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Статті в журналах з теми "Fission products Analysis"

Автор: Grafiati

Опубліковано: 4 червня 2021

Оновлено: 12 лютого 2022

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1

Petrovski, A. M., T. N. Korbut, E. A. Rudak, and 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, no. 4 (January 11, 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

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2

Zhou, Tao, Peng Xu, Tian Qi, Xuemeng Qin, Juan Chen, and 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 (June 3, 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

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3

Auxier, John D., Jacob A. Jordan, S. Adam Stratz, Shayan Shahbazi, Daniel E. Hanson, Derek Cressy, and Howard L. Hall. "Thermodynamic analysis of volatile organometallic fission products." Journal of Radioanalytical and Nuclear Chemistry 307, no. 3 (December 17, 2015): 1621–27. http://dx.doi.org/10.1007/s10967-015-4653-9.

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4

Dietz, N. L., and D. D. Keiser. "TEM Analysis of Corrosion Products From a Radioactive Stainless Steel-based Alloy." Microscopy and Microanalysis 6, S2 (August 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), fro

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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, and 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 powe

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7

Taylor, Zack, Benjamin Collins, and 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

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8

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

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9

Thomas, L. E., and 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

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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 ab

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11

Kilim, Stanisław, Elżbieta Strugalska-Gola, Marcin Szuta, Marcin Bielewicz, Sergej I. Tyutyunnikov, Walter I. Furman, Jindra Adam, and Vladimir I. Stegailov. "Np-237 incineration study in various beams in ADS setup QUINTA." Nukleonika 63, no. 1 (March 1, 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 fissi

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12

Wang, Yizhen, Menglei Cui, Jiong Guo, Jinlin Niu, Yingjie Wu, Baokun Liu, and Fu Li. "Lognormal-Based Sampling for Fission Product Yields Uncertainty Propagation in Pebble-Bed HTGR." Science and Technology of Nuclear Installations 2020 (September 25, 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 eval

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13

Klunder, Gregory L., John E. Andrews, Patrick M. Grant, Brian D. Andresen, and Richard E. Russo. "Analysis of Fission Products Using Capillary Electrophoresis with On-Line Radioactivity Detection." Analytical Chemistry 69, no. 15 (August 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, no. 9 (September 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 epi

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15

Rochman, Dimitri Alexandre, and 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., and E. A. Erin. "Plasma parameters in atomic-emission spectral analysis of phosphate concentrates of the fission products." Industrial laboratory. Diagnostics of materials 85, no. 2 (March 1, 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.

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17

Sidhu, R. S., R. J. Chen, Yu A. Litvinov, and 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, no. 1 (June 1, 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 w

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19

Ngwenya, N., and E. M. N. Chirwa. "Biological removal of cationic fission products from nuclear wastewater." Water Science and Technology 63, no. 1 (January 1, 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 facilita

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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. Th

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21

Huang, Jintao, Bun Tsuchiya, Kenji Konashi, and Michio Yamawaki. "Thermodynamic analysis of chemical states of fission products in uranium–zirconium hydride fuel." Journal of Nuclear Materials 294, no. 1-2 (April 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, and Brienne N. Seiner. "Separation of Lanthanide Isotopes from Mixed Fission Product Samples." Separations 8, no. 7 (July 20, 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 s

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23

Vogt, R., J. Randrup, P. Talou, J. T. Van Dyke, and 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

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24

Mukhamadeev, Ruben, Leonid Parafilo, Yury Baranaev, and 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, no. 2 (November 26, 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 s

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25

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

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26

Hearne, Jason A., and Pavel V. Tsvetkov. "Analysis of the transmutation of long lived fission products using a charged particle beam." Annals of Nuclear Energy 133 (November 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, no. 3 (November 2, 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, no. 1 (March 15, 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

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29

Kilim, S., E. Strugalska-Gola, M. Szuta, M. Bielewicz, S. Tyutyunnikov, J. Adam, and 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 base

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30

Kerkápoly, Anikó, Nóra Vajda, Tamás Pintér, and Pintér Csordás. "Hot particles analysis originating from failed and damaged fuels." Open Chemistry 3, no. 1 (March 1, 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

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31

Apostol, M., M. Constantin, and A. Leca. "Uncertainty analysis for fission products transport in CANDU primary heat transport during a severe accident." Kerntechnik 75, no. 4 (August 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, no. 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, no. 2 (January 1998): 195–207. http://dx.doi.org/10.1080/08929889808426454.

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34

Stankovskiy, A., and 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

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35

Voirin, Brieuc, Grégoire Kessedjian, Abdelaziz Chebboubi, Sylvain Julien-Laferrière, and 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

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36

Jiao, Zengtong, Xiaotong Chen, Chao Fang, Gang Xu, Chi Zhang, Luhao Fan, and Bing Liu. "DFT Study of Cs/Sr/Ag Adsorption on Defective Matrix Graphite." Science and Technology of Nuclear Installations 2020 (August 28, 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 d

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Jaroszewicz, Janusz, Zuzanna Marcinkowska, and Krzysztof Pytel. "Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis." Nukleonika 59, no. 2 (July 8, 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

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38

Leng, B., I. J. van Rooyen, Y. Q. Wu, I. Szlufarska, and K. Sridharan. "STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment." Journal of Nuclear Materials 475 (July 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, and I. Ryzhov. "A ROOT-based analysis tool for measurements of neutron-induced fission products at the IGISOL facility." Physica Scripta T150 (September 28, 2012): 014025. http://dx.doi.org/10.1088/0031-8949/2012/t150/014025.

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Guo, Zan, Shuliang Zou, Wenge Ma, and Haiyin Dai. "HAZOP Analysis and Research of Temporary Acid Adding System for High-Discharge Waste Liquid." Science and Technology of Nuclear Installations 2021 (February 23, 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, and N. Toulhoat. "Thermal Behaviour of Xenon in a Refractory Metal for Gas Fast Reactor Fuel Elements." Defect and Diffusion Forum 272 (March 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 13

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42

Bachhav, Mukesh, Brandon Miller, Jian Gan, Dennis Keiser, Ann Leenaers, S. Van den Berghe, and 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, no. 15 (July 27, 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 Centr

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43

MAMTIMIN, MAYIR, VALERIIA N. STAROVOITOVA, and FRANK HARMON. "LINAC-BASED PHOTONUCLEAR APPLICATIONS AT THE IDAHO ACCELERATOR CENTER." International Journal of Modern Physics: Conference Series 27 (January 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

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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, no. 3 (December 23, 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., and 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, no. 2 (May 1987): 277–83. http://dx.doi.org/10.1007/bf02132360.

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46

Yang-Hyun, Koo, Sohn Dong-Seong, and 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, no. 1 (March 1994): 62–78. http://dx.doi.org/10.1016/0022-3115(94)90248-8.

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47

Moiseenko, V., and S. Chernitskiy. "Nuclear Fuel Cycle with Minimized Waste." Nuclear and Radiation Safety, no. 1(81) (March 12, 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

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48

Vyshemirskyi, M., V. Pustovit, V. Kravchenko, and D. Donskyi. "Analysis of Processes in the Containment Using ATHLET-CD and COCOSYS Codes." Nuclear and Radiation Safety, no. 2(86) (June 12, 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 produc

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Flores y Flores, Alain, Danilo Ferretto, Tereza Marková, and 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, no. 14 (July 16, 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 Europ

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Dzianisau, Siarhei, Jinsu Park, Sooyoung Choi, Alexey Cherezov, Xianan Du, and 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 collap

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