Relevant bibliographies by topics / Fission products Analysis

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  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers
  6. Reports

Academic literature on the topic 'Fission products Analysis'

Author: Grafiati
Published: 4 June 2021
Last updated: 12 February 2022

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Journal articles on the topic "Fission products Analysis"

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 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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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 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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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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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), 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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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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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 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, 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 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, 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 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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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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Dissertations / Theses on the topic "Fission products Analysis"

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EL-HAJJE, REFAAT Safety Science Faculty of Science UNSW. "A SIMULTANEOUS MEASUREMENT OF THE ANGULAR DISTRIBUTION, MASS AND KINETIC ENERGY OF 235U AND 232Th FISSION FRAGMENTS." Awarded by:University of New South Wales. School of Safety Science, 2000. http://handle.unsw.edu.au/1959.4/17612.

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Simultaneous measurements of the angular distribution, mass distribution and average total kinetic energy of fission fragments produced by the neutron-induced fission of 235U and 232Th have been made using a gridded ionisation chamber. The neutron energy range used was thermal to 1.9 MeV for 235U and 1.4 to 1.7 MeV for 232Th. The following topics were investigated: the interdependence of the fission fragment angular and mass distribution; the anomalous behaviour of fragment anisotropy for 235U(n,f) at neutron energies En below 150 keV; the possible existence of a third symmetric mass peak for 232Th(n,f); the mass fine structure in 235U(n,f) and 232Th(n,f); and the dependence of the fission fragment average total kinetic energy on the excitation energy of the fissioning nucleus. For this study, mono-energetic neutrons were produced by the and reactions. Four signals produced by the fission chamber were fed into a data acquisition system and processed by a specially modified comprehensive computer program. The results indicate that there is no interdependence between the angular and mass distributions of fragments for 235U(n,f) and for 232Th(n,f). The angular distribution of 235U fission fragments showed an anisotropy of less than one for En below 150 keV. For 232Th, the expected minimum in the anisotropy near En = 1.6 MeV was confirmed. No evidence for a third peak in the mass symmetry region of 232Th(n,f) was observed, within the yield sensitivity limitation of the chamber. Fine structure was observed in the mass yield distributions for 235U(n,f) and 232Th(n,f) at mass locations predicted by theory. The fission fragment average total kinetic energy for 235U(n,f) and 232Th(n,f) showed no significant dependence on the excitation energy of the fissioning nucleus. Possible reasons for some of these results are advanced.
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2

SATO, IVONE M. "Determinacao dos produtos de fissao em rejeito liquido de atividade alta por fluorescencia de raio-x .Correcao da interferencia espectral pela razao das intensidades." reponame:Repositório Institucional do IPEN, 1988. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9888.

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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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3

Belhabib, Tayeb. "Comportement thermique des défauts lacunaires induits par l’hélium et les gaz de fission dans le dioxyde d’uranium." Thesis, Orléans, 2012. http://www.theses.fr/2012ORLE2071/document.

