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

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.)

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

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

One-, two- and three-dimensional transport codes using multi-group double-differential form cross sections

International Nuclear Information System (INIS)

Mori, Takamasa; Nakagawa, Masayuki; Sasaki, Makoto.

1988-11-01

We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)

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)

Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

International Nuclear Information System (INIS)

Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.

1990-09-01

Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations

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

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

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%

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

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.

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

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)

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)

PROF-DD, Generator of Multigroup Cross-Sections Library DDX for MORSE-DD, ANISN-DD, DOT-DD

International Nuclear Information System (INIS)

Mori, Takamasa; Nakagawa, Masayuki; Ishiguro, Yukio

2002-01-01

1 - Description of program or function: The code system PROF-DD generates a multi-group double-differential cross section library DDX from evaluated data in ENDF/B-IV or ENDF/B-V format. The system consists of the following five modules: PROF-DDX is the main module of the system. It calculates the multigroup DDX and stores them on a master PDS file. MCFILEF generates a control file for PROF-DDX, which contains energy group and angle bin structures. SPINPTF prepares an input data file for PROF-DDX by combining the control file with other input data. DDXLIBMK edits a DDX library from the master PDS file for transport calculations. RESENDD performs resonance cross section and Doppler broadening calculations. 2 - Restrictions on the complexity of the problem: The numbers of energy groups and angle bins are less than 150 and 40, respectively

ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies

International Nuclear Information System (INIS)

Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.

1987-01-01

Multigroup P3 neutron, P0-P3 secondary gamma ray production (SGRP), and P6 gamma ray interaction (GRI) cross section libraries have been generated to support design work on the Advanced Neutron Source (ANS) reactor. The libraries, designated ANSL-V (Advanced Neutron Source Cross-Section Libraries), are data bases in a format suitable for subsequent generation of problem dependent cross sections. The ANSL-V libraries are available on magnetic tape from the Radiation Shielding Information Center at Oak Ridge National Laboratory

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

A computer program with graphical user interface to plot the multigroup cross sections of WIMS-D library

International Nuclear Information System (INIS)

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

2002-01-01

As a result of the IAEA Co-ordinated Research Programme entitled 'Final Stage of the WIMS Library Update Project', new and updated WIMS-D libraries based upon ENDF/B-VI.5, JENDL-3.2 and JEF-2.2 have become available. A project to prepare an exhaustive handbook of WIMS-D cross sections from old and new libraries has been taken up by the authors. As part of this project, we have developed a computer program XnWlup with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualization 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. The current features of the software, on-line help manual and future plans for further development are described in this paper

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)

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

Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

International Nuclear Information System (INIS)

Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

1986-02-01

Multigroup cross sections (66 neutron groups and 22 photon groups) are described for neutron energies from thermal to 400 MeV. The elements considered are hydrogen, 10 B, 11 B, carbon, nitrogen, oxygen, sodium, magnesium, aluminum, silicon, sulfur, potassium, calcium, chromium, iron, nickel, tungsten, and lead. The cross section data presented are a revision of similar data presented previously. In the case of iron, transport calculations using the earlier and the revised cross sections are presented and compared, and significant differences are found. The revised cross sections are available from the Radiation Shielding information Center of the Oak Ridge National Laboratory. 32 refs., 5 figs., 3 tabs

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

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)

Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

International Nuclear Information System (INIS)

Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

1986-01-01

For a variety of applications, e.g., accelerator shielding design, neutrons in radiotherapy, radiation damage studies, etc., it is necessary to carry out transport calculations involving medium-energy (greater than or equal to20 MeV) neutrons. A previous paper described neutron-photon multigroup cross sections in the ANISN format for neutrons from thermal to 400 MeV. In the present paper the cross-section data presented previously have been revised to make them agree with available experimental data. 7 refs., 1 fig

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

MINX, Multigroup Cross-Sections and Self-Shielding Factors from ENDF/B for Program SPHINX

International Nuclear Information System (INIS)

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

1977-01-01

1 - Description of problem or function: MINX calculates fine-group averaged infinitely diluted cross sections and self-shielding factors from ENDF/B-IV data. Its primary purpose is to generate a pseudo-composition-independent multigroup library which is input to the SPHINX space-energy collapse program (2) (PSR-0129) through standard CCCC-III (8) interfaces. MINX incorporates and improves upon the resonance capabilities of existing codes such as ETOX (5) (NESC0388) and ENDRUN (9) and the high-order group-to-group transfer matrices of SUPERTOG (10) (PSR-0013) and ETOG (11). Fine group energy boundaries, Legendre expansion order, gross spectral shape component (in the Bondarenko flux model), temperatures and dilutions can all be used specifically. 2 - Method of solution: Infinitely dilute, un-broadened point cross sections are obtained from resolved resonance parameters using a modified version of the RESEND program (3) (NESC0465). The SIGMA1 (4) (IAEA0854) kernel-broadening method is used to Doppler broaden and thin the tabulated linearized pointwise cross sections at 0 K (outside of the unresolved energy region). Effective temperature- dependent self-shielded pointwise cross sections are derived from the formulation in the ETOX code. The primary modification to the ETOX algorithm is associated with the numerical quadrature scheme used to establish the mean values of the fluctuation intervals. The selection of energy mesh points, at which the effective cross sections are calculated, has been modified to include the energy points given in the ENDF/B file or, if the energy-independent formalism was employed, points at half-lethargy intervals. Infinitely dilute group cross sections and self-shielding factors are generated using the Bondarenko flux weighting model with the gross spectral shape under user control. The integral over energy for each group is divided into a set of panels defined by the union of the grid points describing the total cross section, the

Neutron-photon multigroup cross sections for neutron energies up to 400 MeV: HILO86R

International Nuclear Information System (INIS)

Kotegawa, Hiroshi; Nakane, Yoshihiro; Hasegawa, Akira; Tanaka, Shun-ichi

1993-02-01

A macroscopic multigroup cross section library of 66 neutron and 22 photon groups for neutron energies up to 400 MeV: HILO86R is prepared for 10 typical shielding materials; water, concrete, iron, air, graphite, polyethylene, heavy concrete, lead, aluminum and soil. The library is a revision of the DLC-119/HILO86, in which only the cross sections below 19.6 MeV have been exchanged with a group cross section processed from the JENDL-3 microscopic cross section library. In the HILO86R library, self shielding factors are used to produce effective cross sections for neutrons less than 19.6 MeV considering rather coarse energy meshes. Energy spectra and dose attenuation in water, concrete and iron have been compared among the HILO, HILO86 and HILO86R libraries for different energy neutron sources. Significant discrepancy has been observed in the energy spectra less than a couple of MeV energy in iron among the libraries, resulting large difference in the dose attenuation. The difference was attributed to the effect of self-shielding factor, namely to the difference between infinite dilution and effective cross sections. Even for 400 MeV neutron source the influence of the self-shielding factor is significant, nevertheless only the cross sections below 19.6 MeV are exchanged. (author)

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)

Comparative analysis among several cross section sets

International Nuclear Information System (INIS)

Caldeira, A.D.

1983-01-01

Critical parameters were calculated using the one dimensional multigroup transport theory for several cross section sets. Calculations have been performed for water mixtures of uranium metal, plutonium metal and uranium-thorium oxide, and for metallics systems, to determine the critical dimensions of geometries (sphere and cylinder). For this aim, the following cross section sets were employed: 1) multigroup cross section sets obtained from the GAMTEC-II code; 2) the HANSEN-ROACH cross section sets; 3) cross section sets from the ENDF/B-IV, processed by the NJOY code. Finally, we have also calculated the corresponding critical radius using the one dimensional multigroup transport DTF-IV code. The numerical results agree within a few percent with the critical values obtained in the literature (where the greatest discrepancy occured in the critical dimensions of water mixtures calculated with the values generated by the NJOY code), a very good results in comparison with similar works. (Author) [pt

Analysis of benchmark experiments for testing the IKE multigroup cross-section libraries based on ENDF/B-III and IV

International Nuclear Information System (INIS)

Keinert, J.; Mattes, M.

1975-01-01

Benchmark experiments offer the most direct method for validation of nuclear cross-section sets and calculational methods. For 16 fast and thermal critical assemblies containing uranium and/or plutonium of different compositions we compared our calculational results with measured integral quantities, such as ksub(eff), central reaction rate ratios or fast and thermal activation (dis)advantage factors. Cause of the simple calculational modelling of these assemblies the calculations proved as a good test for the IKE multigroup cross-section libraries essentially based on ENDF/B-IV. In general, our calculational results are in excellent agreement with the measured values. Only with some critical systems the basic ENDF/B-IV data proved to be insufficient in calculating ksub(eff), probably due to Pu neutron data and U 238 fast capture cross-sections. (orig.) [de

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

Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC 2 -2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC 2 -2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC 2 -2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC 2 -2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC 2 -2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC 2 -2, VIM, and NJOY. For almost all nuclides considered, MC 2 -2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC 2 -2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC 2 -2/TWODANT calculations were in good agreement with MCNP solutions within ∼0.25% Δρ, except a few small LANL fast assemblies. Relative to the MCNP solution, the MC 2 -2/TWODANT

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

Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies

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)

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

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

BUGLE-96: A revised multigroup cross section library for LWR applications based on ENDF/B-VI Release 3

International Nuclear Information System (INIS)

White, J.E.; Ingersoll, D.T.; Slater, C.O.; Roussin, R.W.

1996-01-01

A revised multigroup cross-section library based ON ENDF/B-VI Release 3 has been produced for light water reactor shielding and reactor pressure vessel dosimetry applications. This new broad-group library, which is designated BUGLE-96, represents an improvement over the BUGLE-93 library released in February 1994 and is expected to replace te BUGLE-93 data. The cross-section processing methodology is the same as that used for producing BUGLE-93 and is consistent with ANSI/ANS 6.1.2. As an added feature, cross-section sets having upscatter data for four thermal neutron groups are included in the BUGLE-96 package available from the Radiation Shielding Information Center. The upscattering data should improve the application of this library to the calculation of more accurate thermal fluences, although more computer time will be required. The incorporation of feedback from users has resulted in a data library that addresses a wider spectrum of user needs

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)

Generation of multigroup cross-sections from micro-group ones in code system SUHAM-U used for VVER-1000 reactor core calculations with MOX loading

Energy Technology Data Exchange (ETDEWEB)

Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)

2005-07-01

At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.

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

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)

THEMIS-4: a coherent punctual and multigroup cross section library for Monte Carlo and SN codes from ENDF/B4

International Nuclear Information System (INIS)

Dejonghe, G.; Gonnord, J.; Monnier, A.; Nimal, J.C.

1983-05-01

The THEMIS cross section processing system has been developped to produce punctual data for MONTE CARLO and coherent multigroup data for SN codes from ENDF/B. The THEMIS-4 data base has been generated from ENDF/B4 using the system and can be accessed by the 3-D Monte Carlo system TRIPOLI-2 and by the SN codes ANISN and DOT. An interpretation of ORNL fusion shielding benchmark is presented

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

SENSIT: a cross-section and design sensitivity and uncertainty analysis code

International Nuclear Information System (INIS)

Gerstl, S.A.W.

1980-01-01

SENSIT computes the sensitivity and uncertainty of a calculated integral response (such as a dose rate) due to input cross sections and their uncertainties. Sensitivity profiles are computed for neutron and gamma-ray reaction cross sections of standard multigroup cross section sets and for secondary energy distributions (SEDs) of multigroup scattering matrices. In the design sensitivity mode, SENSIT computes changes in an integral response due to design changes and gives the appropriate sensitivity coefficients. Cross section uncertainty analyses are performed for three types of input data uncertainties: cross-section covariance matrices for pairs of multigroup reaction cross sections, spectral shape uncertainty parameters for secondary energy distributions (integral SED uncertainties), and covariance matrices for energy-dependent response functions. For all three types of data uncertainties SENSIT computes the resulting variance and estimated standard deviation in an integral response of interest, on the basis of generalized perturbation theory. SENSIT attempts to be more comprehensive than earlier sensitivity analysis codes, such as SWANLAKE

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)

Mixed Legendre moments and discrete scattering cross sections for anisotropy representation

International Nuclear Information System (INIS)

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

2012-01-01

This paper deals with the resolution of the integro-differential form of the Boltzmann transport equation for neutron transport in nuclear reactors. In multigroup theory, deterministic codes use transfer cross sections which are expanded on Legendre polynomials. This modelling leads to negative values of the transfer cross section for certain scattering angles, and hence, the multigroup scattering source term is wrongly computed. The first part compares the convergence of 'Legendre-expanded' cross sections with respect to the order used with the method of characteristics (MOC) for Pressurised Water Reactor (PWR) type cells. Furthermore, the cross section is developed using piecewise-constant functions, which better models the multigroup transfer cross section and prevents the occurrence of any negative value for it. The second part focuses on the method of solving the transport equation with the above-mentioned piecewise-constant cross sections for lattice calculations for PWR cells. This expansion thereby constitutes a 'reference' method to compare the conventional Legendre expansion to, and to determine its pertinence when applied to reactor physics calculations. (authors)

Neutron cross section libraries for analysis of fusion neutronics experiments

International Nuclear Information System (INIS)

Kosako, Kazuaki; Oyama, Yukio; Maekawa, Hiroshi; Nakamura, Tomoo

1988-03-01

We have prepared two computer code systems producing neutron cross section libraries to analyse fusion neutronics experiments. First system produces the neutron cross section library in ANISN format, i.e., the multi-group constants in group independent format. This library can be obtained by using the multi-group constant processing code system MACS-N and the ANISN format cross section compiling code CROKAS. Second system is for the continuous energy cross section library for the MCNP code. This library can be obtained by the nuclear data processing system NJOY which generates pointwise energy cross sections and the cross section compiling code MACROS for the MCNP library. In this report, we describe the production procedures for both types of the cross section libraries, and show six libraries with different conditions in ANISN format and a library for the MCNP code. (author)

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

Generation and performance of a multigroup coupled neutron-gamma cross-section library for deterministic and Monte Carlo borehole logging analysis

International Nuclear Information System (INIS)

Kodeli, I.; Aldama, D. L.; De Leege, P. F. A.; Legrady, D.; Hoogenboom, J. E.; Cowan, P.

2004-01-01

As part of the IRTMBA (Improved Radiation Transport Modelling for Borehole Applications) project of the EU community's 5. framework program a special purpose multigroup cross-section library was prepared for use in deterministic and Monte Carlo oil well logging particle transport calculations. This library is expected to improve the prediction of the neutron and gamma spectra at the detector positions of the logging tool, and their use for the interpretation of the neutron logging measurements was studied. Preparation and testing of this library is described. (authors)

Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

Energy Technology Data Exchange (ETDEWEB)

Díez, C.J., E-mail: cj.diez@upm.es [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Cabellos, O. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Instituto de Fusión Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Martínez, J.S. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain)

2015-01-15

Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

International Nuclear Information System (INIS)

Díez, C.J.; Cabellos, O.; Martínez, J.S.

2015-01-01

Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties

Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

Science.gov (United States)

Díez, C. J.; Cabellos, O.; Martínez, J. S.

2015-01-01

Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

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.

Effects of space-dependent cross sections on core physics parameters for compact fast spectrum space power reactors

International Nuclear Information System (INIS)

Lell, R.M.; Hanan, N.A.

1987-01-01

Effects of multigroup neutron cross section generation procedures on core physics parameters for compact fast spectrum reactors have been examined. Homogeneous and space-dependent multigroup cross section sets were generated in 11 and 27 groups for a representative fast reactor core. These cross sections were used to compute various reactor physics parameters for the reference core. Coarse group structure and neglect of space-dependence in the generation procedure resulted in inaccurate computations of reactor flux and power distributions and in significant errors regarding estimates of core reactivity and control system worth. Delayed neutron fraction was insensitive to cross section treatment, and computed reactivity coefficients were only slightly sensitive. However, neutron lifetime was found to be very sensitive to cross section treatment. Deficiencies in multigroup cross sections are reflected in core nuclear design and, consequently, in system mechanical design

FENDL/MG. Library of multigroup cross sections in GENDF and MATXS format for neutron-photon transport calculations. Version 1.1 of March 1995. Summary documentation

International Nuclear Information System (INIS)

Pashchenko, A.B.; Wienke, H.; Ganesan, S.

1996-01-01

Selected neutron reaction nuclear data evaluations 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 GENDF and MATXS format using the NJOY system by R.E. MacFarlane, in VITAMIN-J group structure with VITAMIN-E weighting spectrum. This document summarizes the resulting multigroup data library FENDL/MG version 1.1. The data are available costfree, upon request from the IAEA Nuclear Data Section, online or on magnetic tape. (author). 7 refs, 1 tab

System THEMIS. Cross sections processing system from ENDF/B

Energy Technology Data Exchange (ETDEWEB)

Gonnord, J.

1983-09-01

The THEMIS system allowed to prepare a self punctual and multigroup library for codes solving the TRIPOLI-PROMETHEE transport equation, allowing comparisons with different methods and approximations. The contents of the THEMIS data base was fixed from its use by the PROMETHEE system (punctual Monte Carlo calculations, multigroup calculations, uncertainties analysis and sensitivity studies). The main characteristics of the THEMIS cross section processing system are briefly presented.

SENSIT: a cross-section and design sensitivity and uncertainty analysis code. [In FORTRAN for CDC-7600, IBM 360

Energy Technology Data Exchange (ETDEWEB)

Gerstl, S.A.W.

1980-01-01

SENSIT computes the sensitivity and uncertainty of a calculated integral response (such as a dose rate) due to input cross sections and their uncertainties. Sensitivity profiles are computed for neutron and gamma-ray reaction cross sections of standard multigroup cross section sets and for secondary energy distributions (SEDs) of multigroup scattering matrices. In the design sensitivity mode, SENSIT computes changes in an integral response due to design changes and gives the appropriate sensitivity coefficients. Cross section uncertainty analyses are performed for three types of input data uncertainties: cross-section covariance matrices for pairs of multigroup reaction cross sections, spectral shape uncertainty parameters for secondary energy distributions (integral SED uncertainties), and covariance matrices for energy-dependent response functions. For all three types of data uncertainties SENSIT computes the resulting variance and estimated standard deviation in an integral response of interest, on the basis of generalized perturbation theory. SENSIT attempts to be more comprehensive than earlier sensitivity analysis codes, such as SWANLAKE.

System THEMIS. Cross sections processing system from ENDF/B

International Nuclear Information System (INIS)

Gonnord, J.

1983-09-01

The THEMIS system allowed to prepare a self punctual and multigroup library for codes solving the TRIPOLI-PROMETHEE transport equation, allowing comparisons with different methods and approximations. The contents of the THEMIS data base was fixed from its use by the PROMETHEE system (punctual Monte Carlo calculations, multigroup calculations, uncertainties analysis and sensitivity studies). The main characteristics of the THEMIS cross section processing system are briefly presented [fr

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)

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.)

Process of cross section generation for radiation shielding calculations, using the NJOY code

International Nuclear Information System (INIS)

Ono, S.; Corcuera, R.P.

1986-10-01

The process of multigroup cross sections generation for radiation shielding calculations, using the NJOY code, is explained. Photon production cross sections, processed by the GROUPR module, and photon interaction cross sections processed by the GAMINR are given. These data are compared with the data produced by the AMPX system and published data. (author) [pt

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

AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

Energy Technology Data Exchange (ETDEWEB)

Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

1992-10-01

AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available.

AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

International Nuclear Information System (INIS)

Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

1992-10-01

AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available

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

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

Code implementation of partial-range angular scattering cross sections: GAMMER and MORSE

International Nuclear Information System (INIS)

Ward, J.T. Jr.

1978-01-01

A partial-range (finite-element) method has been previously developed for representing multigroup angular scattering in Monte Carlo photon transport. Computer application of the method, with preliminary quantitative results is discussed here. A multigroup photon cross section processing code, GAMMER, was written which utilized ENDF File 23 point data and the Klein--Nishina formula for Compton scattering. The cross section module of MORSE, along with several execution routines, were rewritten to permit use of the method with photon transport. Both conventional and partial-range techniques were applied for comparison to calculating angular and spectral penetration of 6-MeV photons through a six-inch iron slab. GAMMER was found to run 90% faster than SMUG, with further improvement evident for multiple-media situations; MORSE cross section storage was reduced by one-third; cross section processing, greatly simplified; and execution time, reduced by 15%. Particle penetration was clearly more forward peaked, as moment accuracy is retained to extremly high order. This method of cross section treatment offers potential savings in both storage and handling, as well as improved accuracy and running time in the actual execution phase. 3 figures, 4 tables

MIRANDA - a module based on multiregion resonance theory for generating cross sections within the AUS neutronics code system

International Nuclear Information System (INIS)

Robinson, G.S.

1985-12-01

MIRANDA is the cross-section generation module of the AUS neutronics code system used to prepare multigroup cross-section data which are pertinent to a particular study from a general purpose multigroup library of cross sections. Libraries have been prepared from ENDF/B which are suitable for thermal and fast fission reactors and for fusion blanket studies. The libraries include temperature dependent data, resonance cross sections represented by subgroup parameters and may contain photon as well as neutron data. The MIRANDA module includes a multiregion resonance calculation in slab, cylinder or cluster geometry, a homogeneous B L flux solution, and a group condensation facility. This report documents the modifications to an earlier version of MIRANDA and provides a complete user's manual

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)

New Tools to Prepare ACE Cross-section Files for MCNP Analytic Test Problems

International Nuclear Information System (INIS)

Brown, Forrest B.

2016-01-01

Monte Carlo calculations using one-group cross sections, multigroup cross sections, or simple continuous energy cross sections are often used to: (1) verify production codes against known analytical solutions, (2) verify new methods and algorithms that do not involve detailed collision physics, (3) compare Monte Carlo calculation methods with deterministic methods, and (4) teach fundamentals to students. In this work we describe 2 new tools for preparing the ACE cross-section files to be used by MCNP ® for these analytic test problems, simple a ce.pl and simple a ce m g.pl.

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.

MCNP and MATXS cross section libraries based on JENDL-3.3

International Nuclear Information System (INIS)

Kosako, Kazuaki; Konno, Chikara; Fukahori, Tokio; Shibata, Keiichi

2003-01-01

The continuous energy cross section library for the Monte Carlo transport code MCNP-4C, FSXLIB-J33, has been generated from the latest version of JENDL-3.3. The multigroup cross section library with the MATXS format, MATXS-J33, has been generated also from JENDL-3.3. Both libraries contain all nuclides in JENDL-3.3 and are processed at 300 K by the nuclear data processing system NJOY99. (author)

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

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.

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)

Evaluation of cross sections of Th-232 and U-233

International Nuclear Information System (INIS)

Dias, A.M.

1978-01-01

The cross sections in multigroups of Th-232 and U-233 are evaluated by comparison of theoretical results and experimental data obtained through experiments on the fast reactors IBR-I, EBR-II, BR-I and AETR. The deviation between calculated values and experimental results is about 10%. They are therefore satisfatory for neutronic calculations [pt

BUGLE-93 (ENDF/B-VI) cross-section library data testing using shielding benchmarks

International Nuclear Information System (INIS)

Hunter, H.T.; Slater, C.O.; White, J.E.

1994-01-01

Several integral shielding benchmarks were selected to perform data testing for new multigroup cross-section libraries compiled from the ENDF/B-VI data for light water reactor (LWR) shielding and dosimetry. The new multigroup libraries, BUGLE-93 and VITAMIN-B6, were studied to establish their reliability and response to the benchmark measurements by use of radiation transport codes, ANISN and DORT. Also, direct comparisons of BUGLE-93 and VITAMIN-B6 to BUGLE-80 (ENDF/B-IV) and VITAMIN-E (ENDF/B-V) were performed. Some benchmarks involved the nuclides used in LWR shielding and dosimetry applications, and some were sensitive specific nuclear data, i.e. iron due to its dominant use in nuclear reactor systems and complex set of cross-section resonances. Five shielding benchmarks (four experimental and one calculational) are described and results are presented

Shielding Factor Method for producing effective cross sections: MINX/SPHINX and the CCCC interface system

International Nuclear Information System (INIS)

MacFarlane, R.E.; Weisbin, C.R.; Paik, N.C.

1978-01-01

The Shielding Factor Method (SFM) is an economical designer-oriented method for producing the coarse-group space and energy self-shielded cross sections needed for reactor-core analysis. Extensive experience with the ETOX/1DX and ENDRUN/TDOWN systems has made the SFM the method of choice for most US fast-reactor design activities. The MINX/SPHINX system was designed to expand upon the capabilities of the older SFM codes and to incorporate the new standard interfaces for fast-reactor cross sections specified by the Committee for Computer Code Coordination (CCCC). MINX is the cross-section processor. It generates multigroup cross sections, shielding factors, and group-to-group transfer matriccs from ENDF/B-IV and writes them out as CCCC ISOTXS and BRKOXS files. It features detailed pointwise resonance reconstruction, accurate Doppler broadening, and an efficient treatment of anisotropic scattering. SPHINX is the space-and-energy shielding code. It uses specific mixture and geometry information together with equivalence principles to construct shielded macroscopic multigroup cross sections in as many as 240 groups. It then makes a flux calculation by diffusion or transport methods and collapses to an appropriate set of cell-averaged coarse-group effective cross sections. The integration of MINX and SPHINX with the CCCC interface system provides an efficient, accurate, and convenient system for producing effective cross sections for use in fast-reactor problems. The system has also proved useful in shielding and CTR applications. 3 figures, 4 tables

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.

Sensitivity of LWR fuel cycle costs to uncertainties in detailed thermal cross sections

International Nuclear Information System (INIS)

Ryskamp, J.M.; Becker, M.; Harris, D.R.

1979-01-01

Cross sections averaged over the thermal energy (< 1 or 2 eV) group have been shown to have an important economic role for light-water reactors. Cost implications of thermal cross section uncertainties at the few-group level were reported earlier. When it has been determined that costs are sensitive to a specific thermal-group cross section, it becomes desirable to determine how specific energy-dependent cross sections influence fuel cycle costs. Multigroup cross-section sensitivity coefficients vary with fuel exposure. By changing the shape of a cross section displayed on a view-tube through an interactive graphics system, one can compute the change in few-group cross section using the exposure dependent sensitivity coefficients. With the changed exposure dependent few-group cross section, a new fuel cycle cost is computed by a sequence of batch depletion, core analysis, and fuel batch cost code modules. Fuel cycle costs are generally most sensitive to cross section uncertainties near the peak of the hardened Maxwellian flux

FCXSEC: multigroup cross-section libraries for nuclear fuel cycle shielding calculations

International Nuclear Information System (INIS)

Ford, W.E. III; Webster, C.C.; Diggs, B.R.; Pevey, R.E.; Croff, A.G.

1980-05-01

Starting with the pseudo-composition-independent VITAMIN-C cross-sectin library, composition-dependent fine-(171n-36γ) and broad-group (22n-21γ) self-shielded AMPX master, broad-group microscopic ANISN-formatted, and broad-group macroscopic ANISN-formatted cross-section libraries were generated to be used for nuclear fuel cycle shielding calculations. The specifications for the data and the procedure used to prepare the libraries are described

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

Formats and processing of evaluated nuclear data into multigroup cross-sections

International Nuclear Information System (INIS)

Motta, M.

1984-01-01

The first part of these lectures concerns the data in nuclear files and their manipulation. The structure of the data files as divided into the resonance region (subdivided into the resolved and the unresolved regions) and the continuum region is presented. The reactions concerned are the elastic scattering; the radiative capture and the fission methods for averaging the cross sections are given. Then, the group averaging formulas and the self-shielding factors are presented in some detail. The second part concerns a presentation of nuclear data files handling and conversion. The main libraries are listed and several maintenance computer codes presented. The way the conversion among different files is handled is also presented. The listings of several BASIC programs for different cross section calculations are given. These codes are self-guided

12G: code for conversion of isotope-ordered cross-section libraries into group-ordered cross-section libraries

International Nuclear Information System (INIS)

Resnik, W.M. II; Bosler, G.E.

1977-09-01

Many current reactor physics codes accept cross-section libraries in an isotope-ordered form, convert them with internal preprocessing routines to a group-ordered form, and then perform calculations using these group-ordered data. Occasionally, because of storage and time limitations, the preprocessing routines in these codes cannot convert very large multigroup isotope-ordered libraries. For this reason, the I2G code, i.e., ISOTXS to GRUPXS, was written to convert externally isotope-ordered cross section libraries in the standard file format called ISOTXS to group-ordered libraries in the standard format called GRUPXS. This code uses standardized multilevel data management routines which establish a strategy for the efficient conversion of large libraries. The I2G code is exportable contingent on access to, and an intimate familiarization with, the multilevel routines. These routines are machine dependent, and therefore must be provided by the importing facility. 6 figures, 3 tables

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.

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.

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

Implementation of the rapid cross section adjustment approach at General Electric

International Nuclear Information System (INIS)

Cowan, C.L.; Kujawski, E.; Protsik, R.

