Academic literature on the topic 'Nuclear reaction codes'
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Journal articles on the topic "Nuclear reaction codes"
Kataria, S. K., V. S. Ramamurthy, M. Blann, and T. T. Komoto. "Shell-dependent level densities in nuclear reaction codes." Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment 288, no. 2-3 (March 1990): 585–88. http://dx.doi.org/10.1016/0168-9002(90)90155-y.
Full textDenikin, Andrey, Alexander Karpov, Mikhail Naumenko, Vladimir Rachkov, Viacheslav Samarin, and Vycheslav Saiko. "Synergy of Nuclear Data and Nuclear Theory Online." EPJ Web of Conferences 239 (2020): 03021. http://dx.doi.org/10.1051/epjconf/202023903021.
Full textÖzdoğan, H., İsmail Hakki Sarpün, Mert Şekerci, and Abdullah Kaplan. "Production cross-section calculations of 111In via proton and alpha-induced nuclear reactions." Modern Physics Letters A 36, no. 08 (February 18, 2021): 2150051. http://dx.doi.org/10.1142/s0217732321500516.
Full textHilaire, Stephane, Eric Bauge, Pierre Chau Huu-Tai, Marc Dupuis, Sophie Péru, Olivier Roig, Pascal Romain, and Stephane Goriely. "Potential sources of uncertainties in nuclear reaction modeling." EPJ Nuclear Sciences & Technologies 4 (2018): 16. http://dx.doi.org/10.1051/epjn/2018014.
Full textSarpün, İsmail Hakki, Hasan Özdoğan, Kemal Taşdöven, Hüseyin Ali Yalim, and Abdullah Kaplan. "Theoretical photoneutron cross-section calculations on Osmium isotopes by Talys and Empire codes." Modern Physics Letters A 34, no. 26 (August 30, 2019): 1950210. http://dx.doi.org/10.1142/s0217732319502109.
Full textŞekerci, Mert, Hasan Özdoğan, and Abdullah Kaplan. "Level density model effects on the production cross-section calculations of some medical isotopes via (α, xn) reactions where x = 1–3." Modern Physics Letters A 35, no. 24 (June 23, 2020): 2050202. http://dx.doi.org/10.1142/s0217732320502028.
Full textHenning, Greg, Antoine Bacquias, Catalin Borcea, Mariam Boromiza, Roberto Capote, Philippe Dessagne, Jean-Claude Drohé, et al. "MEASUREMENT OF 182,184,186W (N, N’ γ) CROSS SECTIONS AND WHAT WE CAN LEARN FROM IT." EPJ Web of Conferences 247 (2021): 09003. http://dx.doi.org/10.1051/epjconf/202124709003.
Full textSabra, M. S., Robert A. Weller, Marcus H. Mendenhall, Robert A. Reed, Michael A. Clemens, and A. F. Barghouty. "Validation of Nuclear Reaction Codes for Proton-Induced Radiation Effects: The Case for CEM03." IEEE Transactions on Nuclear Science 58, no. 6 (December 2011): 3134–38. http://dx.doi.org/10.1109/tns.2011.2169989.
Full textKorbut, Tamara, Maksim Kravchenko, Ivan Edchik, and Sergey Korneev. "Yalina-thermal facility neutron characteristic computational study 129I, 237Np and 243Am transmutation reaction rates calculations." EPJ Web of Conferences 239 (2020): 22013. http://dx.doi.org/10.1051/epjconf/202023922013.
Full textVoinov, A. V., S. M. Grimes, C. R. Brune, A. Bürger, A. Görgen, M. Guttormsen, A. C. Larsen, T. N. Massey, and S. Siem. "Level Density Inputs in Nuclear Reaction Codes and the Role of the Spin Cutoff Parameter." Nuclear Data Sheets 119 (May 2014): 255–57. http://dx.doi.org/10.1016/j.nds.2014.08.070.
