Relevant bibliographies by topics / Neutron kinetics

Contents

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

Academic literature on the topic 'Neutron kinetics'

Author: Grafiati
Published: 4 June 2021
Last updated: 3 February 2022

Create a spot-on reference in APA, MLA, Chicago, Harvard, and other styles

Select a source type:

Consult the lists of relevant articles, books, theses, conference reports, and other scholarly sources on the topic 'Neutron kinetics.'

Next to every source in the list of references, there is an 'Add to bibliography' button. Press on it, and we will generate automatically the bibliographic reference to the chosen work in the citation style you need: APA, MLA, Harvard, Chicago, Vancouver, etc.

You can also download the full text of the academic publication as pdf and read online its abstract whenever available in the metadata.

Journal articles on the topic "Neutron kinetics"

1

Khalimonchuk, V. A. "Mid-Weighed Kinetic Parameters for Use in the Two-Group Diffusion Model of Reactor Dynamics with Fuel Based on a Mixture of Fission Isotopes." Nuclear and Radiation Safety, no. 1(81) (March 12, 2019): 58–61. http://dx.doi.org/10.32918/nrs.2019.1(81).10.

Full text
Abstract:
In the model of reactor kinetics based on the description of neutron transport in the two-group diffusion approximation, the number of equations describing the change in the concentration of delayed neutron precursors depends not only on the number of groups of delayed neutrons, but also on the number of fissile isotopes present in nuclear fuel. Since each isotope is characterized by six groups of delayed neutrons, the total number of differential equations describing concentrations of delayed neutron precursors is equal to the product of the number of fissile isotopes (M) and the number of groups of delayed neutrons for each isotope (i = 6). This is true provided that the decay constant of the concentrations of delayed neutron precursors that were formed from the division by fast or thermal neutrons can be taken in the same way. In fact, there is a difference, though small, in these values for the two energy groups. Therefore, the number of the corresponding equations is twice as high. In this paper, a mathematical expression is obtained for the weighted average decay constant of delayed neutron predecessors from fission by fast and thermal neutrons in a multiplying medium with several fissile isotopes. This, together with the conventional procedure of weighing the fraction of delayed neutrons from fission by fast or thermal neutrons in a similar multiplying medium, allows the two-group diffusion model of the reactor kinetics to be limited to only six equations for the concentrations of delayed neutron precursors and thus the kinetic model of the reactor to be simplified.
APA, Harvard, Vancouver, ISO, and other styles
2

Kulikov, Gennady G., Anatoly N. Shmelev, Vladimir A. Apse, and Evgeny G. Kulikov. "On a significant slowing-down of the kinetics of fast transient processes in a fast reactor." Nuclear Energy and Technology 6, no. 4 (2020): 295–98. http://dx.doi.org/10.3897/nucet.6.60379.

Full text
Abstract:
The kinetics of nuclear reactors is determined by the average neutron lifetime. When the inserted reactivity is more than the effective delayed neutron fraction, the reactor kinetics becomes very rapid. It is possible to slow down the fast reactor kinetics by increasing the neutron lifetime. The authors consider the possibility of using the lead isotope, 208Pb, as a neutron reflector with specific properties in a lead-cooled fast reactor. To analyze the emerging effects in a reactor of this type, a point kinetics model was selected, which takes into account neutrons returning from the 208Pb reflector to the reactor core. Such specific properties of 208Pb as the high atomic weight and weak neutron absorption allow neutrons from the reactor core to penetrate deeply into the 208Pb reflector, slow down in it, and have a noticeable probability to return to the reactor core and affect the chain fission reaction. The neutrons coming back from the 208Pb reflector have a long ‘dead-time’, i.e., the sum of times when neutrons leave the reactor core, entering the 208Pb reflector, and then diffuse back into the reactor core. During the ‘dead-time’, these neutrons cannot affect the chain fission reaction. In terms of the delay time, the neutrons returning from the deep layers of the 208Pb reflector are close to the delayed neutrons. Moreover, the number of the neutrons coming back from the 208Pb reflector considerably exceeds the number of the delayed neutrons. As a result, the neutron lifetime formed by the prompt neutron lifetime and the ‘dead-time’ of the neutrons from the 208Pb reflector can be substantially increased. This will lead to a longer reactor acceleration period, which will mitigate the effects of prompt supercriticality. Thus, the use of 208Pb as a neutron reflector can significantly improve the fast reactor nuclear safety.
APA, Harvard, Vancouver, ISO, and other styles
3

Aboanber, Ahmed E. "Generalized and Stability Rational Functions for Dynamic Systems of Reactor Kinetics." International Journal of Nuclear Energy 2013 (August 13, 2013): 1–12. http://dx.doi.org/10.1155/2013/903904.

Full text
Abstract:
The base of reactor kinetics dynamic systems is a set of coupled stiff ordinary differential equations known as the point reactor kinetics equations. These equations which express the time dependence of the neutron density and the decay of the delayed neutron precursors within a reactor are first order nonlinear and essentially describe the change in neutron density within the reactor due to a change in reactivity. Outstanding the particular structure of the point kinetic matrix, a semianalytical inversion is performed and generalized for each elementary step resulting eventually in substantial time saving. Also, the factorization techniques based on using temporarily the complex plane with the analytical inversion is applied. The theory is of general validity and involves no approximations. In addition, the stability of rational function approximations is discussed and applied to the solution of the point kinetics equations of nuclear reactor with different types of reactivity. From the results of various benchmark tests with different types of reactivity insertions, the developed generalized Padé approximation (GPA) method shows high accuracy, high efficiency, and stable character of the solution.
APA, Harvard, Vancouver, ISO, and other styles
4

Picca, Paolo, and Roberto Furfaro. "Neutron inverse kinetics via Gaussian Processes." Annals of Nuclear Energy 47 (September 2012): 146–54. http://dx.doi.org/10.1016/j.anucene.2012.03.023.

