov. "Analysis of a severe beyond design basis accident for the EGP-6 reactor of the Bilibino NPP. Radioactive source term determination". Nuclear Energy and Technology 4, n.º 2 (26 de noviembre de 2018): 135–42. http://dx.doi.org/10.3897/nucet.4.30774.

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Analysis was performed of dynamic phase of severe accident of the EGP-6 reactor of the Bilibino NPP, due to uncontrolled reactivity insertion initiated by withdrawal of two pare of automatic control rods with followed by full failure of reactor emergency protection system. This initial event leads to promt increasing of reactor core power up to 450% of nominal value with short period, coupled with rise of temperature of fuel, pressure and temperature of coolant. These factors lead to crisis of heat exchange with subsequent ruptures tubes of fuel assemblies and coolant blow down into graphite stack. All its lead to rise of pressure in reactor shell and damage of it, outflow of steam-water mixture through up-reactor area to ventilation system, communication corridors and reactor hall and further – to atmospheric release. Transient processes were calculated using code RELAP5/Mod3.2. It was considered stages of processes of fuel damage and evaluated dynamic of a number and degree of damaged fuel assembles. They were grouped on burn-up and for each group it was performed analysis of dynamic of damage values. Further it was considered processes of yield of fission products from damaged fuel with models, based on experimental data on yield of fission products from fuel material, used in assembles of Bilibino NPP fuel type (fuel tubes with steel cladding, where fuel material is grits of uranium dioxide in magnesium), under condition of severe accident, especially performed in SSC IPPE. Transport of fission products with steam and air up to release points was evaluated with models, based on experimental data of fission product transport through graphite stack under conditions of severe accident, also especially performed in SSC IPPE. Evaluation of source term was performed in accordance with accident dynamic and assumed modes of release for conservative and most possible approaches. It was noted good self-protection property of EGP-6 reactor under severe beyond design basis accident condition.

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Kontautas, A., E. Babilas y E. Urbonavičius. "COCOSYS analysis for deposition of aerosols and fission products in PHEBUS FPT-2 containment". Nuclear Engineering and Design 247 (junio de 2012): 160–67. http://dx.doi.org/10.1016/j.nucengdes.2012.02.015.

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26

Hearne, Jason A. y Pavel V. Tsvetkov. "Analysis of the transmutation of long lived fission products using a charged particle beam". Annals of Nuclear Energy 133 (noviembre de 2019): 501–10. http://dx.doi.org/10.1016/j.anucene.2019.06.035.

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27

Morrison, Samuel S., Sue B. Clark, Tere A. Eggemeyer, Erin C. Finn, C. Corey Hines, Mathew D. King, Lori A. Metz et al. "Activation product analysis in a mixed sample containing both fission and neutron activation products". Journal of Radioanalytical and Nuclear Chemistry 314, n.º 3 (2 de noviembre de 2017): 2501–6. http://dx.doi.org/10.1007/s10967-017-5563-9.

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28

Rohanda, Anis. "ANALISIS PERUBAHAN MASSA BAHAN FISIL DAN NON FISIL DALAM TERAS PWR 1000 MWe DENGAN ORIGEN-ARP 5.1". JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 17, n.º 1 (15 de marzo de 2015): 13. http://dx.doi.org/10.17146/tdm.2015.17.1.2234.

