Thematische Bibliographien / Nuclear reaction codes

Inhaltsverzeichnis

Zeitschriftenartikel
Dissertationen
Bücher
Buchteile
Konferenzberichte
Berichte der Organisationen

Auswahl der wissenschaftlichen Literatur zum Thema „Nuclear reaction codes“

Autor: Grafiati

Veröffentlicht am 23. Juni 2021

Zuletzt aktualisiert am 1. Februar 2022

Geben Sie eine Quelle nach APA, MLA, Chicago, Harvard und anderen Zitierweisen an

Wählen Sie eine Art der Quelle aus:

Machen Sie sich mit den Listen der aktuellen Artikel, Bücher, Dissertationen, Berichten und anderer wissenschaftlichen Quellen zum Thema "Nuclear reaction codes" bekannt.

Neben jedem Werk im Literaturverzeichnis ist die Option "Zur Bibliographie hinzufügen" verfügbar. Nutzen Sie sie, wird Ihre bibliographische Angabe des gewählten Werkes nach der nötigen Zitierweise (APA, MLA, Harvard, Chicago, Vancouver usw.) automatisch gestaltet.

Sie können auch den vollen Text der wissenschaftlichen Publikation im PDF-Format herunterladen und eine Online-Annotation der Arbeit lesen, wenn die relevanten Parameter in den Metadaten verfügbar sind.

Zeitschriftenartikel zum Thema "Nuclear reaction codes"

Kataria, S. K., V. S. Ramamurthy, M. Blann und 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, Nr. 2-3 (März 1990): 585–88. http://dx.doi.org/10.1016/0168-9002(90)90155-y.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Denikin, Andrey, Alexander Karpov, Mikhail Naumenko, Vladimir Rachkov, Viacheslav Samarin und 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.

Der volle Inhalt der Quelle

Annotation:

The paper describes the NRV web knowledge base on low-energy nuclear physics developed in the Joint Institute for Nuclear Research. The NRV knowledge base working through the Internet integrates a large amount of digitized experimental data on the properties of nuclei and nuclear reaction cross sections with a wide range of computational programs for modeling of nuclear properties and nuclear dynamics. Today, the NRV becomes a powerful instrument for nuclear physics research as well as for educational applications. Advantages of the functioning scheme of the knowledge base provide the synergy of coexistence of the experimental data and computational codes within one platform.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Özdoğan, H., İsmail Hakki Sarpün, Mert Şekerci und Abdullah Kaplan. „Production cross-section calculations of 111In via proton and alpha-induced nuclear reactions“. Modern Physics Letters A 36, Nr. 08 (18.02.2021): 2150051. http://dx.doi.org/10.1142/s0217732321500516.

Der volle Inhalt der Quelle

Annotation:

[Formula: see text], a known gamma emitter, is used for many medical purposes such as imaging of myocardial metastases. It can be produced by using different nuclear reactions. In this study, the reactions of [Formula: see text]Ag([Formula: see text]2n)[Formula: see text], [Formula: see text](p,[Formula: see text]n)[Formula: see text], [Formula: see text](p,[Formula: see text]2n)[Formula: see text], [Formula: see text](p,[Formula: see text]3n)[Formula: see text] and [Formula: see text](p,[Formula: see text]4n)[Formula: see text], which are the production routes of [Formula: see text], were investigated. Production cross-section calculations were performed by using equilibrium and pre-equilibrium models of TALYS 1.95 and EMPIRE 3.2 nuclear reaction codes. Hauser–Feshbach Model was appointed in both codes for calculations of equilibrium approximations. Exciton and Hybrid Monte Carlo Simulation (HMS) models were used in the EMPIRE 3.2, whereas Two-Component Exciton and Geometry Dependent Hybrid Model, which is implemented to TALYS code, has been used in the TALYS 1.95 for pre-equilibrium reactions. Also, a weighting matrix of the nuclear models was obtained by using statistical variance analysis. The optimum beam energy to obtain [Formula: see text] has been determined by using the results obtained from this weighting matrix.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Hilaire, Stephane, Eric Bauge, Pierre Chau Huu-Tai, Marc Dupuis, Sophie Péru, Olivier Roig, Pascal Romain und 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.

