WorldWideScience

Sample records for lafpx-v multigroup fission

  1. On the calculation of multi-group fission spectrum vectors

    International Nuclear Information System (INIS)

    Mueller, E.Z.

    1984-05-01

    In this report, the problem of calculating fission spectrum vectors in a consistent manner is formulated. The practical implications of using fission spectrum vectors in multi-group transport calculations are also addressed. The significance of the weighting spectra used for the calculation of fission spectrum vectors is illustrated for the case of a simple neutronic assembly

  2. Preparation of multigroup lumped fission product cross-sections from ENDF/B-VI for FBRs

    International Nuclear Information System (INIS)

    Devan, K.; Gopalakrishnan, V.; Mohanakrishnan, P.; Sridharan, M.S.

    1997-01-01

    Multigroup pseudo fission product cross-sections were computed from the American evaluated nuclear data library ENDF/B-VI, corresponding to various burnups of the proposed 500 MWe prototype fast breeder reactor (PFBR), in India. The data were derived from the cross-sections of 111 selected fission products that account for almost complete capture of fission products in an FBR. The dependence of burnup on the pseudo fission product cross-sections, and comparison with other data sets, viz. JNDC, ENDF/B-IV and ABBN, are discussed. (author)

  3. Development and testing of multigroup library with correction of self-shielding effects in fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Zou Jun; He Zhaozhong; Zeng Qin; Qiu Yuefeng; Wang Minghuang

    2010-01-01

    A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K eff , neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.

  4. Integral tests of coupled multigroup neutron and gamma cross sections with fission and fusion sources

    International Nuclear Information System (INIS)

    Schriewer, J.; Hehn, G.; Mattes, M.; Pfister, G.; Keinert, J.

    1978-01-01

    Calculations were made for different benchmark experiments in order to test the coupled multigroup neutron and gamma library EURLIB-3 with 100 neutron groups and 20 gamma groups. In cooperation with EURATOM, Ispra, we produced this shielding library recently from ENDF/B-IV data for application in fission and fusion technology. Integral checks were performed for natural lithium, carbon, oxygen, and iron. Since iron is the most important structural material in nuclear technology, we started with calculations of iron benchmark experiments. Most of them are integral experiments of INR, Karlsruhe, but comparisons were also done with benchmark experiments from USA and Japan. For the experiments with fission sources we got satisfying results. All details of the resonances cannot be checked with flux measurements and multigroup cross sections used. But some averaged resonance behaviour of the measured and calculated fluxes can be compared and checked within the error limits given. We get greater differences in the calculations of benchmark experiments with 14 MeV neutron sources. For iron the group cross sections of EURLIB-3 produce an underestimation of the neutron flux in a broad energy region below the source energy. The conclusion is that the energy degradation by inelastic scattering is too strong. For fusion application the anisotropy of the inelastic scatter process must be taken into account, which isn't done by the processing codes at present. If this effect isn't enough, additional corrections have to be applied to the inelastic cross sections of iron in ENDF/B-IV. (author)

  5. FPDCYS and FPSPEC: computer programs for calculating fission-product beta and gamma multigroup spectra from ENDF/B-IV data

    International Nuclear Information System (INIS)

    Stamatelatos, M.G.; England, T.R.

    1977-05-01

    FPDCYS and FPSPEC are two FORTRAN computer programs used at the Los Alamos Scientific Laboratory (LASL), in conjunction with the CINDER-10 program, for calculating cumulative fission-product beta and/or gamma multigroup spectra in arbitrary energy structures, and for arbitrary neutron irradiation periods and cooling times. FPDCYS processes ENDF/B-IV fission-product decay energy data to generate multigroup beta and gamma spectra from individual ENDF/B-IV fission-product nuclides. FPSPEC further uses these spectra and the corresponding nuclide activities calculated by the CINDER-10 code to produce cumulative beta and gamma spectra in the same energy grids in which FPDCYS generates individual isotope decay spectra. The code system consisting of CINDER-10, FPDCYS, and FPSPEC has been used for comparisons with experimental spectra and continues to be used at LASL for generating spectra in special user-oriented group structures. 3 figures

  6. Preparation of lumped fission product (FP) cross sections for a multigroup library

    International Nuclear Information System (INIS)

    Ono, S.; Corcuera, R.P.

    1984-01-01

    A method for the calculation of lumped Fission Product (FP) cross sections has been developed. The group constants fo each nuclide are generated by NJOY code, based on ENDF/B-V data. In this first version, cross section of 28 nuclides are lumped for typical characteristics of Binary Breeder Reactor (BBR). One energy group calculations are made for a 1000 MWe fast reactor to verify the influence of burnup, number of FP and fuel composition on the lumped fission product cross sections. (Author) [pt

  7. Production and testing of the ENEA-Bologna VITJEFF32.BOLIB (JEFF-3.2) multi-group (199 n + 42 γ) cross section library in AMPX format for nuclear fission applications

    Science.gov (United States)

    Pescarini, Massimo; Orsi, Roberto; Frisoni, Manuela

    2017-09-01

    The ENEA-Bologna Nuclear Data Group produced the VITJEFF32.BOLIB multi-group coupled neutron/photon (199 n + 42 γ) cross section library in AMPX format, based on the OECD-NEA Data Bank JEFF-3.2 evaluated nuclear data library. VITJEFF32.BOLIB was conceived for nuclear fission applications as European counterpart of the ORNL VITAMIN-B7 similar library (ENDF/B-VII.0 data). VITJEFF32.BOLIB has the same neutron and photon energy group structure as the former ORNL VITAMIN-B6 reference library (ENDF/B-VI.3 data) and was produced using similar data processing methodologies, based on the LANL NJOY-2012.53 nuclear data processing system for the generation of the nuclide cross section data files in GENDF format. Then the ENEA-Bologna 2007 Revision of the ORNL SCAMPI nuclear data processing system was used for the conversion into the AMPX format. VITJEFF32.BOLIB contains processed cross section data files for 190 nuclides, obtained through the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the ORNL DOORS system, can be generated from VITJEFF32.BOLIB through the cited SCAMPI version. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF32.BOLIB validation.

  8. Production and testing of the ENEA-Bologna VITJEFF32.BOLIB (JEFF-3.2 multi-group (199 n + 42 γ cross section library in AMPX format for nuclear fission applications

    Directory of Open Access Journals (Sweden)

    Pescarini Massimo

    2017-01-01

    Full Text Available The ENEA-Bologna Nuclear Data Group produced the VITJEFF32.BOLIB multi-group coupled neutron/photon (199 n + 42 γ cross section library in AMPX format, based on the OECD-NEA Data Bank JEFF-3.2 evaluated nuclear data library. VITJEFF32.BOLIB was conceived for nuclear fission applications as European counterpart of the ORNL VITAMIN-B7 similar library (ENDF/B-VII.0 data. VITJEFF32.BOLIB has the same neutron and photon energy group structure as the former ORNL VITAMIN-B6 reference library (ENDF/B-VI.3 data and was produced using similar data processing methodologies, based on the LANL NJOY-2012.53 nuclear data processing system for the generation of the nuclide cross section data files in GENDF format. Then the ENEA-Bologna 2007 Revision of the ORNL SCAMPI nuclear data processing system was used for the conversion into the AMPX format. VITJEFF32.BOLIB contains processed cross section data files for 190 nuclides, obtained through the Bondarenko (f-factor method for the treatment of neutron resonance self-shielding and temperature effects. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the ORNL DOORS system, can be generated from VITJEFF32.BOLIB through the cited SCAMPI version. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF32.BOLIB validation.

  9. ENEA-Bologna production and testing of Jeff-3.1 multi-group cross section libraries for nuclear fission applications

    International Nuclear Information System (INIS)

    Pescarini, M.; Orsi, R.; Sinitsa, V.

    2008-01-01

    The ENEA-Bologna Nuclear Data Group produced the JEFF-3.1 VITJEFF31.BOLIB and MATJEFF31. BOLIB fine-group coupled neutron and photon (199 n + 42 γ) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format, with the same specifications and energy group structure of the Endf/B-VI-3 VITAMIN-B6 American library. Each library, containing 181 nuclide cross section files, was generated from the same set of cross section data files in GENDF format, obtained through the Bondarenko (f-factor) method, with an ENEA-Bologna revised version of the GROUPR module of the NJOY-99.160 system. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the DANTSYS and DOORS systems, can be generated from VITJEFF31.BOLIB and MATJEFF31.BOLIB through, respectively, further data processing with an ENEA-Bologna revised version of the SCAMPI system and with the TRANSX code. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF31.BOLIB validation. (authors)

  10. ENEA-Bologna production and testing of JEF-2.2 multi-group cross section libraries for nuclear fission applications

    International Nuclear Information System (INIS)

    Pescarini, M.; Orsi, R.; Martinelli, T.; Sinitsa, V.; Blokhin, A.I.

    2005-01-01

    The ENEA-Bologna Nuclear Data Group produced the VITJEF22.BOLIB (NEA-1699/01 ZZ VITJEF22.BOLIB) and MATJEF22.BOLIB (NEA-1740/01 ZZ MATJEF22.BOLIB) fine-group coupled neutron and photon (199 n + 42 γ) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format and based on the JEF-2.2 European nuclear data file. Both the libraries were produced from the same set of cross section files in GENDF format, generated with the NJOY-94.66 nuclear data processing system. The present libraries can be considered as European counterparts of the VITAMIN-B6 (DLC-0184 ZZ VITAMIN-B6) American library in AMPX format, based on the ENDF/B-VI Release 3 American nuclear data file. In fact they have the same general features and the same neutron and photon energy group structures as VITAMIN-B6. In particular, all these libraries are pseudo-problem-independent and based on the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Each ENEA-Bologna library contains a set of 133 nuclide cross section files processed at 4 temperatures (300 K, 600 K, 1000 K and 2100 K) and obtained for the most part with 6 to 8 values of the background cross section σ 0 . Thermal scattering cross sections were processed at all the temperatures available in the JEF-2.2 thermal scattering law data file for 5 additional bound nuclides: H-1 in light water, H-1 in polyethylene, H-2 in heavy water, C in graphite and Be in beryllium metal. Collapsed working libraries of self-shielded cross sections in the formats used by the deterministic transport codes of the DANTSYS and DOORS systems can be generated from VITJEF22.BOLIB and MATJEF22.BOLIB through, respectively, further problem-dependent data processing with the AMPX or SCAMPI nuclear data processing systems and with the TRANSX code. (authors)

  11. CHARTB multigroup transport package

    International Nuclear Information System (INIS)

    Baker, L.

    1979-03-01

    The physics and numerical implementation of the radiation transport routine used in the CHARTB MHD code are discussed. It is a one-dimensional (Cartesian, cylindrical, and spherical symmetry), multigroup,, diffusion approximation. Tests and applications will be discussed as well

  12. FENDL multigroup libraries

    International Nuclear Information System (INIS)

    Ganesan, S.; Muir, D.W.

    1992-01-01

    Selected neutron reaction nuclear data libraries and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into MATXSR format using the NJOY system on the VAX4000 computer of the IAEA. This document lists the resulting multigroup data libraries. All the multigroup data generated are available cost-free upon request from the IAEA Nuclear Data Section. (author). 9 refs

  13. Elaboration of a nodal method to solve the steady state multigroup diffusion equation. Study and use of the multigroup diffusion code DAHRA

    International Nuclear Information System (INIS)

    Halilou, A.; Lounici, A.

    1981-01-01

    The subject is divided in two parts: In the first part a nodal method has been worked out to solve the steady state multigroup diffusion equation. This method belongs to the same set of nodal methods currently used to calculate the exact fission powers and neutron fluxes in a very short computing time. It has been tested on a two dimensional idealized reactors. The effective multiplication factor and the fission powers for each fuel element have been calculated. The second part consists in studying and mastering the multigroup diffusion code DAHRA - a reduced version of DIANE - a two dimensional code using finite difference method

  14. NDS multigroup cross section libraries

    International Nuclear Information System (INIS)

    DayDay, N.

    1981-12-01

    A summary description and documentation of the multigroup cross section libraries which exist at the IAEA Nuclear Data Section are given in this report. The libraries listed are available either on tape or in printed form. (author)

  15. Multi-Group Library Generation with Explicit Resonance Interference Using Continuous Energy Monte Carlo Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ho Jin; Cho, Jin Young [KAERI, Daejeon (Korea, Republic of); Kim, Kang Seog [Oak Ridge National Laboratory, Oak Ridge (United States); Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    In this study, multi-group cross section libraries for the DeCART code were generated using a new procedure. The new procedure includes generating the RI tables based on the MC calculations, correcting the effective fission product yield calculations, and considering most of the fission products as resonant nuclides. KAERI (Korea Atomic Energy Research Institute) has developed the transport lattice code KARMA (Kernel Analyzer by Ray-tracing Method for fuel Assembly) and DeCART (Deterministic Core Analysis based on Ray Tracing) for a multi-group neutron transport analysis of light water reactors (LWRs). These codes adopt the method of characteristics (MOC) to solve the multi-group transport equation and resonance fixed source problem, the subgroup and the direct iteration method with resonance integral tables for resonance treatment. With the development of the DeCART and KARMA code, KAERI has established its own library generation system for a multi-group transport calculation. In the KAERI library generation system, the multi-group average cross section and resonance integral (RI) table are generated and edited using PENDF (point-wise ENDF) and GENDF (group-wise ENDF) produced by the NJOY code. The new method does not need additional processing because the MC method can handle any geometry information and material composition. In this study, the new method is applied to the dominant resonance nuclide such as U{sup 235} and U{sup 238} and the conventional method is applied to the minor resonance nuclides. To examine the newly generated multi-group cross section libraries, various benchmark calculations such as pin-cell, FA, and core depletion problem are performed and the results are compared with the reference solutions. Overall, the results by the new method agree well with the reference solution. The new procedure based on the MC method were verified and provided the multi-group library that can be used in the SMR nuclear design analysis.

  16. Approximate albedo boundary conditions for energy multigroup X,Y-geometry discrete ordinates nuclear global calculations

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Davi J.M.; Nunes, Carlos E.A.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: davijmsilva@yahoo.com.br, E-mail: ceanunes@yahoo.com.br, E-mail: rcbarros@pq.cnpq.br [Secretaria Municipal de Educacao de Itaborai, RJ (Brazil); Universidade Estacio de Sa (UNESA), Rio de Janeiro, RJ (Brazil); Universidade do Estado do Rio de Janeiro (UERJ), Novra Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional

    2017-11-01

    Discussed here is the accuracy of approximate albedo boundary conditions for energy multigroup discrete ordinates (S{sub N}) eigenvalue problems in two-dimensional rectangular geometry for criticality calculations in neutron fission reacting systems, such as nuclear reactors. The multigroup (S{sub N}) albedo matrix substitutes approximately the non-multiplying media around the core, e.g., baffle and reflector, as we neglect the transverse leakage terms within these non-multiplying regions. Numerical results to a typical model problem are given to illustrate the accuracy versus the computer running time. (author)

  17. An analytical multigroup benchmark for (n, γ) and (n, n', γ) verification of diffusion theory algorithms

    International Nuclear Information System (INIS)

    Ganapol, B.D.

    2011-01-01

    Highlights: → Coupled neutron and gamma transport is considered in the multigroup diffusion approximation. → The model accommodates fission, up- and down-scattering and common neutron-gamma interactions. → The exact solution to the diffusion equation in a heterogeneous media of any number of regions is found. → The solution is shown to parallel the one-group case in a homogeneous medium. → The discussion concludes with a heterogeneous, 2 fuel-plate 93.2% enriched reactor fuel benchmark demonstration. - Abstract: The angular flux for the 'rod model' describing coupled neutron/gamma (n, γ) diffusion has a particularly straightforward analytical representation when viewed from the perspective of a one-group homogeneous medium. Cast in the form of matrix functions of a diagonalizable matrix, the solution to the multigroup equations in heterogeneous media is greatly simplified. We shall show exactly how the one-group homogeneous medium solution leads to the multigroup solution.

  18. Some reciprocity-like relations in multi-group neutron diffusion and transport theory over bare homogeneous regions

    International Nuclear Information System (INIS)

    Modak, R.S.; Sahni, D.C.

    1996-01-01

    Some simple reciprocity-like relations that exist in multi-group neutron diffusion and transport theory over bare homogeneous regions are presented. These relations do not involve the adjoint solutions and are directly related to numerical schemes based on an explicit evaluation of the fission matrix. (author)

  19. Parallel computation of multigroup reactivity coefficient using iterative method

    Science.gov (United States)

    Susmikanti, Mike; Dewayatna, Winter

    2013-09-01

    One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium.

  20. Procedure to Generate the MPACT Multigroup Library

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-17

    The CASL neutronics simulator MPACT is under development for the neutronics and T-H coupled simulation for the light water reactor. The objective of this document is focused on reviewing the current procedure to generate the MPACT multigroup library. Detailed methodologies and procedures are included in this document for further discussion to improve the MPACT multigroup library.

  1. TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticality studies

    International Nuclear Information System (INIS)

    Ermumcu, G.; Gonnord, J.; Nimal, J.C.

    1980-01-01

    TRIMARAN is developed for safety analysis of nuclear components containing fissionable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method, in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies

  2. Delayed fission

    Energy Technology Data Exchange (ETDEWEB)

    Hatsukawa, Yuichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-07-01

    Delayed fission is a nuclear decay process that couples {beta} decay and fission. In the delayed fission process, a parent nucleus undergoes {beta} decay and thereby populates excited states in the daughter. If these states are of energies comparable to or greater than the fission barrier of the daughter, then fission may compete with other decay modes of the excited states in the daughter. In this paper, mechanism and some experiments of the delayed fission will be discussed. (author)

  3. ERRORJ, Multigroup covariance matrices generation from ENDF-6 format

    International Nuclear Information System (INIS)

    Chiba, Go

    2007-01-01

    1 - Description of program or function: ERRORJ produces multigroup covariance matrices from ENDF-6 format following mainly the methods of the ERRORR module in NJOY94.105. New version differs from previous version in the following features: Additional features in ERRORJ with respect to the NJOY94.105/ERRORR module: - expands processing for the covariance matrices of resolved and unresolved resonance parameters; - processes average cosine of scattering angle and fission spectrum; - treats cross-correlation between different materials and reactions; - accepts input of multigroup constants with various forms (user input, GENDF, etc.); - outputs files with various formats through utility NJOYCOVX (COVERX format, correlation matrix, relative error and standard deviation); - uses a 1% sensitivity method for processing of resonance parameters; - ERRORJ can process the JENDL-3.2 and 3.3 covariance matrices. Additional features of the version 2 with respect to the previous version of ERRORJ: - Since the release of version 2, ERRORJ has been modified to increase its reliability and stability, - calculation of the correlation coefficients in the resonance region, - Option for high-speed calculation is implemented, - Perturbation amount is optimised in a sensitivity calculation, - Effect of the resonance self-shielding can be considered, - a compact covariance format (LCOMP=2) proposed by N. M. Larson can be read. Additional features of the version 2.2.1 with respect to the previous version of ERRORJ: - Several routines were modified to reduce calculation time. The new one needs shorter calculation time (50-70%) than the old version without changing results. - In the U-233 and Pu-241 files of JENDL-3.3 an inconsistency between resonance parameters in MF=32 and those in MF=2 was corrected. NEA-1676/06: This version differs from the previous one (NEA-1676/05) in the following: ERRORJ2.2.1 was modified to treat the self-shielding effect accurately. NEA-1676/07: This version

  4. Nuclear fission

    International Nuclear Information System (INIS)

    Kodama, T.

    1981-01-01

    The nuclear fission process is pedagogically reviewed from a macroscopic-microscopic point of view. The Droplet model is considered. The fission dynamics is discussed utilizing path integrals and semiclassical methods. (L.C.) [pt

  5. Range calculations using multigroup transport methods

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

    1979-01-01

    Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of particle range distributions. These techniques are illustrated by analysis of Au-196 atoms recoiling from (n,2n) reactions with gold. The results of these calculations agree very well with range calculations performed with the atomistic code MARLOWE. Although some detail of the atomistic model is lost in the multigroup transport calculations, the improved computational speed should prove useful in the solution of fusion material design problems

  6. A multilevel in space and energy solver for multigroup diffusion eigenvalue problems

    Directory of Open Access Journals (Sweden)

    Ben C. Yee

    2017-09-01

    Full Text Available In this paper, we present a new multilevel in space and energy diffusion (MSED method for solving multigroup diffusion eigenvalue problems. The MSED method can be described as a PI scheme with three additional features: (1 a grey (one-group diffusion equation used to efficiently converge the fission source and eigenvalue, (2 a space-dependent Wielandt shift technique used to reduce the number of PIs required, and (3 a multigrid-in-space linear solver for the linear solves required by each PI step. In MSED, the convergence of the solution of the multigroup diffusion eigenvalue problem is accelerated by performing work on lower-order equations with only one group and/or coarser spatial grids. Results from several Fourier analyses and a one-dimensional test code are provided to verify the efficiency of the MSED method and to justify the incorporation of the grey diffusion equation and the multigrid linear solver. These results highlight the potential efficiency of the MSED method as a solver for multidimensional multigroup diffusion eigenvalue problems, and they serve as a proof of principle for future work. Our ultimate goal is to implement the MSED method as an efficient solver for the two-dimensional/three-dimensional coarse mesh finite difference diffusion system in the Michigan parallel characteristics transport code. The work in this paper represents a necessary step towards that goal.

  7. An energy recondensation method using the discrete generalized multigroup energy expansion theory

    International Nuclear Information System (INIS)

    Zhu Lei; Forget, Benoit

    2011-01-01

    Highlights: → Discrete-generalized multigroup method was implemented as a recondensation scheme. → Coarse group cross-sections were recondensed from core-level solution. → Neighboring effect of reflector and MOX bundle was improved. → Methodology was shown to be fully consistent when a flat angular flux approximation is used. - Abstract: In this paper, the discrete generalized multigroup (DGM) method was used to recondense the coarse group cross-sections using the core level solution, thus providing a correction for neighboring effect found at the core level. This approach was tested using a discrete ordinates implementation in both 1-D and 2-D. Results indicate that 2 or 3 iterations can substantially improve the flux and fission density errors associated with strong interfacial spectral changes as found in the presence of strong absorbers, reflector of mixed-oxide fuel. The methodology is also proven to be fully consistent with the multigroup methodology as long as a flat-flux approximation is used spatially.

  8. Macroscopic multigroup constants for accelerator driven system core calculation

    International Nuclear Information System (INIS)

    Heimlich, Adino; Santos, Rubens Souza dos

    2011-01-01

    The high-level wastes stored in facilities above ground or shallow repositories, in close connection with its nuclear power plant, can take almost 106 years before the radiotoxicity became of the order of the background. While the disposal issue is not urgent from a technical viewpoint, it is recognized that extended storage in the facilities is not acceptable since these ones cannot provide sufficient isolation in the long term and neither is it ethical to leave the waste problem to future generations. A technique to diminish this time is to transmute these long-lived elements into short-lived elements. The approach is to use an Accelerator Driven System (ADS), a sub-critical arrangement which uses a Spallation Neutron Source (SNS), after separation the minor actinides and the long-lived fission products (LLFP), to convert them to short-lived isotopes. As an advanced reactor fuel, still today, there is a few data around these type of core systems. In this paper we generate macroscopic multigroup constants for use in calculations of a typical ADS fuel, take into consideration, the ENDF/BVI data file. Four energy groups are chosen to collapse the data from ENDF/B-VI data file by PREPRO code. A typical MOX fuel cell is used to validate the methodology. The results are used to calculate one typical subcritical ADS core. (author)

  9. Generating and verification of ACE-multigroup library for MCNP

    International Nuclear Information System (INIS)

    Chen Chaobin; Hu Zehua; Chen Yixue; Wu Jun; Yang Shouhai

    2012-01-01

    The Monte Carlo code MCNP can handle multigroup calculations and a sample multigroup set based on ENDF/B-V, MGXSNP, is available for MCNP for coupled neutron-photon transport. However, this library is not suit- able for all problems, and there is a need for users to be able to generate multigroup libraries tailored to their specific applications. For these purposes CSPT (cross section processing tool) is created to generate multigroup library for MCNP from deterministic multigroup cross sections (GENDF or ANISN format at present). Several ACE-multigroup libraries based on ENDF/B-VII.0 converted and verified in this work, we drawn the conclusion that the CSPT code works correctly and the libraries produced are credible. (authors)

  10. Ternary fission

    International Nuclear Information System (INIS)

    Wagemans, C.

    1991-01-01

    Since its discovery in 1946, light (charged) particle accompanied fission (ternary fission) has been extensively studied, for spontaneous as well as for induced fission reactions. The reason for this interest was twofold: the ternary particles being emitted in space and time close to the scission point were expected to supply information on the scission point configuration and the ternary fission process was an important source of helium, tritium, and hydrogen production in nuclear reactors, for which data were requested by the nuclear industry. Significant experimental progress has been realized with the advent of high-resolution detectors, powerful multiparameter data acquisition systems, and intense neutron and photon beams. As far as theory is concerned, the trajectory calculations (in which scission point parameters are deduced from the experimental observations) have been very much improved. An attempt was made to explain ternary particle emission in terms of a Plateau-Rayleigh hydrodynamical instability of a relatively long cylindrical neck or cylindrical nucleus. New results have also been obtained on the so-called open-quotes trueclose quotes ternary fission (fission in three about-equal fragments). The spontaneous emission of charged particles has also clearly been demonstrated in recent years. This chapter discusses the main characteristics of ternary fission, theoretical models, light particle emission probabilities, the dependence of the emission probabilities on experimental variables, light particle energy distributions, light particle angular distributions, correlations between light particle accompanied fission observables, open-quotes trueclose quotes ternary fission, and spontaneous emission of heavy ions. 143 refs., 18 figs., 8 tabs

  11. Introduction of corrections taking into account interdependence of multigroup constants to the results of multigroup perturbation theory calculations

    International Nuclear Information System (INIS)

    Raskach, K. F.

    2012-01-01

    In multigroup calculations of reactivity and sensitivity coefficients, methodical errors can appear if the interdependence of multigroup constants is not taken into account. For this effect to be taken into account, so-called implicit components of the aforementioned values are introduced. A simple technique for computing these values is proposed. It is based on the use of subgroup parameters.

  12. Ternary fission

    Indian Academy of Sciences (India)

    the energy minimization of all possible ternary breakups of a heavy radioactive nucleus. Further, within the TCM we have analysed the competition between different geometries as well as different positioning of the fragments. Also, an attempt was made to calculate the mass distribution of ternary fission process within the ...

  13. A multigroup treatment of radiation transport

    International Nuclear Information System (INIS)

    Tahir, N.A.; Laing, E.W.; Nicholas, D.J.

    1980-12-01

    A multi-group radiation package is outlined which will accurately handle radiation transfer problems in laser-produced plasmas. Bremsstrahlung, recombination and line radiation are included as well as fast electron Bremsstrahlung radiation. The entire radiation field is divided into a large number of groups (typically 20), which diffuse radiation energy in real space as well as in energy space, the latter occurring via electron-radiation interaction. Using this model a radiation transport code will be developed to be incorporated into MEDUSA. This modified version of MEDUSA will be used to study radiative preheat effects in laser-compression experiments at the Central Laser Facility, Rutherford Laboratory. The model is also relevant to heavy ion fusion studies. (author)

  14. Multigroup cross section library; WIMS library

    International Nuclear Information System (INIS)

    Kannan, Umasankari

    2000-01-01

    The WIMS library has been extensively used in thermal reactor calculations. This multigroup constants library was originally developed from the UKNDL in the late 60's and has been updated in 1986. This library has been distributed with the WIMS-D code by NEA data bank. The references to WIMS library in literature are the 'old' which is the original as developed by the AEA Winfrith and the 'new' which is the current 1986 WIMS library. IAEA has organised a CRP where a new and fully updated WIMS library will soon be available. This paper gives an overview of the definitions of the group constants that go into any basic nuclear data library used for reactor calculations. This paper also outlines the contents of the WIMS library and some of its shortcomings

  15. Multi-group neutron transport theory

    International Nuclear Information System (INIS)

    Zelazny, R.; Kuszell, A.

    1962-01-01

    Multi-group neutron transport theory. In the paper the general theory of the application of the K. M. Case method to N-group neutron transport theory in plane geometry is given. The eigenfunctions (distributions) for the system of Boltzmann equations have been derived and the completeness theorem has been proved. By means of general solution two examples important for reactor and shielding calculations are given: the solution of a critical and albedo problem for a slab. In both cases the system of singular integral equations for expansion coefficients into a full set of eigenfunction distributions has been reduced to the system of Fredholm-type integral equations. Some results can be applied also to some spherical problems. (author) [fr

  16. Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, E. [Dept. of Mechanical Engineering, Univ. of Texas at Austin, 1 Univ. Place C2200, Austin, TX 78712 (United States); Deinert, M.; Bingham Cady, K. [Dept. of Theoretical and Applied Mechanics, Cornell Univ., Ithaca, NY 14853 (United States)

    2006-07-01

    Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)

  17. A multi-region boundary element method for multigroup neutron diffusion calculations

    International Nuclear Information System (INIS)

    Ozgener, H.A.; Ozgener, B.

    2001-01-01

    For the analysis of a two-dimensional nuclear system consisting of a number of homogeneous regions (termed cells), first the cell matrices which depend solely on the material composition and geometrical dimension of the cell (hence on the cell type) are constructed using a boundary element formulation based on the multigroup boundary integral equation. For a particular nuclear system, the cell matrices are utilized in the assembly of the global system matrix in block-banded form using the newly introduced concept of virtual side. For criticality calculations, the classical fission source iteration is employed and linear system solutions are by the block Gaussian-elimination algorithm. The numerical applications show the validity of the proposed formulation both through comparison with analytical solutions and assessment of benchmark problem results against alternative methods

  18. Assessment and comparison of different multigroup neutron cross section libraries for dosimetry purposes

    International Nuclear Information System (INIS)

    Erradi, L.; Karouani, K.

    1994-01-01

    Many multigroup neutron cross section libraries have been processed from basic evaluated nuclear data for use in neutron dosimetry, reactor shielding calculation and in the development of fusion reactors. Most of these libraries have been tested only for fission spectra and were not validated for fusion spectra. Fifteen of these libraries such as DOSCROS84, IRDF85 and ENDFB5 have been used along with the neutron spectra unfolding code SAND II to evaluate about fifteen threshold detector saturated activities. The comparison between these computed activities and the measured ones of a set of foils placed in different places along the axis of a paraffin cylinder and irradiated by 14 MeV neutrons generated by a D-T source, hence giving rise to complex spectra, leads to different types of discrepancies. The analysis of these discrepancies allows to select from these libraries the ones that can be recommended. 1 fig., 4 refs. (author)

  19. Fission meter

    Science.gov (United States)

    Rowland, Mark S [Alamo, CA; Snyderman, Neal J [Berkeley, CA

    2012-04-10

    A neutron detector system for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly into information that a first responder can use to discriminate materials. The system comprises counting neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source.

  20. Final report [on solving the multigroup diffusion equations

    International Nuclear Information System (INIS)

    Birkhoff, G.

    1975-01-01

    Progress achieved in the development of variational methods for solving the multigroup neutron diffusion equations is described. An appraisal is made of the extent to which improved variational methods could advantageously replace difference methods currently used

  1. Ideological Fission

    DEFF Research Database (Denmark)

    Christiansen, Steen Ledet

    ; it is a materialisation of an ideological fission which attempts to excise certain ideological constructions, yet paradoxically casting them in a form that is recognizable and familiar. The monstrous metonomy which is used shows us glimpses of a horrid being, intended to vilify the attack on New York City. However......, it is a being which is reminiscent of earlier monsters - from Godzilla to The Blob. It is evident that the Cloverfield monster is a paradoxical construction which attempts to articulate fear and loathing about terrorism, but ends up trapped in an ideological dead-end maze, unable to do anything other than...

  2. Mica fission detectors

    International Nuclear Information System (INIS)

    Wong, C.; Anderson, J.D.; Hansen, L.; Lehn, A.V.; Williamson, M.A.

    1977-01-01

    The present development status of the mica fission detectors is summarized. It is concluded that the techniques have been refined and developed to a state such that the mica fission counters are a reliable and reproducible detector for fission events

  3. On the convergence of multigroup discrete-ordinates approximations

    International Nuclear Information System (INIS)

    Victory, H.D. Jr.; Allen, E.J.; Ganguly, K.

    1987-01-01

    Our analysis is divided into two distinct parts which we label for convenience as Part A and Part B. In Part A, we demonstrate that the multigroup discrete-ordinates approximations are well-defined and converge to the exact transport solution in any subcritical setting. For the most part, we focus on transport in two-dimensional Cartesian geometry. A Nystroem technique is used to extend the discrete ordinates multigroup approximates to all values of the angular and energy variables. Such an extension enables us to employ collectively compact operator theory to deduce stability and convergence of the approximates. In Part B, we perform a thorough convergence analysis for the multigroup discrete-ordinates method for an anisotropically-scattering subcritical medium in slab geometry. The diamond-difference and step-characteristic spatial approximation methods are each studied. The multigroup neutron fluxes are shown to converge in a Banach space setting under realistic smoothness conditions on the solution. This is the first thorough convergence analysis for the fully-discretized multigroup neutron transport equations

  4. Fission product model for lattice calculation of high conversion boiling water reactor

    International Nuclear Information System (INIS)

    Iijima, S.; Yoshida, T.; Yamamoto, T.

    1988-01-01

    A high precision fission product model for boiling water reactor (BWR) lattice calculation was developed, which consists of 45 nuclides to be treated explicitly and one nonsaturating pseudo nuclide. This model is applied to a high conversion BWR lattice calculation code. From a study based on a three-energy-group calculation of fission product poisoning due to full fission products and explicitly treated nuclides, the multigroup capture cross sections and the effective fission yields of the pseudo nuclide are determined, which do not depend on fuel types or reactor operating conditions for a good approximation. Apart from nuclear data uncertainties, the model and the derived pseudo nuclide constants would predict the fission product reactivity within an error of 0.1% Δk at high burnup

  5. Fission theory and actinide fission data

    Energy Technology Data Exchange (ETDEWEB)

    Michaudon, A.

    1975-06-01

    The understanding of the fission process has made great progress recently, as a result of the calculation of fission barriers, using the Strutinsky prescription. Double-humped shapes were obtained for nuclei in the actinide region. Such shapes could explain, in a coherent manner, many different phenomena: fission isomers, structure in near-threshold fission cross sections, intermediate structure in subthreshold fission cross sections and anisotropy in the emission of the fission fragments. A brief review of fission barrier calculations and relevant experimental data is presented. Calculations of fission cross sections, using double-humped barrier shapes and fission channel properties, as obtained from the data discussed previously, are given for some U and Pu isotopes. The fission channel theory of A. Bohr has greatly influenced the study of low-energy fission. However, recent investigation of the yields of prompt neutrons and γ rays emitted in the resonances of {sup 235}U and {sup 239}Pu, together with the spin determination for many resonances of these two nuclei cannot be explained purely in terms of the Bohr theory. Variation in the prompt neutron and γ-ray yields from resonance to resonance does not seem to be due to such fission channels, as was thought previously, but to the effect of the (n,γf) reaction. The number of prompt fission neutrons and the kinetic energy of the fission fragments are affected by the energy balance and damping or viscosity effects in the last stage of the fission process, from saddle point to scission. These effects are discussed for some nuclei, especially for {sup 240}Pu.

  6. Multigroup Moderation Test in Generalized Structured Component Analysis

    Directory of Open Access Journals (Sweden)

    Angga Dwi Mulyanto

    2016-05-01

    Full Text Available Generalized Structured Component Analysis (GSCA is an alternative method in structural modeling using alternating least squares. GSCA can be used for the complex analysis including multigroup. GSCA can be run with a free software called GeSCA, but in GeSCA there is no multigroup moderation test to compare the effect between groups. In this research we propose to use the T test in PLS for testing moderation Multigroup on GSCA. T test only requires sample size, estimate path coefficient, and standard error of each group that are already available on the output of GeSCA and the formula is simple so the user does not need a long time for analysis.

  7. The Suppression of Energy Discretization Errors in Multigroup Transport Calculations

    International Nuclear Information System (INIS)

    Larsen, Edward

    2013-01-01

    The Objective of this project is to develop, implement, and test new deterministric methods to solve, as efficiently as possible, multigroup neutron transport problems having an extremely large number of groups. Our approach was to (i) use the standard CMFD method to 'coarsen' the space-angle grid, yielding a multigroup diffusion equation, and (ii) use a new multigrid-in-space-and-energy technique to efficiently solve the multigroup diffusion problem. The overall strategy of (i) how to coarsen the spatial an energy grids, and (ii) how to navigate through the various grids, has the goal of minimizing the overall computational effort. This approach yields not only the fine-grid solution, but also coarse-group flux-weighted cross sections that can be used for other related problems.

  8. Nuclear fission and reactions

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    The nuclear fission research programs are designed to elucidate basic features of the fission process. Specifically, (1) factors determining how nucleons of a fissioning nucleus are distributed between two fission fragments, (2) factors determining kinetic energy and excitation energies of fragments, and (3) factors controlling fission lifetimes. To these ends, fission studies are reported for several heavy elements and include investigations of spontaneous and neutron-induced fission, heavy ion reactions, and high energy proton reactions. The status of theoretical research is also discussed. (U.S.)

  9. WIMSD5, Deterministic Multigroup Reactor Lattice Calculations

    International Nuclear Information System (INIS)

    2004-01-01

    1 - Description of program or function: The Winfrith improved multigroup scheme (WIMS) is a general code for reactor lattice cell calculation on a wide range of reactor systems. In particular, the code will accept rod or plate fuel geometries in either regular arrays or in clusters and the energy group structure has been chosen primarily for thermal calculations. The basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups, but the user is offered the choice of accurate solutions in many groups or rapid calculations in few groups. Temperature dependent thermal scattering matrices for a variety of scattering laws are included in the library for the principal moderators which include hydrogen, deuterium, graphite, beryllium and oxygen. WIMSD5 is a successor version of WIMS-D/4. 2 - Method of solution: The treatment of resonances is based on the use of equivalence theorems with a library of accurately evaluated resonance integrals for equivalent homogeneous systems at a variety of temperatures. The collision theory procedure gives accurate spectrum computations in the 69 groups of the library for the principal regions of the lattice using a simplified geometric representation of complicated lattice cells. The computed spectra are then used for the condensation of cross-sections to the number of groups selected for solution of the transport equation in detailed geometry. Solution of the transport equation is provided either by use of the Carlson DSN method or by collision probability methods. Leakage calculations including an allowance for streaming asymmetries may be made using either diffusion theory or the more elaborate B1-method. The output of the code provides Eigenvalues for the cases where a simple buckling mode is applicable or cell-averaged parameters for use in overall reactor calculations. Various reaction rate edits are provided for direct comparison with experimental measurements. 3 - Restrictions on the complexity of

  10. Review of multigroup nuclear cross-section processing

    Energy Technology Data Exchange (ETDEWEB)

    Trubey, D.K.; Hendrickson, H.R. (comps.)

    1978-10-01

    These proceedings consist of 18 papers given at a seminar--workshop on ''Multigroup Nuclear Cross-Section Processing'' held at Oak Ridge, Tennessee, March 14--16, 1978. The papers describe various computer code systems and computing algorithms for producing multigroup neutron and gamma-ray cross sections from evaluated data, and experience with several reference data libraries. Separate abstracts were prepared for 13 of the papers. The remaining five have already been cited in ERA, and may be located by referring to the entry CONF-780334-- in the Report Number Index. (RWR)

  11. XNWLUP, Graphical user interface to plot WIMS-D library multigroup cross sections

    International Nuclear Information System (INIS)

    Ganesan, S.; Jagannathan, V.; Thiyagarajan, T.K.

    2005-01-01

    1 - Description of program or function: XnWlup is a computer program with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualisation of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. IAEA1395/05: New features of version 3.0: - Plotting absorption and fission cross sections of resonant nuclide after applying the self-shielding cross section. - Plotting the data of Resonant Integral table, as a function of dilution cross section for a selected temperature and for a given energy group. - Plotting the data of Resonant Integral table, as a function of temperature for a selected background dilution cross section and for a given energy group. - Clearing all the graphs except one graph from the display screen is easily done by using a tool bar button. - Displaying the coordinate of the cursor point with appropriate units. 2 - Methods: XnWlup helps to obtain histogram plots of the values of cross section data of an element/isotope available as 69-group WIMS-D library as a function of energy bins. The software XnWlup is developed with this graphical user interface in order to help those users who frequently refer to the WIMS-D library cross section data of neutron-nuclear reactions. The software also helps to produce handbook of WIMS-D cross sections

  12. Measurements of fission yields

    International Nuclear Information System (INIS)

    Denschlag, H.O.

    2000-01-01

    After some historical introductory remarks on the discovery of nuclear fission and early fission yield determinations, the present status of knowledge on fission yields is briefly reviewed. Practical and fundamental reasons motivating the pursuit of fission yield measurements in the coming century are pointed out. Recent results and novel techniques are described that promise to provide new interesting insights into the fission process during the next century. (author)

  13. Radiochemical studies on fission

    Energy Technology Data Exchange (ETDEWEB)

    None

    1973-07-01

    Research progress is reported on nuclear chemistry; topics considered include: recoil range and kinetic energy distribution in the thermal neutron ftssion of /sup 245/Cm; mass distribution and recoil range measurements in the reactor neutron-induced fission of /sup 232/U; fission yields in the thermal neutron fission of /sup 241/PU highly asymmetric binary fission of uranium induced by reactor neutrons; and nuclear charge distribution in low energy fission. ( DHM)

  14. FINELM: a multigroup finite element diffusion code. Part II

    International Nuclear Information System (INIS)

    Davierwalla, D.M.

    1981-05-01

    The author presents the axisymmetric case in cylindrical coordinates for the finite element multigroup neutron diffusion code, FINELM. The numerical acceleration schemes incorporated viz. the Lebedev extrapolations and the coarse mesh rebalancing, space collapsing, are discussed. A few benchmark computations are presented as validation of the code. (Auth.)

  15. RZ calculations for self shielded multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z. [Commissariat a l' Energie Atomique CEA, Direction de l' Energie Nucleaire, DEN/DM2S/SERMA/LENR, 91191 Gif-sur-Yvette Cedex (France)

    2006-07-01

    A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

  16. RZ calculations for self shielded multigroup cross sections

    International Nuclear Information System (INIS)

    Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z.

    2006-01-01

    A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

  17. Calculation of multigroup reaction rates for the Ghana Research ...

    African Journals Online (AJOL)

    The discrete ordinate spatial model, which pro-vides solution to the differential form of the transport equation by the Carlson-SN (N=4) approach was adopted to solve the Ludwig-Boltzmann multigroup neutron transport equation for this analysis. The results show that for any fissile resonance absorber, the reaction rates ...

  18. Fission products collecting devices

    International Nuclear Information System (INIS)

    Matsumoto, Hiroshi

    1979-01-01

    Purpose: To enable fission products trap with no contamination to coolants and cover gas by the provision of a fission products trap above the upper part of a nuclear power plant. Constitution: Upon fuel failures in a reactor core, nuclear fission products leak into coolants and move along the flow of the coolants to the coolants above the reactor core. The fission products are collected in a trap container and guided along a pipeline into fission products detector. The fission products detector monitors the concentration of the fission products and opens the downstream valve of the detector when a predetermined concentration of the fission products is detected to introduce the fission products into a waste gas processing device and release them through the exhaust pipe. (Seki, T.)

  19. Energy released in fission

    International Nuclear Information System (INIS)

    James, M.F.

    1969-05-01

    The effective energy released in and following the fission of U-235, Pu-239 and Pu-241 by thermal neutrons, and of U-238 by fission spectrum neutrons, is discussed. The recommended values are: U-235 ... 192.9 ± 0.5 MeV/fission; U-238 ... 193.9 ± 0.8 MeV/fission; Pu-239 ... 198.5 ± 0.8 MeV/fission; Pu-241 ... 200.3 ± 0.8 MeV/fission. These values include all contributions except from antineutrinos and very long-lived fission products. The detailed contributions are discussed, and inconsistencies in the experimental data are pointed out. In Appendix A, the contribution to the total useful energy release in a reactor from reactions other than fission are discussed briefly, and in Appendix B there is a discussion of the variations in effective energy from fission with incident neutron energy. (author)

  20. HAC and fission reactors

    International Nuclear Information System (INIS)

    Fujiwara, I.; Moriyama, H.; Tachikawa, E.

    1984-01-01

    In the fission process, newly formed fission products undergo hot atom reactions due to their energetic recoil and abnormal positive charge. The hot atom reactions of the fission products are usually accompanied by secondary effects such as radiation damage, especially in condensed phase. For reactor safety it is valuable to know the chemical behaviour and the release behaviour of these radioactive fission products. Here, the authors study the chemical behaviour and the release behaviour of the fission products from the viewpoint of hot atom chemistry (HAC). They analyze the experimental results concerning fission product behaviour with the help of the theories in HAC and other neighboring fields such as radiation chemistry. (Auth.)

  1. Investigations of safety-related parameters applying a new multi-group diffusion code for HTR transients

    International Nuclear Information System (INIS)

    Kasselmann, S.; Druska, C.; Lauer, A.

    2010-01-01

    The energy spectra of fast and thermal neutrons from fission reactions in the FZJ code TINTE are modelled by two broad energy groups. Present demands for increased numerical accuracy led to the question of how precise the 2-group approximation is compared to a multi-group model. Therefore a new simulation program called MGT (Multi Group TINTE) has recently been developed which is able to handle up to 43 energy groups. Furthermore, an internal spectrum calculation for the determination of cross-sections can be performed for each time step and location within the reactor. In this study the multi-group energy models are compared to former calculations with only two energy groups. Different scenarios (normal operation and design-basis accidents) have been defined for a high temperature pebble bed reactor design with annular core. The effect of an increasing number of energy groups on safety-related parameters like the fuel and coolant temperature, the nuclear heat source or the xenon concentration is studied. It has been found that for the studied scenarios the use of up to 8 energy groups is a good trade-off between precision and a tolerable amount of computing time. (orig.)

  2. Methodological study of the adjustment of multigroup cross sections for fast reactors, by means of integral data

    International Nuclear Information System (INIS)

    Bastos, H.F.B.N.

    1979-01-01

    In this work a study of the methodology of the adjustment of multigroup cross sections by means of integral data is presented. A synthesis of the principal methods existent and the mathematical development of the adaptation of one of them are made. A calculational system is built from this reference method, with the basic conditions for the operation of the process of adjustment. In order to test the system developed and analyze several problems related to the adjustment, a series of trial adjustments was made with the value of the U 235 fission cross section from the infinite dilution library used in the calculational system for fast reactors of the Instituto de Engenharia Nuclear. (author)

  3. HLW disposal by fission reactors; calculation of trans-mutation rate and recycle

    International Nuclear Information System (INIS)

    Mulyanto

    1997-01-01

    Transmutation of MA (Minor actinide) and LLFPS (long-lived fission products) into stable nuclide or short-lived isotopes by fission reactors seem to become an alternative technology for HLW disposal. in this study, transmutation rate and recycle calculation were developed in order to evaluate transmutation characteristics of MA and LLFPs in the fission reactors. inventory of MA and LLFPs in the transmutation reactors were determined by solving of criticality equation with 1-D cylindrical geometry of multigroup diffusion equations at the beginning of cycle (BOC). transmutation rate and burn-up was determined by solving of depletion equation. inventory of MA and LLFPs was calculated for 40 years recycle. From this study, it was concluded that characteristics of MA and LLFPs in the transmutation reactors can be evaluated by recycle calculation. by calculation of transmutation rate, performance of fission reactor for transmutation of MA or LLFPs can be discussed

  4. Sub-cell balanced nodal expansion methods using S4 eigenfunctions for multi-group SN transport problems in slab geometry

    International Nuclear Information System (INIS)

    Hong, Ser Gi; Lee, Deokjung

    2015-01-01

    A highly accurate S 4 eigenfunction-based nodal method has been developed to solve multi-group discrete ordinate neutral particle transport problems with a linearly anisotropic scattering in slab geometry. The new method solves the even-parity form of discrete ordinates transport equation with an arbitrary S N order angular quadrature using two sub-cell balance equations and the S 4 eigenfunctions of within-group transport equation. The four eigenfunctions from S 4 approximation have been chosen as basis functions for the spatial expansion of the angular flux in each mesh. The constant and cubic polynomial approximations are adopted for the scattering source terms from other energy groups and fission source. A nodal method using the conventional polynomial expansion and the sub-cell balances was also developed to be used for demonstrating the high accuracy of the new methods. Using the new methods, a multi-group eigenvalue problem has been solved as well as fixed source problems. The numerical test results of one-group problem show that the new method has third-order accuracy as mesh size is finely refined and it has much higher accuracies for large meshes than the diamond differencing method and the nodal method using sub-cell balances and polynomial expansion of angular flux. For multi-group problems including eigenvalue problem, it was demonstrated that the new method using the cubic polynomial approximation of the sources could produce very accurate solutions even with large mesh sizes. (author)

  5. Fission Research at IRMM

    Directory of Open Access Journals (Sweden)

    Al-Adili A.

    2010-03-01

    Full Text Available Fission Research at JRC-IRMM has a longstanding tradition. The present paper is discussing recent investigations of fission fragment properties of 238 U(n,f, 234 U(n,f, prompt neutron emission in fission of 252 Cf(SF as well as the prompt fission neutron spectrum of 235 U(n,f and is presenting the most important results.

  6. Equilibrium fission model calculations

    International Nuclear Information System (INIS)

    Beckerman, M.; Blann, M.

    1976-01-01

    In order to aid in understanding the systematics of heavy ion fission and fission-like reactions in terms of the target-projectile system, bombarding energy and angular momentum, fission widths are calculated using an angular momentum dependent extension of the Bohr-Wheeler theory and particle emission widths using angular momentum coupling

  7. Optimization of multi-group cross sections for fast reactor analysis

    International Nuclear Information System (INIS)

    Chin, M. R.; Manalo, K. L.; Edgar, C. A.; Paul, J. N.; Molinar, M. P.; Redd, E. M.; Yi, C.; Sjoden, G. E.

    2013-01-01

    The selection of the number of broad energy groups, collapsed broad energy group boundaries, and their associated evaluation into collapsed macroscopic cross sections from a general 238-group ENDF/B-VII library dramatically impacted the k eigenvalue for fast reactor analysis. An analysis was undertaken to assess the minimum number of energy groups that would preserve problem physics; this involved studies using the 3D deterministic transport parallel code PENTRAN, the 2D deterministic transport code SCALE6.1, the Monte Carlo based MCNP5 code, and the YGROUP cross section collapsing tool on a spatially discretized MOX fuel pin comprised of 21% PUO 2 -UO 2 with sodium coolant. The various cases resulted in a few hundred pcm difference between cross section libraries that included the 238 multi-group reference, and cross sections rendered using various reaction and adjoint weighted cross sections rendered by the YGROUP tool, and a reference continuous energy MCNP case. Particular emphasis was placed on the higher energies characteristic of fission neutrons in a fast spectrum; adjoint computations were performed to determine the average per-group adjoint fission importance for the MOX fuel pin. This study concluded that at least 10 energy groups for neutron transport calculations are required to accurately predict the eigenvalue for a fast reactor system to within 250 pcm of the 238 group case. In addition, the cross section collapsing/weighting schemes within YGROUP that provided a collapsed library rendering eigenvalues closest to the reference were the contribution collapsed, reaction rate weighted scheme. A brief analysis on homogenization of the MOX fuel pin is also provided, although more work is in progress in this area. (authors)

  8. PUFF-III: A Code for Processing ENDF Uncertainty Data Into Multigroup Covariance Matrices

    International Nuclear Information System (INIS)

    Dunn, M.E.