Full textDissertations / Theses on the topic "Nuclear reaction codes"
MAI, LUIZ A. "Sistema de obtencao de um pre-projeto otimizado de um nucleo de um reator nuclear." reponame:Repositório Institucional do IPEN, 1988. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9914.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
HIROMOTO, MARIA Y. K. "PSINCO-um programa para calculo da distribuicao de potencia e supervisao do nucleo de reatores nucleares, utilizando sinais de detetores tipo 'SPD'." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10706.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
CARVALHO, LUIZ S. "Frequencia de danos no nucleo por blecaute em reator nuclear de concepcao avancada." reponame:Repositório Institucional do IPEN, 2004. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11147.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
Laufer, Michael Robert. "Granular Dynamics in Pebble Bed Reactor Cores." Thesis, University of California, Berkeley, 2013. http://pqdtopen.proquest.com/#viewpdf?dispub=3593891.
Full textThis study focused on developing a better understanding of granular dynamics in pebble bed reactor cores through experimental work and computer simulations. The work completed includes analysis of pebble motion data from three scaled experiments based on the annular core of the Pebble Bed Fluoride Salt-Cooled High- Temperature Reactor (PB-FHR). The experiments are accompanied by the development of a new discrete element simulation code, GRECO, which is designed to offer a simple user interface and simplified two-dimensional system that can be used for iterative purposes in the preliminary phases of core design. The results of this study are focused on the PB-FHR, but can easily be extended for gas-cooled reactor designs.
Experimental results are presented for three Pebble Recirculation Experiments (PREX). PREX 2 and 3.0 are conventional gravity-dominated granular systems based on the annular PB-FHR core design for a 900 MWth commercial prototype plant and a 16 MWth test reactor, respectively. Detailed results are presented for the pebble velocity field, mixing at the radial zone interfaces, and pebble residence times. A new Monte Carlo algorithm was developed to study the residence time distributions of pebbles in different radial zones. These dry experiments demonstrated the basic viability of radial pebble zoning in cores with diverging geometry before pebbles reach the active core.
Results are also presented from PREX 3.1, a scaled facility that uses simulant materials to evaluate the impact of coupled fluid drag forces on the granular dynamics in the PB-FHR core. PREX 3.1 was used to collect first of a kind pebble motion data in a multidimensional porous media flow field. Pebble motion data were collected for a range of axial and cross fluid flow configurations where the drag forces range from half the buoyancy force up to ten times greater than the buoyancy force. Detailed analysis is presented for the pebble velocity field, mixing behavior, and residence time distributions for each fluid flow configuration.
The axial flow configurations in PREX 3.1 showed small changes in pebble motion compared to a reference case with no fluid flow and showed similar overall behavior to PREX 3.0. This suggests that dry experiments can be used for core designs with uniform one-dimensional coolant flow early in the design process at greatly reduced cost. Significant differences in pebble residence times were observed in the cross fluid flow configurations, but these were not accompanied by an overall horizontal diffusion bias. Radial zones showed only a small shift in position due to mixing in the diverging region and remained stable in the active core. The results from this study support the overall viability of the annular PB-FHR core by demonstrating consistent granular flow behavior in the presence of complex reflector geometries and multidimensional fluid flow fields.
GRECO simulations were performed for each of the experiments in this study in order to develop a preliminary validation basis and to understand for which applications the code can provide useful analysis. Overall, the GRECO simulation results showed excellent agreement with the gravity-dominated PREX experiments. Local velocity errors were found to be generally within 10-15% of the experimental data. Average radial zone interface positions were predicted within two pebble diameters. GRECO simulations over predicted the amount of mixing around the average radial zone interface position and therefore can be treated as a conservative upper bound when used in neutronics analysis. Residence time distributions from the GRECO velocity data based on the Monte Carlo algorithm closely matched those derived from the experiment velocity statistics. GRECO simulation results for PREX 3.1 with coupled drag forces showed larger errors compared to the experimental data, particularly in the cases with cross fluid flow. The large discrepancies suggest that GRECO results in systems with coupled fluid drag forces cannot be used with high confidence at this point and future development work on coupled pebble and fluid dynamics with multidimensional fluid flow fields is required.
Jahn, Gordon James. "Agent-based structural condition monitoring for nuclear reactor cores." Thesis, University of Strathclyde, 2011. http://oleg.lib.strath.ac.uk:80/R/?func=dbin-jump-full&object_id=17400.