Der volle Inhalt der Quelle

Annotation:

Nowadays, reliance on nuclear models to interpolate or extrapolate between experimental data points is very common, for nuclear data evaluation. It is also well known that the knowledge of nuclear reaction mechanisms is at best approximate, and that their modeling relies on many parameters which do not have a precise physical meaning outside of their specific implementations in nuclear model codes: they carry both specific physical information, and effective information that is related to the deficiencies of the model itself. Therefore, to improve the uncertainties associated with evaluated nuclear data, the models themselves must be refined so that their parameters can be rigorously derived from theory. Examples of such a process will be given for a wide sample of models like: detailed theory of compound nucleus decay through multiple nucleon or gamma emission, or refinements to the width fluctuation factor of the Hauser-Feshbach model. All these examples will illustrate the reduction in the effective components of nuclear model parameters, through the reduced dynamics of parameter adjustment needed to account for experimental data. The significant progress, recently achieved for the non-fission channels, also highlights the difficult path ahead to improve our quantitative understanding of fission in a similar way: by relying on microscopic theory.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Sarpün, İsmail Hakki, Hasan Özdoğan, Kemal Taşdöven, Hüseyin Ali Yalim und Abdullah Kaplan. „Theoretical photoneutron cross-section calculations on Osmium isotopes by Talys and Empire codes“. Modern Physics Letters A 34, Nr. 26 (30.08.2019): 1950210. http://dx.doi.org/10.1142/s0217732319502109.

Der volle Inhalt der Quelle

Annotation:

In this study, the level density parameter and the gamma ray strength function effects on photoneutron reaction cross-section calculations for Osmium isotopes were investigated by employing available level density models and gamma ray strength functions within Talys v1.8 and Empire v3.1 nuclear codes. A relative variance analysis was done to determine the best gamma ray strength function. Then, the effect of level density models for the photoneutron reactions was investigated by using the best gamma ray strength function. The results were compared with each other and also with the experimental data taken from the literature.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Şekerci, Mert, Hasan Özdoğan und 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, Nr. 24 (23.06.2020): 2050202. http://dx.doi.org/10.1142/s0217732320502028.

Der volle Inhalt der Quelle

Annotation:

Level density models have an undeniable importance for a better perception on the nature of nuclear reactions, which influences our life via various ways. Many novel and advanced medical application use radioisotopes, which are produced with nuclear reactions. By considering the connection between the level density models and the importance of theoretical calculations for the production routes of medically important isotopes, this study is performed to investigate the level density model effects on the production cross-section calculations of [Formula: see text]Zn, [Formula: see text]Ga, [Formula: see text]Kr, [Formula: see text]Pd, [Formula: see text]In, [Formula: see text]I and [Formula: see text]At radioisotopes via some alpha particle induced and neutron emitting reactions. For theoretical calculations; frequently used computation tools, such as TALYS and EMPIRE codes, are applied. Obtained theoretical results are then compared with the experimental data, taken from Experimental Nuclear Reaction Data (EXFOR) library. For a better interpretation of the results, a mean weighted deviation calculation for each investigated reaction is performed in addition to a visual comparison of the graphical representations of the outcomes.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Henning, 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.

Der volle Inhalt der Quelle

Annotation:

Today’s development of nuclear installations rely on numerical simulation for which the main input are evaluated nuclear data. Inelastic neutron scattering (n, xn) is a reaction of importance because it modifies the neutron population, the neutron energy distribution and may create new isotopes. The study of this reaction on tungsten isotopes is interesting because it is a common structural material. Additionally, tungsten isotopes are a good testing field for theories. The IPHC group started an experimental program with the GRAPhEME setup installed at the neutron beam facility GELINA to measure (n, xn γ) reaction cross sections using prompt gamma spectroscopy and neutron energy determination by time-of-flight. The obtained experimental data provide constraints on nuclear reaction mechanisms models for 182,184,186W. Indeed, to reproduce correctly the experimental (n, n’ γ) cross-sections, the reaction codes must include accurate models of the reaction mechanism, nuclear de-excitation process and use correct nuclear structure information.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Sabra, M. S., Robert A. Weller, Marcus H. Mendenhall, Robert A. Reed, Michael A. Clemens und A. F. Barghouty. „Validation of Nuclear Reaction Codes for Proton-Induced Radiation Effects: The Case for CEM03“. IEEE Transactions on Nuclear Science 58, Nr. 6 (Dezember 2011): 3134–38. http://dx.doi.org/10.1109/tns.2011.2169989.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Korbut, Tamara, Maksim Kravchenko, Ivan Edchik und 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.