1978-01-01

The General Electric rapid cross section adjustment approach was developed to use the shielding factor method for formulating multigroup cross sections. In this approach, space- and composition-dependent cross sections for a particular reactor or shield design are prepared from a generalized cross section library by the use of resonance self-shielding factors, and by the adjustment of elastic scattering cross sections for the local neutron flux spectra. The principal tool in the cross section adjustment package is the data processing code TDOWN. This code was specified to give the user a high degree of flexibility in the analysis of advanced reactor designs. Of particular interest in the analysis of critical experiments is the ability to carry out cell heterogeneity self-shielding calculations using a multiregion equivalence relationship, and the homogenization of the cross sections over the specified cell with the flux weighting obtained from transport theory calculations. Extensive testing of the rapid cross section adjustment approach, including comparisons with Monte Carlo methods, indicated that this approach can be utilized with a high degree of confidence in the design analysis of complex fast reactor systems. 2 figures, 1 table

ZZ BOREHOLE-EB6.8-MG, multi group cross-section library for deterministic and Monte Carlo codes

International Nuclear Information System (INIS)

Kodeli, Ivo; Aldama, Daniel L.; Leege, Piet F.A. de; Legrady, David; Hoogenboom, J. Eduard

2007-01-01

1 - Description: Format: MATXS and ACE; Number of groups: 175 neutron, 45 gamma-ray; Nuclides: H-1, C-12, O-16, Na-23, Mg-nat, Al-27, Si-28, -29, -30, S-nat, Cl-35, -37, K-nat, Ca-nat, Mn-55, Fe-54, -56, -57, -58, I-127, W-nat. Origin: ENDF/B-VI.8; Weighting spectrum: Fission and fusion peak at high energies and a 1/E + thermal Maxwellian extension at low energies. The following materials/nuclides are included in the library: H-1, C-12, O-16, Na-23, Mg-nat, Al-27, Si-28, -29, -30, S-nat, Cl-35, -37, K-nat, Ca-nat, Fe-54, -56, -57, -58, Mn-55, I-127, W-nat. ZZ-BOREHOLE-EB6.8-MG is a multigroup cross section library for deterministic (DOORS, DANTSYS) and Monte Carlo (MCNP) transport codes developed for the oil well logging applications. The library is based on the ENDF/B-VI.8 evaluation and was processed by the NJOY-99 code. The cross sections are given in the 175 neutron and 45 gamma ray group structure. The MATXS format library can be directly used in TRANSX code to prepare the multigroup self-shielded cross sections for deterministic discrete ordinates codes like DOORS and DANTSYS. The data provided in the GROUPR and GAMINR format were converted to the MCNP ACE format by the NSLINK, SCALE and CRSRD codes. IAEA1398/03: Multigroup cross section data for Mn-55 were added in TRANSX format

SCAMPI: A code package for cross-section processing

International Nuclear Information System (INIS)

Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.

1996-01-01

The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis

SCAMPI: A code package for cross-section processing

Energy Technology Data Exchange (ETDEWEB)

Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.

1996-04-01

The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis.

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

Development of Indian cross section data files for Th-232 and U-233 and integral validation studies

International Nuclear Information System (INIS)

Ganesan, S.

1988-01-01

This paper presents an overview of the tasks performed towards the development of Indian cross section data files for Th-232 and U-233. Discrepancies in various neutron induced reaction cross sections in various available evaluated data files have been obtained by processing the basic data into multigroup form and intercomparison of the latter. Interesting results of integral validation studies for capture, fission and (n,2n) cross sections for Th-232 by analyses of selected integral measurements are presented. In the resonance range, energy regions where significant differences in the calculated self-shielding factors for Th-232 occur have been identified by a comparison of self-shielded multigroup cross sections derived from two recent evaluated data files, viz., ENDF/B-V (Rev.2) and JENDL-2, for several dilutions and temperatures. For U-233, the three different basic data files ENDF/B-IV, JENDL-2 and ENDL-84 were intercompared. Interesting observations on the predictional capability of these files for the criticality of the spherical metal U-233 system are given. The current status of Indian data file is presented. (author) 62 ref

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.

Status of standard cross section library and future plan

International Nuclear Information System (INIS)

Zukeran, Atsushi

2001-01-01

JSSTDL-300 multi-group cross section library with 300 neutron energy groups coupled with 104 group γ-ray cross sections was developed for general users in nuclear reactor physics and/or design, whose source data is the evaluated nuclear data library JENDL-3.2. For the purpose of a standard or common use, several famous cross section libraries worldwide used, i.e., ABBN-25, GAM-123, VITAMIN-C/J(E+C), MGCL-137, BERMUDA-12 and FNS-125 for neutron, and LANL-12, -24-, -48, and CSEWG-94 for γ-ray, are consulted about setting the common energy group structure. Furthermore, in order to expand the applicability, the top energy is set on 20 MeV and the lowest energy is 10 -5 eV. In the thermal neutron energy region, the JSSTDL-300 has about 20 energy groups. Besides, many utility codes for group collapsing and for data format transformation are provided for general users. (author)

Integral-capture measurements and cross-section adjustments for Nd, Sm, and Eu

International Nuclear Information System (INIS)

Anderl, R.A.; Schmittroth, F.; Harker, Y.D.

1981-07-01

Integral-capture reaction rates are reported for 143 Nd, 144 Nd, 145 Nd, 147 Sm, 151 Eu, 152 Eu, 153 Eu, and 154 Eu irradiated in different neutron spectra in EBR-II. These reaction rates are based primarily on mass-spectrometric measurements of the isotopic atom ratios of the capture product to the target nuclide. The neutron spectra are characterized using passive neutron dosimetry and spectrum-unfolding with the FERRET least-squares data analysis code. Reaction rates for the neutron spectrum monitors were determined by the radiometric technique using Ge(Li) spectrometers. These rates are also reported here. The integral data for the rare-earth samples and for the spectrum monitors were used in multigroup flux/cross-section adtustment analyses with FERRET to generate adjustments to 47 group representations of the ENDF/B-IV capture cross sections for the rare-earth isotopes. These adjusted cross sections are in good agreement with recent differential data and with adjusted cross sections based on STEK integral data. Examples are given of the use of the adjusted cross sections and covariance matrices for cross-section evaluation

Annotated bibliography covering generation and use of evaluated cross section uncertainty files

International Nuclear Information System (INIS)

Peelle, R.W.; Burrows, T.W.

1983-03-01

Literature references related to definition, generation, and use of evaluated cross section uncertainty (variance-covariance) files are listed with comments intended primarily to guide the reader toward materials of immediate interest. Papers are also cited that cover covariance information for individual experiments and that relate to production and use of multigroup covariance matrices. Titles are divided among several major categories

Generation of broad-group neutron/photon cross-section libraries for shielding applications

International Nuclear Information System (INIS)

Ingersoll, D.T.; Roussin, R.W.; Fu, C.Y.; White, J.E.

1989-01-01

The generation and use of multigroup cross-section libraries with broad energy group structures is primarily for the economy of computer resources. Also, the establishment of reference broad-group libraries is desirable in order to avoid duplication of effort, both in terms of the data generation and verification, and to assure a common data base for all participants in a specific project. Uncertainties are inevitably introduced into the broad-group cross sections due to approximations in the grouping procedure. The dominant uncertainty is generally with regard to the energy weighting function used to average the pointwise or fine-group data within a single broad group. Intelligent choice of the weighting functions can reduce such uncertainties. Also, judicious selection of the energy group structure can help to reduce the sensitivity of the computed responses to the weighting function, at least for a selected set of problems. Two new multigroup cross section libraries have been recently generated from ENDF/B-V data for two specific shielding applications. The first library was prepared for use in sodium-cooled reactor systems and is available in both broad-group structures. The second library, just recently completed, was prepared for use in air-over-ground environments and is available in a broad-group (46-neutron, 23-photon) energy structure. The selection of the specific group structures and weighting functions was an important part of the generation of both libraries

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)

TEMPEST-2, Thermalization Program for Neutron Spectra and Multigroup Cross-Sections

International Nuclear Information System (INIS)

Gowins, G.

1984-01-01

Description of problem or function: TEMPEST2 is a neutron thermalization program based upon the Wigner-Wilkins approximation for light moderators and the Wilkins approximation for heavy moderators. A Maxwellian distribution may also be used. The model used may be selected as a function of energy. The second-order differential equations are integrated directly rather than transformed to the Riccati equation. The program provides microscopic and macroscopic cross-section averages over the thermal neutron spectrum

Nuclear cross section library for oil well logging analysis

International Nuclear Information System (INIS)

Kodeli, I.; Kitsos, S.; Aldama, D.L.; Zefran, B.

2003-01-01

As part of the IRTMBA (Improved Radiation Transport Modelling for Borehole Applications) Project of the EU Community's 5 th Programme a special purpose multigroup cross section library to be used in the deterministic (as well as Monte Carlo) oil well logging particle transport calculations was prepared. This library is expected to improve the prediction of the neutron and gamma spectra at the detector positions of the logging tool, and their use for the interpretation of the neutron logging measurements was studied. Preparation and testing of this library is described. (author)

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.

Qualification of the calculational methods of the fluence in the pressurised water reactors. Improvement of the cross sections treatment by the probability table method; Qualification des methodes de calculs de fluence dans les reacteurs a eau pressurisee. Amelioration du traitement des sections efficaces par la methode des tables de probabilite

Energy Technology Data Exchange (ETDEWEB)

Zheng, S H

1994-01-01

It is indispensable to know the fluence on the nuclear reactor pressure vessel. The cross sections and their treatment have an important rule to this problem. In this study, two ``benchmarks`` have been interpreted by the Monte Carlo transport program TRIPOLI to qualify the calculational method and the cross sections used in the calculations. For the treatment of the cross sections, the multigroup method is usually used but it exists some problems such as the difficulty to choose the weighting function and the necessity of a great number of energy to represent well the cross section`s fluctuation. In this thesis, we propose a new method called ``Probability Table Method`` to treat the neutron cross sections. For the qualification, a program of the simulation of neutron transport by the Monte Carlo method in one dimension has been written; the comparison of multigroup`s results and probability table`s results shows the advantages of this new method. The probability table has also been introduced in the TRIPOLI program; the calculational results of the iron deep penetration benchmark has been improved by comparing with the experimental results. So it is interest to use this new method in the shielding and neutronic calculation. (author). 42 refs., 109 figs., 36 tabs.

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

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)

Cross section recondensation method via generalized energy condensation theory

International Nuclear Information System (INIS)

Douglass, Steven; Rahnema, Farzad

2011-01-01

Highlights: → A new method is presented which corrects for core environment error from specular boundaries at the lattice cell level. → Solution obtained with generalized energy condensation provides improved approximation to the core level fine-group flux. → Iterative recondensation of the cross sections and unfolding of the flux provides on-the-fly updating of the core cross sections. → Precomputation of energy integrals and fine-group cross sections allows for easy implementation and efficient solution. → Method has been implemented in 1D and shown to correct the environment error, particularly in strongly heterogeneous cores. - Abstract: The standard multigroup method used in whole-core reactor analysis relies on energy condensed (coarse-group) cross sections generated from single lattice cell calculations, typically with specular reflective boundary conditions. Because these boundary conditions are an approximation and not representative of the core environment for that lattice, an error is introduced in the core solution (both eigenvalue and flux). As current and next generation reactors trend toward increasing assembly and core heterogeneity, this error becomes more significant. The method presented here corrects for this error by generating updated coarse-group cross sections on-the-fly within whole-core reactor calculations without resorting to additional cell calculations. In this paper, the fine-group core flux is unfolded by making use of the recently published Generalized Condensation Theory and the cross sections are recondensed at the whole-core level. By iteratively performing this recondensation, an improved core solution is found in which the core-environment has been fully taken into account. This recondensation method is both easy to implement and computationally very efficient because it requires precomputation and storage of only the energy integrals and fine-group cross sections. In this work, the theoretical basis and development

FENDL2/A-MCNP, FENDL2/A-VITJE and FENDL2/A-VITJFLAT. The processed FENDL-2 neutron activation cross-section data files. Summary documentation

International Nuclear Information System (INIS)

Pashchenko, A.B.; Wienke, H.

1997-01-01

This document summarizes the libraries of neutron activation cross-section data processed into the following three formats: continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP4A; VITAMIN-J 175 multigroup format weighted with the VITAMIN-E weighting spectrum as used by the transmutation codes REAC*2/3 and FOUR ACES; VITAMIN-J 175 multigroup ENDF-6 format, with a flat weighting spectrum. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape. (author)

Evaluation of ETOG-3Q/ETOG-3, FLANGE-II, XLACS, NJOY and linear/recent/groupie codes for calculations of resonance and reference cross sections

International Nuclear Information System (INIS)

Anaf, J.; Chalhoub, E.S.

1991-01-01

The NJOY and LINEAR/RECENT/GROUPIE calculational procedures for the resolved and unresolved resonance contributions and background cross sections are evaluated. Elastic scattering, fission and capture multigroup cross sections generated by these codes and the previously validated ETOG-3Q, ETOG-3, FLANGE-II and XLACS are compared. Constant weighting function and zero Kelvin temperature are considered. Discrepancies are presented and analyzed. (author)

Qualification of the calculational methods of the fluence in the pressurised water reactors. Improvement of the cross sections treatment by the probability table method

International Nuclear Information System (INIS)

Zheng, S.H.

1994-01-01

It is indispensable to know the fluence on the nuclear reactor pressure vessel. The cross sections and their treatment have an important rule to this problem. In this study, two ''benchmarks'' have been interpreted by the Monte Carlo transport program TRIPOLI to qualify the calculational method and the cross sections used in the calculations. For the treatment of the cross sections, the multigroup method is usually used but it exists some problems such as the difficulty to choose the weighting function and the necessity of a great number of energy to represent well the cross section's fluctuation. In this thesis, we propose a new method called ''Probability Table Method'' to treat the neutron cross sections. For the qualification, a program of the simulation of neutron transport by the Monte Carlo method in one dimension has been written; the comparison of multigroup's results and probability table's results shows the advantages of this new method. The probability table has also been introduced in the TRIPOLI program; the calculational results of the iron deep penetration benchmark has been improved by comparing with the experimental results. So it is interest to use this new method in the shielding and neutronic calculation. (author). 42 refs., 109 figs., 36 tabs

COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

Energy Technology Data Exchange (ETDEWEB)

Woo Y. Yoon; David W. Nigg

2008-09-01

COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete

COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

International Nuclear Information System (INIS)

Yoon, Woo Y.; Nigg, David W.

2008-01-01

COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete

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)

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

ZZ AIRFEWG, Gamma, Neutron Transport Calculation in Air Using FEWG1 Cross-Section

International Nuclear Information System (INIS)

1985-01-01

1 - Description of program or function: Format: ANISN; Number of groups: 37 neutron / 21 gamma-ray; Nuclides: air (79% N and 21% O); Origin: DLC-0031/FEWG1 cross sections (ENDF/B-IV). Weighting spectrum: 1/E. The AIRFEWG library has been generated by an ANISN multigroup calculation of gamma-ray, neutron, and secondary gamma-ray transport in infinite homogeneous air using DLC-0031/FEWG1 cross sections. 2 - Method of solution: The results were generated with a P3, ANISN run with a source in a single energy group. Thus, 58 such runs were required. For sources in the 37 neutron groups, both neutron and secondary gamma-ray fluence results were calculated. For gamma-ray sources only gamma-ray fluences were calculated

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.

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

PUFF-IV, Code System to Generate Multigroup Covariance Matrices from ENDF/B-VI Uncertainty Files

International Nuclear Information System (INIS)

2007-01-01

1 - Description of program or function: The PUFF-IV code system processes ENDF/B-VI formatted nuclear cross section covariance data into multigroup covariance matrices. PUFF-IV is the newest release in this series of codes used to process ENDF uncertainty information and to generate the desired multi-group correlation matrix for the evaluation of interest. This version includes corrections and enhancements over previous versions. It is written in Fortran 90 and allows for a more modular design, thus facilitating future upgrades. PUFF-IV enhances support for resonance parameter covariance formats described in the ENDF standard and now handles almost all resonance parameter covariance information in the resolved region, with the exception of the long range covariance sub-subsections. PUFF-IV is normally used in conjunction with an AMPX master library containing group averaged cross section data. Two utility modules are included in this package to facilitate the data interface. The module SMILER allows one to use NJOY generated GENDF files containing group averaged cross section data in conjunction with PUFF-IV. The module COVCOMP allows one to compare two files written in COVERX format. 2 - Methods: Cross section and flux values on a 'super energy grid,' consisting of the union of the required energy group structure and the energy data points in the ENDF/B-V file, are interpolated from the input cross sections and fluxes. Covariance matrices are calculated for this grid and then collapsed to the required group structure. 3 - Restrictions on the complexity of the problem: PUFF-IV cannot process covariance information for energy and angular distributions of secondary particles. PUFF-IV does not process covariance information in Files 34 and 35; nor does it process covariance information in File 40. These new formats will be addressed in a future version of PUFF

TIMS-1, Multigroup Cross-Sections of Heavy Isotope Mixture with Resonance from ENDF/B

International Nuclear Information System (INIS)

Takano, Hideki; Ishiguro, Yukio; Matsui, Yasushi

1984-01-01

1 - Description of problem or function: TIMS-1 is a code for calculating the group constants of heavy resonant nuclei by using ENDF/ B-4 format data. This code calculates infinitely dilute cross sections and self-shielding factors as a function of composition sigma-0 temperature T and R-parameter, where R is the ratio of ato- mic number density of two different resonant nuclei. 2 - Method of solution: In the unresolved resonance region, a ladder of resonance parameters and levels is generated with Monte Carlo method. The temperature dependent cross sections are calculated with the Breit-Wigner single-level and multi-level formula. The neutron spectrum is accurately calculated by solving numerically the neutron slowing down equation using a recurrence formula for neutron slowing down source. 3 - Restrictions on the complexity of the problem: The maximum numbers of energy groups, temperatures and compositions are 60, 4 and 10 respectively

Mechanized evaluation of neutron cross-sections

International Nuclear Information System (INIS)

Horsley, A.; Parker, J.B.

1967-01-01

The evaluation work to provide accurate and consistent neutron cross-section data for multigroup neutronics calculations is not fully exploiting the available theoretical and experimental results; this has been so particularly since the introduction of on-line data handling techniques enabled experimenters to turn out vast quantities of numbers. This situation can be radically improved only by mechanizing the evaluation processes. Systems such as the SC1SRS tape will not only largely overcome the task of collecting data but will provide speedy access to it; by using computers and graph-plotting machines to tabulate and display this data, the labour of evaluation can be very greatly reduced. With some types of cross-section there is hope that by using modern curve-fitting techniques the actual evaluation and statistical accounting of the data can be performed automatically. Some areas where automatic evaluation would seem likely to succeed are specified and a discussion of the mathematical difficulties incurred, such as the elimination of anomalous data, is given. Particularly promising is the use of splines in the mechanized evaluation of data. Splines are the mathematical analogues of the draughtsman's spline used in drawing smooth curves. Their principal properties are the excellent approximations they give to the derivatives of a function; in contrast to conventional polynomial fitting, this feature ensures good interpolation and, when required, stable extrapolation. Various methods of using splines in data graduation and the problem of marrying these methods to standard statistical procedures are examined. The results of work done at AWRE with cubic splines on the mechanized evaluation of neutron scattering total cross-section and angular distribution data are presented. (author)

Preparation and benchmarking of ANSL-V cross sections for advanced neutron source reactor studies

International Nuclear Information System (INIS)

Arwood, J.W.; Ford, W.E. III; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.

1987-01-01

Research and development for the advanced neutron source (ANS) reactor is being funded by the US Dept. of Energy. This reactor is to provide the world's most intense steady-state source of low-energy neutrons for a national experimental user facility. Pseudo-problem-independent, multigroup cross-section libraries were generated to support ANS design work. The libraries, designated ANSL-V, are data bases in AMPX master format for subsequent generation of problem-dependent cross sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, and MORSE. Included in ANSL-V are 123-material P 3 neutron, 46-material P 0 or P 6 secondary gamma-ray production (SGRP), and 34-material P 6 gamma-ray interaction (GRI) libraries

Evaluation of ETOG-3Q, ETOG-3, FLANGE-II, XLACS, NJOY and LINEAR/RECENT/GROUPIE computer codes concerning to the resonance contribution and background cross sections

International Nuclear Information System (INIS)

Anaf, J.; Chalhoub, E.S.

1988-12-01

The NJOY and LINEAR/RECENT/GROUPIE calculational procedures for the resolved and unresolved resonance contributions and background cross sections are evaluated. Elastic scattering, fission and capture multigroup cross sections generated by these codes and the previously validated ETOG-3Q, ETOG-3, FLANGE-II and XLACS are compared. Constant weighting function and zero Kelvin temperature are considered. Discrepancies are presented and analysed. (author) [pt

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

COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

Energy Technology Data Exchange (ETDEWEB)

Woo Y. Yoon; David W. Nigg

2009-08-01

COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete

COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

International Nuclear Information System (INIS)

Yoon, Woo Y.; Nigg, David W.

2009-01-01

COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete

COMBINE/PC - a portable neutron spectrum and cross-section generation program

International Nuclear Information System (INIS)

Nigg, D.W.; Grimesey, R.A.; Curtis, R.L.

1990-01-01

Use of personal computers and engineering workstations for complex scientific computations has expanded rapidly in the past few years. This trend is expected to continue in the future with the introduction of increasingly sophisticated microprocessors and microcomputer systems. In response to this, an integrated system of neutronics and radiation transport software suitable for operation in an IBM personal computer (PC)-class environment has been under development at the Idaho National Engineering Laboratory (INEL) for the past 3 years. A key component of this system will be module to produce application-specific multigroup cross-section libraries that can be used in various neutron transport and diffusion theory code modules. This software module, referred to as COMBINE/PC, was recently completed at INEL and is the subject of this paper. COMBINE/PC was developed to provide an ENDF/B-based neutron cross-section generation capability of sufficient sophistication to handle a wide variety of practical fission and fusion-related applications while maintaining a compact machine-independent structure

ZZ DLC-11 RITTS, 121-Group Coupled Cross-Section for ANISN, DOT, MORSE

International Nuclear Information System (INIS)

1970-01-01

A - Nature of physical problem solved: Format: ANISN, DTF-4, DOT and MORSE. Number of groups: 100 neutron energy groups (14.92 MeV to thermal) 21 gamma-ray energy groups (14.0 to 0.01 MeV) Nuclides: H, C, O, N, Na, Mg, P, S, Cl, K, and Ca, (microscopic cross sections) and 9 organic materials including 11-element standard man, 4-element standard man, skin, bone, tissue, brain, lung, red marrow, and muscle (macroscopic cross sections). Origin: ENDF/B for H, C, N, O, Na, and Mg; O5R library for Ca, S, and K; GAM-2 library for Cl; Evaluation by J.J. Ritts for P. Weighting spectrum: 1/E for the top 99 groups and Maxwellian for the thermal group values. DLC-11 data is suitable for neutron, gamma-ray, or coupled neutron and gamma-ray transport calculations. It is intended for use in multigroup discrete ordinates or Monte Carlo transport codes which treat anisotropic scattering by Legendre expansion up to order P3. DLC-11 is a collection of multigroup cross section data which were compiled by J. J. Ritts for use in depth-dose calculations in anthropomorphic phantoms. For convenience the data are grouped as follows - 1. A coupled 121-group (100 neutron, 21 gamma-ray) set of data for the 11 elements H, C, O, N. Na, Mg, P, S, Cl, K, and Ca. This set includes P3 coupled 121-group microscopic cross sections plus 121-group kerma factors for the 11 elements. 2. A 100-group set of neutron cross sections for the 11 elements. 3. A coupled 121-group set of macroscopic cross sections for 9 organic materials including 11-element standard man, 4-element standard man, skin, bone, tissue, brain, lung, red marrow, and muscle. B - Method of solution: The basic data sources were ENDF/B for H, C, N, O, Na, and Mg, the O5R library for Ca, S, and K, the GAM-2 library for Cl and an evaluation by Ritts for P. A 1/E spectrum was assumed for averaging the top 99 groups and a Maxwellian for averaging the thermal group values. The gamma-ray cross sections were computed from DLC-3/HPIC using MUG. The

Performance of JEF2.2 based continuous energy cross sections in predicting the multiplication factor of critical systems

International Nuclear Information System (INIS)

John, T.M.; de Leege, P.F.A.; Hoogenboom, J.E.

1996-01-01

The continuous energy representation of cross sections for neutronics calculations avoids the requirement of resonance self shielding and the assumptions about the neutron spectrum used for weighing cross sections, required in the preparation of a multigroup cross sections library. The cross sections library prepared for a particular temperature of the nuclide is valid irrespective of the environment of the nuclide and can be used in calculations for many types of reactors. It is comparatively easier to incorporate them in Monte Carlo simulation of neutron transport. The Monte Carlo code MCNP is capable of using a continuous energy representation of nuclear cross sections in simulation of neutron or photon transport. The ACER module of NJOY is able to generate the continuous energy cross section of any nuclide in a format that can be used by MCNP, from any evaluated data file in ENDF/B format. Continuous energy cross sections prepared from the evaluated data file JEF2.2 was used to analyse some standard critical benchmarks and also the critical configuration of the HOR, a 2 MW research reactor at Delft, the Netherlands. Results show that continuous energy cross sections prepared from JEF2.2 evaluated file predicts the multiplication factor of critical systems very close to unity. (author). 6 refs., 2 tabs., 1 fig

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

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

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)

Calculated neutron-activation cross sections for E/sub n/ /le/ 100 MeV for a range of accelerator materials

International Nuclear Information System (INIS)

Bozoian, M.; Arthur, E.D.; Perry, R.T.; Wilson, W.B.; Young, P.G.

1988-01-01

Activation problems associated with particle accelerators are commonly dominated by reactions of secondary neutrons produced in reactions of beam particles with accelerator or beam stop materials. Measured values of neutron-activation cross sections above a few MeV are sparse. Calculations with the GNASH code have been made for neutrons incident on all stable nuclides of a range of elements common to accelerator materials. These elements include B, C, N, O, Ne, Mg, Al, Si, P, S, Ar, K, Ca, Cr, Mn, Fe, Co, Ni, Cu, Zn, Zr, Mo, Nd, and Sm. Calculations were made for a grid of incident neutron energies extending to 100 MeV. Cross sections leading to the direct production of as many as 87 activation products for each of 84 target nuclide were tabulated on this grid of neutron energies, each beginning with the threshold for the product nuclide's formation. Multigrouped values of these cross sections have been calculated and are being integrated into the cross-section library of the REAC-2 neutron activation code. Illustrative cross sections are presented. 20 refs., 6 figs., 1 tab

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)

MVP/GMVP II, MC Codes for Neutron and Photon Transport Calc. based on Continuous Energy and Multigroup Methods

International Nuclear Information System (INIS)

2005-01-01

A - Description of program or function: (1) Problems to be solved: MVP/GMVP can solve eigenvalue and fixed-source problems. The multigroup code GMVP can solve forward and adjoint problems for neutron, photon and neutron-photon coupled transport. The continuous-energy code MVP can solve only the forward problems. Both codes can also perform time-dependent calculations. (2) Geometry description: MVP/GMVP employs combinatorial geometry to describe the calculation geometry. It describes spatial regions by the combination of the 3-dimensional objects (BODIes). Currently, the following objects (BODIes) can be used. - BODIes with linear surfaces: half space, parallelepiped, right parallelepiped, wedge, right hexagonal prism; - BODIes with quadratic surface and linear surfaces: cylinder, sphere, truncated right cone, truncated elliptic cone, ellipsoid by rotation, general ellipsoid; - Arbitrary quadratic surface and torus. The rectangular and hexagonal lattice geometry can be used to describe the repeated geometry. Furthermore, the statistical geometry model is available to treat coated fuel particles or pebbles for high temperature reactors. (3) Particle sources: The various forms of energy-, angle-, space- and time-dependent distribution functions can be specified. (4) Cross sections: The ANISN-type PL cross sections or the double-differential cross sections can be used in the multigroup code GMVP. On the other hand, the specific cross section libraries are used in the continuous-energy code MVP. The libraries are generated from the evaluated nuclear data (JENDL-3.3, ENDF/B-VI, JEF-3.0 etc.) by using the LICEM code. The neutron cross sections in the unresolved resonance region are described by the probability table method. The neutron cross sections at arbitrary temperatures are available for MVP by just specifying the temperatures in the input data. (5) Boundary conditions: Vacuum, perfect reflective, isotropic reflective (white), periodic boundary conditions can be

Re-evaluation of the neutron scattering dynamics in heavy water, generation of multigroup cross sections for THERM-126

International Nuclear Information System (INIS)

Keinert, J.