Full text
APA, Harvard, Vancouver, ISO, and other styles
5

Hykes, Joshua M., and Rodolfo M. Ferrer. "Validation of CASMO5 neutron kinetics parameters." Annals of Nuclear Energy 136 (February 2020): 107015. http://dx.doi.org/10.1016/j.anucene.2019.107015.

Full text
APA, Harvard, Vancouver, ISO, and other styles
6

Pitot, Samuel, and Nicolas Alborghetti. "ICONE15-10190 COUPLED FULLY 3D NEUTRON KINETICS THERMAL-HYDRAULIC COMPUTATIONS FOR DNB ANALYSIS ON PWRS." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2007.15 (2007): _ICONE1510. http://dx.doi.org/10.1299/jsmeicone.2007.15._icone1510_84.

Full text
APA, Harvard, Vancouver, ISO, and other styles
7

Zhu, Wenzhang, and Qiang ZHAO. "ICONE19-43375 Solution of Point-Reactor Neutron Kinetics Equation by Gauss Precise Time-Integration Method." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2011.19 (2011): _ICONE1943. http://dx.doi.org/10.1299/jsmeicone.2011.19._icone1943_160.

Full text
APA, Harvard, Vancouver, ISO, and other styles
8

Tertychny-Dauri, V. Yu. "Nuclear Electromagnetic Generator: Introduction in Neutron Algebra and Elements of Neutron Kinetics." OALib 05, no. 06 (2018): 1–17. http://dx.doi.org/10.4236/oalib.1104670.

Full text
APA, Harvard, Vancouver, ISO, and other styles
9

Gozani, Tsahi, and Michael J. King. "Neutron kinetics in moderators and SNM detection through epithermal-neutron-induced fissions." Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment 805 (January 2016): 101–15. http://dx.doi.org/10.1016/j.nima.2015.08.031.

Full text
APA, Harvard, Vancouver, ISO, and other styles
10

Kale, Vivek, Rakesh Kumar, K. Obaidurrahman, and Avinash Gaikwad. "Linear stability analysis of a nuclear reactor using the lumped model." Nuclear Technology and Radiation Protection 31, no. 3 (2016): 218–27. http://dx.doi.org/10.2298/ntrp1603218k.

Full text
Abstract:
The stability analysis of a nuclear reactor is an important aspect in the design and operation of the reactor. A stable neutronic response to perturbations is essential from the safety point of view. In this paper, a general methodology has been developed for the linear stability analysis of nuclear reactors using the lumped reactor model. The reactor kinetics has been modelled using the point kinetics equations and the reactivity feedbacks from fuel, coolant and xenon have been modelled through the appropriate time dependent equations. These governing equations are linearized considering small perturbations in the reactor state around a steady operating point. The characteristic equation of the system is used to establish the stability zone of the reactor considering the reactivity coefficients as parameters. This methodology has been used to identify the stability region of a typical pressurized heavy water reactor. It is shown that the positive reactivity feedback from xenon narrows down the stability region. Further, it is observed that the neutron kinetics parameters (such as the number of delayed neutron precursor groups considered, the neutron generation time, the delayed neutron fractions, etc.) do not have a significant influence on the location of the stability boundary. The stability boundary is largely influenced by the parameters governing the evolution of the fuel and coolant temperature and xenon concentration.
APA, Harvard, Vancouver, ISO, and other styles

Dissertations / Theses on the topic "Neutron kinetics"

1

Kotchoubey, Jurij. "POLCA-T Neutron Kinetics Model Benchmarking." Thesis, KTH, Reaktorteknologi, 2015. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-176096.

Full text
Abstract:
The demand for computational tools that are capable to reliably predict the behavior of a nuclear reactor core in a variety of static and dynamic conditions does inevitably require a proper qualification of these tools for the intended purposes. One of the qualification methods is the verification of the code in question. Hereby, the correct implementation of the applied model as well as its flawless implementation in the code are scrutinized. The present work concerns with benchmarking as a substantial part of the verification of the three-dimensional, multigroup neutron kinetics model employed in the transient code POLCA-T. The benchmarking is done by solving some specified and widely used space-time kinetics benchmark problems and comparing the results to those of other, established and well-proven spatial kinetics codes. It is shown that the obtained results are accurate and consistent with corresponding solutions of other codes. In addition, a sensitivity analysis is carried out with the objective to study the sensitivity of the POLCA-T neutronics to variations in different numerical options. It is demonstrated that the model is numerically stable and provide reproducible results for a wide range of various numerical settings. Thus, the model is shown to be rather insensitive to significant variations in input, for example. The other consequence of this analysis is that, depending on the treated transient, the computing costs can be reduced by, for instance, employing larger time-steps during the time-integration process or using a reduced number of iterations. Based on the outcome of this study, one can finally conclude that the POLCA-T neutron kinetics is modeled and implemented correctly and thus, the model is fully capable to perform the assigned tasks.
APA, Harvard, Vancouver, ISO, and other styles
2

Jensen, Jeremy Davis. "The physical kinetics of water in Yucca Mountain zeolites via quasielastic neutron scattering." abstract and full text PDF (free order & download UNR users only), 2005. http://0-gateway.proquest.com.innopac.library.unr.edu/openurl?url_ver=Z39.88-2004&rft_val_fmt=info:ofi/fmt:kev:mtx:dissertation&res_dat=xri:pqdiss&rft_dat=xri:pqdiss:1433390.

Full text
APA, Harvard, Vancouver, ISO, and other styles
3

Peltonen, Joanna. "Development of effective algorithm for coupled thermal-hydraulics : neutron-kinetics analysis of reactivity transient." Licentiate thesis, Stockholm : Skolan för teknikvetenskap, Kungliga Tekniska högskolan, 2009. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-11033.

Full text
APA, Harvard, Vancouver, ISO, and other styles
4

Schramm, Marcelo. "An algorithm for multi-group two-dimensional neutron diffusion kinetics in nuclear reactor cores." reponame:Biblioteca Digital de Teses e Dissertações da UFRGS, 2016. http://hdl.handle.net/10183/142510.