Der volle Inhalt der Quelle

Annotation:

Present work describes Monte-Carlo calculations of the neutron field and minor actinide transmutation reaction rates within the Yalina-Thermal sub-critical assembly of the Joint Institute for Power and Nuclear Research – Sosny of the National Academy of Sciences of Belarus. The computer model of the facility was prepared for the corresponding calculations via MCU-PD and MCNP Monte-Carlo codes. The model neutron characteristics estimations were performed as well as the nuclear safety analysis. The up-to-date ENDF B/VIII, JEFF 3.3 and JENDL 4.0 nuclear data libraries were used during research.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Voinov, A. V., S. M. Grimes, C. R. Brune, A. Bürger, A. Görgen, M. Guttormsen, A. C. Larsen, T. N. Massey und S. Siem. „Level Density Inputs in Nuclear Reaction Codes and the Role of the Spin Cutoff Parameter“. Nuclear Data Sheets 119 (Mai 2014): 255–57. http://dx.doi.org/10.1016/j.nds.2014.08.070.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Mehr Quellen

Dissertationen zum Thema "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.

Der volle Inhalt der Quelle

Annotation:

Made available in DSpace on 2014-10-09T12:32:47Z (GMT). No. of bitstreams: 0
Made available in DSpace on 2014-10-09T14:08:32Z (GMT). No. of bitstreams: 1 03473.pdf: 1345873 bytes, checksum: c84cb3f38a6cd9b74002f62ae4817f88 (MD5)
Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

APA, Harvard, Vancouver, ISO und andere Zitierweisen

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.

Der volle Inhalt der Quelle

Annotation:

Made available in DSpace on 2014-10-09T12:43:12Z (GMT). No. of bitstreams: 0
Made available in DSpace on 2014-10-09T13:56:31Z (GMT). No. of bitstreams: 1 06442.pdf: 6689802 bytes, checksum: 46ba1e2b896e77a4288d24e80ac804ff (MD5)
Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

APA, Harvard, Vancouver, ISO und andere Zitierweisen

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.

Der volle Inhalt der Quelle

Annotation:

Made available in DSpace on 2014-10-09T12:48:53Z (GMT). No. of bitstreams: 0
Made available in DSpace on 2014-10-09T14:01:14Z (GMT). No. of bitstreams: 1 09627.pdf: 6224254 bytes, checksum: 0192b8abd2aed7811607e803516e20a7 (MD5)
Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Laufer, Michael Robert. „Granular Dynamics in Pebble Bed Reactor Cores“. Thesis, University of California, Berkeley, 2013. http://pqdtopen.proquest.com/#viewpdf?dispub=3593891.

Der volle Inhalt der Quelle

Annotation:

This 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.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

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.

Der volle Inhalt der Quelle

Annotation:

A significant proportion of the UK energy needs are currently serviced by a fleet of ageing nuclear reactors. Ensuring that these reactors are operated safely is the highest priority and the structural health of their cores, that provide channels for control rods and coolant gas, is a key aspect. This thesis focuses on the application of structuralhealth monitoring to the graphite reactor cores used in the UK and presents a specification for the use of structural health monitoring (SHM) techniques already es- tablished in bridge and aircraft monitoring, with data obtained through existing reactor monitoring processes. This approach utilises statistical and clustering techniques on monitoring data that can be acquired during online operation of the plant. The use of existing monitoring processes to complement the established inspection regime for nuclear reactors is a novel contribution from this work. As part of proving the SHM approach, this thesis reports on work undertaken to identify suitable data and numerical limits for the cluster analysis. This analysis considers the data with respect to the stated aim of detectin~ core distortion and demonstrates that the chosen data and values are acceptable and conservative in the context of reactor condition monitoring. An assessment of the SHM solution is presented describing the im- plementation of the SHM approach using a multi-agent system (MAS), IMAPS. This implementation required consideration of using MAS tech- nology for condition monitoring, and the novel contribution of a technique for storing and retrieving historical data in a manner concomitant with both MAS and relational database theory is presented.ij The thesis concludes that condition monitoring is feasible on the graphite cores, and that multi-variate analysis through SHM implemented within a MAS offers a storage and analysis platform that can both handle the data volumes and accommodate further extensions as required.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Shuffler, Carter Alexander. „Optimization of hydride fueled pressurized water reactor cores“. Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/33634.

Der volle Inhalt der Quelle

Annotation:

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.
Includes 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.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

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.

Der volle Inhalt der Quelle

Annotation:

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.
Includes 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.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

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.