1982-06-01

In providing THERM-126 with cross section matrices for deuterium bound in heavy water the IKE phonon spectrum was reevaluated. The changes are modifications in the acoustic part and in the frequency of the second oscillator. Contrary to the phonon spectrum model for D in D 2 O in ENDF/B-IV the broad band of hindered rotations is assumed to be temperature dependent taking into account the diffusive motion of the molecule. With the new model scattering law data S (α, β) are generated in the temperature range 293.6 K-673.6 K. The THERM-126 scattering cross section matrices are calculated up to P 3 . As a validity check a lot of differential and integral cross sections are compared to experiments and benchmarks are recalculated. (orig.) [de

A punched-card library of neutron cross-sections and its use in the mechanized preparation of group cross-sections for use in Monte Carlo, Carlson S{sub n} and other multi-group neutronics calculations on high-speed computers

Energy Technology Data Exchange (ETDEWEB)

Parker, K [Atomic Weapons Research Establishment, Aldermaston (United Kingdom)

1962-03-15

The AWRE punched-card library of neutron cross-sections is described together with associated IBM-7090 programmes which process this data to give group-averaged cross-sections for use in Monte Carlo, Carlson S{sub n} and other multi-group neutronics calculations. The methods developed to deal with both isotropic and anisotropic elastic scattering are described. These include the multi-group transport approximation and the full treatment of anisotropic scattering using the Legendre polynomial moments of the scattering transfer matrix. The principles of group-constant formation are considered and illustrated by describing systems of group constants suitable for fast-reactor calculations. Practical problems such as the empirical adjustment of group constants to reproduce integral results and the collapsing of a many-group set of constants to give a few-group set are discussed. (author) [French] L'auteur decrit le fichier de cartes perforees sur lesquelles on enregistre a l'Atomic Weapons Research Establishment (AWRE) les sections efficaces neutroniques ainsi que les programmes IBM-7090 associes qui sont employes pour le traitement de ces informations, en vue d'obtenir des sections efficaces moyennes par groupe pouvant servir aux calculs de neutroniques a plusieurs groupes, effectues a l'aide des methodes de Monte-Carlo, S{sub n} de Carlson et autres methodes. L'auteur expose ensuite les methodes mises au point roda etudier la diffusion elastique, tant isotrope qu'anisotrope. Elles comprennent l'approximation de transport a plusieurs groupes, ainsi que le traitement complet de la diffusion anisotrope par les moments polynomiaux de Legendre de la matrice de transfert de la diffusion. L'auteur examine les principes de la formation des constantes de groupes; a titre d'illustration, il decrit les systemes de constantes de groupes qui se pretent aux calculs de reacteurs a neutrons rapides. Il expose quelques problemes pratiques, tels que l'ajustement empirique des

An Experiment of Robust Parallel Algorithm for the Eigenvalue problem of a Multigroup Neutron Diffusion based on modified FETI-DP

Energy Technology Data Exchange (ETDEWEB)

Chang, Jonghwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Parallelization of Monte Carlo simulation is widely adpoted. There are also several parallel algorithms developed for the SN transport theory using the parallel wave sweeping algorithm and for the CPM using parallel ray tracing. For practical purpose of reactor physics application, the thermal feedback and burnup effects on the multigroup cross section should be considered. In this respect, the domain decomposition method(DDM) is suitable for distributing the expensive cross section calculation work. Parallel transport code and diffusion code based on the Raviart-Thomas mixed finite element method was developed. However most of the developed methods rely on the heuristic convergence of flux and current at the domain interfaces. Convergence was not attained in some cases. Mechanical stress computation community has also work on the DDM to solve the stress-strain equation using the finite element methods. The most successful domain decomposition method in terms of robustness is FETI-DP. We have modified the original FETI-DP to solve the eigenvalue problem for the multigroup diffusion problem in this study.

Cross section and method uncertainties: the application of sensitivity analysis to study their relationship in radiation transport benchmark problems

International Nuclear Information System (INIS)

Weisbi, C.R.; Oblow, E.M.; Ching, J.; White, J.E.; Wright, R.Q.; Drischler, J.

1975-08-01

Sensitivity analysis is applied to the study of an air transport benchmark calculation to quantify and distinguish between cross-section and method uncertainties. The boundary detector response was converged with respect to spatial and angular mesh size, P/sub l/ expansion of the scattering kernel, and the number and location of energy grid boundaries. The uncertainty in the detector response due to uncertainties in nuclear data is 17.0 percent (one standard deviation, not including uncertainties in energy and angular distribution) based upon the ENDF/B-IV ''error files'' including correlations in energy and reaction type. Differences of approximately 6 percent can be attributed exclusively to differences in processing multigroup transfer matrices. Formal documentation of the PUFF computer program for the generation of multigroup covariance matrices is presented. (47 figures, 14 tables) (U.S.)

Energy meshing techniques for processing ENDF/B-VI cross sections using the AMPX code system

International Nuclear Information System (INIS)

Dunn, M.E.; Greene, N.M.; Leal, L.C.

1999-01-01

Modern techniques for the establishment of criticality safety for fissile systems invariably require the use of neutronic transport codes with applicable cross-section data. Accurate cross-section data are essential for solving the Boltzmann Transport Equation for fissile systems. In the absence of applicable critical experimental data, the use of independent calculational methods is crucial for the establishment of subcritical limits. Moreover, there are various independent modern transport codes available to the criticality safety analyst (e.g., KENO V.a., MCNP, and MONK). In contrast, there is currently only one complete software package that processes data from the Version 6 format of the Evaluated Nuclear Data File (ENDF) to a format useable by criticality safety codes. To facilitate independent cross-section processing, Oak Ridge National Laboratory (ORNL) is upgrading the AMPX code system to enable independent processing of Version 6 formats using state-of-the-art procedures. The AMPX code system has been in continuous use at ORNL since the early 1970s and is the premier processor for providing multigroup cross sections for criticality safety analysis codes. Within the AMPX system, the module POLIDENT is used to access the resonance parameters in File 2 of an ENDF/B library, generate point cross-section data, and combine the cross sections with File 3 point data. At the heart of any point cross-section processing code is the generation of a suitable energy mesh for representing the data. The purpose of this work is to facilitate the AMPX upgrade through the development of a new and innovative energy meshing technique for processing point cross-section data

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)

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

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.)

A comparison of the BUGLE-80, SAILOR, and ELXSIR neutron cross-section libraries for PWR pressure vessels surveillance dosimetry and shielding applications

International Nuclear Information System (INIS)

Basha, H.S.; Manahan, M.P.

1992-01-01

In this paper three multigroup neutron cross-section libraries are used in synthesized three-dimensional discrete ordinates transport analyses to investigate their similarities, differences, and results for pressurized water reactor (PWR) pressure vessel surveillance dosimetry and shielding applications. The calculated-to-experimental (C/E) rations and the calculated reaction rates of several fast reactions are compared for the BUGLE-80, SAILOR, and ELXSIR cross-section libraries at the 97-deg surveillance capsule of the San Onofre Nuclear Generation Station Unit 2 (SONGS-2) and at the 90- and 97-deg (C/E ratios only) cavity dosimetry locations for another PWR (referred to as Reactor X)

Generation and Testing of the ENDF/B-VI Continuous-Energy Cross-Section Library for Use with Continuous-Energy Versions of KENO

International Nuclear Information System (INIS)

Goluoglu, Sedat; Dunn, Michael E.; Greene, Norman Maurice; Petrie Jr, Lester M.; Hollenbach, Daniel F.

2007-01-01

KENO V.a and KENO-VI are Monte Carlo codes that solve the multigroup form of the Boltzmann transport equation. These codes are part of the SCALE system of codes and are used for performing criticality calculations of systems with fissionable material. In general, continuous-energy Monte Carlo methods are preferred because such an approach avoids many of the assumptions inherent in the multigroup treatment. On the other hand, continuous-energy treatment is much more demanding in terms of computer storage space for data, memory requirements, and calculation speed. Continuous-energy versions of KENO V.a and KENO-VI have been created and are being extensively tested. Generation of ENDF/B-VI continuous-energy cross sections is explained, and the results of the validation and verification of the codes and the data are presented

Validation of the WIMSD4M cross-section generation code with benchmark results

International Nuclear Information System (INIS)

Deen, J.R.; Woodruff, W.L.; Leal, L.E.

1995-01-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D 2 O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented

Validation of the WIMSD4M cross-section generation code with benchmark results

Energy Technology Data Exchange (ETDEWEB)

Deen, J.R.; Woodruff, W.L. [Argonne National Lab., IL (United States); Leal, L.E. [Oak Ridge National Lab., TN (United States)

1995-01-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D{sub 2}O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.

Solution of multi-group diffusion equation in x-y-z geometry by finite Fourier transformation

International Nuclear Information System (INIS)

Kobayashi, Keisuke

1975-01-01

The multi-group diffusion equation in three-dimensional x-y-z geometry is solved by finite Fourier transformation. Applying the Fourier transformation to a finite region with constant nuclear cross sections, the fluxes and currents at the material boundaries are obtained in terms of the Fourier series. Truncating the series after the first term, and assuming that the source term is piecewise linear within each mesh box, a set of coupled equations is obtained in the form of three-point equations for each coordinate. These equations can be easily solved by the alternative direction implicit method. Thus a practical procedure is established that could be applied to replace the currently used difference equation. This equation is used to solve the multi-group diffusion equation by means of the source iteration method; and sample calculations for thermal and fast reactors show that the present method yields accurate results with a smaller number of mesh points than the usual finite difference equations. (auth.)

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

Extension and Verification of the Cross-Section Library for the VVER-1000 Surveillance Specimen Region

International Nuclear Information System (INIS)

Kirilova, D.; Belousov, S.; Ilieva, K.

2011-01-01

The objective of this work is a generation of new version of the BGL multigroup cross-section to extend the region of its applicability. The existing library version is problem oriented for VVER-1000 type of reactors and was generated by collapsing of the VITAMIN-B6 problem independent cross-section fine-group library applying the VVER-1000 reactor middle plane spectrum in cylindrical geometry. The new version BGLex additionally contains cross-sections averaged on the corresponding spectra of the surveillance specimen's (SS) region for VVER-1000 type of reactors. Comparative analysis of the neutron spectra for different one-dimensional geometry models that could be applied for the cross-section collapsing using the software package SCALE, showed a high sensitivity of the results to the geometry model. That is why a neutron importance assessment was done for the SS region using the adjoint solution calculated by the two-dimensional code DORT and problem-independent library VITAMIN-B6. The one-dimensional geometry model applied to the cross-section collapsing were determined by the material limits above the reactor core in axial direction z as for every material a homogenization in radial direction was done. The material homogenization in radial direction was done by material weighing taking into account the adjoint solution as well as the neutron source. The one-dimensional geometry model comprising the homogenized weighed materials was applied for the cross-section generation of the fine-group library VITAMIN-B6 to the broad-group structure of BGL library. The new version BGLex was extended with cross-sections for the SS region. Verification and validation of the new version BGLex is forthcoming. It includes comparison between the calculated results with the new version BGLex and the libraries BGL and VITAMIN-B6 and comparison with experimental results. (author)

Extension and Verification of the Cross-Section Library for the VVER- 1000 Surveillance Specimen Region

International Nuclear Information System (INIS)

Kirilova, D.; Belousov, S.; Ilieva, K.

2011-01-01

The objective of this work is a generation of new version of the BGL multigroup cross-section to extend the region of its applicability. The existing library version is problem oriented for VVER-1000 type of reactors and was generated by collapsing of the VITAMIN-B6 problem independent cross-section fine-group library applying the VVER-1000 reactor middle plane spectrum in cylindrical geometry. The new version BGLex additionally contains cross-sections averaged on the corresponding spectra of the surveillance specimen's (SS) region for VVER-1000 type of reactors. Comparative analysis of the neutron spectra for different one-dimensional geometry models that could be applied for the cross-section collapsing using the software package SCALE, showed a high sensitivity of the results to the geometry model. That is why a neutron importance assessment was done for the SS region using the adjoint solution calculated by the two-dimensional code DORT and problem-independent library VITAMIN-B6. The one-dimensional geometry model applied to the cross-section collapsing were determined by the material limits above the reactor core in axial direction z as for every material a homogenization in radial direction was done. The material homogenization in radial direction was done by material weighing taking into account the adjoint solution as well as the neutron source. The one-dimensional geometry model comprising the homogenized weighed materials was applied for the cross-section generation of the fine-group library VITAMIN-B6 to the broad-group structure of BGL library. The new version BGLex was extended with cross-sections for the SS region. Verification and validation of the new version BGLex is forthcoming. It includes comparison between the calculated results with the new version BGLex and the libraries BGL and VITAMIN-B6 and comparison with experimental results. (author)

ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies. Supplement 1

Energy Technology Data Exchange (ETDEWEB)

Wright, R.Q.; Renier, J.P.; Bucholz, J.A.

1995-08-01

The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as {sup 6}Li, {sup 7}Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa).

A broad-group cross-section library based on ENDF/B-VII.0 for fast neutron dosimetry Applications

Energy Technology Data Exchange (ETDEWEB)

Alpan, F.A. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2011-07-01

A new ENDF/B-VII.0-based coupled 44-neutron, 20-gamma-ray-group cross-section library was developed to investigate the latest evaluated nuclear data file (ENDF) ,in comparison to ENDF/B-VI.3 used in BUGLE-96, as well as to generate an objective-specific library. The objectives selected for this work consisted of dosimetry calculations for in-vessel and ex-vessel reactor locations, iron atom displacement calculations for reactor internals and pressure vessel, and {sup 58}Ni(n,{gamma}) calculation that is important for gas generation in the baffle plate. The new library was generated based on the contribution and point-wise cross-section-driven (CPXSD) methodology and was applied to one of the most widely used benchmarks, the Oak Ridge National Laboratory Pool Critical Assembly benchmark problem. In addition to the new library, BUGLE-96 and an ENDF/B-VII.0-based coupled 47-neutron, 20-gamma-ray-group cross-section library was generated and used with both SNLRML and IRDF dosimetry cross sections to compute reaction rates. All reaction rates computed by the multigroup libraries are within {+-} 20 % of measurement data and meet the U. S. Nuclear Regulatory Commission acceptance criterion for reactor vessel neutron exposure evaluations specified in Regulatory Guide 1.190. (authors)

Consistency check of iron and sodium cross sections with integral benchmark experiments using a large amount of experimental information

International Nuclear Information System (INIS)

Baechle, R.-D.; Hehn, G.; Pfister, G.; Perlini, G.; Matthes, W.

1984-01-01

Single material benchmark experiments are designed to check neutron and gamma cross-sections of importance for deep penetration problems. At various penetration depths a large number of activation detectors and spectrometers are placed to measure the radiation field as completely as possible. The large amount of measured data in benchmark experiments can be evaluated best by the global detector concept applied to nuclear data adjustment. A new iteration procedure is presented for adjustment of a large number of multigroup cross sections, which has been implemented now in the modular adjustment code ADJUST-EUR. A theoretical test problem has been deviced to check the total program system with high precision. The method and code are going to be applied for validating the new European Data Files (JEF and EFF) in progress. (Auth.)

Evaluation of cross-section uncertainties using physical constraints for 238U, 239Pu

International Nuclear Information System (INIS)

De Saint Jean, Cyrille; Privas, Edwin; Archier, Pascal; Noguere, Gilles; Litaize, Olivier; Leconte, Pierre; Bernard, David

2014-01-01

Neutron-induced reactions between 0 eV and 20 MeV are based on various physical properties such as nuclear reaction models, microscopic and integral measurements. Most of the time, the evaluation work is done independently between the resolved resonance range and the continuum, giving rise to mismatches for the cross-sections, larger uncertainties on boundary and no cross-correlation between high-energy domain and resonance range. In addition the use of integral experiment is sometimes only related to central values (evaluation is 'working fine' on a dedicated set of benchmarks) and reductions of uncertainties are not straightforward on cross-sections themselves: working fine could be mathematically reflected by a reduced uncertainty. As the CIELO initiative is to bring experts in each field to propose/discuss these matters, after having presented the status of 238 U and 239 Pu cross-sections covariances evaluation (for JEFF-3.2 as well as the WPEC SG34 subgroup), this paper will present several methodologies that may be used to avoid such effects on covariances. A first idea based on the use of experiments overlapping two energy domains appeared in the near past. It was reviewed and extended to the use of systematic uncertainties (normalisation for example) and for integral experiments as well. In addition, we propose a methodology taking into account physical constraints on an overlapping energy domain where both nuclear reaction models are used (continuity of both cross-sections and derivatives for example). The use of Lagrange multiplier (related to these constraints) in a classical generalised least square procedure will be exposed. Some academic examples will then be presented for both point-wise and multi-group cross-sections to present the methodologies. In addition, new results for 239 Pu will be presented on resonance range and higher energies to reduce capture and fission cross-section uncertainties by using integral experiments (JEZEBEL experiment as

Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

Energy Technology Data Exchange (ETDEWEB)

Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

2005-07-01

Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)

Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

International Nuclear Information System (INIS)

Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H.

2005-01-01

Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)

Sensitivity study and functionalization of cross section to fuel and moderator temperature

International Nuclear Information System (INIS)

Zee, Sung Quun; Song, Jae Seung; Cho, Young Chul

1995-11-01

A reactor core neutronics code MASTER is under development as a part of Korean Core Design System ADONIS. MASTER solves two-group three-dimensional; neutron diffusion equation which requires fuel assembly-wise group constants, to calculate the neutron flux distribution in the core. The group constants are obtained from the fuel assembly multi-group neutron transport calculation, and inputted as functions of the core operating condition. The functionalization of the group constant requires sensitivity analysis to various core operating conditions. In this report, the sensitivity of group constant to fuel and moderator temperature were analyzed. Lumped higher order macroscopic cross section derivative method was developed to reduce the computer memory and the number of floating point operations to treat group constants in MASTER. 1 fig., 6 tabs., 2 refs. (Author) .new

Verification of the accuracy of Doppler broadened, self-shielded multigroup cross sections for fast power reactor applications

International Nuclear Information System (INIS)

Ganesan, S.; Gopalakrishnan, V.; Ramanadhan, M.M.; Cullen, D.E.

1988-01-01

Verification results for Doppler broadening and self-shielding are presented. One of the important results presented is that the original SIGMA1 method of numerical Doppler broadening has now been demonstrated to be inaccurate and not capable of producing results to within required accuracies. Fortunately, due to this study, the SIGMA1 method has been significantly improved and the new SIGMA1 is now capable of producing results to within required accuracies. Although this paper presents results based upon using only one code system, it is important to realize that the original SIGMA1 method is presently used in many cross-section processing code systems; the results of this paper indicate that unless these other code systems are updated to include the new SIGMA1 method, the results produced by these code systems could be very inaccurate. The objectives of the IAEA nuclear data processing code verification project are reviewed as well as the requirements for the accuracy of calculation of Doppler coefficients and the present status of these calculations. The initial results of Doppler broadening and self-shielding calculations are presented and the inconsistency of the results which led to the discovery of errors in the original SIGMA1 method of Doppler broadening are pointed out. Analysis of the errors found and improvements in the SIGMA1 method are presented. Improved results are presented in order to demonstrate that the new SIGMA1 method can produce results within required accuracies. Guidelines are presented to limit the uncertainty introduced due to cross-section processing in order to balance available computer resources to accuracy requirements. Finally cross-section processing code users are invited to participate in the IAEA processing code verification project in order to verify the accuracy of their calculated results. (author)

Calculation of self-shielding factors for cross-sections in the unresolved resonance region using the GRUCON applied program package

International Nuclear Information System (INIS)

Sinitsa, V.V.

1984-11-01

The author gives a scheme for the calculation of the self-shielding factors in the unresolved resonance region using the GRUCON applied program package. This package is especially created to be used in the conversion of evaluated neutron cross-section data, as available in existing data libraries, into multigroup microscopic constants. A detailed description of the formulae and algorithms used in the programs is given. Some typical examples of calculation are considered and the results are compared with those of other authors. The calculation accuracy is better than 2%

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)

A subroutine for the calculation of resonance cross sections of U-238 in HTR fuel elements

Energy Technology Data Exchange (ETDEWEB)

Cuniberti, R; Marullo, G C

1971-02-15

In this paper, a survey of the codes used at Ispra for the calculations of resonance absorption in HTR fuel elements is presented and a subroutine for the calculation of resonance cross-sections, in a seven groups energy structure, for a HTR lattice of annular type is described. A library of homogeneous resonance integrals and a wide tabulation of lump and kernel Bell factors, and moderators efficiency is given. This paper deals mainly with the problem of taking into account the correct slowing down of neutrons in the graphite and with the derivation of Bell factors to be used in a multigroup calculation scheme.

Resolution of the multigroup scattering equation in a one-dimensional geometry and subsidiary calculations: the MUDE code; Resolution de l'equation multigroupe de la diffusion dans une geometrie a une dimension et calculs annexes: code MUDE

Energy Technology Data Exchange (ETDEWEB)

Bore, C; Dandeu, Y; Saint-Amand, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

MUDE is a nuclear code written in FORTRAN II for IBM 7090-7094. It resolves a system of difference equations approximating to the one-dimensional multigroup neutron scattering problem. More precisely, this code makes it possible to: 1. Calculate the critical condition of a reactor (k{sub eff}, critical radius, critical composition) and the corresponding fluxes; 2. Calculate the associated fluxes and various subsidiary results; 3. Carry out perturbation calculations; 4. Study the propagation of fluxes at a distance; 5. Estimate the relative contributions of the cross sections (macroscopic or microscopic); 6. Study the changes with time of the composition of the reactor. (authors) [French] MUDE est un code nucleaire ecrit en FORTRAN II pour IBM 7090-7094. Il resout un systeme d'equations aux differences approchant le probleme de diffusion neutronique multigroupe a une dimension. Plus precisement ce code permet de: 1. Calculer la condition critique d'un reacteur (k{sub eff}, rayon critique, composition critique) et les flux correspondants; 2. Calculer les flux adjoints et divers resultats connexes; 3. Effectuer des calculs de perturbation; 4. Etudier la propagation des flux a longue distance; 5. Ponderer des sections efficaces (macroscopiques ou microscopiques); 6. Etudier l'evolution de la composition du reacteur au cours du temps. (auteurs)

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)

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.

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.)

FENDL/A-MCNP and FENDL/A-VITJE. The processed neutron activation cross-section data files of the FENDL project. Version 1.1 of March 1995. Summary documentation

International Nuclear Information System (INIS)

Pashchenko, A.B.; Wienke, H.; Ganesan, S.

1996-01-01

This document summarizes a neutron activation cross-section database processed in two formats as generated by F.M. Mann within the project of the Fusion Evaluated Nuclear Data Library (FENDL): in continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP; and in 175 group multigroup format with VIT-E weighting spectrum, as used by the transmutation code REAC*2/3. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape. (author). 2 refs, 1 tab

VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

1975-10-01

The computer code block VENTURE, designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry is described. A variety of types of problems may be solved: the usual eigenvalue problem, a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations, or an indirect criticality search on nuclide concentrations, or on dimensions. First-order perturbation analysis capability is available at the macroscopic cross section level

Condensation and homogenization of cross sections for the deterministic transport codes with Monte Carlo method: Application to the GEN IV fast neutron reactors

International Nuclear Information System (INIS)

Cai, Li

2014-01-01

In the framework of the Generation IV reactors neutronic research, new core calculation tools are implemented in the code system APOLLO3 for the deterministic part. These calculation methods are based on the discretization concept of nuclear energy data (named multi-group and are generally produced by deterministic codes) and should be validated and qualified with respect to some Monte-Carlo reference calculations. This thesis aims to develop an alternative technique of producing multi-group nuclear properties by a Monte-Carlo code (TRIPOLI-4). At first, after having tested the existing homogenization and condensation functionalities with better precision obtained nowadays, some inconsistencies are revealed. Several new multi-group parameters estimators are developed and validated for TRIPOLI-4 code with the aid of itself, since it has the possibility to use the multi-group constants in a core calculation. Secondly, the scattering anisotropy effect which is necessary for handling neutron leakage case is studied. A correction technique concerning the diagonal line of the first order moment of the scattering matrix is proposed. This is named the IGSC technique and is based on the usage of an approximate current which is introduced by Todorova. An improvement of this IGSC technique is then presented for the geometries which hold an important heterogeneity property. This improvement uses a more accurate current quantity which is the projection on the abscissa X. The later current can represent the real situation better but is limited to 1D geometries. Finally, a B1 leakage model is implemented in the TRIPOLI-4 code for generating multi-group cross sections with a fundamental mode based critical spectrum. This leakage model is analyzed and validated rigorously by the comparison with other codes: Serpent and ECCO, as well as an analytical case.The whole development work introduced in TRIPOLI-4 code allows producing multi-group constants which can then be used in the core

FENDL/MG-2.0 and FENDL/MC-2.0. The processed cross-section libraries for neutron photon transport calculations. Version 1, March 1997. Summary documentation

International Nuclear Information System (INIS)

Wienke, H.; Herman, M.

1998-01-01

Evaluated neutron reaction data and photon-atom interaction cross sections for materials contained in the general purpose Fusion Evaluated Nuclear Data Library (FENDL/E2.0) have been processed with the NJOY code system into VITAMIN-J multigroup structure, for use in discrete-ordinates transport codes, and into continuous energy ACE format, for use in the Monte Carlo transport code MCNP. This document summarizes the resulting data libraries FENDL/MG-2.0 version 1 and FENDL/MC-2.0 version 1. The data are available costfree from the IAEA Nuclear Data Section online or on magnetic tape. (author)

ZZ TEMPEST/MUFT, Thermal Neutron and Fast Neutron Multigroup Cross-Section Library for Program LEOPARD

International Nuclear Information System (INIS)

Kim, Jung-Do; Lee, Jong Tai

1986-01-01

Description of problem or function: Format: TEMPEST and MUFT; Number of groups: 246 thermal groups in TEMPEST Format and 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD. Nuclides: H, O, Zr, C, Fe, Ni, Al, Cr, Mn, U, Pu, Th, Pa, Xe, Sm, B and D. Origin: ENDF/B-4; Weighting spectrum: 1/E + U 235 fission spectrum. Data library of thermal and fast neutron group Cross sections to generate input to the program LEOPARD. The data is based on ENDF/B-4 and consists of two parts: (1) 246 thermal groups in TEMPEST Format. (2) 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD (NESC0279)

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

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.

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)

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

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.

Standard cross-section data

International Nuclear Information System (INIS)

Carlson, A.D.

1984-01-01

The accuracy of neutron cross-section measurement is limited by the uncertainty in the standard cross-section and the errors associated with using it. Any improvement in the standard immediately improves all cross-section measurements which have been made relative to that standard. Light element, capture and fission standards are discussed. (U.K.)

Sample problems for the novice user of the AMPX-II system

International Nuclear Information System (INIS)

Ford, W.E. III; Roussin, R.W.; Petrie, L.M.; Diggs, B.R.; Comolander, H.E.

1979-01-01

Contents of the IBM version of the APMX system distributed by the Radiation Shielding Information Center (APMX-II) are described. Sample problems which demonstrate the procedure for implementing AMPX-II modules to generate point cross sections; generate multigroup neutron, photon production, and photon interaction cross sections for various transport codes; collapse multigroup cross sections; check, edit, and punch multigroup cross sections; and execute a one-dimensional discrete ordinates transport calculation are detailed. 25 figures, 9 tables

The considering of the slowing down effect in the formalism of probability tables. Application to the effective cross section calculation

International Nuclear Information System (INIS)

Bouhelal, O.K.A.

1990-01-01

The exact determination of the effective multigroup cross sections imposes the numerical solution of the slowing down equation on a very fine energy mesh. Given the complexity of these calculations, different approximation methods have been developed but without a satisfactory treatment of the slowing-down effect. The usual methods are essentially based on interpolations using precalculated tables. The models that use the probability tables allow to reduce the amount of data and the computational effort. A variety of methods proposed by Soviets, then by Americans, and finally the French method, based on the ''moments of a probability distribution'' are incontestably valid within the framework of the statistical hypothesis. This stipulates that the collision densities do not depend on cross section and there is no ambiguity in the effective cross section calculation. The objective of our work is to show that the non statistical phenomena, such as the slowing-down effect which is taken into account, can be described by probability tables which are able to represent the neutronic values and collision densities. The formalism involved in the statistical hypothesis, is based on the Gauss quadrature of the cross sections moments. In the non-statistical hypothesis we introduce the crossed probability tables using the quadratures of double integrals of cross sections, comments. Moreover, a mathematical formalism allowing to establish a relationship between the crossed probability tables and the collision densities was developed. This method was applied on uranium-238 in the range of resolved resonances where the slowing down effect is significant. Validity of the method and the analysis of the obtained results are studied through a reference calculation based on a solution of a discretized slowing down equation using a very fine mesh in which each microgroup can be correctly defined via the statistical probability tables. 42 figs., 32 tabs., 49 refs. (author)

Development of a one-group cross section data base of the ORIGEN2 computer code for research reactor applications

Energy Technology Data Exchange (ETDEWEB)

Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Hwang, Won Guk [Kyung Hee University, Seoul (Korea, Republic of)

1992-03-01

A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORlGEN2-predicted burnup-dependent actinide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base. (Author).