    2000-01-01

    PUFF-III is an extension of the previous PUFF-II code that was developed in the 1970s and early 1980s. The PUFF codes process the Evaluated Nuclear Data File (ENDF) covariance data and generate multigroup covariance matrices on a user-specified energy grid structure. Unlike its predecessor, PUFF-III can process the new ENDF/B-VI data formats. In particular, PUFF-III has the capability to process the spontaneous fission covariances for fission neutron multiplicity. With regard to the covariance data in File 33 of the ENDF system, PUFF-III has the capability to process short-range variance formats, as well as the lumped reaction covariance data formats that were introduced in ENDF/B-V. In addition to the new ENDF formats, a new directory feature is now available that allows the user to obtain a detailed directory of the uncertainty information in the data files without visually inspecting the ENDF data. Following the correlation matrix calculation, PUFF-III also evaluates the eigenvalues of each correlation matrix and tests each matrix for positive definiteness. Additional new features are discussed in the manual. PUFF-III has been developed for implementation in the AMPX code system, and several modifications were incorporated to improve memory allocation tasks and input/output operations. Consequently, the resulting code has a structure that is similar to other modules in the AMPX code system. With the release of PUFF-III, a new and improved covariance processing code is available to process ENDF covariance formats through Version VI

  9. Fission gas behavior during fast thermal transients

    International Nuclear Information System (INIS)

    Esteves, R.G.

    1976-01-01

    The behavior of non-equilibrium fission in fuel elements undergoing fast thermal transients is analyzed. To facilitate the analysis, a new variable, the equilibrium variable (EV) is defined. This variable, together with bubble radius, completely specifies a bubble with respect to its size and equilibrium condition. The analysis is coded using a two-variable (radius and EV) multigroup numerical approximation that accepts as input the time-temperature history, the time-fission rate history, and the time-thermal gradient history of the fuel element. Studies were performed to test the code for convergence with respect to the time interval and the number of groups chosen. For a series of transient simulation studies, the measurements obtained at HEDL (microscopic examination of intragranular porosity in oxide fuel transient-tested in TREAT) are used. Two different transient histories were selected; the first, a high-temperature transient (HTT) with a peak at 2477 0 K and the second, a low-temperature transient (LTT) with a peak-temperature at 2000 0 K. The LTT was simulated for three different conditions: Bubbles were allowed to move via (a) only biased migration, (b) via random migration, and (c) via both mechanisms. The HTT was also run for both mechanisms. The agreement with HEDL microscopic observations was fair for bubbles smaller than 964 A in diameter, and poor for larger bubbles. Bubbles that grew during the heat-up part of the transient were frozen at a larger size during the cool down

  10. EPRI-LATTICE: a multigroup neutron transport code for light water reactor lattice physics calculations

    International Nuclear Information System (INIS)

    Jones, D.B.

    1986-01-01

    EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated

  11. Cyclotron radiation by a multi-group method

    International Nuclear Information System (INIS)

    Chu, T.C.

    1980-01-01

    A multi-energy group technique is developed to study conditions under which cyclotron radiation emission can shift a Maxwellian electron distribution into a non-Maxwellian; and if the electron distribution is non-Maxwellian, to study the rate of cyclotron radiation emission as compared to that emitted by a Maxwellian having the same mean electron density and energy. The assumptions in this study are: the electrons should be in an isotropic medium and the magnetic field should be uniform. The multi-group technique is coupled into a multi-group Fokker-Planck computer code to study electron behavior under the influence of cyclotron radiation emission in a self-consistent fashion. Several non-Maxwellian distributions were simulated to compare their cyclotron emissions with the corresponding energy and number density equivalent Maxwellian distribtions

  12. The Multigroup Neutron Diffusion Equations/1 Space Dimension

    Energy Technology Data Exchange (ETDEWEB)

    Linde, Sven

    1960-06-15

    A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix.

  13. Nuclear data processing and multigroup cross section generation

    International Nuclear Information System (INIS)

    Kim, Jeong Do; Kil, Chung Sub

    1996-01-01

    The multigroup constants for WIMS/CASMO were updated with ENDF/B-VI and were tested. The continuous energy MCNP library developed last year was validated against the LWR-simulated critical experiments. The MCNP library will be used to design and analyze nuclear and shielding facilities. The system for generation of MATXS multigroup library and TRANSX code, which is able to prepare the data for the discrete ordinates and diffusion codes from the MATXS library, was established. The MATXS libraries for analyses of thermal and fast critical experiments were generated and tested. The MATXS/TRANSX system for the discrete ordinates and diffusion codes will be useful for nuclear analyses. 10 tabs., 5 figs., 17 refs. (Author)

  14. The Multigroup Neutron Diffusion Equations/1 Space Dimension

    International Nuclear Information System (INIS)

    Linde, Sven

    1960-06-01

    A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix

  15. Scalable Multi-group Key Management for Advanced Metering Infrastructure

    OpenAIRE

    Benmalek , Mourad; Challal , Yacine; Bouabdallah , Abdelmadjid

    2015-01-01

    International audience; Advanced Metering Infrastructure (AMI) is composed of systems and networks to incorporate changes for modernizing the electricity grid, reduce peak loads, and meet energy efficiency targets. AMI is a privileged target for security attacks with potentially great damage against infrastructures and privacy. For this reason, Key Management has been identified as one of the most challenging topics in AMI development. In this paper, we propose a new Scalable multi-group key ...

  16. Optimal calculational schemes for solving multigroup photon transport problem

    International Nuclear Information System (INIS)

    Dubinin, A.A.; Kurachenko, Yu.A.

    1987-01-01

    A scheme of complex algorithm for solving multigroup equation of radiation transport is suggested. The algorithm is based on using the method of successive collisions, the method of forward scattering and the spherical harmonics method, and is realized in the FORAP program (FORTRAN, BESM-6 computer). As an example the results of calculating reactor photon transport in water are presented. The considered algorithm being modified may be used for solving neutron transport problems

  17. Multi-group diffusion perturbation calculation code. PERKY (2002)

    Energy Technology Data Exchange (ETDEWEB)

    Iijima, Susumu; Okajima, Shigeaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Perturbation calculation code based on the diffusion theory ''PERKY'' is designed for nuclear characteristic analyses of fast reactor. The code calculates reactivity worth on the multi-group diffusion perturbation theory in two or three dimensional core model and kinetics parameters such as effective delayed neutron fraction, prompt neutron lifetime and absolute reactivity scale factor ({rho}{sub 0} {delta}k/k) for FCA experiments. (author)

  18. Multigroup or multipoint thermal neutron data preparation. Programme SIGMA

    International Nuclear Information System (INIS)

    Matausek, M.V.; Kunc, M.

    1974-01-01

    When calculating the space energy distribution of thermal neutrons in reactor lattices, in either the multigroup or the multipoint approximation, it is convenient to divide the problem into two independent parts. Firstly, for all material regions of the given reactor lattice cell, the group or the point values of cross sections, scattering kernel and the outer source of thermal neutrons are calculated by a data preparation programme. These quantities are then used as input, by the programme which solves multigroup or multipoint transport equations, to generate the space energy neutron spectra in the cell considered and to determine the related integral quantities, namely the different reaction rates. The present report deals with the first part of the problem. An algorithm for constructing a set of thermal neutron input data, to be used with the multigroup or multipoint version of the code MULTI /1,2,3/, is presented and the new version of the programme SIGMA /4/, written in FORTRAN IV for the CDC-3600 computer, is described. For a given reactor cell material, composed of a number of different isotopes, this programme calculates the group or the point values of the scattering macroscopic absorption cross section, macroscopic scattering cross section, kernel and the outer source of thermal neutrons. Numerous options are foreseen in the programme, concerning the energy variation of cross sections and a scattering kernel, concerning the weighting spectrum in multigroup scheme or the procedure for constructing the scattering matrix in the multipoint scheme and, finally, concerning the organization of output. The details of the calculational algorithm are presented in Section 2 of the paper. Section 3 contains the description of the programme and the instructions for its use (author)

  19. CASTRO: A NEW COMPRESSIBLE ASTROPHYSICAL SOLVER. III. MULTIGROUP RADIATION HYDRODYNAMICS

    International Nuclear Information System (INIS)

    Zhang, W.; Almgren, A.; Bell, J.; Howell, L.; Burrows, A.; Dolence, J.

    2013-01-01

    We present a formulation for multigroup radiation hydrodynamics that is correct to order O(v/c) using the comoving-frame approach and the flux-limited diffusion approximation. We describe a numerical algorithm for solving the system, implemented in the compressible astrophysics code, CASTRO. CASTRO uses a Eulerian grid with block-structured adaptive mesh refinement based on a nested hierarchy of logically rectangular variable-sized grids with simultaneous refinement in both space and time. In our multigroup radiation solver, the system is split into three parts: one part that couples the radiation and fluid in a hyperbolic subsystem, another part that advects the radiation in frequency space, and a parabolic part that evolves radiation diffusion and source-sink terms. The hyperbolic subsystem and the frequency space advection are solved explicitly with high-order Godunov schemes, whereas the parabolic part is solved implicitly with a first-order backward Euler method. Our multigroup radiation solver works for both neutrino and photon radiation.

  20. Nuclear data and multigroup methods in fast reactor calculations

    International Nuclear Information System (INIS)

    Gur, Y.

    1975-03-01

    The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)

  1. A numerical method for multigroup slab-geometry discrete ordinates problems with no spatial truncation error

    International Nuclear Information System (INIS)

    Barros, R.C. de; Larsen, E.W.

    1991-01-01

    A generalization of the one-group Spectral Green's Function (SGF) method is developed for multigroup, slab-geometry discrete ordinates (S N ) problems. The multigroup SGF method is free from spatial truncation errors; it generated numerical values for the cell-edge and cell-average angular fluxes that agree with the analytic solution of the multigroup S N equations. Numerical results are given to illustrate the method's accuracy

  2. Fission level densities

    International Nuclear Information System (INIS)

    Maslov, V.M.

    1998-01-01

    Fission level densities (or fissioning nucleus level densities at fission saddle deformations) are required for statistical model calculations of actinide fission cross sections. Back-shifted Fermi-Gas Model, Constant Temperature Model and Generalized Superfluid Model (GSM) are widely used for the description of level densities at stable deformations. These models provide approximately identical level density description at excitations close to the neutron binding energy. It is at low excitation energies that they are discrepant, while this energy region is crucial for fission cross section calculations. A drawback of back-shifted Fermi gas model and traditional constant temperature model approaches is that it is difficult to include in a consistent way pair correlations, collective effects and shell effects. Pair, shell and collective properties of nucleus do not reduce just to the renormalization of level density parameter a, but influence the energy dependence of level densities. These effects turn out to be important because they seem to depend upon deformation of either equilibrium or saddle-point. These effects are easily introduced within GSM approach. Fission barriers are another key ingredients involved in the fission cross section calculations. Fission level density and barrier parameters are strongly interdependent. This is the reason for including fission barrier parameters along with the fission level densities in the Starter File. The recommended file is maslov.dat - fission barrier parameters. Recent version of actinide fission barrier data obtained in Obninsk (obninsk.dat) should only be considered as a guide for selection of initial parameters. These data are included in the Starter File, together with the fission barrier parameters recommended by CNDC (beijing.dat), for completeness. (author)

  3. Sensitivity of 238U resonance absorption to library multigroup structure as calculated by WIMS-AECL

    International Nuclear Information System (INIS)

    Laughton, P.J.; Donnelly, J.V.

    1995-01-01

    In simulations of the TRX-1 experimental lattice, WIMS-AECL overpredicts, relative to MCNP, resonance absorption in neutron-energy groups containing the three large, low-lying resonances of 238 U when a standard ENDF/B-V-based library is used. A total excess in these groups of 4.0 neutron captures by 238 U per thousand fission neutrons has been observed. Similar comparisons are made in this work for the MIT-4 experimental lattice and simplified CANDU lattice cells containing 37-element fuel, with and without heavy-water coolant. Eleven different 89-group cross-section libraries were constructed for WIMS-AECL from ENDF/B-V data: only the neutron-energy-group boundaries used in generating multigroup cross sections and the Goldstein-Cohen correction factors differ from one library to the next. The first library uses the original 89-group structure, and the other ten involve energy groups of varying widths centred on the three large, low-lying resonances of 238 U. For TRX-1, some reduction in total discrepancy in 238 U capture can be achieved by using a new structure, although the improvement is small. The discrepancies in 238 U capture are of the same order for the MIT-4 case as those observed for TRX-1 for both the original group structure and the ten new structures. The WIMS-AECL calculation of 238 U resonance absorption in the same ranges of energy for the simplified CANDU 37-element lattice are in better agreement with MCNP than they are for TRX-1 and MIT-4: when the original structure is used, WIMS-AECL underpredicts total capture rate by 238 U in the energy range of interest by only 0.56 per thousand fission neutrons (coolant present) and 0.88 per thousand fission neutrons (voided coolant channel). The discrepancies are reduced when some of the new structures are used. For almost all of the cases considered here-TRX-1, MIT-4 and CANDU with coolant-better group-by-group agreement of 238 U capture around the 6.67-eV resonance is achieved by using a new library

  4. Multi-Group Covariance Data Generation from Continuous-Energy Monte Carlo Transport Calculations

    International Nuclear Information System (INIS)

    Lee, Dong Hyuk; Shim, Hyung Jin

    2015-01-01

    The sensitivity and uncertainty (S/U) methodology in deterministic tools has been utilized for quantifying uncertainties of nuclear design parameters induced by those of nuclear data. The S/U analyses which are based on multi-group cross sections can be conducted by an simple error propagation formula with the sensitivities of nuclear design parameters to multi-group cross sections and the covariance of multi-group cross section. The multi-group covariance data required for S/U analysis have been produced by nuclear data processing codes such as ERRORJ or PUFF from the covariance data in evaluated nuclear data files. However in the existing nuclear data processing codes, an asymptotic neutron flux energy spectrum, not the exact one, has been applied to the multi-group covariance generation since the flux spectrum is unknown before the neutron transport calculation. It can cause an inconsistency between the sensitivity profiles and the covariance data of multi-group cross section especially in resolved resonance energy region, because the sensitivities we usually use are resonance self-shielded while the multi-group cross sections produced from an asymptotic flux spectrum are infinitely-diluted. In order to calculate the multi-group covariance estimation in the ongoing MC simulation, mathematical derivations for converting the double integration equation into a single one by utilizing sampling method have been introduced along with the procedure of multi-group covariance tally

  5. Fast fission phenomena

    International Nuclear Information System (INIS)

    Gregoire, Christian.

    1982-03-01

    Experimental studies of fast fission phenomena are presented. The paper is divided into three parts. In the first part, problems associated with fast fission processes are examined in terms of interaction potentials and a dynamic model is presented in which highly elastic collisions, the formation of compound nuclei and fast fission appear naturally. In the second part, a description is given of the experimental methods employed, the observations made and the preliminary interpretation of measurements suggesting the occurence of fast fission processes. In the third part, our dynamic model is incorporated in a general theory of the dissipative processes studied. This theory enables fluctuations associated with collective variables to be calculated. It is applied to highly inelastic collisions, to fast fission and to the fission dynamics of compound nuclei (for which a schematic representation is given). It is with these calculations that the main results of the second part can be interpreted [fr

  6. Fission product yields

    International Nuclear Information System (INIS)

    Valenta, V.; Hep, J.

    1978-01-01

    Data are summed up necessary for determining the yields of individual fission products from different fissionable nuclides. Fractional independent yields, cumulative and isobaric yields are presented here for the thermal fission of 235 U, 239 Pu, 241 Pu and for fast fission (approximately 1 MeV) of 235 U, 238 U, 239 Pu, 241 Pu; these values are included into the 5th version of the YIELDS library, supplementing the BIBFP library. A comparison is made of experimental data and possible improvements of calculational methods are suggested. (author)

  7. Fission neutron multiplicity calculations

    International Nuclear Information System (INIS)

    Maerten, H.; Ruben, A.; Seeliger, D.

    1991-01-01

    A model for calculating neutron multiplicities in nuclear fission is presented. It is based on the solution of the energy partition problem as function of mass asymmetry within a phenomenological approach including temperature-dependent microscopic energies. Nuclear structure effects on fragment de-excitation, which influence neutron multiplicities, are discussed. Temperature effects on microscopic energy play an important role in induced fission reactions. Calculated results are presented for various fission reactions induced by neutrons. Data cover the incident energy range 0-20 MeV, i.e. multiple chance fission is considered. (author). 28 refs, 13 figs

  8. Intermediate energy nuclear fission

    International Nuclear Information System (INIS)

    Hylten, G.

    1982-01-01

    Nuclear fission has been investigated with the double-kinetic-energy method using silicon surface barrier detectors. Fragment energy correlation measurements have been made for U, Th and Bi with bremsstrahlung of 600 MeV maximum energy. Distributions of kinetic energy as a function of fragment mass are presented. The results are compared with earlier photofission data and in the case of bismuth, with calculations based on the liquid drop model. The binary fission process in U, Yb, Tb, Ce, La, Sb, Ag and Y induced by 600 MeV protons has been investigated yielding fission cross sections, fragment kinetic energies, angular correlations and mass distributions. Fission-spallation competition calculations are used to deduce values of macroscopic fission barrier heights and nuclear level density parameter values at deformations corresponding to the saddle point shapes. We find macroscopic fission barriers lower than those predicted by macroscopic theories. No indication is found of the Businaro Gallone limit expected to occur somewhere in the mass range A = 100 to A = 140. For Ce and La asymmetric mass distributions similar to those in the actinide region are found. A method is described for the analysis of angular correlations between complementary fission products. The description is mainly concerned with fission induced by medium-energy protons but is applicable also to other projectiles and energies. It is shown that the momentum and excitation energy distributions of cascade residuals leading to fission can be extracted. (Author)

  9. Transmutation analysis considering and explicit fission product treatment based on a coupled Hammer-Technion and Cinder-2 system

    International Nuclear Information System (INIS)

    Abe, A.Y.

    1989-01-01

    This work presents a study about neutron absorption in a typical PWR cell by considering an explicit treatment for the fission products. The proposed methodology to treat fission product neutron absorption in a lattice calculation combines the HAMMER-TECHNION and CINDER-2 codes. The fission product chain treatment considers nearly 99% of all original CINDER-2 neutron absorption chain treatment. Parallel to the explicit treatment, a cross section library in the HAMMER-TECHNION code multigroup structure for the fission products was generated using the ENDF/B-V fission product library and processed by NJOY and AMPX-II processing codes. The methodology validation was investigated against two available benchmarks and it was obtained excellent results for the K-Infinity (IAEA-TECDOC-233) as function of burnup and enrichment and for the aggregate quantity sup(σ)2200 in units of barns/fission cross sections (OKAZAKI and SOKOLOWSKI). This work contributed for a better understanding of the fission product neutron absorption in a typical PWR cell and showed that the explicit fission product treatment can be successfully achieved. Besides that the performance of the ENDF/B-V fission product library was accessed. (author)

  10. From Fourier Transforms to Singular Eigenfunctions for Multigroup Transport

    International Nuclear Information System (INIS)

    Ganapol, B.D.

    2001-01-01

    A new Fourier transform approach to the solution of the multigroup transport equation with anisotropic scattering and isotropic source is presented. Through routine analytical continuation, the inversion contour is shifted from the real line to produce contributions from the poles and cuts in the complex plane. The integrand along the branch cut is then recast in terms of matrix continuum singular eigenfunctions, demonstrating equivalence of Fourier transform inversion and the singular eigenfunction expansion. The significance of this paper is that it represents the initial step in revealing the intimate connection between the Fourier transform and singular eigenfunction approaches as well as serves as a basis for a numerical algorithm

  11. Status of multigroup cross-section data for shielding applications

    International Nuclear Information System (INIS)

    Roussin, R.W.; Maskewitz, B.F.; Trubey, D.K.

    1983-01-01

    Multigroup cross-section libraries for shielding applications in formats for direct use in discrete ordinates or Monte Carlo codes have long been a part of the Data Library Collection (DLC) of the Radiation Shielding Information Center (RSIC). In recent years libraries in more flexible and comprehensive formats, which allow the user to derive his own problem-dependent sets, have been added to the collection. The current status of both types is described, as well as projections for adding data libraries based on ENDF/B-V

  12. Adjustement of multigroup cross sections using fast reactor integral data

    International Nuclear Information System (INIS)

    Renke, C.A.C.

    1982-01-01

    A methodology for the adjustment of multigroup cross section is presented, structured with aiming to compatibility the limitated number of measured values of integral parameters known and disponible, and the great number of cross sections to be adjusted the group of cross section used is that obtained from the Carnaval II calculation system, understanding as formular the sets of calculation methods and data bases. The adjustment is realized, using the INCOAJ computer code, developed in function of one statistical formulation, structural from the bayer considerations, taking in account the measurement processes of cross section and integral parameters defined on statistical bases. (E.G.) [pt

  13. Status of recent fast capture cross section evaluations for important fission product nuclides

    International Nuclear Information System (INIS)

    Gruppelaar, H.

    1982-01-01

    A comparison is made between recent evaluations of fission-product cross sections as given in the CNEN/CEA, ENDF/B-IV, ENDF/V-V, JENDL-1, RCN-2 and RCN-3 data libraries. The intercomparison is restricted to 24 important fission products in a fast power reactor. The evaluation methods used to obtain the various data files are reviewed and possible shortcomings are indicated. A survey is given of the experimental data based used in the various evaluations. Some graphs are included showing the new ENDF/B-V and RCN-3 fastcapture cross-section evaluations. Further intercomparisons are made by means of multi-group and one-group cross sections. It is shown that lumped fission-product cross sections calculated from the most recent versions of the data files are in quite good agreement with each other. This review concludes with a discussion on observed discrepancies and requests for new measurements. 78 references

  14. The nuclear fission

    International Nuclear Information System (INIS)

    Fiorentino, J.

    1983-01-01

    The nuclear fission process considering initially the formation of compound nucleus and finishing with radioactive decay of fission products is studied. The process is divided in three parts which consist of the events associated to the nucleus of intermediate transitional state, the scission configuration, and the phenomenum of post scission. (M.C.K.) [pt

  15. Fission gas detection system

    International Nuclear Information System (INIS)

    Colburn, R.P.

    1984-01-01

    A device for collecting fission gas released by failed fuel rods which device uses a filter adapted to pass coolant but to block passage of fission gas bubbles due to the surface tension of the bubbles. The coolant may be liquid metal. (author)

  16. Muon-induced fission

    International Nuclear Information System (INIS)

    Polikanov, S.

    1980-01-01

    A review of recent experimental results on negative-muon-induced fission, both of 238 U and 232 Th, is given. Some conclusions drawn by the author are concerned with muonic atoms of fission fragments and muonic atoms of the shape isomer of 238 U. (author)

  17. Relativistic Coulomb Fission

    Science.gov (United States)

    Norbury, John W.

    1992-01-01

    Nuclear fission reactions induced by the electromagnetic field of relativistic nuclei are studied for energies relevant to present and future relativistic heavy ion accelerators. Cross sections are calculated for U-238 and Pu-239 fission induced by C-12, Si-28, Au-197, and U-238 projectiles. It is found that some of the cross sections can exceed 10 b.

  18. SERKON program for compiling a multigroup library to be used in BETTY calculation

    International Nuclear Information System (INIS)

    Nguyen Phuoc Lan.

    1982-11-01

    A SERKON-type program was written to compile data sets generated by FEDGROUP-3 into a multigroup library for BETTY calculation. A multigroup library was generated from the ENDF/B-IV data file and tested against the TRX-1 and TRX-2 lattices with good results. (author)

  19. MUXS: a code to generate multigroup cross sections for sputtering calculations

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

    1982-10-01

    This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc

  20. Study of hypernuclei fission

    International Nuclear Information System (INIS)

    Malek, F.

    1990-01-01

    This work is about PS177 experience made on LEAR machine at CERN in 1988. The annihilation reaction of anti protons on a target of Bismuth or Uranium is studied. Lambda particles are produced by this reaction, in the nucleus in 2% of cases 7.1 10 -3 hypernuclei by stopped antiproton in the target are produced. The prompt hypernucleus fission probability of uranium is 75% and that of Bismuth 10%. The mass distribution of fission fragments is symmetrical ((≡ the excitation energy of the nucleus is very high). If the nucleus hasn't fissioned, the non-mesonic lambda decay, gives it an energy of 100 MeV, what allows to fission later. This fission is delayed because the hypernucleus lifetime is 1.3 +0.25 -0.21 10 -10 sec for Bismuth [fr

  1. The nuclear fission process

    International Nuclear Information System (INIS)

    Wagemans, C.

    1991-01-01

    Fifty years after its discovery, the nuclear fission phenomenon is of recurring interest. When its fundamental physics aspects are considered, fission is viewed in a very positive way, which is reflected in the great interest generated by the meetings and large conferences organized for the 50th anniversary of its discovery. From a purely scientific and practical point of view, a new book devoted to the (low energy) nuclear fission phenomenon was highly desirable considering the tremendous amount of new results obtained since the publication of the book Nuclear Fission by Vandenbosch and Huizenga in 1973 (Academic Press). These new results could be obtained thanks to the growth of technology, which enabled the construction of powerful new neutron sources, particle and heavy ion accelerators, and very performant data-acquisition and computer systems. The re-invention of the ionization chamber, the development of large fission fragment spectrometers and sophisticated multiparameter devices, and the production of exotic isotopes also contributed significantly to an improved understanding of nuclear fission. This book is written at a level to introduce graduate students to the exciting subject of nuclear fission. The very complete list of references following each chapter also makes the book very useful for scientists, especially nuclear physicists. The book has 12 chapters covering the fission barrier and the various processes leading to fission as well as the characteristics of the various fission reaction products. In order to guarantee adequate treatment of the very specialized research fields covered, several distinguished scientists actively involved in some of these fields were invited to contribute their expertise as authors or co-authors of the different chapters

  2. Semi-continuous and multigroup models in extended kinetic theory

    International Nuclear Information System (INIS)

    Koller, W.

    2000-01-01

    The aim of this thesis is to study energy discretization of the Boltzmann equation in the framework of extended kinetic theory. In case that external fields can be neglected, the semi- continuous Boltzmann equation yields a sound basis for various generalizations. Semi-continuous kinetic equations describing a three component gas mixture interacting with monochromatic photons as well as a four component gas mixture undergoing chemical reactions are established and investigated. These equations reflect all major aspects (conservation laws, equilibria, H-theorem) of the full continuous kinetic description. For the treatment of the spatial dependence, an expansion of the distribution function in terms of Legendre polynomials is carried out. An implicit finite differencing scheme is combined with the operator splitting method. The obtained numerical schemes are applied to the space homogeneous study of binary chemical reactions and to spatially one-dimensional laser-induced acoustic waves. In the presence of external fields, the developed overlapping multigroup approach (with the spline-interpolation as its extension) is well suited for numerical studies. Furthermore, two formulations of consistent multigroup approaches to the non-linear Boltzmann equation are presented. (author)

  3. Fission 2009 4. International Workshop on Nuclear Fission and Fission Product Spectroscopy - Compilation of slides

    International Nuclear Information System (INIS)

    2009-01-01

    This conference is dedicated to the last achievements in experimental and theoretical aspects of the nuclear fission process. The topics include: mass, charge and energy distribution, dynamical aspect of the fission process, nuclear data evaluation, quasi-fission and fission lifetime in super heavy elements, fission fragment spectroscopy, cross-section and fission barrier, and neutron and gamma emission. This document gathers the program of the conference and the slides of the presentations

  4. Solution of the multigroup neutron diffusion Eigenvalue problem in slab geometry by modified power method

    Energy Technology Data Exchange (ETDEWEB)

    Zanette, Rodrigo [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pós-Graduação em Matemática Aplicada; Petersen, Claudio Z.; Tavares, Matheus G., E-mail: rodrigozanette@hotmail.com, E-mail: claudiopetersen@yahoo.com.br, E-mail: matheus.gulartetavares@gmail.com [Universidade Federal de Pelotas (UFPEL), RS (Brazil). Programa de Pós-Graduação em Modelagem Matemática

    2017-07-01

    We describe in this work the application of the modified power method for solve the multigroup neutron diffusion eigenvalue problem in slab geometry considering two-dimensions for nuclear reactor global calculations. It is well known that criticality calculations can often be best approached by solving eigenvalue problems. The criticality in nuclear reactors physics plays a relevant role since establishes the ratio between the numbers of neutrons generated in successive fission reactions. In order to solve the eigenvalue problem, a modified power method is used to obtain the dominant eigenvalue (effective multiplication factor (K{sub eff})) and its corresponding eigenfunction (scalar neutron flux), which is non-negative in every domain, that is, physically relevant. The innovation of this work is solving the neutron diffusion equation in analytical form for each new iteration of the power method. For solve this problem we propose to apply the Finite Fourier Sine Transform on one of the spatial variables obtaining a transformed problem which is resolved by well-established methods for ordinary differential equations. The inverse Fourier transform is used to reconstruct the solution for the original problem. It is known that the power method is an iterative source method in which is updated by the neutron flux expression of previous iteration. Thus, for each new iteration, the neutron flux expression becomes larger and more complex due to analytical solution what makes propose that it be reconstructed through an polynomial interpolation. The methodology is implemented to solve a homogeneous problem and the results are compared with works presents in the literature. (author)

  5. Fission in a Plasma

    Energy Technology Data Exchange (ETDEWEB)

    Younes, W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-10-26

    A three-year theory project was undertaken to study the fission process in extreme astrophysical environments, such as the crust of neutron stars. In the first part of the project, the effect of electron screening on the fission process was explored using a microscopic approach. For the first time, these calculations were carried out to the breaking point of the nucleus. In the second part of the project, the population of the fissioning nucleus was calculated within the same microscopic framework. These types of calculations are extremely computer-intensive and have seldom been applied to heavy deformed nuclei, such as fissioning actinides. The results, tools and methodologies produced in this work will be of interest to both the basic-science and nuclear-data communities.

  6. Microscopic Theory of Fission

    International Nuclear Information System (INIS)

    Younes, W; Gogny, D

    2008-01-01

    In recent years, the microscopic method has been applied to the notoriously difficult problem of nuclear fission with unprecedented success. In this paper, we discuss some of the achievements and promise of the microscopic method, as embodied in the Hartree-Fock method using the Gogny finite-range effective interaction, and beyond-mean-field extensions to the theory. The nascent program to describe induced fission observables using this approach at the Lawrence Livermore National Laboratory is presented

  7. Fusion-fission dynamics

    International Nuclear Information System (INIS)

    Blocki, J.; Planeta, R.; Brzychczyk, J.; Grotowski, K.

    1992-01-01

    Classical dynamical calculations of the heavy ion induced fission processes have been performed for the reactions 40 Ar+ 141 Pr, 20 Ne+ 165 Ho and 12 C+ 175 Lu leading to the iridium like nucleus. As a result prescission lifetimes were obtained and compared with the experimental values. The comparison between the calculated and experimental lifetimes indicates that the one-body dissipation picture is much more relevant in describing the fusion-fission dynamics than the two-body one. (orig.)

  8. Multigroup P8 - elastic scattering matrices of main reactor elements

    International Nuclear Information System (INIS)

    Garg, S.B.; Shukla, V.K.

    1979-01-01

    To study the effect of anisotropic scattering phenomenon on shielding and neutronics of nuclear reactors multigroup P8-elastic scattering matrices have been generated for H, D, He, 6 Li, 7 Li, 10 B, C, N, O, Na, Cr, Fe, Ni, 233 U, 235 U, 238 U, 239 Pu, 240 Pu, 241 Pu and 242 Pu using their angular distribution, Legendre coefficient and elastic scattering cross-section data from the basic ENDF/B library. Two computer codes HSCAT and TRANS have been developed to complete this task for BESM-6 and CDC-3600 computers. These scattering matrices can be directly used as input to the transport theory codes ANISN and DOT. (auth.)

  9. MORET: Version 4.B. A multigroup Monte Carlo criticality code

    International Nuclear Information System (INIS)

    Jacquet, Olivier; Miss, Joachim; Courtois, Gerard

    2003-01-01

    MORET 4 is a three dimensional multigroup Monte Carlo code which calculates the effective multiplication factor (keff) of any configurations more or less complex as well as reaction rates in the different volumes of the geometry and the leakage out of the system. MORET 4 is the Monte Carlo code of the APOLLO2-MORET 4 standard route of CRISTAL, the French criticality package. It is the most commonly used Monte Carlo code for French criticality calculations. During the last four years, the MORET 4 team has developed or improved the following major points: modernization of the geometry, implementation of perturbation algorithms, source distribution convergence, statistical detection of stationarity, unbiased variance estimation and creation of pre-processing and post-processing tools. The purpose of this paper is not only to present the new features of MORET but also to detail clearly the physical models and the mathematical methods used in the code. (author)

  10. Multi-group dynamic quantum secret sharing with single photons

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Hongwei [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China); Ma, Haiqiang, E-mail: hqma@bupt.edu.cn [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China); Wei, Kejin [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China); Yang, Xiuqing [School of Science, Beijing Jiaotong University, Beijing 100044 (China); Qu, Wenxiu; Dou, Tianqi; Chen, Yitian; Li, Ruixue; Zhu, Wu [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China)

    2016-07-15

    In this letter, we propose a novel scheme for the realization of single-photon dynamic quantum secret sharing between a boss and three dynamic agent groups. In our system, the boss can not only choose one of these three groups to share the secret with, but also can share two sets of independent keys with two groups without redistribution. Furthermore, the security of communication is enhanced by using a control mode. Compared with previous schemes, our scheme is more flexible and will contribute to a practical application. - Highlights: • A multi-group dynamic quantum secret sharing with single photons scheme is proposed. • Any one of the groups can be chosen to share secret through controlling the polarization of photons. • Two sets of keys can be shared simultaneously without redistribution.

  11. The isotope density inverse problem in multigroup neutron transport

    International Nuclear Information System (INIS)

    Zazula, J.M.

    1981-01-01

    The inverse problem for stationary multigroup anisotropic neutron transport is discussed in order to search for isotope densities in multielement medium. The spatial- and angular-integrated form of neutron transport equation, in terms of the flux in a group - density of an element spatial correlation, leads to a set of integral functionals for the densities weighted by the group fluxes. Some methods of approximation to make the problem uniquently solvable are proposed. Particularly P 0 angular flux information and the spherically-symetrical geometry of an infinite medium are considered. The numerical calculation using this method related to sooner evaluated direct problem data gives promising agreement with primary densities. This approach would be the basis for further application in an elemental analysis of a medium, using an isotopic neutron source and a moving, energy-dependent neutron detector. (author)

  12. A Laplace transform method for energy multigroup hybrid discrete ordinates

    International Nuclear Information System (INIS)

    Segatto, C.F.; Vilhena, M.T.; Barros, R.C.

    2010-01-01

    In typical lattice cells where a highly absorbing, small fuel element is embedded in the moderator, a large weakly absorbing medium, high-order transport methods become unnecessary. In this work we describe a hybrid discrete ordinates (S N) method for energy multigroup slab lattice calculations. This hybrid S N method combines the convenience of a low-order S N method in the moderator with a high-order S N method in the fuel. The idea is based on the fact that in weakly absorbing media whose physical size is several neutron mean free paths in extent, even the S 2 method (P 1 approximation), leads to an accurate result. We use special fuel-moderator interface conditions and the Laplace transform (LTS N ) analytical numerical method to calculate the two-energy group neutron flux distributions and the thermal disadvantage factor. We present numerical results for a range of typical model problems.

  13. ANISN-L, a CDC-7600 code which solves the one-dimensional, multigroup, time dependent transport equation by the method of discrete ordinates

    Energy Technology Data Exchange (ETDEWEB)

    Wilcox, T. P.

    1973-09-20

    The code ANISN-L solves the one-dimensional, multigroup, time-independent Boltzmann transport equation by the method of discrete ordinates. In problems involving a fissionable system, it can calculate the system multiplication or alpha. In such cases, it is also capable of determining isotopic concentrations, radii, zone widths, or buckling in order to achieve a given multiplication or alpha. The code may also calculate fluxes caused by a specified fixed source. Neutron, gamma, and coupled neutron--gamma problems may be solved in either the forward or adjoint (backward) modes. Cross sections describing upscatter, as well as the usual downscatter, may be employed. This report describes the use of ANISN-L; this is a revised version of ANISN which handles both large and small problems efficiently on CDC-7600 computers. (RWR)

  14. Potentials of fissioning plasmas

    International Nuclear Information System (INIS)

    Karlheinz, Thom.

    1979-01-01

    Successful experiments with the nuclear pumping of lasers have demonstrated that in gaseous medium the kinetic energy of fission fragments can be converted directly into non-equilibrium optical radiation. This confirms the concept that the fissioning medium in a gas-phase nuclear reactor shows an internal structure such as a plasma in nearly thermal equilibrium varying up to a state of extreme-non-equilibrium. The accompanying variations of temperatures, pressure and radiative spectrum suggest wide ranges of applications. For example, in the gas-phase fission reactor concept enriched uranium hexafluoride or an uranium plasma replaces conventional fuel elements and permits operation above the melting point of solid materials. This potential has been motivation for the US National Aeronautics and Space Administration (NASA) to conduct relevant research for high specific impulse propulsion in space. The need to separate the high temperature gaseous fuel from the surfaces of a containing vessel and to protect them against thermal radiation has led to the concept of an externally moderated reactor in which the fissioning gaseous material is suspended by fluid dynamic means and the flow of opaque buffer gas removes the power. The gaseous nuclear fuel can slowly be circulated through the reactor for continuous on-site reprocessing including the annihilation of transuranium actinides at fission when being fed back into the reactor. An equilibrium of the generation and destruction of such actinides at fission when being fed back into the reactor. An equilibrium of the generation and destruction of such actinides can thus be achieved. These characteristics and the unique radiative properties led to the expectation that the gas-phase fission reactor could feature improved safety, safeguarding and economy, in addition to new technologies such as processing, photochemistry and the transmission of power over large distances in space

  15. RGENDF - An interface program between the NJOY code and codes using multigroup cross-sections

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Anaf, J.

    1988-02-01

    An interface program for reformatting multigroup cross-section libraries generated by NJOY into ENDF/B-V format and the EXPANDA, PFCOND and COMPAR input formats is presented. (author). 7 refs, 1 fig., 1 tab

  16. Kalpakkam multigroup cross section set for fast reactor applications - status and performance

    International Nuclear Information System (INIS)

    Ramanadhan, M.M.; Gopalakrishnan, M.M.

    1986-01-01

    This report documents the status of the presently created set of multigroup constants at Kalpakkam. The list of nuclides processed and the details of multigroup structure are given. Also included are the particulars of dilutions and temperatures for each nuclide in the multigroup cross section set for which self shielding factors have been calculated. Using this new multigroup cross section set, measured integral quantities such as K-eff, central reaction rate ratios, central reactivity worths etc. were calculated for a few fast critical benchmark assemblies and the calculated values of neutronic parameters obtained were compared with those obtained using the available Cadarache cross section library and those published in literature for ENDF/B-IV based set and Japanese evaluated nuclear data library (JENDL). The details of analyses are documented along with the conclusions. (author). 17 refs., 12 tabs

  17. Self-shielding phenomenon modelling in multigroup transport code Apollo-2; Modelisation du phenomene d'autoprotection dans le code de transport multigroupe Apollo 2

    Energy Technology Data Exchange (ETDEWEB)

    Coste-Delclaux, M

    2006-03-15

    This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)

  18. Self-shielding phenomenon modelling in multigroup transport code Apollo-2; Modelisation du phenomene d'autoprotection dans le code de transport multigroupe Apollo 2

    Energy Technology Data Exchange (ETDEWEB)

    Coste-Delclaux, M

    2006-03-15

    This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)

  19. Social comparison and perceived breach of psychological contract: their effects on burnout in a multigroup analysis.

    Science.gov (United States)

    Cantisano, Gabriela Topa; Domínguez, J Francisco Morales; García, J Luis Caeiro

    2007-05-01

    This study focuses on the mediator role of social comparison in the relationship between perceived breach of psychological contract and burnout. A previous model showing the hypothesized effects of perceived breach on burnout, both direct and mediated, is proposed. The final model reached an optimal fit to the data and was confirmed through multigroup analysis using a sample of Spanish teachers (N = 401) belonging to preprimary, primary, and secondary schools. Multigroup analyses showed that the model fit all groups adequately.

  20. Complex of two-dimensional multigroup programs for neutron-physical computations of nuclear reactor

    International Nuclear Information System (INIS)

    Karpov, V.A.; Protsenko, A.N.

    1975-01-01

    Briefly stated mathematical aspects of the two-dimensional multigroup method of neutron-physical computation of nuclear reactor. Problems of algorithmization and BESM-6 computer realisation of multigroup diffuse approximations in hexagonal and rectangular calculated lattices are analysed. The results of computation of fast critical assembly having complicated composition of the core are given. The estimation of computation accuracy of criticality, neutron fields distribution and efficiency of absorbing rods by means of computer programs developed is done. (author)

  1. Application of direct discrete method (DDM) to multigroup neutron transport problems

    International Nuclear Information System (INIS)

    Vosoughi, Naser; Salehi, Ali Akbar; Shahriari, Majid

    2003-01-01

    The Direct Discrete Method (DDM), which produced excellent results for one-group neutron transport problems, has been developed for multigroup energy. A multigroup neutron transport discrete equation has been produced for a cylindrical shape fuel element with and without associated coolant regions with two boundary conditions. The calculations are illustrated for two-group energy by graphs showing the fast and thermal fluxes. The validity of the results are tested against the results obtained by the ANISN code. (author)

  2. RADHEAT-V4: a code system to generate multigroup constants and analyze radiation transport for shielding safety evaluation

    International Nuclear Information System (INIS)

    Yamano, Naoki; Minami, Kazuyoshi; Koyama, Kinji; Naito, Yoshitaka.

    1989-03-01

    A modular code system RADHEAT-V4 has been developed for performing precisely neutron and photon transport analyses, and shielding safety evaluations. The system consists of the functional modules for producing coupled multi-group neutron and photon cross section sets, for analyzing the neutron and photon transport, and for calculating the atom displacement and the energy deposition due to radiations in nuclear reactor or shielding material. A precise method named Direct Angular Representation (DAR) has been developed for eliminating an error associated with the method of the finite Legendre expansion in evaluating angular distributions of cross sections and radiation fluxes. The DAR method implemented in the code system has been described in detail. To evaluate the accuracy and applicability of the code system, some test calculations on strong anisotropy problems have been performed. From the results, it has been concluded that RADHEAT-V4 is successfully applicable to evaluating shielding problems accurately for fission and fusion reactors and radiation sources. The method employed in the code system is very effective in eliminating negative values and oscillations of angular fluxes in a medium having an anisotropic source or strong streaming. Definitions of the input data required in various options of the code system and the sample problems are also presented. (author)

  3. Nuclear fission and neutron-induced fission cross-sections

    Energy Technology Data Exchange (ETDEWEB)

    James, G.D.; Lynn, J.E.; Michaudon, A.; Rowlands, J.; de Saussure, G.

    1981-01-01

    A general presentation of current knowledge of the fission process is given with emphasis on the low energy fission of actinide nuclei and neutron induced fission. The need for and the required accuracy of fission cross section data in nuclear energy programs are discussed. A summary is given of the steps involved in fission cross section measurement and the range of available techniques. Methods of fission detection are described with emphasis on energy dependent changed and detector efficiency. Examples of cross section measurements are given and data reduction is discussed. The calculation of fission cross sections is discussed and relevant nuclear theory including the formation and decay of compound nuclei and energy level density is introduced. A description of a practical computation of fission cross sections is given.

  4. The evaluation for reference fission yield of 238U fission

    International Nuclear Information System (INIS)

    Liang Qichang; Liu Tingjin

    1998-01-01

    In the fission yield data evaluation and measurement, the reference yield is very important, good or poor recommended or measurement values depend upon the reference data to a great extent. According to the CRP's requirement, the evaluation of reference fission yields have been and will be carried out in CNDC, as a part of the whole work (contract No.9504/R 0 /Regular Budget Fund), the evaluation for 29 reference fission yields of 15 product nuclides from 238 U fission have been completed

  5. Fission fragment angular momentum

    International Nuclear Information System (INIS)

    Frenne, D. De

    1991-01-01

    Most of the energy released in fission is converted into translational kinetic energy of the fragments. The remaining excitation energy will be distributed among neutrons and gammas. An important parameter characterizing the scission configuration is the primary angular momentum of the nascent fragments. Neutron emission is not expected to decrease the spin of the fragments by more than one unit of angular momentum and is as such of less importance in the determination of the initial fragment spins. Gamma emission is a suitable tool in studying initial fragment spins because the emission time, number, energy, and multipolarity of the gammas strongly depend on the value of the primary angular momentum. The main conclusions of experiments on gamma emission were that the initial angular momentum of the fragments is large compared to the ground state spin and oriented perpendicular to the fission axis. Most of the recent information concerning initial fragment spin distributions comes from the measurement of isomeric ratios for isomeric pairs produced in fission. Although in nearly every mass chain isomers are known, only a small number are suitable for initial fission fragment spin studies. Yield and half-life considerations strongly limit the number of candidates. This has the advantage that the behavior of a specific isomeric pair can be investigated for a number of fissioning systems at different excitation energies of the fragments and fissioning nuclei. Because most of the recent information on primary angular momenta comes from measurements of isomeric ratios, the global deexcitation process of the fragments and the calculation of the initial fragment spin distribution from measured isomeric ratios are discussed here. The most important results on primary angular momentum determinations are reviewed and some theoretical approaches are given. 45 refs., 7 figs., 2 tabs

  6. Fission product detection

    International Nuclear Information System (INIS)

    Liatard, E.; Akrouf, S.; Bruandet, J.F

    1987-01-01

    The response of photovoltaic cells to heavy ions and fission products have been tested on beam. Their main advantages are their extremely low price, their low sensitivity to energetic light ions with respect to fission products, and the possibility to cut and fit them together to any shape without dead zone. The time output signals of a charge sensitive preamplifier connected to these cells allows fast coincidences. A resolution of 12ns (F.W.H.M.) have been measured between two cells [fr

  7. Low energy nuclear fission

    International Nuclear Information System (INIS)

    Nifenecker, H.

    1982-02-01

    In these lectures we present the liquid drop model of fission and compare some of its prediction with experiment. The liquid drop analogy allows to define in a rather simple and intuitive way a number of useful concepts and possible observables. We then discuss, using the example of the oscillator model, the generality of shell effects. We show how a synthesis of the liquid drop model and of the shell model can be made using the Strutinsky shell averaging procedure. Some experimental data related to the existence of shape isomers are presented and discussed. We conclude by discussing some aspects, both experimental and theoretical, of fission dynamics

  8. Fission of heavy hypernuclei

    International Nuclear Information System (INIS)

    Nifenecker, H.

    1993-01-01

    The results on delayed and prompt fission of heavy hypernuclei obtained by the LEAR PS177 collaboration are recalled and discussed. It is shown that the hypernuclei life-times can be explained in term of a weak strangeness violating lambda-nucleon interaction with a cross section close to 6.0 10 -15 barns. The lambda attachment function is shown to be sensitive to the scission configuration, just before fission, and to the neck dynamics. This function provides a new way to study the nuclear scission process. (author)

  9. Fission gas measuring technology

    International Nuclear Information System (INIS)

    Lee, Hyung Kwon; Kim, Eun Ka; Hwang, Yong Hwa; Lee, Eun Pyo; Chun, Yong Bum; Seo, Ki Seog; Park, Dea Gyu; Chu, Yong Sun; Ahn, Sang Bok.

    1998-02-01

    Safety and economy of nuclear plant are greatly affected by the integrity of nuclear fuels during irradiation reactor core. A series of post-irradiation examination (PIE) including non-destructive and destructive test is to be conducted to evaluate and characterize the nuclear performance. In this report, a principle of the examination equipment to measure and analyse fission gases existing nuclear fuels were described and features of the component and device consisting the fission gas measuring equipment are investigated. (author). 4 refs., 2 tabs., 6 figs

  10. Fission gas measuring technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyung Kwon; Kim, Eun Ka; Hwang, Yong Hwa; Lee, Eun Pyo; Chun, Yong Bum; Seo, Ki Seog; Park, Dea Gyu; Chu, Yong Sun; Ahn, Sang Bok

    1998-02-01

    Safety and economy of nuclear plant are greatly affected by the integrity of nuclear fuels during irradiation reactor core. A series of post-irradiation examination (PIE) including non-destructive and destructive test is to be conducted to evaluate and characterize the nuclear performance. In this report, a principle of the examination equipment to measure and analyse fission gases existing nuclear fuels were described and features of the component and device consisting the fission gas measuring equipment are investigated. (author). 4 refs., 2 tabs., 6 figs.

  11. Fission modelling with FIFRELIN

    International Nuclear Information System (INIS)

    Litaize, Olivier; Serot, Olivier; Berge, Leonie

    2015-01-01

    The nuclear fission process gives rise to the formation of fission fragments and emission of particles (n,γ, e - ). The particle emission from fragments can be prompt and delayed. We present here the methods used in the FIFRELIN code, which simulates the prompt component of the de-excitation process. The methods are based on phenomenological models associated with macroscopic and/or microscopic ingredients. Input data can be provided by experiment as well as by theory. The fission fragment de-excitation can be performed within Weisskopf (uncoupled neutron and gamma emission) or a Hauser-Feshbach (coupled neutron/gamma emission) statistical theory. We usually consider five free parameters that cannot be provided by theory or experiments in order to describe the initial distributions required by the code. In a first step this set of parameters is chosen to reproduce a very limited set of target observables. In a second step we can increase the statistics to predict all other fission observables such as prompt neutron, gamma and conversion electron spectra but also their distributions as a function of any kind of parameters such as, for instance, the neutron, gamma and electron number distributions, the average prompt neutron multiplicity as a function of fission fragment mass, charge or kinetic energy, and so on. Several results related to different fissioning systems are presented in this work. The goal in the next decade will be i) to replace some macroscopic ingredients or phenomenological models by microscopic calculations when available and reliable, ii) to be a support for experimentalists in the design of detection systems or in the prediction of necessary beam time or count rates with associated statistics when measuring fragments and emitted particle in coincidence iii) extend the model to be able to run a calculation when no experimental input data are available, iv) account for multiple chance fission and gamma emission before fission, v) account for the

  12. Low energy nuclear fission

    International Nuclear Information System (INIS)

    Nifenecker, H.

    1980-08-01

    In these lectures the liquid drop model of fission is presented and some of its predictions compared with experiment. The liquid drop analogy allows to define in a rather simple and intuitive way a number of useful concepts and possible observables. It is shown how a synthesis of the liquid drop model and of the shell model can be made using the Strutinsky shell averaging procedure. Some experimental data related to the existence of shape isomers are presented and discussed. We conclude by discussing some aspects, both experimental and theoretical, of fission dynamics

  13. Fission Product Library and Resource

    Energy Technology Data Exchange (ETDEWEB)

    Burke, J. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Padgett, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-09-29

    Fission product yields can be extracted from an irradiated sample by performing gamma ray spectroscopy on the whole sample post irradiation. There are several pitfalls to avoid when trying to determine a specific isotope's fission product yield.