Der volle Inhalt der Quelle

Annotation:

Perhaps surprisingly, the largest experience in operating nuclear power plants has been in nuclear naval propulsion, particularly submarines. This accumulated experience may become the basis of a proposed new generation of compact nuclear power plant designs. In an effort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. Reactor cores for such an application would need to be fundamentally different from land-based power generation systems, which require regular refueling, and from reactors used in military submarines, as the fuel used could not conceivably be as highly enriched. Nuclear-powered propulsion would allow ships to operate with low fuel costs, long refueling intervals, and minimal emissions; however, currently such systems remain largely confined to military vessels. This research project undertakes computational modeling of possible soluble-boron-free (SBF) reactor core designs for this application, with a view to informing design decisions in terms of choices of fuel composition, materials, core geometry and layout. Computational modeling using appropriate reactor physics (e.g. WIMS, MONK, Serpent and PANTHER), thermal-hydraulics etc. codes (e.g. COBRA-EN) is used for this project. With an emphasis on reactor physics, this study investigates possible fuel assembly and core designs for civil marine propulsion applications. In particular, it explores the feasibility of using uranium/thorium-rich fuel in a compact, long-life reactor and seek optimal choices and designs of the fuel composition, reactivity control, assembly geometry, and core loading in order to meet the operational needs of a marine propulsion reactor. In this reactor physics and 3D coupled neutronics/thermal-hydraulics study, we attempt to design a civil marine reactor core that fulfills the objective of providing at least 15 effective full-power-years (EFPY) life at 333 MWth. In order to unleash the benefit of thorium in a long life core, the micro-heterogeneous ThO2-UO2 duplex fuel is well-positioned to be utilized in our proposed civil marine core. Unfortunately, A limited number of studies of duplex fuel are available in the public domain, but its use has never been examined in the context of a SBF environment for long-life small modular rector (SMR) core. Therefore, we assumed micro-heterogeneous ThO2-UO2 duplex fuel for our proposed marine core in order to explore its capability. For the proposed civil marine propulsion core design, this study uses 18% U-235 enriched micro-heterogeneous ThO2-UO2 duplex fuel. To provide a basis for comparison we also evaluate the performance of homogeneously mixed 15% U-235 enriched all-UO2 fuel. This research also attempts to design a high power density core with 14 EFPY while satisfying the neutronic and thermal-hydraulics safety constraints. A core with an average power density of 100 MW/m3 has been successfully designed while obtaining a core life of 14 years. The average core power density for this core is increased by ∼50% compared to the reference core design (63 MW/m3 and is equivalent to Sizewell B PWR (101.6 MW/m3 which means capital costs could be significantly reduced and the economic attractiveness of the marine core commensurately improved. In addition, similar to the standard SMR core, a reference core with a power density of 63 MW/m3 has been successfully designed while obtaining a core life of ∼16 years. One of the most important points that can be drawn from these studies is that a duplex fuel lattice needs less burnable absorber than uranium-only fuel to achieve the same poison performance. The higher initial reactivity suppression and relatively smaller reactivity swing of the duplex can make the task of reactivity control through BP design in a thorium-rich core easier. It is also apparent that control rods have greater worth in a duplex core, reducing the control material requirements and thus potentially the cost of the rods. This research also analyzed the feasibility of using thorium-based duplex fuel in different cases and environments to observe whether this fuel consistently exhibit superior performance compared to the UO2 core in both the assembly and whole-core levels. The duplex fuel/core consistently exhibits superior performance in consideration of all the neutronic and TH constraints specified. It can therefore be concluded from this study that the superior performance of the thorium-based micro-heterogeneous ThO2-UO2 duplex fuel provides enhanced confidence that this fuel can be reliably used in high power density and long-life SBF marine propulsion core systems, offering neutronic advantages compared to the all-UO2 fuel. Last, but not least, considering all these factors, duplex fuel can potentially open the avenue for low-enriched uranium (LEU) SBF cores with different configurations. Motivated by growing environmental concerns and anticipated economic pressures, the overall goal of this study is to examine the technological feasibility of expanding the use of nuclear propulsion to civilian maritime shipping and to identify and propose promising candidate core designs.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

PINTO, 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.

Der volle Inhalt der Quelle

Annotation:

Made available in DSpace on 2014-10-09T12:34:46Z (GMT). No. of bitstreams: 0
Made available in DSpace on 2014-10-09T14:01:16Z (GMT). No. of bitstreams: 0
Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

APA, Harvard, Vancouver, ISO und andere Zitierweisen

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.