Development of a one-group cross section data base of the ORIGEN2 computer code for research reactor applications

International Nuclear Information System (INIS)

Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai; Hwang, Won Guk

1992-01-01

A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORlGEN2-predicted burnup-dependent actinide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base. (Author)

RADHEAT-V3, a code system for generating coupled neutron and gamma-ray group constants and analyzing radiation transport

International Nuclear Information System (INIS)

Koyama, Kinji; Taji, Yukichi; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

1977-07-01

The modular code system RADHEAT is for producing coupled multigroup neutron and gamma-ray cross section sets, analyzing the neutron and gamma-ray transport, and calculating the energy deposition and atomic displacements due to these radiations in a nuclear reactor or shield. The basic neutron cross sections and secondary gamma-ray production data are taken from ENDF/B and POPOP4 libraries respectively. The system (1) generates multigroup neutron cross sections, energy deposition coefficients and atomic displacement factors due to neutron reactions, (2) generates multigroup gamma-ray cross sections and energy transfer coefficients, (3) generates secondary gamma-ray production cross sections, (4) combines these cross sections into the coupled set, (5) outputs and updates the multigroup cross section libraries in convenient formats for other transport codes, (6) analyzes the neutron and gamma-ray transport and calculates the energy deposition and the number density of atomic displacements in a medium, (7) collapses the cross sections to a broad-group structure, by option, using the weighting functions obtained by one-dimensional transport calculation, and (8) plots, by option, multigroup cross sections, and neutron and gamma-ray distributions. Definitions of the input data required in various options of the code system are also given. (auth.)

Calculation of multigroup constants in WIMS format with programs fedgroup and flange and comparison of the results obtained using different evaluated libraries

International Nuclear Information System (INIS)

Trkov, A.; Budnar, M.; Copic, M.; Perdan, A.; Ravnik, M.

1982-01-01

Multigroup constants for 1-H-1, 92-U-235, and 92-U-238 have been calculated. Averaged cross-sections and other constants have been prepared in the WIMS 69-group format. Comparison has been made between group constants obtained with several evaluated libraries (KEDAK-3 1975, 1979, ENDF/B-4, ENDF/B-5) and the WIMS-D library. Observed differences are most pronounced in the resonance and fast region. From test runs on fuel cell with the WIMS program it can be deduced that these differences affect the fewgroup constants significantly. (author)

Cross sections for atmospheric corrections

International Nuclear Information System (INIS)

Meyer, J.P.; Casse, M.; Westergaard, N.

1975-01-01

A set of cross sections for spallation of relativistic nuclei is proposed based on (i) the best available proton cross sections, (ii) an extrapolation to heavier nuclei of the dependence on the number of nucleons lost of the 'target factor' observed for C 12 and O 16 by Lindstrom et al. (1975), in analogy with Rudstam's formalism, and (iii) on a normalization of all cross sections to the total cross sections for production of fragments with Asub(f) >= 6. The obtained cross sections for peripheral interactions are not inconsistent with simple geometrical considerations. (orig.) [de

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.

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

Multitrajectory eikonal cross sections

International Nuclear Information System (INIS)

Turner, R.E.

1983-01-01

With the use of reference and distorted transition operators, a time-correlation-function representation of the inelastic differential cross section has recently been used to obtain distorted eikonal cross sections. These cross sections involve straight-line and reference classical translational trajectories that are unaffected by any internal-state changes which have occurred during the collision. This distorted eikonal theory is now extended to include effects of internal-state changes on the translational motion. In particular, a different classical trajectory is associated with each pair of internal states. Expressions for these inelastic cross sections are obtained in terms of time-ordered cosine and sine memory functions using the Zwanzig-Feshbach projection-operator method. Explicit formulas are obtained in the time-disordered perturbation approximation

Benchmarking of multigroup neutron cross sections libraries on neutron transmission through WWER-440 vessel

International Nuclear Information System (INIS)

Ilieva, K.; Belousov, S.; Apostolov, T.

1998-01-01

The verification of calculated neutron fluence onto the WWER-440/230 pressure vessel is very topical task in particular referring that some of this type of reactors have been operated the major part of its design lifetime. Since the induced activity from the neutron irradiation onto the elements is a simple response of neutron flux the neutron fluence verification usually is done using the measured activity of radionuclides produced during reactor operation. Calculational and experimental results of 54 Mn induced activity of scraps from inner wall of Unit 1 reactor pressure vessel after 18th cycle and detectors irradiated behind the vessel during the 18th cycle of Unit 1 at Kozloduy NPP as well as neutron flux attenuation through the WWER-440/230 pressure vessel are presented. Neutron cross sections libraries generated on the base of ENDF/B-IV and ENDF/B-VI have been used in the calculations. The comparative analysis of evaluated activities and attenuation coefficient demonstrates the better reliability of the neutron fluence calculations by the libraries based on ENDF/B-VI than by ones on ENDF/B-IV. The extreme rarity of data for the activity of scraps from the WWER-440 reactor vessel and its combination with the data for the detectors irradiated behind the vessel makes them especially attractive for verification of calculational methods of neutron fluence onto the WWER-440 vessel with dummy cassettes loading. (author)

FEMA DFIRM Cross Sections

Data.gov (United States)

Minnesota Department of Natural Resources — FEMA Cross Sections are required for any Digital Flood Insurance Rate Map database where cross sections are shown on the Flood Insurance Rate Map (FIRM). Normally...

AFCI-2.0 Neutron Cross Section Covariance Library

Energy Technology Data Exchange (ETDEWEB)

Herman, M.; Herman, M; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.

2011-03-01

The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R&D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78 structural

AFCI-2.0 Neutron Cross Section Covariance Library

International Nuclear Information System (INIS)

Herman, M.; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.

2011-01-01

The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R and D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78

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

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…

Comparison of integral cross section values of several cross section libraries in the SAND-II format

International Nuclear Information System (INIS)

Zijp, W.L.; Nolthenius, H.J.

1976-09-01

A comparison of some integral cross-section values for several cross-section libraries in the SAND-II format is presented. The integral cross-section values are calculated with the aid of the spectrum functions for a Watt fission spectrum, a 1/E spectrum and a Maxwellian spectrum. The libraries which are considered here are CCC-112B, ENDF/B-IV, DETAN74, LAPENAS and CESNEF. These 5 cross-section libraries used have all the SAND-II format. Discrepancies between cross-sections in the different libraries are indicated but not discussed

Relativistic photon-Maxwellian electron cross sections

International Nuclear Information System (INIS)

Wienke, B.R.; Lathrop, B.L.; Devaney, J.J.

1986-01-01

Temperature corrected cross sections, complementing the Klein-Nishina set, are developed for astrophysical, plasma, and transport applications. The set is obtained from a nonlinear least squares fit to the exact photon-Maxwellian electron cross sections, using the static formula as the asymptotic basis. Two parameters are sufficient (two decimal places) to fit the exact cross sections over a range of 0-100 keV in electron temperature, and 0-1 MeV in incident photon energy. The fit is made to the total cross sections, yet the parameters predict both total and differential scattering cross sections well. Corresponding differential energy cross sections are less accurate. An extended fit to (just) the total cross sections, over the temperature and energy range 0-5 MeV, is also described. (author)

Integral nucleus-nucleus cross sections

International Nuclear Information System (INIS)

Barashenkov, V.S.; Kumawat, H.

2003-01-01

Expressions approximating the experimental integral cross sections for elastic and inelastic interactions of light and heavy nuclei at the energies up to several GeV/nucleon are presented. The calculated cross sections are inside the corridor of experimental errors or very close to it. Described in detail FORTRAN code and a numerical example of the cross section approximation are also presented

Photon-splitting cross sections

International Nuclear Information System (INIS)

Johannessen, A.M.; Mork, K.J.; Overbo, I.

1980-01-01

The differential cross section for photon splitting (scattering of one photon into two photons) in a Coulomb field, obtained earlier by Shima, has been integrated numerically to yield various differential cross sections. Energy spectra differential with respect to the energy of one of the outgoing photons are presented for several values of the primary photon energy. Selected examples of recoil momentum distributions and some interesting doubly or multiply differential cross sections are also given. Values for the total cross section are obtained essentially for all energies. The screening effect caused by atomic electrons is also taken into account, and is found to be important for high energies, as in e + e - pair production. Comparisons with various approximate results obtained by previous authors mostly show fair agreement. We also discuss the possibilities for experimental detection and find the most promising candidate to be a measurement of both photons, and their energies, at a moderately high energy

Comparison of integral cross section values of several cross section libraries in the SAND-II format

International Nuclear Information System (INIS)

Zijp, W.L.; Nolthenius, H.J.

1978-01-01

A comparison of some integral cross section values for several cross section libraries in the SAND-II format is presented. The integral cross section values are calculated with aid of the spectrum functions for a Watt fission spectrum, a 1/E spectrum and a Maxwellian spectrum. The libraries which are considered here are CCC-112B, ENDF/B-IV, DETAN74, LAPENAS and CESNEF. These 5 cross section libraries used have all the SAND-II format. (author)

Jet inclusive cross sections

International Nuclear Information System (INIS)

Del Duca, V.

1992-11-01

Minijet production in jet inclusive cross sections at hadron colliders, with large rapidity intervals between the tagged jets, is evaluated by using the BFKL pomeron. We describe the jet inclusive cross section for an arbitrary number of tagged jets, and show that it behaves like a system of coupled pomerons

Electron collision cross sections of mercury

International Nuclear Information System (INIS)

Suzuki, Susumu; Kuzuma, Kiyotaka; Itoh, Haruo

2006-01-01

In this paper, we propose a new collision cross section set for mercury which revises the original set summarized by Hayashi in 1989. Hanne reported three excitation collision cross sections (6 3 P 0 , 6 3 P 1 , 6 3 P 2 ) determined from an electron beam experiment in 1988. As a matter for regret, no attentive consideration was given to combining these three excitation cross sections with the cross section set of Hayashi. Therefore we propose a new set where these three excitation cross sections are included. In this study, other two excitation cross sections (6 1 P 1 , 6 3 D 3 ) except for the three excitation collision cross sections (6 3 P 0 , 6 3 P 1 , 6 3 P 2 ) are taken from the original set of Hayashi. The momentum transfer cross section and the ionization collision cross section are also taken from Hayashi. A Monte Carlo Simulation (MCS) technique is applied for evaluating our new cross section set. The present results of the electron drift velocity and the ionization coefficient are compared to experimental values. Agreement is secured in relation to the electron drift velocity for 1.5 Td 2 ) is the reduced electric field, E (V/cm) is the electric field, N (1/cm 3 ) is the number density of mercury atoms at 0degC, 1 Torr, E/N is also equal to 2.828 x 10 -17 E/p 0 from the relation of the ideal gas equation, p 0 (Torr) is gas pressure at 0degC, 1 Torr=1.33322 x 10 -2 N/cm -2 and 10 -17 V/cm 2 is called 1 Td. Thus it is ensured that our new cross section set is reasonable enough to be used up to 100 eV when considering with the electron drift velocity and the ionization coefficient. (author)

Measurements and analysis of the 127I and 129I neutron capture and total cross sections

International Nuclear Information System (INIS)

Noguere, G.

2005-01-01

Most of the experimental work on the interaction of neutrons with matter has focused on materials important to reactor physics and reactor structures. By comparison, the corresponding data for minor actinides or long-lived fission products are poor. A significant demand has developed for improved neutron cross-section data of these little-studied nuclides due to the surge of interest in the transmutation of nuclear waste. With 400 kg of 129 I produced yearly in the reactors of the EU countries and a very long β - half-life of 1.57 x 10 7 years, iodine requires disposal strategies that will isolate this isotope from the environment for long periods of time. Therefore, 129 I is potentially a key long-lived fission product for transmutation applications, since 129 I transmutes in 130 I after a single neutron capture and decays to 130 Xe with a 12.36 h half-life. Accurate capture cross sections would help to reduce uncertainties in waste management concepts. For that purpose, Time-Of-Flight measurements covering the [0.5 eV-100 keV] energy range have been carried out at the 150 MeV pulsed neutron source GELINA of the Institute for Reference Materials and Measurements (IRMM). Two types of experiments have been performed at the IRMM, namely capture and transmission experiments. They are respectively related to the neutron capture and total cross sections. Since the PbI 2 samples used in this work contain natural and radioactive iodine, extensive measurements of 129 I have been carried out under the same experimental conditions as for the 129 I. The data reduction process was performed with the AGS system, and the resonance parameters were extracted with the SAMMY and REFIT shape analysis codes. In a last step, the parameters have been converted into ENDF-6 format and processed with the NJOY code to produce point-wise and multigroup cross sections, as well as MCNP and ERANOS libraries. (author)

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)

Background-cross-section-dependent subgroup parameters

International Nuclear Information System (INIS)

Yamamoto, Toshihisa

2003-01-01

A new set of subgroup parameters was derived that can reproduce the self-shielded cross section against a wide range of background cross sections. The subgroup parameters are expressed with a rational equation which numerator and denominator are expressed as the expansion series of background cross section, so that the background cross section dependence is exactly taken into account in the parameters. The advantage of the new subgroup parameters is that they can reproduce the self-shielded effect not only by group basis but also by subgroup basis. Then an adaptive method is also proposed which uses fitting procedure to evaluate the background-cross-section-dependence of the parameters. One of the simple fitting formula was able to reproduce the self-shielded subgroup cross section by less than 1% error from the precise evaluation. (author)

Cross-section methodology in SIMMER

International Nuclear Information System (INIS)

Soran, P.D.

1975-11-01

The cross-section methodology incorporated in the SIMMER code is described. Data base for all cross sections is the ENDF/B system with various progressing computer codes to group collapse and modify the group constants which are used in SIMMER. Either infinitely dilute cross sections or the Bondarenko formalism can be used in SIMMER. Presently only a microscopic treatment is considered, but preliminary macroscopic algorithms have been investigated

Cross-section methodology in SIMMER

International Nuclear Information System (INIS)

Soran, P.D.

1976-05-01

The cross-section methodology incorporated in the SIMMER code is described. Data base for all cross sections is the ENDF/B system with various progressing computer codes to group collapse and modify the group constants which are used in SIMMER. Either infinitely dilute cross sections or the Bondarenko formalism can be used in SIMMER. Presently only a microscopic treatment is considered, but preliminary macroscopic algorithms have been investigated

Electron-impact cross sections of Ne

International Nuclear Information System (INIS)

Tsurubuchi, S.; Arakawa, K.; Kinokuni, S.; Motohashi, K.

2000-01-01

Electron-impact absolute emission cross sections were measured for the 3p→3s transitions of Ne. Excitation functions of the 3s→2p first resonance lines were measured in the energy range from the threshold to 1000 eV by a polarization-free optical method and relative cross sections were normalized to the absolute values, (41.0±5.4)x10 -19 cm 2 for the 73.6 nm line and (7.1±1.0)x10 -19 cm 2 for the 74.4 nm line, which were determined at 500 eV. The integrated level-excitation cross sections of Suzuki et al for the 1s 2 and 1s 4 levels were combined with the corresponding 3p→3s cascade cross sections obtained in this paper to give absolute emission cross sections for the resonance lines. The level-excitation cross sections of the 1s 2 and 1s 4 states in Paschen notation were determined from the threshold to 1000 eV by subtracting 3p→3s cascade cross sections from the corresponding 3s→2p emission cross sections of the resonance lines. A large cascade contribution is found in the emission cross section of the resonance lines. It is 28.5% for the 73.6 nm line and 49.6% for the 74.4 nm line at 40 eV, and 17.0 and 61.8%, respectively, at 300 eV. (author)

Neutron cross sections: Book of curves

International Nuclear Information System (INIS)

McLane, V.; Dunford, C.L.; Rose, P.F.

1988-01-01

Neuton Cross Sections: Book of Curves represents the fourth edition of what was previously known as BNL-325, Neutron Cross Sections, Volume 2, CURVES. Data is presented only for (i.e., intergrated) reaction cross sections (and related fission parameters) as a function of incident-neutron energy for the energy range 0.01 eV to 200 MeV. For the first time, isometric state production cross sections have been included. 11 refs., 4 figs

Generation, Testing, and Validation of a WIMS-D/4 Multigroup Cross-Section Library Based on the JENDL-3.2 Nuclear Data

International Nuclear Information System (INIS)

Rahman, Mafizur; Takano, Hideki

2001-01-01

A new 69-group library of multigroup constants for the lattice code WIMS-D/4 has been generated with an improved resonance treatment, processing nuclear data from JENDL-3.2 by NJOY91.108. A parallel ENDF/B-VI based library has also been constructed for intercomparison of results. Benchmark calculations for a number of thermal reactor critical assemblies of both uranium and plutonium fuels have been performed with the code WIMS-D/4.1 with its three different libraries: the original WIMS library (NEA-0329/10) and the new ENDF/B-VI and JENDL-3.2 based libraries. The results calculated with both ENDF and JENDL based libraries show a similar tendency and are found in better agreement with the experimental values. Benchmark parameters are further calculated with the comprehensive lattice code SRAC95. The results from SRAC95 and WIMS-D/4.1 (both using JENDL-3.2 based libraries) agree well with each other. The new library is also verified for its applicability to mixed-oxide cores of varying plutonium contents

Evaluated cross section libraries

International Nuclear Information System (INIS)

Maqurno, B.A.

1976-01-01

The dosimetry tape (ENDF/B-IV tape 412) was issued in a general CSEWG distribution, August 1974. The pointwise cross section data file was tested with specified reference spectra. A group averaged cross section data file (620 groups based on tape 412) was tested with the above spectra and the results are presented in this report

Program GROUPIE (version 79-1): calculation of Bondarenko self-shielded neutron cross sections and multiband parameters from data in the ENDF/B format

International Nuclear Information System (INIS)

Cullen, D.E.

1980-01-01

Program GROUPIE reads evaluated data in the ENDF/B format and uses these data to calculate Bondarenko self-shielded cross sections and multiband parameters. To give as much generality as possible, the program allows the user to specify arbitrary energy groups and an arbitrary energy groups and an arbitrary energy-dependent neutron spectrum (weighing function). To guarantee the accuracy of the results, all integrals are performed analytically; in no case is iteration or any approximate form of integration used. The output from this program includes both listings and multiband parameters suitable for use either in a normal multigroup transport calculation or in a multiband transport calculation. A listing of the source deck is available on request

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)

Nuclear Forensics and Radiochemistry: Cross Sections

Energy Technology Data Exchange (ETDEWEB)

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

2017-11-08

The neutron activation of components in a nuclear device can provide useful signatures of weapon design or sophistication. This lecture will cover some of the basics of neutron reaction cross sections. Nuclear reactor cross sections will also be presented to illustrate the complexity of convolving neutron energy spectra with nuclear excitation functions to calculate useful effective reactor cross sections. Deficiencies in the nuclear database will be discussed along with tools available at Los Alamos to provide new neutron cross section data.

Differential Top Cross-section Measurements

CERN Document Server

Fenton, Michael James; The ATLAS collaboration

2017-01-01

The top quark is the heaviest known fundamental particle. The measurement of the differential top-quark pair production cross-section provides a stringent test of advanced perturbative QCD calculations. The ATLAS collaboration has performed detailed measurements of those differential cross sections at a centre-of-mass energy of 13 TeV. This talk focuses on differential cross-section measurements in the lepton+jets final state, including using boosted top quarks to probe our understanding of top quark production in the TeV regime.

Improvement of Modeling HTGR Neutron Physics by Uncertainty Analysis with the Use of Cross-Section Covariance Information

Science.gov (United States)

Boyarinov, V. F.; Grol, A. V.; Fomichenko, P. A.; Ternovykh, M. Yu

2017-01-01

This work is aimed at improvement of HTGR neutron physics design calculations by application of uncertainty analysis with the use of cross-section covariance information. Methodology and codes for preparation of multigroup libraries of covariance information for individual isotopes from the basic 44-group library of SCALE-6 code system were developed. A 69-group library of covariance information in a special format for main isotopes and elements typical for high temperature gas cooled reactors (HTGR) was generated. This library can be used for estimation of uncertainties, associated with nuclear data, in analysis of HTGR neutron physics with design codes. As an example, calculations of one-group cross-section uncertainties for fission and capture reactions for main isotopes of the MHTGR-350 benchmark, as well as uncertainties of the multiplication factor (k∞) for the MHTGR-350 fuel compact cell model and fuel block model were performed. These uncertainties were estimated by the developed technology with the use of WIMS-D code and modules of SCALE-6 code system, namely, by TSUNAMI, KENO-VI and SAMS. Eight most important reactions on isotopes for MHTGR-350 benchmark were identified, namely: 10B(capt), 238U(n,γ), ν5, 235U(n,γ), 238U(el), natC(el), 235U(fiss)-235U(n,γ), 235U(fiss).

Activation cross section data file, (1)

International Nuclear Information System (INIS)

Yamamuro, Nobuhiro; Iijima, Shungo.

1989-09-01

To evaluate the radioisotope productions due to the neutron irradiation in fission of fusion reactors, the data for the activation cross sections ought to be provided. It is planning to file more than 2000 activation cross sections at final. In the current year, the neutron cross sections for 14 elements from Ni to W have been calculated and evaluated in the energy range 10 -5 to 20 MeV. The calculations with a simplified-input nuclear cross section calculation system SINCROS were described, and another method of evaluation which is consistent with the JENDL-3 were also mentioned. The results of cross section calculation are in good agreement with experimental data and they were stored in the file 8, 9 and 10 of ENDF/B format. (author)

Utilization of cross-section covariance data in FBR core nuclear design and cross-section adjustment

International Nuclear Information System (INIS)

Ishikawa, Makoto

1994-01-01

In the core design of large fast breeder reactors (FBRs), it is essentially important to improve the prediction accuracy of nuclear characteristics from the viewpoint of both reducing cost and insuring reliability of the plant. The cross-section errors, that is, covariance data are one of the most dominant sources for the prediction uncertainty of the core parameters, therefore, quantitative evaluation of covariance data is indispensable for FBR core design. The first objective of the present paper is to introduce how the cross-section covariance data are utilized in the FBR core nuclear design works. The second is to delineate the cross-section adjustment study and its application to an FBR design, because this improved design method markedly enhances the needs and importance of the cross-section covariance data. (author)

Status of pseudo fission product cross sections for fast reactors. Results of the SWG 17, International working party on evaluation coordination of the nuclear science committee, NEA- OECD

International Nuclear Information System (INIS)

Gruppelaar, H.; Kloosterman, J.L.; Pijlgroms, B.J.; Rimpault, G.; Smith, P.; Ignatyuk, A.; Koshcheev, V.; Nikolaev, M.; Thsiboulia, A.; Kawai, M.; Nakagawa, T.; Watanabe, T.; Zukeran, A.; Nakajima, Y.; Matsunobu, H.

1998-08-01

Within the framework of the SWG17 benchmark organized by a Working Party of the Nuclear Science Committee of the Nuclear Energy Agency (NEA), a comparison of lumped or pseudo fission product cross sections for fast reactors has been made. Four institutions participated with data libraries based on the JEF2.2, EAF-4.2, BROND-2, FONDL-2.1, ADL-3 and JENDL-3.2 evaluated nuclear data files. Several parameters have been compared with each other: the one-group cross sections and reactivity worths of the lumped nuclide for several partial absorption and scattering cross sections, and the one-group cross sections of the individual fission products. Also graphs of the multi-group cross sections of the lumped nuclide have been compared, as well as graphs of capture cross sections for 27 nuclides. From two contributions based on JEF2.2, it can be concluded that the data processing influences the capture cross section by about 1% and the inelastic scattering cross section by 2%. The differences between the lumped cross sections of the different data libraries are surprisingly small: maximum 6% for capture and 9% for the inelastic scattering. Similar results are obtained for the reactivity effects. Since the reactivity worth of the lumped nuclide is dominated by the capture reaction, the maximum spread in the total reactivity worth is still only 5.3%. There is a systematic difference between total, elastic and capture cross sections of JENDL-3.2 and JEF2.2 of the same order of magnitude. Possible reasons for this discrepancy have been indicated. The one-group capture and inelastic scattering cross sections of most of the important individual fission products differ by less than 10% (root mean square values). Larger differences are observed for unstable nuclides where there is a lack of experimental data. For the (n,2n) group cross sections, which are rather sensitive to the weighting spectrum in the fast energy range, these differences are several tens of percents. The final

Is the quasielastic pion cross section really bigger than the pion-nucleus reaction cross section

International Nuclear Information System (INIS)

Silbar, R.R.

1979-01-01

It is shown that soft pion charge exchanges may increase the inclusive (π + ,π 0 ') cross section, relative to the total quasielastic (π + ,π + ') cross section, by as much as a factor of two. 4 references

JENDL gas-production cross section file

International Nuclear Information System (INIS)

Nakagawa, Tsuneo; Narita, Tsutomu

1992-05-01

The JENDL gas-production cross section file was compiled by taking cross-section data from JENDL-3 and by using the ENDF-5 format. The data were given to 23 nuclei or elements in light nuclei and structural materials. Graphs of the cross sections and brief description on their evaluation methods are given in this report. (author)

Doppler broadening of cross sections

International Nuclear Information System (INIS)

Buckler, P.A.C.; Pull, I.C.

1962-12-01

Expressions for temperature dependent cross-sections in terms of resonance parameters are obtained, involving generalisations of the conventional Doppler functions, ψ and φ. Descriptions of Fortran sub-routines, which calculate broadened cross-sections in accordance with the derived formulae, are included. (author)

Cross-sectional anatomy for computed tomography

International Nuclear Information System (INIS)

Farkas, M.L.

1988-01-01

This self-study guide recognizes that evaluation and interpretation of CT-images demands a firm understanding of both cross-sectional anatomy and the principles of computed tomography. The objectives of this book are: to discuss the basic principles of CT, to stress the importance of cross-sectional anatomy to CT through study of selected cardinal transverse sections of head, neck, and trunk, to explain orientation and interpretation of CT-images with the aid of corresponding cross-sectional preparations

Density-dependent expressions for photoionization cross-sections

International Nuclear Information System (INIS)

Sun Weiguo; Ma Xiaoguang; Cheng Yansong

2004-01-01

Alternative expressions for photoionization cross-sections and dielectric influence functions are suggested to study the photoionization cross-sections of atoms in solid system. The basic picture is that the photoionization cross-section of atoms in a real system can be described as the coupling between quantum quantity (QQ) and classical quantity (CQ) parts. The QQ part represents the photoionization cross-sections of an isolated particle, while the CQ part may represent most of the important influence of the macroscopic effects (e.g., the interactions of all surrounding polarized particles, and the dielectric property, etc.) on the photoionization cross-sections. The applications to the barium system show that the number-density-dependent new photoionization formula not only obtains the same cross-sections as those from the first order approximation for ideal gas, but also can generate the cross-sections for solid barium by transforming those of ideal gas of the same species using the dielectric influence function

Density-dependent expressions for photoionization cross-sections

Energy Technology Data Exchange (ETDEWEB)

Sun Weiguo; Ma Xiaoguang; Cheng Yansong

2004-06-07

Alternative expressions for photoionization cross-sections and dielectric influence functions are suggested to study the photoionization cross-sections of atoms in solid system. The basic picture is that the photoionization cross-section of atoms in a real system can be described as the coupling between quantum quantity (QQ) and classical quantity (CQ) parts. The QQ part represents the photoionization cross-sections of an isolated particle, while the CQ part may represent most of the important influence of the macroscopic effects (e.g., the interactions of all surrounding polarized particles, and the dielectric property, etc.) on the photoionization cross-sections. The applications to the barium system show that the number-density-dependent new photoionization formula not only obtains the same cross-sections as those from the first order approximation for ideal gas, but also can generate the cross-sections for solid barium by transforming those of ideal gas of the same species using the dielectric influence function.

Cross sections in 25 groups obtained from ENDF/B-IV and ENDL/78 libraries, processed with GALAXY and NJOY computer codes

International Nuclear Information System (INIS)

Chalhoub, E.S.; Corcuera, R.P.

1982-01-01

The discrepancies existing between ENDF/B-IV and ENDL/78 libraries, in diferent energy regions are identified, and the order of the differences in multigroup sections are determined, when GALAXY or NJOY computer codes are used. (E.G.) [pt

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

Total neutron cross section of lead

International Nuclear Information System (INIS)

Kanda, K.; Aizawa, O.