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Dans les nouvelles centrales nucléaires dites 4ème génération, comme d’ailleurs les anciennes, le dioxyde d’uranium devra opérer dans des milieux hostiles de températures et d’irradiation avec la présence des produits de fission (PF) et des particules alpha (α). Le fonctionnement dans ces conditions extrêmes induira des déplacements d’atomes et dégradera les propriétés thermiques et mécaniques du combustible UO2. La compréhension du comportement des défauts lacunaires, des PF et de l’hélium est cruciale pour prévoir le comportement du dioxyde d’uranium au sein de ces futures installations nucléaires. La première partie de cette thèse est consacrée à l’étude des défauts lacunaires induits par l’implantation de krypton et d’iode (quelques MeV) dans l’UO2 polycristallin et leurs stades de recuits. L’analyse par spectroscopie d’annihilation de positons (PAS) a permis de mettre en évidence la création de défauts de Schottky VU-2VO dans le cas des implantations iode et la formation de clusters lacunaires contenant du gaz pour les implantations krypton. L’évolution en température de ces défauts générés dépend des paramètres d’implantation (nature des ions, énergie, fluence). Cette étude a montré les rôles importants que peuvent jouer les défauts lacunaires et la présence des gaz de fission dans l’évolution du matériau UO2. Ensuite, nous nous sommes intéressés à l’étude et à la caractérisation, par PAS et les techniques d’analyse par faisceau d’ions (NRA/C et RBS/C), du comportement de l’hélium dans l’UO2. Les mesures de NRA/C et RBS/C révèlent une localisation d’une grande fraction d’hélium dans les sites interstitiels octaédriques de la matrice UO2. La localisation de l’hélium reste stable dans ces sites pour T< 600°C, évoluent légèrement entre 600 et 700°C et devient aléatoire à 800°C. Les mesures PAS mettent en évidence trois stades d’évolution des défauts lacunaires : la recombinaison par migration des interstitiels d’oxygène, l’agglomération des défauts entre 600 et 800°C et leur dissociation et élimination lorsque la température augmente. Ces résultats suggèrent que le transport d'hélium est assisté par les défauts lacunaires
In the new fourth generation nuclear plants, as in the old ones, uranium dioxide must operate in hostile environments of temperature and irradiation with the presence of fission products (FP) and alpha particles (α). Operation in these extreme conditions will induce atoms displacements and degrade the thermal and mechanical properties of UO2 fuel. Understanding the behavior of induced vacancy defects, FP and helium is crucial to predict the uranium dioxide behavior in the future nuclear reactors. The first part of this thesis is dedicated to the study of vacancy defects induced by krypton and iodine implantation (a few MeV) in the UO2 polycrystalline and of their evolution under annealing. Analysis by positron annihilation spectroscopy (PAS) has highlighted the creation of Schottky defects VU-2VO in the case of iodine implantations and formation of vacancy clusters containing the gas for krypton implantation. The temperature evolution of these defects depends on the implantation parameters (nature of the ion energy, fluence). This study showed the important roles that can play vacancy defects and the presence of fission gases in the evolution of UO2 material. Then we were interested in the study of the helium behavior in UO2 its location and migration, agglomeration and interaction with vacancy defects by using PAS and ion beam analysis (NRA/C and RBS/C). The NRA/C and RBS/C characterizations showed a localization of a large helium fraction in the octahedral interstitial sites of the UO2 matrix. The helium location in these sites remains stable for T <600°C, changing slightly between 600 and 700°C and becomes random at 800°C. Positron annihilation spectroscopy reveals three stages of vacancy defects evolution : The recombination with oxygen interstitial migration, defects agglomeration between 600 and 800°C and their dissociation and elimination when the temperature increases. These results suggest that the He transport is assisted by the vacancy defects
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4

Junior, Iberê Souza Ribeiro. "Determinação de fatores de interferência de produtos de fissão do urânio na Análise por Ativação Neutrônica." Universidade de São Paulo, 2014. http://www.teses.usp.br/teses/disponiveis/85/85131/tde-22092014-144404/.

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A análise por ativação com nêutrons é um método utilizado na determinação de diversos elementos em diferentes tipos de matrizes. Entretanto, quando a amostra contém altos teores de U ocorre o problema de interferência devido aos produtos de fissão do isótopo 235U. Um dos métodos de tratar este problema é fazer a correção usando fatores de interferência devido à fissão do U para os radionuclídeos utilizados nas análises dos elementos. No presente estudo foram determinados os valores dos fatores de interferência devido à fissão do U para os radioisótopos 141Ce, 143Ce,140La, 99Mo, 147Nd, 153Sm e 95Zr no reator nuclear de pesquisas IEA-R1 do IPEN-CNEN/SP. Esses fatores de interferência foram determinados experimentalmente, por meio da irradiação dos padrões sintéticos em uma determinada posição do reator, e teoricamente, determinando a razão dos fluxos de nêutrons epitérmicos e térmicos na mesma posição onde os padrões sintéticos foram irradiados e utilizando parâmetros nucleares da literatura. Os fatores de interferência obtidos foram comparados com os valores reportados em outros estudos. Para avaliar esses fatores de interferência, eles foram aplicados em análises dos elementos alvo deste estudo, nos materiais de referência certificados NIST 8704 Buffalo River Sediment, IRMM BCR-667 Estuarine Sediment e IAEA-SL-1 Lake Sediment.
Neutron activation analysis is a method used in the determination of several elements in different kinds of matrices. However, when the sample contains high U levels the problem of 235U fission interference occurs. A way to solve this problem is to perform the correction using the interference factor due to U fission for the radionuclides used on elemental analysis. In this study, the interference factors due to U fission for the radioisotopes 141Ce, 143Ce, 140La, 99Mo, 147Nd, 153Sm and 95Zr in the research nuclear reactor IEA-R1 at IPEN-CNEN/SP were determined. These interference factors were determined experimentally, by irradiation of synthetic standards in a selected position in the reactor, and theoretically, determining the epithermal to neutron fluxes ratio in the same position where synthetic standards were irradiated and using reported nuclear parameters on the literature. The obtained interference factors were compared with values reported by other works. To evaluate the reliability of these factors they were applied in the analysis of studied elements in the certified reference materials NIST 8704 Buffalo River Sediment, IRMM BCR- 667 Estuarine Sediment e IAEA-SL-1 Lake Sediment.
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GERALDO, BIANCA. "Utilização de métodos radioanalíticos para a determinação de isótopos de urânio, netúnio, plutônio, amerício e cúrio em rejeitos radioativos." reponame:Repositório Institucional do IPEN, 2012. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9942.