Full text
Abstract:
O objetivo desta tese é introduzir uma nova metodologia para a cinética bidimensional multi- grupo de difusão de nêutrons em reatores nucleares. A metodologia apresentada usa uma aproximação polinomial em um domínio homogêneo retangular com condições de contornos não homogêneas. Como ela consiste em uma série de Taylor truncada, sua estimativa de erro varia de acordo com o tamanho do retângulo. Os coeficientes são obtidos principalmente pelas suas relações com o termo independente, que _e determinado pela equação diferencial. Estas relações são obtidas apenas pelas condições de contorno, e é demonstrado serem linearmente independentes. Um esquema numérico é feito para assegurar uma rápida convergência. Estes procedimentos feitos para um retângulo homogêneo são feitos para construir soluções para problemas de autovalor e dependentes do tempo de geometria ortogonal global com parâmetros seccionalmente constantes pelo método iterativo SOR. O autovalor dominante e sua autofunção são obtidos pelo método da potência no problema de autovalor. A solução para casos dependentes do tempo usam o método de Euler modificado na variável tempo. Quatro casos-teste clássicos são considerados para ilustração.<br>The objective of this thesis is to introduce a new methodology for two{dimensional multi{ group neutron diffusion kinetics in a reactor core. The presented methodology uses a polyno- mial approximation in a rectangular homogeneous domain with non{homogeneous boundary conditions. As it consists on a truncated Taylor series, its error estimates varies with the size of the rectangle. The coefficients are obtained mainly by their relations with the independent term, which is determined by the differential equation. These relations are obtained by the boundary conditions only, and these relations are proven linear independent. A numerical scheme is made to assure faster convergence. The procedures done for one homogeneous rectangle are used to construct the solution of global orthogonal geometry with step{wise constant parameters steady state and time dependent problems by the iterative SOR algo- rithm. The dominant eigenvalue and its eigenfunction are obtained by the power method in the eigenvalue problem. The solution for the time dependent cases uses the modi ed Euler method in the time variable. Four classic test cases are considered for illustration.
APA, Harvard, Vancouver, ISO, and other styles
5

Bernstein, Saskia. "Determination of reaction kinetics and mechanisms of 1.13 NM tobermorite by in-situ neutron diffraction." Diss., lmu, 2011. http://nbn-resolving.de/urn:nbn:de:bvb:19-143845.

Full text
APA, Harvard, Vancouver, ISO, and other styles
6

Lalik, Erwin. "Reduction kinetics and thermal behaviour of MoO₃ studed 'in situ' with neutron powder diffraction." Thesis, Birkbeck (University of London), 2003. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.408700.

Full text
APA, Harvard, Vancouver, ISO, and other styles
7

Lee, Bongjoon. "Analysis of the Kinetics of Filler Segregation in Granular Block copolymer Microstructure." Research Showcase @ CMU, 2016. http://repository.cmu.edu/dissertations/705.

Full text
Abstract:
Block copolymers have attracted interests for potential application ranging from dynamic photonic sensors to solid-state ion conductors. However, due to nucleation and growth mechanism, block copolymer inherently forms granular microstructure with defects such as grain boundaries. Understanding the microstructure of block copolymer is thus crucial in many applications because the microstructure determines the transport property of functional fillers such as ions in block copolymer template. Previous research has shown that athermal filler segregated to grain boundary of lamellae block copolymer and retards the grain coarsening. However, the kinetics of this grain boundary segregation during thermal annealing has not been revealed. Polystyrene-b-polyisoprene blended with deuterated polystyrene is used for neutron scattering study on studying the kinetics of grain boundary segregation. Deuterated polystyrene will segregate to grain boundaries, therefore, decorate grain boundary. The filler segregation behavior will be studied by comparing neutron scattering of polystyrene-b-polyisoprene/deuterated polystyrene with different annealing times (at T=130 deg C, duration of 0hr, 3hr, 1day, 3day and 7day, respectively). Invariant (Q) analysis along with grain mapping is conducted to quantitatively analyze the kinetics of grain boundary segregation. This kinetic was in good agreement with the McLean’s kinetic model for grain boundary segregation in metals. By applying Langmuir-Mclean’s segregation isotherm equation, we have predicted the equilibrium concentration of filler in grain boundary by calculating the strain energy stored in grain boundary.
APA, Harvard, Vancouver, ISO, and other styles
8

Peltonen, Joanna. "Effective Spatial Mapping for Coupled Code Analysis of Thermal–Hydraulics/Neutron–Kinetics of Boiling Water Reactors." Doctoral thesis, KTH, Kärnkraftsäkerhet, 2013. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-122088.