1976-01-01

The total thermal-neutron cross section of natural lead under various physical conditions was measured by the transmission method. It became clear that the total cross section at room temperature previously reported is lower than the present data. The total cross section at 400, 500, and 600 0 C, above the melting point of lead, 327 0 C, was also measured, and the changes in the cross section as a function of temperature were examined, especially near and below the melting point. The data obtained for the randomly oriented polycrystalline state at room temperature were in reasonable agreement with the theoretical values calculated by the THRUSH and UNCLE-TOM codes

Top quark production cross-section measurements

CERN Document Server

Chen, Ye; The ATLAS collaboration

2017-01-01

Measurements of the inclusive and differential cross-sections for top-quark pair and single top production cross sections in proton-proton collisions with the ATLAS detector at the Large Hadron Collider are presented at center-of-mass energies of 8 TeV and 13 TeV. The inclusive measurements reach high precision and are compared to the best available theoretical calculations. These measurements, including results using boosted tops, probe our understanding of top-pair production in the TeV regime. The results are compared to Monte Carlo generators implementing LO and NLO matrix elements matched with parton showers and NLO QCD calculations. For the t-channel single top measurement, the single top-quark and anti-top-quark total production cross-sections, their ratio, as well as differential cross sections are also presented. A measurement of the production cross-section of a single top quark in association with a W boson, the second largest single-top production mode, is also presented. Finally, measurements of ...

The NJOY Nuclear Data Processing System: Volume 3, The GROUPR, GAMINR, and MODER modules

International Nuclear Information System (INIS)

MacFarlane, R.E.; Muir, D.W.

1987-10-01

The NJOY Nuclear Data Processing System is a comprehensive computer code package for producing pointwise and multigroup cross sections and related quantities from ENDF/B-IV, V, or VI evaluated nuclear data. A concise description of the code system and references to the ancestors of NJOY are given in Vol. 1 of this report. This volume describes the GROUPR module, which produces multigroup neutron interaction cross sections and group-to-group production cross sections for neutrons and photons; the GAMINR module, which produces multigroup photon-interaction cross sections and group-to-group matrices; and the MODER module, which converts ENDF/B and NJOY interface files back and forth between formatted (i.e., BCD, ASCII) and binary modes and performs several associated editing functions. 34 refs., 13 figs

A lumped parameter method of characteristics approach and multigroup kernels applied to the subgroup self-shielding calculation in MPACT

International Nuclear Information System (INIS)

Stimpson, Shane G.; Liu, Yuxuan; Collins, Benjamin S.; Clarno, Kevin T.

2017-01-01

An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC) is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1) a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications) Progression Problem 2a and (2) a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Furthermore given these performance benefits, these approaches have been adopted as the default in MPACT.

A lumped parameter method of characteristics approach and multigroup kernels applied to the subgroup self-shielding calculation in MPACT

Directory of Open Access Journals (Sweden)

Shane Stimpson

2017-09-01

Full Text Available An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1 a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications Progression Problem 2a and (2 a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Given these performance benefits, these approaches have been adopted as the default in MPACT.

Differences between LASL- and ANL-processed cross sections

International Nuclear Information System (INIS)

Kidman, R.B.; MacFarlane, R.E.; Becker, M.

1978-03-01

As part of the Los Alamos Scientific Laboratory (LASL) cross-section processing development, LASL cross sections and results from MINX/1DX system are compared to the Argonne National Laboratory cross sections and results from the ETOE-2/MC 2 -2 system for a simple reactor problem. Exact perturbation theory is used to establish the eigenvalue effect of every isotope group cross-section difference. Cross sections, cross-section differences, and their eigenvalue effects are clearly and conveniently displayed and compared on a group-by-group basis

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)

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.

(n,{alpha}) cross section measurement of gaseous sample using gridded ionization chamber. Cross section determination

Energy Technology Data Exchange (ETDEWEB)

Sanami, Toshiya; Baba, Mamoru; Saito, Keiichiro; Ibara, Yasutaka; Hirakawa, Naohiro [Tohoku Univ., Sendai (Japan). Faculty of Engineering

1997-03-01

We are developing a method of (n,{alpha}) cross section measurement using gaseous samples in a gridded ionization chamber (GIC). This method enables cross section measurements in large solid angle without the distortion by the energy loss in a sample, but requires a method to estimate the detection efficiency. We solve this problem by using GIC signals and a tight neutron collimation. The validity of this method was confirmed through the {sup 12}C(n,{alpha}{sub 0}){sup 9}Be measurement. We applied this method to the {sup 16}O(n,{alpha}){sup 13}C cross section around 14.1 MeV. (author)

Production and testing of the VITAMIN-B6 fine-group and the BUGLE-93 broad-group neutron/photon cross-section libraries derived from ENDF/B-VI nuclear data

International Nuclear Information System (INIS)

Ingersoll, D.T.; White, J.E.; Wright, R.Q.; Hunter, H.T.; Slater, C.O.; Greene, N.M.; MacFarlane, R.E.

1993-01-01

A new multigroup cross-section library based on ENDF/B-VI data has been produced and tested for light water reactor shielding and reactor pressure vessel dosimetry applications. The broad-group library is designated BUGLE-93. The processing methodology is consistent with ANSI/ANS 6.1.2, since the ENDF data were first processed into a fine-group, ''pseudo problem-independent'' format and then collapsed into the final broad-group format. The fine-group library is designated VITAMIN-B6. An extensive integral data testing effort was also performed. In general, results using the new data show significant improvements relative to earlier ENDF data

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)

Photoionization cross section of atomic and molecular oxygen

International Nuclear Information System (INIS)

Pareek, P.N.

1983-01-01

Photoionization cross sections of atomic oxygen and dissociative photoionization cross sections of molecular oxygen were measured from their respective thresholds to 120 angstrom by use of a photoionization mass spectrometer in conjunction with a spark light source. The photoionization cross sections O 2 + parent ion and O + fragment ion from neutral O 2 were obtained by a technique that eliminated the serious problem of identifying the true abundances of O + ions. These ions are generally formed with considerable kinetic energy and, because most mass spectrometers discriminate against energetic ions, true O + abundances are difficult to obtain. In the present work the relative cross sections for producing O + ions are obtained and normalized against the total cross sections in a spectral region where dissociative ionization is not possible. The fragmentation cross sections for O + were then obtained by subtraction of O 2 + cross sections from the known total photoionization cross sections. The results are compared with the previously published measurements. The absolute photoionization cross section of atomic oxygen sigma 8 /sub +/ was measured at 304 A. The actual number density of oxygen atoms within the ionization region was obtained by measuring the fraction of 0 2 molecules dissociated. This sigma/sub +/ at 304 angstrom was used to convert the relative photoinization cross sections, measured as a function of wavelength using a calibrated photodiode, to absolute cross sections. The results are compared with previous measurements and calculated cross sections. angstrom Rydberg series converging to the OII 4 P state was observed

Measurements of neutron capture cross sections

International Nuclear Information System (INIS)

Nakajima, Yutaka

1984-01-01

A review of measurement techniques for the neutron capture cross sections is presented. Sell transmission method, activation method, and prompt gamma-ray detection method are described using examples of capture cross section measurements. The capture cross section of 238 U measured by three different prompt gamma-ray detection methods (large liquid scintillator, Moxon-Rae detector, and pulse height weighting method) are compared and their discrepancies are resolved. A method how to derive the covariance is described. (author)

XCOM: Photon Cross Sections Database

Science.gov (United States)

SRD 8 XCOM: Photon Cross Sections Database (Web, free access)   A web database is provided which can be used to calculate photon cross sections for scattering, photoelectric absorption and pair production, as well as total attenuation coefficients, for any element, compound or mixture (Z <= 100) at energies from 1 keV to 100 GeV.

Graphs of the cross sections in the recommended Monte Carlo cross-section library at the Los Alamos Scientific Laboratory

International Nuclear Information System (INIS)

Soran, P.D.; Seamon, R.E.

1980-05-01

Graphs of all neutron cross sections and photon production cross sections on the Recommended Monte Carlo Cross Section (RMCCS) library have been plotted along with local neutron heating numbers. Values for anti ν, the average number of neutrons per fission, are also given

Scattering cross section for various potential systems

Directory of Open Access Journals (Sweden)

Myagmarjav Odsuren

2017-08-01

Full Text Available We discuss the problems of scattering in this framework, and show that the applied method is very useful in the investigation of the effect of the resonance in the observed scattering cross sections. In this study, not only the scattering cross sections but also the decomposition of the scattering cross sections was computed for the α–α system. To obtain the decomposition of scattering cross sections into resonance and residual continuum terms, the complex scaled orthogonality condition model and the extended completeness relation are used. Applying the present method to the α–α and α–n systems, we obtained good reproduction of the observed phase shifts and cross sections. The decomposition into resonance and continuum terms makes clear that resonance contributions are dominant but continuum terms and their interference are not negligible. To understand the behavior of observed phase shifts and the shape of the cross sections, both resonance and continuum terms are calculated.

Scattering cross section for various potential systems

Energy Technology Data Exchange (ETDEWEB)

Odsuren, Myagmarjav; Khuukhenkhuu, Gonchigdorj; Davaa, Suren [Nuclear Research Center, School of Engineering and Applied Sciences, National University of Mongolia, Ulaanbaatar (Mongolia); Kato, Kiyoshi [Nuclear Reaction Data Centre, Faculty of Science, Hokkaido University, Sapporo (Japan)

2017-08-15

We discuss the problems of scattering in this framework, and show that the applied method is very useful in the investigation of the effect of the resonance in the observed scattering cross sections. In this study, not only the scattering cross sections but also the decomposition of the scattering cross sections was computed for the α–α system. To obtain the decomposition of scattering cross sections into resonance and residual continuum terms, the complex scaled orthogonality condition model and the extended completeness relation are used. Applying the present method to the α–α and α–n systems, we obtained good reproduction of the observed phase shifts and cross sections. The decomposition into resonance and continuum terms makes clear that resonance contributions are dominant but continuum terms and their interference are not negligible. To understand the behavior of observed phase shifts and the shape of the cross sections, both resonance and continuum terms are calculated.

NNLO jet cross sections by subtraction

International Nuclear Information System (INIS)

Somogyi, G.; Bolzoni, P.; Trocsanyi, Z.

2010-06-01

We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of an earlier NNLO subtraction scheme over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state. (orig.)

NNLO jet cross sections by subtraction

Energy Technology Data Exchange (ETDEWEB)

Somogyi, G.; Bolzoni, P. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Trocsanyi, Z. [CERN, Geneva (Switzerland)

2010-06-15

We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of an earlier NNLO subtraction scheme over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state. (orig.)

NNLO jet cross sections by subtraction

CERN Document Server

Somogyi, Gabor; Trocsanyi, Zoltan

2010-01-01

We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of the NNLO subtraction scheme of [1-4], over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state.

Differential cross sections and cross-section ratios for the electron-impact excitation of the neon 2p53s configuration

International Nuclear Information System (INIS)

Khakoo, M. A.; Wrkich, J.; Larsen, M.; Kleiban, G.; Kanik, I.; Trajmar, S.; Brunger, M.J.; Teubner, P.J.O.; Crowe, A.; Fontes, C.J.; Clark, R.E.H.; Zeman, V.; Bartschat, K.; Madison, D.H.; Srivastava, R.; Stauffer, A.D.

2002-01-01

Electron-impact differential cross-section measurements for the excitation of the 2p 5 3s configuration of Ne are reported. The Ne cross sections are obtained using experimental differential cross sections for the electron-impact excitation of the n=2 levels of atomic hydrogen [Khakoo et al., Phys. Rev. A 61, 012701-1 (1999)], and existing experimental helium differential cross-section measurements, as calibration standards. These calibration measurements were made using the method of gas mixtures (Ne and H followed by Ne and He), in which the gas beam profiles of the mixed gases are found to be the same within our experimental errors. We also present results from calculations of these differential cross sections using the R-matrix and unitarized first-order many-body theory, the distorted-wave Born approximation, and relativistic distorted-wave methods. Comparison with available experimental differential cross sections and differential cross-section ratios is also presented

Capture cross sections on unstable nuclei

Science.gov (United States)

Tonchev, A. P.; Escher, J. E.; Scielzo, N.; Bedrossian, P.; Ilieva, R. S.; Humby, P.; Cooper, N.; Goddard, P. M.; Werner, V.; Tornow, W.; Rusev, G.; Kelley, J. H.; Pietralla, N.; Scheck, M.; Savran, D.; Löher, B.; Yates, S. W.; Crider, B. P.; Peters, E. E.; Tsoneva, N.; Goriely, S.

2017-09-01

Accurate neutron-capture cross sections on unstable nuclei near the line of beta stability are crucial for understanding the s-process nucleosynthesis. However, neutron-capture cross sections for short-lived radionuclides are difficult to measure due to the fact that the measurements require both highly radioactive samples and intense neutron sources. Essential ingredients for describing the γ decays following neutron capture are the γ-ray strength function and level densities. We will compare different indirect approaches for obtaining the most relevant observables that can constrain Hauser-Feshbach statistical-model calculations of capture cross sections. Specifically, we will consider photon scattering using monoenergetic and 100% linearly polarized photon beams. Challenges that exist on the path to obtaining neutron-capture cross sections for reactions on isotopes near and far from stability will be discussed.

Status of neutron dosimetry cross sections

International Nuclear Information System (INIS)

Griffin, P.J.; Kelly, J.G.

1992-01-01

Several new cross section libraries, such as ENDF/B-VI(release 2), IRDF-90,JEF-2.2, and JENDL-3 Dosimetry, have recently been made available to the dosimetry community. the Sandia National Laboratories (SNL) Radiation Metrology Laboratory (RML) has worked with these libraries since pre-release versions were available. this paper summarizes the results of the intercomparison and testing of dosimetry cross sections. As a result of this analysis, a compendium of the best dosimetry cross sections was assembled from the available libraries for use within the SNL RML. this library, referred to as the SNLRML Library, contains 66 general dosimetry sensors and 3 special dosimeters unique to the RML sensor inventory. The SNLRML cross sections have been put into a format compatible with commonly used spectrum determination codes

Capture cross sections on unstable nuclei

Directory of Open Access Journals (Sweden)

Tonchev A.P.

2017-01-01

Full Text Available Accurate neutron-capture cross sections on unstable nuclei near the line of beta stability are crucial for understanding the s-process nucleosynthesis. However, neutron-capture cross sections for short-lived radionuclides are difficult to measure due to the fact that the measurements require both highly radioactive samples and intense neutron sources. Essential ingredients for describing the γ decays following neutron capture are the γ-ray strength function and level densities. We will compare different indirect approaches for obtaining the most relevant observables that can constrain Hauser-Feshbach statistical-model calculations of capture cross sections. Specifically, we will consider photon scattering using monoenergetic and 100% linearly polarized photon beams. Challenges that exist on the path to obtaining neutron-capture cross sections for reactions on isotopes near and far from stability will be discussed.

Graphs of the cross sections in the Alternate Monte Carlo Cross Section library at the Los Alamos Scientific Laboratory

International Nuclear Information System (INIS)

Seamon, R.E.; Soran, P.D.

1980-06-01

Graphs of all neutron cross sections and photon production cross sections on the Alternate Monte Carlo Cross Section (AMCCS) library have been plotted along with local neutron heating numbers. The values of ν-bar, the average number of neutrons per fission, are also plotted for appropriate isotopes

Neutron-capture Cross Sections from Indirect Measurements

Energy Technology Data Exchange (ETDEWEB)

Escher, J E; Burke, J T; Dietrich, F S; Ressler, J J; Scielzo, N D; Thompson, I J

2011-10-18

Cross sections for compound-nuclear reactions play an important role in models of astrophysical environments and simulations of the nuclear fuel cycle. Providing reliable cross section data remains a formidable task, and direct measurements have to be complemented by theoretical predictions and indirect methods. The surrogate nuclear reactions method provides an indirect approach for determining cross sections for reactions on unstable isotopes, which are difficult or impossible to measure otherwise. Current implementations of the method provide useful cross sections for (n,f) reactions, but need to be improved upon for applications to capture reactions.

Atlas of photoneutron cross sections obtained with monoenergetic photons

International Nuclear Information System (INIS)

Dietrich, S.S.; Berman, B.L.

1988-01-01

Photoneutron cross-section and integrated cross-section data obtained with monoenergetic photons are presented in a uniform format. All of the measured partial photoneutron cross sections, the total photoneutron cross section, and the photoneutron yield cross section are plotted as functions of the incident photon energy, as are the integrated photoneutron cross sections and their first and second moments. The values of the integrated cross sections and the moments of the integrated total cross section up to the highest photon energy for which they were measured are tabulated, as are the parameters of Lorentz curves fitted to the total photoneutron cross-section data for medium and heavy nuclei (A>50). This compilation is current as of June 1987. copyright 1988 Academic Press, Inc

Cross-Sectional Analysis of Longitudinal Mediation Processes.

Science.gov (United States)

O'Laughlin, Kristine D; Martin, Monica J; Ferrer, Emilio

2018-01-01

Statistical mediation analysis can help to identify and explain the mechanisms behind psychological processes. Examining a set of variables for mediation effects is a ubiquitous process in the social sciences literature; however, despite evidence suggesting that cross-sectional data can misrepresent the mediation of longitudinal processes, cross-sectional analyses continue to be used in this manner. Alternative longitudinal mediation models, including those rooted in a structural equation modeling framework (cross-lagged panel, latent growth curve, and latent difference score models) are currently available and may provide a better representation of mediation processes for longitudinal data. The purpose of this paper is twofold: first, we provide a comparison of cross-sectional and longitudinal mediation models; second, we advocate using models to evaluate mediation effects that capture the temporal sequence of the process under study. Two separate empirical examples are presented to illustrate differences in the conclusions drawn from cross-sectional and longitudinal mediation analyses. Findings from these examples yielded substantial differences in interpretations between the cross-sectional and longitudinal mediation models considered here. Based on these observations, researchers should use caution when attempting to use cross-sectional data in place of longitudinal data for mediation analyses.

Parametric equations for calculation of macroscopic cross sections

International Nuclear Information System (INIS)

Botelho, Mario Hugo; Carvalho, Fernando

2015-01-01

Neutronic calculations of the core of a nuclear reactor is one thing necessary and important for the design and management of a nuclear reactor in order to prevent accidents and control the reactor efficiently as possible. To perform these calculations a library of nuclear data, including cross sections is required. Currently, to obtain a cross section computer codes are used, which require a large amount of processing time and computer memory. This paper proposes the calculation of macroscopic cross section through the development of parametric equations. The paper illustrates the proposal for the case of macroscopic cross sections of absorption (Σa), which was chosen due to its greater complexity among other cross sections. Parametric equations created enable, quick and dynamic way, the determination of absorption cross sections, enabling the use of them in calculations of reactors. The results show efficient when compared with the absorption cross sections obtained by the ALPHA 8.8.1 code. The differences between the cross sections are less than 2% for group 2 and less than 0.60% for group 1. (author)

School belongingness and mental health functioning across the primary-secondary transition in a mainstream sample: multi-group cross-lagged analyses.

Science.gov (United States)

Vaz, Sharmila; Falkmer, Marita; Parsons, Richard; Passmore, Anne Elizabeth; Parkin, Timothy; Falkmer, Torbjörn

2014-01-01

The relationship between school belongingness and mental health functioning before and after the primary-secondary school transition has not been previously investigated in students with and without disabilities. This study used a prospective longitudinal design to test the bi-directional relationships between these constructs, by surveying 266 students with and without disabilities and their parents, 6-months before and after the transition to secondary school. Cross-lagged multi-group analyses found student perception of belongingness in the final year of primary school to contribute to change in their mental health functioning a year later. The beneficial longitudinal effects of school belongingness on subsequent mental health functioning were evident in all student subgroups; even after accounting for prior mental health scores and the cross-time stability in mental health functioning and school belongingness scores. Findings of the current study substantiate the role of school contextual influences on early adolescent mental health functioning. They highlight the importance for primary and secondary schools to assess students' school belongingness and mental health functioning and transfer these records as part of the transition process, so that appropriate scaffolds are in place to support those in need. Longer term longitudinal studies are needed to increase the understanding of the temporal sequencing between school belongingness and mental health functioning of all mainstream students.

School Belongingness and Mental Health Functioning across the Primary-Secondary Transition in a Mainstream Sample: Multi-Group Cross-Lagged Analyses

Science.gov (United States)

Vaz, Sharmila; Falkmer, Marita; Parsons, Richard; Passmore, Anne Elizabeth; Parkin, Timothy; Falkmer, Torbjörn

2014-01-01

The relationship between school belongingness and mental health functioning before and after the primary-secondary school transition has not been previously investigated in students with and without disabilities. This study used a prospective longitudinal design to test the bi-directional relationships between these constructs, by surveying 266 students with and without disabilities and their parents, 6-months before and after the transition to secondary school. Cross-lagged multi-group analyses found student perception of belongingness in the final year of primary school to contribute to change in their mental health functioning a year later. The beneficial longitudinal effects of school belongingness on subsequent mental health functioning were evident in all student subgroups; even after accounting for prior mental health scores and the cross-time stability in mental health functioning and school belongingness scores. Findings of the current study substantiate the role of school contextual influences on early adolescent mental health functioning. They highlight the importance for primary and secondary schools to assess students’ school belongingness and mental health functioning and transfer these records as part of the transition process, so that appropriate scaffolds are in place to support those in need. Longer term longitudinal studies are needed to increase the understanding of the temporal sequencing between school belongingness and mental health functioning of all mainstream students. PMID:24967580

School belongingness and mental health functioning across the primary-secondary transition in a mainstream sample: multi-group cross-lagged analyses.

Directory of Open Access Journals (Sweden)

Sharmila Vaz

Full Text Available The relationship between school belongingness and mental health functioning before and after the primary-secondary school transition has not been previously investigated in students with and without disabilities. This study used a prospective longitudinal design to test the bi-directional relationships between these constructs, by surveying 266 students with and without disabilities and their parents, 6-months before and after the transition to secondary school. Cross-lagged multi-group analyses found student perception of belongingness in the final year of primary school to contribute to change in their mental health functioning a year later. The beneficial longitudinal effects of school belongingness on subsequent mental health functioning were evident in all student subgroups; even after accounting for prior mental health scores and the cross-time stability in mental health functioning and school belongingness scores. Findings of the current study substantiate the role of school contextual influences on early adolescent mental health functioning. They highlight the importance for primary and secondary schools to assess students' school belongingness and mental health functioning and transfer these records as part of the transition process, so that appropriate scaffolds are in place to support those in need. Longer term longitudinal studies are needed to increase the understanding of the temporal sequencing between school belongingness and mental health functioning of all mainstream students.

Low Energy Neutrino Cross Sections

International Nuclear Information System (INIS)

Zeller, G.P.

2004-01-01

Present atmospheric and accelerator based neutrino oscillation experiments operate at low neutrino energies (Ev ∼ 1 GeV) to access the relevant regions of oscillation parameter space. As such, they require precise knowledge of the cross sections for neutrino-nucleon interactions in the sub-to-few GeV range. At these energies, neutrinos predominantly interact via quasi-elastic (QE) or single pion production processes, which historically have not been as well studied as the deep inelastic scattering reactions that dominate at higher energies.Data on low energy neutrino cross sections come mainly from bubble chamber, spark chamber, and emulsion experiments that collected their data decades ago. Despite relatively poor statistics and large neutrino flux uncertainties, these measurements provide an important and necessary constraint on Monte Carlo models in present use. The following sections discuss the current status of QE, resonant single pion, coherent pion, and single kaon production cross section measurements at low energy

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

Group cross-section processing method and common nuclear group cross-section library based on JENDL-3 nuclear data file

International Nuclear Information System (INIS)

Hasegawa, Akira

1991-01-01

A common group cross-section library has been developed in JAERI. This system is called 'JSSTDL-295n-104γ (neutron:295 gamma:104) group constants library system', which is composed of a common 295n-104γ group cross-section library based on JENDL-3 nuclear data file and its utility codes. This system is applicable to fast and fusion reactors. In this paper, firstly outline of group cross-section processing adopted in Prof. GROUCH-G/B system is described in detail which is a common step for all group cross-section library generation. Next available group cross-section libraries developed in Japan based on JENDL-3 are briefly reviewed. Lastly newly developed JSSTDL library system is presented with some special attention to the JENDL-3 data. (author)

Neutron-capture cross sections from indirect measurements

Directory of Open Access Journals (Sweden)

Scielzo N.D.

2012-02-01

Full Text Available Cross sections for compound-nuclear reactions reactions play an important role in models of astrophysical environments and simulations of the nuclear fuel cycle. Providing reliable cross section data remains a formidable task, and direct measurements have to be complemented by theoretical predictions and indirect methods. The surrogate nuclear reactions method provides an indirect approach for determining cross sections for reactions on unstable isotopes, which are difficult or impossible to measure otherwise. Current implementations of the method provide useful cross sections for (n,f reactions, but need to be improved upon for applications to capture reactions.

Curves and tables of neutron cross sections

International Nuclear Information System (INIS)

Nakagawa, Tsuneo; Asami, Tetsuo; Yoshida, Tadashi

1990-07-01

Neutron cross-section curves from the Japanese Evaluated Nuclear Data Library version 3, JENDL-3, are presented in both graphical and tabular form for users in a wide range of application areas in the nuclear energy field. The contents cover cross sections for all the main reactions induced by neutrons with an energy below 20 MeV including; total, elastic scattering, capture, and fission, (n,n'), (n,2n), (n,3n), (n,α), (n,p) reactions. The 2200 m/s cross-section values, resonance integrals, and Maxwellian- and fission-spectrum averaged cross sections are also tabulated. (author)

Cross-sectional and longitudinal relationship between neuroticism and cognitive ability in advanced old age: the moderating role of severe sensory impairment.

Science.gov (United States)

Wettstein, Markus; Kuźma, Elżbieta; Wahl, Hans-Werner; Heyl, Vera

2016-09-01

Gaining a comprehensive picture of the network of constructs in which cognitive functioning is embedded is crucial across the full lifespan. With respect to personality, previous findings support a relationship between neuroticism and cognitive abilities. However, findings regarding old age are inconsistent. In particular, little is known about potentially moderating variables which might explain some of the inconsistency. Our aim was to examine the moderating effect of severe sensory impairment on cross-sectional and longitudinal associations between neuroticism and cognitive functioning. The study sample consisted of 121 visually impaired (VI), 116 hearing impaired (HI), and 150 sensory unimpaired older adults (UI). Mean age was 82.50 years (SD = 4.71 years). Neuroticism was assessed by the NEO Five Factor Inventory, and multiple established tests were used for the assessment of cognitive performance (e.g., subtests of the revised Wechsler Adult Intelligence Scale). Bivariate correlations and multi-group structural equation models indicated stronger relationships between cognitive abilities and neuroticism in both sensory impaired groups (VI and HI) compared to UI older individuals. This relationship was attenuated but still significant in both sensory impaired groups when controlling for age, education and health (number of chronic conditions). In cross-lagged panel models, higher baseline neuroticism was significantly associated with lower cognitive performance four years later in VI and HI individuals. Our results suggest that sensory impairment moderates both cross-sectional and longitudinal associations between neuroticism and cognitive function in advanced old age.

Experiment on neutron transmission through depleted uranium layers and analysis with DOT 3.5 and MCNP

International Nuclear Information System (INIS)

Oka, Y.; Kodama, T.; Akiyama, M.; Hashikura, H.; Kondo, S.

1987-01-01

The reaction rates in the multi-layers containing depleted uranium were measured by activation foils and micro-fission chambers. The analysis of the experiment was carried out by using the multi-group transport calculation code, DOT 3.5 and the continuous energy Monte Carlo code, MCNP. The multi-group calculation overpredicted the low energy reaction rates in the DU layers, while the continuous energy calculation agreed well. The multi-group and continuous energy calculation was compared for the one-dimensional transmission of iron spheres. The results revealed overprediction of the multi-group calculation near the fast neutron source. The averaging of the resonance shapes in generating the multi-group cross sections made minima of the resonance valleys higher than that of the pointwise cross section. This increased the scattering of the neutrons inside and caused the overprediction of the multi-group calculation

Neutron Cross Sections for Aluminium

Energy Technology Data Exchange (ETDEWEB)

Forsberg, Leif

1963-08-15

Total, elastic, inelastic, (n, 2n), (n, {alpha}), (n, p), and (n, {gamma}) cross sections for aluminium have been compiled from thermal to 100 MeV based upon literature search and theoretical interpolations and estimates. Differential elastic cross sections in the centre of mass system are represented by the Legendre coefficients. This method was chosen in order to obtain the best description of the energy dependence of the anisotropy.

High ET jet cross sections at CDF

International Nuclear Information System (INIS)

Flaugher, B.