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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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RIBEIRO, JUNIOR IBERE S. "Determinação de fatores de interferência de produtos de fissão do urânio na análise por ativação neutrônica." reponame:Repositório Institucional do IPEN, 2014. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11801.

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Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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7

Dickinson, Shirley. "Infrared spectroscopic and mass spectrometric studies of high-temperature molecules relevant to severe nuclear reactor accidents." Thesis, University of Southampton, 1990. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.255768.

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8

Kennedy, William B. (William Blake) 1979. "Analysis of the MIT research reactor fission product and actinide radioactivity inventories." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/32723.

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Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Physics, 2004.
MIT Institute Archives copy: leaves 92-111 bound in reverse order.
Includes bibliographical references (leaf 57).
The current analysis of the MITR core radioactivity inventory eliminates unnecessary assumptions made in previous estimates of the inventory, and revises the list of contributory isotopes to include all actinide and fission product isotopes necessary for a proper accident source term calculation. The result is a power-history-dependent inventory that increases with bum-up, and comprises 41 actinide isotopes and 596 fission product isotopes. The analysis uses the ORIGEN2 depletion code to calculate the activity of actinide and fission product isotopes for eight MITR input models at 32 intervals over a period of 5376MWD. The input models simulate a MITR core loaded with high- enrichment, U-Alx cermet fuel or low-enrichment, monolithic U-Mo fuel, and operated at 6MW with a continuous-burn-up or cyclic-burn-up-and-decay power history. Reorganization of the ORIGEN2 output file, and application of an element reduction criterion creates the condensed matrix file for each MITR input model. This file lists the contribution of each isotope to the core radioactivity inventory at each output interval, and is the basis for all inventory analysis. The inventory analysis yields three important conclusions. First, the assumption of an equilibrium inventory of isotopes in the fuel is accurate to within 3% for all time after 10% fuel bum-up, and conservative over the entire fuel cycle. The equilibrium fuel assumption is invalid for the actinides due to a slow rate of inventory growth. Second, the cyclic-bum-up-and-decay power history yields a lower core inventory than the continuous-burn-up power history for both fuel enrichments. The difference is minimized by increasing the ratio of irradiation time to decay time.
(cont.) Finally, the analysis indicates that conversion to a U-Mo fuel will produce an actinide inventory 18 times greater than that of the current U-Alx fuel, with no significant change in the fission product inventory. However, the actinide inventory is a small fraction of the fission product inventory. The worst-case core inventory available for release is 2.91 E+7Ci for the high-enrichment fuel, and 2.94E+7Ci for the low-enrichment fuel, with a core loading of 24 elements in each case. The best-estimate core inventory available for release is 2.83E+7Ci, and 2.82E+7Ci respectively, and accounts for typical cyclic operation of the MITR.
by William B. Kennedy.
S.B.
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9

Halonen, Kimmo. "Gamma spectrum analysis of fission product release during accidental conditions: focus on ruthenium release during air ingress." Thesis, KTH, Fysik, 2012. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-103715.