  14. Spontaneous fission of superheavy nuclei

    Indian Academy of Sciences (India)

    The fission-like configurations are used for the total deformation energy calculations. A ... oscillator potential for the two fission fragment regions reads as ... Beyond this limit, the contribution of more remote levels is negligible. Once the density ...

  15. Mirror fusion--fission hybrids

    International Nuclear Information System (INIS)

    Lee, J.D.

    1978-01-01

    The fusion-fission concept and the mirror fusion-fission hybrid program are outlined. Magnetic mirror fusion drivers and blankets for hybrid reactors are discussed. Results of system analyses are presented and a reference design is described

  16. Process for treating fission waste

    International Nuclear Information System (INIS)

    Rohrmann, C.A.; Wick, O.J.

    1983-01-01

    A method is described for the treatment of fission waste. A glass forming agent, a metal oxide, and a reducing agent are mixed with the fission waste and the mixture is heated. After melting, the mixture separates into a glass phase and a metal phase. The glass phase may be used to safely store the fission waste, while the metal phase contains noble metals recovered from the fission waste

  17. FINELM: a multigroup finite element diffusion code. Part I

    International Nuclear Information System (INIS)

    Davierwalla, D.M.

    1980-12-01

    The author presents a two dimensional code for multigroup diffusion using the finite element method. It was realized that the extensive connectivity which contributes significantly to the accuracy, results in a matrix which, although symmetric and positive definite, is wide band and possesses an irregular profile. Hence, it was decided to introduce sparsity techniques into the code. The introduction of the R-Z geometry lead to a great deal of changes in the code since the rotational invariance of the removal matrices in X-Y geometry did not carry over in R-Z geometry. Rectangular elements were introduced to remedy the inability of the triangles to model essentially one dimensional problems such as slab geometry. The matter is discussed briefly in the text in the section on benchmark problems. This report is restricted to the general theory of the triangular elements and to the sparsity techniques viz. incomplete disections. The latter makes the size of the problem that can be handled independent of core memory and dependent only on disc storage capacity which is virtually unlimited. (Auth.)

  18. Travelling Wave Solutions in Multigroup Age-Structured Epidemic Models

    Science.gov (United States)

    Ducrot, Arnaut; Magal, Pierre; Ruan, Shigui

    2010-01-01

    Age-structured epidemic models have been used to describe either the age of individuals or the age of infection of certain diseases and to determine how these characteristics affect the outcomes and consequences of epidemiological processes. Most results on age-structured epidemic models focus on the existence, uniqueness, and convergence to disease equilibria of solutions. In this paper we investigate the existence of travelling wave solutions in a deterministic age-structured model describing the circulation of a disease within a population of multigroups. Individuals of each group are able to move with a random walk which is modelled by the classical Fickian diffusion and are classified into two subclasses, susceptible and infective. A susceptible individual in a given group can be crisscross infected by direct contact with infective individuals of possibly any group. This process of transmission can depend upon the age of the disease of infected individuals. The goal of this paper is to provide sufficient conditions that ensure the existence of travelling wave solutions for the age-structured epidemic model. The case of two population groups is numerically investigated which applies to the crisscross transmission of feline immunodeficiency virus (FIV) and some sexual transmission diseases.

  19. Multigroup perturbation model for kinetic analysis of nuclear reactors

    International Nuclear Information System (INIS)

    Souza, G.M.

    1989-01-01

    The scope of this work is the development of a multigroup perturbation theory for the purpose of Kinetic and dynamic analysis of nuclear reactors. The equations that describe the reactor behavior were presented in all generality and written in the shorthand notation of matrices and vectors. In the derivation of those equations indetermined operators and discretizing factors were introduced and then determined by comparision with conventional equations. Fick's Law was developed in higher orders for neutron and importance current density. The solution of the direct and adjoint fields were represented by combination of the eigenfunctions of the B and B* operators and the eigenvalue modulus equality was established mathematically. In the derivation of the reactivity expression the B operator perturbation was split in two non coupled to the flux form and level. The prompt neutrons effective mean life was derived from reactor equations and importance conservation. The establishment of the Nordheim's equation, although modified, was based on Gandini. Finally, a mathematical interpretation of the flux-trap region was avented. (author)

  20. 50 years of nuclear fission

    International Nuclear Information System (INIS)

    Hilscher, D.

    1989-01-01

    The article tells the story of the discovery of nuclear fission in Berlin 50 years ago by Otto Hahn and Fritz Strassmann in cooperation with Lise Meitner. 50 years later nuclear fission is still a subject of research. Some question remain unanswered. Selected new research results are used to discuss the dynamics of the collective movement of the elementary nuclear fission process. (orig.) [de

  1. Fission dynamics of hot nuclei

    Indian Academy of Sciences (India)

    2014-04-05

    Apr 5, 2014 ... across the fission barrier is very small or in other words, the fission barrier is much ... of this shape evolution, the gross features of the fissioning nucleus can be described ..... [7] Y Abe, C Gregoire and H Delagrange, J. Phys.

  2. Status of fission yield measurements

    International Nuclear Information System (INIS)

    Maeck, W.J.

    1979-01-01

    Fission yield measurement and yield compilation activities in the major laboratories of the world are reviewed. In addition to a general review of the effort of each laboratory, a brief summary of yield measurement activities by fissioning nuclide is presented. A new fast reactor fission yield measurement program being conducted in the US is described

  3. The discovery of fission

    International Nuclear Information System (INIS)

    McKay, H.A.C.

    1978-01-01

    In this article by the retired head of the Separation Processes Group of the Chemistry Division, Atomic Energy Research Establishment, Harwell, U.K., the author recalls what he terms 'an exciting drama, the unravelling of the nature of the atomic nucleus' in the years before the Second World War, including the discovery of fission. 12 references. (author)

  4. Elastocapillary Instability in Mitochondrial Fission

    Science.gov (United States)

    Gonzalez-Rodriguez, David; Sart, Sébastien; Babataheri, Avin; Tareste, David; Barakat, Abdul I.; Clanet, Christophe; Husson, Julien

    2015-08-01

    Mitochondria are dynamic cell organelles that constantly undergo fission and fusion events. These dynamical processes, which tightly regulate mitochondrial morphology, are essential for cell physiology. Here we propose an elastocapillary mechanical instability as a mechanism for mitochondrial fission. We experimentally induce mitochondrial fission by rupturing the cell's plasma membrane. We present a stability analysis that successfully explains the observed fission wavelength and the role of mitochondrial morphology in the occurrence of fission events. Our results show that the laws of fluid mechanics can describe mitochondrial morphology and dynamics.

  5. A threshold for dissipative fission

    International Nuclear Information System (INIS)

    Thoennessen, M.; Bertsch, G.F.

    1993-01-01

    The empirical domain of validity of statistical theory is examined as applied to fission data on pre-fission data on pre-fission neutron, charged particle, and γ-ray multiplicities. Systematics are found of the threshold excitation energy for the appearance of nonstatistical fission. From the data on systems with not too high fissility, the relevant phenomenological parameter is the ratio of the threshold temperature T thresh to the (temperature-dependent) fission barrier height E Bar (T). The statistical model reproduces the data for T thresh /E Bar (T) thresh /E Bar (T) independent of mass and fissility of the systems

  6. Barium 139 as Fission Indicator

    Energy Technology Data Exchange (ETDEWEB)

    Broda, E.

    1943-07-01

    This report is based on a measurement performed at the Cavendish Laboratory (Cambridge) by E. Broda in December 1943 where a technique has been worked out for measuring the fission density in a uranium containing medium in relative units by determining the amount of a suitable fission product formed. Generally a given fission product will be formed in natural uranium by slow neutron fission of U235 or by fast neutron fission of either U235 or U238. It is intended to translate the relative units into absolute units by comparison of the Ba yield with the indication of UF6 fission chamber in the same medium. This has to be done separately for fast and slow neutron fission as the yields may be different. Another application of the technique developed is the measurement of thermal neutron density in an uraniferous medium without using a detector subject to variations of sensitivity according to the properties of the medium. (nowak)

  7. Fusion-fission dynamics

    International Nuclear Information System (INIS)

    Blocki, J.; Planeta, R.; Brzychczyk, J.; Grotowski, K.

    1991-04-01

    Classical dynamical calculations of the heavy ion induced fission process for the reactions 40 Ar+ 141 Pr, 20 Ne+ 165 Ho and 12 C+ 175 Lu leading to the iridium like nucleus have been performed. As a result prescission lifetimes were obtained and compared with the experimental values. The agreement between the calculated and experimental lifetimes indicates that the one-body dissipation picture is much more relevant in describing the fusion-fission dynamics than the two-body one. Somewhat bigger calculated times than the experimental ones in case of the C+Lu reaction at 16 MeV/nucleon may be a signal on the energy range applicability of the one-body dissipation model. (author)

  8. The fission track method

    International Nuclear Information System (INIS)

    Hansen, K.

    1990-01-01

    During the last decade fission track (FT) analysis has evolved as an important tool in exploration for hydrocarbon resources. Most important is this method's ability to yield information about temperatures at different times (history), and thus relate oil generation and time independently of other maturity parameters. The purpose of this paper is to introduce the basics of the method and give an example from the author's studies. (AB) (14 refs.)

  9. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system

    International Nuclear Information System (INIS)

    Yang, W.S.; Lee, C.H.

    2008-01-01

    results overestimated the multiplication factor by 0.22 ∼ 0.35% Δρ for these small systems with very hard neutron spectrum. Comparisons of measured and calculated values for the fission reaction rate ratios of Godiva and Jezebel assemblies also showed that the MC 2 -2/TWODANT results agreed well with measurements within 2.7%. From a series of methodology review and ENDF/B-VII.0 data processing, several improvement needs to enhance accuracy were also identified for the ETOE-2/MC 2 -2 code system, including the multigroup slowing-down solution for whole-energy range, proper treatment for anisotropy of inelastic scattering, improved evaluation of inelastic and high-order anisotropic scattering source in RABANL calculations.

  10. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

    Energy Technology Data Exchange (ETDEWEB)

    Yang, W. S.; Lee, C. H. (Nuclear Engineering Division)

    2008-05-16

    . Relative to the MCNP solution, the MC{sup 2}-2/TWODANT results overestimated the multiplication factor by 0.22 {approx} 0.35% {Delta}{rho} for these small systems with very hard neutron spectrum. Comparisons of measured and calculated values for the fission reaction rate ratios of Godiva and Jezebel assemblies also showed that the MC{sup 2}-2/TWODANT results agreed well with measurements within 2.7%. From a series of methodology review and ENDF/B-VII.0 data processing, several improvement needs to enhance accuracy were also identified for the ETOE-2/MC{sup 2}-2 code system, including the multigroup slowing-down solution for whole-energy range, proper treatment for anisotropy of inelastic scattering, improved evaluation of inelastic and high-order anisotropic scattering source in RABANL calculations.

  11. [Fission product yields of 60 fissioning reactions]. Final report

    International Nuclear Information System (INIS)

    Rider, B.F.

    1995-01-01

    In keeping with the statement of work, I have examined the fission product yields of 60 fissioning reactions. In co-authorship with the UTR (University Technical Representative) Talmadge R. England ''Evaluation and Compilation of Fission Product Yields 1993,'' LA-UR-94-3106(ENDF-349) October, (1994) was published. This is an evaluated set of fission product Yields for use in calculation of decay heat curves with improved accuracy has been prepared. These evaluated yields are based on all known experimental data through 1992. Unmeasured fission product yields are calculated from charge distribution, pairing effects, and isomeric state models developed at Los Alamos National Laboratory. The current evaluation has been distributed as the ENDF/B-VI fission product yield data set

  12. Analysis of a multigroup stylized CANDU half-core benchmark

    International Nuclear Information System (INIS)

    Pounders, Justin M.; Rahnema, Farzad; Serghiuta, Dumitru

    2011-01-01

    Highlights: → This paper provides a benchmark that is a stylized model problem in more than two energy groups that is realistic with respect to the underlying physics. → An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core CANDU benchmark problem. → Reference eigenvalues and selected pin and bundle fission rates are included. → 2-, 4- and 47-group Monte Carlo solutions are compared to analyze homogenization-free transport approximations that result from energy condensation. - Abstract: An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core Canadian deuterium uranium (CANDU) reactor benchmark problem. Reference eigenvalues and selected pin and bundle fission rates are also included. This benchmark is intended to provide computational reactor physicists and methods developers with a stylized model problem in more than two energy groups that is realistic with respect to the underlying physics. In addition to transport theory code verification, the 8-group energy structure provides reactor physicist with an ideal problem for examining cross section homogenization and collapsing effects in a full-core environment. To this end, additional 2-, 4- and 47-group full-core Monte Carlo benchmark solutions are compared to analyze homogenization-free transport approximations incurred as a result of energy group condensation.

  13. Multigroup calculations of low-energy neutral transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Gilligan, J.G.; Gralnick, S.L.; Price, W.G. Jr.; Kammash, T.

    1978-01-01

    Multigroup discrete ordinates methods avoid many of the approximations that have been used in previous neutral transport analyses. Of particular interest are the neutral profiles generated as an integral part of larger plasma system simulation codes. To determine the appropriateness of utilizing a particular multigroup code, ANISN, for this purpose, results are compared with the neutral transport module of the Duechs code. For a typical TFTR plasma, predicted neutral densities differ by a maximum factor of three on axis and outfluxes at the plasma boundary by approximately 40%. This is found to be significant for a neutral transport module. Possible sources of the observed discrepancies are indicated from an analysis of the approximations used in the Duechs model. Recommendations are made concerning the future application of the multigroup method. (author)

  14. Proposal to extend CSEWG neutron and photon multigroup structures for wider applications

    International Nuclear Information System (INIS)

    LaBauve, R.J.; Wilson, W.B.

    1976-02-01

    The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented

  15. Proposal to extend CSEWG neutron and photon multigroup structures for wider applications. [Tables

    Energy Technology Data Exchange (ETDEWEB)

    LaBauve, R.J.; Wilson, W.B.

    1976-02-01

    The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented.

  16. Coulomb fission and transfer fission at heavy ion collisions

    International Nuclear Information System (INIS)

    Himmele, G.

    1981-01-01

    In the present thesis the first direct evidence of nuclear fission after inelastic scattering of heavy ions (sup(183,184)W, 152 Sm → 238 U; 184 W → 232 Th; 184 W, 232 Th → 248 Cm) is reported. Experiments which were performed at the UNILAC of the Gesellschaft fuer Schwerionenforschung in Darmstadt show the observed heavy ion induced fission possesses significant properties of the Coulomb fission. The observed dependence of the fission probability for inelastic scattering on the projectile charge proves that the nuclear fission is mediated by the electromagnetic interaction between heavy ions. This result suggests moreover a multiple Coulomb-excitation preceding the fission. Model calculations give a first indication, that the Coulomb fission proceeds mainly from the higher β phonons. In the irradiation with 184 W the fission probability of 232 Th is for all incident energies about 40% smaller that at 238 U. The target dependence of the Coulomb fission however doesn't allow, to give quantitative statements about the position and B(E2)-values of higher lying β phonons. (orig./HSI) [de

  17. Multigroup cross section collapsing optimization of a He-3 detector assembly model using deterministic transport techniques

    International Nuclear Information System (INIS)

    Huang, Mi; Yi, Ce; Manalo, Kevin L.; Sjoden, Glenn E.

    2011-01-01

    Multigroup optimization is performed on a neutron detector assembly to examine the validity of transport response in forward and adjoint modes. For SN transport simulations, we discuss the multigroup collapse of an 80 group library to 40, 30, and 16 groups, constructed from using the 3-D parallel PENTRAN and macroscopic cross section collapsing with YGROUP contribution weighting. The difference in using P_1 and P_3 Legendre order in scattering cross sections is investigated; also, associated forward and adjoint transport responses are calculated. We conclude that for the block analyzed, a 30 group cross section optimizes both computation time and accuracy relative to the 80 group transport calculations. (author)

  18. Multi-level methods for solving multigroup transport eigenvalue problems in 1D slab geometry

    International Nuclear Information System (INIS)

    Anistratov, D. Y.; Gol'din, V. Y.

    2009-01-01

    A methodology for solving eigenvalue problems for the multigroup neutron transport equation in 1D slab geometry is presented. In this paper we formulate and compare different variants of nonlinear multi-level iteration methods. They are defined by means of multigroup and effective one-group low-order quasi diffusion (LOQD) equations. We analyze the effects of utilization of the effective one-group LOQD problem for estimating the eigenvalue. We present numerical results to demonstrate the performance of the iteration algorithms in different types of reactor-physics problems. (authors)

  19. Multi-level nonlinear diffusion acceleration method for multigroup transport k-Eigenvalue problems

    International Nuclear Information System (INIS)

    Anistratov, Dmitriy Y.

    2011-01-01

    The nonlinear diffusion acceleration (NDA) method is an efficient and flexible transport iterative scheme for solving reactor-physics problems. This paper presents a fast iterative algorithm for solving multigroup neutron transport eigenvalue problems in 1D slab geometry. The proposed method is defined by a multi-level system of equations that includes multigroup and effective one-group low-order NDA equations. The Eigenvalue is evaluated in the exact projected solution space of smallest dimensionality, namely, by solving the effective one- group eigenvalue transport problem. Numerical results that illustrate performance of the new algorithm are demonstrated. (author)

  20. Multigroup computation of the temperature-dependent Resonance Scattering Model (RSM) and its implementation

    Energy Technology Data Exchange (ETDEWEB)

    Ghrayeb, S. Z. [Dept. of Mechanical and Nuclear Engineering, Pennsylvania State Univ., 230 Reber Building, Univ. Park, PA 16802 (United States); Ouisloumen, M. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Ougouag, A. M. [Idaho National Laboratory, MS-3860, PO Box 1625, Idaho Falls, ID 83415 (United States); Ivanov, K. N.

    2012-07-01

    A multi-group formulation for the exact neutron elastic scattering kernel is developed. This formulation is intended for implementation into a lattice physics code. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering, which in turn affect the estimation of core reactivity and burnup characteristics. A computer program has been written to test the formulation for various nuclides. Results of the multi-group code have been verified against the correct analytic scattering kernel. In both cases neutrons were started at various energies and temperatures and the corresponding scattering kernels were tallied. (authors)

  1. Multigroup neutron transport equation in the diffusion and P{sub 1} approximation

    Energy Technology Data Exchange (ETDEWEB)

    Obradovic, D [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1970-07-01

    Investigations of the properties of the multigroup transport operator, width and without delayed neutrons in the diffusion and P{sub 1} approximation, is performed using Keldis's theory of operator families as well as a technique . recently used for investigations into the properties of the general linearized Boltzmann operator. It is shown that in the case without delayed neutrons, multigroup transport operator in the diffusion and P{sub 1} approximation possesses a complete set of generalized eigenvectors. A formal solution to the initial value problem is also given. (author)

  2. A small flat fission chamber

    International Nuclear Information System (INIS)

    Li Yijun; Wang Dalun; Chen Suhe

    1999-01-01

    With fission materials of depleted uranium, natural uranium, enriched uranium, 239 Pu, and 237 Np, the authors have designed and made a series of small flat fission chamber. The authors narrated the construction of the fission chamber and its technological process of manufacture, and furthermore, the authors have measured and discussed the follow correct factor, self-absorption, boundary effect, threshold loss factor, bottom scatter and or so

  3. Fission barriers of light nuclei

    International Nuclear Information System (INIS)

    Grotowski, K.; Planeta, R.; Blann, M.; Komoto, T.

    1989-01-01

    Experimental fission excitation functions for compound nuclei /sup 52/Fe, /sup 49/Cr, /sup 46/V, and /sup 44/Ti formed in heavy-ion reactions are analyzed in the Hauser-Feshbach/Bohr-Wheeler formalism using fission barriers based on the rotating liquid drop model of Cohen et al. and on the rotating finite range model of Sierk. We conclude that the rotating finite range approach gives better reproduction of experimental fission yields, consistent with results found for heavier systems

  4. Nuclear Forensics and Radiochemistry: Fission

    Energy Technology Data Exchange (ETDEWEB)

    Rundberg, Robert S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-11-07

    Radiochemistry has been used to study fission since it’ discovery. Radiochemical methods are used to determine cumulative mass yields. These measurements have led to the two-mode fission hypothesis to model the neutron energy dependence of fission product yields. Fission product yields can be used for the nuclear forensics of nuclear explosions. The mass yield curve depends on both the fuel and the neutron spectrum of a device. Recent studies have shown that the nuclear structure of the compound nucleus can affect the mass yield distribution.

  5. Micro plate fission chamber development

    International Nuclear Information System (INIS)

    Wang Mei; Wen Zhongwei; Lin Jufang; Jiang Li; Liu Rong; Wang Dalun

    2014-01-01

    To conduct the measurement of neutron flux and the fission rate distribution at several position in assemblies, the micro plate fission chamber was designed and fabricated. Since the requirement of smaller volume and less structure material was taken into consideration, it is convinient, commercial and practical to use fission chamber to measure neutron flux in specific condition. In this paper, the structure of fission chamber and process of fabrication were introduced and performance test result was presented. The detection efficiency is 91.7%. (authors)

  6. Fission yield measurements at IGISOL

    Science.gov (United States)

    Lantz, M.; Al-Adili, A.; Gorelov, D.; Jokinen, A.; Kolhinen, V. S.; Mattera, A.; Moore, I.; Penttilä, H.; Pomp, S.; Prokofiev, A. V.; Rakopoulos, V.; Rinta-Antila, S.; Simutkin, V.; Solders, A.

    2016-06-01

    The fission product yields are an important characteristic of the fission process. In fundamental physics, knowledge of the yield distributions is needed to better understand the fission process. For nuclear energy applications good knowledge of neutroninduced fission-product yields is important for the safe and efficient operation of nuclear power plants. With the Ion Guide Isotope Separator On-Line (IGISOL) technique, products of nuclear reactions are stopped in a buffer gas and then extracted and separated by mass. Thanks to the high resolving power of the JYFLTRAP Penning trap, at University of Jyväskylä, fission products can be isobarically separated, making it possible to measure relative independent fission yields. In some cases it is even possible to resolve isomeric states from the ground state, permitting measurements of isomeric yield ratios. So far the reactions U(p,f) and Th(p,f) have been studied using the IGISOL-JYFLTRAP facility. Recently, a neutron converter target has been developed utilizing the Be(p,xn) reaction. We here present the IGISOL-technique for fission yield measurements and some of the results from the measurements on proton induced fission. We also present the development of the neutron converter target, the characterization of the neutron field and the first tests with neutron-induced fission.

  7. Fission yield measurements at IGISOL

    Directory of Open Access Journals (Sweden)

    Lantz M.

    2016-01-01

    Full Text Available The fission product yields are an important characteristic of the fission process. In fundamental physics, knowledge of the yield distributions is needed to better understand the fission process. For nuclear energy applications good knowledge of neutroninduced fission-product yields is important for the safe and efficient operation of nuclear power plants. With the Ion Guide Isotope Separator On-Line (IGISOL technique, products of nuclear reactions are stopped in a buffer gas and then extracted and separated by mass. Thanks to the high resolving power of the JYFLTRAP Penning trap, at University of Jyväskylä, fission products can be isobarically separated, making it possible to measure relative independent fission yields. In some cases it is even possible to resolve isomeric states from the ground state, permitting measurements of isomeric yield ratios. So far the reactions U(p,f and Th(p,f have been studied using the IGISOL-JYFLTRAP facility. Recently, a neutron converter target has been developed utilizing the Be(p,xn reaction. We here present the IGISOL-technique for fission yield measurements and some of the results from the measurements on proton induced fission. We also present the development of the neutron converter target, the characterization of the neutron field and the first tests with neutron-induced fission.

  8. Fusion-fission type collisions

    International Nuclear Information System (INIS)

    Oeschler, H.

    1980-01-01

    Three examples of fusion-fission type collisions on medium-mass nuclei are investigated whether the fragment properties are consistent with fission from equilibrated compound nuclei. Only in a very narrow band of angular momenta the data fulfill the necessary criteria for this process. Continuous evolutions of this mechnism into fusion fission and into a deep-inelastic process and particle emission prior to fusion have been observed. Based on the widths of the fragment-mass distributions of a great variety of data, a further criterion for the compound-nucleus-fission process is tentatively proposed. (orig.)

  9. Study on the effect factor of the absolute fission rates measured by depleted uranium fission chamber

    International Nuclear Information System (INIS)

    Jiang Li; Liu Rong; Wang Dalun; Wang Mei; Lin Jufang; Wen Zhongwei

    2003-01-01

    The absolute fission rates was measured by the depleted uranium fission chamber. The efficiency of the fission fragments recorded in the fission chamber was analyzed. The factor influencing absolute fission rates was studied in the experiment, including the disturbing effect between detectors and the effect of the structural of the fission chamber, etc

  10. Energy from nuclear fission()

    Science.gov (United States)

    Ripani, M.

    2015-08-01

    The main features of nuclear fission as physical phenomenon will be revisited, emphasizing its peculiarities with respect to other nuclear reactions. Some basic concepts underlying the operation of nuclear reactors and the main types of reactors will be illustrated, including fast reactors, showing the most important differences among them. The nuclear cycle and radioactive-nuclear-waste production will be also discussed, along with the perspectives offered by next generation nuclear assemblies being proposed. The current situation of nuclear power in the world, its role in reducing carbon emission and the available resources will be briefly illustrated.

  11. Measurement of Fission Product Yields from Fast-Neutron Fission

    Science.gov (United States)

    Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Henderson, R.; Kenneally, J.; Macri, R.; McNabb, D.; Ryan, C.; Sheets, S.; Stoyer, M. A.; Tonchev, A. P.; Bhatia, C.; Bhike, M.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.

    2014-09-01

    One of the aims of the Stockpile Stewardship Program is a reduction of the uncertainties on fission data used for analyzing nuclear test data [1,2]. Fission products such as 147Nd are convenient for determining fission yields because of their relatively high yield per fission (about 2%) and long half-life (10.98 days). A scientific program for measuring fission product yields from 235U,238U and 239Pu targets as a function of bombarding neutron energy (0.1 to 15 MeV) is currently underway using monoenergetic neutron beams produced at the 10 MV Tandem Accelerator at TUNL. Dual-fission chambers are used to determine the rate of fission in targets during activation. Activated targets are counted in highly shielded HPGe detectors over a period of several weeks to identify decaying fission products. To date, data have been collected at neutron bombarding energies 4.6, 9.0, 14.5 and 14.8 MeV. Experimental methods and data reduction techniques are discussed, and some preliminary results are presented.

  12. Tip model of cold fission

    International Nuclear Information System (INIS)

    Goennenwein, F.; Boersig, B.

    1991-01-01

    Cold fission is defined to be the limiting case of nuclear fission where virtually all of the available energy is converted into the total kinetic energy of the fragments. The fragments have, therefore, to be born in or at least close to their respective ground states. Starting from the viewpoint that cold fission corresponds to most compact scission configurations, energy constraints have been exploited to calculate minimum tip distances between the two nascent fragments in binary fission. Crucial input parameters to this tip model of cold fission are the ground-state deformations of fragment nuclei. It is shown that the minimum tip distances being compatible with energy conservation vary strongly with both the mass and charge fragmentation of the fission prone nucleus. The tip distances refer to nuclei with equivalent sharp surfaces. In keeping with the size of the surface width of leptodermous nuclei, only configurations where the tip distances are smaller than a few fm may be considered as valid scission configurations. From a comparison with experimental data on cold fission this critical tip distance appears to be 3.0 fm for the model parameters chosen. Whenever the model calculation yields tip distances being smaller than the critical value, a necessary condition for attaining cold fission is considered to be fulfilled. It is shown that this criterion allows to understand in fair agreement with experiment which mass fragmentations are susceptible to lead to cold fission and which fragment-charge divisions are the most favored in each isobaric mass chain. Being based merely on energy arguments, the model cannot aim at predicting fragment yields in cold fission. However, the tip model proposed appears well suited to delineate the phase space where cold fission phenomena may come into sight. (orig.)

  13. Post-scission fission theory: Neutron emission in fission

    International Nuclear Information System (INIS)

    Madland, D.G.

    1997-01-01

    A survey of theoretical representations of two of the observables in neutron emission in fission is given, namely, the prompt fission neutron spectrum N (E) and the average prompt neutron multiplicity bar ν p . Early representations of the two observables are presented and their deficiencies are discussed. This is followed by summaries and examples of recent theoretical models for the calculation of these quantities. Emphasis is placed upon the predictability and accuracy of the recent models. In particular, the dependencies of N (E) and bar ν p upon the fissioning nucleus and its excitation energy are treated. Recent work in the calculation of the prompt fission neutron spectrum matrix N (E, E n ), where E n is the energy of the neutron inducing fission, is then discussed. Concluding remarks address the current status of our ability to calculate these observables with confidence, the direction of future theoretical efforts, and limitations to current (and future) approaches

  14. Fission fragment driven neutron source

    Science.gov (United States)

    Miller, Lowell G.; Young, Robert C.; Brugger, Robert M.

    1976-01-01

    Fissionable uranium formed into a foil is bombarded with thermal neutrons in the presence of deuterium-tritium gas. The resulting fission fragments impart energy to accelerate deuterium and tritium particles which in turn provide approximately 14 MeV neutrons by the reactions t(d,n).sup.4 He and d(t,n).sup.4 He.

  15. Neutron multiplicity of fission fragments

    Energy Technology Data Exchange (ETDEWEB)

    Abdelrahman, Y S [Physics department, mu` rah university Al-Karak, (Jordan)

    1995-10-01

    The total average neutron multiplicity of the fission fragments produced by the spontaneous fission of {sup 248} Cm has been measured. This measurement has been done by using a new experimental technique. This technique mainly depends on {gamma}-{gamma} coincidence using a very high resolution high purity germanium (HPGe) detector. 2 figs.

  16. Fission throughout the periodic table

    International Nuclear Information System (INIS)

    Moretto, L.G.; Wozniak, G.J.

    1989-04-01

    The dualistic view of fission and evaporation as two distinct compound nucleus processes is substituted with a unified view in which fission, complex fragment emission, and light particle evaporation are seen as different aspects of a single process. 47 refs., 22 figs

  17. Angular finite volume method for solving the multigroup transport equation with piecewise average scattering cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G., E-mail: ansar.calloo@cea.fr, E-mail: jean-francois.vidal@cea.fr, E-mail: romain.le-tellier@cea.fr, E-mail: gerald.rimpault@cea.fr [CEA, DEN, DER/SPRC/LEPh, Saint-Paul-lez-Durance (France)

    2011-07-01

    This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S{sub n} method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)

  18. Finally! A valid test of configural invariance using permutation in multigroup CFA

    NARCIS (Netherlands)

    Jorgensen, T.D.; Kite, B.A.; Chen, P.-Y.; Short, S.D.; van der Ark, L.A.; Wiberg, M.; Culpepper, S.A.; Douglas, J.A.; Wang, W.-C.

    2017-01-01

    In multigroup factor analysis, configural measurement invariance is accepted as tenable when researchers either (a) fail to reject the null hypothesis of exact fit using a χ2 test or (b) conclude that a model fits approximately well enough, according to one or more alternative fit indices (AFIs).

  19. AMZ, multigroup constant library for EXPANDA code, generated by NJOY code from ENDF/B-IV

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Moraes, Marisa de

    1985-01-01

    It is described a library of multigroup constants with 70 energy groups and 37 isotopes to fast reactor calculation. The cross sections, scattering matrices and self-shielding factors were generated by NJOY code and RGENDF interface program, from ENDF/B-IV'S evaluated data. The library is edited in adequated format to be used by EXPANDA code. (M.C.K.) [pt

  20. MCFT: a program for calculating fast and thermal neutron multigroup constants

    International Nuclear Information System (INIS)

    Yang Shunhai; Sang Xinzeng

    1993-01-01

    MCFT is a program for calculating the fast and thermal neutron multigroup constants, which is redesigned from some codes for generation of thermal neutron multigroup constants and for fast neutron multigroup constants adapted on CYBER 825 computer. It uses indifferently as basic input with the evaluated nuclear data contained in the ENDF/B (US), KEDAK (Germany) and UK (United Kingdom) libraries. The code includes a section devoted to the generation of resonant Doppler broadened cross section in the framework of single-or multi-level Breit-Wigner formalism. The program can compute the thermal neutron scattering law S (α, β, T) as the input data in tabular, free gas or diffusion motion form. It can treat up to 200 energy groups and Legendre moments up to P 5 . The output consists of various reaction multigroup constants in all neutron energy range desired in the nuclear reactor design and calculation. Three options in input file can be used by the user. The output format is arbitrary and defined by user with a minimum of program modification. The program includes about 15,000 cards and 184 subroutines. FORTRAN 5 computer language is used. The operation system is under NOS 2 on computer CYBER 825

  1. MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2013-11-08

    The MC2-3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC2-2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (~2000) or hyperfine (~400,000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified isotopic temperatures. The pointwise cross sections are directly used in the hyperfine group calculation whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for two-dimensional whole-core problems to generate region-dependent broad-group cross sections. Multigroup cross sections are written in the ISOTXS format for a user-specified group structure. The code is executable on UNIX, Linux, and PC Windows systems, and its library includes all isotopes of the ENDF/BVII. 0 data.

  2. Angular finite volume method for solving the multigroup transport equation with piecewise average scattering cross sections

    International Nuclear Information System (INIS)

    Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G.

    2011-01-01

    This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S_n method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)

  3. The problem of resonance self-shielding effect in neutron multigroup calculations

    International Nuclear Information System (INIS)

    Wang Qingming; Huang Jinghua

    1991-01-01

    It is not allowed to neglect the resonance self-shielding effect in hybrid blanket and fast reactor neutron designs. The authors discussed the importance as well as the method of considering the resonance self-shielding effect in hybrid blanket and fast reactor neutron multigroup calculations

  4. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

    2014-05-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

  5. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    International Nuclear Information System (INIS)

    Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh

    2014-01-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects

  6. Status of multigroup sensitivity profiles and covariance matrices available from the radiation shielding information center

    International Nuclear Information System (INIS)

    Roussin, R.W.; Drischler, J.D.; Marable, J.H.

    1980-01-01

    In recent years multigroup sensitivity profiles and covariance matrices have been added to the Radiation Shielding Information Center's Data Library Collection (DLC). Sensitivity profiles are available in a single package. DLC-45/SENPRO, and covariance matrices are found in two packages, DLC-44/COVERX and DLC-77/COVERV. The contents of these packages are described and their availability is discussed

  7. A boundary integral equation for boundary element applications in multigroup neutron diffusion theory

    International Nuclear Information System (INIS)

    Ozgener, B.

    1998-01-01

    A boundary integral equation (BIE) is developed for the application of the boundary element method to the multigroup neutron diffusion equations. The developed BIE contains no explicit scattering term; the scattering effects are taken into account by redefining the unknowns. Boundary elements of the linear and constant variety are utilised for validation of the developed boundary integral formulation

  8. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1984-09-01

    This is the tenth issue of a report series on Fission Product Data, which informs us about all the activities in this field, which are planned, ongoing, or have recently been completed. The types of activities included are measurements, compilations and evaluations of: fission product yields (neutron induced and spontaneous fission), neutron reaction cross sections of fission products, data related to the radioactive decay of fission products, delayed neutron data of fission products, lumped fission product data (decay heat, absorption, etc.). There is also a section with recent references relative to fission product nuclear data

  9. Mechanisms of fission neutron emission

    International Nuclear Information System (INIS)

    Maerten, H.

    1991-01-01

    The time evolution in fission is the starting point for discussing not only the main mechanism of fission neutron emission, the evaporation from fully accelerated fragments, but also possible secondary ones connected with dynamical features of nuclear fission. ''Asymptotic'' conditions as relevant for describing the particle release from highly excited, rapidly moving fragments are defined. Corresponding statistical model approaches to fission neutron emission, based on the adequate consideration of the intricate fragment occurrence probability, reproduce most of the experimental data. The remarkable influence of fission modes on neutron observables is analyzed in the framework of a macroscopic-microscopic scission point model consistent with energy conservation. Finally, chances and deficiencies for solving the mechanism puzzle are summarized. (author). 87 refs, 21 figs

  10. Fifty years with nuclear fission

    International Nuclear Information System (INIS)

    Behrens, J.W.; Carlson, A.D.

    1989-01-01

    The news of the discovery of nuclear fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fifieth anniversary of its discovery by holding a topical meeting entitled, ''Fifty Years with Nuclear Fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent development in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicated a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two fully days of sessions (April 27 and 28) at the main site of the NIST in Gaithersburg, Maryland. The wide range of topics covered in this Volume 1 by this topical meeting included plenary invited, and contributed sessions entitled: Preclude to the First Chain Reaction -- 1932 to 1942; Early Fission Research -- Nuclear Structure and Spontaneous Fission; 50 Years of Fission, Science, and Technology; Nuclear Reactors, Secure Energy for the Future; Reactors 1; Fission Science 1; Safeguards and Space Applications; Fission Data; Nuclear Fission -- Its Various Aspects; Theory and Experiments in Support of Theory; Reactors and Safeguards; and General Research, Instrumentation, and By-Product. The individual papers have been cataloged separately

  11. Multigroup and coupled forward-adjoint Monte Carlo calculation efficiencies for secondary neutron doses from proton beams

    International Nuclear Information System (INIS)

    Kelsey IV, Charles T.; Prinja, Anil K.

    2011-01-01

    We evaluate the Monte Carlo calculation efficiency for multigroup transport relative to continuous energy transport using the MCNPX code system to evaluate secondary neutron doses from a proton beam. We consider both fully forward simulation and application of a midway forward adjoint coupling method to the problem. Previously we developed tools for building coupled multigroup proton/neutron cross section libraries and showed consistent results for continuous energy and multigroup proton/neutron transport calculations. We observed that forward multigroup transport could be more efficient than continuous energy. Here we quantify solution efficiency differences for a secondary radiation dose problem characteristic of proton beam therapy problems. We begin by comparing figures of merit for forward multigroup and continuous energy MCNPX transport and find that multigroup is 30 times more efficient. Next we evaluate efficiency gains for coupling out-of-beam adjoint solutions with forward in-beam solutions. We use a variation of a midway forward-adjoint coupling method developed by others for neutral particle transport. Our implementation makes use of the surface source feature in MCNPX and we use spherical harmonic expansions for coupling in angle rather than solid angle binning. The adjoint out-of-beam transport for organs of concern in a phantom or patient can be coupled with numerous forward, continuous energy or multigroup, in-beam perturbations of a therapy beam line configuration. Out-of-beam dose solutions are provided without repeating out-of-beam transport. (author)

  12. AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Lucius, J.L.; Petrie, L.M.; Ford, W.E. III; White, J.E.; Wright, R.Q.

    1976-03-01

    AMPX is a modular system for producing coupled multigroup neutron-gamma cross section sets. Basic neutron and gamma cross-section data for AMPX are obtained from ENDF/B libraries. Most commonly used operations required to generate and collapse multigroup cross-section sets are provided in the system. AMPX is flexibly dimensioned; neutron group structures, and gamma group structures, and expansion orders to represent anisotropic processes are all arbitrary and limited only by available computer core and budget. The basic processes provided will (1) generate multigroup neutron cross sections; (2) generate multigroup gamma cross sections; (3) generate gamma yields for gamma-producing neutron interactions; (4) combine neutron cross sections, gamma cross sections, and gamma yields into final ''coupled sets''; (5) perform one-dimensional discrete ordinates transport or diffusion theory calculations for neutrons and gammas and, on option, collapse the cross sections to a broad-group structure, using the one-dimensional results as weighting functions; (6) plot cross sections, on option, to facilitate the ''evaluation'' of a particular multigroup set of data; (7) update and maintain multigroup cross section libraries in such a manner as to make it not only easy to combine new data with previously processed data but also to do it in a single pass on the computer; and (8) output multigroup cross sections in convenient formats for other codes. (auth)

  13. Research of the application of multi-group libraries based on ENDF/B-VII library in the reactor design

    International Nuclear Information System (INIS)

    Mi Aijun; Li Junjie

    2010-01-01

    In this paper the multi-group libraries were constructed by processing ENDF/B-VII neutron incident files into multi-group structure, and the application of the multi-group libraries in the pressurized-water reactor(PWR) design was studied. The construction of the multi-group library is realized by using the NJOY nuclear data processing system. The code can process the neutron cross section files form ENDF format to MATXS format which was required in SN code. Two dimension transport theory code of discrete ordinates DORT was used to verify the multi-group libraries and the method of the construction by comparing calculations for some representative benchmarks. We made the PWR shielding calculation by using the multi-group libraries and studied the influence of the parameters involved during the construction of the libraries such as group structure, temperatures and weight functions on the shielding design of the PWR. This work is the preparation for the construction of the multi-group library which will be used in PWR shielding design in engineering. (authors)

  14. How fission was discovered

    International Nuclear Information System (INIS)

    Fluegge, S.

    1989-01-01

    After the great survey of neutron induced radioactivity by Fermi and co-workers, the laboratories in Paris and Berlin-Dahlen tried to disentangle the complex results found in uranium. At that time neutron sources were small, activities low, and equipment very simple. Chemistry beyond uranium still was unknown. Hahn and Meitner believed to have observed three transuranic isomeric chains, a doubtful result even then. Early in 1938, Curie and Savic in Paris found an activity interpreted to be actinium, and Hahn and Meitner another to be radium. Both interpretations seemed impossible from energy considerations. Hahn and Strassmann, therefore, continued this work and succeeded to separate the new activity from radium. There remained no doubt that a barium isotope had been produced, the uranium nucleus splitting in the yet-unknown process we now call fission

  15. Fifty years with nuclear fission

    International Nuclear Information System (INIS)

    Behrens, J.W.; Carlson, A.D.

    1989-01-01

    The news of the discovery of nucler fission, by Otto Hahn and Fritz Strassmann in Germany, was brought to the United States by Niels Bohr in January 1939. Since its discovery, the United States, and the world for that matter, has never been the same. It therefore seemed appropriate to acknowledge the fiftieth anniversary of its discovery by holding a topical meeting entitled, ''Fifty years with nuclear fission,'' in the United States during the year 1989. The objective of the meeting was to bring together pioneers of the nuclear industry and other scientists and engineers to report on reminiscences of the past and on the more recent developments in fission science and technology. The conference highlighted the early pioneers of the nuclear industry by dedicating a full day (April 26), consisting of two plenary sessions, at the National Academy of Sciences (NAS) in Washington, DC. More recent developments in fission science and technology in addition to historical reflections were topics for two full days of sessions (April 27 and 28) at the main sites of the NIST in Gaithersburg, Maryland. The wide range of topics covered by Volume 2 of this topical meeting included plenary invited, and contributed sessions entitled, Nuclear fission -- a prospective; reactors II; fission science II; medical and industrial applications by by-products; reactors and safeguards; general research, instrumentation, and by-products; and fission data, astrophysics, and space applications. The individual papers have been cataloged separately

  16. Nuclear fission and fission-product spectroscopy: 3. International workshop on nuclear fission and fission-product spectroscopy

    International Nuclear Information System (INIS)

    Goutte, Heloise; Fioni, Gabriele; Faust, Herbert; Goutte, Dominique

    2005-01-01

    The present book contains the proceedings of the third workshop in a series of workshops previously held in Seyssins in 1994 and 1998. The meeting was jointly organized by different divisions of CEA and two major international laboratories. In the opening address, Prof. B. Bigot, the French High Commissioner for Atomic Energy, outlined France's energy policy for the next few decades. He emphasized the continuing progress of nuclear fission in both technical and economic terms, allowing it to contribute to the energy needs of the planet even more in the future than it does today. Such progress implies a very strong link between fundamental and applied research based on experimental and theoretical approaches. The workshop gathered the different nuclear communities studying the fission process, including topics as the following: - nuclear fission experiments, - spectroscopy of neutron rich nuclei, - fission data evaluation, - theoretical aspects of nuclear fission, - and innovative nuclear systems and new facilities. The scientific program was suggested by an International Advisory Committee. About 100 scientists from 13 different countries attended the conference in the friendly working atmosphere of the Castle of Cadarache in the heart of the Provence. The proceedings of the workshop were divided into 11 sections addressing the following subject matters: 1. Cross sections and resonances (5 papers); 2. Fission at higher energies - I (5 papers); 3. Fission: mass and charge yields (4 papers); 4. Light particles and cluster emission (4 papers); 5. Spectroscopy of neutron rich nuclei (5 papers); 6. Resonances, barriers, and fission times (5 papers); 7. Fragment excitation and neutron emission (4 papers); 8. Mass and energy distributions (4 papers); 9. Needs for nuclear data and new facilities - I (4 papers); 10. Angular momenta and fission at higher Energies - II (3 papers); 11. New facilities - II (2 papers). A poster session of 8 presentations completed the workshop

  17. Gamma Radiation from Fission Fragments

    International Nuclear Information System (INIS)

    Higbie, Jack

    1969-10-01

    The gamma radiation from the fragments of the thermal neutron fission of 235 U has been investigated, and the preliminary data are presented here with suggestions for further lines of research and some possible interpretations of the data. The data have direct bearing on the fission process and the mode of fragment de-excitation. The parameters measured are the radiation decay curve for the time interval (1 - 7) x 10 -10 sec after fission, the photon yield, the total gamma ray energy yield, and the average photon energy. The last three quantities are measured as a function of the fragment mass

  18. Energy partition in nuclear fission

    International Nuclear Information System (INIS)

    Ruben, A.; Maerten, H.; Seeliger, D.

    1990-01-01

    A scission point model (two spheroid model TSM) including semi-empirical temperature-dependent shell correction energies for deformed fragments at scission is presented. It has been used to describe the mass-asymmetry-dependent partition of the total energy release on both fragments from spontaneous and induced fission. Characteristic trends of experimental fragment energy and neutron multiplicity data as function of incidence energy in the Th-Cf region of fissioning nuclei are well reproduced. Based on model applications, information on the energy dissipated during the descent from second saddle of fission barrier to scission point have been deduced. (author). 39 refs, 13 figs

  19. Gamma Radiation from Fission Fragments

    Energy Technology Data Exchange (ETDEWEB)

    Higbie, Jack

    1969-10-15

    The gamma radiation from the fragments of the thermal neutron fission of {sup 235}U has been investigated, and the preliminary data are presented here with suggestions for further lines of research and some possible interpretations of the data. The data have direct bearing on the fission process and the mode of fragment de-excitation. The parameters measured are the radiation decay curve for the time interval (1 - 7) x 10{sup -10} sec after fission, the photon yield, the total gamma ray energy yield, and the average photon energy. The last three quantities are measured as a function of the fragment mass.

  20. Fission-product source terms

    International Nuclear Information System (INIS)

    Lorenz, R.A.

    1981-01-01

    This presentation consists of a review of fission-product source terms for light water reactor (LWR) fuel. A source term is the quantity of fission products released under specified conditions that can be used to calculate the consequences of the release. The source term usually defines release from breached fuel-rod cladding but could also describe release from the primary coolant system, the reactor containment shell, or the site boundary. The source term would be different for each locality, and the chemical and physical forms of the fission products could also differ

  1. The spectroscopy of fission fragments

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, W.R. [Department of Physics and Astronomy, University of Manchester, Manchester, M13 9PL (United Kingdom); Collaboration: La Direction des Sciences de la Matiere du CEA (FR); Le Fonds National de la Recherche Scientifique de Belgique (BE)

    1998-12-31

    High-resolution measurements on {gamma} rays from fission fragments have provided a rich source of information, unobtainable at the moment in any other way, on the spectroscopy of neutron-rich nuclei. In recent years important data have been obtained on the yrast- and near yrast-structure of neutron-rich fission fragments. We discuss the scope of measurements which can be made on prompt gamma rays from secondary fission fragments, the techniques used in the experiments and some results recently obtained. (author) 24 refs., 8 figs., 1 tab.

  2. The spectroscopy of fission fragments

    International Nuclear Information System (INIS)

    Phillips, W.R.

    1998-01-01

    High-resolution measurements on γ rays from fission fragments have provided a rich source of information, unobtainable at the moment in any other way, on the spectroscopy of neutron-rich nuclei. In recent years important data have been obtained on the yrast- and near yrast-structure of neutron-rich fission fragments. We discuss the scope of measurements which can be made on prompt gamma rays from secondary fission fragments, the techniques used in the experiments and some results recently obtained. (author)

  3. Burn-up physics in a coupled Hammer-Technion/Cinder-2 system and ENDF/B-V aggregate fission product thermal cross section validation

    International Nuclear Information System (INIS)

    Santos, A. dos.

    1990-01-01

    The new methodology developed in this work has the following purposes: a) to implement a burnup capability into the HAMMER-TECHNION/9/computer code by using the CINDER-2/10/computer code to perform the transmutation analysis for the actinides and fission products; b) to implement a reduced version of the CINDER-2 fission product chain structure to treat explicity nearly 99% of all original CINDER-2 fission product absorption in a typical PWR unit cell; c) to treat the effect of the fission product neutron absorption in an unit cell in a multigroup basis; d) to develop a tentative validation procedure for the ENOF/C-V stable and long-lived fission product nuclear data based on the available experimental data/11-14/. The analysis will be performed by using the reduce chain in the coupled system CINDER-2 to generate the time dependent effective four group cross sections for actinides and fission products and CINDER-2 to perform the complete transmutation analysis with its built-in chain structure. (author)

  4. Fission dynamics with systems of intermediate fissility

    Indian Academy of Sciences (India)

    results concerning nuclear dissipation and fission time-scale obtained from several of these studies. In particular ... alent to the assumption that fission is delayed, namely, that the fission probability is not .... parameters to be adjusted on the experimental data. ..... (b) Time distribution of all fission events for the 132Ce nucleus.

  5. Fission gas in thoria

    Energy Technology Data Exchange (ETDEWEB)

    Kuganathan, Navaratnarajah, E-mail: n.kuganathan@imperial.ac.uk [Department of Materials, Faculty of Engineering, Imperial College, London, SW7 2AZ (United Kingdom); Ghosh, Partha S. [Material Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Galvin, Conor O.T. [Department of Materials, Faculty of Engineering, Imperial College, London, SW7 2AZ (United Kingdom); Arya, Ashok K. [Material Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Dutta, Bijon K. [Homi Bhabha National Institute, Trombay, Mumbai 400 094 (India); Dey, Gautam K. [Material Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Grimes, Robin W. [Department of Materials, Faculty of Engineering, Imperial College, London, SW7 2AZ (United Kingdom)

    2017-03-15

    The fission gases Xe and Kr, formed during normal reactor operation, are known to degrade fuel performance, particularly at high burn-up. Using first-principles density functional theory together with a dispersion correction (DFT + D), in ThO{sub 2} we calculate the energetics of neutral and charged point defects, the di-vacancy (DV), different neutral tri-vacancies (NTV), the charged tetravacancy (CTV) defect cluster geometries and their interaction with Xe and Kr. The most favourable incorporation point defect site for Xe or Kr in defective ThO{sub 2} is the fully charged thorium vacancy. The lowest energy NTV in larger supercells of ThO{sub 2} is NTV3, however, a single Xe atom is most stable when accommodated within a NTV1. The di-vacancy (DV) is a significantly less favoured incorporation site than the NTV1 but the CTV offers about the same incorporation energy. Incorporation of a second gas atom in a NTV is a high energy process and more unfavourable than accommodation within an existing Th vacancy. The bi-NTV (BNTV) cluster geometry studied will accommodate one or two gas atoms with low incorporation energies but the addition of a third gas atom incurs a high energy penalty. The tri-NTV cluster (TNTV) forms a larger space which accommodates three gas atoms but again there is a penalty to accommodate a fourth gas atom. By considering the energy to form the defect sites, solution energies were generated showing that in ThO{sub 2−x} the most favourable solution equilibrium site is the NTV1 while in ThO{sub 2} it is the DV. - Highlights: • We have considered Xe and Kr in point defects and defect clusters (neutral and charged) using Density Functional Theory (DFT) with a dispersion correction. • The most favourable charge state for a point defect (vacancy or interstitial) is that with full ionic charge and we have found that in all cases gas atoms occupy the fully charged vacancy sites. • The number of fission gas atoms accommodated in ThO{sub 2} is

  6. Collapsing of multigroup cross sections in optimization problems solved by means of the maximum principle of Pontryagin

    International Nuclear Information System (INIS)

    Anton, V.