Full text
Abstract:
Analyses of nuclear reactor safety have increasingly required coupling of full three dimensional neutron kinetics (NK) core models with system transient thermal–hydraulics (TH) codes.  In order to produce results within a reasonable computing time, the coupled codes use two different spatial description of the reactor core.  The TH code uses few, typically 5 to 20 TH channels, which represent the core.  The NK code uses explicit one node for each fuel assembly.  Therefore, a spatial mapping of a coarse grid TH and a fine grid NK domain is necessary.  However, improper mappings may result in loss of valuable information, thus causing inaccurate prediction of safety parameters. The purpose of this thesis is to study the effectiveness of spatial coupling (channel refinement and spatial mapping) and develop recommendations for NK/TH mapping in simulation of safety transients.  Additionally, sensitivity of stability (measured by Decay Ratio and Frequency) to the different types of mapping schemes, is analyzed against OECD/NEA Ringhals–1 Stability Benchmark data. The research methodology consists of spatial coupling convergence study, by increasing the number of TH channels and varying mapping approaches, up to and including the reference case.  The reference case consists of one-to-one mapping: one TH channel per one fuel assembly.  The comparisons of the results are done for steady–state and transient results.  In this thesis mapping (spatial coupling) definition is formed and all the existing mapping approaches were gathered, analyzed and presented.  Additionally, to increase the efficiency and applicability of spatial mapping convergence, a new mapping methodology has been proposed.  The new mapping approach is based on hierarchical clustering method; the method of unsupervised learning that is adopted by many researchers in many different scientific fields, thanks to its flexibility and robustness.  The proposed new mapping method turns out to be very successful for spatial coupling problem and can be fully automatized allowing for significant time reduction in mapping convergence study. The steady–state results obtained from three different plant models for all the investigated cases are presented.  All models achieved well converged steady–state and local parameters were compared and it was concluded that solid basis for further transient analysis was found.  Analyzing the mapping performance, the best predictions for steady–state conditions are the mappings that include the power peaking factor feature alone or with any combination of other features.  Additionally it is of value to keep the core symmetry (symmetry feature).  The big part of this research is devoted to transient analysis.  The selection of transients was done such that it covers a wide range of transients and gathered knowledge may be used for other types of transients.  As a representative of a local perturbation, Control Rod Drop Accident was chosen.  A specially prepared Feedwater Transient was investigated as a regional perturbation and a Turbine Trip is an example of a global one.  In the case of local perturbation, it has been found that a number of TH channels is less important than the type of mapping, so a high number of TH channels does not guarantee improved results.  To avoid unnecessary averaging and to obtain the best prediction, hot channel and core zone where accident happens should be always separated from the rest.  The best performance is achieved with mapping according power peaking factors, and therefore this one is recommended for such type of perturbation. The regional perturbation has been found to be more challenging than the others.  This kind of perturbation is strongly dependent on mapping type that affects the power increase rate, SCRAM time, onset of instability, development of limit cycle, etc.  It has been also concluded that a special effort is needed for input model preparation.   In contrast to the regional perturbation, the global perturbation is found to be the least demanding transient.  Here, the number of TH channels and type of mapping do not have significant impact on average plant behaviour – general plant response is always well recreated.  A special effort has also been paid to investigate the core stability performance, in both global and regional mode.  It has been found that in case of unstable cores, a low number of TH channels significantly suppresses the instability.  For these cases number of TH channels is very important and therefore at least half of the core has to be modeled to have a confidence in predicted DR and FR.  In case of regional instability in order to get correct performance of out-of-phase oscillations, it is recommended to use full-scale model.  If this is not possible, the mapping which is a mixture of 1st power mode and power peaking factors, should be used. The general conclusions and recommendations are summarized at the end of this thesis.  Development of these recommendations was one of the purposes of this investigation and they should be taken into consideration while designing new coupled TH/NK models and choosing mapping strategy for a new transient analysis.<br><p>QC 20130516</p>
APA, Harvard, Vancouver, ISO, and other styles
9

Jackson, Christopher John [Verfasser]. "A dimensionally adaptive neutron kinetics algorithm for efficient nuclear plant safety analysis calculations / Christopher John Jackson." Karlsruhe : KIT-Bibliothek, 1997. http://d-nb.info/1198223820/34.

Full text
APA, Harvard, Vancouver, ISO, and other styles
10

DINIZ, RICARDO. "Obtencao das constantes de decaimento e abundancias relativas de neutrons atrasados atraves da analise de ruido em reatores de potencia zero." reponame:Repositório Institucional do IPEN, 2005. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11247.

Full text
Abstract:
Made available in DSpace on 2014-10-09T12:49:54Z (GMT). No. of bitstreams: 0<br>Made available in DSpace on 2014-10-09T14:02:56Z (GMT). No. of bitstreams: 1 10276.pdf: 7799693 bytes, checksum: 33b179c5ecbae276e3b4235673393d72 (MD5)<br>Tese (Doutoramento)<br>IPEN/T<br>Intituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN-SP
APA, Harvard, Vancouver, ISO, and other styles

Books on the topic "Neutron kinetics"

1

Helmut, Schober, Nagler Stephen E, and SpringerLink (Online service), eds. Studying Kinetics with Neutrons: Prospects for Time-Resolved Neutron Scattering. Springer-Verlag Berlin Heidelberg, 2010.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
2

Freltoft, Torsten. Neutron Study Of Aggregate Structure And Kinetics. Riso National Laboratory, 1986.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
3

Gadek, I͡A. Proverka trekhmernoĭ neĭtronno-kineticheskoĭ programmy HEXDYN3D s pomoshchʹi͡u II. ėtapa ėksperimentov prostranstvenno zavisimoĭ kinetiki na reaktore LR-O: [otchet, Rzhezh, noi͡abrʹ, 1988 g.]. Ústav jaderného v́yzkumu Řež, Informační středisko, 1988.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
4

Dickinson, Clive Fransis. The kinetics of the sodium carbonate-calcium silicate reaction using neutron diffraction and thermogravimetry. University of Salford, 1995.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
5

He dong li fan ying dui zhong zi dong li xue: Nuclear power reactor neutron dynamics. Guo fang gong ye chu ban she, 2005.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
6

Eckold, Götz, Helmut Schober, and Stephen E. Nagler, eds. Studying Kinetics with Neutrons. Springer Berlin Heidelberg, 2010. http://dx.doi.org/10.1007/978-3-642-03309-4.

Full text
APA, Harvard, Vancouver, ISO, and other styles
7

Eckold, Götz, Helmut Schober, and Stephen E. Nagler. Studying Kinetics with Neutrons: Prospects for Time-Resolved Neutron Scattering. Springer, 2012.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
8

S, Rohatgi Upendra, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Systems Technology., and Brookhaven National Laboratory, eds. RAMONA-4B, a computer code with three-dimensional neutron kinetics for BWR and SBWR system transients. Division of Systems Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1998.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
9

S, Rohatgi Upendra, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Systems Technology., and Brookhaven National Laboratory, eds. RAMONA-4B, a computer code with three-dimensional neutron kinetics for BWR and SBWR system transients. Division of Systems Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1998.