1996-08-01

The inclusive jet cross section for p anti p collisions at √s = 1.8 TeV as measured by the CDF collaboration will be presented. Preliminary CDF measurements of the Σ E T cross section at √s = 1.8 TeV and the central inclusive jet cross section at √s = 0.630 TeV will also be shown

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

Recommended evaluation procedure for photonuclear cross section

Energy Technology Data Exchange (ETDEWEB)

Lee, Young-Ouk; Chang, Jonghwa; Fukahori, Tokio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

In order to generate photonuclear cross section library for the necessary applications, data evaluation is combined with theoretical evaluation, since photonuclear cross sections measured cannot provide all necessary data. This report recommends a procedure consisting of four steps: (1) analysis of experimental data, (2) data evaluation, (3) theoretical evaluation and, if necessary, (4) modification of results. In the stage of analysis, data obtained by different measurements are reprocessed through the analysis of their discrepancies to a representative data set. In the data evaluation, photonuclear absorption cross sections are evaluated via giant dipole resonance and quasi-deutron mechanism. With photoabsorption cross sections from the data evaluation, theoretical evaluation is applied to determine various decay channel cross sections and emission spectra using equilibrium and preequilibrium mechanism. After this, the calculated results are compared with measured data, and in some cases the results are modified to better describe measurements. (author)

User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code; Notice d'utilisation du code Tripoli-4, version 4.3: code de transport de particules par la methode de Monte Carlo

Energy Technology Data Exchange (ETDEWEB)

Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B

2003-07-01

This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k{sub eff} (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code

International Nuclear Information System (INIS)

Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.

2003-01-01

This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k eff (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

Recommended activation detector cross sections (RNDL-82)

International Nuclear Information System (INIS)

Bondars, Kh.Ya.; Lapenas, A.A.

1984-01-01

The results of the comparison between measured and calculated average cross sections in 5 benchmark experiments are presented. Calculations have been based on the data from 10 libraries of evaluated cross sections. The recommended library (RNDL-82) of the activation detector cross sections has been created on the basis of the comparison. RNDL-82, including 26 reactions, and the basic characteristics of the detectors are presented. (author)

Activation cross section and isomeric cross section ratios for the (n ,2 n ) reaction on 153Eu

Science.gov (United States)

Luo, Junhua; Jiang, Li; Li, Suyuan

2017-10-01

The 153Eu(n ,2 n ) m1,m2,g152Eu cross section was measured by means of the activation technique at three neutron energies in the range 13-15 MeV. The quasimonoenergetic neutron beam was formed via the 3H(d ,n ) 4He reaction, in the Pd-300 Neutron Generator at the Chinese Academy of Engineering Physics (CAEP). The activities induced in the reaction products were measured using high-resolution γ-ray spectroscopy. The cross section of the population of the second high-spin (8-) isomeric state was measured along with the reaction cross section populating both the ground (3-) and the first isomeric state (0-). Cross sections were also evaluated theoretically using the numerical code TALYS-1.8, with different level density options at neutron energies varying from the reaction threshold to 20 MeV. Results are discussed and compared with the corresponding literature.

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.

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

Activation cross section and isomeric cross-section ratio for the (n,2n) reaction on {sup 132,134}Ba

Energy Technology Data Exchange (ETDEWEB)

Luo, Junhua [Hexi Univ., Zhangye (China). School of Physics and Electromechanical Engineering; Hexi Univ., Zhangye (China). Inst. of New Energy; Wu, Chunlei; Jiang, Li [Chinese Academy of Engineering Physics, Mianyang (China). Inst. of Nuclear Physics and Chemistry; Li, Suyuan [Hexi Univ., Zhangye (China). Inst. of New Energy

2017-07-01

Cross sections of the {sup 132}Ba(n,2n){sup 131m,g}Ba and {sup 134}Ba(n,2n){sup 133m,g}Ba reactions and their isomeric cross section ratios σ{sub m}/σ{sub g} have been measured by means of the activation technique at three neutron energies in the range 13-15 MeV. BaCO{sub 3} samples and Nb monitor foils were activated together to determine the reaction cross section and the incident neutron flux. The quasimonoenergetic neutrons beam were produced via the {sup 3}H(d,n){sup 4}He reaction at the Pd-300 Neutron Generator of the Chinese Academy of Engineering Physics (CAEP). The activities induced in the reaction products were measured using high-resolution γ ray spectroscopy. The pure cross section of the ground-state was derived from the absolute cross section of the metastable state and the residual nuclear decay analysis. Cross sections were also evaluated theoretically using the numerical nuclear model code, TALYS-1.8 with different level density options at neutron energies varying from the reaction threshold to 20 MeV. Results are discussed and compared with the corresponding literature.

Dielectronic recombination cross sections for H-like ions

International Nuclear Information System (INIS)

Pindzola, M.S.; Badnell, N.R.; Griffin, D.C.

1990-01-01

Dielectronic recombination cross sections for several H-like atomic ions are calculated in an isolated-resonance, distorted-wave approximation. Fine-structure and configuration-interaction effects are examined in detail for the O 7+ cross section. Hartree-Fock, intermediate-coupled, multiconfiguration dielectronic recombination cross sections for O 7+ are then compared with the recent experimental measurements obtained with the Test Storage Ring in Heidelberg. The cross-section spectra line up well in energy and the shape of the main resonance structures are comparable. The experimental integrated cross sections differ by up to 20% from theory, but this may be due in part to uncertainties in the electron distribution function

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)

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)

Partial cross sections near the higher resonances

International Nuclear Information System (INIS)

Falk-Vairant, P.; Valladas, G.

1961-07-01

As a continuation of the report given at the 10. Rochester Conference, recent measurements of charge-exchange cross section and π 0 production in π - -p interactions are presented here. Section 1 gives a summary of the known results for the elastic, inelastic, and charge-exchange cross sections. Section 2 presents the behavior of the cross sections in the T=1/2 state, in order to discuss the resonances at 600 and 890 MeV. Section 3 discusses the charge-exchange scattering and the interference term between the T=1/2 and T=3/2 states. Section 4 presents some comments on inelastic processes. This report is reprinted from 'Reviews of Modern Physics', Vol. 33, No. 3, 362-367, July, 1961

Classical scattering cross section in sputtering transport theory

International Nuclear Information System (INIS)

Zhang Zhulin

2002-01-01

For Lindhard scaling interaction potential scattering commonly used in sputtering theory, the authors analyzed the great difference between Sigmund's single power and the double power cross sections calculated. The double power cross sections can give a much better approximation to the Born-Mayer scattering in the low energy region (m∼0.1). In particular, to solve the transport equations by K r -C potential interaction given by Urbassek few years ago, only the double power cross sections (m∼0.1) can yield better approximate results for the number of recoils. Therefore, the Sigmund's single power cross section might be replaced by the double power cross sections in low energy collision cascade theory

Quantification of Back-End Nuclear Fuel Cycle Metrics Uncertainties Due to Cross Sections

International Nuclear Information System (INIS)

Tracy E. Stover Jr.

2007-01-01

models are followed by more rigorous reactor physics and depletion models showing a PWR uranium oxide fuel and various metal fast reactor fuels composed of transuranics. The more rigorous models include multi-group cross sections, multiple burnup steps, neutron transport calculations to update cross sections, and multi-scale multi-physics code sequences to simulate a complete fuel lifetime. Finally, the fast reactor and PWR fuels are combined in a closed fast reactor recycle fuel cycle, and uncertainties on the resulting equilibrium cycle examined

Status of reactor physics activities on cross section generation and functionalization for the prismatic very high temperature reactor, and development of spatially-heterogeneous codes

International Nuclear Information System (INIS)

Lee, C. H.; Zhong, Z.; Taiwo, T. A.; Yang, W. S.; Smith, M. A.; Palmiotti, G.

2006-01-01

The cross section generation methodology and procedure for design and analysis of the prismatic Very High Temperature Gas-cooled Reactor (VHTR) core have been addressed for the DRAGON and REBUS-3/DIF3D code suite. Approaches for tabulation and functionalization of cross sections have been investigated and implemented. The cross sections are provided at different burnup and fuel and moderator temperature states. In the tabulation approach, the multigroup cross sections are tabulated as a function of the state variables so that a cross section file is able to cover the range of core operating conditions. Cross sections for points between tabulated data points are fitted simply by linear interpolation. For the functionalization approach, an investigation of the applicability of quadratic polynomials and linear coupling for fuel and moderator temperature changes has been conducted, based on the observation that cross sections are monotonically changing with fuel or moderator temperatures. Preliminary results show that the functionalization makes it possible to cover a wide range of operating temperature conditions with only six sets of data per burnup, while maintaining a good accuracy and significantly reducing the size of the cross section file. In these approaches, the number of fission products has been minimized to a few nuclides (I/Xe/Pm/Sm and a lumped fission product) to reduce the overall computation time without sacrificing solution accuracy. Discontinuity factors (DFs) based on nodal equivalence theory have been introduced to accurately represent the significant change in neutron spectrum at the interface of the fuel and reflector regions as well as between different fuel blocks (e.g., fuel elements with burnable poisons or control rods). Using the DRAGON code, procedures have been established for generating cross sections for fuel and reflector blocks with and without control absorbers. The preliminary results indicate that the solution accuracy is improved

Cross section data for ionization of important cyanides

International Nuclear Information System (INIS)

Kaur, Jaspreet; Antony, Bobby

2015-01-01

Highlights: • Multi centre spherical complex optical potential formalism used to find the CS. • Effective method (CSP-ic) to derive ionization contribution from inelastic CS. • Result shows excellent accord with previous results and consistent behaviour. • Maiden attempt to find CS for many cyanide molecules. • Strong correlation observed between peak of ionization with target properties. - Abstract: This article presents cross section calculations for interactions of important cyanides with electrons possessing energies beginning from ionization threshold of the target molecule to 5 keV. These data are pursued to meet the ever increasing demand for cross sections by the relevant atomic and molecular community for modelling astrophysical, atmospheric and technological domains. The calculations have been executed using an amalgam of multi centre spherical complex optical potential (MSCOP) formalism and complex scattering potential-ionization contribution (CSP-ic) method. Cross sections are compared with experimental and theoretical data wherever available. Strong correlations are observed for the cross sections which affirms consistent and reliable cross sections. Isomeric effect has been interpreted using variation of cross section with structure and target properties. Our cross sections will be tabulated in atomic collision database for use in modelling various statistical and dynamical quantities.

Cross section data for ionization of important cyanides

Energy Technology Data Exchange (ETDEWEB)

Kaur, Jaspreet; Antony, Bobby, E-mail: bka.ism@gmail.com

2015-11-15

Highlights: • Multi centre spherical complex optical potential formalism used to find the CS. • Effective method (CSP-ic) to derive ionization contribution from inelastic CS. • Result shows excellent accord with previous results and consistent behaviour. • Maiden attempt to find CS for many cyanide molecules. • Strong correlation observed between peak of ionization with target properties. - Abstract: This article presents cross section calculations for interactions of important cyanides with electrons possessing energies beginning from ionization threshold of the target molecule to 5 keV. These data are pursued to meet the ever increasing demand for cross sections by the relevant atomic and molecular community for modelling astrophysical, atmospheric and technological domains. The calculations have been executed using an amalgam of multi centre spherical complex optical potential (MSCOP) formalism and complex scattering potential-ionization contribution (CSP-ic) method. Cross sections are compared with experimental and theoretical data wherever available. Strong correlations are observed for the cross sections which affirms consistent and reliable cross sections. Isomeric effect has been interpreted using variation of cross section with structure and target properties. Our cross sections will be tabulated in atomic collision database for use in modelling various statistical and dynamical quantities.

Accurate Cross Sections for Microanalysis

OpenAIRE

Rez, Peter

2002-01-01

To calculate the intensity of x-ray emission in electron beam microanalysis requires a knowledge of the energy distribution of the electrons in the solid, the energy variation of the ionization cross section of the relevant subshell, the fraction of ionizations events producing x rays of interest and the absorption coefficient of the x rays on the path to the detector. The theoretical predictions and experimental data available for ionization cross sections are limited mainly to K shells of a...

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

AUS module MIRANDA - a data preparation code based on multiregion resonance theory

International Nuclear Information System (INIS)

Robinson, G.S.

1977-07-01

MIRANDA is a data preparation module of the AUS reactor neutronics scheme and is used to prepare multigroup cross-section data which are pertinent to a particular reactor system from a general purpose multigroup library of cross sections. The cross-section library has been prepared from ENDF/B and includes temperature dependent data and resonance cross sections represented by subgroup parameters. The MIRANDA module includes a multiregion resonance calculation in slab, cylinder or cluster geometry, a homogeneous Bsub(L) flux solution, and a group condensation facility. Interaction with other AUS modules, particularly for burnup calculations, is provided. (Author)

Criticality benchmark comparisons leading to cross-section upgrades

International Nuclear Information System (INIS)

Alesso, H.P.; Annese, C.E.; Heinrichs, D.P.; Lloyd, W.R.; Lent, E.M.

1993-01-01

For several years criticality benchmark calculations with COG. COG is a point-wise Monte Carlo code developed at Lawrence Livermore National Laboratory (LLNL). It solves the Boltzmann equation for the transport of neutrons and photons. The principle consideration in developing COG was that the resulting calculation would be as accurate as the point-wise cross-sectional data, since no physics computational approximations were used. The objective of this paper is to report on COG results for criticality benchmark experiments in concert with MCNP comparisons which are resulting in corrections an upgrades to the point-wise ENDL cross-section data libraries. Benchmarking discrepancies reported here indicated difficulties in the Evaluated Nuclear Data Livermore (ENDL) cross-sections for U-238 at thermal neutron energy levels. This led to a re-evaluation and selection of the appropriate cross-section values from several cross-section sets available (ENDL, ENDF/B-V). Further cross-section upgrades anticipated

Cross section homogenization analysis for a simplified Candu reactor

International Nuclear Information System (INIS)

Pounders, Justin; Rahnema, Farzad; Mosher, Scott; Serghiuta, Dumitru; Turinsky, Paul; Sarsour, Hisham

2008-01-01

The effect of using zero current (infinite medium) boundary conditions to generate bundle homogenized cross sections for a stylized half-core Candu reactor problem is examined. Homogenized cross section from infinite medium lattice calculations are compared with cross sections homogenized using the exact flux from the reference core environment. The impact of these cross section differences is quantified by generating nodal diffusion theory solutions with both sets of cross sections. It is shown that the infinite medium spatial approximation is not negligible, and that ignoring the impact of the heterogeneous core environment on cross section homogenization leads to increased errors, particularly near control elements and the core periphery. (authors)

Fission cross section measurements of actinides at LANSCE

Energy Technology Data Exchange (ETDEWEB)

Tovesson, Fredrik [Los Alamos National Laboratory; Laptev, Alexander B [Los Alamos National Laboratory; Hill, Tony S [INL

2010-01-01

Fission cross sections of a range of actinides have been measured at the Los Alamos Neutron Science Center (LANSCE) in support of nuclear energy applications. By combining measurement at two LANSCE facilities, Lujan Center and the Weapons Neutron Research center (WNR), differential cross sections can be measured from sub-thermal energies up to 200 MeV. Incident neutron energies are determined using the time-of-flight method, and parallel-plate ionization chambers are used to measure fission cross sections relative to the {sup 235}U standard. Recent measurements include the {sup 233,238}U, {sup 239,242}Pu and {sup 243}Am neutron-induced fission cross sections. In this paper preliminary results for cross section data of {sup 243}Am and {sup 233}U will be presented.

Methodology series module 3: Cross-sectional studies

Directory of Open Access Journals (Sweden)

Maninder Singh Setia

2016-01-01

Full Text Available Cross-sectional study design is a type of observational study design. In a cross-sectional study, the investigator measures the outcome and the exposures in the study participants at the same time. Unlike in case–control studies (participants selected based on the outcome status or cohort studies (participants selected based on the exposure status, the participants in a cross-sectional study are just selected based on the inclusion and exclusion criteria set for the study. Once the participants have been selected for the study, the investigator follows the study to assess the exposure and the outcomes. Cross-sectional designs are used for population-based surveys and to assess the prevalence of diseases in clinic-based samples. These studies can usually be conducted relatively faster and are inexpensive. They may be conducted either before planning a cohort study or a baseline in a cohort study. These types of designs will give us information about the prevalence of outcomes or exposures; this information will be useful for designing the cohort study. However, since this is a 1-time measurement of exposure and outcome, it is difficult to derive causal relationships from cross-sectional analysis. We can estimate the prevalence of disease in cross-sectional studies. Furthermore, we will also be able to estimate the odds ratios to study the association between exposure and the outcomes in this design.

Methodology Series Module 3: Cross-sectional Studies.

Science.gov (United States)

Setia, Maninder Singh

2016-01-01

Cross-sectional study design is a type of observational study design. In a cross-sectional study, the investigator measures the outcome and the exposures in the study participants at the same time. Unlike in case-control studies (participants selected based on the outcome status) or cohort studies (participants selected based on the exposure status), the participants in a cross-sectional study are just selected based on the inclusion and exclusion criteria set for the study. Once the participants have been selected for the study, the investigator follows the study to assess the exposure and the outcomes. Cross-sectional designs are used for population-based surveys and to assess the prevalence of diseases in clinic-based samples. These studies can usually be conducted relatively faster and are inexpensive. They may be conducted either before planning a cohort study or a baseline in a cohort study. These types of designs will give us information about the prevalence of outcomes or exposures; this information will be useful for designing the cohort study. However, since this is a 1-time measurement of exposure and outcome, it is difficult to derive causal relationships from cross-sectional analysis. We can estimate the prevalence of disease in cross-sectional studies. Furthermore, we will also be able to estimate the odds ratios to study the association between exposure and the outcomes in this design.

Compilation of cross-sections. Pt. 1

International Nuclear Information System (INIS)

Flaminio, V.; Moorhead, W.G.; Morrison, D.R.O.; Rivoire, N.

1983-01-01

A compilation of integral cross-sections for hadronic reactions is presented. This is an updated version of CERN/HERA 79-1, 79-2, 79-3. It contains all data published up to the beginning of 1982, but some more recent data have also been included. Plots of the cross-sections versus incident laboratory momentum are also given. (orig.)

Compilation of cross-sections. Pt. 4

International Nuclear Information System (INIS)

Alekhin, S.I.; Ezhela, V.V.; Lugovsky, S.B.; Tolstenkov, A.N.; Yushchenko, O.P.; Baldini, A.; Cobal, M.; Flaminio, V.; Capiluppi, P.; Giacomelli, G.; Mandrioli, G.; Rossi, A.M.; Serra, P.; Moorhead, W.G.; Morrison, D.R.O.; Rivoire, N.

1987-01-01

This is the fourth volume in our series of data compilations on integrated cross-sections for weak, electromagnetic, and strong interaction processes. This volume covers data on reactions induced by photons, neutrinos, hyperons, and K L 0 . It contains all data published up to June 1986. Plots of the cross-sections versus incident laboratory momentum are also given. (orig.)

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)

Evaluation of fusion-evaporation cross-section calculations

Science.gov (United States)

Blank, B.; Canchel, G.; Seis, F.; Delahaye, P.

2018-02-01

Calculated fusion-evaporation cross sections from five different codes are compared to experimental data. The present comparison extents over a large range of nuclei and isotopic chains to investigate the evolution of experimental and calculated cross sections. All models more or less overestimate the experimental cross sections. We found reasonable agreement by using the geometrical average of the five model calculations and dividing the average by a factor of 11.2. More refined analyses are made for example for the 100Sn region.

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)

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

ENDF/B-5 fission product cross section evaluations

International Nuclear Information System (INIS)

Schenter, R.E.; England, T.R.

1979-12-01

Cross section evaluations were made for the 196 fission product nuclides on the ENDF/B-5 data files. Most of the evaluations involve updating the capture cross sections of the important absorbers for fast and thermal reactor systems. This included updating thermal values, resonance integrals, resonance parameter sets, and fast capture cross sections. For the fast capture results generalized least-squares calculations were made with the computer code FERRET. Input for these cross section adjustments included nuclear models calculations and both integral and differential experimental data results. The differential cross sections and their uncertainties were obtained from the CSIRS library. Integral measurement results came from CFRMF and STEK Assemblies 500, 1000, 2000, 3000, 4000. Comparisons of these evaluations with recent capture measurements are shown. 15 figures, 10 tables

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.

Model cross section calculations using LAHET

International Nuclear Information System (INIS)

Prael, R.E.

1992-01-01

The current status of LAHET is discussed. The effect of a multistage preequilibrium exciton model following the INC is examined for neutron emission benchmark calculations, as is the use of a Fermi breakup model for light nuclei rather than an evaporation model. Comparisons are made also for recent fission cross section experiments, and a discussion of helium production cross sections is presented

Interference analysis of fission cross section

International Nuclear Information System (INIS)

Toshkov, S.A.; Yaneva, N.B.

1976-01-01

The formula for the reaction cross-section based on the R-matrix formalism considering the interference between the two neighbouring resonances, referred to the same value of total momentum was used for the analysis of the cross-section of resonance neutron induced fission of 230Pu. The experimental resolution and thermal motion of the target nuclei were accounted for numerical integration

Total cross sections for electron scattering by He

International Nuclear Information System (INIS)

De Heer, F.J.; Jansen, R.H.J.

1977-01-01

A set of total cross sections for scattering of electrons by He has been evaluated over the energy range of zero to 3000 eV by means of the analysis of experiments and theories on total cross sections for elastic scattering, ionisation and excitation, and on differential cross sections for elastic and inelastic scattering. Between 0 and 19.8 eV, where no inelastic processes occur, the total cross sections for scattering are equal to those for elastic scattering. Above 19.8 eV total cross sections for scattering of electrons have been evaluated by adding those for ionisation, excitation and elastic scattering. The total cross sections thus obtained are probably accurate to about 5% over a large part of the energy range. They appear to be in very good agreement with the recent experimental results of Blaauw et al. (J. Phys. B.; 10:L299 (1977)). The present results have already proved useful for application in the dispersion relation for forward scattering in electron-helium collisions. (author)

Investigation of the response of a neutron moisture meter using a multigroup, two-dimensional diffusion theory code

International Nuclear Information System (INIS)

Ritchie, A.I.M.; Wilson, D.J.

1984-12-01

A multigroup diffusion code has been used to predict the count rate from a neutron moisture meter for a range of values of soil water content ω, thermal neutron absorption cross section Ssub(a) (defined as Σsub(a)/rho) of the soil matrix and soil matrix density rho. Two dimensions adequately approximated the geometry of the source, detector and soil surrounding the detector. Seven energy groups, the data for which were condensed from 128 group data set over the neutron energy spectrum appropriate to the soil-water mixture under study, proved adequate to describe neutron slowing-down and diffusion. The soil-water mixture was an SiO 2 →water mixture, with the absorption cross section of SiO 2 increased to cover the range of Σsub(a) required. The response to changes in matrix density is, in general, linear but the response to changes in water content is not linear over the range of parameter values investigated. Tabular results are presented which allow interpolation of the response for a particular ω, Ssub(a) and rho. It is shown that R(ω, Ssub(a), rho) rho M(Ssub(a)) + C(ω) is a crude representation of the response over a very limited range of variation of ω, and Ssub(a). As the response is a slowly varying function of rho, Ssub(a) and ω, a polynomial fit will provide a better estimate of the response for values of rho, Ssub(a) and ω not tabulated

Cross Sections for Inner-Shell Ionization by Electron Impact

Energy Technology Data Exchange (ETDEWEB)

Llovet, Xavier, E-mail: xavier@ccit.ub.edu [Centres Científics i Tecnològics, Universitat de Barcelona, Lluís Solé i Sabarís 1-3, 08028 Barcelona (Spain); Powell, Cedric J. [Materials Measurement Science Division, National Institute of Standards and Technology, Gaithersburg, Maryland 20899-8370 (United States); Salvat, Francesc [Facultat de Física (ECM and ICC), Universitat de Barcelona, Diagonal 645, 08028 Barcelona (Spain); Jablonski, Aleksander [Institute of Physical Chemistry, Polish Academy of Sciences, ul. Kasprzaka 44/52, 01-224 Warsaw (Poland)

2014-03-15

An analysis is presented of measured and calculated cross sections for inner-shell ionization by electron impact. We describe the essentials of classical and semiclassical models and of quantum approximations for computing ionization cross sections. The emphasis is on the recent formulation of the distorted-wave Born approximation by Bote and Salvat [Phys. Rev. A 77, 042701 (2008)] that has been used to generate an extensive database of cross sections for the ionization of the K shell and the L and M subshells of all elements from hydrogen to einsteinium (Z = 1 to Z = 99) by electrons and positrons with kinetic energies up to 1 GeV. We describe a systematic method for evaluating cross sections for emission of x rays and Auger electrons based on atomic transition probabilities from the Evaluated Atomic Data Library of Perkins et al. [Lawrence Livermore National Laboratory, UCRL-ID-50400, 1991]. We made an extensive comparison of measured K-shell, L-subshell, and M-subshell ionization cross sections and of Lα x-ray production cross sections with the corresponding calculated cross sections. We identified elements for which there were at least three (for K shells) or two (for L and M subshells) mutually consistent sets of cross-section measurements and for which the cross sections varied with energy as expected by theory. The overall average root-mean-square deviation between the measured and calculated cross sections was 10.9% and the overall average deviation was −2.5%. This degree of agreement between measured and calculated ionization and x-ray production cross sections was considered to be very satisfactory given the difficulties of these measurements.

Floodplain Cross Section Lines

Data.gov (United States)

Department of Homeland Security — This table is required for any Digital Flood Insurance Rate Map database where cross sections are shown on the Flood Insurance Rate Map (FIRM). Normally any FIRM...

Group cross sections in the resolved resonance region calculated for a CANDU-PHW reactor operating on closed thorium-uranium and thorium-plutonium-uranium fuel cycles

International Nuclear Information System (INIS)

Hamel, D.; Wilkin, G.B.

1979-09-01

Group cross sections in the resolved resonance region are commonly computed for each nuclide independently of other resonance nuclides present in the fuel mixture. While this technique is usually entirely adequate for uranium fuel cycles, it is necessary to establish its legitimacy for closed thorium fuel cycles topped with fissile uranium or plutonium by analysis of a number of representative cases. At the same time cross sections originating from WIMS (Winfrith Improved Multigroup Scheme) calculations are compared with values computed in this study. In this context, particular attention is paid to the adequacy of the lower boundary for the WIMS resonance treatment. All calculations are based on heavy nuclide cross sections from the ENDF/B-IV data compilaton (Evaluated Nuclear Data File). Appreciable interaction effects have been determined for all nuclides except for 232 Th. In most cases, these are due to the strong 232 Th resonance doublet at 21.8 eV and 23.5 eV but some effects also result from resonances of 234 U (5.19 eV, 48.75 eV), 236 U (5.45 eV), 242 Pu (2.67 eV) and others. The influence of mutual interaction on the infinite lattice multiplicaton factor is very small in comparison to the effects of self-shielding. WIMS cross sections do not always compare well with the values computed in the study, but discrepancies are in most cases related to the different sources of data. Interaction effects are not explicitly taken into account in WIMS. Several nuclides ( 233 Pa, 233 U, 240 Pu, 242 Pu) show appreciable self-shielding below the WIMS resonance region and are therefore not treated adequately. The impact of these discrepancies on the multiplication factor is relatively small, however, because of error cancellation. (author)

A Pebble Bed Reactor cross section methodology

International Nuclear Information System (INIS)

Hudson, Nathanael H.; Ougouag, Abderrafi M.; Rahnema, Farzad; Gougar, Hans

2009-01-01

A method is presented for the evaluation of microscopic cross sections for the Pebble Bed Reactor (PBR) neutron diffusion computational models during convergence to an equilibrium (asymptotic) fuel cycle. This method considers the isotopics within a core spectral zone and the leakages from such a zone as they arise during reactor operation. The randomness of the spatial distribution of fuel grains within the fuel pebbles and that of the fuel and moderator pebbles within the core, the double heterogeneity of the fuel, and the indeterminate burnup of the spectral zones all pose a unique challenge for the computation of the local microscopic cross sections. As prior knowledge of the equilibrium composition and leakage is not available, it is necessary to repeatedly re-compute the group constants with updated zone information. A method is presented to account for local spectral zone composition and leakage effects without resorting to frequent spectrum code calls. Fine group data are pre-computed for a range of isotopic states. Microscopic cross sections and zone nuclide number densities are used to construct fine group macroscopic cross sections, which, together with fission spectra, flux modulation factors, and zone buckling, are used in the solution of the slowing down balance to generate a new or updated spectrum. The microscopic cross-sections are then re-collapsed with the new spectrum for the local spectral zone. This technique is named the Spectral History Correction (SHC) method. It is found that this method accurately recalculates local broad group microscopic cross sections. Significant improvement in the core eigenvalue, flux, and power peaking factor is observed when the local cross sections are corrected for the effects of the spectral zone composition and leakage in two-dimensional PBR test problems.