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10

Belhabib, Tayeb. "Comportement thermique des défauts lacunaires induits par l'hélium et les gaz de fission dans le dioxyde d'uranium." Phd thesis, Université d'Orléans, 2012. http://tel.archives-ouvertes.fr/tel-00831705.

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Dans les nouvelles centrales nucléaires dites 4ème génération, comme d'ailleurs les anciennes, le dioxyde d'uranium devra opérer dans des milieux hostiles de températures et d'irradiation avec la présence des produits de fission (PF) et des particules alpha (α). Le fonctionnement dans ces conditions extrêmes induira des déplacements d'atomes et dégradera les propriétés thermiques et mécaniques du combustible UO2. La compréhension du comportement des défauts lacunaires, des PF et de l'hélium est cruciale pour prévoir le comportement du dioxyde d'uranium au sein de ces futures installations nucléaires. La première partie de cette thèse est consacrée à l'étude des défauts lacunaires induits par l'implantation de krypton et d'iode (quelques MeV) dans l'UO2 polycristallin et leurs stades de recuits. L'analyse par spectroscopie d'annihilation de positons (PAS) a permis de mettre en évidence la création de défauts de Schottky VU-2VO dans le cas des implantations iode et la formation de clusters lacunaires contenant du gaz pour les implantations krypton. L'évolution en température de ces défauts générés dépend des paramètres d'implantation (nature des ions, énergie, fluence). Cette étude a montré les rôles importants que peuvent jouer les défauts lacunaires et la présence des gaz de fission dans l'évolution du matériau UO2. Ensuite, nous nous sommes intéressés à l'étude et à la caractérisation, par PAS et les techniques d'analyse par faisceau d'ions (NRA/C et RBS/C), du comportement de l'hélium dans l'UO2. Les mesures de NRA/C et RBS/C révèlent une localisation d'une grande fraction d'hélium dans les sites interstitiels octaédriques de la matrice UO2. La localisation de l'hélium reste stable dans ces sites pour T< 600°C, évoluent légèrement entre 600 et 700°C et devient aléatoire à 800°C. Les mesures PAS mettent en évidence trois stades d'évolution des défauts lacunaires : la recombinaison par migration des interstitiels d'oxygène, l'agglomération des défauts entre 600 et 800°C et leur dissociation et élimination lorsque la température augmente. Ces résultats suggèrent que le transport d'hélium est assisté par les défauts lacunaires.
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Books on the topic "Fission products Analysis"

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Holliger, Philippe. The new OKLO reaction zones: U-Pb dating and in situ characterization of fission products by ion analysis : report on Progress 1991. Grenoble: Centre d'Études Nucléaires de Grenoble, 1993.

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Leonard, Soffer, and U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Systems Technology., eds. Accident source terms for light-water nuclear power plants: Final report. Washington, DC: Division of Systems Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1995.

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Leonard, Soffer, and U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Systems Technology., eds. Accident source terms for light-water nuclear power plants: Final report. Washington, DC: Division of Systems Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1995.

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Book chapters on the topic "Fission products Analysis"

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Gouget, Karine, Fulvia Verde, and Antoni Barrientos. "In Vivo Labeling and Analysis of Mitochondrial Translation Products in Budding and in Fission Yeasts." In Membrane Trafficking, 113–24. Totowa, NJ: Humana Press, 2008. http://dx.doi.org/10.1007/978-1-59745-261-8_8.

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Zohuri, Bahman. "Fission Product Buildup and Decay." In Neutronic Analysis For Nuclear Reactor Systems, 483–99. Cham: Springer International Publishing, 2019. http://dx.doi.org/10.1007/978-3-030-04906-5_15.

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Zohuri, Bahman. "Fission Product Buildup and Decay." In Neutronic Analysis For Nuclear Reactor Systems, 491–507. Cham: Springer International Publishing, 2016. http://dx.doi.org/10.1007/978-3-319-42964-9_15.