Full textShuffler, Carter Alexander. "Optimization of hydride fueled pressurized water reactor cores." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/33634.
Full textIncludes bibliographical references (leaf 173).
This thesis contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). This pursuit involves implementing an appropriate methodology for design and optimization of hydride and oxide fueled cores. Core design is accomplished for a range of geometries via steady-state and transient thermal hydraulic analyses, which yield the maximum power, and fuel performance and neutronics studies, which provide the achievable discharge burnup. The final optimization integrates the outputs from these separate studies into an economics model to identify geometries offering the lowest cost of electricity, and provide a fair basis for comparing the performance of hydride and oxide fuels. Considerable work has already been accomplished on the project; this thesis builds on this previous work. More specifically, it focuses on the steady-state thermal hydraulic and economic analyses for pressurized water reactor (PWR) cores utilizing UZrH₁.₆ and UO₂. A previous MIT study established the steady-state thermal hydraulic design methodology for determining maximum power from square array PWR core designs.
(cont.) The analysis was not performed for hexagonal arrays under the assumption that the maximum achievable powers for both configurations are the same for matching rod diameters and H/HM ratios. This assumption is examined and verified in this work by comparing the thermal hydraulic performance of a single hexagonal core with its equivalent square counterpart. In lieu of a detailed vibrations analysis, the steady-state thermal hydraulic analysis imposed a single design limit on the axial flow velocity. The wide range of core geometries considered and the large power increases reported by the study makes it prudent to refine this single limit approach. This work accomplishes this by developing and incorporating additional design limits into the thermal hydraulic analysis to prevent excessive rod vibration and wear. The vibrations and wear mechanisms considered are: vortex-induced vibration, fluid-elastic instability, turbulence-induced vibration, fretting wear, and sliding wear. Concomitantly with this work, students at UC Berkeley and MIT have undertaken the neutronics, fuel performance, and transient thermal hydraulic studies.
(cont.) With these results, and the output from the steady-state thermal hydraulic analysis with vibrations and wear imposed design limits, an economics model is employed to determine the optimal geometries for incorporation into existing PWRs. The model also provides a basis for comparing the performance of UZrH₁.₆ to UO₂ for a range of core geometries. Though this analysis focuses only on these fuels, the methodology can easily be extended to additional hydride and oxide fuel types, and will be in the future. Results presented herein do not show significant cost savings for UZrH₁.₆, primarily because the power and energy generation per core loading for both fuels are similar. Furthermore, the most economic geometries typically do not occur where power increases are reported by the thermal hydraulics. As a final note, the economic results in this report require revision to account for recent changes in the fuel performance analysis methodology. The changes, however, are not expected to influence the overall conclusion that UZrH₁.₆ does not outperform UO₂ economically.
by Carter Alexander Shuffler.
S.M.
Trant, Jarrod Michael. "Transient analysis of hydride fueled pressurized water reactor cores." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/33632.
Full textIncludes bibliographical references (leaves 132-133).
This thesis contributes to the hydride nuclear fuel project led by U. C. Berkeley for which MIT is to perform the thermal hydraulic and economic analyses. A parametric study has been performed to determine the optimum combination of lattice pitch, rod diameter, and channel shape-further referred to as geometry-for maximizing power given specific transient conditions for pressurized water reactors (PWR) loaded with either U02 or UZrH1.6 fuel. Several geometries have been examined with the VIPRE subchannel analysis tool along with MATLAB scripts previously developed to automate VIPRE execution. The transients investigated were a large break loss of coolant accident (LBLOCA), am overpower transient, and a complete loss of flow accident. The maximum achievable power for each geometry is defined as the highest power that can be sustained without exceeding any of the steady state or transient limits. The limits were chosen based on technical feasibility and safety of the reference core and compared with the final safely analysis report (FSAR) of the reference core, the South Texas Project Electric Generating Station (STPEGS), whenever possible. This analysis was performed for two separate pressure drop limits of 29 and 60 psia for both a square array with grid spacers and a hexagonal array with wire wraps.
