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Relevant bibliographies by topics / Origen-Arp

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Academic literature on the topic 'Origen-Arp'

Author: Grafiati

Published: 4 June 2021

Last updated: 1 February 2022

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Journal articles on the topic "Origen-Arp"

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McGraw, Carolyn, and Germina Ilas. "Pressurized Water Reactor ENDF/B-VII Cross-Section Libraries for ORIGEN-ARP." Nuclear Technology 183, no. 3 (2013): 436–45. http://dx.doi.org/10.13182/nt13-a19431.

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Bowman, Stephen M., Luiz C. Leal, Otto W. Hermann, and Cecil V. Parks. "ORIGEN-ARP, A Fast and Easy-to-Use Source Term Generation Tool." Journal of Nuclear Science and Technology 37, sup1 (2000): 575–79. http://dx.doi.org/10.1080/00223131.2000.10874953.

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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 (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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Malec, J., K. Ambrožič, and M. Kromar. "DEVELOPMENT OF FUEL CHARACTERIZATION TOOL BASED ON LIBRARY INTERPOLATION." EPJ Web of Conferences 247 (2021): 12004. http://dx.doi.org/10.1051/epjconf/202124712004.

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The feasibility of using Origen+ARP code for depletion and decay calculations for Krško NPP was tested by performing depletion and decay calculations using interpolated libraries and comparing the results to the ones calculated from non-interpolated libraries in order to evaluated the number of libraries needed in order to interpolate fuel properties with sufficient precision for any realistic burnup scenario. For Krško NPP fuel, using three interpolation libraries with different decay heat parameters was enough to bring the approximation error bellow 0:5 % when comparing fuel decay heat throu

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Moon, Hyun. "Radiation release characterization of PWR spent fuel assemblies generated from Korean nuclear power plants." Nuclear Technology and Radiation Protection 24, no. 3 (2009): 183–87. http://dx.doi.org/10.2298/ntrp0903183m.

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Spent nuclear fuel should be kept under safe management until it is disposed of permanently. Because of this, it is important to understand its radiation release characteristics. In this paper, the Monte Carlo method is applied to evaluate the radiation release characteristics of two types of PWR spent fuel assembly generated from the operating plants in Korea: Westinghouse and Korea Standard Nuclear Power Plant. The source terms were calculated using ORIGEN-ARP. The neutron and photon (or gamma) dose distributions along the vertical and horizontal directions of each spent fuel assembly were e

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Lee, Se Geun, and Jae Hak Cheong. "Neutron Activation of Structural Materials of a Dry Storage System for Spent Nuclear Fuel and Implications for Radioactive Waste Management." Energies 13, no. 20 (2020): 5325. http://dx.doi.org/10.3390/en13205325.

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In order to estimate the radiological characteristics of disused dry storage systems for spent nuclear fuel, a stepwise framework to calculate neutron sources (ORIGEN-ARP), incident neutron flux and reaction rate (MCNPX), effective cross-section (hand calculation), and residual activity (ORIGEN-2) was established. Applicability of the framework was demonstrated by comparing the residual activity of a commercialized storage system, HI-STORM 100, listed in the safety analysis report and calculated in this study. For a reference case assuming an impurity-free storage system, the modified effectiv

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Czakoj, Tomas, Jan Frybort, and Martin Lovecky. "CALCULATION UNCERTAINTIES IN SPENT FUEL INVENTORY DETERMINATION." Acta Polytechnica CTU Proceedings 14 (May 17, 2018): 14. http://dx.doi.org/10.14311/app.2018.14.0014.

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In the depletion calculation of the nuclear fuel, the uncertainty is of utmost importance, as it affects the uncertainty of the subsequent calculation, when the calculated composition is used. The calculations are even more important when they are safety related, e. g., when determining the reactivity or emissions of radioactivity to the environment. This work covers the depletion model of Gd-2M+ fuel which was developed in ORIGEN-ARP/TRITON calculation sequences, both being parts of a SCALE 6.2.1 package. The uncertainties of the respective calculation model were determined by comparison with

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Barkauskas, Vytenis, Rita Plukienė, Artūras Plukis, and Vidmantas Remeikis. "generation of RBMK-1500 spent nuclear fuel one-group cross-section libraries and their evaluation against experimental data." Lithuanian Journal of Physics 57, no. 1 (2017): 42–53. http://dx.doi.org/10.3952/physics.v57i1.3455.

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Depletion of RBMK-1500 spent nuclear fuel (SNF) with and without an erbium burnable absorber was modelled, and one-group burn-up dependent cross-section libraries for Origen-ARP were created. Depletion calculations for the generation of cross-section libraries were performed using the SCALE 6.1 code package with the TRITON control module, which employs the NEWT deterministic 2D transport code with the 238-group energy library based on the ENDF-B VII library and the ORIGEN-S nuclide composition calculation code. Concentrations of the most important actinides for criticality safety were calculat

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Valls, Alba, Mireia Grivé, Olga Riba, Maita Morales, and Kastriot Spahiu. "Estimation of the long term helium production in high burn-up spent fuel due to alpha decay and consequences for the canister." MRS Proceedings 1665 (2014): 297–302. http://dx.doi.org/10.1557/opl.2014.658.