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Mondot, J., J. P. Chauvin, and J. P. West. "Validation of Fission Product Capture Cross Sections by the Analysis of Thermal and Epithermal Integral Experiments." In Nuclear Data for Science and Technology, 29–34. Berlin, Heidelberg: Springer Berlin Heidelberg, 1992. http://dx.doi.org/10.1007/978-3-642-58113-7_6.

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Ponomarev-Stepnoi, N. N., and A. A. Khrulev. "Effect of the Annealing Temperature on Escape of Metal Fission Products from Different Media (Features of Experimental Data Analysis)." In Fission Product Transport Processes in Reactor Accidents, 735–62. CRC Press, 2020. http://dx.doi.org/10.1201/9781003070344-61.

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"SAFETY ANALYSIS NEEDS AND MAIN PHENOMENA TO BE STUDIED." In The Phebus Fission Product Project, 115–18. CRC Press, 2003. http://dx.doi.org/10.1201/9781482286779-14.

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"SAFETY ANALYSIS NEEDS AND MAIN PHENOMENA TO BE STUDIED." In The Phebus Fission Product Project, 115–18. CRC Press, 2003. http://dx.doi.org/10.1201/9781482286779-14.

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"SURVEY OF SEVERE ACCIDENT EXPERIMENTS AND ANALYSES IN JAPAN." In The Phebus Fission Product Project, 32–43. CRC Press, 2003. http://dx.doi.org/10.1201/9781482286779-6.

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"SURVEY OF SEVERE ACCIDENT EXPERIMENTS AND ANALYSES IN JAPAN." In The Phebus Fission Product Project, 32–43. CRC Press, 2003. http://dx.doi.org/10.1201/9781482286779-6.

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Arutunjan, R. V., L. A. Bolshov, V. V. Vitukov, V. M. Goloviznin, A. M. Dykhne, V. P. Kiselev, S. V. Klementova, et al. "Theoretical Analysis and Numerical Modeling of Heat Transfer and Fuel Migration in Underlying Soils and Constructive Elements of Nuclear Plants during an Accident Release from the Core." In Fission Product Transport Processes in Reactor Accidents, 789–98. CRC Press, 2020. http://dx.doi.org/10.1201/9781003070344-64.

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Conference papers on the topic "Fission products Analysis"

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WANG, ENHONG, N. T. BREWER, J. H. HAMILTON, A. V. RAMAYYA, J. K. HWANG, Y. X. LUO, J. O. RASMUSSEN, S. J. ZHU, G. M. TER-AKOPIAN, and YU TS OGANESSIAN. "FOUR-FOLD DATA ANALYSIS OF 252Cf FISSION PRODUCTS." In Proceedings of the Fifth International Conference on ICFN5. WORLD SCIENTIFIC, 2013. http://dx.doi.org/10.1142/9789814525435_0084.

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Tian Chenyang, Guo Huiping, Lv Ning, Zhao Kuo, Ma Wenyan, Xu Peng, Zuo Guangxia, Lv Wenhui, and Li Jinjun. "Analysis on delayed gamma spectra of products from uranium fission." In 2015 12th IEEE International Conference on Electronic Measurement & Instruments (ICEMI). IEEE, 2015. http://dx.doi.org/10.1109/icemi.2015.7494172.

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Cao, Jianzhu, Tao Liu, Yuanyu Wu, Hong Li, and Yuanzhong Liu. "Analysis of Radioactive Source Term for Modular HTGR During Normal Operation." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-30075.

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The methods of radioactive source term analysis are introduced in detail for the modular high temperature gas cooled reactor in China. Radioactive fission products and activation products produced in the reactor are described. For fission products, the emphasis is on the process from production through release to the environment for noble gas, iodine and long-lived metallic nuclides. For activation products, it mainly introduces the behaviors of H-3 and C-14. Especially the permeation process from primary circuit to secondary circuit is described for H-3. Using the preliminary design parameters of demonstration HTGR in China, basic prediction of radioactive source term is done and the results are given.
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Yun, J. I., K. Y. Suh, and C. S. Kang. "Heat and Fission Product Transport in a Molten U-Zr-O Pool With Crust." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22438.