    1979-05-01

    A new formulation of multigroup cross section collapsing based on the conservation of point or zone value of hamiltonian is presented. This attempt is proper to optimization problems solved by means of maximum principle of Pontryagin. (author)

  7. COMPAR: system to compare multigroup cross sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 AND XLACS

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1987-11-01

    A system, composed by the computer programs COMPAR and its interfaces, developed for comparing multigroup cross sections calculated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS, is presented. (author)

  8. COMPAR: A system for comparing multigroup cross-sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1988-02-01

    A system consisting of the COMPAR computer program and its interfaces which was developed for comparing multigroup cross-sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS is presented. (author). 13 refs

  9. Chemical effects of fission recoils

    International Nuclear Information System (INIS)

    Meisels, G.G.; Freeman, J.P.; Gregory, J.P.; Richardson, W.C.; Sroka, G.J.

    1978-01-01

    The production of nitrogen from nitrous oxide at high density was employed to investigate the energy deposition efficiency of fission recoils produced from fission of U 235 in uranium-palladium foils clad with platinum. Nitrogen production varied linearly with fission recoil dose from 1.1 x 10 20 to 9.0 x 10 20 eV, and was independent of density between 12.5 and 127.5 g l -1 N 2 O. 16.2 +- 0.8% of the fission recoil energy was deposited external to the foil. Electron microprobe analysis showed some unevenness of new foil and polymer buildup on the surface after irradiation of ethylene-oxygen mixtures. Subsequent irradiation in the presence of nitrous oxide restored some of the original efficiency. This is ascribed to chemical oxidation of the polymer induced by reactive intermediates produced from nitrous oxide. (author)

  10. Extraction chromatography of fission products

    International Nuclear Information System (INIS)

    Bonnevie-Svendsen, M.; Goon, K.

    1978-01-01

    Various cases of using extraction chromatography during analysis of fission products are reviewed. The use of the extraction chromatography method is considered while analysing reprocessed products of nuclear fuel for quantitative radiochemical analysis and control of fission product and actinoide separation during extraction and their chemical state in production solutions. The method is used to obtain pure fractions of typical burnup monitors (neodymium, molybdenum, cerium, cesium, europium, lanthanides) during determination of nuclear fuel burnup degree. While studying the nature of nuclear reactions the method is used to separate quickly short-life isotopes, to purify β-radiator fractions before measuring their half-life periods, to enrich isotopes forming with low output during fission. Examples of using extraction chromatography are given to separate long half-life or stable fission products from spent solutions, to control environment object contamination

  11. Chemical Production using Fission Fragments

    International Nuclear Information System (INIS)

    Dawson, J. K.; Moseley, F.

    1960-01-01

    Some reactor design considerations of the use of fission recoil fragment energy for the production of chemicals of industrial importance have been discussed previously in a paper given at the Second United Nations International Conference on the Peaceful Uses of Atomic Energy [A/Conf. 15/P.76]. The present paper summarizes more recent progress made on this topic at AERE, Harwell. The range-energy relationship for fission fragments is discussed in the context of the choice of fuel system for a chemical production reactor, and the experimental observation of a variation of chemical effect along the length of a fission fragment track is described for the irradiation of nitrogen-oxygen mixtures. Recent results are given on the effect of fission fragments on carbon monoxide-hydrogen gas mixtures and on water vapour. No system investigated to date shows any outstanding promise for large-scale chemical production. (author) [fr

  12. Physics and chemistry of fission

    International Nuclear Information System (INIS)

    1979-01-01

    Full text: In the pleasant and hospitable atmosphere of the Kernforschungsanlage Juelich in the Federal Republic of Germany, the IAEA symposium on the Physics and Chemistry of Fission took place. Almost 200 scientists attended, 154 abstracts were submitted, and 57 papers presented, but more important than the numbers was the quality of the contributions and the progress reported at the symposium. The neutron was discovered almost 50 years ago; 40 years ago the idea of nuclear fission was born. Since then, a number of laboratories have worked hard to explain the phenomenon of fission One would expect that by now scientists would know exactly what happens in a nucleus before and during the process of fission, particularly as there are hundreds of power and research reactors in operation, and fission of uranium isotopes is the basis of their functioning. At first glance, fission seems a simple process: a neutron hits and penetrates the uranium nucleus which becomes excited, i.e. has a surplus of energy. One way to get rid of this energy is for the nucleus to split into two parts; additional products of this process are energy and more neutrons. Nature, however, seems to dislike such straightforward explanations. In the case of fission, scientists have observed a number of phenomena which disagree with a simple model. Sometimes, a nucleus will split into two parts without being 'attacked' by a neutron; this spontaneous fission opens up a new line of fission research and several contributions at the symposium reported on sophisticated experiments designed to unravel some of its specific details. Sometimes, a fissioning nucleus will emit another particle: ternary fission has become a powerful tool for studying the properties of nuclei during the fission process. For the scientist, it is fascinating to observe how the nucleus behaves during fission. They invent models which are supposed to reproduce the most probable course of events leading to fission. In one of these

  13. International handling of fissionable material

    International Nuclear Information System (INIS)

    1975-01-01

    The opinion of the ministry for foreign affairs on international handling of fissionable materials is given. As an introduction a survey is given of the possibilities to produce nuclear weapons from materials used in or produced by power reactors. Principles for international control of fissionable materials are given. International agreements against proliferation of nuclear weapons are surveyed and methods to improve them are proposed. (K.K.)

  14. Theory of nuclear fission: a review

    International Nuclear Information System (INIS)

    Mosel, U.

    1976-01-01

    General properties of nuclear fission are reviewed and related to our present knowledge of fission theory. For this purpose the basic reasons for the shape of the fission barriers are discussed and their consequences compared with experimental results on barrier shapes and structures. Special emphasis is put on the asymmetry of the fission barriers and mass-distributions and its relation to the shells of the nascent fragment shells. Finally the problem of calculating fission cross sections is discussed

  15. Spontaneous fission of 259Md

    International Nuclear Information System (INIS)

    Hulet, E.K.; Wild, J.F.; Lougheed, R.W.; Baisden, P.A.; Landrum, J.H.; Dougan, R.J.; Mustafa, M.; Ghiorso, A.; Nitschke, J.M.

    1979-01-01

    The mass and kinetic energy distributions of fission fragments from the spontaneous fission of th newly discovered nuclide 259 Md were obtained. 259 Md was identified as the E. C. daughter of 259 No, and was found to decay entirely (> 95%) by spontaneous fission with a 95-min half-life. From the kinetic energies measured for 397 pairs of coincident fragments, a mass distribution was derived that is symmetric with sigma = 13 amu. 259 Md, together with 258 Fm and 259 Fm, form a select group of three nuclides whose mass division in spontaneous fission is highly symmetric. Unlike the total-kinetic-energy (TKE) distributions of 258 Fm and 259 Fm, which peak at approx. = to 240 MeV, this distribution for 259 Md is broad and is 50 MeV lower in energy. Analysis of the mass and energy distributions shows that events near mass symmetry also exhibit a broad TKE distribution, with one-third of the symmetric events having TKEs less than 200 MeV. The associated of low TKEs with symmetric mass division in the fission of very heavy actinides is anomalous and inconsistent with theories based upon the emergence of fragment shells near the scission point. Either three-body fragmentation or peculiar fragment shapes are assumed as the cause for the large consumption of Coulomb energy observed for a significant fraction of symmetric fissions in 259 Md. 6 figures

  16. A stochastic approach to fission

    International Nuclear Information System (INIS)

    Boilley, D.; Suraud, E.; Abe, Yasuhisa

    1992-01-01

    A microscopically derived Langevin equation is applied to thermally induced nuclear fission. An important memory effect is pointed out. A strong friction coefficient, calculated from microscopic quantities, tends to decrease the stationary limit of the fission rate and to increase the transient time. Fission was described as a diffusion over a barrier of a collective variable, and a Langevin Equation (LE) was used to study the phenomenon. A study of the stationary flow over the saddle point with a Fokker-Planck Equation (FPE), equivalent to the LE was used to give formula for the stationary fission rate (or reaction rate for the chemistry applications). More recently, a complete study of the fission process was performed numerically with both FPE and LE. A long transient time, that could allow more pre-scission neutrons to evaporate, was pointed out. The derivation of this new LE is recalled, followed by the description of the memory dependence and by the effect of a large friction coefficient on the fission rate. (author) 6 refs., 3 figs

  17. Modelling isothermal fission gas release

    International Nuclear Information System (INIS)

    Uffelen, P. van

    2002-01-01

    The present paper presents a new fission gas release model consisting of two coupled modules. The first module treats the behaviour of the fission gas atoms in spherical grains with a distribution of grain sizes. This module considers single atom diffusion, trapping and fission induced re-solution of gas atoms associated with intragranular bubbles, and re-solution from the grain boundary into a few layers adjacent to the grain face. The second module considers the transport of the fission gas atoms along the grain boundaries. Four mechanisms are incorporated: diffusion controlled precipitation of gas atoms into bubbles, grain boundary bubble sweeping, re-solution of gas atoms into the adjacent grains and gas flow through open porosity when grain boundary bubbles are interconnected. The interconnection of the intergranular bubbles is affected both by the fraction of the grain face occupied by the cavities and by the balance between the bubble internal pressure and the hydrostatic pressure surrounding the bubbles. The model is under validation. In a first step, some numerical routines have been tested by means of analytic solutions. In a second step, the fission gas release model has been coupled with the FTEMP2 code of the Halden Reactor Project for the temperature distribution in the pellets. A parametric study of some steady-state irradiations and one power ramp have been simulated successfully. In particular, the Halden threshold for fission gas release and two simplified FUMEX cases have been computed and are summarised. (author)

  18. Status of fission power

    International Nuclear Information System (INIS)

    Levenson, M.

    1977-01-01

    Fission energy is reviewed from the viewpoints of technology, economics, politics, manufacturers, consumers, and foreign countries. Technically, the reactor program is operating and the light water reactor industry shows signs of maturing, although recent business has been disappointing. Marketing of gas-cooled reactors depends, not on technical, but economic and political issues. Liquid metal fast breeder reactors have been demonstrated worldwide, while the gas-cooled fast breeder remains an undemonstrated option. Nuclear plants, currently costing the same as coal plants with scrubbers, are the cheapest option for utilities because most of the cost is imbedded. The defeat of nuclear initiatives in seven states indicates that public feeling is not as anti-nuclear as opponents to nuclear power claim. The harshness of last winter demonstrated the advantages of a power source that is not so sensitive to the weather for reliable operation and transport, as well as low cost energy. Other nations are proceeding to build a nuclear capability, which the U.S. may jeopardize because of concerns about the fuel cycle, nuclear waste disposal, uranium reserves, and nuclear proliferation

  19. Fission fragment angular distributions and fission cross section validation

    International Nuclear Information System (INIS)

    Leong, Lou Sai

    2013-01-01

    The present knowledge of angular distributions of neutron-induced fission is limited to a maximal energy of 15 MeV, with large discrepancies around 14 MeV. Only 238 U and 232 Th have been investigated up to 100 MeV in a single experiment. The n-TOF Collaboration performed the fission cross section measurement of several actinides ( 232 Th, 235 U, 238 U, 234 U, 237 Np) at the n-TOF facility using an experimental set-up made of Parallel Plate Avalanche Counters (PPAC), extending the energy domain of the incident neutron above hundreds of MeV. The method based on the detection of the 2 fragments in coincidence allowed to clearly disentangle the fission reactions among other types of reactions occurring in the spallation domain. I will show the methods we used to reconstruct the full angular resolution by the tracking of fission fragments. Below 10 MeV our results are consistent with existing data. For example in the case of 232 Th, below 10 MeV the results show clearly the variation occurring at the first (1 MeV) and second (7 MeV) chance fission, corresponding to transition states of given J and K (total spin and its projection on the fission axis), and a much more accurate energy dependence at the 3. chance threshold (14 MeV) has been obtained. In the spallation domain, above 30 MeV we confirm the high anisotropy revealed in 232 Th by the single existing data set. I'll discuss the implications of this finding, related to the low anisotropy exhibited in proton-induced fission. I also explore the critical experiments which is valuable checks of nuclear data. The 237 Np neutron-induced fission cross section has recently been measured in a large energy range (from eV to GeV) at the n-TOF facility at CERN. When compared to previous measurements, the n-TOF fission cross section appears to be higher by 5-7 % beyond the fission threshold. To check the relevance of n-TOF data, we simulate a criticality experiment performed at Los Alamos with a 6 kg sphere of 237 Np. This

  20. Radiochemistry and the Study of Fission

    Energy Technology Data Exchange (ETDEWEB)

    Rundberg, Robert S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-11-14

    These are slides from a lecture given at UC Berkeley. Radiochemistry has been used to study fission since its discovery. Radiochemical methods are used to determine cumulative mass yields. These measurements have led to the two-mode fission hypothesis to model the neutron energy dependence of fission product yields. Fission product yields can be used for the nuclear forensics of nuclear explosions. The mass yield curve depends on both the fuel and the neutron spectrum of a device. Recent studies have shown that the nuclear structure of the compound nucleus can affect the mass yield distribution. The following topics are covered: In the beginning: the discovery of fission; forensics using fission products: what can be learned from fission products, definitions of R-values and Q-values, fission bases, K-factors and fission chambers, limitations; the neutron energy dependence of the mass yield distribution (the two mode fission hypothesis); the influence of nuclear structure on the mass yield distribution. In summary: Radiochemistry has been used to study fission since its discovery. Radiochemical measurement of fission product yields have provided the highest precision data for developing fission models and for nuclear forensics. The two-mode fission hypothesis provides a description of the neutron energy dependence of the mass yield curve. However, data is still rather sparse and more work is needed near second and third chance fission. Radiochemical measurements have provided evidence for the importance of nuclear states in the compound nucleus in predicting the mass yield curve in the resonance region.

  1. Radiochemistry and the Study of Fission

    International Nuclear Information System (INIS)

    Rundberg, Robert S.

    2016-01-01

    These are slides from a lecture given at UC Berkeley. Radiochemistry has been used to study fission since its discovery. Radiochemical methods are used to determine cumulative mass yields. These measurements have led to the two-mode fission hypothesis to model the neutron energy dependence of fission product yields. Fission product yields can be used for the nuclear forensics of nuclear explosions. The mass yield curve depends on both the fuel and the neutron spectrum of a device. Recent studies have shown that the nuclear structure of the compound nucleus can affect the mass yield distribution. The following topics are covered: In the beginning: the discovery of fission; forensics using fission products: what can be learned from fission products, definitions of R-values and Q-values, fission bases, K-factors and fission chambers, limitations; the neutron energy dependence of the mass yield distribution (the two mode fission hypothesis); the influence of nuclear structure on the mass yield distribution. In summary: Radiochemistry has been used to study fission since its discovery. Radiochemical measurement of fission product yields have provided the highest precision data for developing fission models and for nuclear forensics. The two-mode fission hypothesis provides a description of the neutron energy dependence of the mass yield curve. However, data is still rather sparse and more work is needed near second and third chance fission. Radiochemical measurements have provided evidence for the importance of nuclear states in the compound nucleus in predicting the mass yield curve in the resonance region.

  2. Cassandre : a two-dimensional multigroup diffusion code for reactor transient analysis

    International Nuclear Information System (INIS)

    Arien, B.; Daniels, J.

    1986-12-01

    CASSANDRE is a two-dimensional (x-y or r-z) finite element neutronics code with thermohydraulics feedback for reactor dynamics prior to the disassembly phase. It uses the multigroup neutron diffusion theory. Its main characteristics are the use of a generalized quasistatic model, the use of a flexible multigroup point-kinetics algorithm allowing for spectral matching and the use of a finite element description. The code was conceived in order to be coupled with any thermohydraulics module, although thermohydraulics feedback is only considered in r-z geometry. In steady state criticality search is possible either by control rod insertion or by homogeneous poisoning of the coolant. This report describes the main characterstics of the code structure and provides all the information needed to use the code. (Author)

  3. COMPAR, NJOY, GROUPIE, FLANGE-2, ETOG-3, XLACS Multigroup Cross-Sections General Comparison

    International Nuclear Information System (INIS)

    Anaf, Jaime; Chalhoub, E.S.

    1990-01-01

    1 - Description of program or function: A system for comparing multigroup cross sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS. This system comprises the COMPAR program and interface (auxiliary) programs developed for each of the programs under consideration. These are REDCOMP for GROUPIE, FLACOMP for FLANGE-II, ETOCOMP for ETOG-3 and XLACOMP for XLACS. For the NJOY program there is RGENDF, a program developed apart from this system. It is a modular system in which the inclusion of new multigroup cross section generating program requires no more than the development of a new interface module. 2 - Method of solution: Refer to comments in main routine. 3 - Restrictions on the complexity of the problem: Refer to comments in main routine

  4. The LAW Library -- A multigroup cross-section library for use in radioactive waste analysis calculations

    International Nuclear Information System (INIS)

    Greene, N.M.; Arwood, J.W.; Wright, R.Q.; Parks, C.V.

    1994-08-01

    The 238-group LAW Library is a new multigroup neutron cross-section library based on ENDF/B-V data, with five sets of data taken from ENDF/B-VI ( 14 N 7 , 15 N 7 , 16 O 8 , 154Eu 63 , and 155 Eu 63 ). These five nuclides are included because the new evaluations are thought to be superior to those in Version 5. The LAW Library contains data for over 300 materials and will be distributed by the Radiation Shielding Information Center, located at Oak Ridge National Laboratory. It was generated for use in neutronics calculations required in radioactive waste analyses, although it has equal utility in any study requiring multigroup neutron cross sections

  5. Second order time evolution of the multigroup diffusion and P1 equations for radiation transport

    International Nuclear Information System (INIS)

    Olson, Gordon L.

    2011-01-01

    Highlights: → An existing multigroup transport algorithm is extended to be second-order in time. → A new algorithm is presented that does not require a grey acceleration solution. → The two algorithms are tested with 2D, multi-material problems. → The two algorithms have comparable computational requirements. - Abstract: An existing solution method for solving the multigroup radiation equations, linear multifrequency-grey acceleration, is here extended to be second order in time. This method works for simple diffusion and for flux-limited diffusion, with or without material conduction. A new method is developed that does not require the solution of an averaged grey transport equation. It is effective solving both the diffusion and P 1 forms of the transport equation. Two dimensional, multi-material test problems are used to compare the solution methods.

  6. Correction of multigroup cross sections for resolved resonance interference in mixed absorbers

    International Nuclear Information System (INIS)

    Williams, M.L.

    1982-07-01

    The effect that interference between resolved resonances has on averaging multigroup cross sections is examined for thermal reactor-type problems. A simple and efficient numerical scheme is presented to correct a preprocessed multigroup library for interference effects. The procedure is implemented in a design oriented lattice physics computer code and compared with rigorous numerical calculations. The approximate method for computing resonance interference correction factors is applied to obtaining fine-group cross sections for a homogeneous uranium-plutonium mixture and a uranium oxide lattice. It was found that some fine group cross sections are changed by more than 40% due to resonance interference. The change in resonance interference correction factors due to burnup of a PWR fuel pin is examined and found to be small. The effect of resolved resonance interference on collapsed broad-group cross sections for thermal reactor calculations is discussed

  7. MC2-2: a code to calculate fast neutron spectra and multigroup cross sections

    International Nuclear Information System (INIS)

    Henryson, H. II; Toppel, B.J.; Stenberg, C.G.

    1976-06-01

    MC 2 -2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC 2 -2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC 2 -2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC 2 -2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC 2 -2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers

  8. On efficiently computing multigroup multi-layer neutron reflection and transmission conditions

    International Nuclear Information System (INIS)

    Abreu, Marcos P. de

    2007-01-01

    In this article, we present an algorithm for efficient computation of multigroup discrete ordinates neutron reflection and transmission conditions, which replace a multi-layered boundary region in neutron multiplication eigenvalue computations with no spatial truncation error. In contrast to the independent layer-by-layer algorithm considered thus far in our computations, the algorithm here is based on an inductive approach developed by the present author for deriving neutron reflection and transmission conditions for a nonactive boundary region with an arbitrary number of arbitrarily thick layers. With this new algorithm, we were able to increase significantly the computational efficiency of our spectral diamond-spectral Green's function method for solving multigroup neutron multiplication eigenvalue problems with multi-layered boundary regions. We provide comparative results for a two-group reactor core model to illustrate the increased efficiency of our spectral method, and we conclude this article with a number of general remarks. (author)

  9. Verification of KARMA GEOM/TRPT Module with Given Multi-group Cross Sections

    International Nuclear Information System (INIS)

    Koo, Bon Seung; Hong, Ser Gi; Song, Jae Seung

    2009-01-01

    KAERI has developed a two-dimensional multigroup transport theory code KARMA (Kernel Analyzer by Ray-tracing Method for Fuel Assembly). KARMA uses CMFD (Coarse Mesh Finite Difference) accelerated MOC (Method of Characteristics) method for burnup calculation on a single fuel pin, a fuel assembly and a core consisting of rectangular array of fuel pins. KARMA code intends to be employed as a nuclear design tool for the Korean commercial pressurizer water reactor. Prior to the application to actual assembly designs, the code has to be approved by regularity agency. Therefore, it is essential that the reliability of KARMA code should be sufficiently evaluated against well-defined benchmark problems. In this paper, verification of GEOM/TRPT modules of KARMA was performed to confirm a reliability of the KARMA transport solution via comparisons with Monte Carlo calculations by using a consistent set of multi-group macroscopic cross-sections

  10. CLUB - a multigroup integral transport theory code for lattice calculations of PHWR cells

    International Nuclear Information System (INIS)

    Krishnani, P.D.

    1992-01-01

    The computer code CLUB has been developed to calculate lattice parameters as a function of burnup for a pressurised heavy water reactor (PHWR) lattice cell containing fuel in the form of cluster. It solves the multigroup integral transport equation by the method based on combination of small scale collision probability (CP) method and large scale interface current technique. The calculations are performed by using WIMS 69 group cross section library or its condensed versions of 27 or 28 group libraries. It can also compute Keff from the given geometrical buckling in the input using multigroup diffusion theory in fundamental mode. The first order differential burnup equations can be solved by either Trapezoidal rule or Runge-Kutta method. (author). 17 refs., 2 figs

  11. HAMMER, 1-D Multigroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation

    International Nuclear Information System (INIS)

    Honeck, H.C.

    1984-01-01

    1 - Description of problem or function: HAMMER performs infinite lattice, one-dimensional cell multigroup calculations, followed (optionally) by one-dimensional, few-group, multi-region reactor calculations with neutron balance edits. 2 - Method of solution: Infinite lattice parameters are calculated by means of multigroup transport theory, composite reactor parameters by few-group diffusion theory. 3 - Restrictions on the complexity of the problem: - Cell calculations - maxima of: 30 thermal groups; 54 epithermal groups; 20 space points; 20 regions; 18 isotopes; 10 mixtures; 3 thermal up-scattering mixtures; 200 resonances per group; no overlap or interference; single level only. - Reactor calculations - maxima of : 40 regions; 40 mixtures; 250 space points; 4 groups

  12. A code system to generate multigroup cross-sections using basic data

    International Nuclear Information System (INIS)

    Garg, S.B.; Kumar, Ashok

    1978-01-01

    For the neutronic studies of nuclear reactors, multigroup cross-sections derived from the basic energy point data are needed. In order to carry out the design based studies, these cross-sections should also incorporate the temperature and fuel concentration effects. To meet these requirements, a code system comprising of RESRES, UNRES, FIGERO, INSCAT, FUNMO, AVER1 and BGPONE codes has been adopted. The function of each of these codes is discussed. (author)

  13. Young Adults’ Attitude Towards Advertising: a multi-group analysis by ethnicity

    OpenAIRE

    Hiram Ting; Ernest Cyril de Run; Ramayah Thurasamy

    2015-01-01

    Objective – This study aims to investigate the attitude of Malaysian young adults towards advertising. How this segment responds to advertising, and how ethnic/cultural differences moderate are assessed.Design/methodology/approach – A quantitative questionnaire is used to collect data at two universities. Purposive sampling technique is adopted to ensure the sample represents the actual population. Structural equation modelling (SEM) and multi-group analysis (MGA) are utilized in analysis.Fin...

  14. Solution of the multigroup diffusion equation for two-dimensional triangular regions by finite Fourier transformation

    International Nuclear Information System (INIS)

    Takeshi, Y.; Keisuke, K.

    1983-01-01

    The multigroup neutron diffusion equation for two-dimensional triangular geometry is solved by the finite Fourier transformation method. Using the zero-th-order equation of the integral equation derived by this method, simple algebraic expressions for the flux are derived and solved by the alternating direction implicit method. In sample calculations for a benchmark problem of a fast breeder reactor, it is shown that the present method gives good results with fewer mesh points than the usual finite difference method

  15. Survey of computer codes which produce multigroup data from ENDF/B-IV

    International Nuclear Information System (INIS)

    Greene, N.M.

    1975-01-01

    The features of three code systems that produce multigroup neutron data are contrasted. This includes the ETOE-2/MC 2 -2/SDX, MINX/SPHINX and AMPX code packages. These systems all contain a fairly extensive set of processing capabilities with the current evaluated nuclear data files--ENDF/B. They were designed with different goals and applications in mind. This paper discusses some of their differences and the implications for particular situations

  16. Multigroup transport calculations of critical and fuel assemblies with taking into account the scattering anisotropy

    International Nuclear Information System (INIS)

    Rubin, I.E.; Dneprovskaya, N.M.

    2005-01-01

    A technique for calculation of reactor lattices by means of the transmission probabilities with taking into account the scattering anisotropy is generalized for the multigroup case. The errors of the calculated multiplication coefficients and energy release distributions do noe exceed practically the errors, of these values, obtained by the Monte Carlo method. The proposed method is most effective when determining the small difference effects [ru

  17. TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticallity studies

    International Nuclear Information System (INIS)

    Ermuncu, G.; Gonnord, J.; Nimal, J.C.

    1980-04-01

    TRIMARAN is developed for safety analysis of nuclar components containing fissionnable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies

  18. ERRFILS: a preliminary library of 30-group multigroup covariance data for use in CTR sensitivity studies

    International Nuclear Information System (INIS)

    LaBauve, R.J.; Muir, D.W.

    1978-01-01

    A library of 30-group multigroup covariance data was prepared from preliminary ENDF/B-V data with the NJOY code. Data for Fe, Cr, Ni, 10 B, C, Cu, H, and Pb are included in this library. Reactions include total cross sections, elastic and inelastic scattering cross sections, and the most important absorption cross sections. Typical data from the file are shown. 3 tables

  19. Multigroup constants for charged particle elastic nuclear (plus interference) scattering of light isotopes

    International Nuclear Information System (INIS)

    Cullen, D.E.; Perkins, S.T.

    1977-01-01

    Multi-group averaged reaction rates and transfer matrices were calculated for charged particle induced elastic nuclear (plus interference) scattering. Results are presented using a ten group structure for all twenty-five permutations of projectile and target for the following charged particles: p, d, t, 3 He and alpha. Transfer matrices are presented in a simplified form for both incident projectile and the knock-ons; these matrices explicitly conserve energy

  20. Specifications for a two-dimensional multi-group scattering code: ALCI; Specification d'un code de diffusion multigroupe a deux dimensions: ALCI

    Energy Technology Data Exchange (ETDEWEB)

    Bayard, J P; Guillou, A; Lago, B; Bureau du Colombier, M J; Guillou, G; Vasseur, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-02-01

    This report describes the specifications of the ALCI programme. This programme resolves the system of difference equations similar to the homogeneous problem of multigroup neutron scattering, with two dimensions in space, in the three geometries XY, RZ, R{theta}. It is possible with this method to calculate geometric and composition criticalities and also to calculate the accessory problem on demand. The maximum number of points dealt with is 6000. The maximum permissible number of groups is 12. The internal iterations are treated by the method of alternating directions. The external iterations are accelerated using the extrapolation method due to Tchebychev. (authors) [French] Ce rapport decrit les specifications du programme ALCI. Ce programme resout le systeme d'equations aux differences approchant le probleme homogene de la diffusion neutronique multigroupe, a deux dimensions d'espace, dans les trois geometries XY, RZ, R{theta}. Il permet des calculs de criticalite geometrique et de composition et calcule sur demande le probleme adjoint. Le nombre maximum de points traites est de 6000. Le nombre maximum de groupes permis est de 12. Les iterations interieure sont traitees par la methode des directions alternees. Les iterations exterieures sont accelerees par la methode d'extrapolation de Tchebychev. (auteurs)

  1. Depletion Calculations for MTR Core Using MCNPX and Multi-Group Nodal Diffusion Methods

    International Nuclear Information System (INIS)

    Jaradata, Mustafa K.; Park, Chang Je; Lee, Byungchul

    2013-01-01

    In order to maintain a self-sustaining steady-state chain reaction, more fuel than is necessary in order to maintain a steady state chain reaction must be loaded. The introduction of this excess fuel increases the net multiplication capability of the system. In this paper MCNPX and multi-group nodal diffusion theory will be used for depletion calculations for MTR core. The eigenvalue and power distribution in the core will be compared for different burnup. Multi-group nodal diffusion theory with combination of NEWT-TRITON system was used to perform depletion calculations for 3Χ3 MTR core. 2G and 6G approximations were used and compared with MCNPX results for 2G approximation the maximum difference from MCNPX was 40 mk and for 6G approximation was 6 mk which is comparable to the MCNPX results. The calculated power using nodal code was almost the same MCNPX results. Finally the results of the multi-group nodal theory were acceptable and comparable to the calculated using MCNPX

  2. Numerical method for multigroup one-dimensional SN eigenvalue problems with no spatial truncation error

    International Nuclear Information System (INIS)

    Abreu, M.P.; Filho, H.A.; Barros, R.C.

    1993-01-01

    The authors describe a new nodal method for multigroup slab-geometry discrete ordinates S N eigenvalue problems that is completely free from all spatial truncation errors. The unknowns in the method are the node-edge angular fluxes, the node-average angular fluxes, and the effective multiplication factor k eff . The numerical values obtained for these quantities are exactly those of the dominant analytic solution of the S N eigenvalue problem apart from finite arithmetic considerations. This method is based on the use of the standard balance equation and two nonstandard auxiliary equations. In the nonmultiplying regions, e.g., the reflector, we use the multigroup spectral Green's function (SGF) auxiliary equations. In the fuel regions, we use the multigroup spectral diamond (SD) auxiliary equations. The SD auxiliary equation is an extension of the conventional auxiliary equation used in the diamond difference (DD) method. This hybrid characteristic of the SD-SGF method improves both the numerical stability and the convergence rate

  3. Achievement and qualification of multigroup cross-section library for light water reactor calculation

    International Nuclear Information System (INIS)

    Gastaldi, B.

    1986-07-01

    This study intends to improve then to check on integral experiments, the calculation of the main neutronic parameters in light water moderated lattices: Uranium 238 capture and consequently Plutonium 239 build-up, multiplication factor, temperature coefficient. The first part of this work concerns the resonant reaction rate calculation method implemented in the APOLLO code, the so-called LIVOLANT and JEANPIERRE formalism. The errors introduced by the corresponding assumptions are quantified and we propose substitution methods which avoid large biases and supply satisfactory results. The second part is dedicated to the cross-section evaluation of uranium major isotopes and to the achievement of APOLLO multigroup cross-sections. This cross-section set takes into considerations on the one hand the recent differential information and the other hand the various integral information obtained in the French Atomic Energy Commission facilities. The nuclear data file (JEF abd ENDF/B5) processing, for multigroup and self-shielded cross-sections achieving enable us to check the new THEMIS computer code. In the last part, the experimental validation of the proposed procedure (accurate formalism mutuel shielding and new multigroup library) is presented. This qualification is based on the reinterpretation of critical experiments performed in the EOLE reactor at Cadarache and spent fuel analysis. The corresponding results demonstrate that our propositions provide improvements on the computation of the PWR neutronic parameters; calculation-experiment discrepancies are now consistent with experimental uncertainty margins. 46 refs; 31 figs; 23 tabl [fr

  4. Conception and development of an adaptive energy mesher for multigroup library generation of the transport codes

    International Nuclear Information System (INIS)

    Mosca, P.

    2009-12-01

    The deterministic transport codes solve the stationary Boltzmann equation in a discretized energy formalism called multigroup. The transformation of continuous data in a multigroup form is obtained by averaging the highly variable cross sections of the resonant isotopes with the solution of the self-shielding models and the remaining ones with the coarse energy spectrum of the reactor type. So far the error of such an approach could only be evaluated retrospectively. To remedy this, we studied in this thesis a set of methods to control a priori the accuracy and the cost of the multigroup transport computation. The energy mesh optimisation is achieved using a two step process: the creation of a reference mesh and its optimized condensation. In the first stage, by refining locally and globally the energy mesh, we seek, on a fine energy mesh with subgroup self-shielding, a solution equivalent to a reference solver (Monte Carlo or pointwise deterministic solver). In the second step, once fixed the number of groups, depending on the acceptable computational cost, and chosen the most appropriate self-shielding models to the reactor type, we look for the best bounds of the reference mesh minimizing reaction rate errors by the particle swarm optimization algorithm. This new approach allows us to define new meshes for fast reactors as accurate as the currently used ones, but with fewer groups. (author)

  5. Advances on fission chamber modelling

    International Nuclear Information System (INIS)

    Filliatre, Philippe; Jammes, Christian; Geslot, Benoit; Veenhof, Rob

    2013-06-01

    In-vessel, online neutron flux measurements are routinely performed in mock-up and material testing reactors by fission chambers. Those measurements have a wide range of applications, including characterization of experimental conditions, reactor monitoring and safety. Depending on the application, detectors may experience a wide range of constraints, of several magnitudes, in term of neutron flux, gamma-ray flux, temperature. Hence, designing a specific fission chamber and measuring chain for a given application is a demanding task. It can be achieved by a combination of experimental feedback and simulating tools, the latter being based on a comprehensive understanding of the underlying physics. A computation route that simulates fission chambers, named CHESTER, is presented. The retrieved quantities of interest are the neutron-induced charge spectrum, the electronic and ionic pulses, the mean current and variance, the power spectrum. It relies on the GARFIELD suite, originally developed for drift chambers, and makes use of the MAGBOLTZ code to assess the drift parameters of electrons within the filling gas, and the SRIM code to evaluate the stopping range of fission products. The effect of the gamma flux is also estimated. Computations made with several fission chambers exemplify the possibilities of the route. A good qualitative agreement is obtained when comparing the results with the experimental data available to date. In a near future, a comprehensive experimental programme will be undertaken to qualify the route using the known neutron sources, mock-up reactors and wide choice of fission chambers, with a stress on the predictiveness of the Campbelling mode. Depending on the results, a refinement of the modelling and an effort on the accuracy of input data are also to be considered. CHESTER will then make it possible to predict the overall sensitivity of a chamber, and to optimize the design for a given application. Another benefit will be to increase the

  6. Thorium-uranium fission radiography

    Science.gov (United States)

    Haines, E. L.; Weiss, J. R.; Burnett, D. S.; Woolum, D. S.

    1976-01-01

    Results are described for studies designed to develop routine methods for in-situ measurement of the abundance of Th and U on a microscale in heterogeneous samples, especially rocks, using the secondary high-energy neutron flux developed when the 650 MeV proton beam of an accelerator is stopped in a 42 x 42 cm diam Cu cylinder. Irradiations were performed at three different locations in a rabbit tube in the beam stop area, and thick metal foils of Bi, Th, and natural U as well as polished silicate glasses of known U and Th contents were used as targets and were placed in contact with mica which served as a fission track detector. In many cases both bare and Cd-covered detectors were exposed. The exposed mica samples were etched in 48% HF and the fission tracks counted by conventional transmitted light microscopy. Relative fission cross sections are examined, along with absolute Th track production rates, interaction tracks, and a comparison of measured and calculated fission rates. The practicality of fast neutron radiography revealed by experiments to data is discussed primarily for Th/U measurements, and mixtures of other fissionable nuclei are briefly considered.

  7. Energy production using fission fragment rockets

    International Nuclear Information System (INIS)

    Chapline, G.; Matsuda, Y.

    1991-08-01

    Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: Approximately twice as efficient if one can directly convert the fission fragment energy into electricity; by reducing the buildup of a fission fragment inventory in the reactor one could avoid a Chernobyl type disaster; and collecting the fission fragments outside the reactor could simplify the waste disposal problem. 6 refs., 4 figs., 2 tabs

  8. Contribution to the study of nuclear fission

    International Nuclear Information System (INIS)

    Serot, O.

    2009-09-01

    The author proposes an overview of his research activity during the past fifteen years and more particularly that dealing with nuclear fission. The first part reports works on nucleus physics at the scission via the investigation of ternary fission (experimental procedure, influence of fission modes, influence of resonance spin, influence of excitation energy of the fissioning nucleus, emission probabilities, energy spectra of ternary alphas and tritons, emission mechanism). The second part reports measurements and assessments of neutron-induced fission cross sections. The third part reports the investigation of some properties of fission products (efficiencies, branching ratios of the main delayed neutron precursors)

  9. Muon induced fission and fission track dating of minerals

    International Nuclear Information System (INIS)

    Marques, A.

    1988-01-01

    The effects of muon induced fission on geological dating of samples by the fission track method are evaluated for the case of muscovite minerals. It is found a small but significant effect, greater for the longer ages. Since calculations are developped under the hypothesis of constant atmosphere and primary cosmic ray flux it is suggested that any discrepancy found in ages of very old material that cannot be accounted for by well known environmental influences, be taken as an indication of variation on either the atmospheric stopping power or the intensity of cosmic radiation along the ages. (author) [pt

  10. Nuclear fission and neutron-induced fission cross-sections

    CERN Document Server

    James, G D; Michaudon, A; Michaudon, A; Cierjacks, S W; Chrien, R E

    2013-01-01

    Nuclear Fission and Neutron-Induced Fission Cross-Sections is the first volume in a series on Neutron Physics and Nuclear Data in Science and Technology. This volume serves the purpose of providing a thorough description of the many facets of neutron physics in different fields of nuclear applications. This book also attempts to bridge the communication gap between experts involved in the experimental and theoretical studies of nuclear properties and those involved in the technological applications of nuclear data. This publication will be invaluable to those interested in studying nuclear fis

  11. Report of fission study meeting

    International Nuclear Information System (INIS)

    1986-03-01

    This book is the report of fission Study Meeting held from September 19 to 21, 1985 in the Research Center for Nuclear Physics, Osaka University. The objective of this study meeting was to stimulate the research on nuclear physics in Japan, which began to show new development accompanying the advance of the research on heavy ion nuclear reaction, and to make this a new starting point. More than 50 participants from physical, chemical and engineering fields, who have interest in the theory and experiment related to nuclear fission, gathered, and the meeting was a success beyond expectation. The contents covered a wide range including nuclear smashing reaction as well as nuclear fission in a narrow sense. In this book, the gists of 28 papers are collected. (Kako, I.)

  12. Status of fission yield data

    International Nuclear Information System (INIS)

    England, T.R.; Blachot, J.

    1988-01-01

    In this paper we summarize the current status of the recent US evaluation for 34 fissioning nuclides at one or more neutron incident energies and for spontaneous fission. Currently there are 50 yields sets, and for each we have independent and cumulative yields and uncertainties for approximately 1100 fission products. When finalized the recommended data will become part of Version VI of the US ENDF/B. Other major evaluations in progress that are included in a recently formed IAEA Coordinated Research Program are also summarized. In a second part we review two empirical models in use to estimate independent yields. Comparison of model estimates with measured data is presented, including a comparison with some recent data obtained from Lohengrin (Cf-249 T). 18 refs., 13 figs., 3 tabs

  13. Experimental approach to fission process of actinides

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Hiroshi [Osaka Univ., Toyonaka (Japan). Faculty of Science

    1997-07-01

    From experimental views, it seems likely that the mechanism of nuclear fission process remains unsolved even after the Bohr and Weeler`s study in 1939. Especially, it is marked in respect of mass distribution in unsymmetric nuclear fission. The energy dependency of mass distribution can be explained with an assumption of 2-mode nuclear fission. Further, it was demonstrated that the symmetrical fission components and the unsymmetrical ones have different saddle and fission points. Thus, the presence of the 2-mode fission mechanism was confirmed. Here, transition in the nuclear fission mechanism and its cause were investigated here. As the cause of such transition, plausible four causes; a contribution of multiple-chance fission, disappearance of shell effects, beginning of fission following collective excitation due to GDR and nuclear phase transition were examined in the condition of excitation energy of 14.0 MeV. And it was suggested that the transition in the nuclear fission concerned might be related to phase transition. In addition, the mechanism of nuclear fission at a low energy and multi-mode hypothesis were examined by determination of the energy for thermal neutron fission ({sup 233,235}U and {sup 239}Pu) and spontaneous nuclear fission ({sup 252}Cf). (M.N.)

  14. NEACRP thermal fission product benchmark

    International Nuclear Information System (INIS)

    Halsall, M.J.; Taubman, C.J.

    1989-09-01

    The objective of the thermal fission product benchmark was to compare the range of fission product data in use at the present time. A simple homogeneous problem was set with 200 atoms H/1 atom U235, to be burnt up to 1000 days and then decay for 1000 days. The problem was repeated with 200 atoms H/1 atom Pu239, 20 atoms H/1 atom U235 and 20 atoms H/1 atom Pu239. There were ten participants and the submissions received are detailed in this report. (author)

  15. Search for Singlet Fission Chromophores

    Energy Technology Data Exchange (ETDEWEB)

    Havlas, Z.; Akdag, A.; Smith, M. B.; Dron, P.; Johnson, J. C.; Nozik, A. J.; Michl, J.

    2012-01-01

    Singlet fission, in which a singlet excited chromophore shares its energy with a ground-state neighbor and both end up in their triplet states, is of potential interest for solar cells. Only a handful of compounds, mostly alternant hydrocarbons, are known to perform efficiently. In view of the large number of conditions that a successful candidate for a practical cell has to meet, it appears desirable to extend the present list of high performers to additional classes of compounds. We have (i) identified design rules for new singlet fission chromophores and for their coupling to covalent dimers, (ii) synthesized them, and (iii) evaluated their performance as neat solids or covalent dimers.

  16. Fission 99Mo production technology

    International Nuclear Information System (INIS)

    Miao Zengxing; Luo Zhifu; Ma Huimin; Liang Yufu; Yu Ningwen

    2003-01-01

    This paper describes a production technology of fission 99 Mo in the Department Isotope, CIAE. The irradiation target is tubular U-Al alloy containing highly enriched uranium. The target is irradiated in the swimming pool reactor core. The neutron flux is about 4x10 13 /cm 2 .sec. The production scale is 3.7-7.4 TBq (100-200Ci) of fission 99 Mo per batch. Total recovery of 99 Mo is more than 70%. The production practice proves that the process and equipment are safe and reliable. (author)

  17. The wastes of nuclear fission

    International Nuclear Information System (INIS)

    Doubre, H.

    2005-01-01

    In this paper the author presents the problems of the radioactive wastes generated by the nuclear fission. The first part devoted to the fission phenomenon explains the incident neutron energy and the target nuclei role. The second part devoted to the nuclear wastes sources presents the production of wastes upstream of the reactors, in the reactors and why these wastes are dangerous. The third part discusses the radioactive wastes management in France (classification, laws). The last part details the associated research programs: the radionuclides separation, the disposal, the underground storage, the transmutation and the thorium cycle. (A.L.B.)

  18. 40 years of nuclear fission

    International Nuclear Information System (INIS)

    Koch, H.

    1979-01-01

    On the occasion of both the 100th birthday of the discoverer of nuclear fission, Otto Hahn, and the 40th anniversary of this outstanding scientific discovery the historical development is described, which led to nuclear fission. Aspects of scientific life in Berlin and in the whole world at that time are presented, and relations between scientists are characterized by quotations. In particular, stress is laid on the life and activities of Otto Hahn as a human being and as a scientist, and his outstanding scientific achievements are appreciated. (author)

  19. ZZ ANSLV, Multigroup Cross Sections Library for ANS Reactor Design Studies

    International Nuclear Information System (INIS)

    2000-01-01

    A - Description of program or function: - Format: AMPX Master Interface Library format. Number of groups: Fine Group (99 energy groups) General Purpose Neutron Library. Materials: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Broad Group (39 energy groups) General Purpose Neutron Library. Materials: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Gamma-Ray Interaction (GRI) Library in 44-groups. Materials: H, He, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Xe, Sm, Eu, Hf, Ta, Ir, Pb, Th, U, Pu. Origin: ENDF/B-V; LENDL-V evaluations for 12 materials. - Format: AMPX Master Interface Library format. Number of groups: Coupled Library containing (CNG) 99-group neutron and 44-group gamma-ray data. Materials: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Coupled neutron-gamma (CNG) Library containing 39-group, and 44-group gamma-ray data. Materials: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu. Origin: ENDF/B-V. Weighting spectrum: Maxwellian 300 K + 1/(E*sigma-total) + fission spectrum4 types of boundaries have been used depending isotope and library type (see report). Pseudo-problem-independent, multigroup cross section libraries were generated to support the Advanced Neutron source (ANS) reactor design studies. The ANS was

  20. Sommerfeld-Watson transformation for nuclear fission

    International Nuclear Information System (INIS)

    Alexandru, G.

    1978-01-01

    It is proved that the fission matrix element can be written like a Sommerfeld-Watson relation. This leads to a dispersion relation for the fission process in which the substraction term is uniquely determined. (author)

  1. Fission properties of the heaviest elements

    International Nuclear Information System (INIS)

    Moller, P.; Nix, R.

    1995-01-01

    The authors discuss fission properties of the heaviest elements. In particular they focus on stability with respect to spontaneous fission and on the prospects of extending the region of known nuclei beyond the peninsula of currently known nuclides

  2. Fission fragment distributions within dynamical approach

    Energy Technology Data Exchange (ETDEWEB)

    Mazurek, K. [Institute of Nuclear, Physics Polish Academy of Sciences, Krakow (Poland); Nadtochy, P.N. [Omsk State Technical University, Omsk (Russian Federation); Ryabov, E.G.; Adeev, G.D. [Omsk State University, Physics Department, Omsk (Russian Federation)

    2017-04-15

    The review covers recent developments and achievements in the dynamical description of fission process at high excitation energy. It is shown that the dynamical approach based on multidimensional Langevin equations combined with the statistical description of nuclear decay by particles evaporation is capable of fairly well describing the formation of fission fragment mass-energy, charge, and angular distributions of fission fragments in coincidence with the pre- and post-scission particle emission. The final yields of fission and evaporation residues channels products could be obtained. The detailed description of fission dynamics allows studying different stages of fission process, indicating the most important ingredients governing fission process and studying in detail such fundamental nuclear properties as nuclear viscosity and fission timescale. The tasks and perspectives of multidimensional dynamical approach are also discussed. (orig.)

  3. Absolute calibration technique for spontaneous fission sources

    International Nuclear Information System (INIS)

    Zucker, M.S.; Karpf, E.

    1984-01-01

    An absolute calibration technique for a spontaneously fissioning nuclide (which involves no arbitrary parameters) allows unique determination of the detector efficiency for that nuclide, hence of the fission source strength

  4. Investigation of exotic fission modes

    International Nuclear Information System (INIS)

    Poenaru, D. N.; Gherghescu, R. A.; Greiner, W.; Nagame, Y.; Hamilton, J. H.; Ramayya, A. V.

    2002-01-01

    Fission approach to the cluster radioactivities and α-decay has been systematically developed during the last two decades. A more complex process, the ternary fission, was observed since 1946 both in neutron-induced and spontaneous fission. We obtained interesting results concerning the binary fission saddle-point reflection asymmetric nuclear shapes, and we can explain how a possible nuclear quasimolecular state is formed during the 10 Be accompanied cold fission of 252 Cf. The equilibrium nuclear shapes in fission theory are usually determined by minimizing the deformation energy for a given surface equation. We developed a method allowing to obtain a very general saddle-point shape as a solution of a differential equation without an a priori introduction of a shape parametrization. In the approach based on a liquid drop model (LDM), saddle-point shapes are always reflection symmetric: the deformation energy increases with the mass-asymmetry parameter η = (A 1 - A 2 )/(A 1 + A 2 ). By adding the shell corrections to the LDM deformation energy, we obtained minima at a finite mass asymmetry for parent nuclei 238 U, 232,228 Th in agreement with experiments. This correction was calculated phenomenologically. A technique based on the fragment identification by using triple γ coincidences in the large arrays of Ge-detectors, like GAMMASPHERE, was employed at Vanderbilt University to discover new characteristics of the fission process, and new decay modes. The possibility of a whole family of new decay modes, the multicluster accompanied fission, was envisaged. Besides the fission into two or three fragments, a heavy or superheavy nucleus spontaneously breaks into four, five or six nuclei of which two are asymmetric or symmetric heavy fragments and the others are light clusters, e.g. α-particles, 10 Be, 14 C, or combinations of them. Examples were presented for the two-, three- and four cluster accompanied cold fission of 252 Cf and 262 Rf, in which the emitted

  5. Computer program FPIP-REV calculates fission product inventory for U-235 fission

    Science.gov (United States)

    Brown, W. S.; Call, D. W.

    1967-01-01

    Computer program calculates fission product inventories and source strengths associated with the operation of U-235 fueled nuclear power reactor. It utilizes a fission-product nuclide library of 254 nuclides, and calculates the time dependent behavior of the fission product nuclides formed by fissioning of U-235.

  6. Nuclear fission as a macroscopic quantum tunneling

    International Nuclear Information System (INIS)

    Takigawa, N.