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ABSTRACTIn the KBS-3 repository concept and safety analysis, the copper container with a cast iron insert plays a central role in assuring isolation of the waste from the surrounding during long periods of time. All processes that affect its stability are thoroughly analysed, including potential detrimental processes inside the canister. For this reason, an estimation of the helium produced during the long term decay of alpha emitters in the spent fuel is necessary to evaluate if the pressures generated inside can have consequences for the canister.The spent nuclear fuel to be disposed of in S

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Fylonych, Yu, V. Zaporozhan, O. Balashevskyi, and K. Merkotan. "Analysis of the influence of nuclear fuel burnup on the 16N formation rate in the primary coolant of the WWER-1000 reactor." Nuclear Physics and Atomic Energy 22, no. 1 (2021): 48–55. http://dx.doi.org/10.15407/jnpae2021.01.048.

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The developed model of the WWER-1000 reactor using MCNP6.2 (Monte Carlo N-Particle Transport Code) includes the detailed core taking into account the design of the fuel assemblies, as well as the baffle, the lower plenum, the fuel support columns, the core barrel, a downcomer, and the reactor pressure vessel. It allows implementing multifunctional calculations such as recriticality with various fuel configurations, the critical concentration of boric acid, determination of the axial and radial peaking factor in the reactor core, etc. For obtaining the more precise result of the cumulation nitr

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Dissertations / Theses on the topic "Origen-Arp"

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Cotza, Gabriele. "Evoluzione dell’inventario di radioisotopi rilevanti ai fini del trattato CTBT per il bando dei test nucleari, tramite l’utilizzo del codice di calcolo ORIGEN-ARP." Bachelor's thesis, Alma Mater Studiorum - Università di Bologna, 2016.

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L’oggetto di questa tesi è lo studio dei radionuclidi rilevanti che vengono generati dalla detonazione di ordigni nucleari. L’obiettivo principale è la ricerca di parametri legati alle sostanze radioattive derivanti dalla fissione nucleare che permettano di capire se c’è stato un rilascio non autorizzato di radioisotopi legati all’esplosione di una bomba nucleare. Inoltre dall’inventario dei nuclidi rilevati si vuole cercare di capire che tipo di tecnologia è stata utilizzata per la costruzione dell’ordigno facendo riferimento a specifici rapporti isotopici. Per ricavare il probabile inventari

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Chambers, Angela Sue. "A comparison of nuclide production and depletion using MCNPX and ORIGEN-ARP reactor models and a sensitivity study of reactor design parameters using MCNPX for nuclear forensics purposes." Thesis, 2010. http://hdl.handle.net/2152/ETD-UT-2010-05-853.

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The Oak Ridge Isotope Generation and Depletion – Automatic Rapid Proccessing (ORIGEN-ARP) deterministic code has been extensively utilized for determining nuclide concentrations at various specific burnup values for a variety of nuclear reactor designs. Given nuclide concentrations or ratios, such calculations can be used in nuclear forensics and nuclear non-proliferation applications to reverse-calculate the type of reactor and specific burnup of the fuel from which the nuclides originated. Recently, Los Alamos National Laboratory has released a version of its probabilistic radiation transpor

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Conference papers on the topic "Origen-Arp"

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Pavelescu, Alexandru Octavian, Dan Gabriel Cepraga, Konstantina Voukelatou, and Renato Tinti. "Estimation of Clearance Potential Index and Hazard Factors of CANDU Fuel Bundle and Its Validation Based on the Measurements of Radioisotopes Inventories From Pickering Reactor Fuel." In The 11th International Conference on Environmental Remediation and Radioactive Waste Management. ASMEDC, 2007. http://dx.doi.org/10.1115/icem2007-7086.

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This paper is related to the clearance potential levels, ingestion and inhalation hazard factors of the spent nuclear fuel and radioactive wastes. This study required a complex activity that consisted of more steps such as: the acquisition, setting up, validation and application of procedures, codes and libraries. The paper reflects the validation stage of this study. Its objective was to compare the measured inventories of selected actinide and fission products radionuclides in an element from the Pickering CANDU reactor with the inventories predicted using a recent version of the SCALE 5\ORI

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Reports on the topic "Origen-Arp"

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Gauld, I. C. MOX Cross-Section Libraries for ORIGEN-ARP. Office of Scientific and Technical Information (OSTI), 2003. http://dx.doi.org/10.2172/885544.

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Murphy, BD. ORIGEN-ARP Cross-Section Libraries for Magnox, Advanced Gas-Cooled, and VVER Reactor Designs. Office of Scientific and Technical Information (OSTI), 2004. http://dx.doi.org/10.2172/861731.

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Leal, L. C., O. W. Hermann, S. M. Bowman, and C. V. Parks. ARP: Automatic rapid processing for the generation of problem dependent SAS2H/ORIGEN-s cross section libraries. Office of Scientific and Technical Information (OSTI), 1998. http://dx.doi.org/10.2172/296737.

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