(cont.) The square core geometry sustaining the highest power (4820.0 MW) for both the hydride and oxide fueled has a pitch of 9.0 mm and a rod diameter of 6.5 mm and was limited by the complete loss of flow accident. Both of these maximum power geometries occurred at the 60 psia pressure drop case. The maximum power of the 29 psia pressure drop case (4103.9 MW) for both fuel types occurred at a pitch of 9.7 mm and a rod diameter of 6.5 mm. The maximum power for the hexagonal arrayed cores occurred at the same hydrogen to heavy metal ratio as the square cores. The hydride fueled core power (5123.2 MW) was limited by the overpower transient while the oxide fueled core power (4996.1 MW) was limited by the overpower transient. The pressure drop constraint was not limiting for either fuel type for either pressure drop case for the wire wrapped cores.
by Jarrod Michael Trant.
S.M.
Alam, Syed Bahauddin. "The design of reactor cores for civil nuclear marine propulsion." Thesis, University of Cambridge, 2018. https://www.repository.cam.ac.uk/handle/1810/275650.
Full textPINTO, LETICIA N. "Experimentos de efeitos de reatividade no reator nuclear IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2012. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10099.
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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
SANTOS, DIOGO F. dos. "Caracterização dos campos neutrônicos obtidos por meio de armadilhas de nêutrons a partir da utilização de água pesada (D2O) no interior do núcleo do reator nuclear IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/23825.
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Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
Books on the topic "Nuclear reaction codes"
Neighbour, Gareth B. Securing the safe performance of graphite reactor cores. Cambridge, UK: RSC Pub., 2010.
Find full textBoer, Brian. Optimized core design and fuel management of a pebble-bed type nuclear reactor. Amsterdam: IOS Press, 2008.
Find full textRoss, Kyle. MELCOR best practices as applied in the State-of-the-Art Reactor Consequence Analyses (SOARCA) Project. Washington, DC: U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, 2014.
Find full textTurnbull, J. Anthony. Review of nuclear fuel experimental data: Fuel behaviour data available from IFE-OCDE Halden Project for development and validation of computer codes. Paris: Nuclear Energy Agency, Organisation for Economic Co-operation and Development, 1995.
Find full textBilanovic, Z. Neutron-photon energy deposition in CANDU reactor fuel channels: A comparison of modelling techniques using ANISN and MCNP computer codes. Chalk River, Ont: System Chemistry and Corrosion Branch, Chalk River Laboratories, 1994.
Find full textCentre, Bhabha Atomic Research, ed. Operational reactor physics analysis codes (ORPAC). Mumbai: Bhabha Atomic Research Centre, 2007.
Find full textM, Blann, and OECD Nuclear Energy Agency, eds. International code comparison for intermediate energy nuclear data = Comparaison internationale de codes pour le calcul de données nucléaires aux énergies intermédiaires. Paris: Nuclear Energy Agency, Organisation for Economic Co-operation and Development, 1994.
Find full textManagement of ageing in graphite reactor cores. Cambridge: RSC Publishing, 2007.
Find full textB, Murfin W., Johnson Jay D, U.S. Nuclear Regulatory Commission. Division of Safety Issue Resolution., Sandia National Laboratories, Technadyne Engineering Consultants, and Science Applications International Corporation, eds. XSOR codes users manual. Washington, DC: Division of Safety Issue Resolution, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.
Find full textMa, Chang Chun. Study of interfacial condensation in a nuclear reactor core makeup tank. 1993.
Find full textBook chapters on the topic "Nuclear reaction codes"
Troicki, Filip T., Filip T. Troicki, Filip T. Troicki, Carlos A. Perez, Wade L. Thorstad, Brandon J. Fisher, Larry C. Daugherty, et al. "Nuclear Reactor Cores." In Encyclopedia of Radiation Oncology, 564. Berlin, Heidelberg: Springer Berlin Heidelberg, 2013. http://dx.doi.org/10.1007/978-3-540-85516-3_722.
Full textKawano, Toshihiko. "CoH3: The Coupled-Channels and Hauser-Feshbach Code." In Compound-Nuclear Reactions, 27–34. Cham: Springer International Publishing, 2020. http://dx.doi.org/10.1007/978-3-030-58082-7_3.
Full textYamanaka, Masao. "Sensitivity and Uncertainty of Criticality." In Accelerator-Driven System at Kyoto University Critical Assembly, 215–43. Singapore: Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-0344-0_8.