    1995-01-01

    We discuss nuclear fission from the point of view of a macroscopic quantum tunneling, one of whose major interests is to study the effects of environments on the tunneling rate of a macroscopic variable. We show that a vibrational excitation of the fissioning nucleus significantly enhances the fission rate. We show this effect by two different methods. The one is to treat the vibrational excitation as an environmental degree of freedom, the other treats the fission as a two dimensional quantum tunneling. (author)

  7. Verification and validation of multi-group library MUSE1.0 created from ENDF/B-VII.0

    International Nuclear Information System (INIS)

    Chen Yixue; Wu Jun; Yang Shouhai; Zhang Bin; Lu Daogang; Chen Chaobin

    2010-01-01

    A multi-group library set named MUSE1.0 with 172-neutron group and 42-photon group is produced based on ENDF/B-VII.0 using NJOY code. Weight function of the multi-group library set is taken from the Vitanim-e library and the max legendre order of scattering matrix is six. All the nuclides have thermal scattering data created using free-gas scattering law and 10 Bondarenko background cross sections se lected to generate the self-shielded multi-group cross sections. The final libraries have GENDF-format, MATXS-format and ACE-multi-group sub-libraries and each sub-library generated under 4 temperatures(293 K,600 K,800 K and 900 K). This paper provides a summary of the procedure to produce the library set and a detail description of the validation of the multi-group library set by several critical benchmark devices and shielding benchmark devices using MCNP code. The ability to handle the thermal neutron transport and resonance self-shielding problems are investigated specially. In the end, we draw the conclusion that the multi-group libraries produced is credible and can be used in the R and D process of Supercritical Water Reactor Design. (authors)

  8. Neutron gamma competition in fast fission

    International Nuclear Information System (INIS)

    Frehaut, J.

    1989-01-01

    In the present paper we analyse the data we have obtained on the distribution of the gamma-ray energy per fission, as well as on the average energy E-barγ released per fission for the neutron induced fission of several isotopes, in the energy range up to 15 MeV. 6 refs, 9 figs

  9. Fission yield data evaluation system FYDES

    International Nuclear Information System (INIS)

    Liu Tingjin

    1998-01-01

    Taking account of some features of fission yield data, to do the fission yield data evaluation conveniently, a fission yield data evaluation system FYDES has been developed for last two years. Outline of the system, data retrieval and data table standardization, data correction codes, data averaging code, simultaneous evaluation code and data fit programs were introduced

  10. Fission fragment mass and angular distributions

    Indian Academy of Sciences (India)

    2015-07-22

    Jul 22, 2015 ... Synthesis of heavy and superheavy elements is severely hindered by fission and fission-like processes. The probability of these fission-like, non-equilibrium processes strongly depends on the entrance channel parameters. This article attempts to summarize the recent experimental findings and classify the ...

  11. Some aspects of fission and quasifission processes

    Indian Academy of Sciences (India)

    2015-07-22

    Jul 22, 2015 ... In this talk, I will review an incomplete subset of the major milestones in fission research, and briefly discuss some of the topics that I have been involved in during my career. These include studies of vibrational resonances and fission isomers that are caused by the second minimum in the fission barrier in ...

  12. Charged particle-induced nuclear fission reactions

    Indian Academy of Sciences (India)

    The nuclear fission phenomenon continues to be an enigma, even after nearly 75 years of its discovery. Considerable progress has been made towards understanding the fission process. Both light projectiles and heavy ions have been employed to investigate nuclear fission. An extensive database of the properties of ...

  13. Fission approach to cluster radioactivity

    Indian Academy of Sciences (India)

    2015-08-04

    Aug 4, 2015 ... Also, the analytical superasymmetric fission (ASAF) model is successfully employed to make a systematic search and to predict, with other models, cluster ... those of the staff, the journals, various programmes, and Current Science, has changed from 'ias.ernet.in' (or 'academy.ias.ernet.in') to 'ias.ac.in'. Thus ...

  14. Spectroscopy of heavy fissionable nuclei

    Indian Academy of Sciences (India)

    2015-08-05

    Aug 5, 2015 ... Nuclei in the actinide chain and beyond are prone to fission owing to ... mass nuclei are typically more difficult, because the intensity is .... j15/2 neutron alignments in a region where shell stablization effects are crucial.

  15. Space Fission System Test Effectiveness

    International Nuclear Information System (INIS)

    Houts, Mike; Schmidt, Glen L.; Van Dyke, Melissa; Godfroy, Tom; Martin, James; Bragg-Sitton, Shannon; Dickens, Ricky; Salvail, Pat; Harper, Roger

    2004-01-01

    Space fission technology has the potential to enable rapid access to any point in the solar system. If fission propulsion systems are to be developed to their full potential, however, near-term customers need to be identified and initial fission systems successfully developed, launched, and utilized. One key to successful utilization is to develop reactor designs that are highly testable. Testable reactor designs have a much higher probability of being successfully converted from paper concepts to working space hardware than do designs which are difficult or impossible to realistically test. ''Test Effectiveness'' is one measure of the ability to realistically test a space reactor system. The objective of this paper is to discuss test effectiveness as applied to the design, development, flight qualification, and acceptance testing of space fission systems. The ability to perform highly effective testing would be particularly important to the success of any near-term mission, such as NASA's Jupiter Icy Moons Orbiter, the first mission under study within NASA's Project Prometheus, the Nuclear Systems Program

  16. Nuclear fission with inertial confinement

    CERN Document Server

    Koshkarev, D G

    2002-01-01

    The possibility of initiating the explosive fission reaction in a small quantity of fissile material through the heavy ions beam from the powerful accelerator-driver, developed for realization of the thermonuclear synthesis in the deuterium-tritium cylindrical targets with the direct ignition, is considered. The consequences of applying this method in the nuclear engineering are discussed

  17. Dynamical features of nuclear fission

    Indian Academy of Sciences (India)

    Wheeler underestimates several observables in heavy-ion-induced ... excitation energies, there may not be sufficient nuclei near the fission barrier after the .... Dissipation in nuclear dynamics in the mean-field regime accounts for the coupling of the .... barrier for different isotopes of Fr. The lines are drawn to guide the eye.

  18. Group-decoupled multi-group pin power reconstruction utilizing nodal solution 1D flux profiles

    International Nuclear Information System (INIS)

    Yu, Lulin; Lu, Dong; Zhang, Shaohong; Wang, Dezhong

    2014-01-01

    Highlights: • A direct fitting multi-group pin power reconstruction method is developed. • The 1D nodal solution flux profiles are used as the condition. • The least square fit problem is analytically solved. • A slowing down source improvement method is applied. • The method shows good accuracy for even challenging problems. - Abstract: A group-decoupled direct fitting method is developed for multi-group pin power reconstruction, which avoids both the complication of obtaining 2D analytic multi-group flux solution and any group-coupled iteration. A unique feature of the method is that in addition to nodal volume and surface average fluxes and corner fluxes, transversely-integrated 1D nodal solution flux profiles are also used as the condition to determine the 2D intra-nodal flux distribution. For each energy group, a two-dimensional expansion with a nine-term polynomial and eight hyperbolic functions is used to perform a constrained least square fit to the 1D intra-nodal flux solution profiles. The constraints are on the conservation of nodal volume and surface average fluxes and corner fluxes. Instead of solving the constrained least square fit problem numerically, we solve it analytically by fully utilizing the symmetry property of the expansion functions. Each of the 17 unknown expansion coefficients is expressed in terms of nodal volume and surface average fluxes, corner fluxes and transversely-integrated flux values. To determine the unknown corner fluxes, a set of linear algebraic equations involving corner fluxes is established via using the current conservation condition on all corners. Moreover, an optional slowing down source improvement method is also developed to further enhance the accuracy of the reconstructed flux distribution if needed. Two test examples are shown with very good results. One is a four-group BWR mini-core problem with all control blades inserted and the other is the seven-group OECD NEA MOX benchmark, C5G7

  19. Multi-group transport methods for high-resolution neutron activation analysis

    International Nuclear Information System (INIS)

    Burns, K. A.; Smith, L. E.; Gesh, C. J.; Shaver, M. W.

    2009-01-01

    The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of multi-group deterministic methods for the simulation of neutron activation problems. Central to this work is the development of a method for generating multi-group neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so that the key signatures in neutron activation analysis (i.e., the characteristic line energies) are preserved. The mechanics of the cross-section preparation method are described and contrasted with standard neutron-gamma cross-section sets. These custom cross-sections are then applied to several benchmark problems. Multi-group results for neutron and photon flux are compared to MCNP results. Finally, calculated responses of high-resolution spectrometers are compared. Preliminary findings show promising results when compared to MCNP. A detailed discussion of the potential benefits and shortcomings of the multi-group-based approach, in terms of accuracy, and computational efficiency, is provided. (authors)

  20. Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh

    International Nuclear Information System (INIS)

    Aggery, A.

    1999-12-01

    The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

  1. Hydrogen transport in a toroidal plasma using multigroup discrete-ordinates methodology

    International Nuclear Information System (INIS)

    Wienke, B.R.; Miller, W.F. Jr.; Seed, T.J.

    1979-01-01

    Neutral hydrogen transport in a fully ionized two-dimensional tokamak plasma was examined using discrete ordinates and contrasted with earlier analyses. In particular, curvature effects induced by toroidal geometries and ray effects caused by possible source localization were investigated. From an overview of the multigroup discrete-ordinates approximation, methodology in two-dimensional cylindrical geometry is detailed, mesh and plasma zoning procedures are sketched, and the piecewise polynomial solution algorithm on a triangular domain is obtained. Toroidal effects and comparisons as related to reaction rates and perticle spectra are examined for various model and source configurations

  2. Using the probability method for multigroup calculations of reactor cells in a thermal energy range

    International Nuclear Information System (INIS)

    Rubin, I.E.; Pustoshilova, V.S.

    1984-01-01

    The possibility of using the transmission probability method with performance inerpolation for determining spatial-energy neutron flux distribution in cells of thermal heterogeneous reactors is considered. The results of multigroup calculations of several uranium-water plane and cylindrical cells with different fuel enrichment in a thermal energy range are given. A high accuracy of results is obtained with low computer time consumption. The use of the transmission probability method is particularly reasonable in algorithms of the programmes compiled computer with significant reserve of internal memory

  3. Simulate-HEX - The multi-group diffusion equation in hexagonal-z geometry

    International Nuclear Information System (INIS)

    Lindahl, S. O.

    2013-01-01

    The multigroup diffusion equation is solved for the hexagonal-z geometry by dividing each hexagon into 6 triangles. In each triangle, the Fourier solution of the wave equation is approximated by 8 plane waves to describe the intra-nodal flux accurately. In the end an efficient Finite Difference like equation is obtained. The coefficients of this equation depend on the flux solution itself and they are updated once per power/void iteration. A numerical example demonstrates the high accuracy of the method. (authors)

  4. NUMERICAL MULTIGROUP TRANSIENT ANALYSIS OF SLAB NUCLEAR REACTOR WITH THERMAL FEEDBACK

    Directory of Open Access Journals (Sweden)

    Filip Osuský

    2016-12-01

    Full Text Available The paper describes a new numerical code for multigroup transient analyses with thermal feedback. The code is developed at Institute of Nuclear and Physical Engineering. It is necessary to carefully investigate transient states of fast neutron reactors, due to recriticality issues after accident scenarios. The code solves numerical diffusion equation for 1D problem with possible neutron source incorporation. Crank-Nicholson numerical method is used for the transient states. The investigated cases are describing behavior of PWR fuel assembly inside of spent fuel pool and with the incorporated neutron source for better illustration of thermal feedback.

  5. REX1-87, Multigroup Neutron Cross-Sections from ENDF/B

    International Nuclear Information System (INIS)

    Gopalakrishnan, V.; Ganesan, S.

    1988-01-01

    1 - Description of program or function: The program calculates self- shielding factors for reactor applications from a pre-processed (linearized) evaluated nuclear data file in the ENDF/B format. 2 - Method of solution: Bondarenko definition of multigroup self- shielding factors invoking narrow resonance treatment is used. 3 - Restrictions on the complexity of the problem: a) Maximum no. of energy group is 620. b) Only the built-in forms of the weighting functions can be chosen. c) The program is strictly limited to resolved resonance region from physical considerations

  6. Mining the multigroup-discrete ordinates algorithm for high quality solutions

    International Nuclear Information System (INIS)

    Ganapol, B.D.; Kornreich, D.E.

    2005-01-01

    A novel approach to the numerical solution of the neutron transport equation via the discrete ordinates (SN) method is presented. The new technique is referred to as 'mining' low order (SN) numerical solutions to obtain high order accuracy. The new numerical method, called the Multigroup Converged SN (MGCSN) algorithm, is a combination of several sequence accelerators: Romberg and Wynn-epsilon. The extreme accuracy obtained by the method is demonstrated through self consistency and comparison to the independent semi-analytical benchmark BLUE. (authors)

  7. Numerical analysis for multi-group neutron-diffusion equation using Radial Point Interpolation Method (RPIM)

    International Nuclear Information System (INIS)

    Kim, Kyung-O; Jeong, Hae Sun; Jo, Daeseong

    2017-01-01

    Highlights: • Employing the Radial Point Interpolation Method (RPIM) in numerical analysis of multi-group neutron-diffusion equation. • Establishing mathematical formation of modified multi-group neutron-diffusion equation by RPIM. • Performing the numerical analysis for 2D critical problem. - Abstract: A mesh-free method is introduced to overcome the drawbacks (e.g., mesh generation and connectivity definition between the meshes) of mesh-based (nodal) methods such as the finite-element method and finite-difference method. In particular, the Point Interpolation Method (PIM) using a radial basis function is employed in the numerical analysis for the multi-group neutron-diffusion equation. The benchmark calculations are performed for the 2D homogeneous and heterogeneous problems, and the Multiquadrics (MQ) and Gaussian (EXP) functions are employed to analyze the effect of the radial basis function on the numerical solution. Additionally, the effect of the dimensionless shape parameter in those functions on the calculation accuracy is evaluated. According to the results, the radial PIM (RPIM) can provide a highly accurate solution for the multiplication eigenvalue and the neutron flux distribution, and the numerical solution with the MQ radial basis function exhibits the stable accuracy with respect to the reference solutions compared with the other solution. The dimensionless shape parameter directly affects the calculation accuracy and computing time. Values between 1.87 and 3.0 for the benchmark problems considered in this study lead to the most accurate solution. The difference between the analytical and numerical results for the neutron flux is significantly increased in the edge of the problem geometry, even though the maximum difference is lower than 4%. This phenomenon seems to arise from the derivative boundary condition at (x,0) and (0,y) positions, and it may be necessary to introduce additional strategy (e.g., the method using fictitious points and

  8. A self-consistent nodal method in response matrix formalism for the multigroup diffusion equations

    International Nuclear Information System (INIS)

    Malambu, E.M.; Mund, E.H.

    1996-01-01

    We develop a nodal method for the multigroup diffusion equations, based on the transverse integration procedure (TIP). The efficiency of the method rests upon the convergence properties of a high-order multidimensional nodal expansion and upon numerical implementation aspects. The discrete 1D equations are cast in response matrix formalism. The derivation of the transverse leakage moments is self-consistent i.e. does not require additional assumptions. An outstanding feature of the method lies in the linear spatial shape of the local transverse leakage for the first-order scheme. The method is described in the two-dimensional case. The method is validated on some classical benchmark problems. (author)

  9. Multiregion, multigroup collision probability method with white boundary condition for light water reactor thermalization calculations

    International Nuclear Information System (INIS)

    Ozgener, B.; Ozgener, H.A.

    2005-01-01

    A multiregion, multigroup collision probability method with white boundary condition is developed for thermalization calculations of light water moderated reactors. Hydrogen scatterings are treated by Nelkin's kernel while scatterings from other nuclei are assumed to obey the free-gas scattering kernel. The isotropic return (white) boundary condition is applied directly by using the appropriate collision probabilities. Comparisons with alternate numerical methods show the validity of the present formulation. Comparisons with some experimental results indicate that the present formulation is capable of calculating disadvantage factors which are closer to the experimental results than alternative methods

  10. Specifications for a two-dimensional multi-group scattering code: ALCI

    International Nuclear Information System (INIS)

    Bayard, J.P.; Guillou, A.; Lago, B.; Bureau du Colombier, M.J.; Guillou, G.; Vasseur, Ch.

    1965-02-01

    This report describes the specifications of the ALCI programme. This programme resolves the system of difference equations similar to the homogeneous problem of multigroup neutron scattering, with two dimensions in space, in the three geometries XY, RZ, RΘ. It is possible with this method to calculate geometric and composition criticalities and also to calculate the accessory problem on demand. The maximum number of points dealt with is 6000. The maximum permissible number of groups is 12. The internal iterations are treated by the method of alternating directions. The external iterations are accelerated using the extrapolation method due to Tchebychev. (authors) [fr

  11. Development of a polynomial nodal model to the multigroup transport equation in one dimension

    International Nuclear Information System (INIS)

    Feiz, M.

    1986-01-01

    A polynomial nodal model that uses Legendre polynomial expansions was developed for the multigroup transport equation in one dimension. The development depends upon the least-squares minimization of the residuals using the approximate functions over the node. Analytical expressions were developed for the polynomial coefficients. The odd moments of the angular neutron flux over the half ranges were used at the internal interfaces, and the Marshak boundary condition was used at the external boundaries. Sample problems with fine-mesh finite-difference solutions of the diffusion and transport equations were used for comparison with the model

  12. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, G.

    1976-05-01

    The purpose of this series is to inform scientists working on Fission Product Nuclear Data, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. This report consists of reproductions of essentially unaltered original contributions which the authors have sent to IAEA/NDS. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields; neutron cross-section data of fission products; data related to β-, γ-decay of fission products; delayed neutron data; and fission product decay-heat

  13. Fission dynamics in the proton induced fission of heavy nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Rubchenya, V.A. E-mail: rubchen@phys.jyu.fi; Trzaska, W.H.; Itkis, I.M.; Itkis, M.G.; Kliman, J.; Kniajeva, G.N.; Kondratiev, N.A.; Kozulin, E.M.; Krupa, L.; Pokrovski, I.V.; Voskressenski, V.M.; Hanappe, F.; Materna, T.; Dorvaux, O.; Stuttge, L.; Chubarian, G.; Khlebnikov, S.V.; Vakhtin, D.N.; Lyapin, V.G

    2004-04-05

    Multi-parameter correlation study of the reaction {sup 242}Pu(p, f) at E{sub p} 13, 20 and 55 MeV has been carried out. Fission fragment mass and kinetic energy distributions and the double differential neutron spectra have been measured. It was observed that the two-humped shape of mass distributions prevailed up to highest proton energy. Manifestation of the nuclear shell Z 28 near fragment mass A{sub fr} = 70 has been detected. The experimental results were analyzed in the framework of a time-dependent statistical model with inclusion of nuclear friction effects in the fission process. The multi-parameter correlation study of the reaction.

  14. Fission product behaviour in severe accidents

    International Nuclear Information System (INIS)

    Jokiniemi, J.; Auvinen, A.; Maekynen, J.; Valmari, T.

    1998-01-01

    The understanding of fission product (FP) behaviour in severe accidents is important for source term assessment and accident mitigation measures. For example in accident management the operator needs to know the effect of different actions on the behaviour and release of fission products. At VTT fission product behaviour have been studied in different national and international projects. In this presentation the results of projects in EU funded 4th framework programme Nuclear Fission Safety 1994-1998 are reported. The projects are: fission product vapour/aerosol chemistry in the primary circuit (FI4SCT960020), aerosol physics in containment (FI4SCT950016), revaporisation of test samples from Phebus fission products (FI4SCT960019) and assessment of models for fission product revaporisation (FI4SCT960044). Also results from the national project 'aerosol experiments in the Victoria facility' funded by IVO PE and VTT Energy are reported

  15. Material synergism fusion-fission

    International Nuclear Information System (INIS)

    Sankara Rao, K.B.; Raj, B.; Cook, I.; Kohyama, A.; Dudarev, S.

    2007-01-01

    In fission and fusion reactors the common features such as operating temperatures and neutron exposures will have the greatest impact on materials performance and component lifetimes. Developing fast neutron irradiation resisting materials is a common issue for both fission and fusion reactors. The high neutron flux levels in both these systems lead to unique materials problems like void swelling, irradiation creep and helium embitterment. Both fission and fusion rely on ferritic-martensitic steels based on 9%Cr compositions for achieving the highest swelling resistance but their creep strength sharply decreases above ∝ 823K. The use of oxide dispersion strengthened (ODS) alloys is envisaged to increase the operating temperature of blanket systems in the fusion reactors and fuel clad tubes in fast breeder reactors. In view of high operating temperatures, cyclic and steady load conditions and the long service life, properties like creep, low cycle fatigue,fracture toughness and creepfatigue interaction are major considerations in the selection of structural materials and design of components for fission and fusion reactors. Currently, materials selection for fusion systems has to be based upon incomplete experimental database on mechanical properties. The usage of fairly well developed databases, in fission programmes on similar materials, is of great help in the initial design of fusion reactor components. Significant opportunities exist for sharing information on technology of irradiation testing, specimen miniaturization, advanced methods of property measurement, safe windows for metal forming, and development of common materials property data base system. Both fusion and fission programs are being directed to development of clean steels with very low trace and tramp elements, characterization of microstructure and phase stability under irradiation, assessment of irradiation creep and swelling behaviour, studies on compatibility with helium and developing

  16. Linear triangle finite element formulation for multigroup neutron transport analysis with anisotropic scattering

    Energy Technology Data Exchange (ETDEWEB)

    Lillie, R.A.; Robinson, J.C.

    1976-05-01

    The discrete ordinates method is the most powerful and generally used deterministic method to obtain approximate solutions of the Boltzmann transport equation. A finite element formulation, utilizing a canonical form of the transport equation, is here developed to obtain both integral and pointwise solutions to neutron transport problems. The formulation is based on the use of linear triangles. A general treatment of anisotropic scattering is included by employing discrete ordinates-like approximations. In addition, multigroup source outer iteration techniques are employed to perform group-dependent calculations. The ability of the formulation to reduce substantially ray effects and its ability to perform streaming calculations are demonstrated by analyzing a series of test problems. The anisotropic scattering and multigroup treatments used in the development of the formulation are verified by a number of one-dimensional comparisons. These comparisons also demonstrate the relative accuracy of the formulation in predicting integral parameters. The applicability of the formulation to nonorthogonal planar geometries is demonstrated by analyzing a hexagonal-type lattice. A small, high-leakage reactor model is analyzed to investigate the effects of varying both the spatial mesh and order of angular quadrature. This analysis reveals that these effects are more pronounced in the present formulation than in other conventional formulations. However, the insignificance of these effects is demonstrated by analyzing a realistic reactor configuration. In addition, this final analysis illustrates the importance of incorporating anisotropic scattering into the finite element formulation. 8 tables, 29 figures.

  17. Linear triangle finite element formulation for multigroup neutron transport analysis with anisotropic scattering

    International Nuclear Information System (INIS)

    Lillie, R.A.; Robinson, J.C.

    1976-05-01

    The discrete ordinates method is the most powerful and generally used deterministic method to obtain approximate solutions of the Boltzmann transport equation. A finite element formulation, utilizing a canonical form of the transport equation, is here developed to obtain both integral and pointwise solutions to neutron transport problems. The formulation is based on the use of linear triangles. A general treatment of anisotropic scattering is included by employing discrete ordinates-like approximations. In addition, multigroup source outer iteration techniques are employed to perform group-dependent calculations. The ability of the formulation to reduce substantially ray effects and its ability to perform streaming calculations are demonstrated by analyzing a series of test problems. The anisotropic scattering and multigroup treatments used in the development of the formulation are verified by a number of one-dimensional comparisons. These comparisons also demonstrate the relative accuracy of the formulation in predicting integral parameters. The applicability of the formulation to nonorthogonal planar geometries is demonstrated by analyzing a hexagonal-type lattice. A small, high-leakage reactor model is analyzed to investigate the effects of varying both the spatial mesh and order of angular quadrature. This analysis reveals that these effects are more pronounced in the present formulation than in other conventional formulations. However, the insignificance of these effects is demonstrated by analyzing a realistic reactor configuration. In addition, this final analysis illustrates the importance of incorporating anisotropic scattering into the finite element formulation. 8 tables, 29 figures

  18. Variational P1 approximations of general-geometry multigroup transport problems

    International Nuclear Information System (INIS)

    Rulko, R.P.; Tomasevic, D.; Larsen, E.W.

    1995-01-01

    A variational approximation is developed for general-geometry multigroup transport problems with arbitrary anisotropic scattering. The variational principle is based on a functional that approximates a reaction rate in a subdomain of the system. In principle, approximations that result from this functional ''optimally'' determine such reaction rates. The functional contains an arbitrary parameter α and requires the approximate solutions of a forward and an adjoint transport problem. If the basis functions for the forward and adjoint solutions are chosen to be linear functions of the angular variable Ω, the functional yields the familiar multigroup P 1 equations for all values of α. However, the boundary conditions that result from the functional depend on α. In particular, for problems with vacuum boundaries, one obtains the conventional mixed boundary condition, but with an extrapolation distance that depends continuously on α. The choice α = 0 yields a generalization of boundary conditions derived earlier by Federighi and Pomraning for a more limited class of problems. The choice α = 1 yields a generalization of boundary conditions derived previously by Davis for monoenergetic problems. Other boundary conditions are obtained by choosing different values of α. The authors discuss this indeterminancy of α in conjunction with numerical experiments

  19. Interface discontinuity factors in the modal Eigenspace of the multigroup diffusion matrix

    International Nuclear Information System (INIS)

    Garcia-Herranz, N.; Herrero, J.J.; Cuervo, D.; Ahnert, C.

    2011-01-01

    Interface discontinuity factors based on the Generalized Equivalence Theory are commonly used in nodal homogenized diffusion calculations so that diffusion average values approximate heterogeneous higher order solutions. In this paper, an additional form of interface correction factors is presented in the frame of the Analytic Coarse Mesh Finite Difference Method (ACMFD), based on a correction of the modal fluxes instead of the physical fluxes. In the ACMFD formulation, implemented in COBAYA3 code, the coupled multigroup diffusion equations inside a homogenized region are reduced to a set of uncoupled modal equations through diagonalization of the multigroup diffusion matrix. Then, physical fluxes are transformed into modal fluxes in the Eigenspace of the diffusion matrix. It is possible to introduce interface flux discontinuity jumps as the difference of heterogeneous and homogeneous modal fluxes instead of introducing interface discontinuity factors as the ratio of heterogeneous and homogeneous physical fluxes. The formulation in the modal space has been implemented in COBAYA3 code and assessed by comparison with solutions using classical interface discontinuity factors in the physical space. (author)

  20. Continuous energy Monte Carlo method based homogenization multi-group constants calculation

    International Nuclear Information System (INIS)

    Li Mancang; Wang Kan; Yao Dong

    2012-01-01

    The efficiency of the standard two-step reactor physics calculation relies on the accuracy of multi-group constants from the assembly-level homogenization process. In contrast to the traditional deterministic methods, generating the homogenization cross sections via Monte Carlo method overcomes the difficulties in geometry and treats energy in continuum, thus provides more accuracy parameters. Besides, the same code and data bank can be used for a wide range of applications, resulting in the versatility using Monte Carlo codes for homogenization. As the first stage to realize Monte Carlo based lattice homogenization, the track length scheme is used as the foundation of cross section generation, which is straight forward. The scattering matrix and Legendre components, however, require special techniques. The Scattering Event method was proposed to solve the problem. There are no continuous energy counterparts in the Monte Carlo calculation for neutron diffusion coefficients. P 1 cross sections were used to calculate the diffusion coefficients for diffusion reactor simulator codes. B N theory is applied to take the leakage effect into account when the infinite lattice of identical symmetric motives is assumed. The MCMC code was developed and the code was applied in four assembly configurations to assess the accuracy and the applicability. At core-level, A PWR prototype core is examined. The results show that the Monte Carlo based multi-group constants behave well in average. The method could be applied to complicated configuration nuclear reactor core to gain higher accuracy. (authors)

  1. Global dynamics of multi-group SEI animal disease models with indirect transmission

    International Nuclear Information System (INIS)

    Wang, Yi; Cao, Jinde

    2014-01-01

    A challenge to multi-group epidemic models in mathematical epidemiology is the exploration of global dynamics. Here we formulate multi-group SEI animal disease models with indirect transmission via contaminated water. Under biologically motivated assumptions, the basic reproduction number R 0 is derived and established as a sharp threshold that completely determines the global dynamics of the system. In particular, we prove that if R 0 <1, the disease-free equilibrium is globally asymptotically stable, and the disease dies out; whereas if R 0 >1, then the endemic equilibrium is globally asymptotically stable and thus unique, and the disease persists in all groups. Since the weight matrix for weighted digraphs may be reducible, the afore-mentioned approach is not directly applicable to our model. For the proofs we utilize the classical method of Lyapunov, graph-theoretic results developed recently and a new combinatorial identity. Since the multiple transmission pathways may correspond to the real world, the obtained results are of biological significance and possible generalizations of the model are also discussed

  2. Longitudinal multigroup invariance analysis of the satisfaction with food-related life scale in university students.

    Science.gov (United States)

    Schnettler, Berta; Miranda, Horacio; Miranda-Zapata, Edgardo; Salinas-Oñate, Natalia; Grunert, Klaus G; Lobos, Germán; Sepúlveda, José; Orellana, Ligia; Hueche, Clementina; Bonilla, Héctor

    2017-06-01

    This study examined longitudinal measurement invariance in the Satisfaction with Food-related Life (SWFL) scale using follow-up data from university students. We examined this measure of the SWFL in different groups of students, separated by various characteristics. Through non-probabilistic longitudinal sampling, 114 university students (65.8% female, mean age: 22.5) completed the SWFL questionnaire three times, over intervals of approximately one year. Confirmatory factor analysis was used to examine longitudinal measurement invariance. Two types of analysis were conducted: first, a longitudinal invariance by time, and second, a multigroup longitudinal invariance by sex, age, socio-economic status and place of residence during the study period. Results showed that the 3-item version of the SWFL exhibited strong longitudinal invariance (equal factor loadings and equal indicator intercepts). Longitudinal multigroup invariance analysis also showed that the 3-item version of the SWFL displays strong invariance by socio-economic status and place of residence during the study period over time. Nevertheless, it was only possible to demonstrate equivalence of the longitudinal factor structure among students of both sexes, and among those older and younger than 22 years. Generally, these findings suggest that the SWFL scale has satisfactory psychometric properties for longitudinal measurement invariance in university students with similar characteristics as the students that participated in this research. It is also possible to suggest that satisfaction with food-related life is associated with sex and age. Copyright © 2017 Elsevier Ltd. All rights reserved.

  3. Gray and multigroup radiation transport models for two-dimensional binary stochastic media using effective opacities

    International Nuclear Information System (INIS)

    Olson, Gordon L.

    2016-01-01

    One-dimensional models for the transport of radiation through binary stochastic media do not work in multi-dimensions. Authors have attempted to modify or extend the 1D models to work in multidimensions without success. Analytic one-dimensional models are successful in 1D only when assuming greatly simplified physics. State of the art theories for stochastic media radiation transport do not address multi-dimensions and temperature-dependent physics coefficients. Here, the concept of effective opacities and effective heat capacities is found to well represent the ensemble averaged transport solutions in cases with gray or multigroup temperature-dependent opacities and constant or temperature-dependent heat capacities. In every case analyzed here, effective physics coefficients fit the transport solutions over a useful range of parameter space. The transport equation is solved with the spherical harmonics method with angle orders of n=1 and 5. Although the details depend on what order of solution is used, the general results are similar, independent of angular order. - Highlights: • Gray and multigroup radiation transport is done through 2D stochastic media. • Approximate models for the mean radiation field are found for all test problems. • Effective opacities are adjusted to fit the means of stochastic media transport. • Test problems include temperature dependent opacities and heat capacities • Transport solutions are done with angle orders n=1 and 5.

  4. Fission fragment spins and spectroscopy

    International Nuclear Information System (INIS)

    Durell, J.L.

    1988-01-01

    Prompt γ-ray coincidence experiments have been carried out on γ-rays emitted from post-neutron emission fission fragments produced by the aup 19F + 197 Au and 18 O + 232 Th reactions. Decay schemes have been established for even-even nuclei ranging from 78 Se to 148 Nd. Many new states with spin up to ∼ 12h have been observed. Apart from providing a wealth of new information on the spectroscopy of neutron-rich nuclei, the data have been analyzed to determine the average spin of primary fission fragments as a function of fragment mass. The results suggest that the fragment spins are determined by the temperature and shape of the primary fragments at or near to scission

  5. Radiation shielding for fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

    2000-03-01

    Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

  6. Nuclear Dissipation from Fission Time

    International Nuclear Information System (INIS)

    Gontchar, I.; Morjean, M.; Basnary, S.

    2000-01-01

    Fission times, pre-scission neutron multiplicities and GDR pre-scission γ-ray multiplicities measured for uranium or thorium nuclei formed with temperatures T ∼ 1.8 MeV have been compared with calculations performed with CDSM2, a two-dimensional dynamical model combined with a statistical one. Among the three experimental approaches considered, fission times give access to the most precise pieces of information on nuclear dissipation at high excitation energy. For the temperature range under consideration, an agreement between the model and data is achieved if one-body dissipation is used with a strength factor k red ∼ 0.45 ± 0.10 applied to the wall term for the mononuclear configuration. (authors)

  7. The fusion-fission hybrid

    International Nuclear Information System (INIS)

    Teller, E.

    1985-01-01

    As the history of the development of fusion energy shows, a sustained controlled fusion reaction is much more difficult to produce than rapid uncontrolled release of fusion energy. Currently, the ''magnetic bottle'' technique shows sufficient progress that it might applied for the commercial fuel production of /sup 233/U, suitable for use in fission reactors, by developing a fusion-fission hybrid. Such a device would consist of a fusion chamber core surrounded by a region containing cladded uranium pellets cooled by helium, with lithium salts also present to produce tritium to refuel the fusion process. Successful development of this hybrid might be possible within 10 y, and would provide both experience and funds for further development of controlled fusion energy

  8. Sexual differentiation in fission yeast

    DEFF Research Database (Denmark)

    Egel, R; Nielsen, O; Weilguny, D

    1990-01-01

    The regulation of sexual reproduction in yeast constitutes the highest level of differentiation observed in these unicellular organisms. The various ramifications of this system involve DNA rearrangement, transcriptional control, post-translational modification (such as protein phosphorylation) a......) and receptor/signal processing. A few basic similarities are common to both fission and budding yeasts. The wiring of the regulatory circuitry, however, varies considerably between these divergent yeast groups....

  9. Theories of fission gas behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Dias, J W.C. [Companhia Brasileira de Tecnologia Nuclear, Rio de Janeiro (Brazil). Diretoria de Tecnologia e Desenvolvimento; Merckx, K R

    1976-01-01

    A review is presented of the theoretical developments and experimental evidence that have helped to evolve current models used to describe the behavior of inert fission gases created during the irradiation of reactor fuel materials. The phenomena which are stressed relate primarily to steady state behavior of fuel elements but are also relevant to an understanding of transient behavior. The processes considered include gas atom solubility; gas atom diffusivity; bubble nucleation; and bubble growth by bubble coalescence.

  10. Fission tracks dating for obsidian

    International Nuclear Information System (INIS)

    Picon, C.

    1991-01-01

    Obsidian from South America are dated by fission tracks methods. Samples are irradiated in a nuclear reactor with a flux of 10 15 n/cm 2 . Results, corrected by 'Plateau' methods, are the following: obsidian from Bolivia: 4.14 x 10 6 yr., Ecuador: 8.79 x 10 5 yr., Colombia: 3.52 x 10 6 yr., Peru: 6.55 x 10 6 yr., Chile: 1.13 x 10 6 yr. (MMZ). 5 refs., 3 tabs

  11. The VERDI fission fragment spectrometer

    Directory of Open Access Journals (Sweden)

    Frégeau M.O.

    2013-12-01

    Full Text Available The VERDI time-of-flight spectrometer is dedicated to measurements of fission product yields and of prompt neutron emission data. Pre-neutron fission-fragment masses will be determined by the double time-of-flight (TOF technique. For this purpose an excellent time resolution is required. The time of flight of the fragments will be measured by electrostatic mirrors located near the target and the time signal coming from silicon detectors located at 50 cm on both sides of the target. This configuration, where the stop detector will provide us simultaneously with the kinetic energy of the fragment and timing information, significantly limits energy straggling in comparison to legacy experimental setup where a thin foil was usually used as a stop detector. In order to improve timing resolution, neutron transmutation doped silicon will be used. The high resistivity homogeneity of this material should significantly improve resolution in comparison to standard silicon detectors. Post-neutron fission fragment masses are obtained form the time-of-flight and the energy signal in the silicon detector. As an intermediary step a diamond detector will also be used as start detector located very close to the target. Previous tests have shown that poly-crystalline chemical vapour deposition (pCVD diamonds provides a coincidence time resolution of 150 ps not allowing complete separation between very low-energy fission fragments, alpha particles and noise. New results from using artificial single-crystal diamonds (sCVD show similar time resolution as from pCVD diamonds but also sufficiently good energy resolution.

  12. The discovery of uranium fission

    International Nuclear Information System (INIS)

    Brix, P.

    1990-01-01

    Uranium was discovered 200 years ago. Its radioactive character was first demonstrated in 1896 and two years later radium was extracted from uranium minerals. In 1911 studies with alpha rays from radioactive decay led to the unexpected discovery of the atomic nucleus. Exposure of beryllium to alpha rays yielded neutrons, first detected in 1932. Starting in 1934, neutron irradiation of uranium produced radioactive substances erroneously attributed to transuranium elements but with confusing properties. Painstaking experiments by chemists left no doubt on 17 December 1938 that barium was produced by these irradiations: the neutrons had split some uranium nuclei. The physics of the fission process was understood two weeks later; after a few months, neutron multiplication was found to be probable. This review deals with the eminent scientists involved, their successes, errors and disappointments, and the unexpected insights which occurred on the paths and detours of scientific research. It is, therefore, instructive also to discuss how fission was not discovered. The momentous discovery must be considered inevitable; the great tragedy was that Germany started World War II just at the time when the possibility of nuclear chain reactions and bombs became known. The consequences and anxieties that remain after 50 years of nuclear fission demand that mankind act with reason and conscience to maintain peace. (author)

  13. On Use of Multi-Chambered Fission Detectors for In-Core, Neutron Spectroscopy

    Science.gov (United States)

    Roberts, Jeremy A.

    2018-01-01

    Presented is a short, computational study on the potential use of multichambered fission detectors for in-core, neutron spectroscopy. Motivated by the development of very small fission chambers at CEA in France and at Kansas State University in the U.S., it was assumed in this preliminary analysis that devices can be made small enough to avoid flux perturbations and that uncertainties related to measurements can be ignored. It was hypothesized that a sufficient number of chambers with unique reactants can act as a real-time, foilactivation experiment. An unfolding scheme based on maximizing (Shannon) entropy was used to produce a flux spectrum from detector signals that requires no prior information. To test the method, integral, detector responses were generated for singleisotope detectors of various Th, U, Np, Pu, Am, and Cs isotopes using a simplified, pressurized-water reactor spectrum and fluxweighted, microscopic, fission cross sections, in the WIMS-69 multigroup format. An unfolded spectrum was found from subsets of these responses that had a maximum entropy while reproducing the responses considered and summing to one (that is, they were normalized). Several nuclide subsets were studied, and, as expected, the results indicate inclusion of more nuclides leads to better spectra but with diminishing improvements, with the best-case spectrum having an average, relative, group-wise error of approximately 51%. Furthermore, spectra found from minimum-norm and Tihkonov-regularization inversion were of lower quality than the maximum entropy solutions. Finally, the addition of thermal-neutron filters (here, Cd and Gd) provided substantial improvement over unshielded responses alone. The results, as a whole, suggest that in-core, neutron spectroscopy is at least marginally feasible.

  14. Improvement of the efficiency of two-dimensional multigroup transport calculations assuming isotropic reflection with multilevel spatial discretisation

    International Nuclear Information System (INIS)

    Stankovski, Z.; Zmijarevic, I.

    1987-06-01

    This paper presents two approximations used in multigroup two-dimensional transport calculations in large, very homogeneous media: isotropic reflection together with recently proposed group-dependent spatial representations. These approximations are implemented as standard options in APOLLO 2 assembly transport code. Presented example calculations show that significant savings in computational costs are obtained while preserving the overall accuracy

  15. Using Multi-Group Confirmatory Factor Analysis to Evaluate Cross-Cultural Research: Identifying and Understanding Non-Invariance

    Science.gov (United States)

    Brown, Gavin T. L.; Harris, Lois R.; O'Quin, Chrissie; Lane, Kenneth E.

    2017-01-01

    Multi-group confirmatory factor analysis (MGCFA) allows researchers to determine whether a research inventory elicits similar response patterns across samples. If statistical equivalence in responding is found, then scale score comparisons become possible and samples can be said to be from the same population. This paper illustrates the use of…

  16. The Power to Detect Sex Differences in IQ Test Scores Using Multi-Group Covariance and Means Structure Analyses

    Science.gov (United States)

    Molenaar, Dylan; Dolan, Conor V.; Wicherts, Jelle M.

    2009-01-01

    Research into sex differences in general intelligence, g, has resulted in two opposite views. In the first view, a g-difference is nonexistent, while in the second view, g is associated with a male advantage. Past research using Multi-Group Covariance and Mean Structure Analysis (MG-CMSA) found no sex difference in g. This failure raised the…

  17. MULTI - A multigroup or multipoint P{sub 3} programme for calculating thermal neutron spectra in a reactor cell

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M V [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1968-06-15

    Programme MULTI calculates the space energy distribution of thermal neutrons in a multizone, cylindrical, infinitely long reactor lattice by using the multigroup or multipoint P{sub 3} approximation. This report presents a short description of the algorithm and the programme and gives the instructions for its exploitation. (author)

  18. Algorithm development and verification of UASCM for multi-dimension and multi-group neutron kinetics model

    International Nuclear Information System (INIS)

    Si, S.

    2012-01-01

    The Universal Algorithm of Stiffness Confinement Method (UASCM) for neutron kinetics model of multi-dimensional and multi-group transport equations or diffusion equations has been developed. The numerical experiments based on transport theory code MGSNM and diffusion theory code MGNEM have demonstrated that the algorithm has sufficient accuracy and stability. (authors)

  19. Validation of multigroup neutron cross sections for the Advanced Neutron Source against the FOEHN critical experimental measurements

    International Nuclear Information System (INIS)

    Smith, L.A.; Gehin, J.C.; Worley, B.A.; Renier, J.P.

    1994-01-01

    The FOEHN critical experiments were analyzed to validate the use of multigroup cross sections in the design of the Advanced Neutron Source. Eleven critical configurations were evaluated using the KENO, DORT, and VENTURE neutronics codes. Eigenvalue and power density profiles were computed and show very good agreement with measured values

  20. Analysis of sensitive questions across cultures : An application of multigroup item randomized response theory to sexual attitudes and behavior

    NARCIS (Netherlands)

    de Jong, M.G.; Pieters, R.; Stremersch, S.

    2012-01-01

    Answers to sensitive questions are prone to social desirability bias. If not properly addressed, the validity of the research can be suspect. This article presents multigroup item randomized response theory (MIRRT) to measure self-reported sensitive topics across cultures. The method was

  1. Device for measuring fission product density

    International Nuclear Information System (INIS)

    Kaneda, Mitsunori.

    1980-01-01

    Purpose: To determine the fission product density of xenon or the like and enable measurement of real time of fission product density in a reactor by calculating the disintegration and annihilation of the fission product on the basis of neutron detected output. Constitution: The neutron flux in a reactor is detected by a detector, and applied to first and second density calculators. Second fission product density signal of xenon or the like outputted from first device is again inputted to the device to form an annihilation signal due to disintegration to determine the present density of the second fission product of xenon or the like corresponding to the decrease of the neutron due to the poison of xeron or the like. Similarly, second device determines the first fission product density of iodine or the like. (Sekiya, K.)

  2. The resonance neutron fission on heavy nuclei

    International Nuclear Information System (INIS)

    Kopach, Yu.N.; Popov, A.B.; Furman, V.I.; Alfimenkov, V.P.; Lason', L.; Pikel'ner, L.B.; ); Gonin, N.N.; Kozlovskij, L.K.; Tambovtsev, D.I.; Gagarskij, A.M.; Petrov, G.A.; Sokolov, V.E.

    2001-01-01

    A new approach to the description of the fission, similar to the well-known reaction theory and based on the helicity representation for the exit fission channels, is briefly summarized. This approach allows one to connect the multimodal fission representation with A. Bohr's concept of the fission transition states and to obtain formulae for the partial and differential fission cross sections. The formulae are used for analysis of the angular anisotropy of fragments in the neutron resonance induced fission of aligned 235 U nuclei and of the P-even angular forward-backward and right-left correlations of fragments oe the P-odd correlations caused by the interference of s- and p-wave neutron resonances [ru

  3. MPI version of NJOY and its application to multigroup cross-section generation

    Energy Technology Data Exchange (ETDEWEB)

    Alpan, A.; Haghighat, A.

    1999-07-01

    Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances

  4. MPI version of NJOY and its application to multigroup cross-section generation

    International Nuclear Information System (INIS)

    Alpan, A.; Haghighat, A.

    1999-01-01

    Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances, temperatures

  5. Evolvement simulation of the probability of neutron-initiating persistent fission chain

    International Nuclear Information System (INIS)

    Wang Zhe; Hong Zhenying

    2014-01-01

    Background: Probability of neutron-initiating persistent fission chain, which has to be calculated in analysis of critical safety, start-up of reactor, burst waiting time on pulse reactor, bursting time on pulse reactor, etc., is an inherent parameter in a multiplying assembly. Purpose: We aim to derive time-dependent integro-differential equation for such probability in relative velocity space according to the probability conservation, and develop the deterministic code Dynamic Segment Number Probability (DSNP) based on the multi-group S N method. Methods: The reliable convergence of dynamic calculation was analyzed and numerical simulation of the evolvement process of dynamic probability for varying concentration was performed under different initial conditions. Results: On Highly Enriched Uranium (HEU) Bare Spheres, when the time is long enough, the results of dynamic calculation approach to those of static calculation. The most difference of such results between DSNP and Partisn code is less than 2%. On Baker model, over the range of about 1 μs after the first criticality, the most difference between the dynamic and static calculation is about 300%. As for a super critical system, the finite fission chains decrease and the persistent fission chains increase as the reactivity aggrandizes, the dynamic evolvement curve of initiation probability is close to the static curve within the difference of 5% when the K eff is more than 1.2. The cumulative probability curve also indicates that the difference of integral results between the dynamic calculation and the static calculation decreases from 35% to 5% as the K eff increases. This demonstrated that the ability of initiating a self-sustaining fission chain reaction approaches stabilization, while the former difference (35%) showed the important difference of the dynamic results near the first criticality with the static ones. The DSNP code agrees well with Partisn code. Conclusions: There are large numbers of

  6. Fission-energy release for 16 fissioning nuclides. Final report

    International Nuclear Information System (INIS)

    Sher, R.

    1981-03-01

    Results are presented of a least-squares evaluation of the components of energy release per fission in 232 Th, 233 U, 235 U, 238 U, 239 Pu, and 241 Pu. For completeness, older (1978) results based on systematics are presented for these and ten other isotopes of interest. There have been recent indications that the delayed energy components may be somewhat higher than those used previously, but the LSQ results do not seem to change significantly when modest (approx. 1 MeV) increases in the total delayed energy are included in the inputs. Additional measurements of most of the energy components are still needed to resolve remaining discrepancies

  7. What do we learn on the dynamics of fission from α-accompanied fission data

    International Nuclear Information System (INIS)

    Guet, C.; Asghar, M.; Nifenecker, H.; Perrin, P.

    1978-01-01

    Measurements of the angular distribution of α-particles emitted by thermal fission of 236 U are presented. Also the dependence of the angular distribution on the kinetic energy of the fission products is studied. (WL) [de

  8. Comparative evaluation of solar, fission, fusion, and fossil energy resources. Part 2: Power from nuclear fission

    Science.gov (United States)

    Clement, J. D.

    1973-01-01

    Different types of nuclear fission reactors and fissionable materials are compared. Special emphasis is placed upon the environmental impact of such reactors. Graphs and charts comparing reactor facilities in the U. S. are presented.

  9. Nuclear fission with a Langevin equation

    International Nuclear Information System (INIS)

    Boilley, D.; Suraud, E.; Abe, Yasuhisa

    1992-01-01

    A microscopically derived Langevin equation is applied to thermally induced nuclear fission. An important memory effect is pointed out and discussed. A strong friction coefficient, estimated from microscopic quantities, tends to decrease the stationary limit of the fission rate and to increase the transient time. The calculations are performed with a collective mass depending on the collective variable and with a constant mass. Fission rates calculated at different temperatures are shown and compared with previous available results. (author) 23 refs.; 7 figs

  10. Monte carlo sampling of fission multiplicity.

    Energy Technology Data Exchange (ETDEWEB)

    Hendricks, J. S. (John S.)

    2004-01-01

    Two new methods have been developed for fission multiplicity modeling in Monte Carlo calculations. The traditional method of sampling neutron multiplicity from fission is to sample the number of neutrons above or below the average. For example, if there are 2.7 neutrons per fission, three would be chosen 70% of the time and two would be chosen 30% of the time. For many applications, particularly {sup 3}He coincidence counting, a better estimate of the true number of neutrons per fission is required. Generally, this number is estimated by sampling a Gaussian distribution about the average. However, because the tail of the Gaussian distribution is negative and negative neutrons cannot be produced, a slight positive bias can be found in the average value. For criticality calculations, the result of rejecting the negative neutrons is an increase in k{sub eff} of 0.1% in some cases. For spontaneous fission, where the average number of neutrons emitted from fission is low, the error also can be unacceptably large. If the Gaussian width approaches the average number of fissions, 10% too many fission neutrons are produced by not treating the negative Gaussian tail adequately. The first method to treat the Gaussian tail is to determine a correction offset, which then is subtracted from all sampled values of the number of neutrons produced. This offset depends on the average value for any given fission at any energy and must be computed efficiently at each fission from the non-integrable error function. The second method is to determine a corrected zero point so that all neutrons sampled between zero and the corrected zero point are killed to compensate for the negative Gaussian tail bias. Again, the zero point must be computed efficiently at each fission. Both methods give excellent results with a negligible computing time penalty. It is now possible to include the full effects of fission multiplicity without the negative Gaussian tail bias.

  11. Fission-product retention in HTGR fuels

    International Nuclear Information System (INIS)

    Homan, F.J.; Kania, M.J.; Tiegs, T.N.

    1982-01-01

    Retention data for gaseous and metallic fission products are presented for both Triso-coated and Biso-coated HTGR fuel particles. Performance trends are established that relate fission product retention to operating parameters, such as temperature, burnup, and neutron exposure. It is concluded that Biso-coated particles are not adequately retentive of fission gas or metallic cesium, and Triso-coated particles which retain cesium still lose silver. Design implications related to these performance trends are identified and discussed

  12. Measurements of Fission Cross Sections of Actinides

    CERN Multimedia

    Wiescher, M; Cox, J; Dahlfors, M

    2002-01-01

    A measurement of the neutron induced fission cross sections of $^{237}$Np, $^{241},{243}$Am and of $^{245}$Cm is proposed for the n_TOF neutron beam. Two sets of fission detectors will be used: one based on PPAC counters and another based on a fast ionization chamber (FIC). A total of 5x10$^{18}$ protons are requested for the entire fission measurement campaign.

  13. Neutron-induced fission cross sections

    International Nuclear Information System (INIS)

    Weigmann, H.