Der volle Inhalt der Quelle

Annotation:

Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2015-07-23T11:04:17Z No. of bitstreams: 0
Made available in DSpace on 2015-07-23T11:04:17Z (GMT). No. of bitstreams: 0
Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Mehr Quellen

Bücher zum Thema "Nuclear reaction codes"

Neighbour, Gareth B. Securing the safe performance of graphite reactor cores. Cambridge, UK: RSC Pub., 2010.

Den vollen Inhalt der Quelle finden

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Boer, Brian. Optimized core design and fuel management of a pebble-bed type nuclear reactor. Amsterdam: IOS Press, 2008.

Den vollen Inhalt der Quelle finden

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Ross, 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.

Den vollen Inhalt der Quelle finden

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Turnbull, 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.

Den vollen Inhalt der Quelle finden

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Bilanovic, 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.

Den vollen Inhalt der Quelle finden

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Centre, Bhabha Atomic Research, Hrsg. Operational reactor physics analysis codes (ORPAC). Mumbai: Bhabha Atomic Research Centre, 2007.

Den vollen Inhalt der Quelle finden

APA, Harvard, Vancouver, ISO und andere Zitierweisen

M, Blann, und OECD Nuclear Energy Agency, Hrsg. 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.

Den vollen Inhalt der Quelle finden

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Management of ageing in graphite reactor cores. Cambridge: RSC Publishing, 2007.

Den vollen Inhalt der Quelle finden

APA, Harvard, Vancouver, ISO und andere Zitierweisen

B, Murfin W., Johnson Jay D, U.S. Nuclear Regulatory Commission. Division of Safety Issue Resolution., Sandia National Laboratories, Technadyne Engineering Consultants und Science Applications International Corporation, Hrsg. XSOR codes users manual. Washington, DC: Division of Safety Issue Resolution, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.

Den vollen Inhalt der Quelle finden

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Ma, Chang Chun. Study of interfacial condensation in a nuclear reactor core makeup tank. 1993.

Den vollen Inhalt der Quelle finden

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Mehr Quellen

Buchteile zum Thema "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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Kawano, 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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Yamanaka, 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.

Der volle Inhalt der Quelle

Annotation:

AbstractExcess reactivity and control rod worth are generally considered important reactor physics parameters for experimentally examining the neutron characteristics of criticality in a core, and for maintaining safe operation of the reactor core in terms of neutron multiplication in the core. For excess reactivity and control rod worth at KUCA, as well as at the Fast Critical Assembly in the Japan Atomic Energy Agency, special attention is given to analyzing the uncertainty induced by nuclear data libraries based on experimental data of criticality in representative cores (EE1 and E3 cores). Also, the effect of decreasing uncertainty on the accuracy of criticality is discussed in this study. At KUCA, experimental results are accumulated by measurements of excess reactivity and control rod worth. To evaluate the accuracy of experiments for benchmarks, the uncertainty originated from modeling of the core configuration should be discussed in addition to uncertainty induced by nuclear data, since the uncertainty from modeling has a potential to cover the eigenvalue bias more than uncertainty by nuclear data. Here, to investigate the uncertainty of criticality depending on the neutron spectrum of cores, it is very useful to analyze the reactivity of a large number of measurements in typical hard (EE1) and soft (E3) spectrum cores at KUCA.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Oesterle, Ralph G., W. Gene Corley und 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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Rubchenya, 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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Xu, Junying, Lei Zhang, Dekui Zhan, Huiyong Zhang, Yahelle Laroche, Hui Guo und 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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Shi, Chengbin, Maosong Cheng und 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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Kalugin, 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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Rodríguez-Hernandez, Andrés, Armando M. Gómez-Torres, Edmundo del Valle-Gallegos, Javier Jimenez-Escalante, Nico Trost und 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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Papukchiev, Angel, Peter Pandazis, Hristo Hristov und 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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Konferenzberichte zum Thema "Nuclear reaction codes"

Kakavand, Tayeb, Morteza Taghilo und 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.