Find full text
APA, Harvard, Vancouver, ISO, and other styles
10

S, Rohatgi Upendra, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Systems Technology., and Brookhaven National Laboratory, eds. RAMONA-4B, a computer code with three-dimensional neutron kinetics for BWR and SBWR system transients. Division of Systems Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1998.

Find full text
APA, Harvard, Vancouver, ISO, and other styles

Book chapters on the topic "Neutron kinetics"

1

Pyeon, Cheol Ho. "Reactor Kinetics." In Accelerator-Driven System at Kyoto University Critical Assembly. Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-0344-0_3.

Full text
Abstract:
AbstractIn static and kinetic experimental analyses, the reactivity effect of introducing a neutron guide has been examined with various materials and adjustments of the beam window. With the objective of improving the KUCA core characteristics, the implementation of the neutron guide is predicted to increase the fast neutrons in directing the fuel region. With regard to the kinetic characteristics, the subcriticality and the prompt neutron decay constant are monitored for several core configurations and detector positions. The KUCA core is equipped to make locally a hard spectrum core region with the combined use of 235U fuel, a polyethylene moderator, and a Pb–Bi reflector for criticality. In this study, the first attempt is made to examine experimentally the characteristics of kinetics parameters in ADS comprised of 235U-fueled and Pb–Bi-zoned core, and spallation neutrons generated by an injection of 100 MeV protons onto the solid Pb–Bi target. Online monitoring of reactivity has been deduced in real time by the inverse kinetic method on the basis of the one-point kinetic equation with measured neutron signals in the core. Here, measurements by the one-point kinetic equation are validated through the subcriticality evaluation with the PNS histogram and the methodology by the inhour equation.
APA, Harvard, Vancouver, ISO, and other styles
2

Wu, Yican. "Neutron Kinetics." In Neutronics of Advanced Nuclear Systems. Springer Singapore, 2019. http://dx.doi.org/10.1007/978-981-13-6520-1_3.

Full text
APA, Harvard, Vancouver, ISO, and other styles
3

Eckold, Götz, and Helmut Schober. "Introduction to Neutron Techniques." In Studying Kinetics with Neutrons. Springer Berlin Heidelberg, 2009. http://dx.doi.org/10.1007/978-3-642-03309-4_1.

Full text
APA, Harvard, Vancouver, ISO, and other styles
4

Yamanaka, Masao. "Effective Delayed Neutron Fraction." In Accelerator-Driven System at Kyoto University Critical Assembly. Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-0344-0_4.

Full text
Abstract:
AbstractIn kinetic analyses on ADS, although adjoint flux distribution is defined under the existence of an external neutron source, an issue of the proper determination of the weighting function still remains in the definition to obtain the kinetics parameters in the fixed-source calculations. Here, an alternative methodology is proposed with the combined use of the k-ratio method and the reaction rates obtained by the fixed-source calculations, when the subcriticality level or the spectrum of the external neutron source is varied. In ADS experiments, the measurement of βeff is expected to provide complementary verification of the calculation and reliability of nuclear data. Then, the formulation of the Rossi-α method in the pulsed-neutron source has been already available for application to the subcriticality measurement in the pulsed-neutron source (PNS) experiments. Accordingly, the methodology is applied uniquely to deduce the βeff value with the pulsed-neutron source (spallation neutrons), with the combined use of the results of experiments and calculations. Using parameters α and ρ$, the values of βeff/Λ are deduced at near-critical configurations through experimental analyses. To estimate the numerical precision of Λ, the value of βeff/Λ is used as an index of Λ evaluation that is defined by a ratio of Λ values in the super-critical and subcritical states.
APA, Harvard, Vancouver, ISO, and other styles
5

Pyeon, Cheol Ho. "Introduction." In Accelerator-Driven System at Kyoto University Critical Assembly. Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-0344-0_1.

Full text
Abstract:
AbstractAt the Kyoto University Critical Assembly (KUCA), the accelerator-driven system (ADS) is composed of a solid-moderated and solid-reflected core (A-core) and a pulsed-neutron generator (14 MeV neutrons) or the fixed-filed alternating gradient (FFAG) accelerator (100 MeV protons). At KUCA, two external neutron sources, including 14 MeV neutrons and 100 MeV protons, are separately injected into the A-core, and employed for carrying out the ADS experiments. With the combined use of the A-core and two external neutron sources, basic and feasibility studies of ADS have been engaged in the examination of neutronics of ADS, through the measurements of statics and kinetics parameters of reactor physics, including subcritical multiplication factor, subcriticality, prompt neutron decay constant, effective delayed neutron fraction, neutron spectrum, and reaction rates.
APA, Harvard, Vancouver, ISO, and other styles
6

van Dijk, N. H. "Structure Evolution in Materials Studied by Time-Dependent Neutron Scattering." In Studying Kinetics with Neutrons. Springer Berlin Heidelberg, 2009. http://dx.doi.org/10.1007/978-3-642-03309-4_4.

Full text
APA, Harvard, Vancouver, ISO, and other styles
7

Oka, Yoshiaki. "Delayed Neutron and Nuclear Reactor Kinetics." In Nuclear Reactor Kinetics and Plant Control. Springer Japan, 2012. http://dx.doi.org/10.1007/978-4-431-54195-0_1.

Full text
APA, Harvard, Vancouver, ISO, and other styles
8

Riley, D. P., E. H. Kisi, E. Wu, T. Hansen, and P. Henry. "Applications of In Situ Neutron Diffraction to Optimisation of Novel Materials Synthesis." In Studying Kinetics with Neutrons. Springer Berlin Heidelberg, 2009. http://dx.doi.org/10.1007/978-3-642-03309-4_5.

Full text
APA, Harvard, Vancouver, ISO, and other styles
9

Wiedenmann, A., R. Gähler, R. P. May, et al. "Stroboscopic Small Angle Neutron Scattering Investigations of Microsecond Dynamics in Magnetic Nanomaterials." In Studying Kinetics with Neutrons. Springer Berlin Heidelberg, 2009. http://dx.doi.org/10.1007/978-3-642-03309-4_9.