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Teras reaktor merupakan tempat terjadinya reaksi pembelahan (fisi) yang terkendali. Komponen reaktor seperti bahan bakar, kelongsong (cladding) dan air pendingin memiliki peranan penting dalam keberlangsungan reaksi fisi. Reaksi fisi mengakibatkan terbentuknya sejumlah nuklida hasil fisi dan hasil aktivasi. Hasil fisi berasal dari reaksi tangkapan neutron termal dengan bahan fisil sedangkan hasil aktivasi berasal dari interaksi bahan non fisil seperti material kelongsong dan pendingin oleh neutron dan gamma. Pada setiap pengoperasian suatu reaktor, informasi perubahan massa bahan fisil dan non fisil sangat berguna untuk manajemen bahan bakar dalam teras, seperti pengaturan reaktivitas, optimasi dan pemuatan bahan bakar. Untuk itu perlu dilakukan penelitian mengenai perubahan bahan fisil dan non fisil tersebut dalam teras reaktor. Hal ini dapat dilakukan dengan mengamati perubahan massa dari material dalam teras reaktor. Penelitian ini memiliki tujuan untuk mengetahui perubahan massa unsur penyusun material dalam teras, seperti massa dari unsur penyusun elemen bahan bakar nuklir, kelongsong dan air pendingin setelah digunakan dalam teras. Dari perubahan massa tersebut dapat diketahui fraksi bakar atau tingkat konsumsi bahan bakar yang digunakan. Penelitian dilakukan pada basis reaktor PLTN tipe PWR buatan pabrikan asal Amerika Serikat berdaya 1000 MWe dengan menggunakan code penghitung inventori hasil fisi ORIGEN-ARP 5.1, yaitu versi terbaru dari ORIGEN dengan library khusus reaktor daya. Hasil analisis menunjukkan bahwa bahan fisil U-235 mengalami pengurangan massa hingga 58% atau lebih dari separuhnya dari massa U-235 awal untuk tiap kali siklus operasi. Bahan fertil U-238 hanya mengalami pengurangan massa sekitar 2% dari massa awalnya tiap kali siklus operasi. Lain halnya dengan bahan non fisil yang mengalami perubahan massa yang berbeda-beda untuk tiap kali siklus operasinya yang tergantung pada tampang lintang aktivasi serta laju peluruhan dan pembentukan nuklida induk. Kata kunci: bahan fisil, bahan non fisil, PWR, ORIGEN-ARP 5.1 Controlled fission reaction occurs in the reator core. Reactor components such as fuel, cladding and cooling water have an important role in the sustainability of the fission reaction. Fission reaction causes the formation of a number of fission product nuclides and activation products. Fission product nuclides are produced from thermal neutron capture reaction of fissile material while the activation products are originated from interaction of non-fissile materials such as cladding material and coolant by neutron and gamma. At each of reactor operation, the information of fuel material changes in the form of non-fissile or fissile material, is very usefull for the management of core fuel, such as for reactivity control, optimization and loading of fuel. Hence, it needs to perform a research in the fissile and non-fissile material changes in the reactor core. This can be done by observing the change of material mass in the reactor core. The objective of this research is to determine the change in mass of material in the core, such as the mass of the nuclear fuel elements, cladding and cooling water after use in the core. From mass changes can be delivered to burn up calculation or fuel consumption level. The calculation were performed on the basis of the United States PWR 1000 MWe by using a fission inventory computer code of ORIGEN-ARP 5.1, a new version of ORIGEN with specific library for nuclear power plant. The analysis results show that the U-235 fissile material having a mass reduction up to 58% or more than half from the initial U-235 mass for each operation cycle period. Fertile material U-238 was reduced by about 2% only from the initial mass for each operating cycle period. For other cases, the non-fissile material case, mass changes reduced in various for each operation cycle, depend on activation cross-sections and decay and formation rate of parent nuclides. Keyword: fissile material, non fissile material, PWR, ORIGEN-ARP 5.1

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29

Kilim, S., E. Strugalska-Gola, M. Szuta, M. Bielewicz, S. Tyutyunnikov, J. Adam y V. I. Stegailov. "Np-237 transmutation efficiency dependence on beam particle, energy and sample position in QUINTA setup". EPJ Web of Conferences 204 (2019): 04005. http://dx.doi.org/10.1051/epjconf/201920404005.

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Np-237 samples were irradiated with spallation neutrons produced at the ADS setup QUINTA. Six experiments were carried out at the JINR, in Dubna – one in carbon (C6+), three in deuteron, and two in proton beams. The energy in carbon was 24 GeV, in deuteron – 2, 4 and 8 GeV, respectively, and 660 MeV in the proton beam. In five cases the sample was located in a side window in a lead shield. In one case (660 MeV proton beam) two samples were located on the top of the QUINTA setup, one – on the top of section 2, and the second one – on the top of section 4. The transmutation study method was based on gamma-ray spectrometry. During the analysis of the spectra several fission products and one actinide were identified. Fission product activities yielded the number of fissions. The actinide (Np-238), a result of neutron capture by Np-237, yielded the number of captures. The main goal of this work was to find out if and how the transmutation rate depended on the accelerator beam and sample location during the irradiation.

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30

Kerkápoly, Anikó, Nóra Vajda, Tamás Pintér y Pintér Csordás. "Hot particles analysis originating from failed and damaged fuels". Open Chemistry 3, n.º 1 (1 de marzo de 2005): 106–17. http://dx.doi.org/10.2478/bf02476242.