    1991-01-01

    In the history of fission research, neutron-induced fission has always played the most important role. The practical importance of neutron-induced fission rests upon the fact that additional neutrons are produced in the fission process, and thus a chain reaction becomes possible. The practical applications of neutron-induced fission will not be discussed in this chapter, but only the physical properties of one of its characteristics, namely (n,f) cross sections. The most important early summaries on the subject are the monograph edited by Michaudon which also deals with the practical applications, the earlier review article on fission by Michaudon, and the review by Bjornholm and Lynn, in which neutron-induced fission receives major attention. This chapter will attempt to go an intermediate way between the very detailed theoretical treatment in the latter review and the cited monograph which emphasizes the applied aspects and the techniques of fission cross-section measurements. The more recent investigations in the field will be included. Section II will survey the properties of cross sections for neutron-induced fission and also address some special aspects of the experimental methods applied in their measurement. Section Ill will deal with the formal theory of neutron-induced nuclear reactions for the resolved resonance region and the region of statistical nuclear reactions. In Section IV, the fission width, or fission transmission coefficient, will be discussed in detail. Section V will deal with the broader structures due to incompletely damped vibrational resonances, and in particular will address the special case of thorium and neighboring isotopes. Finally, Section VI will briefly discuss parity violation effects in neutron-induced fission. 74 refs., 14 figs., 3 tabs

  14. Fission of nuclei far from stability

    International Nuclear Information System (INIS)

    Schmidt, K.H.; Benlliure, J.; Junghans, A.R.

    2000-11-01

    The secondary-beam facility of GSI provided the technical equipment for a new kind of fission experiment. Fission properties of short-lived neutron-deficient nuclei have been investigated in inverse kinematics. The measured element distributions reveal new kinds of systematics on shell structure and even-odd effects and lead to an improved understanding of structure effects in nuclear fission. Prospects for further experimental studies are discussed. (orig.)

  15. International conference on fifty years research in nuclear fission

    International Nuclear Information System (INIS)

    1989-02-01

    These proceedings contain extended abstracts of the papers presented at the named conference. They deal with static properties of fission, instrumentation for fission studies, fission in compound-nucleus reactions, fission dynamics, fission-like heavy ion reactions, and fusion reactions. See hints under the relevant topics. (HSI)

  16. Status of fission product yield data

    International Nuclear Information System (INIS)

    Cuninghame, J.G.

    1978-01-01

    The topics covered in this paper are: (a) cumulative yields in thermal neutron fission and in fast fission up to 14 MeV incident neutron energy, (b) dependence of the yields on incident neutron energy and spectrum, (c) independent yields, (d) charge dispersion and distribution, and (e) yields of light particles from ternary fission. The paper reviews information on these subjects for fission of actinides from 232 Th upwards with special emphasis on data published since the 1973 Bologna FPND Panel, compares data sets, and discusses the gaps still to be found in them. (author)

  17. True ternary fission in 310126X

    International Nuclear Information System (INIS)

    Banupriya, B.; Vijayaraghavan, K.R.; Balasubramaniam, M.

    2015-01-01

    All possible combinations are minimized by the two dimensional minimization process and minimized with respect to neutron numbers and proton numbers of the fragments. Potential energy is low and Q - value is high at true ternary fission region. It shows that true ternary mode is the dominant mode in the ternary fission of superheavy nuclei. Also, the results show that the fragments with neutron magic numbers are the dominant one in the ternary fission of superheavy nuclei whereas the fragments with proton magic numbers are the dominant one in the ternary fission of heavy nuclei

  18. Attachment of gaseous fission products to aerosols

    International Nuclear Information System (INIS)

    Skyrme, G.

    1985-01-01

    Accidents may occur in which the integrity of fuel cladding is breached and volatile fission products are released to the containment atmosphere. In order to assess the magnitude of the subsequent radiological hazard it is necessary to know the transport behaviour of such fission products. It is frequently assumed that the fission products remain in the gaseous phase. There is a possibility, however, that they may attach themselves to particles and hence substantially modify their transport properties. This paper provides a theoretical assessment of the conditions under which gaseous fission products may be attached to aerosol particles. Specific topics discussed are: the mass transfer of a gaseous fission product to an isolated aerosol particle in an infinite medium; the rate at which the concentration of fission products in the gas phase diminishes within a container as a result of deposition on a population of particles; and the distribution of deposited fission product between different particle sizes in a log-normal distribution. It is shown that, for a given mass, small particles are more efficient for fission product attachment, and that only small concentrations of such particles may be necessary to achieve rapid attachment. Conditions under which gaseous fission products are not attached to particles are also considered, viz, the competing processes of deposition onto the containment walls and onto aerosol particles, and the possibility of the removal of aerosols from the containment by various deposition processes, or agglomeration, before attachment takes place. (author)

  19. Theoretical Description of the Fission Process

    International Nuclear Information System (INIS)

    Witold Nazarewicz

    2003-01-01

    The main goals of the project can be summarized as follows: Development of effective energy functionals that are appropriate for the description of heavy nuclei. Our goal is to improve the existing energy density (Skyrme) functionals to develop a force that will be used in calculations of fission dynamics. Systematic self-consistent calculations of binding energies and fission barriers of actinide and trans-actinide nuclei using modern density functionals. This will be followed by calculations of spontaneous fission lifetimes and mass and charge divisions using dynamic adiabatic approaches based on the WKB approximation. Investigate novel microscopic (non-adiabatic) methods to study the fission process

  20. Cold valleys in fusion and fission

    International Nuclear Information System (INIS)

    Misicu, S.

    2003-01-01

    The cold fission configuration after the preformation of the fragments resembles a short-lived dinuclear or quasi-molecular system. The most conceivable scission configuration is given by two fission fragments in touching with the symmetry axes aligned (pole-pole orientation). This conclusion was based on the simple argument that this configuration offers the optimal tunneling time, i.e. the difference between the Coulomb barrier and the decay energy Q is minimal. Other orientations are apparently precluded in cold spontaneous fission and should be regarded as quasi-fission doorways in the synthesis of superheavy elements by cold fusion. (orig.)

  1. Nuclear fission studies: from LOHENGRIN to FIPPS

    International Nuclear Information System (INIS)

    Chebboubi, Abdelaziz

    2015-01-01

    Nuclear fission consists in splitting a nucleus, in general an actinide, into smaller nuclei. Despite nuclear fission was discovered in 1939 by Hahn and Strassman, fission models cannot predict the fission observables with an acceptable accuracy for nuclear fuel cycle studies for instance. Improvement of fission models is an important issue for the knowledge of the process itself and for the applications. To reduce uncertainties of the nuclear data used in a nuclear reactor simulation, a validation of the models hypothesis is mandatory. In this work, two features of the nuclear fission were investigated in order to test the resistance of the theories. One aspect is the study of the symmetric fission fragments through the measurement of their yield and kinetic energy distribution. The other aspect is the study of the fission fragment angular momentum.Two techniques are available to assess the angular momentum of a fission fragment. The first one is to look at the properties of the prompt gamma. The new spectrometer FIPPS (Fission Product Prompt gamma-ray Spectrometer), is currently under development at the ILL and will combine a fission filter with a large array of gamma and neutron detectors in order to respond to these issues. The first part of this work is dedicated to the study of the properties of a Gas Filled Magnet (GFM) which is the type of fission filter considered for the FIPPS project.The second part of this work deals with the measurement of isomeric yields and evaluations of the angular momentum distribution of fission fragments. The study of the spherical nucleus 132 Sn shed the light on the current limits of fission models. Finally, the last part of this work is about the measurement of the yields and kinetic energy distributions of symmetric fission fragments. Since models predict the existence of fission modes, the symmetry region is a suitable choice to investigate this kind of prediction. In parallel with all these studies, an emphasis on the

  2. Process for the extraction of fission products

    International Nuclear Information System (INIS)

    Anav, M.; Chesne, A.; Leseur, A.; Miquel, P.; Pascard, R.

    1979-01-01

    A process is described for the extraction of fission products contained in irradiated nuclear fuel elements which have been subject to a temperature of at least 1200 0 C during their irradiation prior to dissolving the fuel by the wet process. After mechanically treating the elements in order to decan and/or cut them they are brought into contact with water in order to pass the fission products into aqueous solution. The treated elements are then separated from the thus obtained aqueous solution. At least one of the fission products is then recovered from the aqueous solution. The fission products are iodine, cesium, rubidium and tritium

  3. Theoretical Description of the Fission Process

    Energy Technology Data Exchange (ETDEWEB)

    Witold Nazarewicz

    2003-07-01

    The main goals of the project can be summarized as follows: Development of effective energy functionals that are appropriate for the description of heavy nuclei. Our goal is to improve the existing energy density (Skyrme) functionals to develop a force that will be used in calculations of fission dynamics. Systematic self-consistent calculations of binding energies and fission barriers of actinide and trans-actinide nuclei using modern density functionals. This will be followed by calculations of spontaneous fission lifetimes and mass and charge divisions using dynamic adiabatic approaches based on the WKB approximation. Investigate novel microscopic (non-adiabatic) methods to study the fission process.

  4. A spin exchange model for singlet fission

    Science.gov (United States)

    Yago, Tomoaki; Wakasa, Masanobu

    2018-03-01

    Singlet fission has been analyzed with the Dexter model in which electron exchange occurs between chromophores, conserving the spin for each electron. In the present study, we propose a spin exchange model for singlet fission. In the spin exchange model, spins are exchanged by the exchange interaction between two electrons. Our analysis with simple spin functions demonstrates that singlet fission is possible by spin exchange. A necessary condition for spin exchange is a variation in exchange interactions. We also adapt the spin exchange model to triplet fusion and triplet energy transfer, which often occur after singlet fission in organic solids.

  5. Fission - track age of the Marjalahti Pallasite

    International Nuclear Information System (INIS)

    Bondar, Yu.V.; Perelygin, V.P.

    2006-01-01

    Full text: Investigation of fossil charged-particle tracks in various mineral phases of extraterrestrial samples is a powerful method for research the early stages of the solar system. Over geological time, meteorites crystals have accumulated a record of tracks produced by heavily charged energetic particles from both internal (spontaneous fission of 238U and some other extinct isotopes) and external sources (galactic cosmic rays with Z>20). The fortunate fact that meteorite grains can accumulate latent and very long-lived tracks since soon after the end of nucleosynthesis in the solar nebula enables one to decode their radiation history and to detect any thermal events in the meteorite cosmic history by revealing these tracks through suitable etching procedures. Only a few minerals in meteorites (mainly phosphates) contain small amount of uranium; the fact that 238 U undergoes fission with fission-decay constant λ f ∼ 8.2x10 -17 yr -1 allows one to use this isotope as a chronometer. By measuring the U concentration in the crystals (by reactor irradiation) and the density of the spontaneous-fission tracks it is relatively easy to calculate the 'fission-track age' if 238 U is the main source of fission tracks. However the fission-track dating of extraterrestrial samples compared with the terrestrial ones has some peculiar features due to presence of a number of other potential track sources except the spontaneous fission of 238 U, such as the spontaneous fission of presently extinct 244 Pu, heavy nuclei of cosmic rays and induced fission by cosmic ray primaries. Only tracks from the spontaneous fission of U and Pu are suitable for fission-track dating. The competing effects of these fissioning elements, whose half-lives differ by a factor of ∼50, form a basis for a fission-track chronology for samples older than ∼ 4.0 Gyr. Over small intervals in time (∼ few x10 8 yr ) the track density from spontaneous fission of 238 U is nearly constant. However, the

  6. Solution for the multigroup neutron space kinetics equations by the modified Picard algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Tavares, Matheus G.; Petersen, Claudio Z., E-mail: matheus.gulartetavares@gmail.com [Universidade Federal de Pelotas (UFPEL), Capao do Leao, RS (Brazil). Departamento de Matematica e Estatistica; Schramm, Marcelo, E-mail: schrammmarcelo@gmail.com [Universidade Federal de Pelotas (UFPEL), RS (Brazil). Centro de Engenharias; Zanette, Rodrigo, E-mail: rodrigozanette@hotmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Instituto de Matematica e Estatistica

    2017-07-01

    In this work, we used a modified Picards method to solve the Multigroup Neutron Space Kinetics Equations (MNSKE) in Cartesian geometry. The method consists in assuming an initial guess for the neutron flux and using it to calculate a fictitious source term in the MNSKE. A new source term is calculated applying its solution, and so on, iteratively, until a stop criterion is satisfied. For the solution of the fast and thermal neutron fluxes equations, the Laplace Transform technique is used in time variable resulting in a rst order linear differential matrix equation, which are solved by classical methods in the literature. After each iteration, the scalar neutron flux and the delayed neutron precursors are reconstructed by polynomial interpolation. We obtain the fluxes and precursors through Numerical Inverse Laplace Transform using the Stehfest method. We present numerical simulations and comparisons with available results in literature. (author)

  7. Approximate analytical solution of two-dimensional multigroup P-3 equations

    International Nuclear Information System (INIS)

    Matausek, M.V.; Milosevic, M.

    1981-01-01

    Iterative solution of multigroup spherical harmonics equations reduces, in the P-3 approximation and in two-dimensional geometry, to a problem of solving an inhomogeneous system of eight ordinary first order differential equations. With appropriate boundary conditions, these equations have to be solved for each energy group and in each iteration step. The general solution of the corresponding homogeneous system of equations is known in analytical form. The present paper shows how the right-hand side of the system can be approximated in order to derive a particular solution and thus an approximate analytical expression for the general solution of the inhomogeneous system. This combined analytical-numerical approach was shown to have certain advantages compared to the finite-difference method or the Lie-series expansion method, which have been used to solve similar problems. (orig./RW) [de

  8. Solution of the multilayer multigroup neutron diffusion equation in cartesian geometry by fictitious borders power method

    Energy Technology Data Exchange (ETDEWEB)

    Zanette, Rodrigo; Petersen, Caudio Zen [Univ. Federal de Pelotas, Capao do Leao (Brazil). Programa de Pos Graduacao em Modelagem Matematica; Schramm, Marcello [Univ. Federal de Pelotas (Brazil). Centro de Engenharias; Zabadal, Jorge Rodolfo [Univ. Federal do Rio Grande do Sul, Tramandai (Brazil)

    2017-05-15

    In this paper a solution for the one-dimensional steady state Multilayer Multigroup Neutron Diffusion Equation in cartesian geometry by Fictitious Borders Power Method and a perturbative analysis of this solution is presented. For each new iteration of the power method, the neutron flux is reconstructed by polynomial interpolation, so that it always remains in a standard form. However when the domain is long, an almost singular matrix arises in the interpolation process. To eliminate this singularity the domain segmented in R regions, called fictitious regions. The last step is to solve the neutron diffusion equation for each fictitious region in analytical form locally. The results are compared with results present in the literature. In order to analyze the sensitivity of the solution, a perturbation in the nuclear parameters is inserted to determine how a perturbation interferes in numerical results of the solution.

  9. Program to solve the multigroup discrete ordinates transport equation in (x,y,z) geometry

    International Nuclear Information System (INIS)

    Lathrop, K.D.

    1976-04-01

    Numerical formulations and programming algorithms are given for the THREETRAN computer program which solves the discrete ordinates, multigroup transport equation in (x,y,z) geometry. An efficient, flexible, and general data-handling strategy is derived to make use of three hierarchies of storage: small core memory, large core memory, and disk file. Data management, input instructions, and sample problem output are described. A six-group, S 4 , 18 502 mesh point, 2 800 zone, k/sub eff/ calculation of the ZPPR-4 critical assembly required 144 min of CDC-7600 time to execute to a convergence tolerance of 5 x 10 -4 and gave results in good qualitative agreement with experiment and other calculations. 6 references

  10. MINX: a multigroup interpretation of nuclear X-sections from ENDF/B

    International Nuclear Information System (INIS)

    Weisbin, C.R.; Soran, P.D.; MacFarlane, R.E.; Harris, D.R.; LaBauve, R.J.; Hendricks, J.S.; White, J.E.; Kidman, R.B.

    1976-09-01

    MINX calculates fine-group averaged infinitely dilute cross sections, self-shielding factors, and group-to-group transfer matrices from ENDF/B-IV data. Its primary purpose is to generate pseudo-composition independent multigroup libraries in the standard CCCC-III interface formats for use in the design and analysis of nuclear systems. MINX incorporates and improves upon the resonance capabilities of existing codes such as ETOX and ENDRUN and the high-Legendre-order transfer matrices of ETOG and SUPERTOG. Group structure, Legendre order, weight function, temperature, dilutions, and processing tolerances are all under user control. Paging and variable dimensioning allow very large problems to be run. Both CDC and IBM versions of MINX are available

  11. Solution of the Multigroup-Diffusion equation by the response matrix method

    International Nuclear Information System (INIS)

    Oliveira, C.R.E.

    1980-10-01

    A preliminary analysis of the response matrix method is made, considering its application to the solution of the multigroup diffusion equations. The one-dimensional formulation is presented and used to test some flux expansions, seeking the application of the method to the two-dimensional problem. This formulation also solves the equations that arise from the integro-differential synthesis algorithm. The slow convergence of the power method, used to solve the eigenvalue problem, and its acceleration by means of the Chebyshev polynomial method, are also studied. An algorithm for the estimation of the dominance ratio is presented, based on the residues of two successive iteration vectors. This ratio, which is not known a priori, is fundamental for the efficiency of the method. Some numerical problems are solved, testing the 1D formulation of the response matrix method, its application to the synthesis algorithm and also, at the same time, the algorithm to accelerate the source problem. (Author) [pt

  12. Global dynamics of a novel multi-group model for computer worms

    International Nuclear Information System (INIS)

    Gong Yong-Wang; Song Yu-Rong; Jiang Guo-Ping

    2013-01-01

    In this paper, we study worm dynamics in computer networks composed of many autonomous systems. A novel multi-group SIQR (susceptible-infected-quarantined-removed) model is proposed for computer worms by explicitly considering anti-virus measures and the network infrastructure. Then, the basic reproduction number of worm R 0 is derived and the global dynamics of the model are established. It is shown that if R 0 is less than or equal to 1, the disease-free equilibrium is globally asymptotically stable and the worm dies out eventually, whereas, if R 0 is greater than 1, one unique endemic equilibrium exists and it is globally asymptotically stable, thus the worm persists in the network. Finally, numerical simulations are given to illustrate the theoretical results. (general)

  13. Solution for the multigroup neutron space kinetics equations by the modified Picard algorithm

    International Nuclear Information System (INIS)

    Tavares, Matheus G.; Petersen, Claudio Z.; Schramm, Marcelo; Zanette, Rodrigo

    2017-01-01

    In this work, we used a modified Picards method to solve the Multigroup Neutron Space Kinetics Equations (MNSKE) in Cartesian geometry. The method consists in assuming an initial guess for the neutron flux and using it to calculate a fictitious source term in the MNSKE. A new source term is calculated applying its solution, and so on, iteratively, until a stop criterion is satisfied. For the solution of the fast and thermal neutron fluxes equations, the Laplace Transform technique is used in time variable resulting in a rst order linear differential matrix equation, which are solved by classical methods in the literature. After each iteration, the scalar neutron flux and the delayed neutron precursors are reconstructed by polynomial interpolation. We obtain the fluxes and precursors through Numerical Inverse Laplace Transform using the Stehfest method. We present numerical simulations and comparisons with available results in literature. (author)

  14. The Nodal Polynomial Expansion method to solve the multigroup diffusion equations

    International Nuclear Information System (INIS)

    Ribeiro, R.D.M.

    1983-03-01

    The methodology of the solutions of the multigroup diffusion equations and uses the Nodal Polynomial Expansion Method is covered. The EPON code was developed based upon the above mentioned method for stationary state, rectangular geometry, one-dimensional or two-dimensional and for one or two energy groups. Then, one can study some effects such as the influence of the baffle on the thermal flux by calculating the flux and power distribution in nuclear reactors. Furthermore, a comparative study with other programs which use Finite Difference (CITATION and PDQ5) and Finite Element (CHD and FEMB) Methods was undertaken. As a result, the coherence, feasibility, speed and accuracy of the methodology used were demonstrated. (Author) [pt

  15. Multigroup Path Analysis of the Influence of Healthcare Quality, by Different Health Insurance Types.

    Science.gov (United States)

    Hong, Yong-Rock; Holcomb, Derek; Ballard, Michael; Schwartz, Laurel

    Winds of change have been blowing in the U.S. healthcare system since passage of the Affordable Care Act. Examining differences between individuals covered by different types of insurance is essential if healthcare executives are to develop new strategies in response to the emerging health insurance market. In this study, we used multigroup path analysis models to examine the moderating effects of health insurance on direct and indirect associations with general health status, satisfaction with received care, financial burden, and perceived value of the healthcare system. Data were obtained from the 2012 Medical Expenditure Panel Survey and analyzed according to the types of insurance: private, public, and military. With the satisfactory fit of the model (χ = 2,532.644, df = 96, p spending.

  16. Approximate analytical solution of two-dimensional multigroup P-3 equations

    International Nuclear Information System (INIS)

    Matausek, M.V.; Milosevic, M.

    1981-01-01

    Iterative solution of multigroup spherical harmonics equations reduces, in the P-3 approximation and in two-dimensional geometry, to a problem of solving an inhomogeneous system of eight ordinary first order differential equations. With appropriate boundary conditions, these equations have to be solved for each energy group and in each iteration step. The general solution of the corresponding homogeneous system of equations is known in analytical form. The present paper shows how the right-hand side of the system can be approximated in order to derive a particular solution and thus an approximate analytical expression for the general solution of the inhomogeneous system. This combined analytical-numerical approach was shown to have certain advantages compared to the finite-difference method or the Lie-series expansion method, which have been used to solve similar problems. (author)

  17. ETOA, ABBN Multigroup Constants from ENDF/B for Fast Reactors

    International Nuclear Information System (INIS)

    Nishimura, Hideo

    1977-01-01

    1 - Nature of physical problem solved: Production of ABBN type group constants up to 70 groups for fast reactor calculations, reading ENDF/B library as input. 2 - Method of solution: The multigroup method of Bondarenko et al. is used for processing basic nuclear data. Calculational algorithms for an unresolved resonance region are the same as those in the MC 2 code. For a resolved resonance region, an ultrafine energy structure dependent on a level scheme is adopted. 3 - Restrictions on the complexity of the problem: Maximum number of: energy groups: 70; sigma 0 values: 6; temperatures: 5. Self-shielding factors for an unrealistically low value of sigma 0 are not guaranteed because of the approximations used in the unresolved resonance region

  18. AMPX: a modular system for multigroup cross-section generation and manipulation

    International Nuclear Information System (INIS)

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Diggs, B.R.; Webster, C.C.; Lucius, J.L.; White, J.E.; Wright, R.Q.; Westfall, R.M.

    1978-01-01

    The AMPX system, developed at the Oak Ridge National Laboratory over the past seven years, is a collection of computer programs in a modular arrangement. Starting with ENDF-formatted nuclear data files, the system includes a full range of features needed to produce and use multigroup neutron, gamma-ray production, and gamma-ray interaction cross-section data. The balance between production and analysis is roughly even; thus, the system serves a wide variety of needs. The modularity is particularly attractive, since it allows the user to choose an arbitrary execution sequence from the approximately 40 to 50 modules available in the system. The modularity also allows selection from different treatments; e.g., the Nordheim method, a full-blown integral transport calculation, the Bondarenko method, or other alternative can be selected for resonance shielding. 2 figures

  19. Analytic solutions of the multigroup space-time reactor kinetics equations

    International Nuclear Information System (INIS)

    Lee, C.E.; Rottler, S.

    1986-01-01

    The development of analytical and numerical solutions to the reactor kinetics equations is reviewed. Analytic solutions of the multigroup space-time reactor kinetics equations are developed for bare and reflected slabs and spherical reactors for zero flux, zero current and extrapolated endpoint boundary conditions. The material properties of the reactors are assumed constant in space and time, but spatially-dependent source terms and initial conditions are investigated. The system of partial differential equations is reduced to a set of linear ordinary differential equations by the Laplace transform method. These equations are solved by matrix Green's functions yielding a general matrix solution for the neutron flux and precursor concentration in the Laplace transform space. The detailed pole structure of the Laplace transform matrix solutions is investigated. The temporally- and spatially-dependent solutions are determined from the inverse Laplace transform using the Cauchy residue theorem, the theorem of Frobenius, a knowledge of the detailed pole structure and matrix operators. (author)

  20. Spectrum of the multigroup neutron transport operator for bounded spatial domains

    International Nuclear Information System (INIS)

    Larsen, E.W.

    1979-01-01

    The spectrum of the multigroup neutron transport operator A is studied for bounded spatial regions D which consist of a finite number of material subregions. Our main results provide simple conditions on the material cross sections which guarantee that (1) A possesses eigenvalues in the finite plane; (2) A possesses a ''leading'' eigenvalue lambda 0 which is real, not less than the real part of any other eigenvalue, and to which there corresponds at least one nonnegative eigenfunction psi/sub lambda/0; and (3) A possesses a ''dominant'' eigenvalue lambda 0 which is real, simple, greater than the real part of any other eigenvalue, and whose eigenfunction psi/sub lambda/0 satisfies psi/sub lambda/0> or =0 and ∫psi/sub lambda/0d 2 Ω>0. We give examples to illustrate the results and to show that a leading eigenvalue need not be simple, nor its eigenfunction(s) positive

  1. The solution of the multigroup neutron transport equation using spherical harmonics

    International Nuclear Information System (INIS)

    Fletcher, K.

    1981-01-01

    A solution of the multi-group neutron transport equation in up to three space dimensions is presented. The flux is expanded in a series of unnormalised spherical harmonics. Using the various recurrence formulae a linked set of first order differential equations is obtained for the moments psisup(g)sub(lm)(r), γsup(g)sub(lm)(r). Terms with odd l are eliminated resulting in a second order system which is solved by two methods. The first is a finite difference formulation using an iterative procedure, secondly, in XYZ and XY geometry a finite element solution is given. Results for a test problem using both methods are exhibited and compared. (orig./RW) [de

  2. Recent validation experience with multigroup cross-section libraries and scale

    International Nuclear Information System (INIS)

    Bowman, S.M.; Wright, R.Q.; DeHart, M.D.; Parks, C.V.; Petrie, L.M.

    1995-01-01

    This paper will discuss the results obtained and lessons learned from an extensive validation of new ENDF/B-V and ENDF/B-VI multigroup cross-section libraries using analyses of critical experiments. The KENO V. a Monte Carlo code in version 4.3 of the SCALE computer code system was used to perform the critical benchmark calculations via the automated SCALE sequence CSAS25. The cross-section data were processed by the SCALE automated problem-dependent resonance-processing procedure included in this sequence. Prior to calling KENO V.a, CSAS25 accesses BONAMI to perform resonance self-shielding for nuclides with Bondarenko factors and NITAWL-II to process nuclides with resonance parameter data via the Nordheim Integral Treatment

  3. LABAN-PEL: a two-dimensional, multigroup diffusion, high-order response matrix code

    International Nuclear Information System (INIS)

    Mueller, E.Z.

    1991-06-01

    The capabilities of LABAN-PEL is described. LABAN-PEL is a modified version of the two-dimensional, high-order response matrix code, LABAN, written by Lindahl. The new version extends the capabilities of the original code with regard to the treatment of neutron migration by including an option to utilize full group-to-group diffusion coefficient matrices. In addition, the code has been converted from single to double precision and the necessary routines added to activate its multigroup capability. The coding has also been converted to standard FORTRAN-77 to enhance the portability of the code. Details regarding the input data requirements and calculational options of LABAN-PEL are provided. 13 refs

  4. Analytical synthetic methods of solution of neutron transport equation with diffusion theory approaches energy multigroup

    International Nuclear Information System (INIS)

    Moraes, Pedro Gabriel B.; Leite, Michel C.A.; Barros, Ricardo C.

    2013-01-01

    In this work we developed a software to model and generate results in tables and graphs of one-dimensional neutron transport problems in multi-group formulation of energy. The numerical method we use to solve the problem of neutron diffusion is analytic, thus eliminating the truncation errors that appear in classical numerical methods, e.g., the method of finite differences. This numerical analytical method increases the computational efficiency, since they are not refined spatial discretization necessary because for any spatial discretization grids used, the numerical result generated for the same point of the domain remains unchanged unless the rounding errors of computational finite arithmetic. We chose to develop a computational application in MatLab platform for numerical computation and program interface is simple and easy with knobs. We consider important to model this neutron transport problem with a fixed source in the context of shielding calculations of radiation that protects the biosphere, and could be sensitive to ionizing radiation

  5. SIRIUS - A one-dimensional multigroup analytic nodal diffusion theory code

    Energy Technology Data Exchange (ETDEWEB)

    Forslund, P. [Westinghouse Atom AB, Vaesteraas (Sweden)

    2000-09-01

    In order to evaluate relative merits of some proposed intranodal cross sections models, a computer code called Sirius has been developed. Sirius is a one-dimensional, multigroup analytic nodal diffusion theory code with microscopic depletion capability. Sirius provides the possibility of performing a spatial homogenization and energy collapsing of cross sections. In addition a so called pin power reconstruction method is available for the purpose of reconstructing 'heterogeneous' pin qualities. consequently, Sirius has the capability of performing all the calculations (incl. depletion calculations) which are an integral part of the nodal calculation procedure. In this way, an unambiguous numerical analysis of intranodal cross section models is made possible. In this report, the theory of the nodal models implemented in sirius as well as the verification of the most important features of these models are addressed.

  6. A position sensitive parallel plate avalanche fission detector for use in particle induced fission coincidence measurements

    NARCIS (Netherlands)

    Plicht, J. van der

    1980-01-01

    A parallel plate avalanche detector developed for the detection of fission fragments in particle induced fission reactions is described. The active area is 6 × 10 cm2; it is position sensitive in one dimension with a resolution of 2.5 mm. The detector can withstand a count rate of 25000 fission

  7. Baby fission chambers; Etude de chambres a fission miniatures

    Energy Technology Data Exchange (ETDEWEB)

    Guery, U; Tachon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    The present report is intended, on the one band, as a study of the main types of fission chambers produced to date, and on the other, to deal more generally with this type of detector. Originally, it was with a view to the charting of neutron scatter in 'Proserpine' that the authors undertook the study of these chambers. During the course of the task, it was considered worth tbe trouble of developing its scope to include a more general application: neutron scatter measurement of various energy neutrons within a reduced volume with slight local disturbance. (author) [French] Le present rapport se propose, d'une part, d'exposer les principales realisations de chambres a fission, d'autre part de faire une mise au point a caractere plus general sur ces detecteurs. Au depart, c'est surtout en vue des mesures de densite neutronique dans 'Proserpine' que les auteurs ont etudie ces chambres; au cours de la mise au point, il a paru interessant de developper leur etude pour des applications plus generales: mesures de densites de neutrons de differentes energies dans un element de volume tres reduit et avec faible perturbation locale. (auteur)

  8. Baby fission chambers; Etude de chambres a fission miniatures

    Energy Technology Data Exchange (ETDEWEB)

    Guery, U.; Tachon, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    The present report is intended, on the one band, as a study of the main types of fission chambers produced to date, and on the other, to deal more generally with this type of detector. Originally, it was with a view to the charting of neutron scatter in 'Proserpine' that the authors undertook the study of these chambers. During the course of the task, it was considered worth tbe trouble of developing its scope to include a more general application: neutron scatter measurement of various energy neutrons within a reduced volume with slight local disturbance. (author) [French] Le present rapport se propose, d'une part, d'exposer les principales realisations de chambres a fission, d'autre part de faire une mise au point a caractere plus general sur ces detecteurs. Au depart, c'est surtout en vue des mesures de densite neutronique dans 'Proserpine' que les auteurs ont etudie ces chambres; au cours de la mise au point, il a paru interessant de developper leur etude pour des applications plus generales: mesures de densites de neutrons de differentes energies dans un element de volume tres reduit et avec faible perturbation locale. (auteur)

  9. An accurate solution of point reactor neutron kinetics equations of multi-group of delayed neutrons

    International Nuclear Information System (INIS)

    Yamoah, S.; Akaho, E.H.K.; Nyarko, B.J.B.

    2013-01-01

    Highlights: ► Analytical solution is proposed to solve the point reactor kinetics equations (PRKE). ► The method is based on formulating a coefficient matrix of the PRKE. ► The method was applied to solve the PRKE for six groups of delayed neutrons. ► Results shows good agreement with other traditional methods in literature. ► The method is accurate and efficient for solving the point reactor kinetics equations. - Abstract: The understanding of the time-dependent behaviour of the neutron population in a nuclear reactor in response to either a planned or unplanned change in the reactor conditions is of great importance to the safe and reliable operation of the reactor. In this study, an accurate analytical solution of point reactor kinetics equations with multi-group of delayed neutrons for specified reactivity changes is proposed to calculate the change in neutron density. The method is based on formulating a coefficient matrix of the homogenous differential equations of the point reactor kinetics equations and calculating the eigenvalues and the corresponding eigenvectors of the coefficient matrix. A small time interval is chosen within which reactivity relatively stays constant. The analytical method was applied to solve the point reactor kinetics equations with six-groups delayed neutrons for a representative thermal reactor. The problems of step, ramp and temperature feedback reactivities are computed and the results compared with other traditional methods. The comparison shows that the method presented in this study is accurate and efficient for solving the point reactor kinetics equations of multi-group of delayed neutrons

  10. Testing a new multigroup inference approach to reconstructing past environmental conditions

    Directory of Open Access Journals (Sweden)

    Maria RIERADEVALL

    2008-08-01

    Full Text Available A new, quantitative, inference model for environmental reconstruction (transfer function, based for the first time on the simultaneous analysis of multigroup species, has been developed. Quantitative reconstructions based on palaeoecological transfer functions provide a powerful tool for addressing questions of environmental change in a wide range of environments, from oceans to mountain lakes, and over a range of timescales, from decades to millions of years. Much progress has been made in the development of inferences based on multiple proxies but usually these have been considered separately, and the different numeric reconstructions compared and reconciled post-hoc. This paper presents a new method to combine information from multiple biological groups at the reconstruction stage. The aim of the multigroup work was to test the potential of the new approach to making improved inferences of past environmental change by improving upon current reconstruction methodologies. The taxonomic groups analysed include diatoms, chironomids and chrysophyte cysts. We test the new methodology using two cold-environment training-sets, namely mountain lakes from the Pyrenees and the Alps. The use of multiple groups, as opposed to single groupings, was only found to increase the reconstruction skill slightly, as measured by the root mean square error of prediction (leave-one-out cross-validation, in the case of alkalinity, dissolved inorganic carbon and altitude (a surrogate for air-temperature, but not for pH or dissolved CO2. Reasons why the improvement was less than might have been anticipated are discussed. These can include the different life-forms, environmental responses and reaction times of the groups under study.

  11. Fission fragment excited laser system

    Science.gov (United States)

    McArthur, David A.; Tollefsrud, Philip B.

    1976-01-01

    A laser system and method for exciting lasing action in a molecular gas lasing medium which includes cooling the lasing medium to a temperature below about 150 K and injecting fission fragments through the lasing medium so as to preferentially excite low lying vibrational levels of the medium and to cause population inversions therein. The cooled gas lasing medium should have a mass areal density of about 5 .times. 10.sup.-.sup.3 grams/square centimeter, relaxation times of greater than 50 microseconds, and a broad range of excitable vibrational levels which are excitable by molecular collisions.

  12. Fission tracks diameters in glasses

    International Nuclear Information System (INIS)

    Garzon Ruiperez, L.; Veiguela, J.

    1974-01-01

    Standard glass microscope slides have been irradiated with fission fragments from the uranium. The etching track conditions have been the same for the series, having changed the etching time only for each specimen. For each glass, a minimum of 250 measurements of the tracks diameters have been made, the distributions of which are the bimodal type. Diameters-etching dependence with time is roughly lineal. Energy determinations have been made with the help of the diameters-energy relations. The calculated values agree very well with the know ones. (author) [es

  13. Fission barriers of superheavy nuclei

    International Nuclear Information System (INIS)

    Burvenich, T.

    2001-01-01

    Full text: Self consistent microscopic mean-field models are powerful tools for the description of nuclear structure phenomena in the region of known elements, where they have reached a good quality. Especially the Skyrme-Hartree-Fock (SHF) method and the Relativistic Mean-Field (RMF) model will be considered in the discussion of the properties of these models. When it comes to extrapolation to the region of superheavy elements, though there is agreement concerning the global trends, these model exhibit significant differences in their predictions concerning fission barrier heights and structures. (Author)

  14. Least squares analysis of fission neutron standard fields

    International Nuclear Information System (INIS)

    Griffin, P.J.; Williams, J.G.

    1997-01-01

    A least squares analysis of fission neutron standard fields has been performed using the latest dosimetry cross sections. Discrepant nuclear data are identified and adjusted spectra for 252 Cf spontaneous fission and 235 U thermal fission fields are presented

  15. Neutronics of Laser Fission-Fusion Systems

    International Nuclear Information System (INIS)

    Velarde, G.

    1976-01-01

    Neutronics of Fission-Fusion microsystems inertially confined by Lasers are analysed by transport calculation, both stationary (DTF, TIHOC) and time dependent (TDA, TIHEX), discussing the results obtained for the basic parameters of the fission process (multiplication factor, neutron generation time and Rossi-∞). (Author) 14 refs

  16. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  17. Neutronics of Laser Fission-Fusion Systems

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, G

    1976-07-01

    Neutronics of Fission-Fusion microsystems inertially confined by Lasers are analysed by transport calculation, both stationary (DTF, TIHOC) and time dependent (TDA, TIHEX), discussing the results obtained for the basic parameters of the fission process (multiplication factor, neutron generation time and Rossi-{infinity}). (Author) 14 refs.

  18. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1981-06-01

    This is the seventh issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields (neutron induced and spontaneous fission); neutron reaction cross sections of fission products; data related to the radioactive decay of fission products; delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The sixth issue of this series has been published in June 1980 as INDC(NDS)-113/G+P. The present issue includes contributions which were received by NDS between 1 August 1980 and 25 May 1981

  19. Options for Affordable Fission Surface Power Systems

    International Nuclear Information System (INIS)

    Houts, Mike; Gaddis, Steve; Porter, Ron; Van Dyke, Melissa; Martin, Jim; Godfroy, Tom; Bragg-Sitton, Shannon; Garber, Anne; Pearson, Boise

    2006-01-01

    Fission surface power systems could provide abundant power anywhere on the surface of the moon or Mars. Locations could include permanently shaded regions on the moon and high latitudes on Mars. To be fully utilized, however, fission surface power systems must be safe, have adequate performance, and be affordable. This paper discusses options for the design and development of such systems. (authors)

  20. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1982-07-01

    This is the eighth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. Therefore, the IAEA cannot be held responsible for the information contained nor for any consequences resulting from the use of this information. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); Neutron reaction cross sections of fission products; Data related to the radioactive decay of fission products; Delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The seventh issue of this series has been published in July 1981 as INDC(NDS)-116. The present issue includes contributions which were received by NDS between 1 August 1981 and 15 June 1982

  1. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, M.

    1983-08-01

    This is the ninth issue of a report series on Fission Product Nuclear Data (FPND) which is published by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The purpose of this series is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The main part of this report consists of unaltered original contributions which the authors have sent to IAEA/NDS. The present issue contains also a section with some recent references relative to fission product nuclear data, which were not covered by the contributions submitted. The types of activities being included in this report are measurements, compilations and evaluations of: Fission product yields (neutron induced and spontaneous fission); Neutron reaction cross sections of fission products; Data related to the radioactive decay of fission products; Delayed neutron data of fission products; and lumped fission product data (decay heat, absorption etc.). The eighth issue of this series has been published in July 1982 as INDC(NDS)-130. The present issue includes contributions which were received by NDS between 1 August 1982 and 25 June 1983

  2. Induced-Fission Imaging of Nuclear Material

    International Nuclear Information System (INIS)

    Hausladen, Paul; Blackston, Matthew A.; Mullens, James Allen; McConchie, Seth M.; Mihalczo, John T.; Bingham, Philip R.; Ericson, Milton Nance; Fabris, Lorenzo

    2010-01-01

    This paper presents initial results from development of the induced-fission imaging technique, which can be used for the purpose of measuring or verifying the distribution of fissionable material in an unopened container. The technique is based on stimulating fissions in nuclear material with 14 MeV neutrons from an associated-particle deuterium-tritium (D-T) generator and counting the subsequent induced fast fission neutrons with an array of fast organic scintillation detectors. For each source neutron incident on the container, the neutron creation time and initial trajectory are known from detection of the associated alpha particle of the d + t → α + n reaction. Many induced fissions will lie along (or near) the interrogating neutron path, allowing an image of the spatial distribution of prompt induced fissions, and thereby fissionable material, to be constructed. A variety of induced-fission imaging measurements have been performed at Oak Ridge National Laboratory with a portable, low-dose D-T generator, including single-view radiographic measurements and three-dimensional tomographic measurements. Results from these measurements will be presented along with the neutron transmission images that have been performed simultaneously. This new capability may have applications to a number of areas in which there may be a need to confirm the presence or configuration of nuclear materials, such as nuclear material control and accountability, quality assurance, treaty confirmation, or homeland security applications.

  3. Cumulative fission yield of Ce-148 produced by thermal-neutron fission of U-235

    International Nuclear Information System (INIS)

    Hasan, A.A.

    1984-12-01

    Cumulative fission yield of 148 cesium isotopes and some other fission products produced by thermal-neutron fission of 235 uranium is determined by Germanium/Lithium spectroscopic methods. The measuremets were done at Tsing-Hua open pool reactor using 3 to 4 mg of 93.15% enriched 235 uranium samples. Gamma rays are assigned to the responsible fission products by matching gamma rays energies and half lives. Fission rate is calculated by fission track method. Cumulative fission yields of 148 cesium, 90 krypton, 130 iodine, 144 lanthanum, 89 krypton, 136 xenon, 137 xenon and 140 cesium are calculated. This values are compared with previously predicted values and showed good agreement. 21 Ref

  4. Theoretical Description of the Fission Process

    International Nuclear Information System (INIS)

    Nazarewicz, Witold

    2009-01-01

    Advanced theoretical methods and high-performance computers may finally unlock the secrets of nuclear fission, a fundamental nuclear decay that is of great relevance to society. In this work, we studied the phenomenon of spontaneous fission using the symmetry-unrestricted nuclear density functional theory (DFT). Our results show that many observed properties of fissioning nuclei can be explained in terms of pathways in multidimensional collective space corresponding to different geometries of fission products. From the calculated collective potential and collective mass, we estimated spontaneous fission half-lives, and good agreement with experimental data was found. We also predicted a new phenomenon of trimodal spontaneous fission for some transfermium isotopes. Our calculations demonstrate that fission barriers of excited superheavy nuclei vary rapidly with particle number, pointing to the importance of shell effects even at large excitation energies. The results are consistent with recent experiments where superheavy elements were created by bombarding an actinide target with 48-calcium; yet even at high excitation energies, sizable fission barriers remained. Not only does this reveal clues about the conditions for creating new elements, it also provides a wider context for understanding other types of fission. Understanding of the fission process is crucial for many areas of science and technology. Fission governs existence of many transuranium elements, including the predicted long-lived superheavy species. In nuclear astrophysics, fission influences the formation of heavy elements on the final stages of the r-process in a very high neutron density environment. Fission applications are numerous. Improved understanding of the fission process will enable scientists to enhance the safety and reliability of the nation's nuclear stockpile and nuclear reactors. The deployment of a fleet of safe and efficient advanced reactors, which will also minimize radiotoxic

  5. MC2-2, Calculation of Fast Neutron Spectra and Multigroup Cross-Sections from ENDF/B Data

    International Nuclear Information System (INIS)

    2001-01-01

    1 - Description of program or function: MC 2 -2 solves the neutron slowing-down equations using basic neutron data derived from ENDF/B data files to determine fundamental mode spectra for use in generating multigroup neutron cross sections. The current edition includes the ability to treat all ENDF/B-V and -VI data representations. It accommodates high-order P scattering representations and provides numerous capabilities such as isotope mixing, delayed neutron processing, free-format input, and flexibility in output data selection. This edition supersedes previous releases of the MC22 program and the earlier MC2 program. Improved physics algorithms and increased computational efficiency are incorporated. Input data files required by MC2-2 may be generated from ENDF/B data by the code ETOE-2. The hyper-fine-group integral transport theory module of MC2-2, RABANL, is an improved version of the RABBLE/RABID codes. Many of the MC2-2 modules are used in the SDX code. 2 - Methods: The extended transport P1, B1, consistent P1, and consistent B1 fundamental mode ultra-fine-group equations are solved using continuous slowing-down theory and multigroup methods. Fast and accurate resonance integral methods are used in the narrow resonance resolved and unresolved resonance treatments. A fundamental mode homogeneous unit cell calculation is performed using either a multigroup or a continuous slowing-down treatment. Multigroup neutron homogeneous cross sections are generated in an ISOTXS format for an arbitrary group structure. A hyper-fine-group integral transport slowing down calculation (RABANL) is available as an option. RABANL performs a homogeneous or heterogeneous (pin or slab) unit cell calculation over the resonance region (resolved and unresolved) and generates multigroup neutron cross sections in an ISOTXS format. Neutron cross sections are generated by RABANL for the homogeneous unit cell and for each heterogeneous region in the pin or slab unit cell calculation

  6. Contained fissionly vaporized imploded fission explosive breeder reactor

    International Nuclear Information System (INIS)

    Marwick, E.F.

    1978-01-01

    Disclosed is a nuclear reactor system which produces useful thermal power and breeds fissile isotopes wherein large spherical complex slugs containing fissile and fertile isotopes as well as vaporizing and tamping materials are exploded seriatim in a large containing chamber having walls protected from the effects of the explosion by about two thousand tons of slurry of fissile and fertile isotopes in molten alkali metal. The slug which is slightly sub-critical prior to its entry into the centroid portion of the chamber, then becomes slightly more than prompt-critical because of the near proximity of neutron-reflecting atoms and of fissioning atoms within the slurry. The slurry is heated by explosion of the slugs and serves as a working fluid for extraction of heat energy from the reactor. Explosive debris is precipitated from the slurry and used for the fabrication of new slugs

  7. Fission gas behaviour in water reactor fuels

    International Nuclear Information System (INIS)

    2002-01-01

    During irradiation, nuclear fuel changes volume, primarily through swelling. This swelling is caused by the fission products and in particular by the volatile ones such as krypton and xenon, called fission gas. Fission gas behaviour needs to be reliably predicted in order to make better use of nuclear fuel, a factor which can help to achieve the economic competitiveness required by today's markets. These proceedings communicate the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors. (author)

  8. Physics of neutron emission in fission

    International Nuclear Information System (INIS)

    Lemmel, H.D.

    1989-06-01

    The document contains the proceedings of the IAEA Consultants' Meeting on the Physics of Neutron Emission in Fission, Mito City (Japan), 24-27 May 1988. Included are the conclusions and recommendations reached at the meeting and the papers presented by the meeting participants. These papers cover the following topics: Energy dependence of the number of fission neutrons ν-bar (3 papers), multiplicity distribution of fission neutrons (3 papers), competition between neutron and γ-ray emission (4 papers), the fission neutron yield in resonances (2 papers) and the energy spectrum of fission neutrons in experiment (9 papers), theory (4 papers) and evaluation (1 paper). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  9. Systematics of Fission-Product Yields

    International Nuclear Information System (INIS)

    Wahl, A.C.

    2002-01-01

    Empirical equations representing systematics of fission-product yields have been derived from experimental data. The systematics give some insight into nuclear-structure effects on yields, and the equations allow estimation of yields from fission of any nuclide with atomic number Z F = 90 thru 98, mass number A F = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru ∼200 MeV--the ranges of these quantities for the fissioning nuclei investigated. Calculations can be made with the computer program CYFP. Estimates of uncertainties in the yield estimates are given by equations, also in CYFP, and range from ∼ 15% for the highest yield values to several orders of magnitude for very small yield values. A summation method is used to calculate weighted average parameter values for fast-neutron (∼ fission spectrum) induced fission reactions

  10. Progress in fission product nuclear data

    International Nuclear Information System (INIS)

    Lammer, G.

    1975-01-01

    This is the first issue of a report series on Fission Product Nuclear Data (FPND), published every six months by the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). Its purpose is to inform scientists working on FPND, or using such data, about all activities in this field which are planned, ongoing, or have recently been completed. The types of activities being included in this report are measurements, compilations and evaluations of: fission product yields; neutron cross-section data of fission products; data related to β-, γ-decay of fission products; delayed neutron data; and fission product decay-heat. The present issue includes contributions which were received by NDS before 1 November 1975

  11. Theoretical study of fission dynamics with muons

    International Nuclear Information System (INIS)

    Oberacker, V.E.; Umar, A.S.; Bottcher, C.; Strayer, M.R.; Maruhn, J.A.; Frankfurt Univ.

    1992-01-01

    Following muon capture by actinide atoms, some of the inner shell muonic transitions proceed by inverse internal conversion, i.e. the excitation energy of the muonic atom is transferred to the nucleus. In particular, the muonic E2:(3d→1s) transition energy is close to the peak of the isoscalar giant quadrupole resonance in actinide nuclei which exhibits a large fission width. Prompt fission in the presence of a bound muon allows us to study the dynamics of large-amplitude collective motion. We solve the time-dependent Dirac equation for the muonic spinor wave function in the Coulomb field of the fissioning nucleus on a 3-dimensional lattice and demonstrate that the muon attachment probability to the light fission fragment is a measure of the nuclear energy dissipation between the outer fission barrier and the scission point

  12. Theory of neutron emission in fission

    International Nuclear Information System (INIS)

    Madland, D.G.

    1989-01-01

    Following a summary of the observables in neutron emission in fission, a brief history is given of theoretical representations of the prompt fission neutron spectrum N(E) and average prompt neutron multiplicity /bar /nu///sub p/. This is followed by descriptions, together with examples, of modern approaches to the calculation of these quantities including recent advancements. Emphasis will be placed upon the predictability and accuracy of the modern approaches. In particular, the dependence of N(E) and /bar /nu///sub p/ on the fissioning nucleus and its excitation energy will be discussed, as will the effects of and competition between first-, second- and third-chance fission in circumstances of high excitation energy. Finally, properties of neutron-rich (fission-fragment) nuclei are discussed that must be better known to calculate N(E) and /bar /nu///sub p/ with higher accuracy than is currently possible. 17 refs., 11 figs

  13. Systematics of Fission-Product Yields

    Energy Technology Data Exchange (ETDEWEB)

    A.C. Wahl

    2002-05-01

    Empirical equations representing systematics of fission-product yields have been derived from experimental data. The systematics give some insight into nuclear-structure effects on yields, and the equations allow estimation of yields from fission of any nuclide with atomic number Z{sub F} = 90 thru 98, mass number A{sub F} = 230 thru 252, and precursor excitation energy (projectile kinetic plus binding energies) PE = 0 thru {approx}200 MeV--the ranges of these quantities for the fissioning nuclei investigated. Calculations can be made with the computer program CYFP. Estimates of uncertainties in the yield estimates are given by equations, also in CYFP, and range from {approx} 15% for the highest yield values to several orders of magnitude for very small yield values. A summation method is used to calculate weighted average parameter values for fast-neutron ({approx} fission spectrum) induced fission reactions.

  14. Fission yield covariance generation and uncertainty propagation through fission pulse decay heat calculation

    International Nuclear Information System (INIS)

    Fiorito, L.; Diez, C.J.; Cabellos, O.; Stankovskiy, A.; Van den Eynde, G.; Labeau, P.E.