Full text
APA, Harvard, Vancouver, ISO, and other styles
10

Lund, Reidar. "Small Angle Neutron Scattering as a Tool to Study Kinetics of Block Copolymer Micelles." In Studying Kinetics with Neutrons. Springer Berlin Heidelberg, 2009. http://dx.doi.org/10.1007/978-3-642-03309-4_8.

Full text
APA, Harvard, Vancouver, ISO, and other styles

Conference papers on the topic "Neutron kinetics"

1

Wang, Xin, Kathryn D. Huff, Manuele Aufiero, Per F. Peterson, and Massimiliano Fratoni. "Coupled Reactor Kinetics and Heat Transfer Model for Fluoride Salt-Cooled High-Temperature Reactor Transient Analysis." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60728.

Full text
Abstract:
Coupled reactor kinetics and heat transfer models have been developed at the University of California, Berkeley (UCB) to study Pebble-Bed, Fluoride-salt-cooled, High-temperature Reactors (PB-FHRs) transient behaviors. This paper discusses a coupled point kinetics model and a two-dimensional diffusion model. The former is based on the point kinetics equations with six groups of delayed neutrons and the lumped capacitance heat transfer equations. To account for the reflector effect on neutron lifetime, additional (fictional) groups of delayed neutrons are added in the point kinetics equations to represent the thermalized neutrons coming back from the reflectors. The latter is based on coupled multi-group neutron diffusion and finite element heat transfer model. Multi-group cross sections and diffusion coefficients are generated using the Monte Carlo code Serpent and defined as input in COMSOL 5.0.
APA, Harvard, Vancouver, ISO, and other styles
2

Song, Yingming, Qingyu Gao, Ke Wang, Yaping Guo, Lu Zhang, and Yongwei Yang. "Simulation on Neutron Space-Time Kinetics for Accelerator Driven Sub-Critical System." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66314.

Full text
Abstract:
Monte Carlo transport theory was applied to the variables space and time separated framework of neutron space-time kinetics calculation for Accelerator driven sub-critical reactor. The improved quasi-static approximation was combined with Monte Carlo neutron transport code (IQS/MC) for neutron space-time kinetic process of ADS sub-critical system. Besides, the IQS/MC simulation calculation program with visualization operation platform for ADS sub-critical system was developed simultaneously. The beam transient was analysed simulatedly based on the physical model of CIADS. Three-dimensional grid distributions of relative neutron flux of energy group were separated along time can be obtained by computing energy group separated of neutron flux, meanwhile the totally relative power, fuel temperature and outlet temperature of coolant at the core varied as the time were also obtained. The calculated results of IQS/MC method and point reactor method were compared, which agreed well with the relevant physics laws and verify that the IQS/MC method is applicable to the simulation of ADS neutron space-time kinetics and ADS neutronics transient security analysis.
APA, Harvard, Vancouver, ISO, and other styles
3

Liu, Dong, Xiuchun Luan, Tao Yu, Weining Zhao, and Lei Liu. "The Controllability of the Point Reactor Neutron Kinetics Equations." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30494.

Full text
Abstract:
In this paper, the conditions to ensure the controllability of the point reactor neutron kinetics equations are studied. In a nuclear reactor, due to the delayed neutron precursor concentration and the internal reactivity, the kinetics equations of the nuclear reactor are nonlinear. To solve the problem of the pole placement, the controllability of the point kinetics equations must be guaranteed. Then, a new method to analysis of the controllability conditions of the point kinetics equations of a reactor is carried out here. The method is based on the controllability matrix directly denoted by relevant symbols, and a formula used for controllability analysis is showed with symbols by calculating the determinant of the matrix. First, with using the linearization technique, the equations are linearized with respect to any possible equilibrium point. Subsequently, an analysis of the controllability of the general linear model that includes only one group delayed neutron precursor is performed, obtaining the interesting result that the controllability of the equations are controllable except when the effective precursor radioactive decay constant and the reciprocal of the fuel-to-coolant heat transfer mean time have the same value, which does not occur in practice. Thus, with the same method, the other analysis obtained the conditions to guarantee the controllability of the point kinetics equations with different groups delayed neutron precursor, which includes two-group, three-group and six-group models. Then, the results are compared with that of the numerical controllability matrix, obtaining the final conclusion that the results of the new analysis method give the closer results to the actual situation and list the restrictions that guarantee the controllability of the point reactor neutron kinetics equations.
APA, Harvard, Vancouver, ISO, and other styles
4

Pan, Shibiao. "Research on the Measurement of Deep Subcriticality Based on Pulsed Neutron Source Method." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-15883.

Full text
Abstract:
During the operating of nuclear reactor, the applicable condition of point reactor neutron kinetics is changed by control rods bulk insertion and reactor lattice deviating from the critical state, The measurement of reactivity, with the increase of subcriticality, shows that the results impact on the kinetic distortion effect, along with prompt neutron flux strongly deteriorated. According to the diffusion theory about prompt neutron and delayed neutron, the theoretical analysis and application of point reactor neutron kinetics have been carried out to quantify the kinetic distortion correction factors in subcritical systems, and these indicate that prompt neutron distributions are strongly affected by kinetic distortion. With the self-developed Pulsed Neutron Source measurement system, subcriticality measurement in different configuration of control rods in the zero power reactor of Nuclear Power Institute of China was carried out. In this paper the reactivity of the deep subcritical system are obtained by experiment of Pulsed Neutron Source method, meanwhile, the bias between experiment results of the areas-ratio method and the characteristic decay constant method are analyzed by comparing the condition of the experiment and the theory model, the main factors inducing the bias are found, which supplies helpful reference for other similar design and experiment.
APA, Harvard, Vancouver, ISO, and other styles
5

Shang, Xiaotong, Guanlin Shi, and Kan Wang. "One Step Method for Multigroup Adjoint Neutron Flux Through Continuous Energy Monte Carlo Calculation." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-82185.