NJOY-94, General ENDF/B Processing System for Reactor Design Problems

International Nuclear Information System (INIS)

1997-01-01

1 - Description of program or function: The NJOY nuclear data processing system is a comprehensive computer code system for producing pointwise and multigroup cross sections and related quantities from ENDF/B evaluated nuclear data in the ENDF format, including the latest US library, ENDF/B-VI. The NJOY code works with neutrons, photons, and charged particles and produces libraries for a wide variety of particle transport and reactor analysis codes. It is capable of processing data in ENDF/B-4, ENDF/B-5, and ENDF/B-6 formats for evaluated data. Short descriptions of the different modules follow: RECONR Reconstructs pointwise cross sections from ENDF/B resonance parameters and interpolation schemes. BROADR Doppler broadens and thins pointwise cross sections. UNRESR Computes effective self-shielded pointwise cross sections in the unresolved-resonance region. HEATR Generates pointwise heat production cross sections and radiation-damage-energy production cross sections. THERMR Produces incoherent inelastic energy-to-energy matrices for free or bound scatterers, coherent elastic cross sections for hexagonal materials, and incoherent elastic cross sections. GROUPR Generates self-shielded multigroup cross sections, group- to-group neutron scattering matrices, and photon production matrices from pointwise input. GAMINR Calculates multigroup photon interaction cross sections and KERMA factors and group-to-group photon scattering matrices. ERRORR Produces multigroup covariance matrices from ENDF/B uncertainties. COVR Reads the output of ERRORR and performs covariance plotting and output-formatting operations. DTFR Formats multigroup data for transport codes such as DTF-IV and ANISN. CCCCR Formats multigroup data for the CCCC standard interface files ISOTXS, BRKOXS, and DLAYXS. MATXSR Formats multigroup data for the MATXS cross section interface file. ACER Prepares libraries for the Los Alamos continuous-energy Monte Carlo code MCNP. POWR Prepares libraries for the EPRI

Application of a discrete-energy, discrete-ordinates technique to the study of neutron transport in iron

International Nuclear Information System (INIS)

Ching, J.T.

1975-01-01

An algebraic equivalence between the point-energy and multigroup forms of the Boltzmann transport equation is demonstrated which allows the development of a discrete-energy, discrete-ordinates method for the solution of radiation transport problems. The method utilizes a modified version of a cross section processing scheme devised for the moments method code BMT and the transport equation solution algorithm from the one-dimensional discrete-ordinates transport code ANISN. The combined system, identified as MOMANS, computes fluxes directly from point cross sections in a single operation. In the cross-section processing, the group averaging required for multigroup calculations is replaced by a fast numerical scheme capable of generating a set of transfer cross sections containing all the physical features of interest, thereby increasing the detail in the calculated results. Test calculations in which the discrete-energy method was compared with the multigroup method have shown that for the same energy grid (number of points = number of groups), the discrete-energy method is faster but somewhat less accurate than the multigroup method. However, the accuracy of the discrete-energy method increases rapidly as the spacing between energy points is decreased, approaching that of multigroup calculations. For problems requiring great detail in the energy spectrum the discrete-energy method has therefore proven to be as accurate as, and more economical than, the multigroup technique. This was demonstrated by the application of the method to the study of the transport of neutrons in an iron sphere. Using the capability of the discrete-energy method for rapidly treating changes in cross-section sets, the propagation of neutrons from a 14 MeV source in a 22 cm radius sphere of iron was analyzed for sensitivity to changes in the microscopic scattering mechanisms

Transport cross section for small-angle scattering

International Nuclear Information System (INIS)

D'yakonov, M.I.; Khaetskii, A.V.

1991-01-01

Classical mechanics is valid for describing potential scattering under the conditions (1) λ much-lt α and (2) U much-gt ℎυ/α, where λ is the de Broglie wavelength, α is the characteristic size of the scatterer, U is the characteristic value of the potential energy, and υ is the velocity of the scattered particle. The second of these conditions means that the typical value of the classical scattering angle is far larger than the diffraction angle λ/α. In this paper the authors show that this second condition need not hold in a derivation of the transport cross section. In other words, provided that the condition λ much-lt α holds, it is always possible to calculate the transport cross section from the expressions of classical mechanics, even in the region U approx-lt ℎυ/α, where the scattering is diffractive,and the differential cross section is greatly different from the classical cross section. The transport cross section is found from the classical expression even in the anticlassical case U much-lt ℎυ/α, where the Born approximation can be used

Nonelastic-scattering cross sections of elemental nickel

International Nuclear Information System (INIS)

Smith, A.B.; Guenther, P.T.; Whalen, J.F.

1980-06-01

Neutron total cross sections of elemental nickel were measured from 1.3 to 4.5 MeV, at intervals of approx. 50 keV, with resolutions of 30 to 50 keV and to accuracies of 1 to 2.5%. Neutron differential-elastic-scattering cross sections were measured from 1.45 to 3.8 MeV, at intervals and with resolutions comparable to those of the total cross sections, and to accuracies of 3 to 5%. The nonelastic-scattering cross section is derived from the measured values to accuracies of greater than or equal to 6%. The experimental results are compared with previously reported values as represented by ENDF/B-V, and areas of consistency and discrepancy, noted. The measured results are shown to be in good agreement with the predictions of a model previously reported by the authors. 4 figures, 1 table

Fragmentation cross sections outside the limiting-fragmentation regime

CERN Document Server

Sümmerer, K

2003-01-01

The empirical parametrization of fragmentation cross sections, EPAX, has been successfully applied to estimate fragment production cross sections in reactions of heavy ions at high incident energies. It is checked whether a similar parametrization can be found for proton-induced spallation around 1 GeV, the range of interest for ISOL-type RIB facilities. The validity of EPAX for medium-energy heavy-ion induced reactions is also checked. Only a few datasets are available, but in general EPAX predicts the cross sections rather well, except for fragments close to the projectile, where the experimental cross sections are found to be larger.

Measurements of fission cross-sections. Chapter 4

International Nuclear Information System (INIS)

James, G.D.

1981-01-01

The steps involved in the measurement of fission cross sections are summarized and the range of techniques available are considered. Methods of fission detection are described with particular emphasis on the neutron energy dependent properties of the fission process and the details of fragment energy loss which can lead to energy-dependent changes in detector efficiency. Selected examples of fission cross-section measurements are presented and methods of data reduction, storage, analysis and evaluation, are examined. Finally requested accuracies for fission cross section data are compared to estimated available accuracies. (U.K.)

Vibrational enhancement of total breakup cross sections

International Nuclear Information System (INIS)

Haftel, M.I.; Lim, T.K.

1984-01-01

This paper considers the role of multi-two-body bound states, namely vibrational excitations, on total three-body breakup cross-sections. Total cross-sections are usually easy to measure, and they play a fundamental role in chemical kinetics. (orig.)

Target dependence of K+-nucleus total cross sections

International Nuclear Information System (INIS)

Jiang, M.F.; Ernst, D.J.; Chen, C.M.

1995-01-01

We investigate the total cross section and its target dependence for K + -nucleus scattering using a relativistic momentum-space optical potential model which incorporates relativistically normalized wave functions, invariant two-body amplitudes, covariant kinematics, and an exact full-Fermi averaging integral. The definition of the total cross section in the presence of a Coulomb interaction is reviewed and the total cross section is calculated in a way that is consistent with what is extracted from experiment. In addition, the total cross sections for a nucleus and for the deuteron are calculated utilizing the same theory. This minimizes the dependence of the ratio of these cross sections on the details of the theory. The model dependence of the first-order optical potential calculations is investigated. The theoretical results are found to be systematically below all existing data

Tables of RCN-2 fission-product cross section evaluation

International Nuclear Information System (INIS)

Gruppelaar, H.

1979-05-01

This report (continuation of ECN-13 and ECN-33) describes the third part of the RCN-2 evaluation of neutron cross sections for fission product nuclides in KEDAK format. It contains evaluated data for nine nuclides, i.e. 142 Nd, 143 Nd, 144 Nd, 145 Nd, 146 Nd, 147 Nd, 148 Nd, 150 Nd and 147 Pm. Most emphasis has been given to the evaluation of the radiative capture cross section, in order to provide a data base for adjustment calculations using results of integral measurements. Short evaluation reports are given for this cross section. The evaluated capture cross sections are compared with recent experimental differential and integral data. Graphs are given of the capture cross sections at neutron energies above 1 keV, in which also adjusted point cross sections, based upon integral STEK and CFRMF data have been plotted. Moreover, the results are compared with those of the well-known ENDF/B-IV evaluation for fission product nucleides. Finally, evaluation summaries are given, which include tables of other important neutron cross sections, such as the total, elastic scattering and inelastic scattering cross sections

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

Microscopic cross sections: An utopia?

Energy Technology Data Exchange (ETDEWEB)

Hilaire, S. [CEA Bruyeres-le-Chatel, DIF 91 (France); Koning, A.J. [Nuclear Research and Consultancy Group, PO Box 25, 1755 ZG Petten (Netherlands); Goriely, S. [Institut d' Astronomie et d' Astrophysique, Universite Libre de Bruxelles, Campus de la Plaine, CP 226, 1050 Brussels (Belgium)

2010-07-01

The increasing need for cross sections far from the valley of stability poses a challenge for nuclear reaction models. So far, predictions of cross sections have relied on more or less phenomenological approaches, depending on parameters adjusted to available experimental data or deduced from systematical relations. While such predictions are expected to be reliable for nuclei not too far from the experimentally known regions, it is clearly preferable to use more fundamental approaches, based on sound physical bases, when dealing with very exotic nuclei. Thanks to the high computer power available today, all major ingredients required to model a nuclear reaction can now be (and have been) microscopically (or semi-microscopically) determined starting from the information provided by a nucleon-nucleon effective interaction. We have implemented all these microscopic ingredients in the TALYS nuclear reaction code, and we are now almost able to perform fully microscopic cross section calculations. The quality of these ingredients and the impact of using them instead of the usually adopted phenomenological parameters will be discussed. (authors)

Microscopic cross sections: An utopia?

International Nuclear Information System (INIS)

Hilaire, S.; Koning, A.J.; Goriely, S.

2010-01-01

The increasing need for cross sections far from the valley of stability poses a challenge for nuclear reaction models. So far, predictions of cross sections have relied on more or less phenomenological approaches, depending on parameters adjusted to available experimental data or deduced from systematical relations.While such predictions are expected to be reliable for nuclei not too far from the experimentally known regions, it is clearly preferable to use more fundamental approaches, based on sound physical bases, when dealing with very exotic nuclei. Thanks to the high computer power available today, all major ingredients required to model a nuclear reaction can now be (and have been) microscopically (or semi-microscopically) determined starting from the information provided by a nucleon-nucleon effective interaction. We have implemented all these microscopic ingredients in the TALYS nuclear reaction code, and we are now almost able to perform fully microscopic cross section calculations. The quality of these ingredients and the impact of using them instead of the usually adopted phenomenological parameters will be discussed. (authors)

Multilevel parametrization of fissile nuclei resonance cross sections

International Nuclear Information System (INIS)

Lukyanov, A.A.; Kolesov, V.V.; Janeva, N.

1987-01-01

Because the resonance interference has an important influence on the resonance structure of neutron cross sections energy dependence at lowest energies, multilevel scheme of the cross section parametrization which take into account the resonance interference is used for the description with the same provisions in the regions of the interferential maximum and minimum of the resonance cross sections of the fissile nuclei

Total and ionization cross sections of electron scattering by fluorocarbons

International Nuclear Information System (INIS)

Antony, B K; Joshipura, K N; Mason, N J

2005-01-01

Electron impact total cross sections (50-2000 eV) and total ionization cross sections (threshold to 2000 eV) are calculated for typical plasma etching molecules CF 4 , C 2 F 4 , C 2 F 6 , C 3 F 8 and CF 3 I and the CF x (x 1-3) radicals. The total elastic and inelastic cross sections are determined in the spherical complex potential formalism. The sum of the two gives the total cross section and the total inelastic cross section is used to calculate the total ionization cross sections. The present total and ionization cross sections are found to be consistent with other theories and experimental measurements, where they exist. Our total cross section results for CF x (x = 1-3) radicals presented here are first estimates on these species

Calculation of atom displacement cross section for structure material

International Nuclear Information System (INIS)

Liu Ping; Xu Yiping

2015-01-01

The neutron radiation damage in material is an important consideration of the reactor design. The radiation damage of materials mainly comes from atom displacements of crystal structure materials. The reaction cross sections of charged particles, cross sections of displacements per atom (DPA) and KERMA are the basis of radiation damage calculation. In order to study the differences of DPA cross sections with different codes and different evaluated nuclear data libraries, the DPA cross sections for structure materials were calculated with UNF and NJOY codes, and the comparisons of results were given. The DPA cross sections from different evaluated nuclear data libraries were compared. And the comparison of DPA cross sections between NJOY and Monte Carlo codes was also done. The results show that the differences among these evaluated nuclear data libraries exist. (authors)

Preparation of next generation set of group cross sections. 3

International Nuclear Information System (INIS)

Kaneko, Kunio

2002-03-01

This fiscal year, based on the examination result about the evaluation energy range of heavy element unresolved resonance cross sections, the upper energy limit of the energy range, where ultra-fine group cross sections are produced, was raised to 50 keV, and an improvement of the group cross section processing system was promoted. At the same time, reflecting the result of studies carried out till now, a function producing delayed neutron data was added to the general-purpose group cross section processing system , thus the preparation of general purpose group cross section processing system has been completed. On the other hand, the energy structure, data constitution and data contents of next generation group cross section set were determined, and the specification of a 151 groups next generation group cross section set was defined. Based on the above specification, a concrete library format of the next generation cross section set has been determined. After having carried out the above-described work, using the general-purpose group cross section processing system , which was complete in this study, with use of the JENDL-3. 2 evaluated nuclear data, the 151 groups next generation group cross section of 92 nuclides and the ultra fine group resonance cross section library for 29 nuclides have been prepared. Utilizing the 151 groups next generation group cross section set and the ultra-fine group resonance cross-section library, a bench mark test calculation of fast reactors has been performed by using an advanced lattice calculation code. It was confirmed, by comparing the calculation result with a calculation result of continuous energy Monte Carlo code, that the 151 groups next generation cross section set has sufficient accuracy. (author)

Positive Scattering Cross Sections using Constrained Least Squares

International Nuclear Information System (INIS)

Dahl, J.A.; Ganapol, B.D.; Morel, J.E.

1999-01-01

A method which creates a positive Legendre expansion from truncated Legendre cross section libraries is presented. The cross section moments of order two and greater are modified by a constrained least squares algorithm, subject to the constraints that the zeroth and first moments remain constant, and that the standard discrete ordinate scattering matrix is positive. A method using the maximum entropy representation of the cross section which reduces the error of these modified moments is also presented. These methods are implemented in PARTISN, and numerical results from a transport calculation using highly anisotropic scattering cross sections with the exponential discontinuous spatial scheme is presented

The self shielding module of Apollo.II; Module d`autoprotection du code Apollo.II

Energy Technology Data Exchange (ETDEWEB)

Sanchez, R.

1994-06-01

This note discusses the methods used in the APOLLO.II code for the calculation of self shielded multigroup cross sections. Basically, the calculation consists in characterizing a heterogenous medium with a single parameter: the background cross section, which is in then used to interpolate reaction rates from pre tabulated values. Very fine multigroup slowing down calculations in homogenous media are used to generate these tables, which contain absorption, diffusion and production reaction rates per group, resonant isotope, temperature and background cross section. Multigroup self shielded cross sections are determined from an equivalence that preserves absorption rates at a slowing down problem with given sources. This article gives a detailed description of the PIC and ``dilution matrix`` formalisms that are used in the homogenization step, as well as the utilization of Bell macro-groups and the different quadrature formulas that may be used in the calculations. Self shielding techniques for isotopic resonant mixtures are also discussed. (author). 2 refs., 193 figs., 2 tabs.

Heisenberg rise of total cross sections

International Nuclear Information System (INIS)

Ezhela, V.V.; Yushchenko, O.P.

1988-01-01

It is shown that on the basis of the original idea of Heisenberg on the quasiclassical picture of extended particle interactions one can construct a satisfactory description of the total cross sections, elastic cross sections, elastic diffractive slopes and mean charged multiplicities in the cm energy range from 5 to 900 GeV, and produce reasonable extrapolations up to several tens of TeV. 14 refs.; 7 figs.; 2 tabs

FENDL/E-2.0. Evaluated nuclear data library of neutron-nucleus interaction cross sections and photon production cross sections and photon-atom interaction cross sections for fusion applications. Version 1, March 1997. Summary documentation

International Nuclear Information System (INIS)

Pashchenko, A.B.; Wienke, H.

1998-01-01

This document presents the description of a physical tape containing the basic evaluated nuclear data library of neutron-nucleus interaction cross sections, photon production cross sections and photon-atom interaction cross sections for fusion applications. It is part of the evaluated nuclear data library for fusion applications FENDL-2. The data are available cost-free from the Nuclear Data Section upon request. The data can also be retrieved by the user via online access through international computer networks. (author)

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

Cross sections for hadron and lepton production processes

International Nuclear Information System (INIS)

Bhattacharya, R.

1976-01-01

Charged heavy lepton production in proton-proton collisions is studied. Motivated by recent experimental results from the Stanford Linear Accelerator Center a parton model analysis is given of the reaction p + p → L + + L - + x → μ +- + e/ -+ / + neutrinos + x. Results are presented for the total cross section and the differential cross sections with respect to the invariant mass squared of the final charged leptons and the transverse momenta of each one of them. The two-photon mechanism for pair production in colliding beam exeriments is considered. Through the use of mapped invariant integration variables, a reliable exact numerical calculation of the cross section for the production of muon and pion pairs by the two-photon mechanism is provided. Results are given for the exact total cross sections and also the differential cross sections with respect to the invariant mass squared of the pair. These are compared to the results obtained from the equivalent photon approximation method

Total reaction cross sections and neutron-removal cross sections of neutron-rich light nuclei measured by the COMBAS fragment-separator

Science.gov (United States)

Hue, B. M.; Isataev, T.; Erdemchimeg, B.; Artukh, A. G.; Aznabaev, D.; Davaa, S.; Klygin, S. A.; Kononenko, G. A.; Khuukhenkhuu, G.; Kuterbekov, K.; Lukyanov, S. M.; Mikhailova, T. I.; Maslov, V. A.; Mendibaev, K.; Sereda, Yu M.; Penionzhkevich, Yu E.; Vorontsov, A. N.

2017-12-01

Preliminary results of measurements of the total reaction cross sections σR and neutron removal cross section σ-xn for weakly bound 6He, 8Li, 9Be and 10Be nuclei at energy range (20-35) A MeV with 28Si target is presented. The secondary beams of light nuclei were produced by bombardment of the 22Ne (35 A MeV) primary beam on Be target and separated by COMBAS fragment-separator. In dispersive focal plane a horizontal slit defined the momentum acceptance as 1% and a wedge degrader of 200 μm Al was installed. The Bρ of the second section of the fragment-separator was adjusted for measurements in energy range (20-35) A MeV. Two-neutron removal cross sections for 6He and 10Be and one -neutron removal cross sections 8Li and 9Be were measured.

Microscopic cross-section measurements by thermal neutron activation

International Nuclear Information System (INIS)

Avila L, J.

1987-08-01

Microscopic cross sections measured by thermal neutron activation using RP-0 reactor at the Peruvian Nuclear Energy Institute. The method consists in measuring microscopic cross section ratios through activated samples, requiring being corrected in thermal and epithermal energetic range by Westcott formalism. Furthermore, the comptage ratios measured for each photopeak to its decay fraction should be normalized from interrelation between both processes above, activation microscopic cross sections are obtained

Distorted eikonal cross sections: A time-dependent view

International Nuclear Information System (INIS)

Turner, R.E.

1982-01-01

For Hamiltonians with two potentials, differential cross sections are written as time-correlation functions of reference and distorted transition operators. Distorted eikonal differential cross sections are defined in terms of straight-line and reference classical trajectories. Both elastic and inelastic results are obtained. Expressions for the inelastic cross sections are presented in terms of time-ordered cosine and sine memory functions through the use of the Zwanzig-Feshbach projection-operator method

Discussion of electron cross sections for transport calculations

International Nuclear Information System (INIS)

Berger, M.J.

1983-01-01

This paper deals with selected aspects of the cross sections needed as input for transport calculations and for the modeling of radiation effects in biological materials. Attention is centered mainly on the cross sections for inelastic interactions between electrons and water molecules and the use of these cross sections for the calculation of energy degradation spectra and of ionization and excitation yields. 40 references, 3 figures, 1 table

Tachyonic ionization cross sections of hydrogenic systems

Energy Technology Data Exchange (ETDEWEB)

Tomaschitz, Roman [Department of Physics, Hiroshima University, 1-3-1 Kagami-yama, Higashi-Hiroshima 739-8526 (Japan)

2005-03-11

Transition rates for induced and spontaneous tachyon radiation in hydrogenic systems as well as the transversal and longitudinal ionization cross sections are derived. We investigate the interaction of the superluminal radiation field with matter in atomic bound-bound and bound-free transitions. Estimates are given for Ly-{alpha} transitions effected by superluminal quanta in hydrogen-like ions. The tachyonic photoelectric effect is scrutinized, in the Born approximation and at the ionization threshold. The angular maxima occur at different scattering angles in the transversal and longitudinal cross sections, which can be used to sift out longitudinal tachyonic quanta in a photon flux. We calculate the tachyonic ionization and recombination cross sections for Rydberg states and study their asymptotic scaling with respect to the principal quantum number. At the ionization threshold of highly excited states of order n {approx} 10{sup 4}, the longitudinal cross section starts to compete with photoionization, in recombination even at lower levels.

Symmetric charge transfer cross section of uranium

International Nuclear Information System (INIS)

Shibata, Takemasa; Ogura, Koichi

1995-03-01

Symmetric charge transfer cross section of uranium was calculated under consideration of reaction paths. In the charge transfer reaction a d 3/2 electron in the U atom transfers into the d-electron site of U + ( 4 I 9/2 ) ion. The J value of the U atom produced after the reaction is 6, 5, 4 or 3, at impact energy below several tens eV, only resonant charge transfer in which the product atom is ground state (J=6) takes place. Therefore, the cross section is very small (4-5 x 10 -15 cm 2 ) compared with that considered so far. In the energy range of 100-1000eV the cross section increases with the impact energy because near resonant charge transfer in which an s-electron in the U atom transfers into the d-electron site of U + ion. Charge transfer cross section between U + in the first excited state (289 cm -1 ) and U in the ground state was also obtained. (author)

NNLO jet cross sections by subtraction

Energy Technology Data Exchange (ETDEWEB)

Somogyi, G.; Bolzoni, P. [DESY, Platanenallee 6, D-15738 Zeuthen (Germany); Trocsanyi, Z. [CERN PH-TH, on leave from University of Debrecen and Institute of Nuclear Research of HAS, H-4001 P.O.Box 51 (Hungary)

2010-08-15

We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of the NNLO subtraction scheme of Refs. [G. Somogyi, Z. Trocsanyi, and V. Del Duca, JHEP 06, 024 (2005), (arXiv:hep-ph/0502226); G. Somogyi and Z. Trocsanyi, (2006), (arXiv:hep-ph/0609041); G. Somogyi, Z. Trocsanyi, and V. Del Duca, JHEP 01, 070 (2007), (arXiv:hep-ph/0609042); G. Somogyi and Z. Trocsanyi, JHEP 01, 052 (2007), (arXiv:hep-ph/0609043)] over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state.

NNLO jet cross sections by subtraction

International Nuclear Information System (INIS)

Somogyi, G.; Bolzoni, P.; Trocsanyi, Z.

2010-01-01

We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of the NNLO subtraction scheme of Refs. [G. Somogyi, Z. Trocsanyi, and V. Del Duca, JHEP 06, 024 (2005), (arXiv:hep-ph/0502226); G. Somogyi and Z. Trocsanyi, (2006), (arXiv:hep-ph/0609041); G. Somogyi, Z. Trocsanyi, and V. Del Duca, JHEP 01, 070 (2007), (arXiv:hep-ph/0609042); G. Somogyi and Z. Trocsanyi, JHEP 01, 052 (2007), (arXiv:hep-ph/0609043)] over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state.

FENDL/E. Evaluated nuclear data library of neutron nuclear interaction cross-sections and photon production cross-sections and photon-atom interaction cross sections for fusion applications. Version 1.1 of November 1994

International Nuclear Information System (INIS)

Pashchenko, A.B.; Wienke, H.; Ganesan, S.; McLaughlin, P.K.

1996-01-01

This document presents the description of a physical tape containing the basic evaluated nuclear data library of neutron nuclear interaction cross-sections and photon production cross-sections and photon-atom interaction cross-sections for fusion applications. It is part of FENDL, the evaluated nuclear data library for fusion applications. The nuclear data are available cost-free for distribution to interested scientists upon request. The data can also be retrieved by the user via online access through international computer networks. (author). 11 refs, 1 tab

Average cross sections for the 252Cf neutron spectrum

International Nuclear Information System (INIS)

Dezso, Z.; Csikai, J.

1977-01-01

A number of average cross sections have been measured for 252 Cf neutrons in (n, γ), (n,p), (n,2n), (n,α) reactions by the activation method and for fission by fission chamber. Cross sections have been determined for 19 elements and 45 reactions. The (n,γ) cross section values lie in the interval from 0.3 to 200 mb. The data as a function of target neutron number increases up to about N=60 with minimum near to dosed shells. The values lie between 0.3 mb and 113 mb. These cross sections decrease significantly with increasing the threshold energy. The values are below 20 mb. The data do not exceed 10 mb. Average (n,p) cross sections as a function of the threshold energy and average fission cross sections as a function of Zsup(4/3)/A are shown. The results obtained are summarized in tables

Cross section library DOSCROS77 (in the SAND-II format)

International Nuclear Information System (INIS)

Zijp, W.L.; Nolthenius, H.J.; Borg, N.J.C.M. van der.

1977-08-01

The dosimetry cross section library DOSCROS77 is documented with tables, plots and cross section values averaged over a few reference spectra. This library is based on the ENDF/B-IV dosimetry file, supplemented with some other evaluations. The total number of reaction cross section sets incorporated in this library is 49 (+3 cover cross sections sets). The cross section data are available in a format which is suitable for the program SAND-II

Problem Oriented Neutron-Gamma Cross Sections Libraries for WWER-440 and WWER-1000 Shielding and Reactor Vessel Dosimetry Application

International Nuclear Information System (INIS)

Belousov, S.; Antonov, S.; Ilieva, K.

1997-01-01

The 47 neutron and 20 gamma group libraries BGL-440 and BGL-1000 for the shielding and reactor vessel dosimetry application have been generated for WWER-440 and WWER-1000 by collapsing the VITAMIN-B6 library (199 neutron and 42 gamma groups on the base of ENDF/B-6). The first parts of the libraries for neutron-gamma transport calculation, BGL-440-1 (150 nuclides) and BGL-1000-1 (140 nuclides), have been generated by a modified version of SAS1X control module of the SCALE system. The appropriate zone-average neutron flux had been used for these sub-libraries collapsing. The BGL-440-2 and BGL-1000-2 sub-libraries consist of cross sections for all 120 nuclides of VITAMIN-B6, for calculation of the transport through non-reactor materials of dosimeters, capsules, specimens which may be placed in the cavity behind the reactor vessel. The neutron spectrum just beyond the RPV had been used for this collapsing. As the first test the comparative calculations of the neutron flux on/behind the WWER-1000 reactor vessel have been realised using the libraries BGL-1000 and BUGLE, intended for the American PWR reactors. The integral neutron flux values by BGL-1000 and BUGLE differ by 3% onto the vessel, and 5% behind the vessel. This result shows that the calculations of the neutron flux responses for the WWER vessel surveillance, especially in locations behind the WWER vessel have to be done by the appropriate BGL library. Key words: neutron transport, multigroup neutron cross section libraries

First measurement of the Rayleigh cross section

NARCIS (Netherlands)

Naus, H.; Ubachs, W.

2000-01-01

Rayleigh cross section for N2, Ar and SF6 was performed using the technique of cavity ring-down spectroscopy (CRDS). The experiment was based on the assumption that scattering cross section is equal to the extinction in the absence of absorption. The theory explains the molecular origin of

Parameterized representation of macroscopic cross section for PWR reactor

International Nuclear Information System (INIS)

Fiel, João Cláudio Batista; Carvalho da Silva, Fernando; Senra Martinez, Aquilino; Leal, Luiz C.

2015-01-01

Highlights: • This work describes a parameterized representation of the homogenized macroscopic cross section for PWR reactor. • Parameterization enables a quick determination of problem-dependent cross-sections to be used in few group calculations. • This work allows generating group cross-section data to perform PWR core calculations without computer code calculations. - Abstract: The purpose of this work is to describe, by means of Chebyshev polynomials, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 92 U enrichment. The cross-section data analyzed are fission, scattering, total, transport, absorption and capture. The parameterization enables a quick and easy determination of problem-dependent cross-sections to be used in few group calculations. The methodology presented in this paper will allow generation of group cross-section data from stored polynomials to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by the proposed methodology when compared with results from the SCALE code calculations show very good agreement

The total collision cross section in the glory region

International Nuclear Information System (INIS)

Biesen, J.J.H. van den.