Full textOesterle, Ralph G., W. Gene Corley, and Ahmed Elremaily. "History of Shear Design Provisions in the ASME/ACI Code for Concrete Reactor Vessels and Containments." In Infrastructure Systems for Nuclear Energy, 287–305. Chichester, UK: John Wiley & Sons, Ltd, 2013. http://dx.doi.org/10.1002/9781118536254.ch18.
Full textRubchenya, V. A. "New model and code for calculation of product yields in fusion-fission reactions." In Exotic Nuclei and Atomic Masses, 381. Berlin, Heidelberg: Springer Berlin Heidelberg, 2003. http://dx.doi.org/10.1007/978-3-642-55560-2_145.
Full textXu, Junying, Lei Zhang, Dekui Zhan, Huiyong Zhang, Yahelle Laroche, Hui Guo, and Guillaume Niessen. "Study of Potential for In-Vessel Retention Through External Reactor Vessel Flooding: Code Comparison." In Proceedings of The 20th Pacific Basin Nuclear Conference, 601–15. Singapore: Springer Singapore, 2017. http://dx.doi.org/10.1007/978-981-10-2311-8_56.
Full textShi, Chengbin, Maosong Cheng, and Guimin Liu. "Development and Verification of Liquid-Fueled Molten Salt Reactor Analysis Code Based on RELAP5." In Proceedings of The 20th Pacific Basin Nuclear Conference, 731–39. Singapore: Springer Singapore, 2017. http://dx.doi.org/10.1007/978-981-10-2317-0_69.
Full textKalugin, M. A. "Validation of the MCU-RFFI/A Code for Applications to Plutonium Systems and Use of the MCU-RFFI/A Code for Verification of Physics Design Codes Intended for Calculations of Vver Reactor Performance With Mox Fuel." In Safety Issues Associated with Plutonium Involvement in the Nuclear Fuel Cycle, 147–58. Dordrecht: Springer Netherlands, 1999. http://dx.doi.org/10.1007/978-94-011-4591-6_18.
Full textRodríguez-Hernandez, Andrés, Armando M. Gómez-Torres, Edmundo del Valle-Gallegos, Javier Jimenez-Escalante, Nico Trost, and Victor H. Sanchez-Espinoza. "Accelerating AZKIND Simulations of Light Water Nuclear Reactor Cores Using PARALUTION on GPU." In Communications in Computer and Information Science, 419–31. Cham: Springer International Publishing, 2016. http://dx.doi.org/10.1007/978-3-319-32243-8_29.
Full textPapukchiev, Angel, Peter Pandazis, Hristo Hristov, and Martina Scheuerer. "Validation of Coupled CFD-CSM Methods for Vibration Phenomena in Nuclear Reactor Cores." In Notes on Numerical Fluid Mechanics and Multidisciplinary Design, 55–69. Cham: Springer International Publishing, 2021. http://dx.doi.org/10.1007/978-3-030-55594-8_7.
Full textConference papers on the topic "Nuclear reaction codes"
Kakavand, Tayeb, Morteza Taghilo, and Mahdi Sadeghi. "Determination of 89Zr Production Parameters via Different Reactions Using ALICE and TALYS Codes." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-30298.
Full textKakavand, Tayeb, and Morteza Taghilo. "Calculations of Excitation Functions to Produce 88Y via Various Nuclear Reactions by ALICE/91 and TALYS-1.0 Codes." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-30328.
Full textNoorikalkhoran, Omid, and Massimiliano Gei. "Simulation of Hydrogen Distribution due to In-Vessel Severe Accident in WWER-1000 NPP Containment: A Comparison of CONTAIN and MELCOR Codes Results." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-82635.
Full textWang, Te-Chuan. "Comparison of Severe Accident Results by Using MAAP5 and MAAP4 Codes." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29017.
Full textUchibori, Akihiro, Shin Kikuchi, Akikazu Kurihara, Hirotsugu Hamada, and Hiroyuki Ohshima. "Multiphysics Analysis System for Tube Failure Accident in Steam Generator of Sodium-Cooled Fast Reactor." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16692.