Full text
Abstract:
The adjoint neutron flux is vital in the analysis of reactor kinetics parameters and reactor transient events. Both deterministic and Monte Carlo methods have been developed for the adjoint neutron flux calculation on the basis of multi-group cross sections which may vary significantly among different types of reactors. The iterated fission probability (IFP) method is introduced to calculate the neutron importance which is able to represent the adjoint neutron flux in continuous energy problem and have been applied to the calculation of kinetics parameters. However, the adjoint neutron flux can’t be obtained directly and applied to both Monte Carlo transient event analysis and deterministic methods. In this study, a method based on IFP is studied and implemented in Monte Carlo code RMC. The multi-group adjont neutron flux can be obtained directly through the discretization of energy and space with the modification of fission neutrons through continuous energy Monte Carlo calculations. The obtained multi-group adjoint neutron flux can be used in both Monte Carlo transient analysis and deterministic methods.
APA, Harvard, Vancouver, ISO, and other styles
6

Li, Zhifeng, Hongchun Wu, Chenghui Wan, and Tianliang Hu. "The Fast Three-Dimensional Space-Time Neutron Kinetic Model for Cartesian Geometry and Cylindrical Geometry." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60861.

Full text
Abstract:
In order to raise computation speed on the premise of enough numerical accuracy, the Predictor-Corrector Improved Quasi-Static (PC-IQS) method and Nodal Green’s Function Method (NGFM) were combined to solve the three-dimensional space-time neutron diffusion kinetics problems for Cartesian geometry. In addition, the improved quasi-static method and the Krylov algorithm were applied to solve the three-dimensional space-time neutron diffusion kinetics problems for cylindrical geometry. Based on the proposed model, the program of three-dimensional neutron space-time kinetic code has been tested by the two-dimensional and three-dimensional transient numerical benchmarks. The numerical results obtained by this work were in good agreement with the reference solutions.
APA, Harvard, Vancouver, ISO, and other styles
7

Hosseini, Seyed Abolfazl, and Naser Vosoughi. "Sensitivity Analysis of Kinetics Parameters of Tehran Research Reactor (TRR)." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75523.

Full text
Abstract:
In this research, effective delayed neutron fraction (βeff) and neutron generation time (Λ) of the Tehran Research Reactor (TRR) are calculated for different uranium enrichments from 14.84 w/o to 96.56 w/o U235 in two states of the TRR, (cold fuel, clad and coolant temperature of 20 °C; and hot fuel, clad and coolant temperature of 65, 49 and 44 °C, respectively) using the MTR_PC computer code. Comparative analysis shows that both βeff and Λ increase as fuel enrichment decreases. However, variation rate of βeff is not the same in two conditions. βeff in the hot state is larger than those calculated in the cold state when fuel enrichment goes more than 83.91%, while the situation is reverse for enrichment less than that. The obtained neutron generation time shows normal behavior for all different fuel enrichments. The variables involved in kinetics parameters calculations (i.e., neutron fission cross section, fuel enrichment, etc.) are investigated theoretically to confirm the results of calculations in cold and hot states. Variations of βeff and Λ with fuel burnup are studied too.
APA, Harvard, Vancouver, ISO, and other styles
8

Costa, Antonella Lombardi, Patrícia Amélia de Lima Reis, Claubia Pereira, Maria Auxiliadora Fortini Veloso, and Clarysson Alberto Mello da Silva. "Research Reactor Analysis Using Thermal Hydraulic and Neutron Kinetic Coupling." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30237.

Full text
Abstract:
Simulations of complex scenarios in nuclear power plants have been improved by the utilization of coupled thermal hydraulic (TH) and neutron kinetics (NK) system codes with the development of computer technology and new calculation methodology which made it possible to perform transport calculation schemes with accurate solutions. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0.0 code. By using this code, a multi-dimensional neutron kinetics model based on the NESTLE code can be implemented also. In this way, during a 3D TH/NK coupled simulation, RELAP5-3D calls the appropriate NESTLE subroutines to perform the calculations. The development and the assessment of the thermal hydraulic RELAP5 code model for the IPR-R1 TRIGA have been validated for steady state and transient situations and the results were published in preceding works. The model has been adapted to RELAP5-3D code and was verified to point kinetic calculations. After this, adequate cross sections to the NK code were supplied using the WIMSD5 code. The results of steady state and transient calculations using the 3D neutron modeling to the IPR-R1 are being presented in this paper.
APA, Harvard, Vancouver, ISO, and other styles
9

Verdú, G., D. Ginestar, R. Miró, et al. "NOKIN1D: one-dimensional neutron kinetics based on a nodal collocation method." In SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo, edited by D. Caruge, C. Calvin, C. M. Diop, F. Malvagi, and J. C. Trama. EDP Sciences, 2014. http://dx.doi.org/10.1051/snamc/201402211.

Full text
APA, Harvard, Vancouver, ISO, and other styles
10

Martinez-Quiroga, Victor, Sabahattin Akbas, Fatih Aydogan, Abderrafi M. Ougouag, and Chris Allison. "Coupling of RELAP5-SCDAP MOD4.0 and Neutronic Codes." In ASME 2015 International Mechanical Engineering Congress and Exposition. American Society of Mechanical Engineers, 2015. http://dx.doi.org/10.1115/imece2015-52991.