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AbstractThe increase of activities of fission products and transmutation products in the primary coolant of a nuclear power plant indicates the presence of fuel rod failures. The measurement of the activity concentration of the primary coolant was able to detect fuel failures in the reactor core. Microanalytical methods for examining individual hot particles have been developed and applied to fuel failure detection under normal operation conditions as well as during the severe fuel damage that occurred in the cleaning tank incident at Unit 2 of NPP Paks in April 2003. Several faulty fuel rods can be detected simultaneously by the characterization of individual hot particles originating from the primary water. The analysis of particles originating from the damaged fuels provides information relating to the dissolution process of the fuel debris.

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31

Apostol, M., M. Constantin y A. Leca. "Uncertainty analysis for fission products transport in CANDU primary heat transport during a severe accident". Kerntechnik 75, n.º 4 (agosto de 2010): 170–77. http://dx.doi.org/10.3139/124.110075.

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32

Korotev, Randy L. "Error in neutron activation analysis from recoil-implanted fission products from uranium in aluminum foil." GEOCHEMICAL JOURNAL 22, n.º 3 (1988): 133–37. http://dx.doi.org/10.2343/geochemj.22.133.

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33

Bin, Li. "Analysis of fission products— a method for verification of a CTBT during on‐site inspections". Science & Global Security 7, n.º 2 (enero de 1998): 195–207. http://dx.doi.org/10.1080/08929889808426454.

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34

Stankovskiy, A. y G. Van den Eynde. "Advanced Method for Calculations of Core Burn-Up, Activation of Structural Materials, and Spallation Products Accumulation in Accelerator-Driven Systems". Science and Technology of Nuclear Installations 2012 (2012): 1–12. http://dx.doi.org/10.1155/2012/545103.

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The ALEPH2 Monte Carlo depletion code has two principal features that make it a flexible and powerful tool for reactor analysis. First of all, it uses a nuclear data library covering neutron- and proton-induced reactions, neutron and proton fission product yields, spontaneous fission product yields, radioactive decay data, and total recoverable energies per fission. Secondly, it uses a state-of-the-art numerical solver for the first-order ordinary differential equations describing the isotope balances, namely, a Radau IIA implicit Runge-Kutta method. The versatility of the code allows using it for time behavior simulation of various systems ranging from single pin model to full-scale reactor model, including such specific facilities as accelerator-driven systems. The core burn-up, activation of the structural materials, irradiation of samples, and, in addition, accumulation of spallation products in accelerator-driven systems can be calculated in a single ALEPH2 run. The code is extensively used for the neutronics design of the MYRRHA research facility which will operate in both critical and subcritical modes.

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35

Voirin, Brieuc, Grégoire Kessedjian, Abdelaziz Chebboubi, Sylvain Julien-Laferrière y Olivier Serot. "From fission yield measurements to evaluation: status on statistical methodology for the covariance question". EPJ Nuclear Sciences & Technologies 4 (2018): 26. http://dx.doi.org/10.1051/epjn/2018030.

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Studies on fission yields have a major impact on the characterization and the understanding of the fission process and are mandatory for reactor applications. Fission yield evaluation represents the synthesis of experimental and theoretical knowledge to perform the best estimation of mass, isotopic and isomeric yields. Today, the output of fission yield evaluation is available as a function of isotopic yields. Without the explicitness of evaluation covariance data, mass yield uncertainties are greater than those of isotopic yields. This is in contradiction with experimental knowledge where the abundance of mass yield measurements is dominant. These last years, different covariance matrices have been suggested but the experimental part of those are neglected. The collaboration between the LPSC Grenoble and the CEA Cadarache starts a new program in the field of the evaluation of fission products in addition to the current experimental program at Institut Laue-Langevin. The goal is to define a new methodology of evaluation based on statistical tests to define the different experimental sets in agreement, giving different solutions for different analysis choices. This study deals with the thermal neutron induced fission of 235U. The mix of data is non-unique and this topic will be discussed using the Shannon entropy criterion in the framework of the statistical methodology proposed.

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36

Jiao, Zengtong, Xiaotong Chen, Chao Fang, Gang Xu, Chi Zhang, Luhao Fan y Bing Liu. "DFT Study of Cs/Sr/Ag Adsorption on Defective Matrix Graphite". Science and Technology of Nuclear Installations 2020 (28 de agosto de 2020): 1–11. http://dx.doi.org/10.1155/2020/4921623.