Der volle Inhalt der Quelle

Annotation:

The 89Zr radioisotope is used in the field of tumor diagnostics, tumor therapy and the investigation of the biokinetic. The present work is investigated a suitable reaction to produce 89Zr..The Zirconium-89 excitation function via 89Y(p,n)89Zr, 89Y(d,2n)89Zr, natZr(p,pxn)89Zr, natSr(α,xn)89Zr and 90Zr(n,2n)89Zr reactions were calculated by ALICE-91 and TALYS-1.0 codes and the reaction of 89Y(p,n)89Zr has been selected. The calculated excitation function of 89Y(p,n)89Zr reaction was compared with the reported measurement and evaluations. Requisite thickness of targets was obtained by SRIM code for all above reactions except the 90Zr(n,2n)89Zr reaction. The 89Zr production yield was evaluated with attention to excitation function and stopping power for all above reactions except 90Zr(n,2n)89Zr reaction.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Kakavand, Tayeb, und 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.

Der volle Inhalt der Quelle

Annotation:

Excitation functions were calculated by the ALICE/91 and TALYS-1.0 codes for natRb(a,xn)88Y, natZr(p,pxn)88Y, natSr(a,xn)88Y, 89Y(p,n)88Y and 88Sr(p,n)88Y reactions. The calculated cross sections were compared with the experimental data. The suitable energy ranges for the production of 88Y for each reaction is reported. From the excitation functions, integral yields of the products were calculated. Finally the suitable reaction was selected for the production of 88Y.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Noorikalkhoran, Omid, und 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.

Der volle Inhalt der Quelle

Annotation:

During a severe accident or Beyond Design Basis Accident (BDBA), the reaction of water with zirconium alloy as fuel clad, radiolysis of water, molten corium-concrete interaction (MCCI) and post-accident corrosion can generate a source of hydrogen. In the present work, hydrogen distribution due to in-vessel reaction (between zircaloy and steam) has been simulated inside a WWER-1000 reactor containment. In the first step, the thermal hydraulic parameters of containment have been simulated for a DECL (Double Ended Cold Leg) accident (DBA phase) in both short and long time and the effects of spray as Engineering Safety Features (ESFs) on mitigating the parameters have been studied. In the second step, it has been assumed that the accident developed into an in-vessel core melting accident. While in pre-phase of core melting (severe accident phase), hydrogen will be produced as a result of zircaloy and steam reaction (BDBA phase), the hydrogen distribution has been simulated for 23 cells inside the reactor containment by using CONTAIN 2.0 (Best estimate code) and MELCOR 1.8.6 codes. Finally, the results have been compared to FSAR results. As it can be seen from the comparisons, both CONTAIN and MELCOR codes can predict the results in good agreement with FSAR (ANGAR code) results. CONTAIN shows peak pressure around 0.36 MPa in short-term and this amount is about 0.38 and 0.4 MPa for MELCOR and ANGAR (FSAR) results respectively. All these values are under design pressure that is around 0.46 MPa. Cell 20 has the maximum mole fraction of hydrogen in long-term about 9.5% while the maximum amount of hydrogen takes place in cell 22. The differences between the results of codes are because of different equations, Models, Numerical methods and assumptions that have been considered by the codes. The simulated Hydrogen Distribution Map (HDM) can be used for upgrading the location of HCAV systems and Hydrogen Mitigator features (like the recombiners and ignitors) inside the containment to reduce the risk of hydrogen explosion.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Wang, 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.

Der volle Inhalt der Quelle

Annotation:

MAAP5 (Modular Accident Analysis Program Rev. 5.0.0), developed by Fauske & Associates, Inc.’s (FAI) based on the MAAP4 code, is a severe accident analysis code. It is a computer program capable of simulating the response and mitigation actions of light water reactor nuclear power plants (NPPs), including advanced boiling water reactor (ABWR) during severe accident. A specific loss of all core cooling accident sequence, LCLP-PF-R-N, based on Final Safety Analysis Report (FSAR) of Lungmen (ABWR) NPP, was selected as a based case and simulated by the MAAP5 and MAAP4 codes. The MAAP5 and MAAP4 parameter files for Lungmen NPP were established based on Lungmen NPP design data and the MAAP5 and MAAP4 users’ guides. The main severe accident phenomena and the fission product release fractions associated with the LCLP-PF-R-N sequence were simulated. The purpose of this paper is to compare the analysis results of LCLP-PF-R-N sequence calculated by MAAP5 and MAAP4 codes. The two codes give similar results for important phenomena during the accidents, including core uncovery, core support plate failure, debris relocation to the lower plenum, vessel failure, passive flooder opens, containment overpressure protection system (COPS) activation, noble gases and volatile species (like CsI) release to environment, except for the amount of hydrogen production in core. MAAP5 predicts a greater amount of hydrogen production in core than that of MAAP4. This is because MAAP4 predicts earlier reactor pressure vessel (RPV) depressurization than that of MAAP5. That results in earlier steam exhaustion and oxidation reaction termination in core than those of MAAP5. This paper successfully demonstrates the severe accident of Lungmen NPP, and analysis results can provide useful information for the MAAP5 and MAAP4 users.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Uchibori, Akihiro, Shin Kikuchi, Akikazu Kurihara, Hirotsugu Hamada und 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.