Full textKakavand, T., K. Kamali Moghaddam, M. Sadeghi, and R. Ghasemi. "Design of Tellurium-123 Target for Producing Iodine-123 Radioisotope Using Computer Simulation Techniques." In 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89667.
Full textHorie, Hideki, Yutaka Takeuchi, Kenya Takiwaki, Fumie Sebe, Kazuo Kakiuchi, and Hisaki Sato. "Severe Accident Analysis for Reactor Core Applying SiC to Fuel Claddings and Channel Boxes." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81923.
Full textGosmain, Cécile-Aline, Sylvain Rollet, and Damien Schmitt. "3D Calculations of PWR Vessels Neutron Fluence With EFLUVE 3D Code." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16316.
Full textVechgama, Wasin, and Kampanart Silva. "Study of Fission Product Behavior in Containment Vessel Using Modified ART Mod 2: Update of Cesium and Iodine Compound Models." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-82069.
Full textNasrabadi, M. N., and M. Sepiani. "Study of components and statistical reaction mechanism in simulation of nuclear process for optimized production of 64Cu and 67Ga medical radioisotopes using TALYS, EMPIRE and LISE++ nuclear reaction and evaporation codes." In 4TH INTERNATIONAL CONGRESS IN ADVANCES IN APPLIED PHYSICS AND MATERIALS SCIENCE (APMAS 2014). AIP Publishing LLC, 2015. http://dx.doi.org/10.1063/1.4914267.
Full textReports on the topic "Nuclear reaction codes"
Koi, Tatsumi. Interfacing the JQMD and JAM Nuclear Reaction Codes to Geant4. Office of Scientific and Technical Information (OSTI), June 2003. http://dx.doi.org/10.2172/813352.
Full textPARMA, JR, EDWARD J. BURNCAL: A Nuclear Reactor Burnup Code Using MCNP Tallies. Office of Scientific and Technical Information (OSTI), November 2002. http://dx.doi.org/10.2172/805880.
Full textFrancis, Matthew W., Charles F. Weber, Marco T. Pigni, and Ian C. Gauld. Reactor Fuel Isotopics and Code Validation for Nuclear Applications. Office of Scientific and Technical Information (OSTI), February 2015. http://dx.doi.org/10.2172/1185693.
Full textMurata, K. K., D. C. Williams, R. O. Griffith, R. G. Gido, E. L. Tadios, F. J. Davis, G. M. Martinez, K. E. Washington, and J. Tills. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis. Office of Scientific and Technical Information (OSTI), December 1997. http://dx.doi.org/10.2172/569132.
Full textLittle, W. W. Jr. 1DB, a one-dimensional diffusion code for nuclear reactor analysis. Office of Scientific and Technical Information (OSTI), September 1991. http://dx.doi.org/10.2172/6366280.
Full textClarno, Kevin, Alfred Abraham Lorber, Richard J. Pryor, William F. Spotz, Rodney Cannon Schmidt, Kenneth Belcourt, Russell Warren Hooper, and Larry LaRon Humphries. Foundational development of an advanced nuclear reactor integrated safety code. Office of Scientific and Technical Information (OSTI), February 2010. http://dx.doi.org/10.2172/973349.
Full textOrmand, W., and K. Kravvaris. YAHFC: A Code Framework to Model Nuclear Reactions and Estimate Correlated Uncertainties. Office of Scientific and Technical Information (OSTI), April 2021. http://dx.doi.org/10.2172/1778648.
Full textOrmand, W. Monte Carlo Hauser-Feshbach computer code system to model nuclear reactions: YAHFC. Office of Scientific and Technical Information (OSTI), July 2021. http://dx.doi.org/10.2172/1808762.
Full textMcCollam, K. Analysis of Fe(n,x[gamma]) cross sections using the TNG nuclear reaction model code. Office of Scientific and Technical Information (OSTI), April 1993. http://dx.doi.org/10.2172/6549441.
Full textM. J. Russell. Assessement of Codes and Standards Applicable to a Hydrogen Production Plant Coupled to a Nuclear Reactor. Office of Scientific and Technical Information (OSTI), June 2006. http://dx.doi.org/10.2172/911554.
Full text