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Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the pool. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool was estimated by product of the mass concentration and energy conversion factor of each fission product. For the calculation of heat generation rate in the pool, twenty-nine elements were chosen and classified by their chemical properties. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis was performed for the TMI-2 accident. The pool is assumed to be a partially filled hemispherical geometry and the change of pool geometry during the numerical calculation was neglected. Results of the numerical calculation revealed that the peak temperature of the molten pool significantly decreased and most of the volatile fission products were released from the molten pool during the accident.
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Porcheron, Emmanuel, and Pascal Lemaitre. "Analysis of Aerosol Collection by Droplets: Application to Fission Products Removal in Case of Severe Accident." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48582.

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TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Suˆrete´ Nucle´aire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulate typical accidental thermal hydraulic flow conditions in nuclear Pressurized Water Reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety CFD code validation. The present work is devoted to study the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between the spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we performed detailed characterization of the two-phase flow.
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Blaise, P., S. Cathalau, N. Thiollay, P. Fougeras, V. Laval, H. Philibert, J. M. Girard, and J. P. Hudelot. "Fission Products Particular Peak Measurement for UO2-Gd2O3-MOX γ-Scanning Renormalization in 100% MOX ABWR Mock-Up Cores." In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49464.

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This paper presents the principles of the peak check measurements by gamma spectrometry. One details the main equations used for the analysis of the raw data as the calculation of the different sources of uncertainties and their propagation on the result. The method is illustrated with actual examples from the French-Japanese BASALA ABWR 100% MOX experimental program. • For each individual fission rate, the systematical uncertainty due to the radioactive decay data is smaller than the statistical uncertainty due to the counting process. • The main part of the final uncertainty on the scaling factor is brought by the systematical uncertainty on the average fission yields. This study enables to propose some recommendations for fission products and nuclear data evaluation used. • Only the analysis of the 140La peak at 1596 keV leads to acceptable uncertainties on the fission rate maps renormalization, with a good consistency with the integral γ-scanning results. • The JEF-2.2 evaluation on the fission yields and their associated uncertainties seems more realistic than the ENDF/B-6 and is recommended for the scaling factor analysis.
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Kim, Young Min, M. S. Cho, Y. W. Lee, and W. J. Lee. "Development of a Fuel Performance Analysis Code COPA." In Fourth International Topical Meeting on High Temperature Reactor Technology. ASMEDC, 2008. http://dx.doi.org/10.1115/htr2008-58040.

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A fuel performance analysis code for a very high temperature gas-cooled reactor (VHTR) COPA (Coated Particle) is being developed at the Korea Atomic Energy Research Institute (KAERI). The COPA code consists of nine modules: BURN, TEMTR, TEMPEB, TEMBL, MECH, FAIL, FPREL, ABAQ, and MPRO. The BURN determines neutron flux and fluence at a location of a reactor core, and then calculates a fuel burnup, a fission rate per volume and a fission product inventory throughout a fuel particle and a fuel element. The TEMTR, TEMPEB and TEMBL calculate the temperature distributions in a coated fuel particle, a pebble and a fuel block by using a one-dimensional finite difference method, respectively. The MECH performs mechanical calculations on a coated fuel particle by using a finite element method. The FAIL performs probabilistic calculations to estimate the failure probabilities of the coating layers during an experiment or a reactor operation. The FPREL estimates the migrations of gaseous and metallic fission products through a fuel particle and a fuel element by using a one-dimensional finite difference method. The ABAQ performs the analysis of the crack and debonding in a coated fuel particle. The MPRO calculates the material properties of the kernel, low-density pyrocarbon, high-density pyrocarbon, silicon carbide, matrix graphite, and structural graphite. Each module is used to produce input data for other modules or is inserted into other modules. The COPA code is one of the computer codes taking part in the IAEA-CRP-6 benchmarking program. The stresses and failure fractions calculated by the COPA-MECH and COPA-FAIL showed good agreements with the results by the other countries’ codes. In order to establish a good database of the related material properties, KAERI is participating in an international irradiation experiment, is planning its own irradiation and post-irradiation experiments, and will perform ab-initio calculations on the fuel materials.
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Li, Ran, and Jiyang Yu. "Development of PCCSAP-3D Code for Passive Containment: Models of Noncondensable Gases, Aerosols and Fission Products." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-15606.