1982-01-01

Chapter 1 presents a calculation of approximate total cross sections in the glory region from noble gas potentials. The relations between the main features of the total cross section and the properties of the potential to which these are sensitive are extensively investigated in chapter II. A beam apparatus has been developed, which allows for accurate measurements on the total cross section. All effects due to the finite angular and velocity resolution of the apparatus can be eliminated from the data to yield actual total cross sections as a function of the relative velocity. This facilitates a comparison to total cross sections predicted by potentials available in the literature. A brief description of the apparatus and of the data reduction is given in chapter III. The total cross section data obtained for various noble gas combinations are presented and analysed in chapter IV, where also a large number of potentials proposed in the literature is tested. In chapter V the quenching of the glories in the case of a non-spherical interaction is analysed. Subsequently, total cross section data for some atom-molecule systems are discussed. (Auth.)

Hardon cross sections at ultra high energies

International Nuclear Information System (INIS)

Yodh, G.B.

1987-01-01

A review of results on total hadronic cross sections at ultra high energies obtained from a study of longitudinal development of cosmic ray air showers is given. The experimental observations show that proton-air inelastic cross section increases from 275 mb to over 500 mb as the collision energy in the center of mass increases from 20 GeV to 20 TeV. The proton-air inelastic cross section, obtained from cosmic ray data at √s = 30 TeV, is compared with calculations using various different models for the energy variation of the parameters of the elementary proton-proton interaction. Three conclusions are derived

Total cross section results for deuterium electrodisintegration

International Nuclear Information System (INIS)

Skopik, D.M.; Murphy, J.J. II; Shin, Y.M.

1976-01-01

Theoretical total cross sections for deuterium electrodisintegration are presented as a function of incident electron energy. The cross section has been calculated using virtual photon theory with Partovi's photodisintegration calculation for E/subx/ > 10 MeV and effective range theory for E/subx/ 2 H(e, n) reaction in Tokamak reactors

Electron-impact ionization cross section of rubidium

International Nuclear Information System (INIS)

Kim, Y.; Migdalek, J.; Siegel, W.; Bieron, J.

1998-01-01

A theoretical model for electron-impact ionization cross section has been applied to Rb and the theoretical cross section (from the threshold to 1 keV in incident energy) is in good agreement with the recent experimental data obtained using Rb atoms trapped in a magneto-optical trap. The theoretical model, called the binary-encounter endash dipole (BED) model, combines a modified Mott cross section with the high-energy behavior of Born cross sections. To obtain the continuum dipole oscillator strength df/dE of the 5s electron required in the BED model, we used Dirac-Fock continuum wave functions with a core polarization potential that reproduced the known position of the Cooper minimum in the photoionization cross section. For inner-shell ionization, we used a simpler version of df/dE, which retained the hydrogenic shape. The contributions of the 4p→4d, 5s, and 5p autoionizing excitations were estimated using the plane-wave Born approximation. As a by-product, we also present the dipole oscillator strengths for the 5s→np 1/2 and 5s→np 3/2 transitions for high principal quantum numbers n near the ionization threshold obtained from the Dirac-Fock wave functions with the same core polarization potential as that used for the continuum wave functions. copyright 1998 The American Physical Society

Measurement cross sections for radioisotopes production

International Nuclear Information System (INIS)

Garrido, E.

2011-01-01

New radioactive isotopes for nuclear medicine can be produced using particle accelerators. This is one goal of Arronax, a high energy - 70 MeV - high intensity - 2*350 μA - cyclotron set up in Nantes. A priority list was established containing β - - 47 Sc, 67 Cu - β + - 44 Sc, 64 Cu, 82 Sr/ 82 Rb, 68 Ge/ 68 Ga - and α emitters - 211 At. Among these radioisotopes, the Scandium 47 and the Copper 67 have a strong interest in targeted therapy. The optimization of their productions required a good knowledge of their cross-sections but also of all the contaminants created during irradiation. We launched on Arronax a program to measure these production cross-sections using the Stacked-Foils' technique. It consists in irradiating several groups of foils - target, monitor and degrader foils - and in measuring the produced isotopes by γ-spectrometry. The monitor - nat Cu or nat Ni - is used to correct beam loss whereas degrader foils are used to lower beam energy. We chose to study the nat Ti(p,X) 47 Sc and 68 Zn(p,2p) 67 Cu reactions. Targets are respectively natural Titanium foil - bought from Goodfellow - and enriched Zinc 68 deposited on Silver. In the latter case, Zn targets were prepared in-house - electroplating of 68 Zn - and a chemical separation between Copper and Gallium isotopes has to be made before γ counting. Cross-section values for more than 40 different reactions cross-sections have been obtained from 18 MeV to 68 MeV. A comparison with the Talys code is systematically done. Several parameters of theoretical models have been studied and we found that is not possible to reproduce faithfully all the cross-sections with a given set of parameters. (author)

Temperature-dependent absorption cross sections for hydrogen peroxide vapor

Science.gov (United States)

Nicovich, J. M.; Wine, P. H.

1988-01-01

Relative absorption cross sections for hydrogen peroxide vapor were measured over the temperature ranges 285-381 K for lambda = 230 nm-295 nm and 300-381 K for lambda = 193 nm-350 nm. The well established 298 K cross sections at 202.6 and 228.8 nm were used as an absolute calibration. A significant temperature dependence was observed at the important tropospheric photolysis wavelengths lambda over 300 nm. Measured cross sections were extrapolated to lower temperatures, using a simple model which attributes the observed temperature dependence to enhanced absorption by molecules possessing one quantum of O-O stretch vibrational excitation. Upper tropospheric photodissociation rates calculated using the extrapolated cross sections are about 25 percent lower than those calculated using currently recommended 298 K cross sections.

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.

Calculation of the intermediate energy activation cross section

Energy Technology Data Exchange (ETDEWEB)

Furihata, Shiori; Yoshizawa, Nobuaki [Mitsubishi Research Inst., Inc., Tokyo (Japan)

1997-03-01

We discussed the activation cross section in order to predict accurately the activation of soil around an accelerator with high energy and strong intensity beam. For the assessment of the accuracy of activation cross sections estimated by a numerical model, we compared the calculated cross section with various experimental data, for Si(p,x){sup 22}Na, Al(p,x){sup 22}Na, Fe(p,x){sup 22}Na, Si(p,x){sup 7}Be, O(p,x){sup 3}H, Al(p,x){sup 3}H and Si(p,x){sup 3}H reactions. We used three computational codes, i.e., quantum molecular dynamics (QMD) plus statistical decay model (SDM), HETC-3STEP and the semiempirical method developed by Silberberg et.al. It is observed that the codes are accurate above 1GeV, except for {sup 7}Be production. We also discussed the difference between the activation cross sections of proton- and neutron-induced reaction. For the incident energy at 40MeV, it is found that {sup 3}H production cross sections of neutron-induced reaction are ten times as large as those of proton-induced reaction. It is also observed that the choice of the activation cross sections seriously affects to the estimate of saturated radioactivity, if the maximum energy of neutron flux is below 100MeV. (author)

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

Validation of evaluated neutron standard cross sections

International Nuclear Information System (INIS)

Badikov, S.; Golashvili, T.

2008-01-01

Some steps of the validation and verification of the new version of the evaluated neutron standard cross sections were carried out. In particular: -) the evaluated covariance data was checked for physical consistency, -) energy-dependent evaluated cross-sections were tested in most important neutron benchmark field - 252 Cf spontaneous fission neutron field, -) a procedure of folding differential standard neutron data in group representation for preparation of specialized libraries of the neutron standards was verified. The results of the validation and verification of the neutron standards can be summarized as follows: a) the covariance data of the evaluated neutron standards is physically consistent since all the covariance matrices of the evaluated cross sections are positive definite, b) the 252 Cf spectrum averaged standard cross-sections are in agreement with the evaluated integral data (except for 197 Au(n,γ) reaction), c) a procedure of folding differential standard neutron data in group representation was tested, as a result a specialized library of neutron standards in the ABBN 28-group structure was prepared for use in reactor applications. (authors)

View-CXS neutron and photon cross-sections viewer

International Nuclear Information System (INIS)

Subbaiah, K.V.; Sunil Sunny, C.

2004-01-01

A graphical user-friendly interface is developed in Visual Basic (VB)-6 to view the variation of neutron and photon interaction cross-sections of different isotopes as a function of energy. VB subroutines developed read the binary data files of cross-sections created in MCNP-ACE (Briesmeister, J.F., 1993. MCNP - a general purpose Monte Carlo N-Particle Transport code. Version 4A. LANL, USA), ANISN-DLC (Engle W.W. Jr., 1967, A User's Manual for ANISN, K-1693; ORNL, 1974. 100 group neutron cross section data based on ENDF/B-III. Oak Ridge National Laboratory, USA) and KENO-AMPX (Petrie, L.M., Landers, N.F., 1984 KENO-Va- An Improved Monte Carlo Criticality Program with Super Grouping. RSICC-CCC-548, USA) formats using LAHEY-77 Fortran Compiler. The information on isotopes present in each library will be displayed with the help of database files prepared using Micro-Soft ACESS. The cross-section data can be viewed in different presentation styles namely, line graphs, bar graphs, histograms etc., with different color and symbol options. The cross-section plots generated can be saved as Bit-Map file to embed in any other text files. This software enables inter comparison of cross-sections from different type of libraries for isotopes as well as mixtures. Provision is made to view the cross-sections for nuclear reactions such as (n,γ), (n,f), (n,α), etc. The software can be obtained from Radiation Safety Information and Computational Centre (RSICC), ORNL, USA with the code package identification number PSR-514. The software package needs a hard disk space of about 80 MB when installed and works in WINDOWS-95/98/2000 operating systems

Optical Model and Cross Section Uncertainties

Energy Technology Data Exchange (ETDEWEB)

Herman,M.W.; Pigni, M.T.; Dietrich, F.S.; Oblozinsky, P.

2009-10-05

Distinct minima and maxima in the neutron total cross section uncertainties were observed in model calculations using spherical optical potential. We found this oscillating structure to be a general feature of quantum mechanical wave scattering. Specifically, we analyzed neutron interaction with 56Fe from 1 keV up to 65 MeV, and investigated physical origin of the minima.We discuss their potential importance for practical applications as well as the implications for the uncertainties in total and absorption cross sections.

Pion-nucleus cross sections

International Nuclear Information System (INIS)

Barashenkov, V.S.

1990-01-01

The tables of inelastic and total cross sections of π ± mesons interactions with nuclei 4 He- 238 U are presented. The tables are obtained by theoretical analysis of known experimental data for energies higher some tens of MeV. 1 ref.; 1 tab

Proton-air and proton-proton cross sections

Directory of Open Access Journals (Sweden)

Ulrich Ralf

2013-06-01

Full Text Available Different attempts to measure hadronic cross sections with cosmic ray data are reviewed. The major results are compared to each other and the differences in the corresponding analyses are discussed. Besides some important differences, it is crucial to see that all analyses are based on the same fundamental relation of longitudinal air shower development to the observed fluctuation of experimental observables. Furthermore, the relation of the measured proton-air to the more fundamental proton-proton cross section is discussed. The current global picture combines hadronic proton-proton cross section data from accelerator and cosmic ray measurements and indicates a good consistency with predictions of models up to the highest energies.

Fission-neutron displacement cross sections in metals

International Nuclear Information System (INIS)

Takamura, Saburo; Aruga, Takeo; Nakata, Kiyotomo

1985-01-01

The sensitivity damage rates for 22 metals were measured after fission-spectrum neutron irradiation at low temperature and the experimental damage rates were compared with the theoretical calculation. The relation between the theoretical displacement cross section and the atomic weight of metals can be written by two curves; one is for fcc and hcp metals, and another is for bcc metals. On the other hand, the experimental displacement cross section versus atomic weight is shown approximately by a curve for both fcc and bcc metals, and the cross section for hcp metals deviates from the curve. The defect production efficiency is 0.3-0.4 for fcc metals and 0.6-0.8 for bcc metals. (orig.)

Compact fitting formulas for electron-impact cross sections

International Nuclear Information System (INIS)

Kim, Y.K.

1992-01-01

Compact fitting formulas, which contain four fitting constants, are presented for electron-impact excitation and ionization cross sections of atoms and ions. These formulas can fit experimental and theoretical cross sections remarkably well, when resonant structures are smoothed out, from threshold to high incident electron energies (<10 keV), beyond which relativistic formulas are more appropriate. Examples of fitted cross sections for some atoms and ions are presented. The basic form of the formula is valid for both atoms and molecules

Single-level resonance parameters fit nuclear cross-sections

Science.gov (United States)

Drawbaugh, D. W.; Gibson, G.; Miller, M.; Page, S. L.

1970-01-01

Least squares analyses of experimental differential cross-section data for the U-235 nucleus have yielded single level Breit-Wigner resonance parameters that fit, simultaneously, three nuclear cross sections of capture, fission, and total.

Total cross section of highly excited strings

International Nuclear Information System (INIS)

Lizzi, F.; Senda, I.

1990-01-01

The unpolarized total cross section for the joining of two highly excited strings is calculated. The calculation is performed by taking the average overall states in the given excitation levels of the initial strings. We find that the total cross section grows with the energy and momentum of the initial states. (author). 8 refs, 1 fig

MXS cross-section preprocessor user's manual

International Nuclear Information System (INIS)

Parker, F.; Ishikawa, M.; Luck, L.

1987-03-01

The MXS preprocessor has been designed to reduce the execution time of programs using isotopic cross-section data and to both reduce the execution time and improve the accuracy of shielding-factor interpolation in the SIMMER-II accident analysis program. MXS is a dual-purpose preprocessing code to: (1) mix isotopes into materials and (2) fit analytic functions to the shelf-shielding data. The program uses the isotope microscopic neutron cross-section data from the CCCC standard interface file ISOTXS and the isotope Bondarenko self-shielding data from the CCCC standard interface file BRKOXS to generate cross-section and self-shielding data for materials. The materials may be a mixture of several isotopes. The self-shielding data for the materials may be the actual shielding factors or a set of coefficients for functions representing the background dependence of the shielding factors. A set of additional data is given to describe the functions necessary to interpolate the shielding factors over temperature

Neutron capture cross sections of Kr

Directory of Open Access Journals (Sweden)

Fiebiger Stefan

2017-01-01

Full Text Available Neutron capture and β− -decay are competing branches of the s-process nucleosynthesis path at 85Kr [1], which makes it an important branching point. The knowledge of its neutron capture cross section is therefore essential to constrain stellar models of nucleosynthesis. Despite its importance for different fields, no direct measurement of the cross section of 85Kr in the keV-regime has been performed. The currently reported uncertainties are still in the order of 50% [2, 3]. Neutron capture cross section measurements on a 4% enriched 85Kr gas enclosed in a stainless steel cylinder were performed at Los Alamos National Laboratory (LANL using the Detector for Advanced Neutron Capture Experiments (DANCE. 85Kr is radioactive isotope with a half life of 10.8 years. As this was a low-enrichment sample, the main contaminants, the stable krypton isotopes 83Kr and 86Kr, were also investigated. The material was highly enriched and contained in pressurized stainless steel spheres.

NNLO jet cross sections by subtraction

Science.gov (United States)

Somogyi, G.; Bolzoni, P.; Trócsányi, Z.

2010-08-01

We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of the NNLO subtraction scheme of Refs. [G. Somogyi, Z. Trócsányi, and V. Del Duca, JHEP 06, 024 (2005), arXiv:hep-ph/0502226; G. Somogyi and Z. Trócsányi, (2006), arXiv:hep-ph/0609041; G. Somogyi, Z. Trócsányi, and V. Del Duca, JHEP 01, 070 (2007), arXiv:hep-ph/0609042; G. Somogyi and Z. Trócsányi, JHEP 01, 052 (2007), arXiv:hep-ph/0609043] over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state.

Mid-IR Absorption Cross-Section Measurements of Hydrocarbons

KAUST Repository

Alrefae, Majed Abdullah

2013-01-01

-known at combustion-relevant conditions. Absorption cross-section is an important spectroscopic quantity and has direct relation to the species concentration. In this work, the absorption cross-sections of basic hydrocarbons are measured using Fourier Transform

LHCb cross-section measurements with heavy flavour jets

CERN Multimedia

Michielin, Emanuele

2017-01-01

Cross-section measurements of jets originating from the hadronization of beauty ($b$) and charm ($c$) quarks at LHCb give the unique opportunity to probe Parton Distribution Functions (PDFs) at low and large momentum fraction and to test the Standard Model in the forward region. In this poster the production of $t\\bar{t}$ pairs in the forward region, the measurement of the $W+b\\bar{b}$ and $W+c\\bar{c}$ cross-section and the measurement of the $Z\\rightarrow b\\bar{b}$ cross-section are presented.

Neutron cross section measurements for the Fast Breeder Program

International Nuclear Information System (INIS)

Block, R.C.

1979-06-01

This research was concerned with the measurement of neutron cross sections of importance to the Fast Breeder Reactor. The capture and total cross sections of fission products ( 101 102 104 Ru, 143 145 Nd, 149 Sm, 95 97 Mo, Cs, Pr, Pd, 107 Pd, 99 Tc) and tag gases (Kr, 78 80 Kr) were measured up to 100 keV. Filtered neutron beams were used to measure the capture cross section of 238 U (with an Fe filter) and the total cross section of Na (with a Na filter). A radioactive neutron capture detector was developed. A list of publications is included

Developing Scientific Reasoning Through Drawing Cross-Sections

Science.gov (United States)

Hannula, K. A.

2012-12-01

Cross-sections and 3D models of subsurface geology are typically based on incomplete information (whether surface geologic mapping, well logs, or geophysical data). Creating and evaluating those models requires spatial and quantitative thinking skills (including penetrative thinking, understanding of horizontality, mental rotation and animation, and scaling). However, evaluating the reasonableness of a cross-section or 3D structural model also requires consideration of multiple possible geometries and geologic histories. Teaching students to create good models requires application of the scientific methods of the geosciences (such as evaluation of multiple hypotheses and combining evidence from multiple techniques). Teaching these critical thinking skills, especially combined with teaching spatial thinking skills, is challenging. My Structural Geology and Advanced Structural Geology courses have taken two different approaches to developing both the abilities to visualize and to test multiple models. In the final project in Structural Geology (a 3rd year course with a pre-requisite sophomore mapping course), students create a viable cross-section across part of the Wyoming thrust belt by hand, based on a published 1:62,500 geologic map. The cross-section must meet a number of geometric criteria (such as the template constraint), but is not required to balance. Each student tries many potential geometries while trying to find a viable solution. In most cases, the students don't visualize the implications of the geometries that they try, but have to draw them and then erase their work if it does not meet the criteria for validity. The Advanced Structural Geology course used Midland Valley's Move suite to test the cross-sections that they made in Structural Geology, mostly using the flexural slip unfolding algorithm and testing whether the resulting line lengths balanced. In both exercises, students seemed more confident in the quality of their cross-sections when the

Applications of the BEam Cross section Analysis Software (BECAS)

DEFF Research Database (Denmark)

Blasques, José Pedro Albergaria Amaral; Bitsche, Robert; Fedorov, Vladimir

2013-01-01

A newly developed framework is presented for structural design and analysis of long slender beam-like structures, e.g., wind turbine blades. The framework is based on the BEam Cross section Analysis Software – BECAS – a finite element based cross section analysis tool. BECAS is used for the gener......A newly developed framework is presented for structural design and analysis of long slender beam-like structures, e.g., wind turbine blades. The framework is based on the BEam Cross section Analysis Software – BECAS – a finite element based cross section analysis tool. BECAS is used...... for the generation of beam finite element models which correctly account for effects stemming from material anisotropy and inhomogeneity in cross sections of arbitrary geometry. These type of modelling approach allows for an accurate yet computationally inexpensive representation of a general class of three...

Absolute photoionization cross-section of the methyl radical.

Science.gov (United States)

Taatjes, Craig A; Osborn, David L; Selby, Talitha M; Meloni, Giovanni; Fan, Haiyan; Pratt, Stephen T

2008-10-02

The absolute photoionization cross-section of the methyl radical has been measured using two completely independent methods. The CH3 photoionization cross-section was determined relative to that of acetone and methyl vinyl ketone at photon energies of 10.2 and 11.0 eV by using a pulsed laser-photolysis/time-resolved synchrotron photoionization mass spectrometry method. The time-resolved depletion of the acetone or methyl vinyl ketone precursor and the production of methyl radicals following 193 nm photolysis are monitored simultaneously by using time-resolved synchrotron photoionization mass spectrometry. Comparison of the initial methyl signal with the decrease in precursor signal, in combination with previously measured absolute photoionization cross-sections of the precursors, yields the absolute photoionization cross-section of the methyl radical; sigma(CH3)(10.2 eV) = (5.7 +/- 0.9) x 10(-18) cm(2) and sigma(CH3)(11.0 eV) = (6.0 +/- 2.0) x 10(-18) cm(2). The photoionization cross-section for vinyl radical determined by photolysis of methyl vinyl ketone is in good agreement with previous measurements. The methyl radical photoionization cross-section was also independently measured relative to that of the iodine atom by comparison of ionization signals from CH3 and I fragments following 266 nm photolysis of methyl iodide in a molecular-beam ion-imaging apparatus. These measurements gave a cross-section of (5.4 +/- 2.0) x 10(-18) cm(2) at 10.460 eV, (5.5 +/- 2.0) x 10(-18) cm(2) at 10.466 eV, and (4.9 +/- 2.0) x 10(-18) cm(2) at 10.471 eV. The measurements allow relative photoionization efficiency spectra of methyl radical to be placed on an absolute scale and will facilitate quantitative measurements of methyl concentrations by photoionization mass spectrometry.

Cross-section crushing behaviour of hat-sections (Part II: Analytical modelling)

NARCIS (Netherlands)

Hofmeyer, H.

2005-01-01

Hat-sections are often used to experimentally investigate building sheeting subject to a concentrated load and bending. In car doors, hat-sections are used for side-impact protection. Their crushing behaviour can partly be explained by only observing their cross-sectional behaviour [1]. This

Asymptotic behaviour of pion-pion total cross-sections

Energy Technology Data Exchange (ETDEWEB)

Greynat, David [Dipartimento di Scienze Fisiche, Universita di Napoli “Federico II”,Via Cintia, 80126 Napoli (Italy); Rafael, Eduardo de [Aix-Marseille Université, CNRS,CPT, UMR 7332, 13288 Marseille (France); Université de Toulon, CNRS,CPT, UMR 7332, 83957 La Garde (France); Vulvert, Grégory [Departament de Física Teórica, IFIC,CSIC - Universitat de València, Apt. Correus 22085, E-46071 València (Spain)

2014-03-24

We derive a sum rule which shows that the Froissart-Martin bound for the asymptotic behaviour of the ππ total cross sections at high energies, if modulated by the Lukaszuk-Martin coefficient of the leading log{sup 2} s behaviour, cannot be an optimal bound in QCD. We next compute the total cross sections for π{sup +}π{sup −}, π{sup ±}π{sup 0} and π{sup 0}π{sup 0} scattering within the framework of the constituent chiral quark model (CχQM) in the limit of a large number of colours N{sub c} and discuss their asymptotic behaviours. The same ππ cross sections are also discussed within the general framework of Large-N{sub c} QCD and we show that it is possible to make an Ansatz for the isospin I=1 and I=0 spectrum which satisfy the Froissart-Martin bound with coefficients which, contrary to the Lukaszuk-Martin coefficient, are not singular in the chiral limit and have the correct Large-N{sub c} counting. We finally propose a simple phenomenological model which matches the low energy behaviours of the σ{sub π{sup ±}π{sup 0total}}(s) cross section predicted by the CχQM with the high energy behaviour predicted by the Large-N{sub c} Ansatz. The magnitude of these cross sections at very high energies is of the order of those observed for the pp and pp-bar scattering total cross sections.

Asymptotic behaviour of pion-pion total cross-sections

International Nuclear Information System (INIS)

Greynat, David; Rafael, Eduardo de; Vulvert, Grégory

2014-01-01

We derive a sum rule which shows that the Froissart-Martin bound for the asymptotic behaviour of the ππ total cross sections at high energies, if modulated by the Lukaszuk-Martin coefficient of the leading log 2  s behaviour, cannot be an optimal bound in QCD. We next compute the total cross sections for π + π − , π ± π 0 and π 0 π 0 scattering within the framework of the constituent chiral quark model (CχQM) in the limit of a large number of colours N c and discuss their asymptotic behaviours. The same ππ cross sections are also discussed within the general framework of Large-N c QCD and we show that it is possible to make an Ansatz for the isospin I=1 and I=0 spectrum which satisfy the Froissart-Martin bound with coefficients which, contrary to the Lukaszuk-Martin coefficient, are not singular in the chiral limit and have the correct Large-N c counting. We finally propose a simple phenomenological model which matches the low energy behaviours of the σ π ± π 0 total (s) cross section predicted by the CχQM with the high energy behaviour predicted by the Large-N c Ansatz. The magnitude of these cross sections at very high energies is of the order of those observed for the pp and pp-bar scattering total cross sections

Neutron total scattering cross sections of elemental antimony

Energy Technology Data Exchange (ETDEWEB)

Smith, A.B.; Guenther, P.T.; Whalen, J.F.

1982-11-01

Neutron total cross sections are measured from 0.8 to 4.5 MeV with broad resolutions. Differential-neutron-elastic-scattering cross sections are measured from 1.5 to 4.0 MeV at intervals of 50 to 200 keV and at scattering angles distributed between 20 and 160 degrees. Lumped-level neutron-inelastic-scattering cross sections are measured over the same angular and energy range. The exPerimental results are discussed in terms of an optical-statistical model and are compared with respective values given in ENDF/B-V.

Neutron total scattering cross sections of elemental antimony

International Nuclear Information System (INIS)

Smith, A.B.; Guenther, P.T.; Whalen, J.F.

1982-11-01

Neutron total cross sections are measured from 0.8 to 4.5 MeV with broad resolutions. Differential-neutron-elastic-scattering cross sections are measured from 1.5 to 4.0 MeV at intervals of 50 to 200 keV and at scattering angles distributed between 20 and 160 degrees. Lumped-level neutron-inelastic-scattering cross sections are measured over the same angular and energy range. The exPerimental results are discussed in terms of an optical-statistical model and are compared with respective values given in ENDF/B-V

Total cross sections for heavy flavour production at HERA

CERN Document Server

Frixione, Stefano; Nason, P; Ridolfi, G; Frixione, S; Mangano, M L; Nason, P; Ridolfi, G

1995-01-01

We compute total cross sections for charm and bottom photoproduction at HERA energies, and discuss the relevant theoretical uncertainties. In particular we discuss the problems arising from the small-x region, the uncertainties in the gluon parton density, and the uncertainties in the hadronic component of the cross section. Total electroproduction cross sections, calculated in the Weizs\\"acker-Williams approximation, are also given.

Determination of Intermediate Resonance Parameter with RMET21 for nTRACER

International Nuclear Information System (INIS)

Sohail, Muhammad; Kim, Myung Hyun

2012-01-01

Ray Tracing based code nTRACER is being developed in Seoul National University that has the capability of 3-dimensional whole core neutron transport calculation. As a part of development of multi-group neutron cross section library for nTRACER, the current work is intended to accurately determine intermediate resonance parameters. Beside the systematic calculation of subgroup parameters for resonance self shielding calculation, intermediate resonance parameters itself can be as important as the multi-group neutron cross section in the library and its overall accuracy. In this paper lambda factors were computed using RMET21 from ENDF/B-VII.1 for nTRACER to investigate its dependence on temperature and background cross section and replaced with lambda factors from HELIOS multi-group library. The procedure used for determining the intermediate resonance parameter for the isotope under study is introduced in the next section. Oxygen being one of the primary nuclide in PWR fuel has been selected for intermediate resonance parameters calculation

Heavy flavour hadro-production cross-sections

CERN Document Server

Wöhri, H K

2003-01-01

Hadro-production data on charm and beauty absolute cross-sections, collected by experiments at CERN, DESY and Fermilab, are reviewed. The measurements, corrected for the 'time evolution' of the branching ratios, are compared to calculations done with Pythia, as a function of the collision energy, using the latest parametrizations of the parton densities. We then estimate some charm and beauty production cross-sections relevant for future measurements, including nuclear effectes in the PDFs. We finish by briefly addressing the relevance, in heavy-ion collisions, of beauty production as feed-down for J/psi production.

Evaluated activation cross-sections and intercomparison of the ...

Indian Academy of Sciences (India)

mental data cross-section with the theoretical codes, to study the quality of the theoretical ... the cross-section, angular distribution, double differential data, gamma ..... TALYS. TENDL. Figure 6. Excitation function of the 87Sr(p, 2n)86Y reaction.

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