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The geometries, adsorption energies, and electronic structures of Cs, Sr, and Ag atoms on matrix graphite surface with point defects were calculated and analyzed using the density functional theory (DFT) and the Perdew–Burke–Ernzerhof (PBE) formulation of the generalized gradient approximation (GGA). Three different types of point defects, i.e., single vacancy and “bridge” and “spiro” interstitials are considered using approximate van der Waals (vdW) correction methods. The results of adsorption energies show that the metal fission products of Cs, Sr, and Ag are more stable on single vacancy defects than “bridge” or “spiro” interstitial defects. This is further confirmed by the analysis of electronic structures, such as charge density difference (CDD) and density of state (DOS). All these results indicate that dangling bonds play an important role in the adsorption behaviors of metallic fission products on matrix graphite.

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37

Jaroszewicz, Janusz, Zuzanna Marcinkowska y Krzysztof Pytel. "Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis". Nukleonika 59, n.º 2 (8 de julio de 2014): 43–52. http://dx.doi.org/10.2478/nuka-2014-0009.

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Abstract The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.

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38

Leng, B., I. J. van Rooyen, Y. Q. Wu, I. Szlufarska y K. Sridharan. "STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment". Journal of Nuclear Materials 475 (julio de 2016): 62–70. http://dx.doi.org/10.1016/j.jnucmat.2016.03.008.

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39

Mattera, A., D. Gorelov, M. Lantz, B. Lourdel, H. Penttilä, S. Pomp y I. Ryzhov. "A ROOT-based analysis tool for measurements of neutron-induced fission products at the IGISOL facility". Physica Scripta T150 (28 de septiembre de 2012): 014025. http://dx.doi.org/10.1088/0031-8949/2012/t150/014025.

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40

Guo, Zan, Shuliang Zou, Wenge Ma y Haiyin Dai. "HAZOP Analysis and Research of Temporary Acid Adding System for High-Discharge Waste Liquid". Science and Technology of Nuclear Installations 2021 (23 de febrero de 2021): 1–6. http://dx.doi.org/10.1155/2021/6633916.

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To prevent the formation of salt from fission products of high-level radioactive liquid wastes (HLWs), a certain amount of acid is added to maintain the acidity of liquid waste. This study analyzes the accidents associated with the addition of acids in a factory by using the hazard and operability analysis (HAZOP), while elucidating the corresponding defects and risks of this approach. By improving the design of the system, the possibility of an accident is significantly reduced. This study can provide guidance for adding acids to treat other high-level waste liquids.

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41

Viaud, C., G. Carlot, P. Garcia, P. Martin, N. Millard-Pinard, N. Moncoffre, C. Peaucelle, Thierry Sauvage y N. Toulhoat. "Thermal Behaviour of Xenon in a Refractory Metal for Gas Fast Reactor Fuel Elements". Defect and Diffusion Forum 272 (marzo de 2008): 25–30. http://dx.doi.org/10.4028/www.scientific.net/ddf.272.25.

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Helium cooled Gas Fast Reactors (GFR) are designed for producing energy more efficiently and improving safety features such as a total retention of fission products (Xe, I, Cs). This study deals with the diffusion of xenon in refractory liners dedicated to the retention of fission products produced in GFR fuels. The material (W, Mo, W-Re, Mo-Re) will be located in the heart of the nuclear fuel element, where the operating temperature is in the 1000°C- 1600°C range. For the investigation of thermally activated rare gas behaviour, a γ-spectrometry analysis experiment has been performed on the 133Xenon implanted refractory liner. Preliminary results on the 133Xenon release at 1600°C from a tungsten single crystal is presented. In spite of the low concentration of implanted gas (~ppm) and simple microstructure, the prevailing mechanism appears to be complex. One and two dimensional diffusion models are used to characterize or discriminate the highlighted phenomena: burst release, diffusion and trapping of rare gas atoms.

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42

Bachhav, Mukesh, Brandon Miller, Jian Gan, Dennis Keiser, Ann Leenaers, S. Van den Berghe y Mitchell K. Meyer. "Microstructural Changes and Chemical Analysis of Fission Products in Irradiated Uranium-7 wt.% Molybdenum Metallic Fuel Using Atom Probe Tomography". Applied Sciences 11, n.º 15 (27 de julio de 2021): 6905. http://dx.doi.org/10.3390/app11156905.