Der volle Inhalt der Quelle

Annotation:

Multiphysics analysis system was newly developed to evaluate possibility of failure propagation occurrence under heat transfer tube failure accident in a steam generator of sodium-cooled fast reactors. The analysis system consists of the computer codes, SERAPHIM, TACT, RELAP5, which are based on the mechanistic numerical models. The SERAPHIM code calculates the multicomponent multiphase flow involving sodium-water chemical reaction. In this study, numerical models for the chemical reaction about production of a sodium monoxide and its transport process were constructed to enable evaluation of a wastage environment. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The TACT code was integrated by the numerical models of the fluid-structure thermal coupling, the temperature and stress evaluation, the wastage evaluation and the failure judgment. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the rapidly heated tube in the present work.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Kakavand, T., K. Kamali Moghaddam, M. Sadeghi und 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.

Der volle Inhalt der Quelle

Annotation:

Iodine-123 is one of the most famous radioisotopes for Single Photon Emission Computed Tomography (SPECT) use, so, for 123I production, the 123Te has been chosen as a target through 123Te (p,n) 123I reaction. The various enriched targets (%99.9, %91, %85.4 and %70.1) have been used for the present calculations. In the current work, by using computer codes; ALICE & SRIM and doing a sort of calculations, we are going to demonstrate our latest effort for feasibility study of producing 123I by the above mentioned reaction. By using proton beam energy of less than 30 MeV, the mentioned codes give more accurate results. The cross section of all Tellurium reactions with proton has been calculated at 0–30 MeV proton beam energy with ALICE code. In the present work, the yield of 123I has been calculated by analytical methods. Our prediction for producing 123I yield via bombardment of 123Te (%99.9) with proton beam energy at 5–15 MeV is about 7.2 mCi/μAh. The present work shows that, the 123I yield is proportional to abundance of 123Te. Thermodynamical calculations with various current beams of up to 900 μA have been done, and the proper cooling system for the above purpose has been designed.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Horie, Hideki, Yutaka Takeuchi, Kenya Takiwaki, Fumie Sebe, Kazuo Kakiuchi und 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.

Der volle Inhalt der Quelle

Annotation:

Development of a fuel cladding or a channel box applying silicon carbide (SiC), which has high accident tolerance, in place of zircaloy (Zry) or Steel Use Stainless (SUS) composing current light water reactors, has being proceeded with after the accident of Fukushima Daiichi Nuclear Power Plant (1F). When applying SiC to core structures of a nuclear power plant such as fuel cladding, it is expected that the difference of high temperature oxidation characteristics in the severe accident (SA) conditions would mitigate progression of core damage comparing with the current Zry fuel core. This study performed SA analyses considering high temperature chemical reaction characteristics of SiC by using SA analysis code “MAAP”, and thermal hydraulics analysis code “TRAC Toshiba version (TRAC)”, and compared the difference between SiC and Zry. Both codes originally have no model of oxidation reaction for SiC. Hence, a new model for SiC in addition to the current model for Zry was incorporated into “MAAP”. On the other hand, “TRAC” adjusted reaction rate by changing oxidation reaction coefficients in the current Zry oxidation reaction models such as Baker-Just and Cathcart correlations in order to simulate SiC-water/steam reaction. In analysis using “MAAP”, seven accident sequences from representative Probabilistic Risk Assessment ones were selected to evaluate the difference of SA behavior between two materials. As a result, in the case of replacing current Zry of fuel claddings and channel boxes into SiC, an amount of hydrogen generation reduced to about 1/6 than the case of Zry. In addition to that, in the case of replacing SUS structures in the reactor core into SiC, an amount of hydrogen generation moreover reduced to about 1/6 than the above result, which means just about 2% of an amount in the original case. On the other hand, in analysis using “TRAC”, the accident sequence for unit 3 of 1F (1F3) was selected, and reaction rate in the oxidation reaction model was examined as parameter. In the case of 1.0 time of the reaction rate, which means an original reaction rate, maximum fuel cladding temperature exceeded 2000K in 50 hour after reactor scram. However, using the reaction rate below 0.01 to the original one, the fuel cladding temperature didn’t exceed 1,600K.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Gosmain, Cécile-Aline, Sylvain Rollet und 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.