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PCCSAC-3D is a code originally developed for AC600 containment thermo-hydraulic analysis. Its validated capabilities include simulating the behaviors of steam-air mixture and liquid water under the unique conditions of an AC600/AP1000 containment after a DBA. The film-tracking model applied gives it the ability to simulate the liquid film both outside and inside the steel containment. Refined with some new models, the new version of the code, named PCCSAP-3D, can cover hydrogen behavior, fission products behavior (in the form of gas and aerosol) and iodine behavior. In the module of noncondensable gas, diffusion of up to 11 species are taken into consideration. A user-definable recombiner/ignitor model is developed to accommodate different types of hydrogen recombiners and ignitors. Given the source term as a boundary condition, the fission products model would be able to track up to 64 radio-isotopes after a LOCA. The leakage and spontaneous decay is accounted for all of these nuclides. Besides, the noble gases, gaseous iodine and fission product aerosols are treated separately. There is no removal mechanism of noble gases. Whereas removal mechanisms of radio-aerosols considered include spray, gravitational sedimentation, diffusio-phoresis and thermo-phoresis. A simple model for gaseous iodine comprises organic iodine and elemental iodine, in which the effects of spray and liquid adsorption are treated integrally. To evaluate the radioactivity consequences of a certain accident, a radioactivity calculation model is brought out to convert the molar concentration or mass concentration of radioactive material into radioactivity concentration. The new version of PCCSAP-3D code with models aforementioned is preliminarily validated by comparing the simulation results with safety analysis results reported in AP1000 Design Control Document. The accident scenario is set as a design basic LOCA with core melt.
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Kienzler, Bernhard, and Ernesto González-Robles. "State-of-the-Art on Instant Release of Fission Products From Spent Nuclear Fuel." In ASME 2013 15th International Conference on Environmental Remediation and Radioactive Waste Management. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icem2013-96044.

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Within the EURATOM FP7 Collaborative Project “Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides)”, a State-of-the-Art Report was prepared. The fast / instant release fraction (IRF) is defined as a fraction of the inventory of radionuclides that may be rapidly released from the fuel and fuel assembly materials at the time of canister breaching. In the context of safety analysis for a repository, the time span for mobilization of this fraction can be considered instantaneously, even if the process takes some time in experiments. Radionuclides contributing to the fast release are fission gases (Kr and Xe), easily soluble elements such as cesium and iodine, and other elements which are hardly incorporated in the UO2 crystal lattice. The present contribution summarizes the results obtained from published studies focused on rapid release experiments carried out with different spent nuclear fuel (SNF), samples, sizes, techniques (batch and flow-through), and durations. A total of 80 experiments cover the study of UO2 SNF from pressure water reactors (PWR) of different initial enrichments and burn-up, while 20 experiments were performed with UO2 SNF from boiling water reactors (BWR) and 8 with MOX fuel.
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Tang, Changbing, Yongjun Jiao, Wenjie Li, Tao Qing, Yifei Miao, and Ping Chen. "Numerical Simulation of Different Sizes Missing Pellet Surface Effects on Thermal-Mechanical Behaviors in Nuclear Fuel Rods." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60116.