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Understanding the microstructural and phase changes occurring during irradiation and their impact on metallic fuel behavior is integral to research and development of nuclear fuel programs. This paper reports systematic analysis of as-fabricated and irradiated low-enriched U-Mo (uranium-molybdenum metal alloy) fuel using atom probe tomography (APT). This study is carried out on U-7 wt.% Mo fuel particles coated with a ZrN layer contained within an Al matrix during irradiation. The dispersion fuel plates from which the fuel samples were extracted are irradiated at Belgian Nuclear Research Centre (SCK CEN) with burn-up of 52% and 66% in the framework of the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) project. The APT studies on U-Mo particles from as-fabricated fuel plates enriched to 19.8% revealed predominantly γ-phase U-Mo, along with a network of the cell boundary decorated with α-U, γ’-U2Mo, and UC precipitates along the grain boundaries. The corresponding APT characterization of irradiated fuel samples showed formation of fission gas bubbles enriched with solid fission products. The intermediate burnup sample showed a uniform distribution of the typical bubble superlattice with a radius of 2 nm arranged in a regular lattice, while the high burnup sample showed a non-uniform distribution of bubbles in grain-refined regions. There was no evidence of remnant α-U, γ’-U2Mo, and UC phases in the irradiated U-7 wt.% Mo samples.

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43

MAMTIMIN, MAYIR, VALERIIA N. STAROVOITOVA y FRANK HARMON. "LINAC-BASED PHOTONUCLEAR APPLICATIONS AT THE IDAHO ACCELERATOR CENTER". International Journal of Modern Physics: Conference Series 27 (enero de 2014): 1460146. http://dx.doi.org/10.1142/s201019451460146x.

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In this paper, current Idaho Accelerator Center (IAC) activities based on the exploitation of high energy bremsstrahlung photons generated by linear electron accelerators will be reviewed. These beams are used to induce photonuclear interactions for a wide variety of applications in materials science, activation analysis, medical research, and nuclear technology. Most of the exploited phenomena are governed by the familiar giant dipole resonance cross section in nuclei. By proper target and converter design, optimization of photon and photoneutron production can be achieved, allowing radiation fields produced with both photons and neutrons to be used for medical isotope production and for fission product transmutation. The latter provides a specific application example that supports long-term fission product waste management. Using high-energy, highpower electron accelerators, we can demonstrate transmutation of radio-toxic, long-lived fission products (LLFP) such as 99Tc and 129I into short lived species. The latest experimental and simulation results will be presented.

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44

Chiang, Ren-Tai. "ANALYSIS OF CS-137 TO CS-134 ACTIVITY RATIO FOR FAILED FUEL EXPOSURE ESTIMATION". Indonesian Journal of Physics and Nuclear Applications 3, n.º 3 (23 de diciembre de 2018): 76–82. http://dx.doi.org/10.24246/ijpna.v3i3.76-82.

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The Cs-134 to Cs-137 activity ratio of the Cs-134 and Cs-137 fission products released from failed fuel rods into primary coolant is very useful to identify the exposure along with the fuel batch of the failed fuel. The calculated and measured Cs-137 to Cs-134 radioactivity ratios of failed BWR and PWR fuels are compared and analyzed for better understanding of their relationship. Moreover, the impacts of power uprate and fuel reload outage on calculated Cs-137 to Cs-134 activity ratios are studied and the physics behind the impacts are provided.

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45

Al-Mugrabi, M. y N. M. Spyrou. "The determination of uranium using short-lived fission products by cyclic and other modes of activation analysis". Journal of Radioanalytical and Nuclear Chemistry Articles 112, n.º 2 (mayo de 1987): 277–83. http://dx.doi.org/10.1007/bf02132360.

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46

Yang-Hyun, Koo, Sohn Dong-Seong y Yoon Young-Ku. "An analysis method for the fuel rod gap inventory of unstable fission products during steady-state operation". Journal of Nuclear Materials 209, n.º 1 (marzo de 1994): 62–78. http://dx.doi.org/10.1016/0022-3115(94)90248-8.

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47

Moiseenko, V. y S. Chernitskiy. "Nuclear Fuel Cycle with Minimized Waste". Nuclear and Radiation Safety, n.º 1(81) (12 de marzo de 2019): 30–35. http://dx.doi.org/10.32918/nrs.2019.1(81).05.