Der volle Inhalt der Quelle

Annotation:

In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel. EDF R&D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte Carlo calculation with nearly the same accuracy. These EFLUVE3D calculations does the Source*Importance product which allows the calculation of the flux, the neutronic fluence (flux over 1MeV integrated on time) received at any point of the interface between the skin and the pressure vessel but also at the capsules of the pressurized water reactor vessels surveillance program and the dpa and reaction rates at different axial positions and different azimuthal positions of the vessel as well as at the surveillance capsules. Moreover, these calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rates calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source). The code needs as input: - for each reaction rate, the geometric importance matrix produced for a 3D pin by pin mesh on the basis of Green’s functions calculated by the Monte Carlo code TRIPOLI; - the neutron sources calculated on assemblies data (enrichment, position, fission fraction as a function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by use of the EDF core calculation scheme and a pin by pin power reconstruction methodology. This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI results. Theses comparisons show a discrepancy of less than 0,5% between the two codes equivalent to the order of magnitude of the stochastic convergence of Monte Carlo results.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Vechgama, Wasin, und 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.

Der volle Inhalt der Quelle

Annotation:

From the Fukushima accident, Thailand has studied fission product behavior in containment vessel using ART Mod 2 code. Cesium iodide (CsI), cesium hydroxide (CsOH) and iodine (I2) behaviors are studied using modified ART Mod 2 code. However, there are other compounds which are not included in the codes especially cesium and iodine compounds such as from Phébus FPT3 experiment including cesium molybdate (Cs2MoO4), cesium telluride (Cs2Te), methyl iodide (CH3I) and iodine pentoxide (I2O5). The paper objective is to add the four compounds in the codes in order to enlarge the coverage of the code in evaluation fission product behavior in the containment vessel. Physical parameters and models of the four compounds are updated in the codes. It is found that deposition phenomena of Cs2MoO4, Cs2Te CH3I and I2O5 are close to the experiment in case of no chemical reaction.

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Nasrabadi, M. N., und 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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Berichte der Organisationen zum Thema "Nuclear reaction codes"

Koi, Tatsumi. Interfacing the JQMD and JAM Nuclear Reaction Codes to Geant4. Office of Scientific and Technical Information (OSTI), Juni 2003. http://dx.doi.org/10.2172/813352.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

PARMA, 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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Francis, Matthew W., Charles F. Weber, Marco T. Pigni und Ian C. Gauld. Reactor Fuel Isotopics and Code Validation for Nuclear Applications. Office of Scientific and Technical Information (OSTI), Februar 2015. http://dx.doi.org/10.2172/1185693.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Murata, K. K., D. C. Williams, R. O. Griffith, R. G. Gido, E. L. Tadios, F. J. Davis, G. M. Martinez, K. E. Washington und J. Tills. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis. Office of Scientific and Technical Information (OSTI), Dezember 1997. http://dx.doi.org/10.2172/569132.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Little, 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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Clarno, Kevin, Alfred Abraham Lorber, Richard J. Pryor, William F. Spotz, Rodney Cannon Schmidt, Kenneth Belcourt, Russell Warren Hooper und Larry LaRon Humphries. Foundational development of an advanced nuclear reactor integrated safety code. Office of Scientific and Technical Information (OSTI), Februar 2010. http://dx.doi.org/10.2172/973349.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Ormand, W., und 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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Ormand, W. Monte Carlo Hauser-Feshbach computer code system to model nuclear reactions: YAHFC. Office of Scientific and Technical Information (OSTI), Juli 2021. http://dx.doi.org/10.2172/1808762.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

McCollam, 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.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

M. 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), Juni 2006. http://dx.doi.org/10.2172/911554.

Der volle Inhalt der Quelle

APA, Harvard, Vancouver, ISO und andere Zitierweisen

Wir bieten Rabatte auf alle Premium-Pläne für Autoren, deren Werke in thematische Literatursammlungen aufgenommen wurden. Kontaktieren Sie uns, um einen einzigartigen Promo-Code zu erhalten!