Full text
Abstract:
High-fidelity and accurate nuclear system codes play a key role in the design and analysis of complex nuclear power plants, which consist of multiple subsystems, such as the reactor core (and its fuel, burnable poisons, control elements, etc.), the reactor internal structures, the vessel, and the energy conversion subsystem and beyond to grid demand. Most commonly the interplay between these various subsystems is modeled using coupled codes, each of which represents one of the subsystems. And the most common direct coupling is that of thermal-hydraulics and neutronics codes. The subject of this paper is the coupling of codes that model not only thermal-hydraulics and neutronics, but also structural components damage. Furthermore, the neutronic component is not limited to the sole core solver. The coupled code system encompasses thermal-hydraulics, material performance of the fuel, neutronic solver, and neutronic data preparation. Thus, this paper presents a framework for coupling RELAP5/SCDAPSIM/MOD4.0 with a suite of neutron kinetics codes that includes NESTLE, DRAGON and a version of the ENDF library. The version of the RELAP5/SCDAPSIM/MOD4.0 code used in this work is one developed by Innovate System Software (ISS) as part of the international SCDAP Development and Training Program (SDTP) for best-estimate analysis to model reactor transients including severe accident phenomena. This RELAP5/SCDAPSIM/MOD4.0 code version is also capable of predicting nuclear fuel performance. It uses nodal power distributions to calculate mechanical and thermal parameters such as heat-up, oxidation and meltdown of fuel rods and control rods, the ballooning and rupture of fuel rod cladding, the release of fission products from fuel rods, and the disintegration of fuel rods into porous debris and molten material. On the neutronics side, this work uses the NESTLE and DRAGON codes. NESTLE is a multi-dimensional static and kinetic neutronic code developed at North Carolina State University. It solves up to four energy groups neutron diffusion equations utilizing the Nodal Expansion Method (NEM) in Cartesian or hexagonal geometry. The DRAGON code, developed at Ecole Polytechnique de Montreal, performs lattice physics calculations based on the neutron transport equation and is capable of using very fine energy group structures. In this work, we have developed a coupling approach to exchange data among the various modules. In the coupling process, the generated nuclear data (in fine multigroup energy structure) are collapsed down into two- or four-group energy structures for use in NESTLE. The neutron kinetics and thermal-hydraulics modules are coupled at each time step by using the cross-section data. The power distribution results of the neutronic calculations are transmitted to the thermal-hydraulics code. The spatial distribution of coolant density and the fuel-moderator temperature, which result from the thermal-hydraulic calculations, are transmitted back to the neutron kinetics codes and then the loop is closed using new neutronics results. Details of the actual data transfers will be described in the full length paper.
APA, Harvard, Vancouver, ISO, and other styles

Reports on the topic "Neutron kinetics"

1

McKenzie, IV, George Espy, Joetta Marie Goda, Travis Justin Grove, and Rene Gerardo Sanchez. Comparison Of A Neutron Kinetics Parameter For A Polyethylene Moderated Highly Enriched Uranium System. Office of Scientific and Technical Information (OSTI), 2017. http://dx.doi.org/10.2172/1352410.

Full text
APA, Harvard, Vancouver, ISO, and other styles
2

Grebennikov, A. N., A. K. Zhitnik, and O. A. Zvenigorodskaya. Results of comparative RBMK neutron computation using VNIIEF codes (cell computation, 3D statics, 3D kinetics). Final report. Office of Scientific and Technical Information (OSTI), 1995. http://dx.doi.org/10.2172/219464.

Full text
APA, Harvard, Vancouver, ISO, and other styles
3

Rohatgi, U. S., H. S. Cheng, H. J. Khan, A. N. Mallen, and L. Y. Neymotin. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual. Office of Scientific and Technical Information (OSTI), 1998. http://dx.doi.org/10.2172/576035.

Full text
APA, Harvard, Vancouver, ISO, and other styles
4

Rohatgi, U. S., H. S. Cheng, H. J. Khan, A. N. Mallen, and L. Y. Neymotin. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - models and correlations. Office of Scientific and Technical Information (OSTI), 1998. http://dx.doi.org/10.2172/576037.

Full text
APA, Harvard, Vancouver, ISO, and other styles
5

Tendler, M., and D. Heifetz. Neutral particle kinetics in fusion devices. Office of Scientific and Technical Information (OSTI), 1986. http://dx.doi.org/10.2172/5851515.

Full text
APA, Harvard, Vancouver, ISO, and other styles
6

Stotler, D. P., C. S. Chang, S. H. Ku, J. Lang, and G. Park. Energy Conservation Tests of a Coupled Kinetic-kinetic Plasma-neutral Transport Code. Office of Scientific and Technical Information (OSTI), 2012. http://dx.doi.org/10.2172/1056658.

Full text
APA, Harvard, Vancouver, ISO, and other styles
7

Higgins, Daniel James, Kyle Thomas Schmitt, Shea Morgan Mosby, and Fredrik Tovesson. Total Kinetic Energy and Fragment Mass Distribution of Neutron-Induced Fission of U-233. Office of Scientific and Technical Information (OSTI), 2017. http://dx.doi.org/10.2172/1396153.

Full text
APA, Harvard, Vancouver, ISO, and other styles
8

Joseph, Ilon. Code Coupling via Jacobian-Free Newton-Krylov Algorithms with Application to Magnetized Fluid Plasma and Kinetic Neutral Models. Office of Scientific and Technical Information (OSTI), 2014. http://dx.doi.org/10.2172/1249135.

Full text
APA, Harvard, Vancouver, ISO, and other styles
9

Pogorelov, Nikolai, and Ming Zhang. Collaborative Research: A Model of Partially Ionized Plasma Flows with Kinetic Treatment of Neutral Atoms and Nonthermal Ions. Office of Scientific and Technical Information (OSTI), 2016. http://dx.doi.org/10.2172/1326403.

Full text
APA, Harvard, Vancouver, ISO, and other styles
10

Pogorelov, Nikolai, Ming Zhang, Sergey Borovikov, et al. Collaborative Research: A Model of Partially Ionized Plasma Flows with Kinetic Treatment of Neutral Atoms and Nonthermal Ions. Office of Scientific and Technical Information (OSTI), 2016. http://dx.doi.org/10.2172/1326821.

Full text
APA, Harvard, Vancouver, ISO, and other styles
We offer discounts on all premium plans for authors whose works are included in thematic literature selections. Contact us to get a unique promo code!

To the bibliography