    2014-01-01

    Highlights: • Fission yield data and uncertainty comparison between major nuclear data libraries. • Fission yield covariance generation through Bayesian technique. • Study of the effect of fission yield correlations on decay heat calculations. • Covariance information contribute to reduce fission pulse decay heat uncertainty. - Abstract: Fission product yields are fundamental parameters in burnup/activation calculations and the impact of their uncertainties was widely studied in the past. Evaluations of these uncertainties were released, still without covariance data. Therefore, the nuclear community expressed the need of full fission yield covariance matrices to be able to produce inventory calculation results that take into account the complete uncertainty data. State-of-the-art fission yield data and methodologies for fission yield covariance generation were researched in this work. Covariance matrices were generated and compared to the original data stored in the library. Then, we focused on the effect of fission yield covariance information on fission pulse decay heat results for thermal fission of 235 U. Calculations were carried out using different libraries and codes (ACAB and ALEPH-2) after introducing the new covariance values. Results were compared with those obtained with the uncertainty data currently provided by the libraries. The uncertainty quantification was performed first with Monte Carlo sampling and then compared with linear perturbation. Indeed, correlations between fission yields strongly affect the uncertainty of decay heat. Eventually, a sensitivity analysis of fission product yields to fission pulse decay heat was performed in order to provide a full set of the most sensitive nuclides for such a calculation

  15. Contribution to the solution of the multigroup Boltzmann equation by the determinist methods and the Monte Carlo method; Contribution a la resolution de l`equation de Bolztmann en multigroupe par les methodes deterministes et Monte-Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Li, M

    1998-08-01

    In this thesis, two methods for solving the multigroup Boltzmann equation have been studied: the interface-current method and the Monte Carlo method. A new version of interface-current (IC) method has been develop in the TDT code at SERMA, where the currents of interface are represented by piecewise constant functions in the solid angle space. The convergence of this method to the collision probability (CP) method has been tested. Since the tracking technique is used for both the IC and CP methods, it is necessary to normalize he collision probabilities obtained by this technique. Several methods for this object have been studied and implemented in our code, we have compared their performances and chosen the best one as the standard choice. The transfer matrix treatment has been a long-standing difficulty for the multigroup Monte Carlo method: when the cross-sections are converted into multigroup form, important negative parts will appear in the angular transfer laws represented by low-order Legendre polynomials. Several methods based on the preservation of the first moments, such as the discrete angles methods and the equally-probable step function method, have been studied and implemented in the TRIMARAN-II code. Since none of these codes has been satisfactory, a new method, the non equally-probably step function method, has been proposed and realized in our code. The comparisons for these methods have been done in several aspects: the preservation of the moments required, the calculation of a criticality problem and the calculation of a neutron-transfer in water problem. The results have showed that the new method is the best one in all these comparisons, and we have proposed that it should be a standard choice for the multigroup transfer matrix. (author) 76 refs.

  16. Downstream behavior of fission products

    International Nuclear Information System (INIS)

    Johnson, I.; Farahat, M.K.; Settle, J.L.; Johnson, C.E.; Ritzman, R.

    1986-01-01

    The downstream behavior of fission products has been investigated by injecting mixtures of CsOH, CsI, and Te into a flowing steam/hydrogen stream and determining the physical and chemical changes that took place as the gaseous mixture flowed down a reaction duct on which a temperature gradient (1000 0 to 200 0 C) had been imposed. Deposition on the wall of the duct occurred by vapor condensation in the higher temperature regions and by aerosol deposition in the remainder of the duct. Reactions in the gas stream between CsOH and CsI and between CsOH and Te had an effect on the vapor condensation. The aerosol was characterized by the use of impingement tabs placed in the gas stream

  17. Laser driven fusion fission hybrids

    International Nuclear Information System (INIS)

    Hansen, L.F.; Maniscalco, J.A.

    1977-11-01

    The role of the fusion-fission hybrid reactor (FFHR) as a fissile fuel and/or power producer is discussed. As long range options to supply the world energy needs, hybrid-fueled thermal-burner reactors are compared to liquid metal fast breeder reactors (LMFBR). A discussion of different fuel cycles (thorium, depleted uranium, and spent fuel) is presented in order to compare the energy multiplication, the production of fissile fuel, the laser efficiency and pellet gain requirements of the hybrid reactor. Lawrence Livermore Laboratory (LLL) has collaborated with Bechtel Corporation and with Westinghouse in two engineering design studies of laser fusion driven hybrid power plants. The hybrid designs which have resulted from these two studies are briefly described and analyzed by considering operational parameters, such as energy multiplication, power density, burn-up and plutonium production as a function time

  18. Aerosols and fission product transport

    International Nuclear Information System (INIS)

    Megaw, W.J.

    1987-12-01

    A survey is presented of current knowledge of the possible role of aerosols in the consequences of in- and out-of-core LOCAs and of end fitting failures in CANDU reactors. An extensive literature search has been made of research on the behaviour of aerosols in possible accidents in water moderated and cooled reactors and the results of various studies compared. It is recommended that further work should be undertaken on the formation of aerosols during these possible accidents and to study their subsequent behaviour. It is also recommended that the fission products behaviour computer code FISSCON II should be re-examined to determine whether it reflects the advances incorporated in other codes developed for light water reactors which have been extensively compared. 47 refs

  19. Geology behind nuclear fission technology

    International Nuclear Information System (INIS)

    Dhana Raju, R.

    2005-01-01

    Geology appears to have played an important role of a precursor to Nuclear Fission Technology (NFT), in the latter's both birth from the nucleus of an atom of and most important application as nuclear power extracted from Uranium (U), present in its minerals. NFT critically depends upon the availability of its basic raw material, viz., nuclear fuel as U and/ or Th, extracted from U-Th minerals of specific rock types in the earth's crust. Research and Development of the Nuclear Fuel Cycle (NFC) depends heavily on 'Geology'. In this paper, a brief review of the major branches of geology and their contributions during different stages of NFC, in the Indian scenario, is presented so as to demonstrate the important role played by 'Geology' behind the development of NFT, in general, and NFC, in particular. (author)

  20. Singlet fission in pentacene dimers

    Science.gov (United States)

    Zirzlmeier, Johannes; Lehnherr, Dan; Coto, Pedro B.; Chernick, Erin T.; Casillas, Rubén; Basel, Bettina S.; Thoss, Michael; Tykwinski, Rik R.; Guldi, Dirk M.

    2015-01-01

    Singlet fission (SF) has the potential to supersede the traditional solar energy conversion scheme by means of boosting the photon-to-current conversion efficiencies beyond the 30% Shockley–Queisser limit. Here, we show unambiguous and compelling evidence for unprecedented intramolecular SF within regioisomeric pentacene dimers in room-temperature solutions, with observed triplet quantum yields reaching as high as 156 ± 5%. Whereas previous studies have shown that the collision of a photoexcited chromophore with a ground-state chromophore can give rise to SF, here we demonstrate that the proximity and sufficient coupling through bond or space in pentacene dimers is enough to induce intramolecular SF where two triplets are generated on one molecule. PMID:25858954

  1. Prompt fission neutron spectra of n + 235U above the (n, nf) fission threshold

    International Nuclear Information System (INIS)

    Shu Nengchuan; Chen Yongjing; Liu Tingjin; Jia Min

    2015-01-01

    Calculations of prompt fission neutron spectra (PFNS) from the 235 U(n, f) reaction were performed with a semi-empirical method for En = 7.0 and 14.7 MeV neutron energies. The total PFNS were obtained as a superposition of (n, xnf) pre-fission neutron spectra and post-fission spectra of neutrons which were evaporated from fission fragments, and these two kinds of spectra were taken as an expression of the evaporation spectrum. The contributions of (n, xnf) fission neutron spectra on the calculated PFNS were discussed. The results show that emission of one or two neutrons in the (n, nf) or (n, 2nf) reactions influences the PFNS shape, and the neutron spectra of the (n, xnf) fission-channel are soft compared with the neutron spectra of the (n, f) fission channel. In addition, analysis of the multiple-chance fission component showed that second-chance fission dominates the PFNS with an incident neutron energy of 14.7 MeV whereas first-chance fission dominates the 7 MeV case. (authors)

  2. Validation of multigroup neutron cross sections and calculational methods for the advanced neutron source against the FOEHN critical experiments measurements

    International Nuclear Information System (INIS)

    Smith, L.A.; Gallmeier, F.X.; Gehin, J.C.

    1995-05-01

    The FOEHN critical experiment was analyzed to validate the use of multigroup cross sections and Oak Ridge National Laboratory neutronics computer codes in the design of the Advanced Neutron Source. The ANSL-V 99-group master cross section library was used for all the calculations. Three different critical configurations were evaluated using the multigroup KENO Monte Carlo transport code, the multigroup DORT discrete ordinates transport code, and the multigroup diffusion theory code VENTURE. The simple configuration consists of only the fuel and control elements with the heavy water reflector. The intermediate configuration includes boron endplates at the upper and lower edges of the fuel element. The complex configuration includes both the boron endplates and components in the reflector. Cross sections were processed using modules from the AMPX system. Both 99-group and 20-group cross sections were created and used in two-dimensional models of the FOEHN experiment. KENO calculations were performed using both 99-group and 20-group cross sections. The DORT and VENTURE calculations were performed using 20-group cross sections. Because the simple and intermediate configurations are azimuthally symmetric, these configurations can be explicitly modeled in R-Z geometry. Since the reflector components cannot be modeled explicitly using the current versions of these codes, three reflector component homogenization schemes were developed and evaluated for the complex configuration. Power density distributions were calculated with KENO using 99-group cross sections and with DORT and VENTURE using 20-group cross sections. The average differences between the measured values and the values calculated with the different computer codes range from 2.45 to 5.74%. The maximum differences between the measured and calculated thermal flux values for the simple and intermediate configurations are ∼ 13%, while the average differences are < 8%

  3. AMZ, library of multigroup constants for EXPANDA computer codes, generated by NJOY computer code from ENDF/B-IV

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Moraes, M. de.

    1984-01-01

    A 70-group, 37-isotope library of multigroup constants for fast reactor nuclear design calculations is described. Nuclear cross sections, transfer matrices, and self-shielding factors were generated with NJOY code and an auxiliary program RGENDF using evaluated data from ENDF/B-IV. The output is being issued in a format suitable for EXPANDA code. Comparisons with JFS-2 library, as well as, test resuls for 14 CSEWG benchmark critical assemblies are presented. (Author) [pt

  4. ESELEM 4: a code for calculating fine neutron spectrum and multi-group cross sections in plate lattice

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Katsuragi, Satoru; Narita, Hideo.

    1976-07-01

    The multi-group treatment has been used in the design study of fast reactors and analysis of experiments at fast critical assemblies. The accuracy of the multi-group cross sections therefore affects strongly the results of these analyses. The ESELEM 4 code has been developed to produce multi-group cross sections with an advanced method from the nuclear data libraries used in the JAERI Fast set. ESELEM 4 solves integral transport equation by the collision probability method in plate lattice geometry to obtain the fine neutron spectrum. A typical fine group mesh width is 0.008 in lethargy unit. The multi-group cross sections are calculated by weighting the point data with the fine structure neutron flux. Some devices are applied to reduce computation time and computer core storage required for the calculation. The slowing down sources are calculated with the use of a recurrence formula derived for elastic and inelastic scattering. The broad group treatment is adopted above 2 MeV for dealing with both light any heavy elements. Also the resonance cross sections of heavy elements are represented in a broad group structure, for which we use the values of the JAERI Fast set. The library data are prepared by the PRESM code from ENDF/A type nuclear data files. The cross section data can be compactly stored in the fast computer core memory for saving the core storage and data processing time. The programme uses the variable dimensions to increase its flexibility. The users' guide for ESELEM 4 and PRESM is also presented in this report. (auth.)

  5. Application of the Laplace transform method for the albedo boundary conditions in multigroup neutron diffusion eigenvalue problems in slab geometry

    International Nuclear Information System (INIS)

    Petersen, Claudio Zen; Vilhena, Marco T.; Barros, Ricardo C.

    2009-01-01

    In this paper the application of the Laplace transform method is described in order to determine the energy-dependent albedo matrix that is used in the boundary conditions multigroup neutron diffusion eigenvalue problems in slab geometry for nuclear reactor global calculations. In slab geometry, the diffusion albedo substitutes without approximation the baffle-reflector system around the active domain. Numerical results to typical test problems are shown to illustrate the accuracy and the efficiency of the Chebysheff acceleration scheme. (orig.)

  6. CINESP - computational program of spatial kinetics for nuclear reactors in the one-two dimension multigroup diffusion theory

    International Nuclear Information System (INIS)

    Santos, R.S. dos

    1993-01-01

    This paper presents a computational program to solve numerically the reactor kinetics equations in the multigroup diffusion theory. One or two-dimensional problems in cylindrical or Cartesian geometries, with any number of energy and delayed-neutron precursors groups are dealt with. The main input and output of the program are briefly discussed. Various results demonstrate the accuracy and versatility of the program, when compared with other kinetics programs. (author)

  7. A multi-group and preemptable scheduling of cloud resource based on HTCondor

    Science.gov (United States)

    Jiang, Xiaowei; Zou, Jiaheng; Cheng, Yaodong; Shi, Jingyan

    2017-10-01

    and LHAASO. The result indicates that multi-group and preemptable resource scheduling is efficient to support multi-group and soft preemption. Additionally, the permission controlling component has been used in the local computing cluster, supporting for experiment JUNO, CMS and LHAASO, and the scale will be expanded to more experiments at the first half year, including DYW, BES and so on. Its evidence that the permission controlling is efficient.

  8. A new modelling of the multigroup scattering cross section in deterministic codes for neutron transport

    International Nuclear Information System (INIS)

    Calloo, A.A.

    2012-01-01

    In reactor physics, calculation schemes with deterministic codes are validated with respect to a reference Monte Carlo code. The remaining biases are attributed to the approximations and models induced by the multigroup theory (self-shielding models and expansion of the scattering law using Legendre polynomials) to represent physical phenomena (resonant absorption and scattering anisotropy respectively). This work focuses on the relevance of a polynomial expansion to model the scattering law. Since the outset of reactor physics, the latter has been expanded on a truncated Legendre polynomial basis. However, the transfer cross sections are highly anisotropic, with non-zero values for a very small range of the cosine of the scattering angle. Besides, the finer the energy mesh and the lighter the scattering nucleus, the more exacerbated is the peaked shape of this cross section. As such, the Legendre expansion is less suited to represent the scattering law. Furthermore, this model induces negative values which are non-physical. In this work, various scattering laws are briefly described and the limitations of the existing model are pointed out. Hence, piecewise-constant functions have been used to represent the multigroup scattering cross section. This representation requires a different model for the diffusion source. The discrete ordinates method which is widely employed to solve the transport equation has been adapted. Thus, the finite volume method for angular discretization has been developed and implemented in Paris environment which hosts the S n solver, Snatch. The angular finite volume method has been compared to the collocation method with Legendre moments to ensure its proper performance. Moreover, unlike the latter, this method is adapted for both the Legendre moments and the piecewise-constant functions representations of the scattering cross section. This hybrid-source method has been validated for different cases: fuel cell in infinite lattice

  9. Obtaining incremental multigroup cross sections for CANDU super cells with reactivity devices

    International Nuclear Information System (INIS)

    Balaceanu, V.; Constantin, M.

    2001-01-01

    In the last 20 years a multigroup methodology WIMS - PIJXYZ (WP) was developed and validated at INR Pitesti for obtaining incremental cross sections for reactivity devices in CANDU reactors. This is an alternate methodology to the CANDU classic methodology (experimentally adjusted) based on the POWDERPUFS and MULTICELL computer codes. The 2D supercell calculation performed with the WIMS code, that is a NEA Data Bank transport code, and which produces multigroup cross sections (on 18 energy groups) for CANDU supercell material (standard and perturbed, with and without reactivity devices). To obtain an as correct as possible 3D modelling for the CANDU supercells containing reactivity devices, the WIMS cross sections are used as input data for the PIJXYZ code, thus obtaining homogenized cross sections for CANDU supercells. PIJXYZ is an integral transport code based on the formalism of the first collision probabilities. It is analogue to the SHETAN code and it was created for neutron analyzes at cell level for CANDU type reactors were the reactivity devices are perpendicular to the fuel channels. The coordinate system used in PIJXYZ is a mixed one, namely a rectangular-cylindrical system. The geometric model used in PIJXYZ is presented. The fuel beam is represented by a horizontal cylinder and the reactivity device by a vertical one both cylinders being immersed in the moderator. Two supercell types were considered: a perturbed supercell (containing a reactivity device) and the standard supercell were the place of reactivity device is occupied by the moderator. The incremental cross sections for reactivity device are obtained as differences between the homogenized over supercell cross sections (with reactivity device) and homogenized over standards supercell (without device) cross sections. The PIJXYZ computation may be done on an energy cutting with 2 up to 18 groups. The validation of VIMS - PIJXYZ was done on the basis of several benchmark and by comparison with

  10. On the mechanism of fission neutron emission

    International Nuclear Information System (INIS)

    Maerten, H.; Richter, D.; Seeliger, D.

    1986-01-01

    This review represents the present knowledge of the mechanism of prompt fission neutron emission. Starting with a brief fission process characterization related with neutron emission, possible emission mechanisms are discussed. It is emphasized that the experimental study of special mechanisms, i.e. scission neutron emission processes, requires a sufficiently correct description of emission probabilities on the base of the main mechanism, i.e. the evaporation from fully accelerated fragments. Adequate statistical-model approaches have to account for the complexity of nuclear fission reflected by an intricate fragment distribution. The present picture of scission neutron emission is not clarified neither experimentally nor theoretically. Deduced data are contradictory and depend on the used analysis procedures often involving rough discriptions of evaporated-neutron distributions. The contribution of two secondary mechanisms of fission neutron emission, i.e. the neutron evaporation during fragment acceleration and neutron emission due to the decay of 5 He after ternary fission, is estimated. We summarize the recent progress of the theoretical description of fission neutron spectra in the framework of statistical models considering the standard spectrum of 252 Cf(sf) neutrons especially. The main experimental basis for the study of fission neutron emission is the accurate measurement of emission probabilities as a function of emission energy and angle (at least) as well as fragment parameters (mass number ratio and kinetic energy). The present status is evaluated. (author)

  11. Determination of the fission barrier height in fission of heavy radioactive beams induced by the (d,p)-transfer

    CERN Multimedia

    A theoretical framework is described, allowing to determine the fission barrier height using the observed cross sections of fission induced by the (d,p)-transfer with accuracy, which is not achievable in another type of low-energy fission of neutron-deficient nuclei, the $\\beta$-delayed fission. The primary goal is to directly determine the fission barrier height of proton-rich fissile nuclei, preferably using the radio-active beams of isotopes of odd elements, and thus confirm or exclude the low values of fission barrier heights, typically extracted using statistical calculations in the compound nucleus reactions at higher excitation energies. Calculated fission cross sections in transfer reactions of the radioactive beams show sufficient sensitivity to fission barrier height. In the probable case that fission rates will be high enough, mass asymmetry of fission fragments can be determined. Results will be relevant for nuclear astrophysics and for production of super-heavy nuclei. Transfer induced fission of...

  12. Thermodynamic analysis of volatile organometallic fission products

    International Nuclear Information System (INIS)

    Auxier II, J.D.; Hall, H.L.; Cressy, Derek

    2016-01-01

    The ability to perform rapid separations in a post nuclear weapon detonation scenario is an important aspect of national security. In the past, separations of fission products have been performed using solvent extraction, precipitation, etc. The focus of this work is to explore the feasibility of using thermochromatography, a technique largely employed in superheavy element chemistry, to expedite the separation of fission products from fuel components. A series of fission product complexes were synthesized and the thermodynamic parameters were measured using TGA/DSC methods. Once measured, these parameters were used to predict their retention times using thermochromatography. (author)

  13. Rearrangement of cluster structure during fission processes

    DEFF Research Database (Denmark)

    Lyalin, Andrey G.; Obolensky, Oleg I.; Solov'yov, Andrey V.

    2004-01-01

    Results of molecular dynamics simulations of fission reactions $Na_10^2+ -->Na_7^++ Na_3^+ and Na_18^2+--> 2Na_9^+ are presented. The dependence of the fission barriers on the isomer structure of the parent cluster is analysed. It is demonstrated that the energy necessary for removing homothetic...... groups of atoms from the parent cluster is largely independent of the isomer form of the parent cluster. The importance of rearrangement of the cluster structure during the fission process is elucidated. This rearrangement may include transition to another isomer state of the parent cluster before actual...

  14. Fusion-fission of heavy systems

    International Nuclear Information System (INIS)

    Rivet, M.F.; Alami, R.; Borderie, B.; Fuchs, H.; Gardes, D.; Gauvin, H.

    1988-01-01

    The influence of the entrance channel on fission processes was studied by forming the same composite system by two different target-projectile combinations ( 40 Ar + 209 Bi and 56 Fe + 187 Re, respectively). Compound nucleus fission and quasi fission were observed and the analysis was performed in the framework of the extra-extra-push model, which provides a qualitative interpretation of the results; limits for the extra-extra-push threshold are given, but problems with quantitative predictions for the extra-push are noted. (orig.)

  15. Fission cross section measurements for minor actinides

    Energy Technology Data Exchange (ETDEWEB)

    Fursov, B. [IPPE, Obninsk (Russian Federation)

    1997-03-01

    The main task of this work is the measurement of fast neutron induced fission cross section for minor actinides of {sup 238}Pu, {sup 242m}Am, {sup 243,244,245,246,247,248}Cm. The task of the work is to increase the accuracy of data in MeV energy region. Basic experimental method, fissile samples, fission detectors and electronics, track detectors, alpha counting, neutron generation, fission rate measurement, corrections to the data and error analysis are presented in this paper. (author)

  16. Fusion barrier distributions and fission anisotropies

    International Nuclear Information System (INIS)

    Hinde, D.J.; Morton, C.R.; Dasgupta, M.; Leigh, J.R.; Lestone, J.P.; Lemmon, R.C.; Mein, J.C.; Newton, J.O.; Timmers, H.; Rowley, N.; Kruppa, A.T.

    1995-01-01

    Fusion excitation functions for 16,17 O+ 144 Sm have been measured to high precision. The extracted fusion barrier distributions show a double-peaked structure interpreted in terms of coupling to inelastic collective excitations of the target. The effect of the positive Q-value neutron stripping channel is evident in the reaction with 17 O. Fission and evaporation residue cross-sections and excitation functions have been measured for the reaction of 16 O+ 208 Pb and the fusion barrier distribution and fission anisotropies determined. It is found that the moments of the fusion l-distribution determined from the fusion and fission measurements are in good agreement. ((orig.))

  17. Fission track geochronology of Eastern Ghats

    Energy Technology Data Exchange (ETDEWEB)

    Virk, H S; Koul, S L; Singh, S [Punjabi Univ., Patiala (India). Dept. of Physics

    1978-12-01

    Fission track geochronology of Eastern Ghats has been established by using fission track technique in dating micaceous and accessory minerals occurring in the region. Annealing studies confirm that radiation damage fossil tracks can be erased in minerals under intense metamorphic episodes thus resetting the geological clock. The fission track ages of the minerals range between 450 +- 5 m.y. to 622 +- 148 m.y. It is concluded that f.t. ages of the minerals date the last metamorphic event of the Eastern Ghats, known as the Indian Ocean Cycle.

  18. Application of Gauss quadratures to the calculation of resolved resonance contribution in multigroup cross sections

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1989-01-01

    A program (RESQ) based on quadratures that evaluates, from ENDF/B data, the resolved resonance contribution in group-averaged cross sections (capture, fission and scattering) was developed. Single and Multilevel Breit-Wigner parameters are accepted. Constant weighting function and zero Kelvin were considered. To assure convergence, different quadrature orders may be analysed. Results are compared with other codes' reconstruction and integration methods. (author) [pt

  19. Effects of fissioning nuclei distributions on fragment mass distributions for high energy fission

    Directory of Open Access Journals (Sweden)

    Rossi P C R

    2012-02-01

    Full Text Available We study the effects of fissioning nuclei mass- and energy-distributions on the formation of fragments for fission induced by high energy probes. A Monte Carlo code called CRISP was used for obtaining mass distributions and spectra of the fissioning nuclei for reactions induced by 660 MeV protons on 241Am and on 239Np, by 500 MeV protons on 208Pb, and by Bremsstrahlung photons with end-point energies at 50 MeV and 3500 MeV on 238U. The results show that even at high excitation energies, asymmetric fission may still contribute significantly to the fission cross section of actinide nuclei, while it is the dominante mode in the case of lead. However, more precise data for high energy fission on actinide are necessary in order to allow definite conclusions.

  20. Investigation of short-living fission products from the spontaneous fission of Cf-252

    International Nuclear Information System (INIS)

    Klonk, H.

    1976-01-01

    In this paper, a method of separating and measuring fission products of Cf-252 is presented. The measurement was achieved by means of γ-spectrometry and thus provides a quantitative analysis with a good separation of the fission products with respect to both atomic number Z and mass number A. The separation of the fission products from the fission source was achieved by means of solid traps. An automatic changing apparatus made it possible to keep irradiation and measuring times short, so even very short-lived fission products could be registered. The quantitative evaluation of primary fission products was made possible by correction according to Bateman equations. With that, the yields of single nuclides and the dispersion of charge can be determined. (orig./WL) [de

  1. Recovery and use of fission product noble metals

    International Nuclear Information System (INIS)

    Jensen, G.A.; Rohmann, C.A.; Perrigo, L.D.

    1980-06-01

    Noble metals in fission products are of strategic value. Market prices for noble metals are rising more rapidly than recovery costs. A promising concept has been developed for recovery of noble metals from fission product waste. Although the assessment was made only for the three noble metal fission products (Rh, Pd, Ru), there are other fission products and actinides which have potential value

  2. Towards a microscopic description of the fission process

    CERN Document Server

    Goutte, H; Berger, J F

    2010-01-01

    One major issue in nuclear physics is to develop a consistent model able to describe on the same footing the different aspects of the fission process, i.e. properties of the fissioning system, fission dynamics and fragment distributions. Microscopic fission studies based on the mean-field approximation are here presented.

  3. Development of multi-group xs libraries for the gfr 2400 reactor

    International Nuclear Information System (INIS)

    Cerba, Š.; Vrban, B.; Lüley, J.; Necas, V.

    2016-01-01

    GFR 2400 is considered as a conceptual design of the large scale GEN IV Gas-Cooled Fast Reactor. In general, the GEN IV technologies are seen as reliable but also very challenging reactor concepts. Since GFR 2400 lacks any experimental data, the questions on its safety are even more complex and the assessment of its performance could be made only based on computational experience. The paper deals with the development process of multi-group XS libraries based on a hybrid deterministic-Stochastic methodology, using the NJOY99, TRANSX, DIF3D, PARTISN and MCNP5 codes. A new optimized 25 group SBJ E 71 2 5G cross section library was developed based on ENDF/B-VII.1 evaluated data, ZZ-KAFAX-E70 background cross sections and GFR 2400 neutron spectrum. The created library was validated through integral experiments evaluated on the HEX-Z deterministic models in DIF3D. The results were also compared with MCNP5 calculations. (authors)

  4. A consistent multigroup model for radiative transfer and its underlying mean opacities

    International Nuclear Information System (INIS)

    Turpault, Rodolphe

    2005-01-01

    In some regimes, such as in plasma physics or in super orbital atmospheric entry of space objects, the effects of radiation are crucial and can tremendously modify the hydrodynamics of the gas. In such cases, it is therefore important to have a good prediction of the radiative variables. However, full transport solutions of these multi-dimensional, time-dependent problems are too expensive to get to be involved in a coupled configuration. It is hence necessary to develop other models for radiation that are cheap, yet accurate enough to give good predictions of the radiative effects. We will herein introduce the multigroup-M1 model and look at its characteristics and in particular try to separate the angular error from the frequential one since these two approximation play very different roles. The angular behaviour of the model will be tested on a case proposed by Su and Olson and used by Olson et al. to compare various moments and (flux-limited) diffusion models. For the frequency behaviour, we use a simplified flame test-case and show the importance of taking good mean opacities

  5. Offensive Strategy in the 2D Soccer Simulation League Using Multi-Group Ant Colony Optimization

    Directory of Open Access Journals (Sweden)

    Shengbing Chen

    2016-02-01

    Full Text Available The 2D soccer simulation league is one of the best test beds for the research of artificial intelligence (AI. It has achieved great successes in the domain of multi-agent cooperation and machine learning. However, the problem of integral offensive strategy has not been solved because of the dynamic and unpredictable nature of the environment. In this paper, we present a novel offensive strategy based on multi-group ant colony optimization (MACO-OS. The strategy uses the pheromone evaporation mechanism to count the preference value of each attack action in different environments, and saves the values of success rate and preference in an attack information tree in the background. The decision module of the attacker then selects the best attack action according to the preference value. The MACO-OS approach has been successfully implemented in our 2D soccer simulation team in RoboCup competitions. The experimental results have indicated that the agents developed with this strategy, along with related techniques, delivered outstanding performances.

  6. Stability analysis of multi-group deterministic and stochastic epidemic models with vaccination rate

    International Nuclear Information System (INIS)

    Wang Zhi-Gang; Gao Rui-Mei; Fan Xiao-Ming; Han Qi-Xing

    2014-01-01

    We discuss in this paper a deterministic multi-group MSIR epidemic model with a vaccination rate, the basic reproduction number ℛ 0 , a key parameter in epidemiology, is a threshold which determines the persistence or extinction of the disease. By using Lyapunov function techniques, we show if ℛ 0 is greater than 1 and the deterministic model obeys some conditions, then the disease will prevail, the infective persists and the endemic state is asymptotically stable in a feasible region. If ℛ 0 is less than or equal to 1, then the infective disappear so the disease dies out. In addition, stochastic noises around the endemic equilibrium will be added to the deterministic MSIR model in order that the deterministic model is extended to a system of stochastic ordinary differential equations. In the stochastic version, we carry out a detailed analysis on the asymptotic behavior of the stochastic model. In addition, regarding the value of ℛ 0 , when the stochastic system obeys some conditions and ℛ 0 is greater than 1, we deduce the stochastic system is stochastically asymptotically stable. Finally, the deterministic and stochastic model dynamics are illustrated through computer simulations. (general)

  7. Multi-Group Reductions of LTE Air Plasma Radiative Transfer in Cylindrical Geometries

    Science.gov (United States)

    Scoggins, James; Magin, Thierry Edouard Bertran; Wray, Alan; Mansour, Nagi N.

    2013-01-01

    Air plasma radiation in Local Thermodynamic Equilibrium (LTE) within cylindrical geometries is studied with an application towards modeling the radiative transfer inside arc-constrictors, a central component of constricted-arc arc jets. A detailed database of spectral absorption coefficients for LTE air is formulated using the NEQAIR code developed at NASA Ames Research Center. The database stores calculated absorption coefficients for 1,051,755 wavelengths between 0.04 µm and 200 µm over a wide temperature (500K to 15 000K) and pressure (0.1 atm to 10.0 atm) range. The multi-group method for spectral reduction is studied by generating a range of reductions including pure binning and banding reductions from the detailed absorption coefficient database. The accuracy of each reduction is compared to line-by-line calculations for cylindrical temperature profiles resembling typical profiles found in arc-constrictors. It is found that a reduction of only 1000 groups is sufficient to accurately model the LTE air radiation over a large temperature and pressure range. In addition to the reduction comparison, the cylindrical-slab formulation is compared with the finite-volume method for the numerical integration of the radiative flux inside cylinders with varying length. It is determined that cylindrical-slabs can be used to accurately model most arc-constrictors due to their high length to radius ratios.

  8. Study on the Control Strategy of Shifting Time Involving Multigroup Clutches

    Directory of Open Access Journals (Sweden)

    Zhen Zhu

    2016-01-01

    Full Text Available This paper focuses on the control strategy of shifting time involving multigroup clutches for a hydromechanical continuously variable transmission (HMCVT. The dynamic analyses of mathematical models are presented in this paper, and the simulation models are used to study the control strategy of HMCVT. Simulations are performed in Simulation X platform to investigate the shifting time of clutches under different operating conditions. On this basis, simulation analysis and test verification of two typical conditions, which play the decisive roles for the shifting quality, are carried out. The results show that there are differences in the shifting time of the two typical conditions. In the shifting process from the negative transmission of hydromechanical ranges to the positive transmission of hydromechanical ranges, the control strategy based on the shifting time is switching the clutches of shifting mechanism firstly and then disengaging a group of clutches of planetary gear mechanism and engaging another group of the clutches of planetary gear mechanism lastly. In the shifting process from the hydraulic range to the hydromechanical range, the control strategy based on the shifting time is switching the clutches of hydraulic shifting mechanism and planetary gear mechanism at first and then engaging the clutch of shifting mechanism.

  9. JSD1000: multi-group cross section sets for shielding materials

    International Nuclear Information System (INIS)

    Yamano, Naoki

    1984-03-01

    A multi-group cross section library for shielding safety analysis has been produced by using ENDF/B-IV. The library consists of ultra-fine group cross sections, fine-group cross sections, secondary gamma-ray production cross sections and effective macroscopic cross sections for typical shielding materials. Temperature dependent data at 300, 560 and 900 K have been also provided. Angular distributions of the group to group transfer cross section are defined by a new method of ''Direct Angular Representation'' (DAR) instead of the method of finite Legendre expansion. The library designated JSD1000 are stored in a direct access data base named DATA-POOL and data manipulations are available by using the DATA-POOL access package. The 3824 neutron group data of the ultra-fine group cross sections and the 100 neutron, 20 photon group cross sections are applicable to shielding safety analyses of nuclear facilities. This report provides detailed specifications and the access method for the JSD1000 library. (author)

  10. Three-dimensional h-adaptivity for the multigroup neutron diffusion equations

    KAUST Repository

    Wang, Yaqi

    2009-04-01

    Adaptive mesh refinement (AMR) has been shown to allow solving partial differential equations to significantly higher accuracy at reduced numerical cost. This paper presents a state-of-the-art AMR algorithm applied to the multigroup neutron diffusion equation for reactor applications. In order to follow the physics closely, energy group-dependent meshes are employed. We present a novel algorithm for assembling the terms coupling shape functions from different meshes and show how it can be made efficient by deriving all meshes from a common coarse mesh by hierarchic refinement. Our methods are formulated using conforming finite elements of any order, for any number of energy groups. The spatial error distribution is assessed with a generalization of an error estimator originally derived for the Poisson equation. Our implementation of this algorithm is based on the widely used Open Source adaptive finite element library deal.II and is made available as part of this library\\'s extensively documented tutorial. We illustrate our methods with results for 2-D and 3-D reactor simulations using 2 and 7 energy groups, and using conforming finite elements of polynomial degree up to 6. © 2008 Elsevier Ltd. All rights reserved.

  11. MENDF71x. Multigroup Neutron Cross Section Data Tables Based upon ENDF/B-VII.1

    International Nuclear Information System (INIS)

    Conlin, Jeremy Lloyd; Parsons, Donald Kent; Gardiner, Steven J.; Gray, Mark Girard; Lee, Mary Beth; White, Morgan Curtis

    2015-01-01

    A new multi-group neutron cross section library has been released along with the release of NDI version 2.0.20. The library is named MENDF71x and is based upon the evaluations released in ENDF/B-VII.1 which was made publicly available in December 2011. ENDF/B-VII.1 consists of 423 evaluations of which ten are excited states evaluations and 413 are ground state evaluations. MENDF71x was created by processing the 423 evaluations into 618-group, downscatter only NDI data tables. The ENDF/B evaluation files were processed using NJOY version 99.393 with the exception of 35 Cl and 233 U. Those two isotopes had unique properties that required that we process the evaluation using NJOY version 2012. The MENDF71x library was only processed to room temperature, i.e., 293.6 K. In the future, we plan on producing a multi-temperature library based on ENDF/B-VII.1 and compatible with MENDF71x.

  12. General solution of the multigroup spherical harmonics equations in R-Z geometry

    International Nuclear Information System (INIS)

    Matausek, M.

    1983-01-01

    In the present paper the generalization is performed of the procedure to solve multigroup spherical harmonics equations, which has originally been proposed and developed foe one-dimensional systems in cylindrical or spherical geometry, and later extended for special case of a two-dimensional system in r-z geometry. The expressions are derived for the axial and the radial dependence of the group values of the neutron flux moments, in the P-3 approximation of the spherical harmonics method, in a cylindrically symmetrical system with an arbitrary number of material regions in both r and z directions. In the special case of an axially homogeneous system, these expressions reduce to the relations derived previously. The analysis is performed of the possibilities to satisfy the boundary conditions in the case when the system considered represents an elementary reactor lattice cell and in the case when the system represents a reactor as a whole. The computational effort is estimated for system of a given configuration. (author)

  13. Young Adults’ Attitude Towards Advertising: a multi-group analysis by ethnicity

    Directory of Open Access Journals (Sweden)

    Hiram Ting

    2015-08-01

    Full Text Available Objective – This study aims to investigate the attitude of Malaysian young adults towards advertising. How this segment responds to advertising, and how ethnic/cultural differences moderate are assessed. Design/methodology/approach – A quantitative questionnaire is used to collect data at two universities. Purposive sampling technique is adopted to ensure the sample represents the actual population. Structural equation modelling (SEM and multi-group analysis (MGA are utilized in analysis. Findings - The findings show that product information, hedonism, and good for economy are significant predictors of attitude towards advertising among young adults. Additionally, falsity is found to be significant among the Chinese, while social role and materialism among the Dayaks. No difference is observed in the effect of attitude on intention towards advertising by ethnicity. While homogeneity in advertising beliefs is assumed across ethnic groups, the Chinese and Dayak young adults are different in some of their advertising beliefs. Practical implications – Despite cultural effect being well-documented, young adults today seem to have similar beliefs and attitude towards advertising. Knowing what is shared and what is not for this segment is essential. Hence, it is imperative to keep track of their values in diversified communities to ensure effective communication process in advertising. Originality/value – In addition to the theory of reasoned action, MGA is utilized to assess the moderating effect of ethnic/culture on the whole model. This affords a more comprehensive understanding on the subject matter in multi-ethnic and cultural countries.

  14. Two-dimensional semi-analytic nodal method for multigroup pin power reconstruction

    International Nuclear Information System (INIS)

    Seung Gyou, Baek; Han Gyu, Joo; Un Chul, Lee

    2007-01-01

    A pin power reconstruction method applicable to multigroup problems involving square fuel assemblies is presented. The method is based on a two-dimensional semi-analytic nodal solution which consists of eight exponential terms and 13 polynomial terms. The 13 polynomial terms represent the particular solution obtained under the condition of a 2-dimensional 13 term source expansion. In order to achieve better approximation of the source distribution, the least square fitting method is employed. The 8 exponential terms represent a part of the analytically obtained homogeneous solution and the 8 coefficients are determined by imposing constraints on the 4 surface average currents and 4 corner point fluxes. The surface average currents determined from a transverse-integrated nodal solution are used directly whereas the corner point fluxes are determined during the course of the reconstruction by employing an iterative scheme that would realize the corner point balance condition. The outgoing current based corner point flux determination scheme is newly introduced. The accuracy of the proposed method is demonstrated with the L336C5 benchmark problem. (authors)

  15. TRIDENT: a two-dimensional, multigroup, triangular mesh discrete ordinates, explicit neutron transport code

    International Nuclear Information System (INIS)

    Seed, T.J.; Miller, W.F. Jr.; Brinkley, F.W. Jr.

    1977-03-01

    TRIDENT solves the two-dimensional-multigroup-transport equations in rectangular (x-y) and cylindrical (r-z) geometries using a regular triangular mesh. Regular and adjoint, inhomogeneous and homogeneous (k/sub eff/ and eigenvalue searches) problems subject to vacuum, reflective, white, or source boundary conditions are solved. General anisotropic scattering is allowed and anisotropic-distributed sources are permitted. The discrete-ordinates approximation is used for the neutron directional variables. An option is included to append a fictitious source to the discrete-ordinates equations that is defined such that spherical-harmonics solutions (in x-y geometry) or spherical-harmonics-like solutions (in r-z geometry) are obtained. A spatial-finite-element method is used in which the angular flux is expressed as a linear polynomial in each triangle that is discontinous at triangle boundaries. Unusual Features of the program: Provision is made for creation of standard interface output files for S/sub N/ constants, angle-integrated (scalar) fluxes, and angular fluxes. Standard interface input files for S/sub N/ constants, inhomogeneous sources, cross sections, and the scalar flux may be read. Flexible edit options as well as a dump and restart capability are provided

  16. Use of CPXSD for generation of effective fast multigroup libraries for pressure vessel fluence calculations

    International Nuclear Information System (INIS)

    Alpan, F. Arzu; Haghighat, Alireza

    2008-01-01

    Multigroup (i.e., broad-group) libraries play a significant role in the accuracy of transport calculations. There are several broad-group libraries available for particular applications. For example the 47-neutron (26 fast groups), 20-gamma-group BUGLE libraries are commonly used for light water reactor shielding and pressure vessel dosimetry problems. However, there is no publicly available methodology to construct group structures for a problem and objective of interest. Therefore, we have developed the Contribution and Point-wise Cross-Section Driven (CPXSD) methodology, which constructs effective fine-and broad-group structures. In this paper, we use the CPXSD methodology to construct broad-group structures for fast neutron dosimetry problems. It is demonstrated that the broad-group libraries generated from CPXSD constructed group structures, while only 14 groups (rather than 26 groups) in the fast energy range are in good agreement (similar to 1 %-2 %) with the fine-group library from which they were derived, in reaction rate calculations.

  17. Review of uncertainty files and improved multigroup cross section files for FENDL

    International Nuclear Information System (INIS)

    Ganesan, S.

    1994-03-01

    The IAEA Nuclear Data Section, in co-operation with several national nuclear data centers and research groups, is creating an internationally available Fusion Evaluated Nuclear Data Library (FENDL), which will serve as a comprehensive source of processed and tested nuclear data tailored to the requirements of the Engineering and Development Activities (EDA) of the International Thermonuclear Experimental Reactor (ITER) Project and other fusion-related development projects. The FENDL project of the International Atomic Energy Agency has the task of coordination with the goal of assembling, processing and testing a comprehensive, fusion-relevant Fusion Evaluated Nuclear Data Library with unrestricted international distribution. The present report contains the summary of the IAEA Advisory Group Meeting on ''Review of Uncertainty Files and Improved Multigroup Cross Section Files for FENDL'', held during 8-12 November 1993 at the Tokai Research Establishment, JAERI, Japan, organized in cooperation with the Japan Atomic Energy Research Institute. The report presents the current status of the FENDL activity and the future work plans in the form of conclusions and recommendations of the four Working Groups of the Advisory Group Meeting on (1) experimental and calculational benchmarks, (2) preparation processed libraries for FENDL/ITER, (3) specifying procedures for improving FENDL and (4) selection of activation libraries for FENDL. (author). 1 tab

  18. TASK, 1-D Multigroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron

    International Nuclear Information System (INIS)

    Buhl, A.R.; Hermann, O.W.; Hinton, R.J.; Dodds, H.L. Jr.; Robinson, J.C.; Lillie, R.A.

    1975-01-01

    1 - Description of problem or function: TASK solves the one-dimensional multigroup form of the reactor kinetics equations, using either transport or diffusion theory and allowing an arbitrary number of delayed neutron groups. The program can also be used to solve standard static problems efficiently such as eigenvalue problems, distributed source problems, and boundary source problems. Convergence problems associated with sources in highly multiplicative media are circumvented, and such problems are readily calculable. 2 - Method of solution: TASK employs a combination scattering and transfer matrix method to eliminate certain difficulties that arise in classical finite difference approximations. As such, within-group (inner) iterations are eliminated and solution convergence is independent of spatial mesh size. The time variable is removed by Laplace transformation. (A later version will permit direct time solutions.) The code can be run either in an outer iteration mode or in closed (non-iterative) form. The running mode is dictated by the number of groups times the number of angles, consistent with available storage. 3 - Restrictions on the complexity of the problem: The principal restrictions are available storage and computation time. Since the code is flexibly-dimensioned and has an outer iteration option there are no internal restrictions on group structure, quadrature, and number of ordinates. The flexible-dimensioning scheme allows optional use of core storage. The generalized cylindrical geometry option is not complete in Version I of the code. The feedback options and omega-mode search options are not included in Version I

  19. Determination of the fission products yields, lanthanide and yttrium, in the fission of 238U with neutrons of fission spectra

    International Nuclear Information System (INIS)

    Nicoli, I.G.

    1981-06-01

    A radiochemical investigation is performed to measure the cumulative fission product yields of several lantanides and yttrium nuclides in the 238 U by fission neutron spectra. Natural and depleted uranium are irradiated under the same experimental conditions in order to find a way to subtract the contribution of the 235 U fission. 235 U percentage in the natural uranium was 3.5 times higher than in the depleted uranium. Uranium oxides samples are irradiated inside the core of the Argonaut Reactor, at the Instituto de Engenharia Nuclear, and the lantanides and yttrium are chemically separated. The fission products gamma activities were detected, counted and analysed in a system constituted by a high resolution Ge(Li) detector, 4096 multichannel analyser and a PDP-11 computer. Cumulative yields for fission products with half-lives between 1 to 33 hours are measured: 93 Y, 141 La, 142 La, 143 Ce and 149 Nd. The chain total yields are calculated. The cumulative fission yields measured for 93 Y, 141 La, 142 La, 143 Ce and 149 Nd are 4,49%, 4,54%, 4,95%, 4,16% and 1,37% respectively and they are in good agreement with the values found in the literature. (Author) [pt

  20. Neutron emission as a probe of fusion-fission and quasi-fission dynamics

    International Nuclear Information System (INIS)

    Hinde, D.J.

    1991-01-01

    Pre- and post scission neutron yeilds have been measured as a function of projectile mass, compound nucleus fissility, and fission mass-split and total kinetic energy (TKE) for 27 fusion-fission and quasi-fission reactions induced by beams of 16,18 O, 40 Ar and 64 Ni. A new method of interpretation of experimental pre-scission neutron multiplicities ν-pre and mean kinetic energies ε ν allows the extraction of fission time scales with much less uncertainty than previously, all fusion-fission results being consistent with a dynamical time scale of (35±15) x 10 -21 s for symmetric fission. All reactions show that ν-pre falls quite rapidly with increasing mass-asymmetry; evidence is presented that for fusion-fission reactions this is partly due to a reduction of the dynamical fission time scale with mass-asymmetry. For quasi-fission, the data indicate that the pre-scission multiplicity and mean neutron kinetic energy are very sensitive to the final mass-asymmetry, but that the time scale is virtually independent of mass-asymmetry. It is concluded that for fusion-fission there is no dependence of ν-pre on TKE, whilst for 64 Ni-induced quasi-fission reactions, a strong increase of ν-pre with decreasing TKE is observed, probably largely caused by neutron emission during the acceleration time of the fission fragments in these fast reactions. Interpretation of post-scission multiplicities in terms of fragment excitation energies leads to deduced time scales consistent with those determined from the pre-scission data. 54 refs., 17 tabs., 25 figs

  1. (d,p)-transfer induced fission of heavy radioactive beams

    CERN Document Server

    Veselsky, Martin

    2012-01-01

    (d,p)-transfer induced fission is proposed as a tool to study low energy fission of exotic heavy nuclei. Primary goal is to directly determine the fission barrier height of proton-rich fissile nuclei, preferably using the radio-active beams of isotopes of odd elements, and thus confirm or exclude the low values of fission barrier heights, typically extracted using statistical calculations in the compound nucleus reactions at higher excitation energies. Calculated fission cross sections in transfer reactions of the radioactive beams show sufficient sensitivity to fission barrier height. In the probable case that fission rates will be high enough, mass asymmetry of fission fragments can be determined. Results will be relevant for nuclear astrophysics and for production of super-heavy nuclei. Transfer induced fission offers a possibility for systematic study the low energy fission of heavy exotic nuclei at the ISOLDE.

  2. A fission gas release model

    Energy Technology Data Exchange (ETDEWEB)

    Denis, A; Piotrkowski, R [Argentine Atomic Energy Commission, Buenos Aires (Argentina)

    1997-08-01

    The hypothesis contained in the model developed in this work are as follows. The UO{sub 2} is considered as a collection of spherical grains. Nuclear reactions produce fission gases, mainly Xe and Kr, within the grains. Due to the very low solubility of these gases in UO{sub 2}, intragranular bubbles are formed, of a few nanometers is size. The bubbles are assumed to be immobile and to act as traps which capture gas atoms. Free atoms diffuse towards the grain boundaries, where they give origin to intergranular, lenticular bubbles, of the order of microns. The gas atoms in bubbles, either inter or intragranular, can re-enter the matrix through the mechanism of resolution induced by fission fragment impact. The amount of gas stored in intergranular bubbles grows up to a saturation value. Once saturation is reached, intergranular bubbles inter-connect and the gas in excess is released through different channels to the external surface of the fuel. The resolution of intergranular bubbles particularly affects the region of the grain adjacent to the grain boundary. During grain growth, the grain boundary traps the gas atoms, either free or in intragranular bubbles, contained in the swept volume. The grain boundary is considered as a perfect sink, i.e. the gas concentration is zero at that surface of the grain. Due to the spherical symmetry of the problem, the concentration gradient is null at the centre of the grain. The diffusion equation was solved using the implicit finite difference method. The initial solution was analytically obtained by the Laplace transform. The calculations were performed at different constant temperatures and were compared with experimental results. They show the asymptotic growth of the grain radius as a function of burnup, the gas distribution within the grain at every instant, the growth of the gas content at the grain boundary up to the saturation value and the fraction of gas released by the fuel element referred to the total gas generated

  3. Independent fission yields of Rb and Cs from thermal-neutron-induced fission of 239Pu

    International Nuclear Information System (INIS)

    Balestrini, S.J.; Forman, L.

    1975-01-01

    The relative independent fission yields of Rb and Cs from thermal-neutron-induced fission of 239 Pu have been measured on line using a mass spectrograph and thermalized neutrons from a burst reactor. Independent yields were derived by normalizing the measurements to products of chain yields and fractional independent yields, estimating the latter from measured cumulative yields of Kr and Xe. Comparing the independent yields with those from 238 U fission, the 239 Pu results show shifts in isotopic yield distribution toward lower mass for both Rb and Cs and also toward the production of more Cs and less Rb when 239 Pu is fissioned

  4. Angular momenta of fission fragments in the {alpha}-accompanied fission of {sup 252}Cf

    Energy Technology Data Exchange (ETDEWEB)

    Jandel, M.; Kliman, J.; Krupa, L.; Morhac, M. [Slovak Academy of Sciences, Department of Nuclear Physics, Bratislava (Slovakia); Joint Institute for Nuclear Research, Flerov Laboratory for Nuclear Reactions, Dubna (Russian Federation); Hamilton, J.H.; Kormicki, J.; Ramayya, A.V.; Hwang, J.K.; Luo, Y.X.; Fong, D.; Gore, P. [Vanderbilt University, Department of Physics, Nashville, TN (United States); Ter-Akopian, G.M.; Oganessian, Yu.Ts.; Rodin, A.M.; Fomichev, A.S.; Popeko, G.S. [Joint Institute for Nuclear Research, Flerov Laboratory for Nuclear Reactions, Dubna (Russian Federation); Daniel, A.V. [Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Rasmussen, J.O.; Macchiavelli, A.O.; Stoyer, M.A. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Donangelo, R.; Cole, J.D.