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Nuclear fuel rods is mainly composed of uranium dioxide pellets and zirconium alloy cladding, there is a gap between pellets and cladding, which is filled with helium. Under the reactor operation conditions, pellets produce a lot of heat by nuclear fission reactions and at the same time also produce lots of radioactive fission products. Cladding serve as the first barrier to accommodate radioactive fission product, needs to maintain its structural integrity under the reactor operation conditions. Cladding stresses can be effectively limited by controlling power increase rates. However, pellet manufacturing defects such as missing pellet surface (MPS), may lead to cladding local stress significantly high to cause cladding failure. Simulating the impact of these defects correctly can help prevent these types of failure. MPS defects are 3D phenomenon, needs 3D modeling method to study the influence of these defects on the cladding .In this paper, stress update algorithm is derived, with the help of ABAQUS (a commercial finite element software), simulated the thermal-mechanical behaviors of the MPS defects fuel rod with a 3D FEM and completed the sensitivity analysis of MPS defects size for the fuel performance. The models included in this simulation, including pellet irradiation swelling (fission gas products induced swelling and fission solid products induced swelling), pellet densification, pellet relocation, pellet thermal expansion, pellet irradiation creep, pellet irradiation hardening, cladding irradiation growth, cladding thermal expansion, cladding thermal creep, cladding irradiation creep, cladding irradiation hardening and gap heat transfer (gas heat conduction, radiation heat transfer and contact heat conduction) etc. Furthermore, considering the effects of irradiation and temperature on the material parameters such as thermal conductivity, specific heat and young’s modulus etc. According to the simulation result, showing that MPS defects have a large impact on the performance of fuel rods, this impact will be more obvious with the size of MPS defects increase. The MPS defects cause larger gap distance between pellet and cladding, higher gap distance causes smaller gap conductance, and then causes elevated temperature at the center of the pellet and in the region of the pellet adjacent to the defect. The cladding temperature is reduced in the area immediately across from the defect, and is elevated in neighboring areas. Meanwhile, MPS defects clearly have a significant effect on stress distribution and maximum stress of the cladding, cause high tensile stresses in the inner surface of the cladding and high compressive stresses on the outer surface of the cladding at the center of the defect. Around the boundaries of the defect, the stresses are reversed, with high compressive stresses on the cladding interior and high tensile stresses on the cladding exterior.
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Reports on the topic "Fission products Analysis"

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Demkowicz, Paul A., Jason M. Harp, Philip L. Winston, and Scott A. Ploger. Analysis of Fission Products on the AGR-1 Capsule Components. Office of Scientific and Technical Information (OSTI), March 2013. http://dx.doi.org/10.2172/1097137.

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Petrova, Petya H., and Pavlin P. Groudev. Analysis of the Fission Products Behaviour in the Phebus FPT1 Experiment by Using the ASTEC V2.1 Code. "Prof. Marin Drinov" Publishing House of Bulgarian Academy of Sciences, May 2019. http://dx.doi.org/10.7546/crabs.2019.05.05.

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Simpson, Michael F., Supathorn Phongikaroon, and Jinsuo Zhang. Development and Optimization of Voltammetric Methods for Real Time Analysis of Electrorefiner Salt with High Concentrations of Actinides and Fission Products. Office of Scientific and Technical Information (OSTI), March 2018. http://dx.doi.org/10.2172/1432793.

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Chapman, Carolyn R. Analysis of the Fission Yeast Rad3+ Gene Product. Fort Belvoir, VA: Defense Technical Information Center, January 1999. http://dx.doi.org/10.21236/ada368445.

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Horne, Steven M., and Kevin R. Jackman. Multispectral Gamma-Ray Analysis Using Clover Detectors with Application to Uranium Fission Product Analysis. Office of Scientific and Technical Information (OSTI), April 2013. http://dx.doi.org/10.2172/1077011.

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Martin, R. C. ANALYSIS OF FISSION PRODUCT RELEASE DATA FOR GERMAN FUEL SPHERE HFR-K3/3. Office of Scientific and Technical Information (OSTI), September 1993. http://dx.doi.org/10.2172/10199681.

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Suh, K. Y. Modeling of in-vessel fission product release including fuel morphology effects for severe accident analyses. Office of Scientific and Technical Information (OSTI), October 1989. http://dx.doi.org/10.2172/7261581.

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Harp, Jason M. Analysis of Individual Compact Fission Product Inventory and Burnup for the AGR-1 TRISO Experiment using Gamma Spectrometry. Office of Scientific and Technical Information (OSTI), December 2010. http://dx.doi.org/10.2172/1494149.

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Rest, J. The DART dispersion analysis research tool: A mechanistic model for predicting fission-product-induced swelling of aluminum dispersion fuels. User`s guide for mainframe, workstation, and personal computer applications. Office of Scientific and Technical Information (OSTI), August 1995. http://dx.doi.org/10.2172/149983.

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McCartney, M. A., and M. G. Plys. Modifications for the development of the MAAP-DOE code: Volume 1, A mechanistic model for core-concrete interactions and fission product release in integrated accident analysis Task 3. 4. 3. Office of Scientific and Technical Information (OSTI), November 1988. http://dx.doi.org/10.2172/6300751.

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