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A uranium-based nuclear fuel and fuel cycle are proposed for energy production. The fuel composition is chosen so that during reactor operation the amount of each transuranic component remains unchanged since the production rate and nuclear reaction rate are balanced. In such a ‘balanced’ fuel only uranium-238 content has a tendency to decrease and, to be kept constant, must be sustained by continuous supply. The major fissionable component of the fuel is plutonium is chosen. This makes it possible to abandon the use of uranium-235, whose reserves are quickly exhausted. The spent nuclear fuel of such a reactor should be reprocessed and used again after separation of fission products and adding depleted uranium. This feature simplifies maintaining the closed nuclear fuel cycle and provides its periodicity. In the fuel balance calculations, nine isotopes of uranium, neptunium, plutonium and americium are used. This number of elements is not complete, but is quite sufficient for calculations which are used for conceptual analysis. For more detailed consideration, this set may be substantially expanded. The variation of the fuel composition depending on the reactor size is not too big. The model accounts for fission, neutron capture and decays. Using MCNPX numerical Monte-Carlo code, the neutron calculations are performed for the reactor of industrial nuclear power plant size with MOX fuel and for a small reactor with metallic fuel. The calculation results indicate that it is possible to achieve criticality of the reactor in both cases and that production and consuming rates are balanced for the transuranic fuel components. In this way, it can be assumed that transuranic elements will constantly return to such a reactor, and only fission products will be sent to storage. This will significantly reduce the radioactivity of spent nuclear fuel. It is important that the storage time for the fission products is much less than for the spent nuclear fuel, just about 300 years.

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48

Vyshemirskyi, M., V. Pustovit, V. Kravchenko y D. Donskyi. "Analysis of Processes in the Containment Using ATHLET-CD and COCOSYS Codes". Nuclear and Radiation Safety, n.º 2(86) (12 de junio de 2020): 27–37. http://dx.doi.org/10.32918/nrs.2020.2(86).04.

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A brief description of performed input deck modifications and results of stand-alone and coupled calculations of Dn 200 mm loss of coolant accident with simultaneous total station blackout accident scenario for Rivne Nuclear Power Plant Unit 1 (WWER‑440/V-213) with application of ATHLET-CD 3.1A and COCOSYS 2.4 codes are presented in the paper. ATHLET-CD stand-alone calculation was performed with constant containment pressure (a time dependent volume with constant pressure and temperature was used as a boundary volume for leakage). Further, mass and energy release and fission products from the primary system obtained during ATHLET‑CD stand-alone calculation were used to perform COCOSYS stand-alone calculation. In addition, coupled ATHLET-CD and COCOSYS calculation was performed. All the computer analyzes were performed until the lower head failure. ATHLET‑CD model was extended with core degradation module (ECORE), which allowed calculation of scenario until reactor pressure vessel failure. According to the results of comparative analysis, nearly the same behavior of the main parameters in the stand-alone and coupled calculation at an early phase of scenario was obtained. Some small differences occur due to distinction in behavior of water and steam mass flows released through the break and due to existence of heat transfer from the primary system structures to the containment compartments during coupled calculation of transient. As for middle and late phases of the accident, some differences between stand-alone and coupled calculation results for analyzed scenario are present. These differences are caused by different total fission products and aerosols release from the reactor coolant system to the containment compartments. The above information allows recommending application of coupled code/model versions for performing the computer severe accident analyses.

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49

Flores y Flores, Alain, Danilo Ferretto, Tereza Marková y Guido Mazzini. "Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2". Sustainability 13, n.º 14 (16 de julio de 2021): 7964. http://dx.doi.org/10.3390/su13147964.

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The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

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50

Dzianisau, Siarhei, Jinsu Park, Sooyoung Choi, Alexey Cherezov, Xianan Du y Deokjung Lee. "DEVELOPMENT OF DECAY HEAT MODEL FOR RAST-K". EPJ Web of Conferences 247 (2021): 07009. http://dx.doi.org/10.1051/epjconf/202124707009.

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Decay heat (DH) is the heat produced through a radioactive decay of fission products during or after a reactor operation. It is known as the second largest source of power in the core after fission. Being such a strong contributor to reactor power, it should be accurately determined at any time of reactor operation. Currently, there are two main approaches for DH estimation used in reactor simulation codes. One approach is based on careful inventorying of all produced target nuclides and their individual contributions to total power. Alternatively, the other popular approach is based on collapsing all target fission products into a small number of groups similar to delayed neutron estimation techniques. However, the last (multigroup) method currently has limitations when used in some transient scenarios such as transients occurred in fresh fuel. In this study, the multigroup method was further developed for reducing limitations while retaining the advantage in computation speed. Then, it was implemented into Reactor Analysis code for Steady state and Transient (RAST-K) and tested against other codes. As a result, the improved method was found capable of determining DH power at all tested stages of reactor operation under any tested operation scenario. In particular, the test simulations using the improved method showed better results in those scenarios that were under accuracy limitations of the original multigroup method. Overall, the quality of transient calculations in RAST-K was improved when using the newly implemented DH module.

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