    2005-06-01

    For the first time, average angular momenta of the ternary fission fragments {sup 100,102}Zr, {sup 106}Mo, {sup 144,146}Ba and {sup 138,140,142}Xe from the {alpha}-accompanied fission of {sup 252}Cf were obtained from relative intensities of prompt {gamma}-ray transitions with the use of the statistical model calculation. Average values of the angular momenta were compared with the corresponding values for the same fission fragments from the binary fission of {sup 252}Cf. Results indicate the presence of a decreasing trend in the average values of angular momenta induced in ternary fission fragments compared to the same binary fission fragments. On the average, the total angular momentum extracted for ternary fission fragments is {proportional_to}1.4{Dirac_h} lower than in binary fission. Consequently, results indicate that the mechanism of the ternary {alpha}-particles emission may directly effect an induction of angular momenta of fission fragments, and possible scenarios of such mechanisms are discussed. Further, the dependence of the angular momenta of {sup 106}Mo and {sup 140}Xe on the number of emitted neutrons from correlated pairs of primary fragments was obtained also showing a decreasing dependence of average angular momenta with increasing number of emitted neutrons. Consequences are briefly discussed. (orig.)

  5. Measurement of prompt fission gamma-ray spectra in fast neutron-induced fission

    International Nuclear Information System (INIS)

    Laborie, J.M.; Belier, G.; Taieb, J.

    2012-01-01

    Knowledge of prompt fission gamma-ray emission has been of major interest in reactor physics for a few years. Since very few experimental spectra were ever published until now, new measurements would be also valuable to improve our understanding of the fission process. An experimental method is currently being developed to measure the prompt fission gamma-ray spectrum from some tens keV up to 10 MeV at least. The mean multiplicity and total energy could be deduced. In this method, the gamma-rays are measured with a bismuth germanate (BGO) detector which has the advantage to present a high P/T ratio and a high efficiency compared to other gamma-ray detectors. The prompt fission neutrons are rejected by the time of flight technique between the BGO detector and a fission trigger given by a fission chamber or a scintillating active target. Energy and efficiency calibration of the BGO detector were carried out up to 10.76 MeV by means of the Al-27(p, gamma) reaction. First prompt fission gamma-ray spectrum measurements performed for the spontaneous fission of Cf-252 and for 1.7 and 15.6 MeV neutron-induced fission of U-238 at the CEA, DAM, DIF Van de Graaff accelerator, will be presented. (authors)

  6. Decay and fission of the oriented nuclei

    CERN Document Server

    Kadmenskij, S G

    2002-01-01

    The fragment angular distributions for binary decay of oriented spherical and deformed nuclei with taking into account the correct transformational properties of wave functions under time inversion have been investigated. It has been shown that for description of fragment angular distributions the adiabatic approximation for collective rotational nuclear degrees of freedom is not correct. It has been demonstrated that this approximation is valid for description of spontaneous and induced low-energy nuclear fission. The dependence of partial fission widths on the orientation of the internal axes spins, projections of spins, and relative angular moments of fission fragments has been analyzed. It has been shown that the adiabatic approximation results in coherent interference of wave functions of fragments relative movement. This interference forms fragments the universal angular distributions of fission fragments for oriented nuclei. For these distributions the deviations from A. Bohr's formula have been invest...

  7. Aqueous cutting fluid for machining fissionable materials

    Science.gov (United States)

    Duerksen, Walter K.; Googin, John M.; Napier, Jr., Bradley

    1984-01-01

    The present invention is directed to a cutting fluid for machining fissionable material. The cutting fluid is formed of glycol, water and boron compound in an adequate concentration for effective neutron attenuation so as to inhibit criticality incidents during machining.

  8. "UCx fission targets oxidation test stand"

    CERN Document Server

    Lacroix, Rachel

    2014-01-01

    "Set up a rig dedicated to the oxidation of UCx and define a procedure for repeatable, reliable and safe method for converting UC2 fission targets into an acceptable uranium carbide oxide waste for subsequent disposal by the Swiss Authorities."

  9. Fission-product release during accidents

    International Nuclear Information System (INIS)

    Hunt, C.E.L.; Cox, D.S.

    1991-09-01

    One of the aims when managing a reactor accident is to minimize the release of radioactive fission products. Release is dependent not only on the temperature, but also on the partial pressure of oxygen. Strongly oxidizing atmospheres, such as those that occurred during the Chernobyl accident, released semi-volatile elements like ruthenium, which has volatile oxides. At low temperatures, UO 2 oxidization to U 3 O 8 can result in extensive breakup of the fuel, resulting in the release of non-volatile fission products as aerosols. Under less oxidizing conditions, when hydrogen accumulates from the zirconium-water reaction, the resulting low oxygen partial pressure can significantly reduce these reactions. At TMI-2, only the noble gases and volatile fission products were released in significant quantities. A knowledge of the effect of atmosphere as well as temperature on the release of fission products from damaged reactor cores is therefore a useful, if not necessary, component of information required for accident management

  10. Vitrification processes for fission product solutions

    International Nuclear Information System (INIS)

    Bonniaud, R.; Jouan, A.; Moncouyoux, J.P.; Sombret, C.

    1982-10-01

    The different processes for fission product vitrification in the world are reviewed. Continuous or discontinuous processes, induction or arc heating, in can melting or casting, tests with radioactive or simulated wastes and industrial realizations are described [fr

  11. Separation of short-lived fission products

    International Nuclear Information System (INIS)

    Tamai, Tadaharu; Ohyoshi, Emiko; Ohyoshi, Akira; Kiso, Yoshiyuki; Shinagawa, Mutsuaki.

    1976-01-01

    A rbief review is presented on the various methods of separation available for both gaseous and liquid states, for the separation of short-lived fission products formed by binary fission of neutron irradiated uranium. The means available for gaseous state are the hot atom reaction, the hydride method and on-line mass separation. For liquid state, use can be made of precipitation, ionic or atomic exchange, solvent extraction and paper electrophoresis. Particular reference is made to electrophoretic separation of ions produced by fission in aqueous solution of uranium. The principle of electrophoretic separation and the procedures for separating the element of interest from the other fission products are outlined, with reference made to the results obtained with the method by the present authors. The elements in question are alkalines, alkaline earths, rare earths, halogens, selenium and

  12. Fission cross section measurements at intermediate energies

    International Nuclear Information System (INIS)

    Laptev, Alexander

    2005-01-01

    The activity in intermediate energy particle induced fission cross-section measurements of Pu, U isotopes, minor actinides and sub-actinides in PNPI of Russia is reviewed. The neutron-induced fission cross-section measurements are under way in the wide energy range of incident neutrons from 0.5 MeV to 200 MeV at the GNEIS facility. In number of experiments at the GNEIS facility, the neutron-induced fission cross sections were obtained for many nuclei. In another group of experiments the proton-induced fission cross-section have been measured for proton energies ranging from 200 to 1000 MeV at 100 MeV intervals using the proton beam of PNPI synchrocyclotron. (author)

  13. Fission track method for uranium ore exploration

    International Nuclear Information System (INIS)

    Guo Shilun; Deng Xinlu; Sun Shengfen; Meng Wu; Zhang Pengfa; Hao Xiuhong

    1986-01-01

    The uranium concentrations in natural water collected in the fields of uranium ore exploration with fission track method have been determined. It shows that the results of fission track method are consistent with that of fluoro-colorimetry and laser fluorometry for the same samples of water with uranium concentration in the region of 10 -4 to 10 -8 g/l. For water samples with lower uranium concentration (≤10 -8 g/l), the fission track method can still give accurate or referential results, but the other two methods failed. The reproducibility of fission track method was checked and discussed by using samples collected in the same fields of uranium ore exploration. The effects of the concentration of the impurities in natural water on determination of uranium concentration were analysed and discussed as well

  14. Discovery of nuclear fission in Berlin 1938

    International Nuclear Information System (INIS)

    Hilscher, D.

    1989-01-01

    The story of the discovery of nuclear fission, one of the most exciting stories of how a scientific puzzle was finally solved and how the scientists involved were blind to many obvious indications, is described. (author). 29 refs

  15. Bimodality in macroscopic dynamics of nuclear fission

    International Nuclear Information System (INIS)

    Bastrukov, S.I.; Salamatin, V.S.; Strteltsova, O.I.; Molodtsova, I.V.; Podgainy, D.V.; )

    2000-01-01

    The elastodynamic collective model of nuclear fission is outlined whose underlying idea is that the stiff structure of nuclear shells imparts to nucleus properties typical of a small piece of an elastic solid. Emphasis is placed on the macroscopic dynamics of nuclear deformations resulting in fission by two energetically different modes. The low-energy S-mode is the fission due to disruption of elongated quadrupole spheroidal shape. The characteristic features of the high-energy T-mode of division by means of torsional shear deformations is the compact scission configuration. Analytic and numerical estimates for the macroscopic fission-barrier heights are presented, followed by discussion of fingerprints of the above dynamical bimodality in the available data [ru

  16. A new neutron counter for fission research

    Energy Technology Data Exchange (ETDEWEB)

    Laurent, B., E-mail: benoit.laurent@cea.fr [CEA, DAM, DIF, F-91297 Arpajon (France); Granier, T.; Bélier, G.; Chatillon, A.; Martin, J.-F.; Taieb, J. [CEA, DAM, DIF, F-91297 Arpajon (France); Hambsch, F.-J. [EC-JRC Institute for Reference Materials and Measurements (IRMM), Retieseweg, 2440 Geel (Belgium); Tovesson, F.; Laptev, A.B.; Haight, R.C.; Nelson, R.O.; O' Donnell, J.M. [Los Alamos Neutron Science Center, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2014-05-01

    A new neutron counter for research experiments on nuclear fission has been developed. This instrument is designed for the detection of prompt fission neutrons within relatively high levels of gamma and neutron background. It is composed of a set of {sup 3}He proportional counters arranged within a block of polyethylene which serves as moderator. The detection properties have been studied by means of Monte Carlo simulations and experiments with radioactive sources. These properties are confirmed by an experiment on neutron-induced fission of {sup 238}U at the WNR facility of the Los Alamos Neutron Science Center during which the mean prompt fission neutron multiplicity, or ν{sup ¯} has been measured from 1 to 20 MeV of incident neutron energy.

  17. Uranium deposits obtention for fission chambers

    International Nuclear Information System (INIS)

    Artacho Saviron, E.

    1972-01-01

    The obtention of uranium deposits of the required quality for small cylindrical fission chambers presents some difficulties. With the method of electroplating here described the uniformity, reproducibility and adherence of the obtained deposits were satisfactory. (Author) 6 refs

  18. Natural fission reactors - the Oklo phenomenon

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Overview describes the discovery of the site, location of the reactors and site geology and discusses the permanence of fission products, nuclear reaction control mechanisms and trace concentrations of elements that act as poisons. (Author)

  19. Development of fission Mo-99 production technology

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Ho; Choung, W. M.; Lee, K. I. and others

    2000-05-01

    Fission Mo-99 is the only parent nuclide of Tc-99m, an extremely useful tool for mdeical diagnosis, with an estimated usage of greater than 80% of nuclear medicine applicatons. HEU and LEU targets to optimize in HANARO irradiation condition suggested and designed for domestic production of fission Mo-99. The optimum process conditions are established in each unit process to meet quality requirements of fission Mo-99 products, and the results of performance test in combined process show Mo separation and purification yield of the above 97%. The concept of Tc generator production process is established, and the result of performance test show Tc production yield of 98.4% in Tc generator procuction process. The drafts is prepared for cooperation of technical cooperation and business investment with foreign country. Evaluation on economic feasibility is accompanied for fission Mo-99 and Tc-99m generator production.

  20. Development of fission Mo-99 production technology

    International Nuclear Information System (INIS)

    Park, Jin Ho; Choung, W. M.; Lee, K. I. and others

    2000-05-01

    Fission Mo-99 is the only parent nuclide of Tc-99m, an extremely useful tool for mdeical diagnosis, with an estimated usage of greater than 80% of nuclear medicine applicatons. HEU and LEU targets to optimize in HANARO irradiation condition suggested and designed for domestic production of fission Mo-99. The optimum process conditions are established in each unit process to meet quality requirements of fission Mo-99 products, and the results of performance test in combined process show Mo separation and purification yield of the above 97%. The concept of Tc generator production process is established, and the result of performance test show Tc production yield of 98.4% in Tc generator procuction process. The drafts is prepared for cooperation of technical cooperation and business investment with foreign country. Evaluation on economic feasibility is accompanied for fission Mo-99 and Tc-99m generator production

  1. Dynamical chaos and induced nuclear fission

    Energy Technology Data Exchange (ETDEWEB)

    Bolotin, Yu L; Krivoshej, I V

    1985-01-01

    It is shown that the exponential instability of trajectories, which arises at negative curvature of the potential energy surface, leads to diffusion of the image point through the barrier and determines real time delays in induced nuclear fission.

  2. Energy from nuclear fission an introduction

    CERN Document Server

    De Sanctis, Enzo; Ripani, Marco

    2016-01-01

    This book provides an overview on nuclear physics and energy production from nuclear fission. It serves as a readable and reliable source of information for anyone who wants to have a well-balanced opinion about exploitation of nuclear fission in power plants. The text is divided into two parts; the first covers the basics of nuclear forces and properties of nuclei, nuclear collisions, nuclear stability, radioactivity, and provides a detailed discussion of nuclear fission and relevant topics in its application to energy production. The second part covers the basic technical aspects of nuclear fission reactors, nuclear fuel cycle and resources, safety, safeguards, and radioactive waste management. The book also contains a discussion of the biological effects of nuclear radiation and of radiation protection, and a summary of the ten most relevant nuclear accidents. The book is suitable for undergraduates in physics, nuclear engineering and other science subjects. However, the mathematics is kept at a level that...

  3. Seventy-five years of nuclear fission

    Indian Academy of Sciences (India)

    technology can play such a vital role in a nation's development subsequently motivated ... fragments with a broad mass distribution is a unique nuclear phenomenon ... low energy and spontaneous fission of actinide nuclei and how these ...

  4. Feasibility study on fission moly target development

    International Nuclear Information System (INIS)

    Kim, Byung Ku; Kim, Seong Nyun; Shon, Dong Seong; Choi, Chang Beom; Lee, Jae Kuk; Park, Jin Ho; Jeong, Won Myung; Jeon, Kwan Sik; You, Jae Hyung; Kang, Kyung Chul; Ahn, Jong Hwan; Ju, Po Kuk

    1996-01-01

    A multi-purpose research reactor, HANARO has been operated on the beginning of 1995 and can be utilized for production of various radioisotopes. And a R and D program for fission Mo production was established, and the technical and economical feasibility study has been performed for fission Mo production in Korea. In this study the process for fission Mo production was recommended as follows; 1. Target : UO 2 of annulus type. 2. Separation and purification : Nitric acid dissolution → Alumina adsorption → Benzoin oxime precipitation → Alumina adsorption. And more desirable plan for steady supply of fission Mo were suggested in following viewpoints; 1. Technical collaboration with foreign company. 2. Backup supply system. 3. Marketing arrangement. (Author)

  5. Fission, fusion and photonuclear physics. Chapter 2

    International Nuclear Information System (INIS)

    Mazur, C.; Ribrag, M.

    Pronounced structures in the time of flight distribution of fission fragments, having a given energy, were recently reported. This experiment has been reproduced with a better time resolution and structures are not observed [fr

  6. Ternary fission induced by polarized neutrons

    Directory of Open Access Journals (Sweden)

    Gönnenwein Friedrich

    2013-12-01

    Full Text Available Ternary fission of (e,e U- and Pu- isotopes induced by cold polarized neutrons discloses some new facets of the process. In the so-called ROT effect shifts in the angular distributions of ternary particles relative to the fission fragments show up. In the so-called TRI effect an asymmetry in the emission of ternary particles relative to a plane formed by the fragment momentum and the spin of the neutron appear. The two effects are shown to be linked to the components of angular momentum perpendicular and parallel to the fission axis at the saddle point of fission. Based on theoretical models the spectroscopic properties of the collective transitional states at the saddle point are inferred from experiment.

  7. Hyperfission - a new mode of nuclear fission

    International Nuclear Information System (INIS)

    Ion, D.B.; Ivascu, M.; Ion-Mihai, R.

    1988-02-01

    In this paper the nuclear hyperfission as a new mode of fission, possible for heavy elements with Z > 92, is investigated. The Q-systematics, hyperfissibility parameters, hyperfission barrier as well as the essential hindrance factors are presented. The hyperfission hindrance factor relative to that of fission is found to be in the interval 1.0x10 -17 - 3.4x10 -16 for the parent nuclei with Z = 92-108. (orig.)

  8. Overview of tritium fast-fission yields

    International Nuclear Information System (INIS)

    Tanner, J.E.

    1981-03-01

    Tritium production rates are very important to the development of fast reactors because tritium may be produced at a greater rate in fast reactors than in light water reactors. This report focuses on tritium production and does not evaluate the transport and eventual release of the tritium in a fast reactor system. However, if an order-of-magnitude increase in fast fission yields for tritium is confirmed, fission will become the dominant production source of tritium in fast reactors

  9. Fission track dating of zircon: a multichronometer

    International Nuclear Information System (INIS)

    Carpena, J.

    1992-01-01

    Scattering in Fission Track ages of zircons of a single rock is possible when they present morphological and geochemical variations, if the greatest care is not taken in the choice of the etching conditions and the counting of tracks. The Fission Track study of two heterogeneous populations of zircons from the Mont Blanc granite and from the Gran Paradiso gneisses allows to show that zircon may work as a multichronometer

  10. Transport properties of fission product vapors

    International Nuclear Information System (INIS)

    Im, K.H.; Ahluwalia, R.K.

    1983-07-01

    Kinetic theory of gases is used to calculate the transport properties of fission product vapors in a steam and hydrogen environment. Provided in tabular form is diffusivity of steam and hydrogen, viscosity and thermal conductivity of the gaseous mixture, and diffusivity of cesium iodide, cesium hydroxide, diatomic tellurium and tellurium dioxide. These transport properties are required in determining the thermal-hydraulics of and fission product transport in light water reactors

  11. A revised calculational model for fission

    Energy Technology Data Exchange (ETDEWEB)

    Atchison, F

    1998-09-01

    A semi-empirical parametrization has been developed to calculate the fission contribution to evaporative de-excitation of nuclei with a very wide range of charge, mass and excitation-energy and also the nuclear states of the scission products. The calculational model reproduces measured values (cross-sections, mass distributions, etc.) for a wide range of fissioning systems: Nuclei from Ta to Cf, interactions involving nucleons up to medium energy and light ions. (author)

  12. Modeling Fission Product Sorption in Graphite Structures

    International Nuclear Information System (INIS)

    Szlufarska, Izabela; Morgan, Dane; Allen, Todd

    2013-01-01

    The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high-temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributions of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission products

  13. Apparatus for storing and processing fissionable substances

    International Nuclear Information System (INIS)

    Dubovsky, B.G.; Bogatyrev, V.K.; Vladykov, G.M.; Sviridenko, V.Y.

    1974-01-01

    An apparatus is described for storing and processing fissionable substances in which there is provided a protective shield in the form of a layer of neutron absorbing material located in direct proximity to a vessel with a fissionable substance contained therein. The layer of neutron retarding material according to the present invention has alternating projections and depressions facing the layer of neutron-absorbing material. (author)

  14. JNDC nuclear data library of fission products

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Ihara, Hitoshi; Akiyama, Masatsugu; Yoshida, Tadashi; Matumoto, Zyun-itiro; Nakasima, Ryuzo

    1983-10-01

    The JNDC (Japanese Nuclear Data Committee) FP (Fission Product) nuclear data library for 1172 fission products is described in this report. The gross theory of beta decay has been used extensively for estimating unknown decay data and also some of known decay data with poor accuracy. The calculated decay powers of fission products using the present library show excellent agreement with the latest measurements at ORNL (Oak Ridge National Laboratory), LANL (Los Alamos National Laboratory) and UTT (University of Tokyo, Tokai) for cooling times shorter than 10 3 s after irradiation. The calculated decay powers by the existing libraries showed systematic deviations at short cooling times; the calculated beta and gamma decay powers after burst fission were smaller than the experimental results for cooling times shorter than 10 s, and in the cooling time range 10 to 10 3 s the beta-decay power was larger than the measured values and the gamma decay power smaller than the measured results. The present JNDC FP nuclear data library resolved these discrepancies in the short cooling time ranges. The decay power of fission products has been calculated for ten fission types and the results have been fitted by an analytical function with 31 exponentials. This permits the easy application of the present results of decay power calculations to a LOCA (Loss-of-Coolant Accident) analysis of a light water reactor and so on. (author)

  15. Modelisation of the fission cross section

    International Nuclear Information System (INIS)

    Morariu, Claudia

    2013-03-01

    The neutron cross sections of four nuclear systems (n+ 235 U, n+ 233 U, n+ 241 Am and n+ 237 Np) are studied in the present document. The target nuclei of the first case, like 235 U and 239 Pu, have a large fission cross section after the absorption of thermal neutrons. These nuclei are called 'fissile' nuclei. The other type of nuclei, like 237 Np and 241 Am, fission mostly with fast neutrons, which exceed the fission threshold energy. These types of nuclei are called 'fertile'. The compound nuclei of the fertile nuclei have a binding energy higher than the fission barrier, while for the fissile nuclei the binding energy is lower than the fission barrier. In this work, the neutron induced cross sections for both types of nuclei are evaluated in the fast energy range. The total, reaction and shape-elastic cross sections are calculated by the coupled channel method of the optical model code ECIS, while the compound nucleus mechanism are treated by the statistical models implemented in the codes STATIS, GNASH and TALYS. The STATIS code includes a refined model of the fission process. Results from the theoretical calculations are compared with data retrieved from the experimental data base EXFOR. (author) [fr

  16. Dynamics in heavy ion fusion and fission

    International Nuclear Information System (INIS)

    Bjoernholm, S.

    1972-01-01

    Dynamical aspects of heavy ion fussion and fission, mainly the aspect of damping which is meant as the dissipation of kinetic energy and the aspect of the effective mass of the fission motion, are discussed. Two categories of evidence of damping effects are given. One relates to the damping of the fission motion for the ground state shape and for the isomeric more elongated shape. The other relates to the damping of the fission motion from the last barrier to the scission point. The dependence of the effective mass associated with the fission motion on the deormation of nucleus is shown. As the elongation of the nucleus increases the effective mass of the fission motion varies strongly from being about forty times greater than the reduced mass in the beta-vibrational state of the ground state shape to being equal to the reduced mass in the moment of scission. Damping effects are expected to be propartional to the difference between the effective mass and the reduced mass. It is concluded that the damping in fussion reactions is relatively weak for lighter products and quite strong for superheavy products like 236 U or 252 Cf. (S.B.)

  17. Theoretical descriptions of neutron emission in fission

    International Nuclear Information System (INIS)

    Madland, D.G.

    1991-01-01

    Brief descriptions are given of the observables in neutron emission in fission together with early theoretical representations of two of these observables, namely, the prompt fission neutron spectrum N(E) and the average prompt neutron multiplicity ν-bar p . This is followed by summaries, together with examples, of modern approaches to the calculation of these two quantities. Here, emphasis is placed upon the predictability and accuracy of the new approaches. In particular, the dependencies of N(E) and ν-bar p upon the fissioning nucleus and its excitation energy are discussed. Then, recent work in multiple-chance fission and other recent work involving new measurements are presented and discussed. Following this, some properties of fission fragments are mentioned that must be better known and better understood in order to calculate N(E) and ν-bar p with higher accuracy than is currently possible. In conclusion, some measurements are recommended for the purpose of benchmarking simultaneous calculations of neutron emission and gamma emission in fission. (author). 32 refs, 26 figs

  18. Fusion-fission hybrid reactors

    International Nuclear Information System (INIS)

    Greenspan, E.

    1984-01-01

    This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

  19. Fission of Polyanionic Metal Clusters

    Science.gov (United States)

    König, S.; Jankowski, A.; Marx, G.; Schweikhard, L.; Wolfram, M.

    2018-04-01

    Size-selected dianionic lead clusters Pbn2 -, n =34 - 56 , are stored in a Penning trap and studied with respect to their decay products upon photoexcitation. Contrary to the decay of other dianionic metal clusters, these lead clusters show a variety of decay channels. The mass spectra of the fragments are compared to the corresponding spectra of the monoanionic precursors. This comparison leads to the conclusion that, in the cluster size region below about n =48 , the fission reaction Pbn2 -→Pbn-10 -+Pb10- is the major decay process. Its disappearance at larger cluster sizes may be an indication of a nonmetal to metal transition. Recently, the pair of Pb10- and Pbn-10 - were observed as pronounced fragments in electron-attachment studies [S. König et al., Int. J. Mass Spectrom. 421, 129 (2017), 10.1016/j.ijms.2017.06.009]. The present findings suggest that this combination is the fingerprint of the decay of doubly charged lead clusters. With this assumption, the dianion clusters have been traced down to Pb212 -, whereas the smallest size for the direct observation was as high as n =28 .

  20. Fission Evaluation on Th-232

    International Nuclear Information System (INIS)

    Lee, Yong-Deok; Lee, Young-Ouk; Park, Joo-Hwan

    2007-01-01

    In recent years, several studies of neutron induced reaction on thorium were carried out in the framework of an IAEA coordinate research project involving a US contribution. The importance of Th-232 is for an innovative fuel cycle concept based on thorium fuel. Thorium fuels are also considered in accelerator driven system (ADS) to produce the power and radioactive waste transmutation. Therefore, the accurate neutron cross section for fission is crucially important for the design of various reactor systems. On December 2006, the ENDF/B-VII involving the new evaluation of actinides for Th-U fuel cycle was released. From the current environmental change, increasing oil price, air pollution by carbon dioxide, drain of oil resource, increasing demand of electricity, and energy independence, nuclear power is slowly to start to be reconsidered recently and it might be an alternative proposal as a production facility of energy and a reuse of resources. Even though it produces the nuclear wastes, it has an advantage in the emission of greenhouse gases. Therefore, new concept of nuclear technology to be developed for power production is subject to the condition of increased safety, reduction of nuclear wastes, resistance to nuclear material proliferation, Thorium fuel cycle is the most feasible option to satisfy the condition. Specially, thorium reserves are much larger than those of uranium

  1. Correlated prompt fission data in transport simulations

    Science.gov (United States)

    Talou, P.; Vogt, R.; Randrup, J.; Rising, M. E.; Pozzi, S. A.; Verbeke, J.; Andrews, M. T.; Clarke, S. D.; Jaffke, P.; Jandel, M.; Kawano, T.; Marcath, M. J.; Meierbachtol, K.; Nakae, L.; Rusev, G.; Sood, A.; Stetcu, I.; Walker, C.

    2018-01-01

    Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n - n, n - γ, and γ - γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ rays from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX - PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in

  2. Correlated prompt fission data in transport simulations

    Energy Technology Data Exchange (ETDEWEB)

    Talou, P.; Jaffke, P.; Kawano, T.; Stetcu, I. [Los Alamos National Laboratory, Nuclear Physics Group, Theoretical Division, Los Alamos, NM (United States); Vogt, R. [Lawrence Livermore National Laboratory, Nuclear and Chemical Sciences Division, Livermore, CA (United States); University of California, Physics Department, Davis, CA (United States); Randrup, J. [Lawrence Berkeley National Laboratory, Nuclear Science Division, Berkeley, CA (United States); Rising, M.E.; Andrews, M.T.; Sood, A. [Los Alamos National Laboratory, Monte Carlo Methods, Codes, and Applications Group, Los Alamos, NM (United States); Pozzi, S.A.; Clarke, S.D.; Marcath, M.J. [University of Michigan, Department of Nuclear Engineering and Radiological Sciences, Ann Arbor, MI (United States); Verbeke, J.; Nakae, L. [Lawrence Livermore National Laboratory, Nuclear and Chemical Sciences Division, Livermore, CA (United States); Jandel, M. [Los Alamos National Laboratory, Nuclear and Radiochemistry Group, Los Alamos, NM (United States); University of Massachusetts, Department of Physics and Applied Physics, Lowell, MA (United States); Meierbachtol, K. [Los Alamos National Laboratory, Nuclear Engineering and Nonproliferation, Los Alamos, NM (United States); Rusev, G.; Walker, C. [Los Alamos National Laboratory, Nuclear and Radiochemistry Group, Los Alamos, NM (United States)

    2018-01-15

    Detailed information on the fission process can be inferred from the observation, modeling and theoretical understanding of prompt fission neutron and γ-ray observables. Beyond simple average quantities, the study of distributions and correlations in prompt data, e.g., multiplicity-dependent neutron and γ-ray spectra, angular distributions of the emitted particles, n-n, n-γ, and γ-γ correlations, can place stringent constraints on fission models and parameters that would otherwise be free to be tuned separately to represent individual fission observables. The FREYA and CGMF codes have been developed to follow the sequential emissions of prompt neutrons and γ rays from the initial excited fission fragments produced right after scission. Both codes implement Monte Carlo techniques to sample initial fission fragment configurations in mass, charge and kinetic energy and sample probabilities of neutron and γ emission at each stage of the decay. This approach naturally leads to using simple but powerful statistical techniques to infer distributions and correlations among many observables and model parameters. The comparison of model calculations with experimental data provides a rich arena for testing various nuclear physics models such as those related to the nuclear structure and level densities of neutron-rich nuclei, the γ-ray strength functions of dipole and quadrupole transitions, the mechanism for dividing the excitation energy between the two nascent fragments near scission, and the mechanisms behind the production of angular momentum in the fragments, etc. Beyond the obvious interest from a fundamental physics point of view, such studies are also important for addressing data needs in various nuclear applications. The inclusion of the FREYA and CGMF codes into the MCNP6.2 and MCNPX-PoliMi transport codes, for instance, provides a new and powerful tool to simulate correlated fission events in neutron transport calculations important in nonproliferation

  3. APPLE, Plot of 1-D Multigroup Neutron Flux and Gamma Flux and Reaction Rates from ANISN

    International Nuclear Information System (INIS)

    Kawasaki, Hiromitsu; Seki, Yasushi

    1983-01-01

    A - Description of problem or function: The APPLE-2 code has the following functions: (1) It plots multi-group energy spectra of neutron and/or gamma ray fluxes calculated by ANISN, DOT-3.5, and MORSE. (2) It gives an overview plot of multi-group neutron fluxes calculated by ANISN and DOT-3.5. The scalar neutron flux phi(r,E) is plotted with the spatial parameter r linear along the Y-axis, logE along the X-axis and log phi(r,E) in the Z direction. (3) It calculates the spatial distribution and region volume integrated values of reaction rates using the scalar flux calculated with ANISN and DOT-3.5. (4) Reaction rate distribution along the R or Z direction may be plotted. (5) An overview plot of reaction rates or scalar fluxes summed over specified groups may be plotted. R(ri,zi) or phi(ri,zi) is plotted with spatial parameters r and z along the X- and Y-axes in an orthogonal coordinate system. (6) Angular flux calculated by ANISN is rearranged and a shell source at any specified spatial mesh point may be punched out in FIDO format. The shell source obtained may be employed in solving deep penetration problems with ANISN, when the entire reactor system is divided into two or more parts and the neutron fluxes in two adjoining parts are connected by using the shell source. B - Method of solution: (a) The input data specification is made as simple as possible by making use of the input data required in the radiation transport code. For example, geometry related data in ANISN and DOT are transmitted to APPLE-2 along with scalar flux data so as to reduce duplicity and errors in reproducing these data. (b) Most the input data follow the free form FIDO format developed at Oak Ridge National Laboratory and used in the ANISN code. Furthermore, the mixture specifying method used in ANISN is also employed by APPLE-2. (c) Libraries for some standard response functions required in fusion reactor design have been prepared and are made available to users of the 42-group neutron

  4. Dual-fission chamber and neutron beam characterization for fission product yield measurements using monoenergetic neutrons

    Science.gov (United States)

    Bhatia, C.; Fallin, B.; Gooden, M. E.; Howell, C. R.; Kelley, J. H.; Tornow, W.; Arnold, C. W.; Bond, E. M.; Bredeweg, T. A.; Fowler, M. M.; Moody, W. A.; Rundberg, R. S.; Rusev, G.; Vieira, D. J.; Wilhelmy, J. B.; Becker, J. A.; Macri, R.; Ryan, C.; Sheets, S. A.; Stoyer, M. A.; Tonchev, A. P.

    2014-09-01

    A program has been initiated to measure the energy dependence of selected high-yield fission products used in the analysis of nuclear test data. We present out initial work of neutron activation using a dual-fission chamber with quasi-monoenergetic neutrons and gamma-counting method. Quasi-monoenergetic neutrons of energies from 0.5 to 15 MeV using the TUNL 10 MV FM tandem to provide high-precision and self-consistent measurements of fission product yields (FPY). The final FPY results will be coupled with theoretical analysis to provide a more fundamental understanding of the fission process. To accomplish this goal, we have developed and tested a set of dual-fission ionization chambers to provide an accurate determination of the number of fissions occurring in a thick target located in the middle plane of the chamber assembly. Details of the fission chamber and its performance are presented along with neutron beam production and characterization. Also presented are studies on the background issues associated with room-return and off-energy neutron production. We show that the off-energy neutron contribution can be significant, but correctable, while room-return neutron background levels contribute less than <1% to the fission signal.

  5. Asymmetry in ternary fission induced by polarized neutrons and fission mechanism

    International Nuclear Information System (INIS)

    Bunakov, V.E.; Gennenvajn, F.; Dzhessinger, P.; Mutterer, M.; Petrov, G.A.

    2003-01-01

    The results of measuring the P-odd, P-even (right-left) and T-odd asymmetries of the charged particles emission in the double and ternary fission, induced by the polarized neutrons, are considered. It is shown, what kind of information on the mechanism of the ternary nuclear fission may be obtained from the theoretical analysis of these data [ru

  6. Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh; Calculs de reference avec un maillage multigroupe fin sur des assemblages critiques par Apollo2

    Energy Technology Data Exchange (ETDEWEB)

    Aggery, A

    1999-12-01

    The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

  7. Analysis of coupled neutron-gamma radiations, applied to shieldings in multigroup albedo method

    International Nuclear Information System (INIS)

    Dunley, Leonardo Souza

    2002-01-01

    The principal mathematical tools frequently available for calculations in Nuclear Engineering, including coupled neutron-gamma radiations shielding problems, involve the full Transport Theory or the Monte Carlo techniques. The Multigroup Albedo Method applied to shieldings is characterized by following the radiations through distinct layers of materials, allowing the determination of the neutron and gamma fractions reflected from, transmitted through and absorbed in the irradiated media when a neutronic stream hits the first layer of material, independently of flux calculations. Then, the method is a complementary tool of great didactic value due to its clarity and simplicity in solving neutron and/or gamma shielding problems. The outstanding results achieved in previous works motivated the elaboration and the development of this study that is presented in this dissertation. The radiation balance resulting from the incidence of a neutronic stream into a shielding composed by 'm' non-multiplying slab layers for neutrons was determined by the Albedo method, considering 'n' energy groups for neutrons and 'g' energy groups for gammas. It was taken into account there is no upscattering of neutrons and gammas. However, it was considered that neutrons from any energy groups are able to produce gammas of all energy groups. The ANISN code, for an angular quadrature order S 2 , was used as a standard for comparison of the results obtained by the Albedo method. So, it was necessary to choose an identical system configuration, both for ANISN and Albedo methods. This configuration was six neutron energy groups and eight gamma energy groups, using three slab layers (iron aluminum - manganese). The excellent results expressed in comparative tables show great agreement between the values determined by the deterministic code adopted as standard and, the values determined by the computational program created using the Albedo method and the algorithm developed for coupled neutron

  8. Symmetry breaking in the opinion dynamics of a multi-group project organization

    International Nuclear Information System (INIS)

    Zhu Zhen-Tao; Zhou Jing; Chen Xing-Guang; Li Ping

    2012-01-01

    A bounded confidence model of opinion dynamics in multi-group projects is presented in which each group's opinion evolution is driven by two types of forces: (i) the group's cohesive force which tends to restore the opinion back towards the initial status because of its company culture; and (ii) nonlinear coupling forces with other groups which attempt to bring opinions closer due to collaboration willingness. Bifurcation analysis for the case of a two-group project shows a cusp catastrophe phenomenon and three distinctive evolutionary regimes, i.e., a deadlock regime, a convergence regime, and a bifurcation regime in opinion dynamics. The critical value of initial discord between the two groups is derived to discriminate which regime the opinion evolution belongs to. In the case of a three-group project with a symmetric social network, both bifurcation analysis and simulation results demonstrate that if each pair has a high initial discord, instead of symmetrically converging to consensus with the increase of coupling scale as expected by Gabbay's result (Physica A 378 (2007) p. 125 Fig. 5), project organization (PO) may be split into two distinct clusters because of the symmetry breaking phenomenon caused by pitchfork bifurcations, which urges that apart from divergence in participants' interests, nonlinear interaction can also make conflict inevitable in the PO. The effects of two asymmetric level parameters are tested in order to explore the ways of inducing dominant opinion in the whole PO. It is found that the strong influence imposed by a leader group with firm faith on the flexible and open minded follower groups can promote the formation of a positive dominant opinion in the PO

  9. Symmetry breaking in the opinion dynamics of a multi-group project organization

    Science.gov (United States)

    Zhu, Zhen-Tao; Zhou, Jing; Li, Ping; Chen, Xing-Guang

    2012-10-01

    A bounded confidence model of opinion dynamics in multi-group projects is presented in which each group's opinion evolution is driven by two types of forces: (i) the group's cohesive force which tends to restore the opinion back towards the initial status because of its company culture; and (ii) nonlinear coupling forces with other groups which attempt to bring opinions closer due to collaboration willingness. Bifurcation analysis for the case of a two-group project shows a cusp catastrophe phenomenon and three distinctive evolutionary regimes, i.e., a deadlock regime, a convergence regime, and a bifurcation regime in opinion dynamics. The critical value of initial discord between the two groups is derived to discriminate which regime the opinion evolution belongs to. In the case of a three-group project with a symmetric social network, both bifurcation analysis and simulation results demonstrate that if each pair has a high initial discord, instead of symmetrically converging to consensus with the increase of coupling scale as expected by Gabbay's result (Physica A 378 (2007) p. 125 Fig. 5), project organization (PO) may be split into two distinct clusters because of the symmetry breaking phenomenon caused by pitchfork bifurcations, which urges that apart from divergence in participants' interests, nonlinear interaction can also make conflict inevitable in the PO. The effects of two asymmetric level parameters are tested in order to explore the ways of inducing dominant opinion in the whole PO. It is found that the strong influence imposed by a leader group with firm faith on the flexible and open minded follower groups can promote the formation of a positive dominant opinion in the PO.

  10. Reliability generalization of the Multigroup Ethnic Identity Measure-Revised (MEIM-R).

    Science.gov (United States)

    Herrington, Hayley M; Smith, Timothy B; Feinauer, Erika; Griner, Derek

    2016-10-01

    [Correction Notice: An Erratum for this article was reported in Vol 63(5) of Journal of Counseling Psychology (see record 2016-33161-001). The name of author Erika Feinauer was misspelled as Erika Feinhauer. All versions of this article have been corrected.] Individuals' strength of ethnic identity has been linked with multiple positive indicators, including academic achievement and overall psychological well-being. The measure researchers use most often to assess ethnic identity, the Multigroup Ethnic Identity Measure (MEIM), underwent substantial revision in 2007. To inform scholars investigating ethnic identity, we performed a reliability generalization analysis on data from the revised version (MEIM-R) and compared it with data from the original MEIM. Random-effects weighted models evaluated internal consistency coefficients (Cronbach's alpha). Reliability coefficients for the MEIM-R averaged α = .88 across 37 samples, a statistically significant increase over the average of α = .84 for the MEIM across 75 studies. Reliability coefficients for the MEIM-R did not differ across study and participant characteristics such as sample gender and ethnic composition. However, consistently lower reliability coefficients averaging α = .81 were found among participants with low levels of education, suggesting that greater attention to data reliability is warranted when evaluating the ethnic identity of individuals such as middle-school students. Future research will be needed to ascertain whether data with other measures of aspects of personal identity (e.g., racial identity, gender identity) also differ as a function of participant level of education and associated cognitive or maturation processes. (PsycINFO Database Record (c) 2016 APA, all rights reserved).

  11. Fission-product yields for thermal-neutron fission of curium-243

    International Nuclear Information System (INIS)

    Breederland, D.G.

    1982-01-01

    Cumulative fission yields for 25 gamma rays emitted during the decay of 23 fission products produced by thermal-neutron fission of 243 Cm have been determined. Using Ge(Li) spectroscopy, 33 successive pulse-height spectra of gamma rays emitted from a 77-ng sample of 243 Cm over a period of approximately two and one-half months were analyzed. Reduction of these spectra resulted in the identification and matching of gamma-ray energies and half-lives to specific radionuclides. Using these results, 23 cumulative fission-product yields were calculated. Only those radionuclides having half-lives between 6 hours and 65 days were observed. Prior to this experiment, no fission-product yields had been recorded for 243 Cm

  12. Fission Surface Power Technology Development Update

    Science.gov (United States)

    Palac, Donald T.; Mason, Lee S.; Houts, Michael G.; Harlow, Scott

    2011-01-01

    Power is a critical consideration in planning exploration of the surfaces of the Moon, Mars, and places beyond. Nuclear power is an important option, especially for locations in the solar system where sunlight is limited or environmental conditions are challenging (e.g., extreme cold, dust storms). NASA and the Department of Energy are maintaining the option for fission surface power for the Moon and Mars by developing and demonstrating technology for a fission surface power system. The Fission Surface Power Systems project has focused on subscale component and subsystem demonstrations to address the feasibility of a low-risk, low-cost approach to space nuclear power for surface missions. Laboratory demonstrations of the liquid metal pump, reactor control drum drive, power conversion, heat rejection, and power management and distribution technologies have validated that the fundamental characteristics and performance of these components and subsystems are consistent with a Fission Surface Power preliminary reference concept. In addition, subscale versions of a non-nuclear reactor simulator, using electric resistance heating in place of the reactor fuel, have been built and operated with liquid metal sodium-potassium and helium/xenon gas heat transfer loops, demonstrating the viability of establishing system-level performance and characteristics of fission surface power technologies without requiring a nuclear reactor. While some component and subsystem testing will continue through 2011 and beyond, the results to date provide sufficient confidence to proceed with system level technology readiness demonstration. To demonstrate the system level readiness of fission surface power in an operationally relevant environment (the primary goal of the Fission Surface Power Systems project), a full scale, 1/4 power Technology Demonstration Unit (TDU) is under development. The TDU will consist of a non-nuclear reactor simulator, a sodium-potassium heat transfer loop, a power

  13. Correlation of recent fission product release data

    International Nuclear Information System (INIS)

    Kress, T.S.; Lorenz, R.A.; Nakamura, T.; Osborne, M.F.

    1989-01-01

    For the calculation of source terms associated with severe accidents, it is necessary to model the release of fission products from fuel as it heats and melts. Perhaps the most definitive model for fission product release is that of the FASTGRASS computer code developed at Argonne National Laboratory. There is persuasive evidence that these processes, as well as additional chemical and gas phase mass transport processes, are important in the release of fission products from fuel. Nevertheless, it has been found convenient to have simplified fission product release correlations that may not be as definitive as models like FASTGRASS but which attempt in some simple way to capture the essence of the mechanisms. One of the most widely used such correlation is called CORSOR-M which is the present fission product/aerosol release model used in the NRC Source Term Code Package. CORSOR has been criticized as having too much uncertainty in the calculated releases and as not accurately reproducing some experimental data. It is currently believed that these discrepancies between CORSOR and the more recent data have resulted because of the better time resolution of the more recent data compared to the data base that went into the CORSOR correlation. This document discusses a simple correlational model for use in connection with NUREG risk uncertainty exercises. 8 refs., 4 figs., 1 tab

  14. Investigation of delayed fission gas release

    International Nuclear Information System (INIS)

    Cayet, Nicolas

    1996-05-01

    The study of the fission gas release process in the high burnup rig IFA-562 has revealed a particular fuel behaviour: a delay in the fission gas release process. It appeared that an important release of gas was measured by the pressure transducers once the power had decreased, whereas, during steady-state operation, the pressure did not increase very much. After examinations, the gap size has been concluded to be the main parameter involving this delay. However the burnup could have been a potential factor, its role is mainly to close the gap by swelling. The observations of low burnup rods have shown the same delayed fission gas release, the gap being small by design and closed essentially by thermal expansion. The study of the kinetics has demonstrated the time-independency of the phenomenon. Thus the proposed mechanism driving this delayed fission gas release would involve three consecutives stages. During steady-state, the gas is released into the interlinkage network of grain boundary bubbles and cracks. Due to the closed gap, the gas is trapped in some void volumes, unable to escape the pellet. During power reduction, the gap and some old/new cracks open, immediately providing a path for the gas to the pressure transducers and explaining this delay in the fission gas release. (author)

  15. An improved technique for fission track dating

    International Nuclear Information System (INIS)

    Zhao Yunlong; Wu Zhaohui; Xia Yuliang

    1996-01-01

    The necessity of improving the fission track dating (FTD) technique both at home and abroad is illustrated. The ways of making such improvement are also proposed. It is suggested to calibrate the constant b value of the uranium standard glass by using the method of fission products activity. The 3 kinds of uranium standard glass which have been calibrated are NBS SRM962a, UB 1 and UB 2 . An established new method σ·Φ ρ d /b, to measure neutron fluence, avoids the influence of the varying neutron spectrum on measuring neutron fluence. The improved etching technique for fission tracks in zircon adopted a two-step method which includes the molten alkali system etching using NaOH + KOH and the mixed acid system etching using HNO 3 + HF; this technique results in adequate track etching, increased track clarity and less interference. In this way the intensity of tracks is authentically reflected. Dividing angular zone in accordance with the angular distribution of spontaneous fission track on the crystal surface of minerals to count the tracks and using the improved etching technique to remove the non-uniform angular distribution of spontaneous fission tracks in zircon, ensure the accuracy of tracks count. The improved FTD techniques were used to finish Laboratory Standardized Calibration. The tests using international FTD age standards samples have proved that above mentioned techniques are reliable and practical in obtaining the accurate FTD data. (8 tabs.; 3 figs.)

  16. Nuclear fission induced by heavy ions

    International Nuclear Information System (INIS)

    Newton, J.O.

    1988-09-01

    Because the accelerators of the 50's and 60's mostly provided beams of light ions, well suited for studying individual quantum states of low angular momentum or reactions involving the transfer of one or two nucleons, the study of fission, being an example of large-scale collective motion, has until recently been outside of the mainstream of nuclear research. This situation has changed in recent years, due to the new generation of accelerators capable of producing beams of heavy ions with energies high enough to overcome the Coulomb barriers of all stable nuclei. These have made possible the study of new examples of large-scale collective motions, involving major rearrangements of nuclear matter, such as deep-inelastic collisions and heavy-ion fusion. Perhaps the most exciting development in the past few years is the discovery that dissipative effects (nuclear viscosity) play an important role in fission induced by heavy ions, contrary to earlier assumptions that the viscosity involved in fission was very weak and played only a minor role. This review will be mainly concerned with developments in heavy-ion induced fission during the last few years and have an emphasis on the very recent results on dissipative effects. Since heavy-ion bombardment usually results in compound systems with high excitation energies and angular momenta, shell effects might be expected to be small, and the subject of low energy fission, where they are important, will not be addressed. 285 refs., 58 figs

  17. Some aspects of the nuclear fission process

    International Nuclear Information System (INIS)

    Netter, F.

    1961-01-01

    In the following report one can find first a short general view on the present situation of our knowledge concerning the nuclear fission process, namely on the nucleus going through the saddle-point. Then there are some aspects connected with the excitation energy of the fissioning nucleus. The measurements made at Saclay on the fast neutron fission cross-section of U 233 , U 235 , Pu 239 , U 238 are described at the beginning of this work. It appears that for U 233 there is some characteristic shape modulation of the cross-section curve, in relation with the collective excited state of the deformed nucleus at the saddle-point. Good evidence of this is also given by the study of the relative fission rate with emission of long-range particles; it appears also that this ternary fission rate does not change substantially for neutron between thermal energy and 2 MeV, but that is very lower for the compound nucleus U 239 than for even-even compound nuclei. At the end there are some experiments on the strong 4,5 MeV gamma-ray originated by slow neutron absorption in U 235 . Time-of-flight device is used to establish that this 4,5 MeV gamma-ray seems mostly connected with radiative capture. (author) [fr

  18. Rupture of the neck in nuclear fission

    International Nuclear Information System (INIS)

    Davies, K.T.R.; Managan, R.A.; Nix, J.R.; Sierk, A.J.

    1977-01-01

    We introduce a degree of freedom to describe the rupture of the neck in nuclear fission and calculate the point at which the neck ruptures as the nucleus descends dynamically from its fission saddle point. This is done by mentally slicing the system into two portions at its minimum neck radius and calculating the force required to separate the two portions while keeping their shapes fixed. This force is obtained by differentiating with respect to separation the sum of the Coulomb and nuclear interaction energies between the two portions. For nuclei throughout the Periodic Table we calculate this force along dynamical paths leading from the fission saddle point. The force is initially attractive but becomes repulsive when the neck reaches a critical size. For actinide nuclei the neck radius at which rupture occurs is about 2 fm. This increases the calculated translational kinetic energy of the fission fragments at infinity relative to that calculated for scission occurring at zero neck radius. With the effect of neck rupture taken into account, we calculate and compare with experimental results fission-fragment kinetic energies for two types of nuclear dissipation: ordinary two-body viscosity and one-body dissipation

  19. Theory of neutron emission in fission

    International Nuclear Information System (INIS)

    Madland, D.G.

    1998-01-01

    A survey of theoretical representations of two of the observables in neutron emission in fission is given, namely, the prompt fission neutron spectrum N(E) and the average prompt neutron multiplicity bar ν p . Early representations of the two observables are presented and their deficiencies are discussed. This is followed by summaries and some examples of recent theoretical models for the calculation of these quantities. Emphasis is placed upon the predictability and accuracy of the recent models. In particular, the dependencies of N(E) and bar ν p upon the fissioning nucleus and its excitation energy are treated in detail for the Los Alamos model. Recent work in the calculation of the prompt fission neutron spectrum matrix N(E, E n ), where E n is the energy of the neutron inducing fission, is then discussed. Concluding remarks address the current status of the ability to calculate these observables with confidence, the direction of future theoretical efforts, and limitations to current (and future) approaches. This paper is an extension of a similar paper presented at the International Centre for Theoretical Physics in 1996

  20. Theoretical models of neutron emission in fission

    International Nuclear Information System (INIS)

    Madland, D.G.

    1992-01-01

    A brief survey of theoretical representations of two of the observables in neutron emission in fission is given, namely, the prompt fission neutron spectrum N(E) and the average prompt neutron multiplicity bar v p . Early representations of the two observables are presented and their deficiencies are discussed. This is followed by summaries and examples of recent theoretical models for the calculation of these quantities. Emphasis is placed upon the predictability and accuracy of the new models. In particular, the dependencies of N(E) and bar v p upon the fissioning nucleus and its excitation energy are treated. Recent work in the calculation of the prompt fission neutron spectrum matrix N(E,E n ), where E n is the energy of the neutron inducing fission, is then discussed. Concluding remarks address the current status of our ability to calculate these observables with confidence, the direction of future theoretical efforts, and limititations to current and future calculations. Finally, recommendations are presented as to which model should be used currently and which model should be pursued in future efforts